WorldWideScience

Sample records for neutron simulation capabilities

  1. Simulated workplace neutron fields

    International Nuclear Information System (INIS)

    Lacoste, V.; Taylor, G.; Rottger, S.

    2011-01-01

    The use of simulated workplace neutron fields, which aim at replicating radiation fields at practical workplaces, is an alternative solution for the calibration of neutron dosemeters. They offer more appropriate calibration coefficients when the mean fluence-to-dose equivalent conversion coefficients of the simulated and practical fields are comparable. Intensive Monte Carlo modelling work has become quite indispensable for the design and/or the characterization of the produced mixed neutron/photon fields, and the use of Bonner sphere systems and proton recoil spectrometers is also mandatory for a reliable experimental determination of the neutron fluence energy distribution over the whole energy range. The establishment of a calibration capability with a simulated workplace neutron field is not an easy task; to date only few facilities are available as standard calibration fields. (authors)

  2. PHISICS multi-group transport neutronic capabilities for RELAP5

    Energy Technology Data Exchange (ETDEWEB)

    Epiney, A.; Rabiti, C.; Alfonsi, A.; Wang, Y.; Cogliati, J.; Strydom, G. [Idaho National Laboratory (INL), 2525 N. Fremont Ave., Idaho Falls, ID 83402 (United States)

    2012-07-01

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

  3. Neutron transportation simulator

    International Nuclear Information System (INIS)

    Uenohara, Yuzo.

    1995-01-01

    In the present invention, problems in an existent parallelized monte carlo method is solved, and behaviors of neutrons in a large scaled system are accurately simulated at a high speed. Namely, a neutron transportation simulator according to the monte carlo method simulates movement of each of neutrons by using a parallel computer. In this case, the system to be processed is divided based on a space region and an energy region to which neutrons belong. Simulation of neutrons in the divided regions is allotted to each of performing devices of the parallel computer. Tarry data and nuclear data of the neutrons in each of the regions are memorized dispersedly to memories of each of the performing devices. A transmission means for simulating the behaviors of the neutrons in the region by each of the performing devices, as well as transmitting the information of the neutrons, when the neutrons are moved to other region, to the performing device in a transported portion are disposed to each of the performing devices. With such procedures, simulation for the neutrons in the allotted region can be conducted with small capacity of memories. (I.S.)

  4. Assessment of cold neutron radiography capability

    International Nuclear Information System (INIS)

    McDonald, T.E. Jr.; Roberts, J.A.

    1998-01-01

    This is the final report of a one-year, Laboratory Directed Research and Development (LDRD) project at Los Alamos National Laboratory (LANL). The authors goals were to demonstrate and assess cold neutron radiography techniques at the Los Alamos Neutron Science Center (LANSCE), Manual Lujan Neutron Scattering Center (Lujan Center), and to investigate potential applications of the capability. The authors have obtained images using film and an amorphous silicon detector. In addition, a new technique they have developed allows neutron radiographs to be made using only a narrow range of neutron energies. Employing this approach and the Bragg cut-off phenomena in certain materials, they have demonstrated material discrimination in radiography. They also demonstrated the imaging of cracks in a sample of a fire-set case that was supplied by Sandia National Laboratory, and they investigated whether the capability could be used to determine the extent of coking in jet engine nozzles. The LANSCE neutron radiography capability appears to have applications in the DOE stockpile maintenance and science-based stockpile stewardship (SBSS) programs, and in industry

  5. The neutron instrument simulation package, NISP

    International Nuclear Information System (INIS)

    Seeger, P.A.; Daemen, L.L.

    2004-01-01

    The Neutron Instrument Simulation Package (NISP) performs complete source-to-detector simulations of neutron instruments, including neutrons that do not follow the expected path. The original user interface (MC( ) Web) is a web-based application, http://strider.lansce.lanl.gov/NISP/Welcome.html. This report describes in detail the newer standalone Windows version, NISP( ) Win. Instruments are assembled from menu-selected elements, including neutron sources, collimation and transport elements, samples, analyzers, and detectors. Magnetic field regions may also be specified for the propagation of polarized neutrons including spin precession. Either interface writes a geometry file that is used as input to the Monte Carlo engine (MC( ) Run) in the user's computer. Both the interface and the engine rely on a subroutine library, MCLIB. The package is completely open source. New features include capillary optics, temperature dependence of Al and Be, revised source files for ISIS, and visualization of neutron trajectories at run time. Also, a single-crystal sample type has been successfully imported from McStas (with more generalized geometry), demonstrating the capability of including algorithms from other sources, and NISP( ) Win may render the instrument in a virtual reality file. Results are shown for two instruments under development.

  6. Building Airport Surface HITL Simulation Capability

    Science.gov (United States)

    Chinn, Fay Cherie

    2016-01-01

    FutureFlight Central is a high fidelity, real-time simulator designed to study surface operations and automation. As an air traffic control tower simulator, FFC allows stakeholders such as the FAA, controllers, pilots, airports, and airlines to develop and test advanced surface and terminal area concepts and automation including NextGen and beyond automation concepts and tools. These technologies will improve the safety, capacity and environmental issues facing the National Airspace system. FFC also has extensive video streaming capabilities, which combined with the 3-D database capability makes the facility ideal for any research needing an immersive virtual and or video environment. FutureFlight Central allows human in the loop testing which accommodates human interactions and errors giving a more complete picture than fast time simulations. This presentation describes FFCs capabilities and the components necessary to build an airport surface human in the loop simulation capability.

  7. A workshop on enhanced national capability for neutron scattering

    Energy Technology Data Exchange (ETDEWEB)

    Hurd, Alan J [Los Alamos National Laboratory; Rhyne, James J [Los Alamos National Laboratory; Lewis, Paul S [Los Alamos National Laboratory

    2009-01-01

    This two-day workshop will engage the international neutron scattering community to vet and improve the Lujan Center Strategic Plan 2007-2013 (SP07). Sponsored by the LANL SC Program Office and the University of California, the workshop will be hosted by LANSCE Professor Sunny Sinha (UCSD). Endorsement by the Spallation Neutron Source will be requested. The discussion will focus on the role that the Lujan Center will play in the national neutron scattering landscape assuming full utilization of beamlines, a refurbished LANSCE, and a 1.4-MW SNS. Because the Lujan Strategic Plan is intended to set the stage for the Signature Facility era at LANSCE, there will be some discussion of the long-pulse spallation source at Los Alamos. Breakout groups will cover several new instrument concepts, upgrades to present instruments, expanded sample environment capabilities, and a look to the future. The workshop is in keeping with a request by BES to update the Lujan strategic plan in coordination with the SNS and the broader neutron community. Workshop invitees will be drawn from the LANSCE User Group and a broad cross section of the US, European, and Pacific Rim neutron scattering research communities.

  8. A three-dimensional nodal neutron kinetics capability for relaps

    International Nuclear Information System (INIS)

    Judd, J.L.; Weaver, W.L.

    1996-01-01

    The incorporation of a three-dimensional neutron kinetics capability into the DOE version of the RELAP5/MOD3.2 reactor safety code is discussed. A brief discussion of the kinetics method is given along with a discussion of the cross section parameterization models available in RELAP5/MOD3.2. The RELAP5/MOD3.2 code is then used to perform calculations of the NEACRP rod ejection and rod withdrawal benchmarks, and results are presented

  9. Overview of US fast-neutron facilities and testing capabilities

    International Nuclear Information System (INIS)

    Evans, E.A.; Cox, C.M.; Jackson, R.J.

    1982-01-01

    Rather than attempt a cataloging of the various fast neutron facilities developed and used in this country over the last 30 years, this paper will focus on those facilities which have been used to develop, proof test, and explore safety issues of fuels, materials and components for the breeder and fusion program. This survey paper will attempt to relate the evolution of facility capabilities with the evolution of development program which use the facilities. The work horse facilities for the breeder program are EBR-II, FFTF and TREAT. For the fusion program, RTNS-II and FMIT were selected

  10. Binary neutron star merger simulations

    Energy Technology Data Exchange (ETDEWEB)

    Bruegmann, Bernd [Jena Univ. (Germany)

    2016-11-01

    Our research focuses on the numerical tools necessary to solve Einstein's equations. In recent years we have been particularly interested in spacetimes consisting of two neutron stars in the final stages of their evolution. Because of the emission of gravitational radiation, the objects are driven together to merge; the emitted gravitational wave signal is visualized. This emitted gravitational radiation carries energy and momentum away from the system and contains information about the system. Late last year the Laser Interferometer Gravitational-wave Observatory (LIGO) began searches for these gravitational wave signals at a sensitivity at which detections are expected. Although such systems can radiate a significant amount of their total mass-energy in gravitational waves, the gravitational wave signals one expects to receive on Earth are not strong, since sources of gravitational waves are often many millions of light years away. Therefore one needs accurate templates for the radiation one expects from such systems in order to be able to extract them out of the detector's noise. Although analytical models exist for compact binary systems when the constituents are well separated, we need numerical simulation to investigate the last orbits before merger to obtain accurate templates and validate analytical approximations. Due to the strong nonlinearity of the equations and the large separation of length scales, these simulations are computationally demanding and need to be run on large supercomputers. When matter is present the computational cost as compared to pure black hole (vacuum) simulations increases even more due to the additional matter fields. But also more interesting astrophysical phenomena can happen. In fact, there is the possibility for a strong electromagnetic signal from the merger (e.g., a short gamma-ray burst or lower-energy electromagnetic signatures from the ejecta) and significant neutrino emission. Additionally, we can expect that

  11. Monte Carlo simulations of neutron scattering instruments

    International Nuclear Information System (INIS)

    Aestrand, Per-Olof; Copenhagen Univ.; Lefmann, K.; Nielsen, K.

    2001-01-01

    A Monte Carlo simulation is an important computational tool used in many areas of science and engineering. The use of Monte Carlo techniques for simulating neutron scattering instruments is discussed. The basic ideas, techniques and approximations are presented. Since the construction of a neutron scattering instrument is very expensive, Monte Carlo software used for design of instruments have to be validated and tested extensively. The McStas software was designed with these aspects in mind and some of the basic principles of the McStas software will be discussed. Finally, some future prospects are discussed for using Monte Carlo simulations in optimizing neutron scattering experiments. (R.P.)

  12. Advanced research capabilities for neutron science and technology: Neutron polarizers for neutron scattering

    International Nuclear Information System (INIS)

    Penttila, S.I.; Fitzsimmons, M.R.; Delheij, P.J.

    1998-01-01

    The authors describe work on the development of polarized gaseous 3 He cells, which are intended for use as neutron polarizers. Laser diode arrays polarize Rb vapor in a sample cell and the 3 He is polarized via collisions. They describe development and tests of such a system at LANSCE

  13. Validation of the simulator neutronics model

    International Nuclear Information System (INIS)

    Gregory, M.V.

    1984-01-01

    The neutronics model in the SRP reactor training simulator computes the variation with time of the neutron population in the reactor core. The power output of a reactor is directly proportional to the neutron population, thus in a very real sense the neutronics model determines the response of the simulator. The geometrical complexity of the reactor control system in SRP reactors requires the neutronics model to provide a detailed, 3D representation of the reactor core. Existing simulator technology does not allow such a detailed representation to run in real-time in a minicomputer environment, thus an entirely different approach to the problem was required. A prompt jump method has been developed in answer to this need

  14. Polarized Neutron Reflectivity Simulation of Ferromagnet/ Antiferromagnet Thin Films

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Ki Yeon; Lee, Jeong Soo

    2008-02-15

    This report investigates the current simulating and fitting programs capable of calculating the polarized neutron reflectivity of the exchange-biased ferromagnet/antiferromagnet magnetic thin films. The adequate programs are selected depending on whether nonspin flip and spin flip reflectivities of magnetic thin films and good user interface are available or not. The exchange-biased systems such as Fe/Cr, Co/CoO, CoFe/IrMn/Py thin films have been simulated successfully with selected programs.

  15. Software for simulation and design of neutron scattering instrumentation

    DEFF Research Database (Denmark)

    Bertelsen, Mads

    designed using the software. The Union components uses a new approach to simulation of samples in McStas. The properties of a sample are split into geometrical and material, simplifying user input, and allowing the construction of complicated geometries such as sample environments. Multiple scattering...... from conventional choices. Simulation of neutron scattering instrumentation is used when designing instrumentation, but also to understand instrumental effects on the measured scattering data. The Monte Carlo ray-tracing package McStas is among the most popular, capable of simulating the path of each...... neutron through the instrument using an easy to learn language. The subject of the defended thesis is contributions to the McStas language in the form of the software package guide_bot and the Union components.The guide_bot package simplifies the process of optimizing neutron guides by writing the Mc...

  16. Recent developments and capabilities in neutron radiography at HEDL

    International Nuclear Information System (INIS)

    Tobin, J.E.; Karnesky, R.A.

    1981-01-01

    Radiographic imaging of irradiated fuels has routinely been qualitative in nature, providing information on the general condition of the internal components of an irradiated fuel pin. Conventional x-ray and neutron radiographic techniques, applicable to irradiated fuels, have not been of high enough resolution to be useful for quantitative image analysis. Recently, however, high resolution neutron radiographic techniques suitable for quantitative image analysis, using precision microdensitometry, have been developed at HEDL. This paper describes the modifications made to a previously developed system to obtain quantitative measurements from high resolution neutron radiographs and evaluates the precision of these measurements

  17. Efficiency simulation of long neutron counter

    International Nuclear Information System (INIS)

    Hu Qingyuan; Li Bojun; Zhang De; Guo Hongsheng; Wang Dong; Yang Gaozhao; Si Fenni; Liu Jian

    2008-01-01

    In order to achieve the high efficiency and uniform sensitivity for neutrons with widely different energies, the efficiency of long boron trifluoride proportional counter imbedded in polyethylene moderator was simulated by MCNP code. The result shows that detective efficiency would increase with increasing moderator radius and response curve at higher energy would be ameliorated through adjusting the thickness of front moderator. Also we calculated the relative efficiencies for different energy of a detector whose efficiencies were calibrated on an accelerator. The simulated efficiency for D-D neutrons (2.4 MeV) is 75% of the efficiency for D-T neutrons (14.1 MeV), which is approximately agreed with experimental data, 61%. The validity of the simulated model was proved by the consistent results between calculation and experiment data. (authors)

  18. High Fidelity Ion Beam Simulation of High Dose Neutron Irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Was, Gary; Wirth, Brian; Motta, Athur; Morgan, Dane; Kaoumi, Djamel; Hosemann, Peter; Odette, Robert

    2018-04-30

    Project Objective: The objective of this proposal is to demonstrate the capability to predict the evolution of microstructure and properties of structural materials in-reactor and at high doses, using ion irradiation as a surrogate for reactor irradiations. “Properties” includes both physical properties (irradiated microstructure) and the mechanical properties of the material. Demonstration of the capability to predict properties has two components. One is ion irradiation of a set of alloys to yield an irradiated microstructure and corresponding mechanical behavior that are substantially the same as results from neutron exposure in the appropriate reactor environment. Second is the capability to predict the irradiated microstructure and corresponding mechanical behavior on the basis of improved models, validated against both ion and reactor irradiations and verified against ion irradiations. Taken together, achievement of these objectives will yield an enhanced capability for simulating the behavior of materials in reactor irradiations

  19. A Comparison of Simulation Capabilities for Ducts

    Energy Technology Data Exchange (ETDEWEB)

    Miller, William A [ORNL; Smith, Matt K [ORNL; Gu, Lixing [Florida Solar Energy Center (FSEC); New, Joshua Ryan [ORNL

    2014-11-01

    Typically, the cheapest way to install a central air conditioning system in residential buildings is to place the ductwork in the attic. Energy losses due to duct-attic interactions can be great, but current whole-house models are unable to capture the dynamic multi-mode physics of the interactions. The building industry is notoriously fragmented and unable to devote adequate research resources to solve this problem. Builders are going to continue to put ducts in the attic because floor space is too expensive to closet them within living space, and there are both construction and aesthetic issues with other approaches such as dropped ceilings. Thus, there is a substantial need to publicly document duct losses and the cost of energy used by ducts in attics so that practitioners, builders, homeowners and state and federal code officials can make informed decisions leading to changes in new construction and additional retrofit actions. Thus, the goal of this study is to conduct a comparison of AtticSim and EnergyPlus simulation algorithms to identify specific features for potential inclusion in EnergyPlus that would allow higher-fidelity modeling of HVAC operation and duct transport of conditioned air. It is anticipated that the resulting analysis from these simulation tools will inform energy decisions relating to the role of ducts in future building energy codes and standards.

  20. Are Hydrostatic Models Still Capable of Simulating Oceanic Fronts

    Science.gov (United States)

    2016-11-10

    Hydrostatic Models Still Capable of Simulating Oceanic Fronts Yalin Fan Zhitao Yu Ocean Dynamics and Prediction Branch Oceanography Division FengYan Shi...OF PAGES 17. LIMITATION OF ABSTRACT Are Hydrostatic Models Still Capable of Simulating Oceanic Fronts? Yalin Fan, Zhitao Yu, and, Fengyan Shi1 Naval...mixed layer and thermocline simulations as well as large scale circulations. Numerical experiments are conducted using hydrostatic (HY) and

  1. Monte Carlo simulation of neutron scattering instruments

    International Nuclear Information System (INIS)

    Seeger, P.A.; Daemen, L.L.; Hjelm, R.P. Jr.

    1998-01-01

    A code package consisting of the Monte Carlo Library MCLIB, the executing code MC RUN, the web application MC Web, and various ancillary codes is proposed as an open standard for simulation of neutron scattering instruments. The architecture of the package includes structures to define surfaces, regions, and optical elements contained in regions. A particle is defined by its vector position and velocity, its time of flight, its mass and charge, and a polarization vector. The MC RUN code handles neutron transport and bookkeeping, while the action on the neutron within any region is computed using algorithms that may be deterministic, probabilistic, or a combination. Complete versatility is possible because the existing library may be supplemented by any procedures a user is able to code. Some examples are shown

  2. Capabilities of a DT tokamak fusion neutron source for driving a spent nuclear fuel transmutation reactor

    International Nuclear Information System (INIS)

    Stacey, W.M.

    2001-01-01

    The capabilities of a DT fusion neutron source for driving a spent nuclear fuel transmutation reactor are characterized by identifying limits on transmutation rates that would be imposed by tokamak physics and engineering limitations on fusion neutron source performance. The need for spent nuclear fuel transmutation and the need for a neutron source to drive subcritical fission transmutation reactors are reviewed. The likely parameter ranges for tokamak neutron sources that could produce an interesting transmutation rate of 100s to 1000s of kg/FPY (where FPY stands for full power year) are identified (P fus ∼ 10-100 MW, β N ∼ 2-3, Q p ∼ 2-5, R ∼ 3-5 m, I ∼ 6-10 MA). The electrical and thermal power characteristics of transmutation reactors driven by fusion and accelerator spallation neutron sources are compared. The status of fusion development vis-a-vis a neutron source is reviewed. (author)

  3. Progress Towards an Indirect Neutron Capture Capability at LANSCE

    Energy Technology Data Exchange (ETDEWEB)

    Koehler, Paul E. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Ullmann, John Leonard [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Couture, Aaron Joseph [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Mosby, Shea Morgan [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-09-20

    There are many neutron-capture cross sections of importance to radiochemical diagnostics and nuclear forensics which are beyond the reach of direct measurements. Hence, we have been developing an apparatus on flight path (FP) 13 at target 1 at LANSCE for tightly constraining these cross sections via determination of the underlying physical quantities. FP-13 was initially a cold-neutron beam line for materials science and therefore required substantial modification for use for nuclear physics. In FY17, we made several improvements to FP-13, demonstrated improved performance due to these changes via measurements on a variety of samples, identified a few more needed improvements, and reconfigured the beam line to implement the most important of these. New measurements to assess the impact of the most recent improvement will commence when beam is restored to LANSCE. Although FP-13 has not yet reached the performance required for small radioactive samples, measurements on a gold sample have led to an important science result which we are preparing for publication.

  4. Some neutron measurements with simulated ING targets

    International Nuclear Information System (INIS)

    Walker, J.

    1966-01-01

    Thermal neutron fluxes in the vicinity of a simulated Intense Neutron Generator target have been measured using Mn and Au foils, and a small BF 3 detector. The target was a Pb cylinder either 4-inch or 8-inch in diameter with a 1.2 g Ra-Be neutron source at its centre. This was centrally mounted in a 5' diam. x 5' high tank which was filled with either H 2 O or D 2 O moderator. Various gaps and absorbing annuli were placed around the target, and air-filled aluminum 'beam tubes' were mounted radially or tangentially from the target to simulate typical ING conditions. The measured thermal neutron fluxes were less than calculated at all radii. The single-age computation clearly gives large errors at large radii, but the multi-energy approach seems to give a useful indication of the thermal flux distribution in spite of the extreme simplicity of the model. The fall in measured fluxes at small radii in both D 2 O and H 2 O is most likely caused by absorption in the target material which is not allowed for in the computational model. (author)

  5. Some neutron measurements with simulated ING targets

    Energy Technology Data Exchange (ETDEWEB)

    Walker, J

    1966-07-01

    Thermal neutron fluxes in the vicinity of a simulated Intense Neutron Generator target have been measured using Mn and Au foils, and a small BF{sub 3} detector. The target was a Pb cylinder either 4-inch or 8-inch in diameter with a 1.2 g Ra-Be neutron source at its centre. This was centrally mounted in a 5' diam. x 5' high tank which was filled with either H{sub 2}O or D{sub 2}O moderator. Various gaps and absorbing annuli were placed around the target, and air-filled aluminum 'beam tubes' were mounted radially or tangentially from the target to simulate typical ING conditions. The measured thermal neutron fluxes were less than calculated at all radii. The single-age computation clearly gives large errors at large radii, but the multi-energy approach seems to give a useful indication of the thermal flux distribution in spite of the extreme simplicity of the model. The fall in measured fluxes at small radii in both D{sub 2}O and H{sub 2}O is most likely caused by absorption in the target material which is not allowed for in the computational model. (author)

  6. Mobile Smog Simulator: New Capabilities to Study Urban Mixtures

    Science.gov (United States)

    A smog simulator developed by EPA scientists and engineers has unique capabilities that will provide information for assessing the health impacts of relevant multipollutant atmospheres and identify contributions of specific sources.

  7. The upgraded cold neutron triple-axis spectrometer FLEXX – enhanced capabilities by new instrumental options

    Directory of Open Access Journals (Sweden)

    Habicht Klaus

    2015-01-01

    Full Text Available The upgrade of the cold neutron triple axis spectrometer FLEXX, a work-horse instrument for inelastic neutron scattering matching the sample environment capabilities at Helmholtz-Zentrum Berlin, has been successfully accomplished. Experiments confirmed an order of magnitude gain in flux now allowing for intensity demanding options to be fully exploited at FLEXX. In this article, we describe the layout and design of two newly available FLEXX instrument options in detail. The new Heusler analyzer gives an increase of the detected polarized neutron flux due to its superior focusing properties, significantly improving the feasibility of future polarized and neutron resonance spin echo experiments. The MultiFLEXX option provides simultaneous access to large regions in wavevector and energy space for inelastic excitations thus adding mapping capabilities to the spectrometer.

  8. Design and Rationale for an In Situ Cryogenic Deformation Capability at a Neutron Source

    International Nuclear Information System (INIS)

    Livescu, V.; Clausen, B.; Sisneros, T.; Bourke, M.A.M.; Woodruff, T.R.; Vaidyanathan, R.; Notardonato, W.U.

    2004-01-01

    When performed in conjunction with neutron diffraction, in situ loading offers unique insights on microstructural deformation mechanisms. This is by virtue of the penetration and phase sensitivity of neutrons. At Los Alamos National Laboratory room and high temperature (up to 1500 deg. C) polycrystalline constitutive response is modeled using finite element and self-consistent models. The models are compared to neutron diffraction measurements. In doing so the implications of slip and creep to microstructural response have been explored. Recently we have been considering low temperature phenomena. This includes changes in deformation mechanisms such as the increased predilection for twinning over slip. Since this is associated with measurable texture changes as well as microstructural strain effects, it is well suited for study using neutron diffraction. This paper outlines the design and rationale for a cryogenic loading capability that will be used on the Spectrometer for MAterials Research at Temperature and Stress (SMARTS) at the Los Alamos Neutron Science Center (LANSCE)

  9. Plant model of KIPT neutron source facility simulator

    International Nuclear Information System (INIS)

    Cao, Yan; Wei, Thomas Y.; Grelle, Austin L.; Gohar, Yousry

    2016-01-01

    Argonne National Laboratory (ANL) of the United States and Kharkov Institute of Physics and Technology (KIPT) of Ukraine are collaborating on constructing a neutron source facility at KIPT, Kharkov, Ukraine. The facility has 100-kW electron beam driving a subcritical assembly (SCA). The electron beam interacts with a natural uranium target or a tungsten target to generate neutrons, and deposits its power in the target zone. The total fission power generated in SCA is about 300 kW. Two primary cooling loops are designed to remove 100-kW and 300-kW from the target zone and the SCA, respectively. A secondary cooling system is coupled with the primary cooling system to dispose of the generated heat outside the facility buildings to the atmosphere. In addition, the electron accelerator has a low efficiency for generating the electron beam, which uses another secondary cooling loop to remove the generated heat from the accelerator primary cooling loop. One of the main functions the KIPT neutron source facility is to train young nuclear specialists; therefore, ANL has developed the KIPT Neutron Source Facility Simulator for this function. In this simulator, a Plant Control System and a Plant Protection System were developed to perform proper control and to provide automatic protection against unsafe and improper operation of the facility during the steady-state and the transient states using a facility plant model. This report focuses on describing the physics of the plant model and provides several test cases to demonstrate its capabilities. The plant facility model uses the PYTHON script language. It is consistent with the computer language of the plant control system. It is easy to integrate with the simulator without an additional interface, and it is able to simulate the transients of the cooling systems with system control variables changing on real-time.

  10. Plant model of KIPT neutron source facility simulator

    Energy Technology Data Exchange (ETDEWEB)

    Cao, Yan [Argonne National Lab. (ANL), Argonne, IL (United States); Wei, Thomas Y. [Argonne National Lab. (ANL), Argonne, IL (United States); Grelle, Austin L. [Argonne National Lab. (ANL), Argonne, IL (United States); Gohar, Yousry [Argonne National Lab. (ANL), Argonne, IL (United States)

    2016-02-01

    Argonne National Laboratory (ANL) of the United States and Kharkov Institute of Physics and Technology (KIPT) of Ukraine are collaborating on constructing a neutron source facility at KIPT, Kharkov, Ukraine. The facility has 100-kW electron beam driving a subcritical assembly (SCA). The electron beam interacts with a natural uranium target or a tungsten target to generate neutrons, and deposits its power in the target zone. The total fission power generated in SCA is about 300 kW. Two primary cooling loops are designed to remove 100-kW and 300-kW from the target zone and the SCA, respectively. A secondary cooling system is coupled with the primary cooling system to dispose of the generated heat outside the facility buildings to the atmosphere. In addition, the electron accelerator has a low efficiency for generating the electron beam, which uses another secondary cooling loop to remove the generated heat from the accelerator primary cooling loop. One of the main functions the KIPT neutron source facility is to train young nuclear specialists; therefore, ANL has developed the KIPT Neutron Source Facility Simulator for this function. In this simulator, a Plant Control System and a Plant Protection System were developed to perform proper control and to provide automatic protection against unsafe and improper operation of the facility during the steady-state and the transient states using a facility plant model. This report focuses on describing the physics of the plant model and provides several test cases to demonstrate its capabilities. The plant facility model uses the PYTHON script language. It is consistent with the computer language of the plant control system. It is easy to integrate with the simulator without an additional interface, and it is able to simulate the transients of the cooling systems with system control variables changing on real-time.

  11. A large-area, position-sensitive neutron detector with neutron/γ-ray discrimination capabilities

    International Nuclear Information System (INIS)

    Zecher, P.D.; Galonsky, A.; Kruse, J.J.; Gaff, S.J.; Ottarson, J.; Wang, J.; Seres, Z.; Ieki, K.; Iwata, Y.; Schelin, H.

    1997-01-01

    To further study neutron-rich halo nuclei, we have constructed a neutron detector array. The array consists of two separate banks of detectors, each of area 2 x 2 m 2 and containing 250 l of liquid scintillator. Each bank is position-sensitive to better than 10 cm. For neutron time-of-flight measurements, the time resolution of the detector has been demonstrated to be about 1 ns. By using the scintillator NE-213, we are able to distinguish between neutron and γ-ray signals above 1 MeV electron equivalent energy. Although the detector array was constructed for a particular experiment it has also been used in a number of other experiments. (orig.)

  12. Improving Neutron Measurement Capabilities; Expanding the Limits of Correlated Neutron Counting

    Energy Technology Data Exchange (ETDEWEB)

    Santi, Peter Angelo [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Geist, William H. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Dougan, Arden [National Nuclear Security Administration (NNSA), Washington, DC (United States)

    2015-11-05

    A number of technical and practical limitations exist within the neutron correlated counting techniques used in safeguards, especially within the algorithms that are used to process and analyze the detected neutron signals. A multi-laboratory effort is underway to develop new and improved analysis and data processing algorithms based on fundamental physics principles to extract additional or more accurate information about nuclear material bearing items.

  13. Electro-Thermal-Mechanical Simulation Capability Final Report

    International Nuclear Information System (INIS)

    White, D

    2008-01-01

    This is the Final Report for LDRD 04-ERD-086, 'Electro-Thermal-Mechanical Simulation Capability'. The accomplishments are well documented in five peer-reviewed publications and six conference presentations and hence will not be detailed here. The purpose of this LDRD was to research and develop numerical algorithms for three-dimensional (3D) Electro-Thermal-Mechanical simulations. LLNL has long been a world leader in the area of computational mechanics, and recently several mechanics codes have become 'multiphysics' codes with the addition of fluid dynamics, heat transfer, and chemistry. However, these multiphysics codes do not incorporate the electromagnetics that is required for a coupled Electro-Thermal-Mechanical (ETM) simulation. There are numerous applications for an ETM simulation capability, such as explosively-driven magnetic flux compressors, electromagnetic launchers, inductive heating and mixing of metals, and MEMS. A robust ETM simulation capability will enable LLNL physicists and engineers to better support current DOE programs, and will prepare LLNL for some very exciting long-term DoD opportunities. We define a coupled Electro-Thermal-Mechanical (ETM) simulation as a simulation that solves, in a self-consistent manner, the equations of electromagnetics (primarily statics and diffusion), heat transfer (primarily conduction), and non-linear mechanics (elastic-plastic deformation, and contact with friction). There is no existing parallel 3D code for simulating ETM systems at LLNL or elsewhere. While there are numerous magnetohydrodynamic codes, these codes are designed for astrophysics, magnetic fusion energy, laser-plasma interaction, etc. and do not attempt to accurately model electromagnetically driven solid mechanics. This project responds to the Engineering R and D Focus Areas of Simulation and Energy Manipulation, and addresses the specific problem of Electro-Thermal-Mechanical simulation for design and analysis of energy manipulation systems

  14. National power grid simulation capability : need and issues

    Energy Technology Data Exchange (ETDEWEB)

    Petri, Mark C. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2009-06-02

    On December 9 and 10, 2008, the Department of Homeland Security (DHS) Science and Technology Directorate sponsored a national workshop at Argonne National Laboratory to explore the need for a comprehensive modeling and simulation capability for the national electric power grid system. The workshop brought together leading electric power grid experts from federal agencies, the national laboratories, and academia to discuss the current state of power grid science and engineering and to assess if important challenges are being met. The workshop helped delineate gaps between grid needs and current capabilities and identify issues that must be addressed if a solution is to be implemented. This report is a result of the workshop and highlights power grid modeling and simulation needs, the barriers that must be overcome to address them, and the benefits of a national power grid simulation capability.

  15. Simulation of Neutron Backscattering applied to organic material detection

    International Nuclear Information System (INIS)

    Forero, N. C.; Cruz, A. H.; Cristancho, F.

    2007-01-01

    The Neutron Backscattering technique is tested when performing the task of localizing hydrogenated explosives hidden in soil. Detector system, landmine, soil and neutron source are simulated with Geant4 in order to obtain the number of neutrons detected when several parameters like mine composition, relative position mine-source and soil moisture are varied

  16. Comparative study of structural properties of trehalose water solutions by neutron diffraction, synchrotron radiation and simulation

    Energy Technology Data Exchange (ETDEWEB)

    Cesaro, A.; Magazu, V.; Migliardo, F.; Sussich, F.; Vadala, M

    2004-07-15

    Neutron diffraction measurements combined with H/D substitution have been performed on trehalose aqueous solutions as a function of temperature and concentration by using the SANDALS diffractometer at ISIS Facility (UK). The findings point out a high capability of trehalose to strongly affect the tetrahedral hydrogen bond network of water. The neutron diffraction results are also compared with simulation and experimental data obtained by synchrotron radiation on the phospholipid bilayer membranes (DPPC)/trehalose/H{sub 2}O ternary system.

  17. Advanced Simulation Capability for Environmental Management: Development and Demonstrations - 12532

    Energy Technology Data Exchange (ETDEWEB)

    Freshley, Mark D.; Freedman, Vicky; Gorton, Ian [Pacific Northwest National Laboratory, MSIN K9-33, P.O. Box 999, Richland, WA 99352 (United States); Hubbard, Susan S. [Lawrence Berkeley National Laboratory, 1 Cyclotron Road, MS 50B-4230, Berkeley, CA 94720 (United States); Moulton, J. David; Dixon, Paul [Los Alamos National Laboratory, MS B284, P.O. Box 1663, Los Alamos, NM 87544 (United States)

    2012-07-01

    The U.S. Department of Energy Office of Environmental Management (EM), Technology Innovation and Development is supporting development of the Advanced Simulation Capability for Environmental Management (ASCEM). ASCEM is a state-of-the-art scientific tool and approach for understanding and predicting contaminant fate and transport in natural and engineered systems. The modular and open source high-performance computing tool facilitates integrated approaches to modeling and site characterization that enable robust and standardized assessments of performance and risk for EM cleanup and closure activities. The ASCEM project continues to make significant progress in development of capabilities, which are organized into Platform and Integrated Tool-sets and a High-Performance Computing Multi-process Simulator. The Platform capabilities target a level of functionality to allow end-to-end model development, starting with definition of the conceptual model and management of data for model input. The High-Performance Computing capabilities target increased functionality of process model representations, tool-sets for interaction with Platform, and verification and model confidence testing. The new capabilities are demonstrated through working groups, including one focused on the Hanford Site Deep Vadose Zone. The ASCEM program focused on planning during the first year and executing a prototype tool-set for an early demonstration of individual components. Subsequently, ASCEM has focused on developing and demonstrating an integrated set of capabilities, making progress toward a version of the capabilities that can be used to engage end users. Demonstration of capabilities continues to be implemented through working groups. Three different working groups, one focused on EM problems in the deep vadose zone, another investigating attenuation mechanisms for metals and radionuclides, and a third focusing on waste tank performance assessment, continue to make progress. The project

  18. Simulation and Modeling Capability for Standard Modular Hydropower Technology

    Energy Technology Data Exchange (ETDEWEB)

    Stewart, Kevin M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Smith, Brennan T. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Witt, Adam M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); DeNeale, Scott T. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Bevelhimer, Mark S. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Pries, Jason L. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Burress, Timothy A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Kao, Shih-Chieh [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Mobley, Miles H. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Lee, Kyutae [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Curd, Shelaine L. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Tsakiris, Achilleas [Univ. of Tennessee, Knoxville, TN (United States); Mooneyham, Christian [Univ. of Tennessee, Knoxville, TN (United States); Papanicolaou, Thanos [Univ. of Tennessee, Knoxville, TN (United States); Ekici, Kivanc [Univ. of Tennessee, Knoxville, TN (United States); Whisenant, Matthew J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Welch, Tim [US Department of Energy, Washington, DC (United States); Rabon, Daniel [US Department of Energy, Washington, DC (United States)

    2017-08-01

    Grounded in the stakeholder-validated framework established in Oak Ridge National Laboratory’s SMH Exemplary Design Envelope Specification, this report on Simulation and Modeling Capability for Standard Modular Hydropower (SMH) Technology provides insight into the concepts, use cases, needs, gaps, and challenges associated with modeling and simulating SMH technologies. The SMH concept envisions a network of generation, passage, and foundation modules that achieve environmentally compatible, cost-optimized hydropower using standardization and modularity. The development of standardized modeling approaches and simulation techniques for SMH (as described in this report) will pave the way for reliable, cost-effective methods for technology evaluation, optimization, and verification.

  19. Simulation modeling on the growth of firm's safety management capability

    Institute of Scientific and Technical Information of China (English)

    LIU Tie-zhong; LI Zhi-xiang

    2008-01-01

    Aiming to the deficiency of safety management measure, established simulation model about firm's safety management capability(FSMC) based on organizational learning theory. The system dynamics(SD) method was used, in which level and rate system, variable equation and system structure flow diagram was concluded. Simulation model was verified from two aspects: first, model's sensitivity to variable was tested from the gross of safety investment and the proportion of safety investment; second, variables dependency was checked up from the correlative variable of FSMC and organizational learning. The feasibility of simulation model is verified though these processes.

  20. Finite size effects in neutron star and nuclear matter simulations

    Energy Technology Data Exchange (ETDEWEB)

    Giménez Molinelli, P.A., E-mail: pagm@df.uba.ar; Dorso, C.O.

    2015-01-15

    In this work we study molecular dynamics simulations of symmetric nuclear and neutron star matter using a semi-classical nucleon interaction model. Our aim is to gain insight on the nature of the so-called “finite size effects”, unavoidable in this kind of simulations, and to understand what they actually affect. To do so, we explore different geometries for the periodic boundary conditions imposed on the simulation cell: cube, hexagonal prism and truncated octahedron. For nuclear matter simulations we show that, at sub-saturation densities and low temperatures, the solutions are non-homogeneous structures reminiscent of the “nuclear pasta” phases expected in neutron star matter simulations, but only one structure per cell and shaped by specific artificial aspects of the simulations—for the same physical conditions (i.e. number density and temperature) different cells yield different solutions. The particular shape of the solution at low enough temperature and a given density can be predicted analytically by surface minimization. We also show that even if this behavior is due to the imposition of periodic boundary conditions on finite systems, this does not mean that it vanishes for very large systems, and it is actually independent of the system size. We conclude that, for nuclear matter simulations, the cells' size sets the only characteristic length scale for the inhomogeneities, and the geometry of the periodic cell determines the shape of those inhomogeneities. To model neutron star matter we add a screened Coulomb interaction between protons, and perform simulations in the three cell geometries. Our simulations indeed produce the well known nuclear pasta, with (in most cases) several structures per cell. However, we find that for systems not too large results are affected by finite size in different ways depending on the geometry of the cell. In particular, at the same certain physical conditions and system size, the hexagonal prism yields a

  1. Capability and limitation study of the DDT passive-active neutron waste assay instrument

    International Nuclear Information System (INIS)

    Nicholas, N.J.; Coop, K.L.; Estep, R.J.

    1992-05-01

    The differential-dieaway-technique passive-active neutron assay system is widely used by transuranic waste generators to certify their drummed waste for eventual shipment to the Waste Isolation Pilot Plant (WIPP). Stricter criteria being established for waste emplacement at the WIPP site has led to a renewed interest in improvements to and a better understanding of current nondestructive assay (NDA) techniques. Our study includes the effects of source position, extreme matrices, high neutron backgrounds, and source self-shielding to explore the system's capabilities and limitations and to establish a basis for comparison with other NDA systems. 11 refs

  2. Development of a coupled neutronic/thermal-hydraulic tool with multi-scale capabilities and applications to HPLWR core analysis

    International Nuclear Information System (INIS)

    Monti, Lanfranco; Starflinger, Joerg; Schulenberg, Thomas

    2011-01-01

    Highlights: → Advanced analysis and design techniques for innovative reactors are addressed. → Detailed investigation of a 3 pass core design with a multi-physics-scales tool. → Coupled 40-group neutron transport/equivalent channels TH core analyses methods. → Multi-scale capabilities: from equivalent channels to sub-channel pin-by-pin study. → High fidelity approach: reduction of conservatism involved in core simulations. - Abstract: The High Performance Light Water Reactor (HPLWR) is a thermal spectrum nuclear reactor cooled and moderated with light water operated at supercritical pressure. It is an innovative reactor concept, which requires developing and applying advanced analysis tools as described in the paper. The relevant water density reduction associated with the heat-up, together with the multi-pass core design, results in a pronounced coupling between neutronic and thermal-hydraulic analyses, which takes into account the strong natural influence of the in-core distribution of power generation and water properties. The neutron flux gradients within the multi-pass core, together with the pronounced dependence of water properties on the temperature, require to consider a fine spatial resolution in which the individual fuel pins are resolved to provide precise evaluation of the clad temperature, currently considered as one of the crucial design criteria. These goals have been achieved considering an advanced analysis method based on the usage of existing codes which have been coupled with developed interfaces. Initially neutronic and thermal-hydraulic full core calculations have been iterated until a consistent solution is found to determine the steady state full power condition of the HPLWR core. Results of few group neutronic analyses might be less reliable in case of HPLWR 3-pass core than for conventional LWRs because of considerable changes of the neutron spectrum within the core, hence 40 groups transport theory has been preferred to the

  3. Environments for online maritime simulators with cloud computing capabilities

    Science.gov (United States)

    Raicu, Gabriel; Raicu, Alexandra

    2016-12-01

    This paper presents the cloud computing environments, network principles and methods for graphical development in realistic naval simulation, naval robotics and virtual interactions. The aim of this approach is to achieve a good simulation quality in large networked environments using open source solutions designed for educational purposes. Realistic rendering of maritime environments requires near real-time frameworks with enhanced computing capabilities during distance interactions. E-Navigation concepts coupled with the last achievements in virtual and augmented reality will enhance the overall experience leading to new developments and innovations. We have to deal with a multiprocessing situation using advanced technologies and distributed applications using remote ship scenario and automation of ship operations.

  4. Stochastic and simulation models of maritime intercept operations capabilities

    OpenAIRE

    Sato, Hiroyuki

    2005-01-01

    The research formulates and exercises stochastic and simulation models to assess the Maritime Intercept Operations (MIO) capabilities. The models focus on the surveillance operations of the Maritime Patrol Aircraft (MPA). The analysis using the models estimates the probability with which a terrorist vessel (Red) is detected, correctly classified, and escorted for intensive investigation and neutralization before it leaves an area of interest (AOI). The difficulty of obtaining adequate int...

  5. Design and implementation of a multiaxial loading capability during heating on an engineering neutron diffractometer

    International Nuclear Information System (INIS)

    Benafan, O.; Padula, S. A.; Skorpenske, H. D.; An, K.; Vaidyanathan, R.

    2014-01-01

    A gripping capability was designed, implemented, and tested for in situ neutron diffraction measurements during multiaxial loading and heating on the VULCAN engineering materials diffractometer at the spallation neutron source at Oak Ridge National Laboratory. The proposed capability allowed for the acquisition of neutron spectra during tension, compression, torsion, and/or complex loading paths at elevated temperatures. The design consisted of age-hardened, Inconel ® 718 grips with direct attachment to the existing MTS load frame having axial and torsional capacities of 100 kN and 400 N·m, respectively. Internal cooling passages were incorporated into the gripping system for fast cooling rates during high temperature experiments up to ∼1000 K. The specimen mounting couplers combined a threaded and hexed end-connection for ease of sample installation/removal without introducing any unwanted loads. Instrumentation of this capability is documented in this work along with various performance parameters. The gripping system was utilized to investigate deformation in NiTi shape memory alloys under various loading/control modes (e.g., isothermal, isobaric, and cyclic), and preliminary results are presented. The measurements facilitated the quantification of the texture, internal strain, and phase fraction evolution in NiTi shape memory alloys under various loading/control modes

  6. Design and implementation of a multiaxial loading capability during heating on an engineering neutron diffractometer

    Energy Technology Data Exchange (ETDEWEB)

    Benafan, O., E-mail: othmane.benafan@nasa.gov [NASA Glenn Research Center, Structures and Materials Division, Cleveland, Ohio 44135 (United States); Advanced Materials Processing and Analysis Center, Materials Science and Engineering Department, University of Central Florida, Orlando, Florida 32816 (United States); Padula, S. A. [NASA Glenn Research Center, Structures and Materials Division, Cleveland, Ohio 44135 (United States); Skorpenske, H. D.; An, K. [Spallation Neutron Source, Oak Ridge National Laboratory, Oak Ridge, Tennessee 37831 (United States); Vaidyanathan, R. [Advanced Materials Processing and Analysis Center, Materials Science and Engineering Department, University of Central Florida, Orlando, Florida 32816 (United States)

    2014-10-01

    A gripping capability was designed, implemented, and tested for in situ neutron diffraction measurements during multiaxial loading and heating on the VULCAN engineering materials diffractometer at the spallation neutron source at Oak Ridge National Laboratory. The proposed capability allowed for the acquisition of neutron spectra during tension, compression, torsion, and/or complex loading paths at elevated temperatures. The design consisted of age-hardened, Inconel{sup ®} 718 grips with direct attachment to the existing MTS load frame having axial and torsional capacities of 100 kN and 400 N·m, respectively. Internal cooling passages were incorporated into the gripping system for fast cooling rates during high temperature experiments up to ~1000 K. The specimen mounting couplers combined a threaded and hexed end-connection for ease of sample installation/removal without introducing any unwanted loads. Instrumentation of this capability is documented in this work along with various performance parameters. The gripping system was utilized to investigate deformation in NiTi shape memory alloys under various loading/control modes (e.g., isothermal, isobaric, and cyclic), and preliminary results are presented. The measurements facilitated the quantification of the texture, internal strain, and phase fraction evolution in NiTi shape memory alloys under various loading/control modes.

  7. Monte Carlo simulation of neutron transport phenomena

    International Nuclear Information System (INIS)

    Srinivasan, P.

    2009-01-01

    Neutron transport is one of the central problems in nuclear reactor related studies and other applied sciences. Some of the major applications of neutron transport include nuclear reactor design and safety, criticality safety of fissile material handling, neutron detector design and development, nuclear medicine, assessment of radiation damage to materials, nuclear well logging, forensic analysis etc. Most reactor and dosimetry studies assume that neutrons diffuse from regions of high to low density just like gaseous molecules diffuse to regions of low concentration or heat flow from high to low temperature regions. However while treatment of gaseous or heat diffusion is quite accurately modeled, treatment of neutron transport as simple diffusion is quite limited. In simple diffusion, the neutron trajectories are irregular, random and zigzag - where as in neutron transport low reaction cross sections (1 barn- 10 -24 cm 2 ) lead to long mean free paths which again depend on the nature and irregularities of the medium. Hence a more accurate representation of the neutron transport evolved based on the Boltzmann equation of kinetic gas theory. In fact the neutron transport equation is a linearized version of the Boltzmann gas equation based on neutron conservation with seven independent variables. The transport equation is difficult to solve except in simple cases amenable to numerical methods. The diffusion (equation) approximation follows from removing the angular dependence of the neutron flux

  8. Advanced Simulation Capability for Environmental Management (ASCEM) Phase II Demonstration

    Energy Technology Data Exchange (ETDEWEB)

    Freshley, M. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Hubbard, S. [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Flach, G. [Savannah River National Lab. (SRNL), Aiken, SC (United States); Freedman, V. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Agarwal, D. [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Andre, B. [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Bott, Y. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Chen, X. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Davis, J. [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Faybishenko, B. [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Gorton, I. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Murray, C. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Moulton, D. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Meyer, J. [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Rockhold, M. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Shoshani, A. [LBNL; Steefel, C. [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Wainwright, H. [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Waichler, S. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2012-09-28

    In 2009, the National Academies of Science (NAS) reviewed and validated the U.S. Department of Energy Office of Environmental Management (EM) Technology Program in its publication, Advice on the Department of Energy’s Cleanup Technology Roadmap: Gaps and Bridges. The NAS report outlined prioritization needs for the Groundwater and Soil Remediation Roadmap, concluded that contaminant behavior in the subsurface is poorly understood, and recommended further research in this area as a high priority. To address this NAS concern, the EM Office of Site Restoration began supporting the development of the Advanced Simulation Capability for Environmental Management (ASCEM). ASCEM is a state-of-the-art scientific approach that uses an integration of toolsets for understanding and predicting contaminant fate and transport in natural and engineered systems. The ASCEM modeling toolset is modular and open source. It is divided into three thrust areas: Multi-Process High Performance Computing (HPC), Platform and Integrated Toolsets, and Site Applications. The ASCEM toolsets will facilitate integrated approaches to modeling and site characterization that enable robust and standardized assessments of performance and risk for EM cleanup and closure activities. During fiscal year 2012, the ASCEM project continued to make significant progress in capabilities development. Capability development occurred in both the Platform and Integrated Toolsets and Multi-Process HPC Simulator areas. The new Platform and Integrated Toolsets capabilities provide the user an interface and the tools necessary for end-to-end model development that includes conceptual model definition, data management for model input, model calibration and uncertainty analysis, and model output processing including visualization. The new HPC Simulator capabilities target increased functionality of process model representations, toolsets for interaction with the Platform, and model confidence testing and verification for

  9. Cosmic-ray neutron simulations and measurements in Taiwan

    International Nuclear Information System (INIS)

    Chen, Wei-Lin; Jiang, Shiang-Huei; Sheu, Rong-Jiun

    2014-01-01

    This study used simulations of galactic cosmic ray in the atmosphere to investigate the neutron background environment in Taiwan, emphasising its altitude dependence and spectrum variation near interfaces. The calculated results were analysed and compared with two measurements. The first measurement was a mobile neutron survey from sea level up to 3275 m in altitude conducted using a car-mounted high-sensitivity neutron detector. The second was a previous measured result focusing on the changes in neutron spectra near air/ground and air/water interfaces. The attenuation length of cosmic-ray neutrons in the lower atmosphere was estimated to be 163 g cm -2 in Taiwan. Cosmic-ray neutron spectra vary with altitude and especially near interfaces. The determined spectra near the air/ground and air/water interfaces agree well with measurements for neutrons below 10 MeV. However, the high-energy portion of spectra was observed to be much higher than our previous estimation. Because high-energy neutrons contribute substantially to a dose evaluation, revising the annual sea-level effective dose from cosmic-ray neutrons at ground level in Taiwan to 35 μSv, which corresponds to a neutron flux of 5.30 x 10 -3 n cm -2 s -1 , was suggested. The cosmic-ray neutron background in Taiwan was studied using the FLUKA simulations and field measurements. A new measurement was performed using a car-mounted high-efficiency neutron detector, re-coding real-time neutron counting rates from sea level up to 3275 m. The attenuation of cosmic-ray neutrons in the lower atmosphere exhibited an effective attenuation length of 163 g cm -2 . The calculated neutron counting rates over predicted the measurements by ∼32 %, which leaded to a correction factor for the FLUKA-calculated cosmic-ray neutrons in the lower atmosphere in Taiwan. In addition, a previous measurement regarding neutron spectrum variation near the air/ground and air/water interfaces was re-evaluated. The results showed that the

  10. Simulations for the neutron detector TETRA with MCNP

    International Nuclear Information System (INIS)

    Testov, D.; Kuznetsova, E.; Wilson, Jh.

    2013-01-01

    To study the nuclear structure of β-delayed neutron precursors at ALTO ISOL-facility at IPN (Orsay), the high efficiency 4π neutron detector TETRA with 3 He filled counters built at JINR (Dubna) was modified. The MCNP simulations to optimize the future configuration were necessary. The details of the calculations and the major results obtained are discussed

  11. Contrasting the capabilities of building energy performance simulation programs

    Energy Technology Data Exchange (ETDEWEB)

    Crawley, Drury B. [US Department of Energy, Washington, DC (United States); Hand, Jon W. [University of Strathclyde, Glasgow, Scotland (United Kingdom). Energy Systems Research Unit; Kummert, Michael [University of Wisconsin-Madison (United States). Solar Energy Laboratory; Griffith, Brent T. [National Renewable Energy Laboratory, Golden, CO (United States)

    2008-04-15

    For the past 50 years, a wide variety of building energy simulation programs have been developed, enhanced and are in use throughout the building energy community. This paper is an overview of a report, which provides up-to-date comparison of the features and capabilities of twenty major building energy simulation programs. The comparison is based on information provided by the program developers in the following categories: general modeling features; zone loads; building envelope and daylighting and solar; infiltration, ventilation and multizone airflow; renewable energy systems; electrical systems and equipment; HVAC systems; HVAC equipment; environmental emissions; economic evaluation; climate data availability, results reporting; validation; and user interface, links to other programs, and availability. (author)

  12. Simulation of Resistive Plate Chamber sensitivity to neutrons

    Energy Technology Data Exchange (ETDEWEB)

    Altieri, S. E-mail: saverio.altieri@pv.infn.it; Belli, G.; Bruno, G.; Merlo, M.; Ratti, S.P.; Riccardi, C.; Torre, P.; Vitulo, P.; Abbrescia, M.; Colaleo, A.; Iaselli, G.; Loddo, F.; Maggi, M.; Marangelli, B.; Natali, S.; Nuzzo, S.; Pugliese, G.; Ranieri, A.; Romano, F

    2001-04-01

    The Resistive Plate Chambers (RPCs) sensitivity to neutrons has been simulated using GEANT code with MICAP and FLUKA interfaces. The calculations have been performed as a function of the neutrons energy in the range 0.02 eV-1 GeV. To evaluate the response of the detector in the LHC background environment, the neutron energy spectrum expected in the CMS muon barrel has been taken into account; a hit rate due to neutrons of about 0.6 Hz cm{sup -2} has been estimated for a 250x250 cm{sup 2} RPC in the RB1 station.

  13. MCNP capabilities at the dawn of the 21st century: Neutron-gamma applications

    International Nuclear Information System (INIS)

    Selcow, E.C.; McKinney, G.W.

    2000-01-01

    The Los Alamos National Laboratory Monte Carlo N-Particle radiation transport code, MCNP, has become an international standard for a wide spectrum of neutron-gamma radiation transport applications. These include nuclear criticality safety, radiation shielding, nuclear safeguards, nuclear well-logging, fission and fusion reactor design, accelerator target design, detector design and analysis, health physics, medical radiation therapy and imaging, radiography, decontamination and decommissioning, and waste storage and disposal. The latest version of the code, MCNP4C, was released to the Radiation Safety Information Computational Center (RSICC) in February 2000.This paper described the new features and capabilities of the code, and discusses the specific applicability to neutron-gamma problems. We will also discuss the future directions for MCNP code development, including rewriting the code in Fortran 90

  14. Predictive Capability Maturity Model for computational modeling and simulation.

    Energy Technology Data Exchange (ETDEWEB)

    Oberkampf, William Louis; Trucano, Timothy Guy; Pilch, Martin M.

    2007-10-01

    The Predictive Capability Maturity Model (PCMM) is a new model that can be used to assess the level of maturity of computational modeling and simulation (M&S) efforts. The development of the model is based on both the authors experience and their analysis of similar investigations in the past. The perspective taken in this report is one of judging the usefulness of a predictive capability that relies on the numerical solution to partial differential equations to better inform and improve decision making. The review of past investigations, such as the Software Engineering Institute's Capability Maturity Model Integration and the National Aeronautics and Space Administration and Department of Defense Technology Readiness Levels, indicates that a more restricted, more interpretable method is needed to assess the maturity of an M&S effort. The PCMM addresses six contributing elements to M&S: (1) representation and geometric fidelity, (2) physics and material model fidelity, (3) code verification, (4) solution verification, (5) model validation, and (6) uncertainty quantification and sensitivity analysis. For each of these elements, attributes are identified that characterize four increasing levels of maturity. Importantly, the PCMM is a structured method for assessing the maturity of an M&S effort that is directed toward an engineering application of interest. The PCMM does not assess whether the M&S effort, the accuracy of the predictions, or the performance of the engineering system satisfies or does not satisfy specified application requirements.

  15. Simulation and prototyping of 2 m long resistive plate chambers for detection of fast neutrons and multi-neutron event identification

    Energy Technology Data Exchange (ETDEWEB)

    Elekes, Z., E-mail: z.elekes@hzdr.de [Helmholtz-Zentrum Dresden-Rossendorf, Dresden (Germany); Aumann, T. [GSI Helmholtzzentrumfür Schwerionenforschung, Darmstadt (Germany); Technische Universität Darmstadt, Darmstadt (Germany); Bemmerer, D. [Helmholtz-Zentrum Dresden-Rossendorf, Dresden (Germany); Boretzky, K. [GSI Helmholtzzentrumfür Schwerionenforschung, Darmstadt (Germany); Caesar, C. [GSI Helmholtzzentrumfür Schwerionenforschung, Darmstadt (Germany); Technische Universität Darmstadt, Darmstadt (Germany); Cowan, T.C. [Helmholtz-Zentrum Dresden-Rossendorf, Dresden (Germany); Technische Universität Dresden, Dresden (Germany); Hehner, J.; Heil, M. [GSI Helmholtzzentrumfür Schwerionenforschung, Darmstadt (Germany); Kempe, M. [Helmholtz-Zentrum Dresden-Rossendorf, Dresden (Germany); Rossi, D. [GSI Helmholtzzentrumfür Schwerionenforschung, Darmstadt (Germany); Röder, M. [Helmholtz-Zentrum Dresden-Rossendorf, Dresden (Germany); Technische Universität Dresden, Dresden (Germany); Simon, H. [GSI Helmholtzzentrumfür Schwerionenforschung, Darmstadt (Germany); Sobiella, M.; Stach, D. [Helmholtz-Zentrum Dresden-Rossendorf, Dresden (Germany); Reinhardt, T. [Helmholtz-Zentrum Dresden-Rossendorf, Dresden (Germany); Technische Universität Dresden, Dresden (Germany); Wagner, A.; Yakorev, D. [Helmholtz-Zentrum Dresden-Rossendorf, Dresden (Germany); Zilges, A. [Universität zu Köln, Köln (Germany); Zuber, K. [Technische Universität Dresden, Dresden (Germany)

    2013-02-11

    Resistive plate chamber (RPC) prototypes of 2 m length were simulated and built. The experimental tests using a 31 MeV electron beam, discussed in details, showed an efficiency higher than 90% and an excellent time resolution of around σ=100ps. Furthermore, comprehensive simulations were performed by GEANT4 toolkit in order to study the possible use of these RPCs for fast neutron (200 MeV–1 GeV) detection and multi-neutron event identification. The validation of simulation parameters was carried out via a comparison to experimental data. A possible setup for invariant mass spectroscopy of multi-neutron emission is presented and the characteristics are discussed. The results show that the setup has a high detection efficiency. Its capability of determining the momentum of the outgoing neutrons and reconstructing the relative energy between the fragments from nuclear reactions is demonstrated for different scenarios.

  16. SIMULATED 8 MeV NEUTRON RESPONSE FUNCTIONS OF A THIN SILICON NEUTRON SENSOR.

    Science.gov (United States)

    Takada, Masashi; Matsumoto, Tetsuro; Masuda, Akihiko; Nunomiya, Tomoya; Aoyama, Kei; Nakamura, Takashi

    2017-12-22

    Neutron response functions of a thin silicon neutron sensor are simulated using PHITS2 and MCNP6 codes for an 8 MeV neutron beam at angles of incidence of 0°, 30° and 60°. The contributions of alpha particles created from the 28Si(n,α)25Mg reaction and the silicon nuclei scattered elastically by neutrons in the silicon sensor have not been well reproduced using the MCNP6 code. The 8 MeV neutron response functions simulated using the PHITS2 code with an accurate event generator mode are in good agreement with experimental results and include the contributions of the alpha particles and silicon nuclei. © The Author(s) 2017. Published by Oxford University Press. All rights reserved. For Permissions, please email: journals.permissions@oup.com.

  17. A combined neutron scattering and simulation study on bioprotectant systems

    Energy Technology Data Exchange (ETDEWEB)

    Affouard, F. [Laboratoire de Dynamique et Structure des Materiaux Moleculaires UMR 8024, Universite Lille I - 59655 Villeneuve d' Ascq cedex (France); Bordat, P. [Laboratoire de Dynamique et Structure des Materiaux Moleculaires UMR 8024, Universite Lille I - 59655 Villeneuve d' Ascq cedex (France); Descamps, M. [Laboratoire de Dynamique et Structure des Materiaux Moleculaires UMR 8024, Universite Lille I - 59655 Villeneuve d' Ascq cedex (France); Lerbret, A. [Laboratoire de Dynamique et Structure des Materiaux Moleculaires UMR 8024, Universite Lille I - 59655 Villeneuve d' Ascq cedex (France); Magazu, S. [Dipartimento di Fisica and INFM, Universita di Messina, P.O. Box 55, I-98166 Messina (Italy); Migliardo, F. [Laboratoire de Dynamique et Structure des Materiaux Moleculaires UMR 8024, Universite Lille I - 59655 Villeneuve d' Ascq cedex (France); Dipartimento di Fisica and INFM, Universita di Messina, P.O. Box 55, I-98166 Messina (Italy)], E-mail: fmigliardo@unime.it; Ramirez-Cuesta, A.J. [ISIS Facility, Rutherford Appleton Laboratory, Chilton, Didcot (United Kingdom); Telling, M.F.T. [ISIS Facility, Rutherford Appleton Laboratory, Chilton, Didcot (United Kingdom)

    2005-10-31

    The present work shows quasi elastic neutron scattering, neutron spin echo and inelastic neutron scattering results on a class of bioprotectant systems, such as homologous disaccharides (i.e., trehalose and sucrose)/water solutions, as a function of temperature. The whole set of findings indicates a noticeable 'kosmotrope' character of the disaccharides, and in particular of trehalose, which is able to strongly modify both the structural and dynamical properties of water. This superior capability of trehalose can be linked to its higher bioprotective effectiveness in respect with the other disaccharides.

  18. Frameworks for Assessing the Quality of Modeling and Simulation Capabilities

    Science.gov (United States)

    Rider, W. J.

    2012-12-01

    The importance of assuring quality in modeling and simulation has spawned several frameworks for structuring the examination of quality. The format and content of these frameworks provides an emphasis, completeness and flow to assessment activities. I will examine four frameworks that have been developed and describe how they can be improved and applied to a broader set of high consequence applications. Perhaps the first of these frameworks was known as CSAU [Boyack] (code scaling, applicability and uncertainty) used for nuclear reactor safety and endorsed the United States' Nuclear Regulatory Commission (USNRC). This framework was shaped by nuclear safety practice, and the practical structure needed after the Three Mile Island accident. It incorporated the dominant experimental program, the dominant analysis approach, and concerns about the quality of modeling. The USNRC gave it the force of law that made the nuclear industry take it seriously. After the cessation of nuclear weapons' testing the United States began a program of examining the reliability of these weapons without testing. This program utilizes science including theory, modeling, simulation and experimentation to replace the underground testing. The emphasis on modeling and simulation necessitated attention on the quality of these simulations. Sandia developed the PCMM (predictive capability maturity model) to structure this attention [Oberkampf]. PCMM divides simulation into six core activities to be examined and graded relative to the needs of the modeling activity. NASA [NASA] has built yet another framework in response to the tragedy of the space shuttle accidents. Finally, Ben-Haim and Hemez focus upon modeling robustness and predictive fidelity in another approach. These frameworks are similar, and applied in a similar fashion. The adoption of these frameworks at Sandia and NASA has been slow and arduous because the force of law has not assisted acceptance. All existing frameworks are

  19. Neutron reflector design with Californium 252 neutron for Boron neutron chapter therapy facility using MCNP5 simulation method

    International Nuclear Information System (INIS)

    Muhammad Fakhrurreza; Kusminanto; Y Sardjono

    2014-01-01

    In this research has made a reflector design to provide beams of Neutron for BNCT with Californium-252 radioactive source. This collimator is useful to obtain optimum epithermal neutron flux with the smallest impurity radiation (thermal neutron, fast neutron, and gamma). The design process is done using Monte Carlo N-Particle simulation version 5 (MCNP5) code to calculate the neutron flux tally form. The chosen reflector design is the reflectors which use material such as BeO ceramic with 13 cm thick. Moderator use sulfur material with the slope angle of the cone is 30°. From the calculation result, it is obtained that Reflector with 1 gram Californium-252 source can produce a neutron output thermal which has thermal neutron specification 2.23189 x 10 9 n/s.cm 2 , epithermal neutron 3.51548 x 10 9 n/s.cm 2 , and fast neutron 4.82241 x 10 9 n/s.cm 2 From the result, it needs additional collimator because the BNCT requirement. (author)

  20. Simulation of merging neutron stars in full general relativity

    International Nuclear Information System (INIS)

    Shibata, M.

    2001-01-01

    We have performed 3D numerical simulations for merger of equal mass binary neutron stars in full general relativity. We adopt a Γ-law equation of state in the form P = (Γ - 1)ρε where P, ρ, ε and Γ are the pressure, rest mass density, specific internal energy, and the adiabatic constant. As initial conditions, we adopt models of irrotational binary neutron stars in a quasiequilibrium state. Simulations have been carried out for a wide range of Γ and compactness of neutron stars, paying particular attention to the final product and gravitational waves. We find that the final product depends sensitively on the initial compactness of the neutron stars: In a merger between sufficiently compact neutron stars, a black hole is formed in a dynamical timescale. As the compactness is decreased, the formation timescale becomes longer and longer. It is also found that a differentially rotating massive neutron star is formed instead of a black hole for less compact binary cases. In the case of black hole formation, the disk mass around the black hole appears to be very small; less than 1% of the total rest mass. It is indicated that waveforms of high-frequency gravitational waves after merger depend strongly on the compactness of neutron stars before the merger. We point out importance of detecting such gravitational waves of high frequency to constrain the maximum allowed mass of neutron stars. (author)

  1. PYRO - new capability for isotopic mass tracking in pyroprocess simulation

    International Nuclear Information System (INIS)

    Liaw, J.R.; Ackerman, J.P.

    1990-01-01

    A new computational code package called PYRO has been developed to support the IFR fuel recycle demonstration project in the HFEF/S facility at ANL-W. The basic pyrochemical code PYRO1 1 models the atomic mass flows and phase compositions of 48 essential chemical elements involved in the pyroprocess. It has been extended to PYRO1 2 by linking with the ORIGEN code to track more than 1000 isotopic species, their radioactive decays, and related phenomena. This paper first describes the pyroprocess to be modeled and the pyrochemical capability that has been implemented in PYRO1 1 , and then gives a full account on the algorithm of extending it to PYRO1 2 for isotopic mass tracking. Results from several scoping and simulation runs will be discussed to illustrate the significance of modeling in process radioactive decays

  2. PYRO: New capability for isotopic mass tracking in pyroprocess simulation

    International Nuclear Information System (INIS)

    Liaw, J.R.; Ackerman, J.P.

    1990-01-01

    A new computational code package called PYRO has been developed to support the IFR fuel recycle demonstration project in the HFEF/S facility at ANL-W. The basic pyrochemical code PYRO1-1 models the atomic mass flows and phase compositions of 48 essential chemical elements involved in the pyroprocess. It has been extended to PYRO1-2 by linking with the ORIGEN code to track more than 1000 isotopic species, their radioactive decays, and related phenomena. This paper first describes the pyroprocess to be modeled and the pyrochemical capability that has been implemented in PYRO1-1, and then gives a full account on the algorithm of extending it to PYRO1-2 for isotopic mass tracking. Results from several scoping and simulation runs will be discussed to illustrate the significance of modeling in-process radioactive decays. 16 refs., 8 figs., 2 tabs

  3. Advanced Simulation Capability for Environmental Management (ASCEM): Early Site Demonstration

    International Nuclear Information System (INIS)

    Meza, Juan; Hubbard, Susan; Freshley, Mark D.; Gorton, Ian; Moulton, David; Denham, Miles E.

    2011-01-01

    The U.S. Department of Energy Office of Environmental Management, Technology Innovation and Development (EM-32), is supporting development of the Advanced Simulation Capability for Environmental Management (ASCEM). ASCEM is a state-of-the-art scientific tool and approach for understanding and predicting contaminant fate and transport in natural and engineered systems. The modular and open source high performance computing tool will facilitate integrated approaches to modeling and site characterization that enable robust and standardized assessments of performance and risk for EM cleanup and closure activities. As part of the initial development process, a series of demonstrations were defined to test ASCEM components and provide feedback to developers, engage end users in applications, and lead to an outcome that would benefit the sites. The demonstration was implemented for a sub-region of the Savannah River Site General Separations Area that includes the F-Area Seepage Basins. The physical domain included the unsaturated and saturated zones in the vicinity of the seepage basins and Fourmile Branch, using an unstructured mesh fit to the hydrostratigraphy and topography of the site. The calculations modeled variably saturated flow and the resulting flow field was used in simulations of the advection of non-reactive species and the reactive-transport of uranium. As part of the demonstrations, a new set of data management, visualization, and uncertainty quantification tools were developed to analyze simulation results and existing site data. These new tools can be used to provide summary statistics, including information on which simulation parameters were most important in the prediction of uncertainty and to visualize the relationships between model input and output.

  4. Simulation of Light Collection for Neutron Electrical Dipole Moment measurement

    Science.gov (United States)

    Ji, Pan; nEDM Collaboration

    2017-09-01

    nEDM (Neutron Electrical Dipole moment) measurement addresses a critical topic in particle physics and Standard Model, that is CPT violation in neutron electrical dipole moment if detected in which the Time reversal violation is connected to the matter/antimatter imparity of the universe. The neutron electric dipole moment was first measured in 1950 by Smith, Purcell, and Ramsey at the Oak Ridge Reactor - the first intense neutron source. This measurement showed that the neutron was very nearly round (to better than one part in a million). The goal of the nEDM experiment is to further improve the precision of this measurement by another factor of 100. The signal from the experiment is detected by collecting the photons generated when neutron beams were captured by liquid helium 3. The Geant4 simulation project that I participate simulates the process of light collection to improve the design for higher capture efficiency. The simulated geometry includes light source, reflector, wavelength shifting fibers, wavelength shifting TPB and acrylic as in real experiment. The UV photons exiting from Helium go through two wavelength-shifting processes in TPB and fibers to be finally captured. Oak Ridge National Laboratory Neutron Electric Dipole Moment measurement project.

  5. Advance simulation capability for environmental management (ASCEM) - 59065

    International Nuclear Information System (INIS)

    Dixon, Paul; Keating, Elizabeth; Moulton, David; Williamson, Mark; Collazo, Yvette; Gerdes, Kurt; Freshley, Mark; Gorton, Ian; Meza, Juan

    2012-01-01

    The United States Department Energy (DOE) Office of Environmental Management (EM) determined that uniform application of advanced modeling in the subsurface could help reduce the cost and risks associated with its environmental cleanup mission. In response to this determination, the EM Office of Technology Innovation and Development (OTID), Groundwater and Soil Remediation (GW and S) began the program Advanced Simulation Capability for Environmental Management (ASCEM). ASCEM is a state-of-the-art scientific tool and approach for integrating data and scientific understanding to enable prediction of contaminant fate and transport in natural and engineered systems. This initiative supports the reduction of uncertainties and risks associated with EM?s environmental cleanup and closure programs through better understanding and quantifying the subsurface flow and contaminant transport behavior in complex geological systems. This involves the long-term performance of engineered components, including cementitious materials in nuclear waste disposal facilities that may be sources for future contamination of the subsurface. This paper describes the ASCEM tools and approach and the ASCEM programmatic accomplishments completed in 2010 including recent advances and technology transfer. The US Department of Energy Office of Environmental Management has begun development of an Advanced Simulation Capability for Environmental Management, (ASCEM). This program will provide predictions of the end states of contaminated areas allowing for cost and risk reduction of EM remedial activities. ASCEM will provide the tools and approaches necessary to standardize risk and performance assessments across the DOE complex. Through its Phase One demonstration, the ASCEM team has shown value to the EM community in the areas of High Performance Computing, Data Management, Visualization, and Uncertainty Quantification. In 2012, ASCEM will provide an initial limited release of a community code for

  6. Design, implementation, and testing of a cryogenic loading capability on an engineering neutron diffractometer

    Energy Technology Data Exchange (ETDEWEB)

    Woodruff, T. R.; Krishnan, V. B.; Vaidyanathan, R. [Department of Mechanical, Materials, and Aerospace Engineering, Advanced Materials Processing and Analysis Center (AMPAC), University of Central Florida, Orlando, Florida 32816 (United States); Clausen, B.; Sisneros, T.; Livescu, V.; Brown, D. W.; Bourke, M. A. M. [Los Alamos National Laboratory, Los Alamos, New Mexico 87545 (United States)

    2010-06-15

    A novel capability was designed, implemented, and tested for in situ neutron diffraction measurements during loading at cryogenic temperatures on the spectrometer for materials research at temperature and stress at Los Alamos National Laboratory. This capability allowed for the application of dynamic compressive forces of up to 250 kN on standard samples controlled at temperatures between 300 and 90 K. The approach comprised of cooling thermally isolated compression platens that in turn conductively cooled the sample in an aluminum vacuum chamber which was nominally transparent to the incident and diffracted neutrons. The cooling/heat rate and final temperature were controlled by regulating the flow of liquid nitrogen in channels inside the platens that were connected through bellows to the mechanical actuator of the load frame and by heaters placed on the platens. Various performance parameters of this system are reported here. The system was used to investigate deformation in Ni-Ti-Fe shape memory alloys at cryogenic temperatures and preliminary results are presented.

  7. Simulation of a complete inelastic neutron scattering experiment

    DEFF Research Database (Denmark)

    Edwards, H.; Lefmann, K.; Lake, B.

    2002-01-01

    A simulation of an inelastic neutron scattering experiment on the high-temperature superconductor La2-xSrxCuO4 is presented. The complete experiment, including sample, is simulated using an interface between the experiment control program and the simulation software package (McStas) and is compared...... with the experimental data. Simulating the entire experiment is an attractive alternative to the usual method of convoluting the model cross section with the resolution function, especially if the resolution function is nontrivial....

  8. Optimization of a neutron detector design using adjoint transport simulation

    International Nuclear Information System (INIS)

    Yi, C.; Manalo, K.; Huang, M.; Chin, M.; Edgar, C.; Applegate, S.; Sjoden, G.

    2012-01-01

    A synthetic aperture approach has been developed and investigated for Special Nuclear Materials (SNM) detection in vehicles passing a checkpoint at highway speeds. SNM is postulated to be stored in a moving vehicle and detector assemblies are placed on the road-side or in chambers embedded below the road surface. Neutron and gamma spectral awareness is important for the detector assembly design besides high efficiencies, so that different SNMs can be detected and identified with various possible shielding settings. The detector assembly design is composed of a CsI gamma-ray detector block and five neutron detector blocks, with peak efficiencies targeting different energy ranges determined by adjoint simulations. In this study, formulations are derived using adjoint transport simulations to estimate detector efficiencies. The formulations is applied to investigate several neutron detector designs for Block IV, which has its peak efficiency in the thermal range, and Block V, designed to maximize the total neutron counts over the entire energy spectrum. Other Blocks detect different neutron energies. All five neutron detector blocks and the gamma-ray block are assembled in both MCNP and deterministic simulation models, with detector responses calculated to validate the fully assembled design using a 30-group library. The simulation results show that the 30-group library, collapsed from an 80-group library using an adjoint-weighting approach with the YGROUP code, significantly reduced the computational cost while maintaining accuracy. (authors)

  9. Simulating Makrofol as a detector for neutron-induced recoils

    International Nuclear Information System (INIS)

    Zhang, G.; Becker, F.; Urban, M.; Xuan, Y.

    2011-01-01

    The response of solid-state nuclear track detector is extremely dependent on incident angles of neutrons, which determine the angular distribution of secondary particles. In this paper, the authors present a method to investigate the angular response of Makrofol detectors. Using the C++-based Monte-Carlo tool-kit Geant4 in combination with SRIM and our MATLAB codes, we simulated the angular response of Makrofol. The simulations were based on the restricted energy loss model, and the concept of energy threshold and critical angle. Experiments were carried out with 252 Cf neutrons to verify the simulation results. (authors)

  10. Monte Carlo simulation of neutron scattering instruments

    International Nuclear Information System (INIS)

    Seeger, P.A.

    1995-01-01

    A library of Monte Carlo subroutines has been developed for the purpose of design of neutron scattering instruments. Using small-angle scattering as an example, the philosophy and structure of the library are described and the programs are used to compare instruments at continuous wave (CW) and long-pulse spallation source (LPSS) neutron facilities. The Monte Carlo results give a count-rate gain of a factor between 2 and 4 using time-of-flight analysis. This is comparable to scaling arguments based on the ratio of wavelength bandwidth to resolution width

  11. Simulation of neutron background for DINO experiment

    International Nuclear Information System (INIS)

    Meghna, K.K.; Bhattacharjee, Pijushpani; Bhattacharya, Satyaki

    2017-01-01

    Various cosmological observations such as rotation curve of galaxies, gravitational lensing etc. establish the existence of a non-luminous matter known as Dark Matter which constitutes about 27% of the matter content of the universe. Despite the evidence for the existence of dark matter, its constituents are still unknown. In underground laboratories, neutrons can be generated mainly by spontaneous fission of U and radiogenic processes, such as by U / Th (α;n) reactions on the rock materials and by cosmogenic processes, such as interaction of cosmic ray muons with rock and shielding materials. We have estimated the flux of both the cosmogenic and the radiogenic neutrons for Jaduguda laboratory facility

  12. Simulations Of Neutron Beam Optic For Neutron Radiography Collimator Using Ray Tracing Methodology

    International Nuclear Information System (INIS)

    Norfarizan Mohd Said; Muhammad Rawi Mohamed Zin

    2014-01-01

    Ray- tracing is a technique for simulating the performance of neutron instruments. McStas, the open-source software package based on a meta-language, is a tool for carrying out ray-tracing simulations. The program has been successfully applied in investigating neutron guide design, flux optimization and other related areas with high complexity and precision. The aim of this paper is to discuss the implementation of ray-tracing technique with McStas for simulating the performance of neutron collimation system developed for imaging system of TRIGA RTP reactor. The code for the simulation was developed and the results are presented. The analysis of the performance is reported and discussed. (author)

  13. Testing the new stochastic neutronic code ANET in simulating safety important parameters

    International Nuclear Information System (INIS)

    Xenofontos, T.; Delipei, G.-K.; Savva, P.; Varvayanni, M.; Maillard, J.; Silva, J.; Catsaros, N.

    2017-01-01

    Highlights: • ANET is a new neutronics stochastic code. • Criticality calculations in both subcritical and critical nuclear systems of conventional design were conducted. • Simulations of thermal, lower epithermal and fast neutron fluence rates were performed. • Axial fission rate distributions in standard and MOX fuel pins were computed. - Abstract: ANET (Advanced Neutronics with Evolution and Thermal hydraulic feedback) is an under development Monte Carlo code for simulating both GEN II/III reactors as well as innovative nuclear reactor designs, based on the high energy physics code GEANT3.21 of CERN. ANET is built through continuous GEANT3.21 applicability amplifications, comprising the simulation of particles’ transport and interaction in low energy along with the accessibility of user-provided libraries and tracking algorithms for energies below 20 MeV, as well as the simulation of elastic and inelastic collision, capture and fission. Successive testing applications performed throughout the ANET development have been utilized to verify the new code capabilities. In this context the ANET reliability in simulating certain reactor parameters important to safety is here examined. More specifically the reactor criticality as well as the neutron fluence and fission rates are benchmarked and validated. The Portuguese Research Reactor (RPI) after its conversion to low enrichment in U-235 and the OECD/NEA VENUS-2 MOX international benchmark were considered appropriate for the present study, the former providing criticality and neutron flux data and the latter reaction rates. Concerning criticality benchmarking, the subcritical, Training Nuclear Reactor of the Aristotle University of Thessaloniki (TNR-AUTh) was also analyzed. The obtained results are compared with experimental data from the critical infrastructures and with computations performed by two different, well established stochastic neutronics codes, i.e. TRIPOLI-4.8 and MCNP5. Satisfactory agreement

  14. Monte Carlo simulation of mixed neutron-gamma radiation fields and dosimetry devices

    International Nuclear Information System (INIS)

    Zhang, Guoqing

    2011-01-01

    Monte Carlo methods based on random sampling are widely used in different fields for the capability of solving problems with a large number of coupled degrees of freedom. In this work, Monte Carlos methods are successfully applied for the simulation of the mixed neutron-gamma field in an interim storage facility and neutron dosimeters of different types. Details are discussed in two parts: In the first part, the method of simulating an interim storage facility loaded with CASTORs is presented. The size of a CASTOR is rather large (several meters) and the CASTOR wall is very thick (tens of centimeters). Obtaining the results of dose rates outside a CASTOR with reasonable errors costs usually hours or even days. For the simulation of a large amount of CASTORs in an interim storage facility, it needs weeks or even months to finish a calculation. Variance reduction techniques were used to reduce the calculation time and to achieve reasonable relative errors. Source clones were applied to avoid unnecessary repeated calculations. In addition, the simulations were performed on a cluster system. With the calculation techniques discussed above, the efficiencies of calculations can be improved evidently. In the second part, the methods of simulating the response of neutron dosimeters are presented. An Alnor albedo dosimeter was modelled in MCNP, and it has been simulated in the facility to calculate the calibration factor to get the evaluated response to a Cf-252 source. The angular response of Makrofol detectors to fast neutrons has also been investigated. As a kind of SSNTD, Makrofol can detect fast neutrons by recording the neutron induced heavy charged recoils. To obtain the information of charged recoils, general-purpose Monte Carlo codes were used for transporting incident neutrons. The response of Makrofol to fast neutrons is dependent on several factors. Based on the parameters which affect the track revealing, the formation of visible tracks was determined. For

  15. Monte Carlo simulation of mixed neutron-gamma radiation fields and dosimetry devices

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, Guoqing

    2011-12-22

    Monte Carlo methods based on random sampling are widely used in different fields for the capability of solving problems with a large number of coupled degrees of freedom. In this work, Monte Carlos methods are successfully applied for the simulation of the mixed neutron-gamma field in an interim storage facility and neutron dosimeters of different types. Details are discussed in two parts: In the first part, the method of simulating an interim storage facility loaded with CASTORs is presented. The size of a CASTOR is rather large (several meters) and the CASTOR wall is very thick (tens of centimeters). Obtaining the results of dose rates outside a CASTOR with reasonable errors costs usually hours or even days. For the simulation of a large amount of CASTORs in an interim storage facility, it needs weeks or even months to finish a calculation. Variance reduction techniques were used to reduce the calculation time and to achieve reasonable relative errors. Source clones were applied to avoid unnecessary repeated calculations. In addition, the simulations were performed on a cluster system. With the calculation techniques discussed above, the efficiencies of calculations can be improved evidently. In the second part, the methods of simulating the response of neutron dosimeters are presented. An Alnor albedo dosimeter was modelled in MCNP, and it has been simulated in the facility to calculate the calibration factor to get the evaluated response to a Cf-252 source. The angular response of Makrofol detectors to fast neutrons has also been investigated. As a kind of SSNTD, Makrofol can detect fast neutrons by recording the neutron induced heavy charged recoils. To obtain the information of charged recoils, general-purpose Monte Carlo codes were used for transporting incident neutrons. The response of Makrofol to fast neutrons is dependent on several factors. Based on the parameters which affect the track revealing, the formation of visible tracks was determined. For

  16. Improvement of analytical capabilities of neutron activation analysis laboratory at the Colombian Geological Survey

    Science.gov (United States)

    Parrado, G.; Cañón, Y.; Peña, M.; Sierra, O.; Porras, A.; Alonso, D.; Herrera, D. C.; Orozco, J.

    2016-07-01

    The Neutron Activation Analysis (NAA) laboratory at the Colombian Geological Survey has developed a technique for multi-elemental analysis of soil and plant matrices, based on Instrumental Neutron Activation Analysis (INAA) using the comparator method. In order to evaluate the analytical capabilities of the technique, the laboratory has been participating in inter-comparison tests organized by Wepal (Wageningen Evaluating Programs for Analytical Laboratories). In this work, the experimental procedure and results for the multi-elemental analysis of four soil and four plant samples during participation in the first round on 2015 of Wepal proficiency test are presented. Only elements with radioactive isotopes with medium and long half-lives have been evaluated, 15 elements for soils (As, Ce, Co, Cr, Cs, Fe, K, La, Na, Rb, Sb, Sc, Th, U and Zn) and 7 elements for plants (Br, Co, Cr, Fe, K, Na and Zn). The performance assessment by Wepal based on Z-score distributions showed that most results obtained |Z-scores| ≤ 3.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    International Nuclear Information System (INIS)

    Page, R.; Jones, J.R.

    1997-01-01

    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

  19. VESUVIO. A project to provide enhanced neutron scattering capabilities at the highest energy transfers

    International Nuclear Information System (INIS)

    Tomkinson, J.; Bowden, Z.A.; Mayers, J.; Norris, J.; Rhodes, N.J.; Colognesi, D.; Fielding, A.L.; Praitano, M.

    1999-01-01

    Complete text of publication follows. The VESUVIO project is financed within the TMR-Access to Large Scale Facility (RTD project) of the European Community. It will provide unique prototype instrumentation at the ISIS neutron source which will build on the success and experience of the eVS spectrometer in measuring single particle dynamics of a wide range of condensed matter systems. The instrumentation is designed for high momentum (20A -1 -1 ) and energy (ℎω>1eV) transfer inelastic neutron scattering studies of microscopic dynamical properties such as, single particle kinetic energies and momentum distributions. Specific objectives are: a) to optimize and construct a high efficiency, high area detector, 6 Li doped scintillator glasses are being tested; b) to construct a sample tank capable of operating with either a cold, or room temperature, filter analyzers; c) to develop new electronics and data acquisition to handle the high count-rates which will be generated in the azimuthal detectors. Some examples of applications performed during the first year of the project will be presented. (author)

  20. New Three-Dimensional Neutron Transport Calculation Capability in STREAM Code

    Energy Technology Data Exchange (ETDEWEB)

    Zheng, Youqi [Xi' an Jiaotong University, Xi' an (China); Choi, Sooyoung; Lee, Deokjung [UNIST, Ulsan (Korea, Republic of)

    2016-10-15

    The method of characteristics (MOC) is one of the best choices for its powerful capability in the geometry modeling. To reduce the large computational burden in 3D MOC, the 2D/1D schemes were proposed and have achieved great success in the past 10 years. However, such methods have some instability problems during the iterations when the neutron leakage for axial direction is large. Therefore, full 3D MOC methods were developed. A lot of efforts have been devoted to reduce the computational costs. However, it still requires too much memory storage and computational time for the practical modeling of a commercial size reactor core. Recently, a new approach for the 3D MOC calculation without transverse integration has been implemented in the STREAM code. In this approach, the angular flux is expressed as a basis function expansion form of only axial variable z. A new approach based on the axial expansion and 2D MOC sweeping to solve the 3D neutron transport equation is implemented in the STREAM code. This approach avoids using the transverse integration in the traditional 2D/1D scheme of MOC calculation. By converting the 3D equation into the 2D form of angular flux expansion coefficients, it also avoids the complex 3D ray tracing. Current numerical tests using two benchmarks show good accuracy of the new method.

  1. Improvement of analytical capabilities of neutron activation analysis laboratory at the Colombian Geological Survey

    Energy Technology Data Exchange (ETDEWEB)

    Parrado, G., E-mail: gparrado@sgc.gov.co; Cañón, Y.; Peña, M., E-mail: mlpena@sgc.gov.co; Sierra, O., E-mail: osierra@sgc.gov.co; Porras, A.; Alonso, D.; Herrera, D. C., E-mail: dherrera@sgc.gov.co; Orozco, J. [Colombian Geological Survey, Nuclear Affairs Technical Division, Neutron Activation Analysis Laboratory, Bogota D. C. (Colombia)

    2016-07-07

    The Neutron Activation Analysis (NAA) laboratory at the Colombian Geological Survey has developed a technique for multi-elemental analysis of soil and plant matrices, based on Instrumental Neutron Activation Analysis (INAA) using the comparator method. In order to evaluate the analytical capabilities of the technique, the laboratory has been participating in inter-comparison tests organized by Wepal (Wageningen Evaluating Programs for Analytical Laboratories). In this work, the experimental procedure and results for the multi-elemental analysis of four soil and four plant samples during participation in the first round on 2015 of Wepal proficiency test are presented. Only elements with radioactive isotopes with medium and long half-lives have been evaluated, 15 elements for soils (As, Ce, Co, Cr, Cs, Fe, K, La, Na, Rb, Sb, Sc, Th, U and Zn) and 7 elements for plants (Br, Co, Cr, Fe, K, Na and Zn). The performance assessment by Wepal based on Z-score distributions showed that most results obtained |Z-scores| ≤ 3.

  2. Specific Heat Capacity of Alloy 690 for Simulating Neutron Irradiation

    International Nuclear Information System (INIS)

    Park, Dae Gyu; Kim, Hee Moon; Song, Woong Sub; Baik, Seung Je; Joo, Young Sun; Ahn, Sang Bok; Park, Jin Seok; Lee, Won Jae; Ryu, Woo Seok

    2011-01-01

    The KAERI(Korea Atomic Energy Research Institute) is developing new type of nuclear reactor, so called 'SMART'(System Integrated Modular Advanced Reactor) which has many features of small power and system integrated modular type. Alloy 690 was selected as the candidate material for the heat exchanger tube of the steam generator of SMART. The SMART R and D is now facing the stage of engineering verification and approval of standard design to apply to DEMO reactors. Therefore, the material performance under the relevant environment is required to be evaluated. The important material performance issues are mechanical properties i.e. (fracture toughness, tensile and hardness) and thermal properties i.e. (thermal diffusivity, specific heat capacity and thermal conductivity) for which the engineering database is necessary to design a steam generator. However, the neutron post irradiation characteristics of the alloy 690 are barely known. As a result, PIE(Post Irradiation Examination) of thermal properties are planed and performed successfully. But specific heat capacity measurement is not performed because of not having proper test system for irradiated materials. Therefore in order to verify the effect of neutron irradiation for alloy 690, simulation method is adopted. In general, high energy neutron bombardment in material bring about lattice defects i.e. void, pore and dislocation. Dominant factor to impact to heat capacity is mainly dislocation in material. Therefore, simulation of neutron irradiation is devised by material rolling method in order to make artificial dislocation in alloy 690 as same effect of neutron irradiation. After preparing test specimens, heat capacity measurements are performed and results are compared with rolled materials and un-rolled materials to verify the effect of neutron irradiation simulation. Main interest of simulation is that heat capacity value is changed by neutron irradiation

  3. Determination Of Simulated Pellet To Pellet Gap Using Neutron Radiography

    International Nuclear Information System (INIS)

    Kusnowo, A.

    1996-01-01

    The defect on the irradiated fuel element could be detected using neutron radiography. The defect could occurred in pellet to pellet gap, cladding, or even cladding to pellet gap. An investigations has been performed to detect pellet to pellet gap defect that might occur in an irradiated fuel element. An Al foil of 0,1; 0,2; 0,3; und 0,4 mm was inserted between pellets to simulate various pellet to pellet gap. The neutron radiography used had power of 700 kW. The result showed that this simulation represented well enough problems that irradiated fuel element may experience

  4. A simulation report for the neutron guide development at HANARO

    International Nuclear Information System (INIS)

    Cho, S. J.; Soo, J. Y.; Seong, B. S.; Lee, C. H.; Kim, H. R.

    2006-04-01

    Lately, a demand of the measurement technique on atomic scale has been exceedingly increased over the whole field of the basic and technical science such as biotechnology, nano-technology, solid state physics, solid chemistry etc. Therefore a project called 'infrastructure construction for cold neutron research and utilization technique development' was launched in KAERI in July 2003, in order to raise a domestic basic science with an international level and elevate a international competitiveness for the bio-, nano- and informatics technology area through a wide contribution in a material structure research field. In order to accomplish this project until 2008, some important developments were launched at a same time such as a cold neutron source which shifts neutrons from a short wavelength range to a long wavelength, a system driving part for a smooth operation of a cold neutron source, and a neutron guide tube to be able to send neutrons to spectrometers located over a long distance. The guide simulation should be preferentially performed for an effective use of expensive neutron to meet the requirements such as wavelengths and type of instruments, experimental space, interferences with other instruments. The objective of this study was to decide guide shape, dimension, amount, curvature and instrument layout

  5. Monte Carlo simulations of the particle transport in semiconductor detectors of fast neutrons

    International Nuclear Information System (INIS)

    Sedlačková, Katarína; Zaťko, Bohumír; Šagátová, Andrea; Nečas, Vladimír

    2013-01-01

    Several Monte Carlo all-particle transport codes are under active development around the world. In this paper we focused on the capabilities of the MCNPX code (Monte Carlo N-Particle eXtended) to follow the particle transport in semiconductor detector of fast neutrons. Semiconductor detector based on semi-insulating GaAs was the object of our investigation. As converter material capable to produce charged particles from the (n, p) interaction, a high-density polyethylene (HDPE) was employed. As the source of fast neutrons, the 239 Pu–Be neutron source was used in the model. The simulations were performed using the MCNPX code which makes possible to track not only neutrons but also recoiled protons at all interesting energies. Hence, the MCNPX code enables seamless particle transport and no other computer program is needed to process the particle transport. The determination of the optimal thickness of the conversion layer and the minimum thickness of the active region of semiconductor detector as well as the energy spectra simulation were the principal goals of the computer modeling. Theoretical detector responses showed that the best detection efficiency can be achieved for 500 μm thick HDPE converter layer. The minimum detector active region thickness has been estimated to be about 400 μm. -- Highlights: ► Application of the MCNPX code for fast neutron detector design is demonstrated. ► Simulations of the particle transport through conversion film of HDPE are presented. ► Simulations of the particle transport through detector active region are presented. ► The optimal thickness of the HDPE conversion film has been calculated. ► Detection efficiency of 0.135% was reached for 500 μm thick HDPE conversion film

  6. Monte Carlo simulation of fast neutron scattering experiments including DD-breakup neutrons

    Energy Technology Data Exchange (ETDEWEB)

    Schmidt, D.; Siebert, B.R.L.

    1993-06-01

    The computational simulation of the deuteron breakup in a scattering experiment has been investigated. Experimental breakup spectra measured at 16 deuteron energies and at 7 angles for each energy served as the data base. Analysis of these input data and of the conditions of the scattering experiment made it possible to reduce the input data. The use of one weighted breakup spectrum is sufficient to simulate the scattering spectra at one incident neutron energy. A number of tests were carried out to prove the validity of this result. The simulation of neutron scattering on carbon, including the breakup, was compared with measured spectra. Differences between calculated and measured spectra were for the most part within the experimental uncertainties. Certain significant deviations can be attributed to erroneous scattering cross sections taken from an evaluation and used in the simulation. Scattering on higher-lying states in [sup 12]C can be analyzed by subtracting the simulated breakup-scattering from the experimental spectra. (orig.)

  7. Monte Carlo simulation of fast neutron scattering experiments including DD-breakup neutrons

    International Nuclear Information System (INIS)

    Schmidt, D.; Siebert, B.R.L.

    1993-06-01

    The computational simulation of the deuteron breakup in a scattering experiment has been investigated. Experimental breakup spectra measured at 16 deuteron energies and at 7 angles for each energy served as the data base. Analysis of these input data and of the conditions of the scattering experiment made it possible to reduce the input data. The use of one weighted breakup spectrum is sufficient to simulate the scattering spectra at one incident neutron energy. A number of tests were carried out to prove the validity of this result. The simulation of neutron scattering on carbon, including the breakup, was compared with measured spectra. Differences between calculated and measured spectra were for the most part within the experimental uncertainties. Certain significant deviations can be attributed to erroneous scattering cross sections taken from an evaluation and used in the simulation. Scattering on higher-lying states in 12 C can be analyzed by subtracting the simulated breakup-scattering from the experimental spectra. (orig.)

  8. Modularized Parallel Neutron Instrument Simulation on the TeraGrid

    International Nuclear Information System (INIS)

    Chen, Meili; Cobb, John W.; Hagen, Mark E.; Miller, Stephen D.; Lynch, Vickie E.

    2007-01-01

    In order to build a bridge between the TeraGrid (TG), a national scale cyberinfrastructure resource, and neutron science, the Neutron Science TeraGrid Gateway (NSTG) is focused on introducing productive HPC usage to the neutron science community, primarily the Spallation Neutron Source (SNS) at Oak Ridge National Laboratory (ORNL). Monte Carlo simulations are used as a powerful tool for instrument design and optimization at SNS. One of the successful efforts of a collaboration team composed of NSTG HPC experts and SNS instrument scientists is the development of a software facility named PSoNI, Parallelizing Simulations of Neutron Instruments. Parallelizing the traditional serial instrument simulation on TeraGrid resources, PSoNI quickly computes full instrument simulation at sufficient statistical levels in instrument de-sign. Upon SNS successful commissioning, to the end of 2007, three out of five commissioned instruments in SNS target station will be available for initial users. Advanced instrument study, proposal feasibility evaluation, and experiment planning are on the immediate schedule of SNS, which pose further requirements such as flexibility and high runtime efficiency on fast instrument simulation. PSoNI has been redesigned to meet the new challenges and a preliminary version is developed on TeraGrid. This paper explores the motivation and goals of the new design, and the improved software structure. Further, it describes the realized new features seen from MPI parallelized McStas running high resolution design simulations of the SEQUOIA and BSS instruments at SNS. A discussion regarding future work, which is targeted to do fast simulation for automated experiment adjustment and comparing models to data in analysis, is also presented

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-04-01

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

  10. Monte Carlo simulation of a coded-aperture thermal neutron camera

    International Nuclear Information System (INIS)

    Dioszegi, I.; Salwen, C.; Forman, L.

    2011-01-01

    We employed the MCNPX Monte Carlo code to simulate image formation in a coded-aperture thermal-neutron camera. The camera, developed at Brookhaven National Laboratory (BNL), consists of a 20 x 17 cm"2 active area "3He-filled position-sensitive wire chamber in a cadmium enclosure box. The front of the box is a coded-aperture cadmium mask (at present with three different resolutions). We tested the detector experimentally with various arrangements of moderated point-neutron sources. The purpose of using the Monte Carlo modeling was to develop an easily modifiable model of the device to predict the detector's behavior using different mask patterns, and also to generate images of extended-area sources or large numbers (up to ten) of them, that is important for nonproliferation and arms-control verification, but difficult to achieve experimentally. In the model, we utilized the advanced geometry capabilities of the MCNPX code to simulate the coded aperture mask. Furthermore, the code simulated the production of thermal neutrons from fission sources surrounded by a thermalizer. With this code we also determined the thermal-neutron shadow cast by the cadmium mask; the calculations encompassed fast- and epithermal-neutrons penetrating into the detector through the mask. Since the process of signal production in "3He-filled position-sensitive wire chambers is well known, we omitted this part from our modeling. Simplified efficiency values were used for the three (thermal, epithermal, and fast) neutron-energy regions. Electronic noise and the room's background were included as a uniform irradiation component. We processed the experimental- and simulated-images using identical LabVIEW virtual instruments. (author)

  11. Progress in the development of the neutron flux monitoring system of the French GEN-IV SFR: simulations and experimental validations [ANIMMA--2015-IO-98

    Energy Technology Data Exchange (ETDEWEB)

    Jammes, C.; Filliatre, P.; De Izarra, G. [CEA, DEN, DER, Instrumentation, Sensors and Dosimetry Laboratory, Cadarache, F-13108 Saint-Paul-lez-Durance, (France); Elter, Zs.; Pazsit, I. [Chalmers University of Technology, Department of Applied Physics, Division of Nuclear Engineering, SE-412 96 Goteborg, (Sweden); Verma, V.; Hellesen, C.; Jacobsson, S. [Division of Applied Nuclear Physics, Uppsala University, Box 516, SE-75120 Uppsala, (Sweden); Hamrita, H.; Bakkali, M. [CEA, DRT, LIST, Sensors and Electronic Architecture Laboratory, Saclay, F-91191 Gif Sur Yvette, (France); Chapoutier, N.; Scholer, A-C.; Verrier, D. [AREVA NP, 10 rue Juliette Recamier F-69456 Lyon, (France); Cantonnet, B.; Nappe, J-C. [PHONIS France S.A.S, Nuclear Instrumentation, Avenue Roger Roncier, B.P. 520, F-19106 Brive Cedex, (France); Molinie, P.; Dessante, P.; Hanna, R.; Kirkpatrick, M.; Odic, E. [Supelec, Department of Power and Energy System, F-91192 Gif Sur Yvette, (France); Jadot, F. [CEA, DEN, DER, ASTRID Project Group, Cadarache, F-13108 Saint-Paul-lez-Durance, (France)

    2015-07-01

    The neutron flux monitoring system of the French GEN-IV sodium-cooled fast reactor will rely on high temperature fission chambers installed in the reactor vessel and capable of operating over a wide-range neutron flux. The definition of such a system is presented and the technological solutions are justified with the use of simulation and experimental results. (authors)

  12. Simulating Neutronic Core Parameters in a Research and Test Reactor

    International Nuclear Information System (INIS)

    Selim, H.K.; Amin, E.A.; Koutb, M.E.

    2011-01-01

    The present study proposes an Artificial Neural Network (ANN) modeling technique that predicts the control rods positions in a nuclear research reactor. The neutron, flux in the core of the reactor is used as the training data for the neural network model. The data used to train and validate the network are obtained by modeling the reactor core with the neutronic calculation code: CITVAP. The type of the network used in this study is the feed forward multilayer neural network with the backpropagation algorithm. The results show that the proposed ANN has good generalization capability to estimate the control rods positions knowing neutron flux for a research and test reactor. This method can be used to predict critical control rods positions to be used for reactor operation after reload

  13. Improving the neutron-to-photon discrimination capability of detectors used for neutron dosimetry in high energy photon beam radiotherapy

    International Nuclear Information System (INIS)

    Irazola, L.; Terrón, J.A.; Bedogni, R; Pola, A.; Lorenzoli, M.; Sánchez-Nieto, B.; Gómez, F.; Sánchez-Doblado, F.

    2016-01-01

    The increasing interest of the medical community to radioinduced second malignancies due to photoneutrons in patients undergoing high-energy radiotherapy, has stimulated in recent years the study of peripheral doses, including the development of some dedicated active detectors. Although these devices are designed to respond to neutrons only, their parasitic photon response is usually not identically zero and anisotropic. The impact of these facts on measurement accuracy can be important, especially in points close to the photon field-edge. A simple method to estimate the photon contribution to detector readings is to cover it with a thermal neutron absorber with reduced secondary photon emission, such as a borated rubber. This technique was applied to the TNRD (Thermal Neutron Rate Detector), recently validated for thermal neutron measurements in high-energy photon radiotherapy. The positive results, together with the accessibility of the method, encourage its application to other detectors and different clinical scenarios. - Highlights: • Neutron-to-photon discrimination of a thermal neutron detector used in radiotherapy. • Photon and anisotropic response study with distance and beam incidence of thermal neutron detector. • Borated rubber for estimating photon contribution in any thermal neutron detector.

  14. Thermal expansion study of simulated DUPIC fuel using neutron diffraction

    International Nuclear Information System (INIS)

    Kang, Kweon Ho; Ryu, H. J.; Bae, J. H.; Kim, H. S.; Song, K. C.; Yang, M. S.; Choi, Y. N.; Han, Y. S.; Oh, H. S.

    2001-07-01

    The lattice parameters of simulated DUPIC fuel and UO2 were measured from room temperature to 1273 K using neutron diffraction to investigate the thermal expansion and density variation with temperature. The lattice parameter of simulated DUPIC fuel is lower than that of UO2 and the linear thermal expansion of simulated DUPIC fuel is higher than that of UO2. For the temperature range from 298 to 1273 K, the average linear thermal expansion coefficients for UO2 and simulated DUPIC fuel are 10.471 ''10-6 and 10.751 ''10-6 K-1, respectively

  15. A methodology of neutronic-thermodynamics simulation for fast reactor

    International Nuclear Information System (INIS)

    Waintraub, M.

    1986-01-01

    Aiming at a general optimization of the project, controlled fuel depletion and management, this paper develop a neutronic thermodynamics simulator, SIRZ, which besides being sufficiently precise, is also economic. That results in a 75% reduction in CPU time, for a startup calculation, when compared with the same calculation at the CITATION code. The simulation system by perturbation calculations, applied to fast reactors, which produce errors smaller than 1% in all components of the reference state given by the CITATION code was tested. (author)

  16. Capability of NIPAM polymer gel in recording dose from the interaction of 10B and thermal neutron in BNCT

    International Nuclear Information System (INIS)

    Khajeali, Azim; Reza Farajollahi, Ali; Kasesaz, Yaser; Khodadadi, Roghayeh; Khalili, Assef; Naseri, Alireza

    2015-01-01

    The capability of N-isopropylacrylamide (NIPAM) polymer gel to record the dose resulting from boron neutron capture reaction in BNCT was determined. In this regard, three compositions of the gel with different concentrations of 10 B were prepared and exposed to gamma radiation and thermal neutrons. Unlike irradiation with gamma rays, the boron-loaded gels irradiated by neutron exhibited sensitivity enhancement compared with the gels without 10 B. It was also found that the neutron sensitivity of the gel increased by the increase of concentration of 10 B. It can be concluded that NIPAM gel might be suitable for the measurement of the absorbed dose enhancement due to 10 B and thermal neutron reaction in BNCT. - Highlights: • Three compositions of NIPAM gel with different concentration of 10 B have been exposed by gamma and thermal neutron. • The vials containing NIPAM gel have been irradiated by an automatic system capable of providing for dose uniformity. • Suitability of NIPAM polymer gel in measuring radiation doses in BNCT has been investigated.

  17. Simulation of neutron fluxes around the W7-X Stellarator

    Energy Technology Data Exchange (ETDEWEB)

    Andersson, Jenny

    1999-12-01

    A new fusion experiment, the WENDELSTEIN 7-X Stellarator (W7-X), will be undertaken in Greifswald in Germany. Measurements of the neutron flux will provide information on fusion reaction rates and possibly also on ion temperatures as function of time. For this purpose moderating neutron counters will be designed, tested, calibrated and eventually used at W7-X. Extensive Monte-Carlo simulations have been performed in order to select the most suitable detector and moderator combination with a flat response function and highest achievable efficiency. Different detector configurations with different moderating materials have been tried out, showing that a 32 cm thick graphite moderating BF{sub 3} -counter gives the desired flat response and sufficient efficiency. Neutron spectra calculations have been made for different torus models and the influence of floor, walls and ceiling (i.e. reactor hall) have been investigated. Presented results suggest that a more detailed torus model significantly reduces the number of neutron counts at the detector. Calculations including the reactor hall indicate a tendency of shifting the neutron spectra towards the thermal region. The main part of the scattered neutrons are back-scattered from the floor. Finally, calculations on the graphite moderating BF{sub 3} -counter in the detailed torus environment were performed in order to assess the absolute response function under the influence of the reactor hall. The results show that the detector count rate will increase by only 5-7 % when the reactor hall is taken into account. With a stellarator generating 10{sup 12} to 10{sup 16} neutrons per second the detector count rate will be 2x10{sup 5} to 2x10{sup 9} neutrons per second.

  18. Simulation of neutron fluxes around the W7-X Stellarator

    International Nuclear Information System (INIS)

    Andersson, Jenny

    1999-12-01

    A new fusion experiment, the WENDELSTEIN 7-X Stellarator (W7-X), will be undertaken in Greifswald in Germany. Measurements of the neutron flux will provide information on fusion reaction rates and possibly also on ion temperatures as function of time. For this purpose moderating neutron counters will be designed, tested, calibrated and eventually used at W7-X. Extensive Monte-Carlo simulations have been performed in order to select the most suitable detector and moderator combination with a flat response function and highest achievable efficiency. Different detector configurations with different moderating materials have been tried out, showing that a 32 cm thick graphite moderating BF 3 -counter gives the desired flat response and sufficient efficiency. Neutron spectra calculations have been made for different torus models and the influence of floor, walls and ceiling (i.e. reactor hall) have been investigated. Presented results suggest that a more detailed torus model significantly reduces the number of neutron counts at the detector. Calculations including the reactor hall indicate a tendency of shifting the neutron spectra towards the thermal region. The main part of the scattered neutrons are back-scattered from the floor. Finally, calculations on the graphite moderating BF 3 -counter in the detailed torus environment were performed in order to assess the absolute response function under the influence of the reactor hall. The results show that the detector count rate will increase by only 5-7 % when the reactor hall is taken into account. With a stellarator generating 10 12 to 10 16 neutrons per second the detector count rate will be 2x10 5 to 2x10 9 neutrons per second

  19. MCNPX Simulation Study of STRAW Neutron Detectors - Summary Paper

    International Nuclear Information System (INIS)

    Mukhopadhyay, Sanjoy; Maurer, Richard; Mitchell, Stephen

    2010-01-01

    of the straws provides imaging capability with high enough resolution for radiation emitting sources. The prototype will provide the first aerial neutron detection system with directional sensitivity.

  20. Continuous energy Neutron Transport Monte Carlo Simulator Project: Decomposition of the neutron energy spectrum by target nuclei tagging

    Energy Technology Data Exchange (ETDEWEB)

    Barcellos, Luiz Felipe F.C.; Bodmann, Bardo E.J.; Vilhena, Marco T.M.B., E-mail: luizfelipe.fcb@gmail.com, E-mail: bardo.bodmann@ufrgs.br, E-mail: mtmbvilhena@gmail.com [Universidade Federal do Rio Grande do Sul (UFRGS), Porto Alegre, RS (Brazil). Grupo de Estudos Nucleares; Leite, Sergio Q. Bogado, E-mail: sbogado@ibest.com.br [Comissao Nacional de Energia Nuclear (CNEN), Rio de Janeiro, RJ (Brazil)

    2017-07-01

    In this work a Monte Carlo simulator with continuous energy is used. This simulator distinguishes itself by using the sum of three probability distributions to represent the neutron spectrum. Two distributions have known shape, but have varying population of neutrons in time, and these are the fission neutron spectrum (for high energy neutrons) and the Maxwell-Boltzmann distribution (for thermal neutrons). The third distribution has an a priori unknown and possibly variable shape with time and is determined from parametrizations of Monte Carlo simulation. It is common practice in neutron transport calculations, e.g. multi-group transport, to consider that the neutrons only lose energy with each scattering reaction and then to use a thermal group with a Maxwellian distribution. Such an approximation is valid due to the fact that for fast neutrons up-scattering occurrence is irrelevant, being only appreciable at low energies, i.e. in the thermal energy region, in which it can be regarded as a Maxwell-Boltzmann distribution for thermal equilibrium. In this work the possible neutron-matter interactions are simulated with exception of the up-scattering of neutrons. In order to preserve the thermal spectrum, neutrons are selected stochastically as being part of the thermal population and have an energy attributed to them taken from a Maxwellian distribution. It is then shown how this procedure can emulate the up-scattering effect by the increase in the neutron population kinetic energy. Since the simulator uses tags to identify the reactions it is possible not only to plot the distributions by neutron energy, but also by the type of interaction with matter and with the identification of the target nuclei involved in the process. This work contains some preliminary results obtained from a Monte Carlo simulator for neutron transport that is being developed at Federal University of Rio Grande do Sul. (author)

  1. Numerical simulation for neutron pinhole imaging in ICF

    International Nuclear Information System (INIS)

    Chen Faxin; Yang Jianlun; Wen Shuhuai

    2005-01-01

    Pinhole imaging of the neutron production in laser-driven inertial confinement fusion experiments can provide important information about performance of various capsule designs. In order to get good results in experiments, it is needed to judge performance of various pinhole designs qualitatively or quantitatively before experiment. Calculation of imaging can be simply separated into pinhole imaging and image spectral analysis. In this paper, pinhole imaging is discussed, codes for neutron pinhole imaging and image showing is programed. The codes can be used to provide theoretical foundation for pinhole designing and simulating data for image analysing. (authors)

  2. Simulation report for neutron guide and spectrometer layout at HANARO

    International Nuclear Information System (INIS)

    Cho, S. J.; Cho, Y. G.; Ryu, J. S.; Seong, B. S.; Lee, C. H.; Shin, J. W.

    2006-01-01

    A project called 'infrastructure construction for cold neutron research and utilization technique development' was launched in KAERI in July 2003, in order to raise a domestic basic science with an international level and elevate a international competitiveness for the bio-, nano- and informatics technology area. At the end of this project, 3 new instruments and 3 instruments to be moved will be installed in the guide hall. In order to accomplish this project until 2008, guide simulation should be performed for an effective use of expensive neutron to meet the requirements such as wavelengths and type of instruments, experimental space, interferences with other instruments

  3. Monte Carlo simulation of neutron counters for safeguards applications

    International Nuclear Information System (INIS)

    Looman, Marc; Peerani, Paolo; Tagziria, Hamid

    2009-01-01

    MCNP-PTA is a new Monte Carlo code for the simulation of neutron counters for nuclear safeguards applications developed at the Joint Research Centre (JRC) in Ispra (Italy). After some preliminary considerations outlining the general aspects involved in the computational modelling of neutron counters, this paper describes the specific details and approximations which make up the basis of the model implemented in the code. One of the major improvements allowed by the use of Monte Carlo simulation is a considerable reduction in both the experimental work and in the reference materials required for the calibration of the instruments. This new approach to the calibration of counters using Monte Carlo simulation techniques is also discussed.

  4. Monte Carlo Simulations of Neutron Oil well Logging Tools

    International Nuclear Information System (INIS)

    Azcurra, Mario

    2002-01-01

    Monte Carlo simulations of simple neutron oil well logging tools into typical geological formations are presented.The simulated tools consist of both 14 MeV pulsed and continuous Am-Be neutron sources with time gated and continuous gamma ray detectors respectively.The geological formation consists of pure limestone with 15% absolute porosity in a wide range of oil saturation.The particle transport was performed with the Monte Carlo N-Particle Transport Code System, MCNP-4B.Several gamma ray spectra were obtained at the detector position that allow to perform composition analysis of the formation.In particular, the ratio C/O was analyzed as an indicator of oil saturation.Further calculations are proposed to simulate actual detector responses in order to contribute to understand the relation between the detector response with the formation composition

  5. Monte Carlo simulations of neutron oil well logging tools

    International Nuclear Information System (INIS)

    Azcurra, Mario O.; Zamonsky, Oscar M.

    2003-01-01

    Monte Carlo simulations of simple neutron oil well logging tools into typical geological formations are presented. The simulated tools consist of both 14 MeV pulsed and continuous Am-Be neutron sources with time gated and continuous gamma ray detectors respectively. The geological formation consists of pure limestone with 15% absolute porosity in a wide range of oil saturation. The particle transport was performed with the Monte Carlo N-Particle Transport Code System, MCNP-4B. Several gamma ray spectra were obtained at the detector position that allow to perform composition analysis of the formation. In particular, the ratio C/O was analyzed as an indicator of oil saturation. Further calculations are proposed to simulate actual detector responses in order to contribute to understand the relation between the detector response with the formation composition. (author)

  6. Moving Up the CMMI Capability and Maturity Levels Using Simulation

    National Research Council Canada - National Science Library

    Raffo, David Mitchell; Wakeland, Wayne

    2008-01-01

    Process Simulation Modeling (PSIM) technology can be used to evaluate issues related to process strategy, process improvement, technology and tool adoption, project management and control, and process design...

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

    Energy Technology Data Exchange (ETDEWEB)

    Seeger, P.A.

    1995-09-01

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

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

    International Nuclear Information System (INIS)

    Seeger, P.A.

    1995-01-01

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

  9. A multi-group neutron noise simulator for fast reactors

    International Nuclear Information System (INIS)

    Tran, Hoai Nam; Zylbersztejn, Florian; Demazière, Christophe; Jammes, Christian; Filliatre, Philippe

    2013-01-01

    Highlights: • The development of a neutron noise simulator for fast reactors. • The noise equation is solved fully in a frequency-domain. • A good agreement with ERANOS on the static calculations. • Noise calculations induced by a localized perturbation of absorption cross section. - Abstract: A neutron noise simulator has been developed for fast reactors based on diffusion theory with multi-energy groups and several groups of delayed neutron precursors. The tool is expected to be applicable for core monitoring of fast reactors and also for other reactor types with hexagonal fuel assemblies. The noise sources are modeled through small stationary fluctuations of macroscopic cross sections, and the induced first order noise is solved fully in the frequency domain. Numerical algorithms are implemented for solving both the static and noise equations using finite differences for spatial discretization, where a hexagonal assembly is radially divided into finer triangular meshes. A coarse mesh finite difference (CMFD) acceleration has been used for accelerating the convergence of both the static and noise calculations. Numerical calculations have been performed for the ESFR core with 33 energy groups and 8 groups of delayed neutron precursors using the cross section data generated by the ERANOS code. The results of the static state have been compared with those obtained using ERANOS. The results show an adequate agreement between the two calculations. Noise calculations for the ESFR core have also been performed and demonstrated with an assumption of the perturbation of the absorption cross section located at the central fuel ring

  10. Investigation of propagation algorithms for ray-tracing simulation of polarized neutrons

    DEFF Research Database (Denmark)

    Bergbäck Knudsen, Erik; Tranum-Rømer, A.; Willendrup, Peter Kjær

    2014-01-01

    Ray-tracing of polarized neutrons faces a challenge when the neutron propagates through an inhomogeneous magnetic field. This affects simulations of novel instruments using encoding of energy or angle into the neutron spin. We here present a new implementation of propagation of polarized neutrons...

  11. Water Hammer Simulations of MMH Propellant - New Capability Demonstration of the Generalized Fluid Flow Simulation Program

    Science.gov (United States)

    Burkhardt, Z.; Ramachandran, N.; Majumdar, A.

    2017-01-01

    Fluid Transient analysis is important for the design of spacecraft propulsion system to ensure structural stability of the system in the event of sudden closing or opening of the valve. Generalized Fluid System Simulation Program (GFSSP), a general purpose flow network code developed at NASA/MSFC is capable of simulating pressure surge due to sudden opening or closing of valve when thermodynamic properties of real fluid are available for the entire range of simulation. Specifically GFSSP needs an accurate representation of pressure-density relationship in order to predict pressure surge during a fluid transient. Unfortunately, the available thermodynamic property programs such as REFPROP, GASP or GASPAK does not provide the thermodynamic properties of Monomethylhydrazine (MMH). This paper will illustrate the process used for building a customized table of properties of state variables from available properties and speed of sound that is required by GFSSP for simulation. Good agreement was found between the simulations and measured data. This method can be adopted for modeling flow networks and systems with other fluids whose properties are not known in detail in order to obtain general technical insight. Rigorous code validation of this approach will be done and reported at a future date.

  12. Simulations of a PSD Plastic Neutron Collar for Assaying Fresh Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Hausladen, Paul [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Newby, Jason [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); McElroy, Robert Dennis [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2016-11-01

    The potential performance of a notional active coincidence collar for assaying uranium fuel based on segmented detectors constructed from the new PSD plastic fast organic scintillator with pulse shape discrimination capability was investigated in simulation. Like the International Atomic Energy Agency's present Uranium Neutron Collar for LEU (UNCL), the PSD plastic collar would also function by stimulating fission in the 235U content of the fuel with a moderated 241Am/Li neutron source and detecting instances of induced fission via neutron coincidence counting. In contrast to the moderated detectors of the UNCL, the fast time scale of detection in the scintillator eliminates statistical errors due to accidental coincidences that limit the performance of the UNCL. However, the potential to detect a single neutron multiple times historically has been one of the properties of organic scintillator detectors that has prevented their adoption for international safeguards applications. Consequently, as part of the analysis of simulated data, a method was developed by which true neutron-neutron coincidences can be distinguished from inter-detector scatter that takes advantage of the position and timing resolution of segmented detectors. Then, the performance of the notional simulated coincidence collar was evaluated for assaying a variety of fresh fuels, including some containing burnable poisons and partial defects. In these simulations, particular attention was paid to the analysis of fast mode measurements. In fast mode, a Cd liner is placed inside the collar to shield the fuel from the interrogating source and detector moderators, thereby eliminating the thermal neutron flux that is most sensitive to the presence of burnable poisons that are ubiquitous in modern nuclear fuels. The simulations indicate that the predicted precision of fast mode measurements is similar to what can be achieved by the present UNCL in thermal mode. For example, the

  13. Expanding the modeling capabilities of the cognitive environment simulation

    International Nuclear Information System (INIS)

    Roth, E.M.; Mumaw, R.J.; Pople, H.E. Jr.

    1991-01-01

    The Nuclear Regulatory Commission has been conducting a research program to develop more effective tools to model the cognitive activities that underlie intention formation during nuclear power plant (NPP) emergencies. Under this program an artificial intelligence (AI) computer simulation called Cognitive Environment Simulation (CES) has been developed. CES simulates the cognitive activities involved in responding to a NPP accident situation. It is intended to provide an analytic tool for predicting likely human responses, and the kinds of errors that can plausibly arise under different accident conditions to support human reliability analysis. Recently CES was extended to handle a class of interfacing loss of coolant accidents (ISLOCAs). This paper summarizes the results of these exercises and describes follow-on work currently underway

  14. Capabilities of current wildfire models when simulating topographical flow

    Science.gov (United States)

    Kochanski, A.; Jenkins, M.; Krueger, S. K.; McDermott, R.; Mell, W.

    2009-12-01

    Accurate predictions of the growth, spread and suppression of wild fires rely heavily on the correct prediction of the local wind conditions and the interactions between the fire and the local ambient airflow. Resolving local flows, often strongly affected by topographical features like hills, canyons and ridges, is a prerequisite for accurate simulation and prediction of fire behaviors. In this study, we present the results of high-resolution numerical simulations of the flow over a smooth hill, performed using (1) the NIST WFDS (WUI or Wildland-Urban-Interface version of the FDS or Fire Dynamic Simulator), and (2) the LES version of the NCAR Weather Research and Forecasting (WRF-LES) model. The WFDS model is in the initial stages of development for application to wind flow and fire spread over complex terrain. The focus of the talk is to assess how well simple topographical flow is represented by WRF-LES and the current version of WFDS. If sufficient progress has been made prior to the meeting then the importance of the discrepancies between the predicted and measured winds, in terms of simulated fire behavior, will be examined.

  15. Reliability analysis of neutron transport simulation using Monte Carlo method

    International Nuclear Information System (INIS)

    Souza, Bismarck A. de; Borges, Jose C.

    1995-01-01

    This work presents a statistical and reliability analysis covering data obtained by computer simulation of neutron transport process, using the Monte Carlo method. A general description of the method and its applications is presented. Several simulations, corresponding to slowing down and shielding problems have been accomplished. The influence of the physical dimensions of the materials and of the sample size on the reliability level of results was investigated. The objective was to optimize the sample size, in order to obtain reliable results, optimizing computation time. (author). 5 refs, 8 figs

  16. Dynamic Simulation of Human Gait Model With Predictive Capability.

    Science.gov (United States)

    Sun, Jinming; Wu, Shaoli; Voglewede, Philip A

    2018-03-01

    In this paper, it is proposed that the central nervous system (CNS) controls human gait using a predictive control approach in conjunction with classical feedback control instead of exclusive classical feedback control theory that controls based on past error. To validate this proposition, a dynamic model of human gait is developed using a novel predictive approach to investigate the principles of the CNS. The model developed includes two parts: a plant model that represents the dynamics of human gait and a controller that represents the CNS. The plant model is a seven-segment, six-joint model that has nine degrees-of-freedom (DOF). The plant model is validated using data collected from able-bodied human subjects. The proposed controller utilizes model predictive control (MPC). MPC uses an internal model to predict the output in advance, compare the predicted output to the reference, and optimize the control input so that the predicted error is minimal. To decrease the complexity of the model, two joints are controlled using a proportional-derivative (PD) controller. The developed predictive human gait model is validated by simulating able-bodied human gait. The simulation results show that the developed model is able to simulate the kinematic output close to experimental data.

  17. CFD Simulation on Cooling Down of Beryllium Filters for Neutron Conditioning for Small Angle Neutron Scattering

    International Nuclear Information System (INIS)

    Azraf Azman; Shahrir Abdullah; Mohd Rizal Mamat

    2011-01-01

    The cryogenic system for cooling Beryllium filter utilizing liquid nitrogen was designed, fabricated, tested and installed at SANS instrument of TRIGA MARK II PUSPATI research reactor. A computational fluid dynamics (CFD) modeling was used to predict the cooling performance of the beryllium for optimization of neutron beam resolution and transmission. This paper presents the transient CFD results of temperature distributions via the thermal link to the beryllium and simulation of heat flux. The simulation data are also compared with the experimental results for the cooling time and distribution to the beryllium. (author)

  18. Simulation study of a hydrostat design for detecting underground leakage of water supply using neutron backscattering

    International Nuclear Information System (INIS)

    Kurosawa, Tadahiro; Nakamura, Takashi; Suzuki, Takashi; Okano, Yasuhiro; Chisaka, Haruo

    1998-01-01

    We have embarked upon the development of a new detection method for underground water leakage using a neutron backscattering system. We have estimated the performance capabilities of such a system using Monte Carlo simulation. It is indicated that a leak which results in 40% water content in the surrounding soils could be detected at depths of up to 40 cm from the surface to the center of the source of leakage. This new detection system could be useful as a hydrostat of underground water supply in noisy areas such as Tokyo, in place of presently-used hydrostats which are based on detection of changes in sound

  19. Dehydration of moulding sand in simulated casting process examined with neutron radiography

    Energy Technology Data Exchange (ETDEWEB)

    Schillinger, B., E-mail: Burkhard.Schillinger@frm2.tum.de [Technische Universitaet Muenchen, FRM II and Faculty for Physics E21, Lichtenbergstr. 1, 85748 Garching (Germany); Calzada, E. [Technische Universitaet Muenchen, FRM II and Faculty for Physics E21, Lichtenbergstr. 1, 85748 Garching (Germany); Eulenkamp, C.; Jordan, G.; Schmahl, W.W. [Ludwig-Maximilians-Universitaet Muenchen, Department fuer Geo- und Umweltwissenschaften, Sektion Kristallographie, Theresienstr. 41, 80333 Muenchen (Germany)

    2011-09-21

    Natural bentonites are an important material in the casting industry. Smectites as the main component of bentonites plasticize and stabilise sand moulds. Pore water as well as interlayer water within the smectites are lost as a function of time, location and temperature. Although rehydration of the smectites should be a reversible process, the industrially dehydrated smectites lose their capability to reabsorb water. This limits the number of possible process cycles of the mould material. A full understanding of the dehydration process would help to optimise the amount of fresh material to be added and thus save resources. A simulated metal casting was investigated with neutron radiography at the ANTARES neutron imaging facility of the FRM II reactor of Technische Universitaet Muenchen, Germany.

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

    NARCIS (Netherlands)

    Maleka, Peane P.; Maucec, Marko; de Meijer, Robert J.

    2011-01-01

    The deficiency in Monte Carlo simulations of coupled neutron-gamma-ray field was investigated by benchmarking two simulation codes with experimental data. Simulations showed better correspondence with the experimental data for gamma-ray transport only. In simulations, the neutron interactions with

  1. Preliminary investigations of Monte Carlo Simulations of neutron energy and LET spectra for fast neutron therapy facilities

    International Nuclear Information System (INIS)

    Kroc, T.K.

    2009-01-01

    No fast neutron therapy facility has been built with optimized beam quality based on a thorough understanding of the neutron spectrum and its resulting biological effectiveness. A study has been initiated to provide the information necessary for such an optimization. Monte Carlo studies will be used to simulate neutron energy spectra and LET spectra. These studies will be bench-marked with data taken at existing fast neutron therapy facilities. Results will also be compared with radiobiological studies to further support beam quality ptimization. These simulations, anchored by this data, will then be used to determine what parameters might be optimized to take full advantage of the unique LET properties of fast neutron beams. This paper will present preliminary work in generating energy and LET spectra for the Fermilab fast neutron therapy facility.

  2. Virtual experiments: the ultimate aim of neutron ray-tracing simulations

    Czech Academy of Sciences Publication Activity Database

    Lefmann, K.; Willendrup, P.K.; Šaroun, Jan

    2008-01-01

    Roč. 16, 3 & 4 (2008), s. 97-111 ISSN 1023-8166 Institutional research plan: CEZ:AV0Z10480505 Keywords : Monte Carlo simulations * neutron scattering * neutron instrumentation Subject RIV: BM - Solid Matter Physics ; Magnetism

  3. Use of high voltage electron microscope to simulate radiation damage by neutrons

    International Nuclear Information System (INIS)

    Mayer, R.M.

    1976-01-01

    The use of the high voltage electron microscope to simulate radiation damage by neutrons is briefly reviewed. This information is important in explaining how alloying affects void formation during neutron irradiation

  4. The pdk-100 enhances interpretation capabilities for pulsed neutron capture logs

    International Nuclear Information System (INIS)

    Randall, R.R.; Oliver, D.W.; Ferti, W.H.

    1986-01-01

    The PDK-100 is a new pulsed neutron logging system designed to measure Sigma (Σ), the macroscopic thermal neutron capture cross section. In addition to determining Σ, the system provides logging curves which are a measure of formation porosity and which furnish information concerning borehole conditions. This paper reviews the principles of operation of the PDK-100, and presents examples which illustrate the utility of the logging system. In addition, the progress of investigations into new parameters which can be derived with pulsed neutron logging data will be reported

  5. Capabilities of a mechanistic model for containment condenser simulation

    International Nuclear Information System (INIS)

    Broxtermann, Philipp; Cron, Daniel von der; Allelein, Hans-Josef

    2011-01-01

    In this paper the first application of the new containment COndenser MOdule (COMO) is being presented. COMO, just under development, represents a newly introduced part of the German containment code system COCOSYS and evaluates the contribution of containment condensers (CC) to passive containment cooling. Several next-generation Light Water Reactors (LWR) will feature innovative systems for passive heat removal during accidents from both the primary circuit and the containment. The passivity is based on natural driving forces only, such as gravity and natural circulation. To investigate the complex thermal-hydraulics and their propagation within a containment during accidents, containment code systems have been developed and validated against a wide variety of experiments. Furthermore, these codes are constantly improved to meet the intensified interest and knowledge in single phenomena and the technological evolution of the actual containment systems, e.g. the installation of passive devices. Accordingly, COCOSYS has been developed and is being continuously enhanced by the Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) and other partners like LRST. As has been shown by GRS and the calculation of the PANDA BC4 experiment, the interaction between a CC and the atmosphere of the surrounding vessel can be reproduced with the help of COCOSYS. However, up to now this has only been achieved on account of good knowledge of the outcome of the experiment, the user's skills and a complex input deck. The main goal of the newly introduced COMO is to improve the simulation of physical processes of CCs. This will be achieved by considering the passive driving forces which the CCs are based on. Up to now, a natural circulation within the CC tubes has been realized. The simulation of boiling conditions and the impact on the flow will be addressed in future studies. Additionally, the application of CCs to reactor simulation is being simplified and thus is supposed to reduce

  6. A primary simulation for residual stress neutron diffractometer

    International Nuclear Information System (INIS)

    Wang Shuying; Liu Lijuan; Sun Liangwei

    2012-01-01

    At present, neutron diffraction method is the unique and nondestructive method that can directly measure the residual stress distribution in deep materials and engineering components. It has an important application in engineering. A simulation of the flux at the position of the sample table was reported, and the resolution of the residual stress instrument was computed at the same time. The effect of the parameters of the second collimator on the flux at the sample position and the resolution of the instrument have been analyzed. The result indicated that the second collimator empress much on the neutron flux and the instrument resolution is well when the sample's diffraction angle is less than 120°. (authors)

  7. Simulation of processes in a SSNTD exposed by monoenergetic neutrons

    Science.gov (United States)

    Meyer, P.; Fromm, M.; Groetz, J. E.; Torrealba, F.; Chambaudet, A.

    1998-03-01

    The nuclear track technique is based on the registering of latent track ions in a solid state dielectric material (Solid State Nuclear Track Detector) which will be chemically etched in order to be observed and analysed using microscopic analysis tools. The purpose of this study is to observe how a polymeric detector irradiated by a neutron fluence can explain the intensity of ionizing radiations to which structures, electronic components or living organism are exposed. To do this, it is necessary to take into account on the one hand the production of ionizing particles in the detector, and on the other hand the efficacy of their detection. The model we propose must be able to link the number of etched tracks and their parameters to the neutron fluence. The use of both a Monte Carlo code which determines the number of recoil nuclei and a computed model of etched track parameters calculations is needed. The Monte Carlo code performs a simulation of neutron-nucleus elastic collision in the detector. We compute the number of recoil nuclei produced between the surface and a depth of 20 μm in the detector. The computer code assigns six pseudo-random numbers ( xE, zE, a, b, c, α) to each neutron having an energy EN and an angle of incidence and give the results of calculations. We used statistical tests (the moments of the origin of order n and of the average of order n, the χ2 test, the gap test, the Pearson test, the run-up run-down test and the serial test) to check the quality of our pseudo-random number generator. Finally, we will compare the calculated results of the response of the SSNTD (tracks cm -2) with the experimental ones to validate the simulation.

  8. Development of moderated neutron calibration fields simulating workplaces of MOX fuel facilities

    International Nuclear Information System (INIS)

    Tsujimura, Norio; Yoshida, Tadayoshi; Takada, Chie

    2005-01-01

    It is important for the MOX fuel facilities to control neutrons produced by the spontaneous fission of plutonium isotopes and those from (α,n) reactions between 18 O and α particles emitted by 238 Pu. Neutron dose meters should be calibrated for measuring these neutrons. We have developed moderated-neutron calibration fields employing a 252 Cf neutron source and moderators mainly for the characteristics evaluation and the calibration of neutron detectors used in MOX fuel facilities. Neutron energy spectrum can be adjusted by changing the position of the 252 Cf neutron source and combining different moderators to simulate the neutron field of the MOX fuel facility. This performance is realized owing to using an existing neutron irradiation room. (K. Yoshida)

  9. Evaluation of the current fast neutron flux monitoring instrumentation applied to LFR demonstrator ALFRED. Capabilities and limitations

    International Nuclear Information System (INIS)

    Lepore, Luigi; Remetti, Romolo; Cappelli, Mauro

    2015-01-01

    Among Gen IV projects for future nuclear power plants, Lead Fast Reactors (LFR) seem to be a very interesting solution due to their benefits in terms of fuel cycle, coolant-safety and waste management. The novelty of the matter causes some open issues about coolant chemical aspect, structural aspects, monitoring instrumentation, etc. Particularly hard neutron flux spectra would make traditional neutron instrumentation unfit to all reactor conditions, i.e. source, intermediate, and power range. Identification of new models of nuclear instrumentation specialized for LFR neutron flux monitoring asks for an accurate evaluation of the environment the sensor will work in. In this study, thermal-hydraulics and chemical conditions for LFR core environment will be assumed, as the neutron flux will be studied extensively by means of the Monte Carlo transport code MCNPX. The core coolant’s high temperature drastically reduces the candidate instrumentation, because only some kind of fission chambers and Self Powered Neutron Detectors can be operated in such an environment. This work aims to evaluate the capabilities of the available instrumentation (usually designed for Sodium Fast Reactors, SFRs) when exposed to the neutron spectrum derived from ALFRED, a pool-type small-power LFR project to demonstrate the feasibility of this technology into the European framework. This paper shows that such instruments do follow the power evolution, but they are not completely suitable to detect the whole range of reactor power. Some improvements are then possible in order to increase the signal-to-noise ratio, by optimizing each instrument in the range of reactor power, such to get the best solution. Some new detector designs are here proposed, and the possibilities for prototyping and testing by means of a fast reactor investigated. (author)

  10. Improvement of neutronic calculations on a Masurca core using adaptive mesh refinement capabilities

    International Nuclear Information System (INIS)

    Fournier, D.; Archier, P.; Le Tellier, R.; Suteau, C.

    2011-01-01

    The simulation of 3D cores with homogenized assemblies in transport theory remains time and memory consuming for production calculations. With a multigroup discretization for the energy variable and a discrete ordinate method for the angle, a system of about 10"4 coupled hyperbolic transport equations has to be solved. For these equations, we intend to optimize the spatial discretization. In the framework of the SNATCH solver used in this study, the spatial problem is dealt with by using a structured hexahedral mesh and applying a Discontinuous Galerkin Finite Element Method (DGFEM). This paper shows the improvements due to the development of Adaptive Mesh Refinement (AMR) methods. As the SNATCH solver uses a hierarchical polynomial basis, p−refinement is possible but also h−refinement thanks to non conforming capabilities. Besides, as the flux spatial behavior is highly dependent on the energy, we propose to adapt differently the spatial discretization according to the energy group. To avoid dealing with too many meshes, some energy groups are joined and share the same mesh. The different energy-dependent AMR strategies are compared to each other but also with the classical approach of a conforming and highly refined spatial mesh. This comparison is carried out on different quantities such as the multiplication factor, the flux or the current. The gain in time and memory is shown for 2D and 3D benchmarks coming from the ZONA2B experimental core configuration of the MASURCA mock-up at CEA Cadarache. (author)

  11. The MCUCN simulation code for ultracold neutron physics

    Science.gov (United States)

    Zsigmond, G.

    2018-02-01

    Ultracold neutrons (UCN) have very low kinetic energies 0-300 neV, thereby can be stored in specific material or magnetic confinements for many hundreds of seconds. This makes them a very useful tool in probing fundamental symmetries of nature (for instance charge-parity violation by neutron electric dipole moment experiments) and contributing important parameters for the Big Bang nucleosynthesis (neutron lifetime measurements). Improved precision experiments are in construction at new and planned UCN sources around the world. MC simulations play an important role in the optimization of such systems with a large number of parameters, but also in the estimation of systematic effects, in benchmarking of analysis codes, or as part of the analysis. The MCUCN code written at PSI has been extensively used for the optimization of the UCN source optics and in the optimization and analysis of (test) experiments within the nEDM project based at PSI. In this paper we present the main features of MCUCN and interesting benchmark and application examples.

  12. Benchmark of the neutronic model used in Maanshan compact simulator

    International Nuclear Information System (INIS)

    Hu, C.-H.; Gone, J.-K.; Ko, H.-T.

    2004-01-01

    The Maanshan compact simulator has adopted a three dimensional kinetic model CONcERT, which was developed by GP International Inc. (GPI) in 1991 for real-time neutronic analysis. Maanshan Nuclear Power Plant utilizes a Westinghouse nuclear steam supply system with three-loop pressurized water reactor. There are 157 fuel assemblies and 52 full-length Rod Cluster Control Assemblies in the reactor core. The control of excess reactivity and power peaking is provided by soluble boron in moderator and burnable absorber rods in fuel assemblies. The neutronic model of CONcERT is based on solving a modified time-dependent two-group diffusion equations coupled to the equations of six-group delayed neutron precursor concentrations. The validation of CONcERT for the Maanshan plant is separated into two groups. The first group compared (1) boron endpoints for different control bank inserted conditions, (2) control rod differential and integral worths and (3) temperature coefficients with the measurements in the Low Power Physical Test (LPPT). The second group compared critical boron concentration and power distribution in high power condition with the measurements. In addition, xenon and samarium equilibrium worths at different power levels as well as the time dependent changes of their worth after the reactor scram are illustrated. (author)

  13. Cronos 2: a neutronic simulation software for reactor core calculations

    International Nuclear Information System (INIS)

    Lautard, J.J.; Magnaud, C.; Moreau, F.; Baudron, A.M.

    1999-01-01

    The CRONOS2 software is that part of the SAPHYR code system dedicated to neutronic core calculations. CRONOS2 is a powerful tool for reactor design, fuel management and safety studies. Its modular structure and great flexibility make CRONOS2 an unique simulation tool for research and development for a wide variety of reactor systems. CRONOS2 is a versatile tool that covers a large range of applications from very fast calculations used in training simulators to time and memory consuming reference calculations needed to understand complex physical phenomena. CRONOS2 has a procedure library named CPROC that allows the user to create its own application environment fitted to a specific industrial use. (authors)

  14. A test-type hyper-thermal neutron generator for neutron capture therapy - estimation of neutron energy spectrum by simulation calculations and TOF experiments

    International Nuclear Information System (INIS)

    Sakurai, Yoshinori; Kobayashi, Tooru; Kobayashi, Katsuhei

    1999-01-01

    In order to clarify the irradiation characteristics of hyper-thermal neutrons and the feasibility of a hyper-thermal neutron irradiation field for neutron capture therapy, a 'test-type' hyper-thermal neutron generator was designed and made. Graphite of 6 cm thickness and 21 cm diameter was selected as the high temperature scatterer. The scatterer is heated up to 1200 deg. C maximum using molybdenum heaters. The radiation heat is shielded by reflectors of molybdenum and stainless steel. The temperature is measured using three R-type thermo-couples and controlled by a program controller. The total thickness of the generator is designed to be as thin as possible, 20 cm in maximum, in the standing point of the neutron beam intensity. The thermal stability, controllability and safety of the generator at high temperature employment were confirmed by the heating tests. As one of the experiments for the characteristics estimation, the neutron energy spectrum dependent on the scatterer temperature was measured by the TOF (time of flight) method using the LINAC neutron generator. The estimations by simulation calculations were also performed. From the experiment and calculation results, it was confirmed that the neutron temperature shifted higher as the scatterer temperature was higher. The prospect of the feasibility of the 'hyper-thermal neutron irradiation field for NCT' was opened from the estimation results of the generator characteristics by the simulation calculations and experiments

  15. Numerical relativity simulations of precessing binary neutron star mergers

    Science.gov (United States)

    Dietrich, Tim; Bernuzzi, Sebastiano; Brügmann, Bernd; Ujevic, Maximiliano; Tichy, Wolfgang

    2018-03-01

    We present the first set of numerical relativity simulations of binary neutron mergers that include spin precession effects and are evolved with multiple resolutions. Our simulations employ consistent initial data in general relativity with different spin configurations and dimensionless spin magnitudes ˜0.1 . They start at a gravitational-wave frequency of ˜392 Hz and cover more than 1 precession period and about 15 orbits up to merger. We discuss the spin precession dynamics by analyzing coordinate trajectories, quasilocal spin measurements, and energetics, by comparing spin aligned, antialigned, and irrotational configurations. Gravitational waveforms from different spin configuration are compared by calculating the mismatch between pairs of waveforms in the late inspiral. We find that precession effects are not distinguishable from nonprecessing configurations with aligned spins for approximately face-on binaries, while the latter are distinguishable from nonspinning configurations. Spin precession effects are instead clearly visible for approximately edge-on binaries. For the parameters considered here, precession does not significantly affect the characteristic postmerger gravitational-wave frequencies nor the mass ejection. Our results pave the way for the modeling of spin precession effects in the gravitational waveform from binary neutron star events.

  16. Challenges in coupled thermal-hydraulics and neutronics simulations for LWR safety analysis

    International Nuclear Information System (INIS)

    Ivanov, Kostadin; Avramova, Maria

    2007-01-01

    The simulation of nuclear power plant accident conditions requires three-dimensional (3D) modeling of the reactor core to ensure a realistic description of physical phenomena. The operational flexibility of Light Water Reactor (LWR) plants can be improved by utilizing accurate 3D coupled neutronics/thermal-hydraulics calculations for safety margins evaluations. There are certain requirements to the coupling of thermal-hydraulic system codes and neutron-kinetics codes that ought to be considered. The objective of these requirements is to provide accurate solutions in a reasonable amount of CPU time in coupled simulations of detailed operational transient and accident scenarios. These requirements are met by the development and implementation of six basic components of the coupling methodologies: ways of coupling (internal or external coupling); coupling approach (integration algorithm or parallel processing); spatial mesh overlays; coupled time-step algorithms; coupling numerics (explicit, semi-implicit and implicit schemes); and coupled convergence schemes. These principles of the coupled simulations are discussed in details along with the scientific issues associated with the development of appropriate neutron cross-section libraries for coupled code transient modeling. The current trends in LWR nuclear power generation and regulation as well as the design of next generation LWR reactor concepts along with the continuing computer technology progress stimulate further development of these coupled code systems. These efforts have been focused towards extending the analysis capabilities as well as refining the scale and level of detail of the coupling. This article analyses the coupled phenomena and modeling challenges on both global (assembly-wise) and local (pin-wise) levels. The issues related to the consistent qualification of coupled code systems as well as their application to different types of LWR transients are presented. Finally, the advances in numerical

  17. Simulated and measured neutron/gamma light output distribution for poly-energetic neutron/gamma sources

    Science.gov (United States)

    Hosseini, S. A.; Zangian, M.; Aghabozorgi, S.

    2018-03-01

    In the present paper, the light output distribution due to poly-energetic neutron/gamma (neutron or gamma) source was calculated using the developed MCNPX-ESUT-PE (MCNPX-Energy engineering of Sharif University of Technology-Poly Energetic version) computational code. The simulation of light output distribution includes the modeling of the particle transport, the calculation of scintillation photons induced by charged particles, simulation of the scintillation photon transport and considering the light resolution obtained from the experiment. The developed computational code is able to simulate the light output distribution due to any neutron/gamma source. In the experimental step of the present study, the neutron-gamma discrimination based on the light output distribution was performed using the zero crossing method. As a case study, 241Am-9Be source was considered and the simulated and measured neutron/gamma light output distributions were compared. There is an acceptable agreement between the discriminated neutron/gamma light output distributions obtained from the simulation and experiment.

  18. Testing the characteristics of a neutron detector array by Monte-Carlo simulations

    International Nuclear Information System (INIS)

    Timis, C.; Cruceru, I.; Sandu, M.; Borcea, C.; Buta, A.; Negoita, F.; Angelique, J.C.; Martin, T.; Peter, J.; Grevy, S.; Lienard, E.; Orr, N.A.

    1998-01-01

    The characteristics of the neutron detector array TONNERRE have been determined experimentally via preliminary tests with a 252 Cf source and by means of simulation using a modified version of the Monte-Carlo program of Cecil et al. Of particular interest is the intrinsic detection efficiency. As it is well known, the neutron detection efficiency for one element of the detector array, depends on the threshold for the light collection (bias) expressed in energy electron equivalent. The experimental efficiencies for five neutron energies and for a bias of 80 KeV ee are presented. The efficiencies for three thresholds and neutron energies between 1-10 MeV are simulated. The neutron energy is determined by TOF over a flight path, s, and the relative energy resolution is given as a function of σ s and σ t (the uncertainties in the flight path), s (uniform as a function of depth) and flight time, t. The mean time resolution was 1.13 ns which gives a TOF resolution of 1.48 ns. That gives a relative energy resolution which increases slowly from 2% at E n =1 MeV to 3.5% at 5 MeV. Position resolution along one module is 12 cm. To help boosting the efficiency, the elements can be arranged in two layers, but that complicates the analysis by enhancing the effects of cross-talk and out-scattering. Cross-talk is the familiar problem of one neutron creating signals in two separate detectors. In out-scattering, a neutron scatters from the non-active part of a detector and is then detected in a different detector with incorrect position and TOF. While methods exist for identifying and eliminating cross-talk events, there are no methods available for identifying out-scattered events. For the case of two layers and a bias of 80 KeV ee, simulated efficiency of two superposed elements versus neutron energy, the out-scattering probability and the probability of cross-talk are presented. The out-scattering probability comes mainly from events when neutrons scatter first on carbon nuclei

  19. Surface roughness effect on ultracold neutron interaction with a wall and implications for computer simulations

    OpenAIRE

    Steyerl, A.; Malik, S. S.; Desai, A. M.; Kaufman, C.

    2009-01-01

    We review the diffuse scattering and the loss coefficient in ultracold neutron reflection from slightly rough surfaces, report a surprising reduction in loss coefficient due to roughness, and discuss the possibility of transition from quantum treatment to ray optics. The results are used in a computer simulation of neutron storage in a recent neutron lifetime experiment that re-ported a large discrepancy of neutron lifetime with the current particle data value. Our partial re-analysis suggest...

  20. Monte Carlo Simulation on Compensated Neutron Porosity Logging in LWD With D-T Pulsed Neutron Generator

    International Nuclear Information System (INIS)

    Zhang Feng; Hou Shuang; Jin Xiuyun

    2010-01-01

    The process of neutron interaction induced by D-T pulsed neutron generator and 241 Am-Be source was simulated by using Monte Carlo method. It is concluded that the thermal neutron count descend exponentially as the spacing increasing. The smaller porosity was, the smaller the differences between the two sources were. When the porosity reached 40%, the ratio of thermal neutron count generated by D-T pulsed neutron source was much larger than that generated by 241 Am-Be neutron source, and its distribution range was wider. The near spacing selected was 20-30 cm, and that of far spacing was about 60-70 cm. The detection depth by using D-T pulsed neutron source was almost unchanged under condition of the same sapcing, and the sensitivity of measurement to the formation porosity decreases. The results showed that it can not only guarantee the statistic of count, but also improve detection sensitivity and depth at the same time of increasing spacing. Therefore, 241 Am-Be neutron source can be replaced by D-T neutron tube in LWD tool. (authors)

  1. COUPLED FREE AND DISSOLVED PHASE TRANSPORT: NEW SIMULATION CAPABILITIES AND PARAMETER INVERSION

    Science.gov (United States)

    The vadose zone free-phase simulation capabilities of the US EPA Hydrocarbon Spill Screening Model (HSSM) (Weaver et al., 1994) have been linked with the 3-D multi-species dissolved-phase contaminant transport simulator MT3DMS (Zheng and Wang, 1999; Zheng, 2005). The linkage pro...

  2. Cryostat insert with gas loading capabilities for use in neutron diffraction studies

    International Nuclear Information System (INIS)

    Koehler, C. F. III; Larese, J. Z.

    2000-01-01

    We describe a versatile insert for use with the ''Standard Orange cryostat'' commonly found at neutron scattering facilities worldwide. Its design permits condensable gases to be introduced into a low-temperature sample cell through a vacuum-insulated, fill-line capillary. It also allows the top-flange-to-sample-cell distance to be easily adjusted and provides for auxillary heating of the fill-line capillary. (c) 2000 American Institute of Physics

  3. Cryostat insert with gas loading capabilities for use in neutron diffraction studies

    Energy Technology Data Exchange (ETDEWEB)

    Koehler, C. F. III [Chemistry Department, Brookhaven National Laboratory, Upton, New York 11973-5000 (United States); Larese, J. Z. [Chemistry Department, Brookhaven National Laboratory, Upton, New York 11973-5000 (United States)

    2000-01-01

    We describe a versatile insert for use with the ''Standard Orange cryostat'' commonly found at neutron scattering facilities worldwide. Its design permits condensable gases to be introduced into a low-temperature sample cell through a vacuum-insulated, fill-line capillary. It also allows the top-flange-to-sample-cell distance to be easily adjusted and provides for auxillary heating of the fill-line capillary. (c) 2000 American Institute of Physics.

  4. Design and performance of vacuum capable detector electronics for linear position sensitive neutron detectors

    International Nuclear Information System (INIS)

    Riedel, R.A.; Cooper, R.G.; Funk, L.L.; Clonts, L.G.

    2012-01-01

    We describe the design and performance of electronics for linear position sensitive neutron detectors. The eight tube assembly requires 10 W of power and can be controlled via digital communication links. The electronics can be used without modification in vacuum. Using a transimpedance amplifier and gated integration, we achieve a highly linear system with coefficient of determinations of 0.9999 or better. Typical resolution is one percent of tube length.

  5. Design and performance of vacuum capable detector electronics for linear position sensitive neutron detectors

    Energy Technology Data Exchange (ETDEWEB)

    Riedel, R.A., E-mail: riedelra@ornl.gov [Oak Ridge National Laboratories, Oak Ridge, TN 37830 (United States); Cooper, R.G.; Funk, L.L.; Clonts, L.G. [Oak Ridge National Laboratories, Oak Ridge, TN 37830 (United States)

    2012-02-01

    We describe the design and performance of electronics for linear position sensitive neutron detectors. The eight tube assembly requires 10 W of power and can be controlled via digital communication links. The electronics can be used without modification in vacuum. Using a transimpedance amplifier and gated integration, we achieve a highly linear system with coefficient of determinations of 0.9999 or better. Typical resolution is one percent of tube length.

  6. Los Alamos neutron science center nuclear weapons stewardship and unique national scientific capabilities

    Energy Technology Data Exchange (ETDEWEB)

    Schoenberg, Kurt F [Los Alamos National Laboratory

    2010-12-15

    This presentation gives an overview of the Los Alamos Neutron Science Center (LANSCE) and its contributions to science and the nuclear weapons program. LANSCE is made of multiple experimental facilities (the Lujan Center, the Weapons Neutron Research facility (WNR), the Ultra-Cold Neutron facility (UCN), the proton Radiography facility (pRad) and the Isotope Production Facility (IPF)) served by the its kilometer long linear accelerator. Several research areas are supported, including materials and bioscience, nuclear science, materials dynamics, irradiation response and medical isotope production. LANSCE is a national user facility that supports researchers worldwide. The LANSCE Risk Mitigation program is currently in progress to update critical accelerator equipment to help extend the lifetime of LANSCE as a key user facility. The Associate Directorate of Business Sciences (ADBS) plays an important role in the continued success of LANSCE. This includes key procurement support, human resource support, technical writing support, and training support. LANSCE is also the foundation of the future signature facility MARIE (Matter-Radiation Interactions in Extremes).

  7. Los Alamos neutron science center nuclear weapons stewardship and unique national scientific capabilities

    International Nuclear Information System (INIS)

    Schoenberg, Kurt F.

    2010-01-01

    This presentation gives an overview of the Los Alamos Neutron Science Center (LANSCE) and its contributions to science and the nuclear weapons program. LANSCE is made of multiple experimental facilities (the Lujan Center, the Weapons Neutron Research facility (WNR), the Ultra-Cold Neutron facility (UCN), the proton Radiography facility (pRad) and the Isotope Production Facility (IPF)) served by the its kilometer long linear accelerator. Several research areas are supported, including materials and bioscience, nuclear science, materials dynamics, irradiation response and medical isotope production. LANSCE is a national user facility that supports researchers worldwide. The LANSCE Risk Mitigation program is currently in progress to update critical accelerator equipment to help extend the lifetime of LANSCE as a key user facility. The Associate Directorate of Business Sciences (ADBS) plays an important role in the continued success of LANSCE. This includes key procurement support, human resource support, technical writing support, and training support. LANSCE is also the foundation of the future signature facility MARIE (Matter-Radiation Interactions in Extremes).

  8. Strong neutron sources - How to cope with weapon material production capabilities of fusion and spallation neutron sources?

    International Nuclear Information System (INIS)

    Englert, M.; Franceschini, G.; Liebert, W.

    2013-01-01

    In this article we investigate the potential and relevance for weapon material production in future fusion power plants and spallation neutron sources (SNS) and sketch what should be done to strengthen these technologies against a non-peaceful use. It is shown that future commercial fusion reactors may have military implications: first, they provide an easy source of tritium for weapons, an element that does not fall under safeguards and for which diversion from a plant could probably not be detected even if some tritium accountancy is implemented. Secondly, large fusion reactors - even if not designed for fissile material breeding - could easily produce several hundred kg Pu per year with high weapon quality and very low source material requirements. If fusion-only reactors will prevail over fission-fusion hybrids in the commercialization phase of fusion technology, the safeguard challenge will be more of a legal than of a technical nature. In pure fusion reactors (and in most SNS) there should be no nuclear material present at any time by design. The presence of undeclared nuclear material would indicate a military use of the plant. This fact offers a clear-cut detection criterion for a covert use of a declared facility. Another important point is that tritium does not fall under the definition of 'nuclear material', so a pure fusion reactor or a SNS that do not use nuclear materials are not directly falling under any international non-proliferation treaty requirements. Non-proliferation treaties have to be amended to take into account that fact. (A.C.)

  9. Monte Carlo simulation study of the muon-induced neutron flux at LNGS

    International Nuclear Information System (INIS)

    Persiani, R.; Garbini, M.; Massoli, F.; Sartorelli, G; Selvi, M.

    2011-01-01

    Muon-induced neutrons are ultimate background for all the experiments searching for rare events in underground laboratories. Several measurements and simulations were performed concerning the neutron production and propagation but there are disagreements between experimental data and simulations. In this work we present our Monte-Carlo simulation study, based on Geant4, to estimate the muon-induced neutron flux at LNGS. The obtained integral flux of neutrons above 1 MeV is 2.31 x 10 -10 n/cm 2 /s.

  10. Simulation for sodium-24 production using cyclic neutron activation analysis

    International Nuclear Information System (INIS)

    Ahamed, O. M. H.

    2012-04-01

    The cyclic neutron activation analysis is a method for elemental analysis which is preferred to use short-lived radio-nuclides. In recent years this method became a new application for radioisotope production especially in low power research reactors. In this study instrumental cyclic neutron activation analysis was used for 2 4N a production using 2 3N a (n, γ) 2 4N a reaction. The simplified westcott convention method is used neutron activation analysis in a research reactor. The method takes into account all corrections that can affect that yield created. In this work a model was devolving for calculations through this simplified Westcott convention method by using C++ program and selecting good parameters that can produce the expected activity. The simulation is used for the theoretical yield calculated has been validated by data from the 1 77L u production used for theoretical yield which it gives the approached result had been obtained by FORTRAN 90 from literature (8). The results were achieved for expected activity at full power (1 x 10 1 2n cm -2 s - 1 ) and half power (5 x 10 11 cm -2 s -1 ) for research reactor MNSR. The activity at full power was equal to about twice the activity at half power ( 49±7, 24.9 ± MBq/g), respectively. The irradiation parameters selected were irradiation time 4 min and decay time 12 min. Sample weight was 50 mg at 12 numbers of cycles, when the K-factor was equal to 1.74. This work is considered as first step for production of 2 4N a which can use such parameters experimentally. It is then possible to compare the expected activity with measured activity. (Author)

  11. Monte Carlo N-particle simulation of neutron-based sterilisation of anthrax contamination.

    Science.gov (United States)

    Liu, B; Xu, J; Liu, T; Ouyang, X

    2012-10-01

    To simulate the neutron-based sterilisation of anthrax contamination by Monte Carlo N-particle (MCNP) 4C code. Neutrons are elementary particles that have no charge. They are 20 times more effective than electrons or γ-rays in killing anthrax spores on surfaces and inside closed containers. Neutrons emitted from a (252)Cf neutron source are in the 100 keV to 2 MeV energy range. A 2.5 MeV D-D neutron generator can create neutrons at up to 10(13) n s(-1) with current technology. All these enable an effective and low-cost method of killing anthrax spores. There is no effect on neutron energy deposition on the anthrax sample when using a reflector that is thicker than its saturation thickness. Among all three reflecting materials tested in the MCNP simulation, paraffin is the best because it has the thinnest saturation thickness and is easy to machine. The MCNP radiation dose and fluence simulation calculation also showed that the MCNP-simulated neutron fluence that is needed to kill the anthrax spores agrees with previous analytical estimations very well. The MCNP simulation indicates that a 10 min neutron irradiation from a 0.5 g (252)Cf neutron source or a 1 min neutron irradiation from a 2.5 MeV D-D neutron generator may kill all anthrax spores in a sample. This is a promising result because a 2.5 MeV D-D neutron generator output >10(13) n s(-1) should be attainable in the near future. This indicates that we could use a D-D neutron generator to sterilise anthrax contamination within several seconds.

  12. Monte-Carlo simulation on the cold neutron guides at CARR

    International Nuclear Information System (INIS)

    Guo Liping; Wang Hongli; Yang Tonghua; Cheng Zhixu; Liu Yi

    2003-01-01

    The designs of the two cold neutron guides to be built at China Advanced Research Reactor (CARR) are simulated with Monte-Carlo simulation software VITESS. Various parameters of the guides, e.g. transmission efficiency, neutron flux, divergence, etc., are obtained. (author)

  13. Cadmium-emitter self-powered thermal neutron detector performance characterization & reactor power tracking capability experiments performed in ZED-2

    Energy Technology Data Exchange (ETDEWEB)

    LaFontaine, M.W., E-mail: physics@execulink.com [LaFontaine Consulting, Kitchener, Ontario (Canada); Zeller, M.B. [Canadian Nuclear Laboratories, Chalk River, Ontario (Canada); Nielsen, K. [Royal Military College of Canada, SLOWPOKE-2 Reactor, Kingston, Ontario (Canada)

    2014-07-01

    Cadmium-emitter self-powered thermal neutron flux detectors (SPDs), are typically used for flux monitoring and control applications in low temperature, test reactors such as the SLOWPOKE-2. A collaborative program between Atomic Energy of Canada, academia (Royal Military College of Canada (RMCC)) and industry (LaFontaine Consulting) was initiated to characterize the incore performance of a typical Cd-emitter SPD; and to obtain a definitive measure of the capability of the detector to track changes in reactor power in real time. Prior to starting the experiment proper, Chalk River Laboratories' ZED-2 was operated at low power (5 watts nominal) to verify the predicted moderator critical height. Test measurements were then performed with the vertical center of the SPD emitter positioned at the vertical mid-plane of the ZED-2 reactor core. Measurements were taken with the SPD located at lattice position L0 (near center), and repeated at lattice position P0 (in D{sub 2}O reflector). An ionization chamber (part of the ZED-2 control instrumentation) monitored reactor power at a position located on the south side of the outside wall of the reactor's calandria. These experiments facilitated measurement of the absolute thermal neutron sensitivity of the subject Cd-emitter SPD, and validated the power tracking capability of said SPD. Procedural details of the experiments, data, calculations and associated graphs, are presented and discussed. (author)

  14. Simulation of a high energy neutron irradiation facility at beamline 11 of the China Spallation Neutron Source

    Energy Technology Data Exchange (ETDEWEB)

    Tairan, Liang [School of Physics and Electronic Information Inner Mongolia University for the Nationalities, Tongliao 028043 (China); Zhiduo, Li [Dongguan Branch, Institute of High Energy Physics, CAS, Beijing 100049 (China); Wen, Yin, E-mail: wenyin@aphy.iphy.ac.cn [Dongguan Branch, Institute of High Energy Physics, CAS, Beijing 100049 (China); Institute of Physics, CAS, P.O. Box 603, Beijing 100190 (China); Fei, Shen [Dongguan Branch, Institute of High Energy Physics, CAS, Beijing 100049 (China); Quanzhi, Yu [Dongguan Branch, Institute of High Energy Physics, CAS, Beijing 100049 (China); Institute of Physics, CAS, P.O. Box 603, Beijing 100190 (China); Tianjiao, Liang [Dongguan Branch, Institute of High Energy Physics, CAS, Beijing 100049 (China)

    2017-07-11

    The China Spallation Neutron Source (CSNS) will accommodate 20 neutron beamlines at its first target station. These beamlines serve different purposes, and beamline 11 is designed to analyze the degraded models and damage mechanisms, such as Single Event Effects in electronic components and devices for aerospace electronic systems. This paper gives a preliminary discussion on the scheme of a high energy neutron irradiation experiment at the beamline 11 shutter based on the Monte Carlo simulation method. The neutron source term is generated by calculating the neutrons scattering into beamline 11 with a model that includes the target-moderator-reflector area. Then, the neutron spectrum at the sample position is obtained. The intensity of neutrons with energy of hundreds of MeV is approximately 1E8 neutron/cm{sup 2}/s, which is useful for experiments. The displacement production rate and gas productions are calculated for common materials such as tungsten, tantalum and SS316. The results indicate that the experiment can provide irradiation dose rate ranges from 1E-5 to 1E-4 dpa per operating year. The residual radioactivity is also calculated for regular maintenance work. These results give the basic reference for the experimental design.

  15. Design and simulation of an optimized e-linac based neutron source for BNCT research

    International Nuclear Information System (INIS)

    Durisi, E.; Alikaniotis, K.; Borla, O.; Bragato, F.; Costa, M.; Giannini, G.; Monti, V.; Visca, L.; Vivaldo, G.; Zanini, A.

    2015-01-01

    The paper is focused on the study of a novel photo-neutron source for BNCT preclinical research based on medical electron Linacs. Previous studies by the authors already demonstrated the possibility to obtain a mixed thermal and epithermal neutron flux of the order of 10"7 cm"−"2 s"−"1. This paper investigates possible Linac’s modifications and a new photo-converter design to rise the neutron flux above 5 10"7 cm"−"2 s"−"1, also reducing the gamma contamination. - Highlights: • Proposal of a mixed thermal and epithermal (named hyperthermal) neutron source based on medical high energy electron Linac. • Photo-neutron production via Giant Dipole Resonance on high Z materials. • MCNP4B-GN simulations to design the photo-converter geometry maximizing the hyperthermal neutron flux and minimizing the fast neutron and gamma contaminations. Hyperthermal neutron field suitable for BNCT preclinical research.

  16. ADVANCED SIMULATION CAPABILITY FOR ENVIRONMENTAL MANAGEMENT- CURRENT STATUS AND PHASE II DEMONSTRATION RESULTS

    Energy Technology Data Exchange (ETDEWEB)

    Seitz, R.

    2013-02-26

    The U.S. Department of Energy (USDOE) Office of Environmental Management (EM), Office of Soil and Groundwater, is supporting development of the Advanced Simulation Capability for Environmental Management (ASCEM). ASCEM is a state-of-the-art scientific tool and approach for understanding and predicting contaminant fate and transport in natural and engineered systems. The modular and open source high-performance computing tool facilitates integrated approaches to modeling and site characterization that enable robust and standardized assessments of performance and risk for EM cleanup and closure activities. The ASCEM project continues to make significant progress in development of computer software capabilities with an emphasis on integration of capabilities in FY12. Capability development is occurring for both the Platform and Integrated Toolsets and High-Performance Computing (HPC) Multiprocess Simulator. The Platform capabilities provide the user interface and tools for end-to-end model development, starting with definition of the conceptual model, management of data for model input, model calibration and uncertainty analysis, and processing of model output, including visualization. The HPC capabilities target increased functionality of process model representations, toolsets for interaction with Platform, and verification and model confidence testing. The Platform and HPC capabilities are being tested and evaluated for EM applications in a set of demonstrations as part of Site Applications Thrust Area activities. The Phase I demonstration focusing on individual capabilities of the initial toolsets was completed in 2010. The Phase II demonstration completed in 2012 focused on showcasing integrated ASCEM capabilities. For Phase II, the Hanford Site deep vadose zone (BC Cribs) served as an application site for an end-to-end demonstration of capabilities, with emphasis on integration and linkages between the Platform and HPC components. Other demonstrations

  17. Feasibility of the integration of CRONOS, a 3-D neutronics code, into real-time simulators

    International Nuclear Information System (INIS)

    Ragusa, J.C.

    2001-01-01

    In its effort to contribute to nuclear power plant safety, CEA proposes the integration of an engineering grade 3-D neutronics code into a real-time plant analyser. This paper describes the capabilities of the neutronics code CRONOS to achieve a fast running performance. First, we will present current core models in simulators and explain their drawbacks. Secondly, the mean features of CRONOS's spatial-kinetics methods will be reviewed. We will then present an optimum core representation with respect to mesh size, choice of finite elements (FE) basis and execution time, for accurate results as well as the multi 1-D thermal-hydraulics (T/H) model developed to take into account 3-D effects in updating the cross-sections. A Main Steam Line Break (MSLB) End-of-Life (EOL) Hot-Zero-Power (HZP) accident will be used as an example, before we conclude with the perspectives of integrating CRONOS's 3-D core model into real-time simulators. (author)

  18. Feasibility of the integration of CRONOS, a 3-D neutronics code, into real-time simulators

    Energy Technology Data Exchange (ETDEWEB)

    Ragusa, J.C. [CEA Saclay, Dept. de Mecanique et de Technologie, 91 - Gif-sur-Yvette (France)

    2001-07-01

    In its effort to contribute to nuclear power plant safety, CEA proposes the integration of an engineering grade 3-D neutronics code into a real-time plant analyser. This paper describes the capabilities of the neutronics code CRONOS to achieve a fast running performance. First, we will present current core models in simulators and explain their drawbacks. Secondly, the mean features of CRONOS's spatial-kinetics methods will be reviewed. We will then present an optimum core representation with respect to mesh size, choice of finite elements (FE) basis and execution time, for accurate results as well as the multi 1-D thermal-hydraulics (T/H) model developed to take into account 3-D effects in updating the cross-sections. A Main Steam Line Break (MSLB) End-of-Life (EOL) Hot-Zero-Power (HZP) accident will be used as an example, before we conclude with the perspectives of integrating CRONOS's 3-D core model into real-time simulators. (author)

  19. General relativistic magnetohydrodynamic simulations of binary neutron star mergers forming a long-lived neutron star

    Science.gov (United States)

    Ciolfi, Riccardo; Kastaun, Wolfgang; Giacomazzo, Bruno; Endrizzi, Andrea; Siegel, Daniel M.; Perna, Rosalba

    2017-03-01

    Merging binary neutron stars (BNSs) represent the ultimate targets for multimessenger astronomy, being among the most promising sources of gravitational waves (GWs), and, at the same time, likely accompanied by a variety of electromagnetic counterparts across the entire spectrum, possibly including short gamma-ray bursts (SGRBs) and kilonova/macronova transients. Numerical relativity simulations play a central role in the study of these events. In particular, given the importance of magnetic fields, various aspects of this investigation require general relativistic magnetohydrodynamics (GRMHD). So far, most GRMHD simulations focused the attention on BNS mergers leading to the formation of a hypermassive neutron star (NS), which, in turn, collapses within few tens of ms into a black hole surrounded by an accretion disk. However, recent observations suggest that a significant fraction of these systems could form a long-lived NS remnant, which will either collapse on much longer time scales or remain indefinitely stable. Despite the profound implications for the evolution and the emission properties of the system, a detailed investigation of this alternative evolution channel is still missing. Here, we follow this direction and present a first detailed GRMHD study of BNS mergers forming a long-lived NS. We consider magnetized binaries with different mass ratios and equations of state and analyze the structure of the NS remnants, the rotation profiles, the accretion disks, the evolution and amplification of magnetic fields, and the ejection of matter. Moreover, we discuss the connection with the central engine of SGRBs and provide order-of-magnitude estimates for the kilonova/macronova signal. Finally, we study the GW emission, with particular attention to the post-merger phase.

  20. Desert Rats 2011 Mission Simulation: Effects of Microgravity Operational Modes on Fields Geology Capabilities

    Science.gov (United States)

    Bleacher, Jacob E.; Hurtado, J. M., Jr.; Meyer, J. A.

    2012-01-01

    Desert Research and Technology Studies (DRATS) is a multi-year series of NASA tests that deploy planetary surface hardware and exercise mission and science operations in difficult conditions to advance human and robotic exploration capabilities. DRATS 2011 (Aug. 30-Sept. 9, 2011) tested strategies for human exploration of microgravity targets such as near-Earth asteroids (NEAs). Here we report the crew perspective on the impact of simulated microgravity operations on our capability to conduct field geology.

  1. Numerical simulation of binary black hole and neutron star mergers

    Energy Technology Data Exchange (ETDEWEB)

    Kastaun, W.; Rezzolla, L. [Albert Einstein Institut, Potsdam-Golm (Germany)

    2016-11-01

    One of the last predictions of general relativity that still awaits direct observational confirmation is the existence of gravitational waves. Those fluctuations of the geometry of space and time are expected to travel with the speed of light and are emitted by any accelerating mass. Only the most violent events in the universe, such as mergers of two black holes or neutron stars, produce gravitational waves strong enough to be measured. Even those waves are extremely weak when arriving at Earth, and their detection is a formidable technological challenge. In recent years sufficiently sensitive detectors became operational, such as GEO600, Virgo, and LIGO. They are expected to observe around 40 events per year. To interpret the observational data, theoretical modeling of the sources is a necessity, and requires numerical simulations of the equations of general relativity and relativistic hydrodynamics. Such computations can only be carried out on large scale supercomputers, given that many scenarios need to be simulated, each of which typically occupies hundreds of CPU cores for a week. Our main goal is to predict the gravitational wave signal from the merger of two compact objects. Comparison with future observations will provide important insights into the fundamental forces of nature in regimes that are impossible to recreate in laboratory experiments. The waveforms from binary black hole mergers would allow one to test the correctness of general relativity in previously inaccessible regimes. The signal from binary neutron star mergers will provide input for nuclear physics, because the signal depends strongly on the unknown properties of matter at the ultra high densities inside neutron stars, which cannot be observed in any other astrophysical scenario. Besides mergers, we also want to improve the theoretical models of close encounters between black holes. A gravitational wave detector with even higher sensitivity, the Einstein Telescope, is already in the

  2. Direct ion storage dosimetry systems for photon, beta and neutron radiation with instant readout capabilities

    International Nuclear Information System (INIS)

    Wernli, C.; Kahilainen, J.

    2001-01-01

    The direct ion storage (DIS) dosemeter is a new type of electronic dosemeter from which the dose information for both H p (10) and H p (0.07) can be obtained instantly at the workplace by using an electronic reader unit. The number of readouts is unlimited and the stored information is not affected by the readout procedure. The accumulated dose can also be electronically reset by authorised personnel. The DIS dosemeter represents a potential alternative for replacing the existing film and thermoluminescence dosemeters (TLDs) used in occupational monitoring due to its ease of use and low operating costs. The standard version for normal photon and beta dosimetry, as well as a developmental version for neutron dosimetry, have been characterised in several field studies. Two new small size variations are also introduced including a contactless readout device and a militarised version optimised for field use. (author)

  3. Simulation of neutrons and gamma pulse signal and research on the pulse shape discrimination technology

    International Nuclear Information System (INIS)

    Zuo Guangxia; He Bin; Xu Peng; Qiu Xiaolin; Ma Wenyan; Li Sufen

    2012-01-01

    In neutrons detection, it is important to discriminate the neutron signals from the gamma-ray background. In this article, simulation of neutrons and gamma pulse signals is developed based on the LabVIEW platform. Two digital algorithms of the charge comparison method and the pulse duration time method are realized using 10000 simulation signals. Experimental results show that neutron and gamma pulse signals can be discriminated by the two methods, and the pulse duration time method is better than the charge comparison method. (authors)

  4. Light ions cyclotron bombardment to simulate fast neutron radiation damage in nuclear materials

    International Nuclear Information System (INIS)

    Segura, E.; Lucki, G.; Aguiar, D.

    1984-01-01

    The applicability and limitations of the use of cyclotron light ions bombardment to simulate the effects of the neutron irradiation are presented. Light ions with energies of about 10 MeV are capable to produce homogeneous damage in specimens suitable for measuring bulk mechanical properties although their low damage rate of 10 -5 dpa.sec -1 limit the dose range to a few dpa. On the other hand, cyclotron alpha particle implantation provides a fast and convenient way of introducing helium with a minimum of side effects so that we can take advantage of this technique to get better understanding of the mechanism by which this insoluble gas produces high temperature embrittlement. Some experimental details such as dimensions and cooling techniques are described. Finally a description of the infrastructure for cyclotron alpha particle implantation and a creep-test facility of the Division of Radiation Damage at IPEN-CNEN/SP are presented. (Author) [pt

  5. Use of Aria to simulate laser weld pool dynamics for neutron generator production.

    Energy Technology Data Exchange (ETDEWEB)

    Noble, David R.; Notz, Patrick K.; Martinez, Mario J.; Kraynik, Andrew Michael

    2007-09-01

    This report documents the results for the FY07 ASC Integrated Codes Level 2 Milestone number 2354. The description for this milestone is, 'Demonstrate level set free surface tracking capabilities in ARIA to simulate the dynamics of the formation and time evolution of a weld pool in laser welding applications for neutron generator production'. The specialized boundary conditions and material properties for the laser welding application were implemented and verified by comparison with existing, two-dimensional applications. Analyses of stationary spot welds and traveling line welds were performed and the accuracy of the three-dimensional (3D) level set algorithm is assessed by comparison with 3D moving mesh calculations.

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

    Energy Technology Data Exchange (ETDEWEB)

    Gerhard Strydom; Cristian Rabiti; Andrea Alfonsi

    2012-10-01

    PHISICS is a neutronics code system currently under development at the Idaho National Laboratory (INL). Its goal is to provide state of the art simulation capability to reactor designers. The different modules for PHISICS currently under development are a nodal and semi-structured transport core solver (INSTANT), a depletion module (MRTAU) and a cross section interpolation (MIXER) module. The INSTANT module is the most developed of the mentioned above. Basic functionalities are ready to use, but the code is still in continuous development to extend its capabilities. This paper reports on the effort of coupling the nodal kinetics code package PHISICS (INSTANT/MRTAU/MIXER) to the thermal hydraulics system code RELAP5-3D, to enable full core and system modeling. This will enable the possibility to model coupled (thermal-hydraulics and neutronics) problems with more options for 3D neutron kinetics, compared to the existing diffusion theory neutron kinetics module in RELAP5-3D (NESTLE). In the second part of the paper, an overview of the OECD/NEA MHTGR-350 MW benchmark is given. This benchmark has been approved by the OECD, and is based on the General Atomics 350 MW Modular High Temperature Gas Reactor (MHTGR) design. The benchmark includes coupled neutronics thermal hydraulics exercises that require more capabilities than RELAP5-3D with NESTLE offers. Therefore, the MHTGR benchmark makes extensive use of the new PHISICS/RELAP5-3D coupling capabilities. The paper presents the preliminary results of the three steady state exercises specified in Phase I of the benchmark using PHISICS/RELAP5-3D.

  7. Soft error rate simulation and initial design considerations of neutron intercepting silicon chip (NISC)

    Science.gov (United States)

    Celik, Cihangir

    -scale technologies. Prevention of SEEs has been studied and applied in the semiconductor industry by including radiation protection precautions in the system architecture or by using corrective algorithms in the system operation. Decreasing 10B content (20%of natural boron) in the natural boron of Borophosphosilicate glass (BPSG) layers that are conventionally used in the fabrication of semiconductor devices was one of the major radiation protection approaches for the system architecture. Neutron interaction in the BPSG layer was the origin of the SEEs because of the 10B (n,alpha) 7Li reaction products. Both of the particles produced have the capability of ionization in the silicon substrate region, whose thickness is comparable to the ranges of these particles. Using the soft error phenomenon in exactly the opposite manner of the semiconductor industry can provide a new neutron detection system based on the SERs in the semiconductor memories. By investigating the soft error mechanisms in the available semiconductor memories and enhancing the soft error occurrences in these devices, one can convert all memory using intelligent systems into portable, power efficient, directiondependent neutron detectors. The Neutron Intercepting Silicon Chip (NISC) project aims to achieve this goal by introducing 10B-enriched BPSG layers to the semiconductor memory architectures. This research addresses the development of a simulation tool, the NISC Soft Error Analysis Tool (NISCSAT), for soft error modeling and analysis in the semiconductor memories to provide basic design considerations for the NISC. NISCSAT performs particle transport and calculates the soft error probabilities, or SER, depending on energy depositions of the particles in a given memory node model of the NISC. Soft error measurements were performed with commercially available, off-the-shelf semiconductor memories and microprocessors to observe soft error variations with the neutron flux and memory supply voltage. Measurement

  8. Simulation of neutron transport equation using parallel Monte Carlo for deep penetration problems

    International Nuclear Information System (INIS)

    Bekar, K. K.; Tombakoglu, M.; Soekmen, C. N.

    2001-01-01

    Neutron transport equation is simulated using parallel Monte Carlo method for deep penetration neutron transport problem. Monte Carlo simulation is parallelized by using three different techniques; direct parallelization, domain decomposition and domain decomposition with load balancing, which are used with PVM (Parallel Virtual Machine) software on LAN (Local Area Network). The results of parallel simulation are given for various model problems. The performances of the parallelization techniques are compared with each other. Moreover, the effects of variance reduction techniques on parallelization are discussed

  9. Simulations of neutron transport at low energy: a comparison between GEANT and MCNP.

    Science.gov (United States)

    Colonna, N; Altieri, S

    2002-06-01

    The use of the simulation tool GEANT for neutron transport at energies below 20 MeV is discussed, in particular with regard to shielding and dose calculations. The reliability of the GEANT/MICAP package for neutron transport in a wide energy range has been verified by comparing the results of simulations performed with this package in a wide energy range with the prediction of MCNP-4B, a code commonly used for neutron transport at low energy. A reasonable agreement between the results of the two codes is found for the neutron flux through a slab of material (iron and ordinary concrete), as well as for the dose released in soft tissue by neutrons. These results justify the use of the GEANT/MICAP code for neutron transport in a wide range of applications, including health physics problems.

  10. MCViNE – An object oriented Monte Carlo neutron ray tracing simulation package

    Energy Technology Data Exchange (ETDEWEB)

    Lin, Jiao Y.Y., E-mail: linjiao@ornl.gov [Caltech Center for Advanced Computing Research, California Institute of Technology (United States); Department of Applied Physics and Materials Science, California Institute of Technology (United States); Neutron Data Analysis and Visualization Division, Oak Ridge National Laboratory (United States); Smith, Hillary L. [Department of Applied Physics and Materials Science, California Institute of Technology (United States); Granroth, Garrett E., E-mail: granrothge@ornl.gov [Neutron Data Analysis and Visualization Division, Oak Ridge National Laboratory (United States); Abernathy, Douglas L.; Lumsden, Mark D.; Winn, Barry; Aczel, Adam A. [Quantum Condensed Matter Division, Oak Ridge National Laboratory (United States); Aivazis, Michael [Caltech Center for Advanced Computing Research, California Institute of Technology (United States); Fultz, Brent, E-mail: btf@caltech.edu [Department of Applied Physics and Materials Science, California Institute of Technology (United States)

    2016-02-21

    MCViNE (Monte-Carlo VIrtual Neutron Experiment) is an open-source Monte Carlo (MC) neutron ray-tracing software for performing computer modeling and simulations that mirror real neutron scattering experiments. We exploited the close similarity between how instrument components are designed and operated and how such components can be modeled in software. For example we used object oriented programming concepts for representing neutron scatterers and detector systems, and recursive algorithms for implementing multiple scattering. Combining these features together in MCViNE allows one to handle sophisticated neutron scattering problems in modern instruments, including, for example, neutron detection by complex detector systems, and single and multiple scattering events in a variety of samples and sample environments. In addition, MCViNE can use simulation components from linear-chain-based MC ray tracing packages which facilitates porting instrument models from those codes. Furthermore it allows for components written solely in Python, which expedites prototyping of new components. These developments have enabled detailed simulations of neutron scattering experiments, with non-trivial samples, for time-of-flight inelastic instruments at the Spallation Neutron Source. Examples of such simulations for powder and single-crystal samples with various scattering kernels, including kernels for phonon and magnon scattering, are presented. With simulations that closely reproduce experimental results, scattering mechanisms can be turned on and off to determine how they contribute to the measured scattering intensities, improving our understanding of the underlying physics.

  11. An Advanced Neutronic Analysis Toolkit with Inline Monte Carlo capability for BHTR Analysis

    Energy Technology Data Exchange (ETDEWEB)

    William R. Martin; John C. Lee

    2009-12-30

    Monte Carlo capability has been combined with a production LWR lattice physics code to allow analysis of high temperature gas reactor configurations, accounting for the double heterogeneity due to the TRISO fuel. The Monte Carlo code MCNP5 has been used in conjunction with CPM3, which was the testbench lattice physics code for this project. MCNP5 is used to perform two calculations for the geometry of interest, one with homogenized fuel compacts and the other with heterogeneous fuel compacts, where the TRISO fuel kernels are resolved by MCNP5.

  12. An Advanced Neutronic Analysis Toolkit with Inline Monte Carlo capability for VHTR Analysis

    International Nuclear Information System (INIS)

    Martin, William R.; Lee, John C.

    2009-01-01

    Monte Carlo capability has been combined with a production LWR lattice physics code to allow analysis of high temperature gas reactor configurations, accounting for the double heterogeneity due to the TRISO fuel. The Monte Carlo code MCNP5 has been used in conjunction with CPM3, which was the testbench lattice physics code for this project. MCNP5 is used to perform two calculations for the geometry of interest, one with homogenized fuel compacts and the other with heterogeneous fuel compacts, where the TRISO fuel kernels are resolved by MCNP5.

  13. New capabilities for Monte Carlo simulation of deuteron transport and secondary products generation

    International Nuclear Information System (INIS)

    Sauvan, P.; Sanz, J.; Ogando, F.

    2010-01-01

    Several important research programs are dedicated to the development of facilities based on deuteron accelerators. In designing these facilities, the definition of a validated computational approach able to simulate deuteron transport and evaluate deuteron interactions and production of secondary particles with acceptable precision is a very important issue. Current Monte Carlo codes, such as MCNPX or PHITS, when applied for deuteron transport calculations use built-in semi-analytical models to describe deuteron interactions. These models are found unreliable in predicting neutron and photon generated by low energy deuterons, typically present in those facilities. We present a new computational tool, resulting from an extension of the MCNPX code, which improve significantly the treatment of problems where any secondary product (neutrons, photons, tritons, etc.) generated by low energy deuterons reactions could play a major role. Firstly, it handles deuteron evaluated data libraries, which allow describing better low deuteron energy interactions. Secondly, it includes a reduction variance technique for production of secondary particles by charged particle-induced nuclear interactions, which allow reducing drastically the computing time needed in transport and nuclear response calculations. Verification of the computational tool is successfully achieved. This tool can be very helpful in addressing design issues such as selection of the dedicated neutron production target and accelerator radioprotection analysis. It can be also helpful to test the deuteron cross-sections under development in the frame of different international nuclear data programs.

  14. A two-dimensional simulator of the neutronic behaviour of low power fast reactors

    International Nuclear Information System (INIS)

    Penha, M.A.V.R. da.

    1984-01-01

    A model to simulate the temporal neutronic behaviour of fast breeder reactors was developed. The effective cross-sections are corrected, whenever the reactor state change; by using linear correlations and interpolation schemes with data contained in a library previously compiled. This methodology was coupled with a simplified spatial neutronic calculation to investigate the temporal behaviour of neutronic parameters such as breeding gain, flux and power. (Author) [pt

  15. Simulations and developments of the Low Energy Neutron detector Array LENA

    International Nuclear Information System (INIS)

    Langer, C.; Algora, A.; Couture, A.; Csatlós, M.; Gulyás, J.; Heil, M.; Krasznahorkay, A.; O'Donnell, J.M.; Plag, R.; Reifarth, R.; Stuhl, L.; Sonnabend, K.; Tornyi, T.; Tovesson, F.

    2011-01-01

    Prototypes of the Low Energy Neutron detector Array (LENA) have been tested and compared with detailed GEANT simulations. LENA will consist of plastic scintillation bars with the dimensions 1000×45×10 mm 3 . The tests have been performed with γ-ray sources and neutrons originating from the neutron-induced fission of 235 U. The simulations agreed very well with the measured response and were therefore used to simulate the response to mono-energetic neutrons with different detection thresholds. LENA will be used to detect low-energy neutrons from (p,n)-type reactions with low momentum transfer foreseen at the R 3 B and EXL setups at FAIR, Darmstadt.

  16. IB: A Monte Carlo simulation tool for neutron scattering instrument design under PVM and MPI

    International Nuclear Information System (INIS)

    Zhao Jinkui

    2011-01-01

    Design of modern neutron scattering instruments relies heavily on Monte Carlo simulation tools for optimization. IB is one such tool written in C++ and implemented under Parallel Virtual Machine and the Message Passing Interface. The program was initially written for the design and optimization of the EQ-SANS instrument at the Spallation Neutron Source. One of its features is the ability to group simple instrument components into more complex ones at the user input level, e.g. grouping neutron mirrors into neutron guides and curved benders. The simulation engine manages the grouped components such that neutrons entering a group are properly operated upon by all components, multiple times if needed, before exiting the group. Thus, only a few basic optical modules are needed at the programming level. For simulations that require higher computer speeds, the program can be compiled and run in parallel modes using either the PVM or the MPI architectures.

  17. Interfacing MCNPX and McStas for simulation of neutron transport

    DEFF Research Database (Denmark)

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

    2013-01-01

    Stas[4, 5, 6, 7]. The coupling between the two simulation suites typically consists of providing analytical fits of MCNPX neutron spectra to McStas. This method is generally successful but has limitations, as it e.g. does not allow for re-entry of neutrons into the MCNPX regime. Previous work to resolve......Simulations of target-moderator-reflector system at spallation sources are conventionally carried out using Monte Carlo codes such as MCNPX[1] or FLUKA[2, 3] whereas simulations of neutron transport from the moderator and the instrument response are performed by neutron ray tracing codes such as Mc...... geometries, backgrounds, interference between beam-lines as well as shielding requirements along the neutron guides....

  18. Simulation analysis of radiation fields inside phantoms for neutron irradiation

    International Nuclear Information System (INIS)

    Satoh, Daiki; Takahashi, Fumiaki; Endo, Akira; Ohmachi, Y.; Miyahara, N.

    2007-01-01

    Radiation fields inside phantoms have been calculated for neutron irradiation. Particle and heavy-ion transport code system PHITS was employed for the calculation. Energy and size dependences of neutron dose were analyzed using tissue equivalent spheres of different size. A voxel phantom of mouse was developed based on CT images of an 8-week-old male C3H/HeNs mouse. Deposition energy inside the mouse was calculated for 2- and 10-MeV neutron irradiation. (author)

  19. Evaluation of the Neutron Detector Response for Cosmic Ray Energy Spectrum by Monte Carlo Transport Simulation

    International Nuclear Information System (INIS)

    Pazianotto, Mauricio T.; Carlson, Brett V.; Federico, Claudio A.; Gonzalez, Odair L.

    2011-01-01

    Neutrons generated by the interaction of cosmic rays with the atmosphere make an important contribution to the dose accumulated in electronic circuits and aircraft crew members at flight altitude. High-energy neutrons are produced in spallation reactions and intranuclear cascade processes by primary cosmic-ray particle interactions with atoms in the atmosphere. These neutrons can produce secondary neutrons and also undergo a moderation process due to atmosphere interactions, resulting in a wider energy spectrum, ranging from thermal energies (0.025 eV) to energies of several hundreds of MeV. The Long-Counter (LC) detector is a widely used neutron detector designed to measure the directional flux of neutrons with about constant response over a wide energy range (thermal to 20 MeV). ). Its calibration process and the determination of its energy response for the wide-energy of cosmic ray induced neutron spectrum is a very difficult process due to the lack of installations with these capabilities. The goal of this study is to assess the behavior of the response of a Long Counter using the Monte Carlo (MC) computational code MCNPX (Monte Carlo N-Particle eXtended). The dependence of the Long Counter response on the angle of incidence, as well as on the neutron energy, will be carefully investigated, compared with the experimental data previously obtained with 241 Am-Be and 252 Cf neutron sources and extended to the neutron spectrum produced by cosmic rays. (Author)

  20. High resolution real time capable combustion chamber simulation; Zeitlich hochaufloesende echtzeitfaehige Brennraumsimulation

    Energy Technology Data Exchange (ETDEWEB)

    Piewek, J. [Volkswagen AG, Wolfsburg (Germany)

    2008-07-01

    The article describes a zero-dimensional model for the real time capable combustion chamber pressure calculation with analogue pressure sensor output. The closed-loop-operation of an Engine Control Unit is shown at the hardware-in-the-loop-simulator (HiL simulator) for a 4-cylinder common rail diesel engine. The presentation of the model focuses on the simulation of the load variation which does not depend on the injection system and thus the simulated heat release rate. Particular attention is paid to the simulation and the resulting test possibilities regarding to full-variable valve gears. It is shown that black box models consisting in the HiL mean value model for the aspirated gas mass, the exhaust gas temperature after the outlet valve and the mean indicated pressure can be replaced by calculations from the high-resolution combustion chamber model. (orig.)

  1. Monte Carlo Simulation of the Time-Of-Flight Technique for the Measurement of Neutron Cross-section in the Pohang Neutron Facility

    Energy Technology Data Exchange (ETDEWEB)

    An, So Hyun; Lee, Young Ouk; Lee, Cheol Woo [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Lee, Young Seok [National Fusion Research Institute, Daejeon (Korea, Republic of)

    2007-10-15

    It is essential that neutron cross sections are measured precisely for many areas of research and technique. In Korea, these experiments have been performed in the Pohang Neutron Facility (PNF) with the pulsed neutron facility based on the 100 MeV electron linear accelerator. In PNF, the neutron energy spectra have been measured for different water levels inside the moderator and compared with the results of the MCNPX calculation. The optimum size of the water moderator has been determined on the base of these results. In this study, Monte Carlo simulations for the TOF technique were performed and neutron spectra of neutrons were calculated to predict the measurements.

  2. McStas 1.1: A tool for building neutron Monte Carlo simulations

    DEFF Research Database (Denmark)

    Lefmann, K.; Nielsen, K.; Tennant, D.A.

    2000-01-01

    McStas is a project to develop general tools for the creation of simulations of neutron scattering experiments. In this paper, we briefly introduce McStas and describe a particular application of the program: the Monte Carlo calculation of the resolution function of a standard triple-axis neutron...

  3. Simulations of muon-induced neutron flux at large depths underground

    International Nuclear Information System (INIS)

    Kudryavtsev, V.A.; Spooner, N.J.C.; McMillan, J.E.

    2003-01-01

    The production of neutrons by cosmic-ray muons at large depths underground is discussed. The most recent versions of the muon propagation code MUSIC, and particle transport code FLUKA are used to evaluate muon and neutron fluxes. The results of simulations are compared with experimental data

  4. Beam dynamics simulation of the Spallation Neutron Source linear accelerator

    International Nuclear Information System (INIS)

    Takeda, H.; Billen, J.H.; Bhatia, T.S.

    1998-01-01

    The accelerating structure for Spallation Neutron Source (SNS) consists of a radio-frequency-quadrupole-linac (RFQ), a drift-tube-linac (DTL), a coupled-cavity-drift-tube-linac (CCDTL), and a coupled-cavity-linac (CCL). The linac is operated at room temperature. The authors discuss the detailed design of linac which accelerates an H - pulsed beam coming out from RFQ at 2.5 MeV to 1000 MeV. They show a detailed transition from 402.5 MHz DTL with a 4 βλ structure to a CCDTL operated at 805 MHz with a 12 βλ structure. After a discussion of overall feature of the linac, they present an end-to-end particle simulation using the new version of the PARMILA code for a beam starting from the RFQ entrance through the rest of the linac. At 1000 MeV, the beam is transported to a storage ring. The storage ring requires a large (±500-keV) energy spread. This is accomplished by operating the rf-phase in the last section of the linac so the particles are at the unstable fixed point of the separatrix. They present zero-current phase advance, beam size, and beam emittance along the entire linac

  5. Advanced Simulation Capability for Environmental Management - Current Status and Phase II Demonstration Results - 13161

    Energy Technology Data Exchange (ETDEWEB)

    Seitz, Roger R.; Flach, Greg [Savannah River National Laboratory, Savannah River Site, Bldg 773-43A, Aiken, SC 29808 (United States); Freshley, Mark D.; Freedman, Vicky; Gorton, Ian [Pacific Northwest National Laboratory, MSIN K9-33, P.O. Box 999, Richland, WA 99352 (United States); Dixon, Paul; Moulton, J. David [Los Alamos National Laboratory, MS B284, P.O. Box 1663, Los Alamos, NM 87544 (United States); Hubbard, Susan S.; Faybishenko, Boris; Steefel, Carl I.; Finsterle, Stefan [Lawrence Berkeley National Laboratory, 1 Cyclotron Road, MS 50B-4230, Berkeley, CA 94720 (United States); Marble, Justin [Department of Energy, 19901 Germantown Road, Germantown, MD 20874-1290 (United States)

    2013-07-01

    The U.S. Department of Energy (US DOE) Office of Environmental Management (EM), Office of Soil and Groundwater, is supporting development of the Advanced Simulation Capability for Environmental Management (ASCEM). ASCEM is a state-of-the-art scientific tool and approach for understanding and predicting contaminant fate and transport in natural and engineered systems. The modular and open source high-performance computing tool facilitates integrated approaches to modeling and site characterization that enable robust and standardized assessments of performance and risk for EM cleanup and closure activities. The ASCEM project continues to make significant progress in development of computer software capabilities with an emphasis on integration of capabilities in FY12. Capability development is occurring for both the Platform and Integrated Tool-sets and High-Performance Computing (HPC) Multi-process Simulator. The Platform capabilities provide the user interface and tools for end-to-end model development, starting with definition of the conceptual model, management of data for model input, model calibration and uncertainty analysis, and processing of model output, including visualization. The HPC capabilities target increased functionality of process model representations, tool-sets for interaction with Platform, and verification and model confidence testing. The Platform and HPC capabilities are being tested and evaluated for EM applications in a set of demonstrations as part of Site Applications Thrust Area activities. The Phase I demonstration focusing on individual capabilities of the initial tool-sets was completed in 2010. The Phase II demonstration completed in 2012 focused on showcasing integrated ASCEM capabilities. For Phase II, the Hanford Site deep vadose zone (BC Cribs) served as an application site for an end-to-end demonstration of capabilities, with emphasis on integration and linkages between the Platform and HPC components. Other demonstrations

  6. Advanced Simulation Capability for Environmental Management - Current Status and Phase II Demonstration Results - 13161

    International Nuclear Information System (INIS)

    Seitz, Roger R.; Flach, Greg; Freshley, Mark D.; Freedman, Vicky; Gorton, Ian; Dixon, Paul; Moulton, J. David; Hubbard, Susan S.; Faybishenko, Boris; Steefel, Carl I.; Finsterle, Stefan; Marble, Justin

    2013-01-01

    The U.S. Department of Energy (US DOE) Office of Environmental Management (EM), Office of Soil and Groundwater, is supporting development of the Advanced Simulation Capability for Environmental Management (ASCEM). ASCEM is a state-of-the-art scientific tool and approach for understanding and predicting contaminant fate and transport in natural and engineered systems. The modular and open source high-performance computing tool facilitates integrated approaches to modeling and site characterization that enable robust and standardized assessments of performance and risk for EM cleanup and closure activities. The ASCEM project continues to make significant progress in development of computer software capabilities with an emphasis on integration of capabilities in FY12. Capability development is occurring for both the Platform and Integrated Tool-sets and High-Performance Computing (HPC) Multi-process Simulator. The Platform capabilities provide the user interface and tools for end-to-end model development, starting with definition of the conceptual model, management of data for model input, model calibration and uncertainty analysis, and processing of model output, including visualization. The HPC capabilities target increased functionality of process model representations, tool-sets for interaction with Platform, and verification and model confidence testing. The Platform and HPC capabilities are being tested and evaluated for EM applications in a set of demonstrations as part of Site Applications Thrust Area activities. The Phase I demonstration focusing on individual capabilities of the initial tool-sets was completed in 2010. The Phase II demonstration completed in 2012 focused on showcasing integrated ASCEM capabilities. For Phase II, the Hanford Site deep vadose zone (BC Cribs) served as an application site for an end-to-end demonstration of capabilities, with emphasis on integration and linkages between the Platform and HPC components. Other demonstrations

  7. Simulation and preliminary experimental results for an active neutron counter using a neutron generator for a fissile material accounting

    International Nuclear Information System (INIS)

    Ahn, Seong-Kyu; Lee, Tae-Hoon; Shin, Hee-Sung; Kim, Ho-Dong

    2009-01-01

    An active neutron coincidence counter using a neutron generator as an interrogation source has been suggested. Because of the high energy of the interrogation neutron source, 2.5 MeV, the induced fission rate is strongly affected by the moderator design. MCNPX simulation has been performed to evaluate the performance achieved with these moderators. The side- and bottom-moderator are significantly important to thermalize neutrons to induce fission. Based on the simulation results, the moderators are designed to be adapted to the experimental system. Their preliminary performance has been tested by using natural uranium oxide powder samples. For a sample of up to 3.5 kg, which contains 21.7 g of 235 U, 2.64 cps/g- 235 U coincidence events have been measured. Mean background error was 9.57 cps and the resultant coincidence error was 13.8 cps. The experimental result shows the current status of an active counting using a neutron generator which still has some challenges to overcome. However, the controllability of an interrogation source makes this system more applicable for a variety of combinations with other non-destructive methods like a passive coincidence counting especially under a harsh environment such as a hot cell. More precise experimental setup and tests with higher enriched samples will be followed to develop a system to apply it to an active measurement for the safeguards of a spent fuel treatment process.

  8. Surface roughness effect on ultracold neutron interaction with a wall and implications for computer simulations

    International Nuclear Information System (INIS)

    Steyerl, A.; Malik, S. S.; Desai, A. M.; Kaufman, C.

    2010-01-01

    We review the diffuse scattering and the loss coefficient in ultracold neutron reflection from slightly rough surfaces, report a surprising reduction in loss coefficient due to roughness, and discuss the possibility of transition from quantum treatment to ray optics. The results are used in a computer simulation of neutron storage in a recent neutron lifetime experiment that reported a large discrepancy of neutron lifetime with the current particle data value. Our partial reanalysis suggests the possibility of systematic effects that were not included in this publication.

  9. Accurate simulation of neutrons in less than one minute Pt. 1: Acceptance diagram shading

    International Nuclear Information System (INIS)

    Bentley, Phillip M.; Andersen, Ken H.

    2009-01-01

    We describe an algorithm for modelling neutron guides based on an extension of acceptance diagram methods: neutron acceptance diagram shading (NADS), using a computer graphical approach. It has a similar black-box, modular strategy to Monte-Carlo simulations, but often outperforms Monte-Carlo by orders of magnitude in terms of calculation speed without sacrificing significant detail. NADS is expected to increase the productivity in the design stage of neutron instrument development, and facilitate the widespread application of advanced, modern artificial-intelligence-based optimisation algorithms for the design of neutron instrumentation.

  10. Comparison of Experiment and Simulation of the triple GEM-Based Fast Neutron Detector

    International Nuclear Information System (INIS)

    Wang Xiao-Dong; Luo Wen; Zhang Jun-Wei; Yang He-Run; Duan Li-Min; Lu Chen-Gui; Hu Rong-Jiang; Hu Bi-Tao; Zhang Chun-Hui; Yang Lei; Zhou Jian-Rong; An Lv-Xing

    2015-01-01

    A detector for fast neutrons based on a 10 × 10 cm"2 triple gas electron multiplier (GEM) device is developed and tested. A neutron converter, which is a high density polyethylene (HDPE) layer, is combined with the triple GEM detector cathode and placed inside the detector, in the path of the incident neutrons. The detector is tested by obtaining the energy deposition spectrum with an Am Be neutron source in the Institute of Modern Physics (IMP) at Lanzhou. In the present work we report the results of the tests and compare them with those of simulations. The transport of fast neutrons and their interactions with the different materials in the detector are simulated with the GEANT4 code, to understand the experimental results. The detector displays a clear response to the incident fast neutrons. However, an unexpected disagreement in the energy dependence of the response between the simulated and measured spectra is observed. The neutron sources used in our simulation include deuterium-tritium (DT, 14 MeV), deuterium-deuterium (DD, 2.45 MeV), and Am Be sources. The simulation results also show that among the secondary particles generated by the incident neutron, the main contributions to the total energy deposition are from recoil protons induced in hydrogen-rich HDPE or Kapton (GEM material), and activation photons induced by neutron interaction with Ar atoms. Their contributions account for 90% of the total energy deposition. In addition, the dependence of neutron deposited energy spectrum on the composition of the gas mixture is presented. (paper)

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

    DEFF Research Database (Denmark)

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

    2012-01-01

    Simulations of target-moderator-reflector system at spallation sources are conventionally carried out using MCNP/X whereas simulations of neutron transport and instrument performance are carried out by neutron ray tracing codes such as McStas. The coupling between the two simulations suites...... typically consists of providing analytical fits from MCNP/X neutron spectra to McStas. This method is generally successful, but as will be discussed in the this paper, there are limitations and a more direct coupling between MCNP/X andMcStas could allow for more accurate simulations of e.g. complex...... moderator geometries, interference between beamlines as well as shielding requirements along the neutron guides. In this paper different possible interfaces between McStas and MCNP/X are discussed and first preliminary performance results are shown....

  12. Sensitivity to Nuclear Data and Neutron Source Type in Calculations of Transmutation Capabilities of the Energy Amplifier Demonstration Facility

    International Nuclear Information System (INIS)

    Dahlfors, Marcus

    2003-05-01

    This text is a summary of two studies the author has performed within the field of 3-D Monte Carlo calculations of Accelerator Driven Systems (ADS) for transmutation of nuclear waste. The simulations were carried out with the state-of-the-art computer code package EA-MC, developed by C. Rubbia and his group at CERN. The concept studied is ANSALDOs 80 MWth Energy Amplifier Demonstration Facility based on classical MOX-fuel technology and on molten Lead-Bismuth Eutectic cooling. A review of neutron cross section sensitivity in numerical calculations of an ADS and a comparative assessment relevant to the transmutation efficiency of plutonium and minor actinides in fusion/fission hybrids and ADS are presented

  13. Sensitivity to Nuclear Data and Neutron Source Type in Calculations of Transmutation Capabilities of the Energy Amplifier Demonstration Facility

    Energy Technology Data Exchange (ETDEWEB)

    Dahlfors, Marcus

    2003-05-01

    This text is a summary of two studies the author has performed within the field of 3-D Monte Carlo calculations of Accelerator Driven Systems (ADS) for transmutation of nuclear waste. The simulations were carried out with the state-of-the-art computer code package EA-MC, developed by C. Rubbia and his group at CERN. The concept studied is ANSALDOs 80 MWth Energy Amplifier Demonstration Facility based on classical MOX-fuel technology and on molten Lead-Bismuth Eutectic cooling. A review of neutron cross section sensitivity in numerical calculations of an ADS and a comparative assessment relevant to the transmutation efficiency of plutonium and minor actinides in fusion/fission hybrids and ADS are presented.

  14. Precise calculations in simulations of the interaction of low energy neutrons with nano-dispersed media

    International Nuclear Information System (INIS)

    Artem’ev, V. A.; Nezvanov, A. Yu.; Nesvizhevsky, V. V.

    2016-01-01

    We discuss properties of the interaction of slow neutrons with nano-dispersed media and their application for neutron reflectors. In order to increase the accuracy of model simulation of the interaction of neutrons with nanopowders, we perform precise quantum mechanical calculation of potential scattering of neutrons on single nanoparticles using the method of phase functions. We compare results of precise calculations with those performed within first Born approximation for nanodiamonds with the radius of 2–5 nm and for neutron energies 3 × 10 -7 –10 -3 eV. Born approximation overestimates the probability of scattering to large angles, while the accuracy of evaluation of integral characteristics (cross sections, albedo) is acceptable. Using Monte-Carlo method, we calculate albedo of neutrons from different layers of piled up diamond nanopowder

  15. Precise calculations in simulations of the interaction of low energy neutrons with nano-dispersed media

    Science.gov (United States)

    Artem'ev, V. A.; Nezvanov, A. Yu.; Nesvizhevsky, V. V.

    2016-01-01

    We discuss properties of the interaction of slow neutrons with nano-dispersed media and their application for neutron reflectors. In order to increase the accuracy of model simulation of the interaction of neutrons with nanopowders, we perform precise quantum mechanical calculation of potential scattering of neutrons on single nanoparticles using the method of phase functions. We compare results of precise calculations with those performed within first Born approximation for nanodiamonds with the radius of 2-5 nm and for neutron energies 3 × 10-7-10-3 eV. Born approximation overestimates the probability of scattering to large angles, while the accuracy of evaluation of integral characteristics (cross sections, albedo) is acceptable. Using Monte-Carlo method, we calculate albedo of neutrons from different layers of piled up diamond nanopowder.

  16. Kinetic simulation of neutron production in a deuterium z-pinch

    International Nuclear Information System (INIS)

    Mostrom, C.; Stygar, William A.; Thoma, Carsten; Welch, Dale Robert; Clark, R.E.; Leeper, Ramon Joe; Rose, David V.

    2010-01-01

    We have found computationally that, at sufficiently high currents, half of the neutrons produced by a deuterium z pinch are thermonuclear in origin. Early experiments below 1-MA current found that essentially all of the neutrons produced by a deuterium pinch are not thermonuclear, but are initiated by an instability that creates beam-target neutrons. Many subsequent authors have supported this result while others have claimed that pinch neutrons are thermonuclear. To resolve this issue, we have conducted fully kinetic, collisional, and electromagnetic simulations of the complete time evolution of a deuterium pinch. We find that at 1-MA pinch currents, most of the neutrons are, indeed, beam-target in origin. At much higher current, half of the neutrons are thermonuclear and half are beam-target driven by instabilities that produce a power law fall off in the ion energy distribution function at large energy. The implications for fusion energy production with such pinches are discussed.

  17. Precise calculations in simulations of the interaction of low energy neutrons with nano-dispersed media

    Energy Technology Data Exchange (ETDEWEB)

    Artem’ev, V. A., E-mail: niitm@inbox.ru [Research Institute of Materials Technology (Russian Federation); Nezvanov, A. Yu. [Moscow State Industrial University (Russian Federation); Nesvizhevsky, V. V. [Institut Max von Laue—Paul Langevin (France)

    2016-01-15

    We discuss properties of the interaction of slow neutrons with nano-dispersed media and their application for neutron reflectors. In order to increase the accuracy of model simulation of the interaction of neutrons with nanopowders, we perform precise quantum mechanical calculation of potential scattering of neutrons on single nanoparticles using the method of phase functions. We compare results of precise calculations with those performed within first Born approximation for nanodiamonds with the radius of 2–5 nm and for neutron energies 3 × 10{sup -7}–10{sup -3} eV. Born approximation overestimates the probability of scattering to large angles, while the accuracy of evaluation of integral characteristics (cross sections, albedo) is acceptable. Using Monte-Carlo method, we calculate albedo of neutrons from different layers of piled up diamond nanopowder.

  18. Relativistic modeling capabilities in PERSEUS extended MHD simulation code for HED plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Hamlin, Nathaniel D., E-mail: nh322@cornell.edu [438 Rhodes Hall, Cornell University, Ithaca, NY, 14853 (United States); Seyler, Charles E., E-mail: ces7@cornell.edu [Cornell University, Ithaca, NY, 14853 (United States)

    2014-12-15

    We discuss the incorporation of relativistic modeling capabilities into the PERSEUS extended MHD simulation code for high-energy-density (HED) plasmas, and present the latest hybrid X-pinch simulation results. The use of fully relativistic equations enables the model to remain self-consistent in simulations of such relativistic phenomena as X-pinches and laser-plasma interactions. By suitable formulation of the relativistic generalized Ohm’s law as an evolution equation, we have reduced the recovery of primitive variables, a major technical challenge in relativistic codes, to a straightforward algebraic computation. Our code recovers expected results in the non-relativistic limit, and reveals new physics in the modeling of electron beam acceleration following an X-pinch. Through the use of a relaxation scheme, relativistic PERSEUS is able to handle nine orders of magnitude in density variation, making it the first fluid code, to our knowledge, that can simulate relativistic HED plasmas.

  19. Research on neutron noise analysis stochastic simulation method for α calculation

    International Nuclear Information System (INIS)

    Zhong Bin; Shen Huayun; She Ruogu; Zhu Shengdong; Xiao Gang

    2014-01-01

    The prompt decay constant α has significant application on the physical design and safety analysis in nuclear facilities. To overcome the difficulty of a value calculation with Monte-Carlo method, and improve the precision, a new method based on the neutron noise analysis technology was presented. This method employs the stochastic simulation and the theory of neutron noise analysis technology. Firstly, the evolution of stochastic neutron was simulated by discrete-events Monte-Carlo method based on the theory of generalized Semi-Markov process, then the neutron noise in detectors was solved from neutron signal. Secondly, the neutron noise analysis methods such as Rossia method, Feynman-α method, zero-probability method, and cross-correlation method were used to calculate a value. All of the parameters used in neutron noise analysis method were calculated based on auto-adaptive arithmetic. The a value from these methods accords with each other, the largest relative deviation is 7.9%, which proves the feasibility of a calculation method based on neutron noise analysis stochastic simulation. (authors)

  20. Neutron electric dipole moment in the instanton vacuum: Quenched versus unquenched simulations

    International Nuclear Information System (INIS)

    Faccioli, P.; Guadagnoli, D.; Simula, S.

    2004-01-01

    We investigate the role played by the fermionic determinant in the evaluation of the CP-violating neutron electric dipole moment (EDM) adopting the Instanton Liquid Model. Significant differences between quenched and unquenched calculations are found. In the case of unquenched simulations the neutron EDM decreases linearly with the quark mass and is expected to vanish in the chiral limit. On the contrary, within the quenched approximation, the neutron EDM increases as the quark mass decreases and is expected to diverge as 1/m N f in the chiral limit. We argue that such a qualitatively different behavior is a parameter-free, semiclassical prediction and occurs because the neutron EDM is sensitive to the topological structure of the vacuum. The present analysis suggests that quenched and unquenched lattice QCD simulations of the neutron EDM as well as of other observables governed by topology might show up important differences in the quark mass dependence for m q QCD

  1. Measurement and simulation of neutron response function of organic liquid scintillator detector

    International Nuclear Information System (INIS)

    Gohil, M.; Banerjee, K.; Bhattacharya, S.; Bhattacharya, C.; Kundu, S.; Rana, T.K.; Mukherjee, G.; Meena, J.K.; Pandey, R.; Pai, H.; Ghosh, T.K.; Dey, A.; Mukhopadhyay, S.; Pandit, D.; Pal, S.; Banerjee, S.R.; Bandhopadhyay, T.

    2012-01-01

    Response functions of monoenergetic neutrons at various energies, corresponding to a measured neutron energy spectrum have been extracted. The experimental response functions for neutron energies in the range of ∼2-20 MeV have been compared with the respective GEANT4 predictions. It has been found that, there is some discrepancy between the experimental and the GEANT4 simulated neutron response functions at lower pulse height regions, which increases with the increase of neutron energy. This might be due to the incompleteness of the physics processes used in the present GEANT4 simulations. In particular, higher order reaction processes which become more significant at higher energies should be properly taken into account in the calculation of response function.

  2. Optimization of the SNS magnetism reflectometer neutron-guide optics using Monte Carlo simulations

    CERN Document Server

    Klose, F

    2002-01-01

    The magnetism reflectometer at the spallation neutron source SNS will employ advanced neutron optics to achieve high data rate, improved resolution, and extended dynamic range. Optical components utilized will include a multi-channel polygonal curved bender and a tapered neutron-focusing guide section. The results of a neutron beam interacting with these devices are rather complex. Additional complexity arises due to the spectral/time-emission profile of the moderator and non-perfect neutron optical coatings. While analytic formulae for the individual components provide some design guidelines, a realistic performance assessment of the whole instrument can only be achieved by advanced simulation methods. In this contribution, we present guide optics optimizations for the magnetism reflectometer using Monte Carlo simulations. We compare different instrument configurations and calculate the resulting data rates. (orig.)

  3. An Enhanced GINGER Simulation Code with Harmonic Emission and HDF5 IO Capabilities

    International Nuclear Information System (INIS)

    Fawley, William M.

    2006-01-01

    GINGER [1] is an axisymmetric, polychromatic (r-z-t) FEL simulation code originally developed in the mid-1980's to model the performance of single-pass amplifiers. Over the past 15 years GINGER's capabilities have been extended to include more complicated configurations such as undulators with drift spaces, dispersive sections, and vacuum chamber wakefield effects; multi-pass oscillators; and multi-stage harmonic cascades. Its coding base has been tuned to permit running effectively on platforms ranging from desktop PC's to massively parallel processors such as the IBM-SP. Recently, we have made significant changes to GINGER by replacing the original predictor-corrector field solver with a new direct implicit algorithm, adding harmonic emission capability, and switching to the HDF5 IO library [2] for output diagnostics. In this paper, we discuss some details regarding these changes and also present simulation results for LCLS SASE emission at λ = 0.15 nm and higher harmonics

  4. Neutron transport simulation in high speed moving media using Geant4

    Science.gov (United States)

    Li, G.; Ciungu, B.; Harrisson, G.; Rogge, R. B.; Tun, Z.; van der Ende, B. M.; Zwiers, I.

    2017-12-01

    A method using Geant4 to simulate neutron transport in moving media is described. The method is implanted in the source code of the software since Geant4 does not intrinsically support a moving object. The simulation utilizes the existing physical model and data library in Geant4, combined with frame transformations to account for the effect of relative velocity between neutrons and the moving media. An example is presented involving a high speed rotating cylinder to verify this method and show the effect of moving media on neutron transport.

  5. Neutron shielding verification measurements and simulations for a 235-MeV proton therapy center

    International Nuclear Information System (INIS)

    Newhauser, W.D.; Titt, U.; Dexheimer, D.; Yan, X.; Nill, S.

    2002-01-01

    The neutron shielding at the Massachusetts General Hospital's 235-MeV proton therapy facility was investigated with measurements, analytical calculations, and realistic three-dimensional Monte Carlo simulations. In 37 of 40 cases studied, the analytical calculations predicted higher neutron dose equivalent rates outside the shielding than the measured, typically by more than a factor of 10, and in some cases more than 100. Monte Carlo predictions of dose equivalent at three locations are, on average, 1.1 times the measured values. Except at one location, all of the analytical model predictions and Monte Carlo simulations overestimate neutron dose equivalent

  6. Burnup Estimation of Rhodium Self-Powered Neutron Detector Emitter in VVER Reactor Core Using Monte Carlo Simulations

    OpenAIRE

    Khrutchinsky, А. А.; Kuten, S. A.; Babichev, L. F.

    2011-01-01

    Estimation of burn-up in a rhodium-103 emitter of self-powered neutron detector in VVER-1000 reactor core has been performed using Monte Carlo simulations within approximation of a constant neutron flux.

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

    International Nuclear Information System (INIS)

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

    2012-01-01

    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)

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

    International Nuclear Information System (INIS)

    Ondis, L.A. II; Tyburski, L.J.; Moskowitz, B.S.

    2000-01-01

    The RCP01 Monte Carlo program is used to analyze many geometries of interest in nuclear design and analysis of light water moderated reactors such as the core in its pressure vessel with complex piping arrangement, fuel storage arrays, shipping and container arrangements, and neutron detector configurations. Written in FORTRAN and in use on a variety of computers, it is capable of estimating steady state neutron or photon reaction rates and neutron multiplication factors. The energy range covered in neutron calculations is that relevant to the fission process and subsequent slowing-down and thermalization, i.e., 20 MeV to 0 eV. The same energy range is covered for photon calculations

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

    Energy Technology Data Exchange (ETDEWEB)

    Ondis, L.A., II; Tyburski, L.J.; Moskowitz, B.S.

    2000-03-01

    The RCP01 Monte Carlo program is used to analyze many geometries of interest in nuclear design and analysis of light water moderated reactors such as the core in its pressure vessel with complex piping arrangement, fuel storage arrays, shipping and container arrangements, and neutron detector configurations. Written in FORTRAN and in use on a variety of computers, it is capable of estimating steady state neutron or photon reaction rates and neutron multiplication factors. The energy range covered in neutron calculations is that relevant to the fission process and subsequent slowing-down and thermalization, i.e., 20 MeV to 0 eV. The same energy range is covered for photon calculations.

  10. Using simulation to improve the capability of undergraduate nursing students in mental health care.

    Science.gov (United States)

    Kunst, Elicia L; Mitchell, Marion; Johnston, Amy N B

    2017-03-01

    Mental health care is an increasing component of acute patient care and yet mental health care education can be limited in undergraduate nursing programs. The aim of this study was to establish if simulation learning can be an effective method of improving undergraduate nurses' capability in mental health care in an acute care environment. Undergraduate nursing students at an Australian university were exposed to several high-fidelity high-technology simulation activities that incorporated elements of acute emergency nursing practice and acute mental health intervention, scaffolded by theories of learning. This approach provided a safe environment for students to experience clinical practice, and develop their skills for dealing with complex clinical challenges. Using a mixed method approach, the primary domains of interest in this study were student confidence, knowledge and ability. These were self-reported and assessed before and after the simulation activities (intervention) using a pre-validated survey, to gauge the self-rated capacity of students to initiate and complete effective care episodes. Focus group interviews were subsequently held with students who attended placement in the emergency department to explore the impact of the intervention on student performance in this clinical setting. Students who participated in the simulation activity identified and reported significantly increased confidence, knowledge and ability in mental health care post-intervention. They identified key features of the intervention included the impact of its realism on the quality of learning. There is some evidence to suggest that the intervention had an impact on the performance and reflection of students in the clinical setting. This study provides evidence to support the use of simulation to enhance student nurses' clinical capabilities in providing mental health care in acute care environments. Nursing curriculum development should be based on best-evidence to ensure that

  11. The Application of Neutron Transport Green's Functions to Threat Scenario Simulation

    Science.gov (United States)

    Thoreson, Gregory G.; Schneider, Erich A.; Armstrong, Hirotatsu; van der Hoeven, Christopher A.

    2015-02-01

    Radiation detectors provide deterrence and defense against nuclear smuggling attempts by scanning vehicles, ships, and pedestrians for radioactive material. Understanding detector performance is crucial to developing novel technologies, architectures, and alarm algorithms. Detection can be modeled through radiation transport simulations; however, modeling a spanning set of threat scenarios over the full transport phase-space is computationally challenging. Previous research has demonstrated Green's functions can simulate photon detector signals by decomposing the scenario space into independently simulated submodels. This paper presents decomposition methods for neutron and time-dependent transport. As a result, neutron detector signals produced from full forward transport simulations can be efficiently reconstructed by sequential application of submodel response functions.

  12. Numerical Simulations of Pillar Structured Solid State Thermal Neutron Detector Efficiency and Gamma Discrimination

    Energy Technology Data Exchange (ETDEWEB)

    Conway, A; Wang, T; Deo, N; Cheung, C; Nikolic, R

    2008-06-24

    This work reports numerical simulations of a novel three-dimensionally integrated, {sup 10}boron ({sup 10}B) and silicon p+, intrinsic, n+ (PIN) diode micropillar array for thermal neutron detection. The inter-digitated device structure has a high probability of interaction between the Si PIN pillars and the charged particles (alpha and {sup 7}Li) created from the neutron - {sup 10}B reaction. In this work, the effect of both the 3-D geometry (including pillar diameter, separation and height) and energy loss mechanisms are investigated via simulations to predict the neutron detection efficiency and gamma discrimination of this structure. The simulation results are demonstrated to compare well with the measurement results. This indicates that upon scaling the pillar height, a high efficiency thermal neutron detector is possible.

  13. Measurement and simulation of thermal neutron flux distribution in the RTP core

    Science.gov (United States)

    Rabir, Mohamad Hairie B.; Jalal Bayar, Abi Muttaqin B.; Hamzah, Na'im Syauqi B.; Mustafa, Muhammad Khairul Ariff B.; Karim, Julia Bt. Abdul; Zin, Muhammad Rawi B. Mohamed; Ismail, Yahya B.; Hussain, Mohd Huzair B.; Mat Husin, Mat Zin B.; Dan, Roslan B. Md; Ismail, Ahmad Razali B.; Husain, Nurfazila Bt.; Jalil Khan, Zareen Khan B. Abdul; Yakin, Shaiful Rizaide B. Mohd; Saad, Mohamad Fauzi B.; Masood, Zarina Bt.

    2018-01-01

    The in-core thermal neutron flux distribution was determined using measurement and simulation methods for the Malaysian’s PUSPATI TRIGA Reactor (RTP). In this work, online thermal neutron flux measurement using Self Powered Neutron Detector (SPND) has been performed to verify and validate the computational methods for neutron flux calculation in RTP calculations. The experimental results were used as a validation to the calculations performed with Monte Carlo code MCNP. The detail in-core neutron flux distributions were estimated using MCNP mesh tally method. The neutron flux mapping obtained revealed the heterogeneous configuration of the core. Based on the measurement and simulation, the thermal flux profile peaked at the centre of the core and gradually decreased towards the outer side of the core. The results show a good agreement (relatively) between calculation and measurement where both show the same radial thermal flux profile inside the core: MCNP model over estimation with maximum discrepancy around 20% higher compared to SPND measurement. As our model also predicts well the neutron flux distribution in the core it can be used for the characterization of the full core, that is neutron flux and spectra calculation, dose rate calculations, reaction rate calculations, etc.

  14. Resolved-particle simulation by the Physalis method: Enhancements and new capabilities

    Energy Technology Data Exchange (ETDEWEB)

    Sierakowski, Adam J., E-mail: sierakowski@jhu.edu [Department of Mechanical Engineering, Johns Hopkins University, 3400 North Charles Street, Baltimore, MD 21218 (United States); Prosperetti, Andrea [Department of Mechanical Engineering, Johns Hopkins University, 3400 North Charles Street, Baltimore, MD 21218 (United States); Faculty of Science and Technology and J.M. Burgers Centre for Fluid Dynamics, University of Twente, P.O. Box 217, 7500 AE Enschede (Netherlands)

    2016-03-15

    We present enhancements and new capabilities of the Physalis method for simulating disperse multiphase flows using particle-resolved simulation. The current work enhances the previous method by incorporating a new type of pressure-Poisson solver that couples with a new Physalis particle pressure boundary condition scheme and a new particle interior treatment to significantly improve overall numerical efficiency. Further, we implement a more efficient method of calculating the Physalis scalar products and incorporate short-range particle interaction models. We provide validation and benchmarking for the Physalis method against experiments of a sedimenting particle and of normal wall collisions. We conclude with an illustrative simulation of 2048 particles sedimenting in a duct. In the appendix, we present a complete and self-consistent description of the analytical development and numerical methods.

  15. Investigating the Mobility of Light Autonomous Tracked Vehicles using a High Performance Computing Simulation Capability

    Science.gov (United States)

    Negrut, Dan; Mazhar, Hammad; Melanz, Daniel; Lamb, David; Jayakumar, Paramsothy; Letherwood, Michael; Jain, Abhinandan; Quadrelli, Marco

    2012-01-01

    This paper is concerned with the physics-based simulation of light tracked vehicles operating on rough deformable terrain. The focus is on small autonomous vehicles, which weigh less than 100 lb and move on deformable and rough terrain that is feature rich and no longer representable using a continuum approach. A scenario of interest is, for instance, the simulation of a reconnaissance mission for a high mobility lightweight robot where objects such as a boulder or a ditch that could otherwise be considered small for a truck or tank, become major obstacles that can impede the mobility of the light autonomous vehicle and negatively impact the success of its mission. Analyzing and gauging the mobility and performance of these light vehicles is accomplished through a modeling and simulation capability called Chrono::Engine. Chrono::Engine relies on parallel execution on Graphics Processing Unit (GPU) cards.

  16. Assessing reactor physics codes capabilities to simulate fast reactors on the example of the BN-600 benchmark

    Energy Technology Data Exchange (ETDEWEB)

    Ivanov, Vladimir [Scientific and Engineering Centre for Nuclear and Radiation Safety (SES NRS), Moscow (Russian Federation); Bousquet, Jeremy [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) gGmbH, Garching (Germany)

    2016-11-15

    This work aims to assess the capabilities of reactor physics codes (initially validated for thermal reactors) to simulate fast sodium cooled reactors. The BFS-62-3A critical experiment from the BN-600 Hybrid Core Benchmark Analyses was chosen for the investigation. Monte-Carlo codes (KENO from SCALE and SERPENT 2.1.23) and the deterministic diffusion code DYN3D-MG are applied to calculate the neutronic parameters. It was found that the multiplication factor and reactivity effects calculated by KENO and SERPENT using the ENDF/B-VII.0 continuous energy library are in a good agreement with each other and with the measured benchmark values. Few-groups macroscopic cross sections, required for DYN3D-MG, were prepared in applying different methods implemented in SCALE and SERPENT. The DYN3D-MG results of a simplified benchmark show reasonable agreement with results from Monte-Carlo calculations and measured values. The former results are used to justify DYN3D-MG implementation for sodium cooled fast reactors coupled deterministic analysis.

  17. Development of new techniques and enhancement of automatic capability of neutron activation analysis at the Dalat Research Reactor

    International Nuclear Information System (INIS)

    Ho Manh Dung; Ho Van Doanh; Tran Quang Thien; Pham Ngoc Tuan; Pham Ngoc Son; Tran Quoc Duong; Nguyen Van Cuong; Nguyen Minh Tuan; Nguyen Giang; Nguyen Thi Sy

    2017-01-01

    The techniques of neutron activation analysis (NAA) including cyclic, epithermal and prompt-gamma (CNAA, ENAA and PGNAA, respectively) have been developed at the Dalat research reactor (DRR). In addition, the efforts has been spent to improve the automatic capability of irradiation, measurement and data processing of NAA. The renewal of necessary devices/tools for sample preparation have also been done. Eventually, the performance and the utility in terms of sensitivity, accuracy and stability of the analytical results generated by NAA at DRR have significantly been improved. The main results of the project are: 1) Upgrading of the fast irradiation system on Channel 13-2/TC to allow the cyclic irradiations; 2) Development of CNAA; 3) Development of ENAA; 4) Application of k0-method for PGNAA; 5) Investigation of the automatic sample changer (ASC2); 6) Upgrading of Ko-DALAT software for ENAA and modification of k0-IAEA software for CNAA and PGNAA; and 7) Optimization of irradiation and measurement facilities as well as sample preparation devices/tools. A set of procedures of relevant developed techniques in the project were established. The procedures have been evaluated by analysis of the reference materials for which they are meeting the requirements of multi-element analysis for the intended applications. (author)

  18. Design of 6 Mev linear accelerator based pulsed thermal neutron source: FLUKA simulation and experiment

    Energy Technology Data Exchange (ETDEWEB)

    Patil, B.J., E-mail: bjp@physics.unipune.ac.in [Department of Physics, University of Pune, Pune 411 007 (India); Chavan, S.T.; Pethe, S.N.; Krishnan, R. [SAMEER, IIT Powai Campus, Mumbai 400 076 (India); Bhoraskar, V.N. [Department of Physics, University of Pune, Pune 411 007 (India); Dhole, S.D., E-mail: sanjay@physics.unipune.ac.in [Department of Physics, University of Pune, Pune 411 007 (India)

    2012-01-15

    The 6 MeV LINAC based pulsed thermal neutron source has been designed for bulk materials analysis. The design was optimized by varying different parameters of the target and materials for each region using FLUKA code. The optimized design of thermal neutron source gives flux of 3 Multiplication-Sign 10{sup 6}ncm{sup -2}s{sup -1} with more than 80% of thermal neutrons and neutron to gamma ratio was 1 Multiplication-Sign 10{sup 4}ncm{sup -2}mR{sup -1}. The results of prototype experiment and simulation are found to be in good agreement with each other. - Highlights: Black-Right-Pointing-Pointer The optimized 6 eV linear accelerator based thermal neutron source using FLUKA simulation. Black-Right-Pointing-Pointer Beryllium as a photonuclear target and reflector, polyethylene as a filter and shield, graphite as a moderator. Black-Right-Pointing-Pointer Optimized pulsed thermal neutron source gives neutron flux of 3 Multiplication-Sign 10{sup 6} n cm{sup -2} s{sup -1}. Black-Right-Pointing-Pointer Results of the prototype experiment were compared with simulations and are found to be in good agreement. Black-Right-Pointing-Pointer This source can effectively be used for the study of bulk material analysis and activation products.

  19. Adaptive Planning: Understanding Organizational Workload to Capability/ Capacity through Modeling and Simulation

    Science.gov (United States)

    Hase, Chris

    2010-01-01

    In August 2003, the Secretary of Defense (SECDEF) established the Adaptive Planning (AP) initiative [1] with an objective of reducing the time necessary to develop and revise Combatant Commander (COCOM) contingency plans and increase SECDEF plan visibility. In addition to reducing the traditional plan development timeline from twenty-four months to less than twelve months (with a goal of six months)[2], AP increased plan visibility to Department of Defense (DoD) leadership through In-Progress Reviews (IPRs). The IPR process, as well as the increased number of campaign and contingency plans COCOMs had to develop, increased the workload while the number of planners remained fixed. Several efforts from collaborative planning tools to streamlined processes were initiated to compensate for the increased workload enabling COCOMS to better meet shorter planning timelines. This paper examines the Joint Strategic Capabilities Plan (JSCP) directed contingency planning and staffing requirements assigned to a combatant commander staff through the lens of modeling and simulation. The dynamics of developing a COCOM plan are captured with an ExtendSim [3] simulation. The resulting analysis provides a quantifiable means by which to measure a combatant commander staffs workload associated with development and staffing JSCP [4] directed contingency plans with COCOM capability/capacity. Modeling and simulation bring significant opportunities in measuring the sensitivity of key variables in the assessment of workload to capability/capacity analysis. Gaining an understanding of the relationship between plan complexity, number of plans, planning processes, and number of planners with time required for plan development provides valuable information to DoD leadership. Through modeling and simulation AP leadership can gain greater insight in making key decisions on knowing where to best allocate scarce resources in an effort to meet DoD planning objectives.

  20. The Advanced Modeling, Simulation and Analysis Capability Roadmap Vision for Engineering

    Science.gov (United States)

    Zang, Thomas; Lieber, Mike; Norton, Charles; Fucik, Karen

    2006-01-01

    This paper summarizes a subset of the Advanced Modeling Simulation and Analysis (AMSA) Capability Roadmap that was developed for NASA in 2005. The AMSA Capability Roadmap Team was chartered to "To identify what is needed to enhance NASA's capabilities to produce leading-edge exploration and science missions by improving engineering system development, operations, and science understanding through broad application of advanced modeling, simulation and analysis techniques." The AMSA roadmap stressed the need for integration, not just within the science, engineering and operations domains themselves, but also across these domains. Here we discuss the roadmap element pertaining to integration within the engineering domain, with a particular focus on implications for future observatory missions. The AMSA products supporting the system engineering function are mission information, bounds on information quality, and system validation guidance. The Engineering roadmap element contains 5 sub-elements: (1) Large-Scale Systems Models, (2) Anomalous Behavior Models, (3) advanced Uncertainty Models, (4) Virtual Testing Models, and (5) space-based Robotics Manufacture and Servicing Models.

  1. Simulation of complete neutron scattering experiments: from model systems to liquid germanium

    International Nuclear Information System (INIS)

    Hugouvieux, V.

    2004-11-01

    In this thesis, both theoretical and experimental studies of liquids are done. Neutron scattering enables structural and dynamical properties of liquids to be investigated. On the theoretical side, molecular dynamics simulations are of great interest since they give positions and velocities of the atoms and the forces acting on each of them. They also enable spatial and temporal correlations to be computed and these quantities are also available from neutron scattering experiments. Consequently, the comparison can be made between results from molecular dynamics simulations and from neutron scattering experiments, in order to improve our understanding of the structure and dynamics of liquids. However, since extracting reliable data from a neutron scattering experiment is difficult, we propose to simulate the experiment as a whole, including both instrument and sample, in order to gain understanding and to evaluate the impact of the different parasitic contributions (absorption, multiple scattering associated with elastic and inelastic scattering, instrument resolution). This approach, in which the sample is described by its structure and dynamics as computed from molecular dynamics simulations, is presented and tested on isotropic model systems. Then liquid germanium is investigated by inelastic neutron scattering and both classical and ab initio molecular dynamics simulations. This enables us to simulate the experiment we performed and to evaluate the influence of the contributions from the instrument and from the sample on the detected signal. (author)

  2. Consideration of a ultracold neutron source in two-dimensional cylindrical geometry by taking simulated boundaries

    Energy Technology Data Exchange (ETDEWEB)

    Gheisari, R., E-mail: gheisari@pgu.ac.ir [Physics Department, Persian Gulf University, Bushehr 75169 (Iran, Islamic Republic of); Nuclear Energy Research Center, Persian Gulf University, Bushehr 75169 (Iran, Islamic Republic of); Firoozabadi, M. M.; Mohammadi, H. [Department of Physics, University of Birjand, Birjand 97175 (Iran, Islamic Republic of)

    2014-01-15

    A new idea to calculate ultracold neutron (UCN) production by using Monte Carlo simulation method to calculate the cold neutron (CN) flux and an analytical approach to calculate the UCN production from the simulated CN flux was given. A super-thermal source (UCN source) was modeled based on an arrangement of D{sub 2}O and solid D{sub 2} (sD{sub 2}). The D{sub 2}O was investigated as the neutron moderator, and sD{sub 2} as the converter. In order to determine the required parameters, a two-dimensional (2D) neutron balance equation written in Matlab was combined with the MCNPX simulation code. The 2D neutron-transport equation in cylindrical (ρ − z) geometry was considered for 330 neutron energy groups in the sD{sub 2}. The 2D balance equation for UCN and CN was solved using simulated CN flux as boundary value. The UCN source dimensions were calculated for the development of the next UCN source. In the optimal condition, the UCN flux and the UCN production rate (averaged over the sD{sub 2} volume) equal to 6.79 × 10{sup 6} cm{sup −2}s{sup −1} and 2.20 ×10{sup 5} cm{sup −3}s{sup −1}, respectively.

  3. Simulation of the Production of Secondary Particles from a Neutron Beam on Polyethylene Targets using the GEANT4 Simulation Tool

    CERN Document Server

    Ilgner, C

    2003-01-01

    In view of a beam test of RadFET semiconductor detectors and optically stimulated luminescence (OSL) detectors as on-line dosimeters for radiation monitoring purposes in the caverns of the Large Hadron Collider (LHC) experiments, a simulation on the production of secondary particles from a neutron beam on a polyethylene target was carried out. We describe the yield of recoil protons, scattered neutrons as well as electrons, positrons and photons, when neutrons of an average energy of 20 MeV hit polyethylene targets of several thicknesses. The simulation was carried out using the latest release 5.2 of the GEANT4 detector description and simulation tool, including advanced hadron interaction models.

  4. Extending the Capabilities of Closed-loop Distributed Engine Control Simulations Using LAN Communication

    Science.gov (United States)

    Aretskin-Hariton, Eliot D.; Zinnecker, Alicia Mae; Culley, Dennis E.

    2014-01-01

    Distributed Engine Control (DEC) is an enabling technology that has the potential to advance the state-of-the-art in gas turbine engine control. To analyze the capabilities that DEC offers, a Hardware-In-the-Loop (HIL) test bed is being developed at NASA Glenn Research Center. This test bed will support a systems-level analysis of control capabilities in closed-loop engine simulations. The structure of the HIL emulates a virtual test cell by implementing the operator functions, control system, and engine on three separate computers. This implementation increases the flexibility and extensibility of the HIL. Here, a method is discussed for implementing these interfaces by connecting the three platforms over a dedicated Local Area Network (LAN). This approach is verified using the Commercial Modular Aero-Propulsion System Simulation 40k (C-MAPSS40k), which is typically implemented on one computer. There are marginal differences between the results from simulation of the typical and the three-computer implementation. Additional analysis of the LAN network, including characterization of network load, packet drop, and latency, is presented. The three-computer setup supports the incorporation of complex control models and proprietary engine models into the HIL framework.

  5. Simulation study of accelerator based quasi-mono-energetic epithermal neutron beams for BNCT.

    Science.gov (United States)

    Adib, M; Habib, N; Bashter, I I; El-Mesiry, M S; Mansy, M S

    2016-01-01

    Filtered neutron techniques were applied to produce quasi-mono-energetic neutron beams in the energy range of 1.5-7.5 keV at the accelerator port using the generated neutron spectrum from a Li (p, n) Be reaction. A simulation study was performed to characterize the filter components and transmitted beam lines. The feature of the filtered beams is detailed in terms of optimal thickness of the primary and additive components. A computer code named "QMNB-AS" was developed to carry out the required calculations. The filtered neutron beams had high purity and intensity with low contamination from the accompanying thermal, fast neutrons and γ-rays. Copyright © 2015 Elsevier Ltd. All rights reserved.

  6. Simulation study on the cold neutron guides in China advanced research reactor

    International Nuclear Information System (INIS)

    Guo Liping; Yang Tonghua; Wang Hongli; Sun Kai; Zhao Zhixiang

    2003-01-01

    The designs of the two cold neutron guides, CNG1 and CNG2, to be built in China advanced research reactor (CARR) are studied with Monte-Carlo simulation technique. The neutron flux density at the exit of the both guides can reach above 1 x10 9 cm -2 ·s -1 under the assumed flux spectrum of the cold neutron source. The transmission efficiency is 50% and 42%, and the maximum divergence is about 2.2 degree and 1.9 degree, respectively for CNG1 and CNG2. Neutron distribution along horizontal direction is quite uniform for both guides, with maximum fluctuation of less than 3%. Gravity can affect neutron distribution along vertical direction considerably

  7. GEANT4 simulation study of a gamma-ray detector for neutron resonance densitometry

    International Nuclear Information System (INIS)

    Tsuchiya, Harufumi; Harada, Hideo; Koizumi, Mitsuo; Kitatani, Fumito; Takamine, Jun; Kureta, Masatoshi; Iimura, Hideki

    2013-01-01

    A design study of a gamma-ray detector for neutron resonance densitometry was made with GEANT4. The neutron resonance densitometry, combining neutron resonance transmission analysis and neutron resonance capture analysis, is a non-destructive technique to measure amounts of nuclear materials in melted fuels of the Fukushima Daiichi nuclear power plants. In order to effectively quantify impurities in the melted fuels via prompt gamma-ray measurements, a gamma-ray detector for the neutron resonance densitometry consists of cylindrical and well type LaBr 3 scintillators. The present simulation showed that the proposed gamma-ray detector suffices to clearly detect the gamma rays emitted by 10 B(n, αγ) reaction in a high environmental background due to 137 Cs radioactivity with its Compton edge suppressed at a considerably small level. (author)

  8. Advanced simulation capability for environmental management - current status and future applications

    Energy Technology Data Exchange (ETDEWEB)

    Freshley, Mark; Scheibe, Timothy [Pacific Northwest National Laboratory, Richland, Washington (United States); Robinson, Bruce; Moulton, J. David; Dixon, Paul [Los Alamos National Laboratory, Los Alamos, New Mexico (United States); Marble, Justin; Gerdes, Kurt [U.S. Department of Energy, Office of Environmental Management, Washington DC (United States); Stockton, Tom [Neptune and Company, Inc, Los Alamos, New Mexico (United States); Seitz, Roger [Savannah River National Laboratory, Aiken, South Carolina (United States); Black, Paul [Neptune and Company, Inc, Lakewood, Colorado (United States)

    2013-07-01

    The U.S. Department of Energy (US DOE) Office of Environmental Management (EM), Office of Soil and Groundwater (EM-12), is supporting development of the Advanced Simulation Capability for Environmental Management (ASCEM). ASCEM is a state-of-the-art scientific tool and approach that is currently aimed at understanding and predicting contaminant fate and transport in natural and engineered systems. ASCEM is a modular and open source high-performance computing tool. It will be used to facilitate integrated approaches to modeling and site characterization, and provide robust and standardized assessments of performance and risk for EM cleanup and closure activities. The ASCEM project continues to make significant progress in development of capabilities, with current emphasis on integration of capabilities in FY12. Capability development is occurring for both the Platform and Integrated Tool-sets and High-Performance Computing (HPC) multi-process simulator. The Platform capabilities provide the user interface and tools for end-to-end model development, starting with definition of the conceptual model, management of data for model input, model calibration and uncertainty analysis, and processing of model output, including visualization. The HPC capabilities target increased functionality of process model representations, tool-sets for interaction with Platform, and verification and model confidence testing. The integration of the Platform and HPC capabilities were tested and evaluated for EM applications in a set of demonstrations as part of Site Applications Thrust Area activities in 2012. The current maturity of the ASCEM computational and analysis capabilities has afforded the opportunity for collaborative efforts to develop decision analysis tools to support and optimize radioactive waste disposal. Recent advances in computerized decision analysis frameworks provide the perfect opportunity to bring this capability into ASCEM. This will allow radioactive waste

  9. Monte Carlo simulations to advance characterisation of landmines by pulsed fast/thermal neutron analysis

    NARCIS (Netherlands)

    Maucec, M.; Rigollet, C.

    The performance of a detection system based on the pulsed fast/thermal neutron analysis technique was assessed using Monte Carlo simulations. The aim was to develop and implement simulation methods, to support and advance the data analysis techniques of the characteristic gamma-ray spectra,

  10. Development and simulation of various methods for neutron activation analysis

    International Nuclear Information System (INIS)

    Otgooloi, B.

    1993-01-01

    Simple methods for neutron activation analysis have been developed. The results on the studies of installation for determination of fluorine in fluorite ores directly on the lorry by fast neutron activation analysis have been shown. Nitrogen in organic materials was shown by N 14 and N 15 activation. The description of the new equipment 'FLUORITE' for fluorate factory have been shortly given. Pu and Be isotope in organic materials, including in wheat, was measured. 25 figs, 19 tabs. (Author, Translated by J.U)

  11. TMCC: a transient three-dimensional neutron transport code by the direct simulation method - 222

    International Nuclear Information System (INIS)

    Shen, H.; Li, Z.; Wang, K.; Yu, G.

    2010-01-01

    A direct simulation method (DSM) is applied to solve the transient three-dimensional neutron transport problems. DSM is based on the Monte Carlo method, and can be considered as an application of the Monte Carlo method in the specific type of problems. In this work, the transient neutronics problem is solved by simulating the dynamic behaviors of neutrons and precursors of delayed neutrons during the transient process. DSM gets rid of various approximations which are always necessary to other methods, so it is precise and flexible in the requirement of geometric configurations, material compositions and energy spectrum. In this paper, the theory of DSM is introduced first, and the numerical results obtained with the new transient analysis code, named TMCC (Transient Monte Carlo Code), are presented. (authors)

  12. Benchmarking shielding simulations for an accelerator-driven spallation neutron source

    Directory of Open Access Journals (Sweden)

    Nataliia Cherkashyna

    2015-08-01

    Full Text Available The shielding at an accelerator-driven spallation neutron facility plays a critical role in the performance of the neutron scattering instruments, the overall safety, and the total cost of the facility. Accurate simulation of shielding components is thus key for the design of upcoming facilities, such as the European Spallation Source (ESS, currently in construction in Lund, Sweden. In this paper, we present a comparative study between the measured and the simulated neutron background at the Swiss Spallation Neutron Source (SINQ, at the Paul Scherrer Institute (PSI, Villigen, Switzerland. The measurements were carried out at several positions along the SINQ monolith wall with the neutron dosimeter WENDI-2, which has a well-characterized response up to 5 GeV. The simulations were performed using the Monte-Carlo radiation transport code geant4, and include a complete transport from the proton beam to the measurement locations in a single calculation. An agreement between measurements and simulations is about a factor of 2 for the points where the measured radiation dose is above the background level, which is a satisfactory result for such simulations spanning many energy regimes, different physics processes and transport through several meters of shielding materials. The neutrons contributing to the radiation field emanating from the monolith were confirmed to originate from neutrons with energies above 1 MeV in the target region. The current work validates geant4 as being well suited for deep-shielding calculations at accelerator-based spallation sources. We also extrapolate what the simulated flux levels might imply for short (several tens of meters instruments at ESS.

  13. Detection of explosives and other illicit materials by a single nanosecond neutron pulses - Monte-Carlo simulations of the detection process

    International Nuclear Information System (INIS)

    Miklaszewski, R.; Drozdowicz, K.; Wiacek, U.; Dworak, D.; Gribkov, V.

    2011-01-01

    Recent progress in the development of a single-pulse Nanosecond Impulse Neutron Investigation System (NINIS) intended for interrogation of hidden objects (explosives and other illicit materials) by means of measuring elastically scattered neutrons is presented in this paper. The method is based on the well know fact that nuclide-specific information is present in the scattered neutron field. The method uses very bright neutron pulses having duration of the order of few nanoseconds, generated by a dense plasma focus (DPF) devices filled with a pure deuterium or deuterium-tritium mixture as a working gas. Very short duration of the neutron pulse, its high brightness and mono-chromaticity allow to use the time-of-flight method with bases of about few meters to distinguish signals from neutrons scattered by different elements. Results of the Monte Carlo simulations of the scattered neutron field from several compounds (explosives and everyday use materials) are presented in the paper. The MCNP5 code has been used to get information on the angular and energy distributions of the neutrons scattered by the above mentioned compounds assuming the initial neutron energy equal to 2.45 MeV (D-D). A new input has been elaborated that allows the modelling of not only a spectrum of the neutrons scattered at different angles but also their time history from the moment of generation up to detection. Such an approach allows getting approximate signals as registered by scintillator + photomultiplier probes placed at various distances from the scattering object, demonstrating a principal capability of the method to identify an elemental content of the inspected objects. Preliminary results of the MCNP modelling of the interrogation process of the airport luggage containing several illicit objects are presented as well. (authors)

  14. Measurement and simulation of the neutron detection efficiency with a Pb-scintillating fiber calorimeter

    Energy Technology Data Exchange (ETDEWEB)

    Anelli, M; Bertolucci, S; Curceanu, C; Giovannella, S; Happacher, F; Iliescu, M; Martini, M; Miscetti, S [Laboratori Nazionali di Frascati, INFN (Italy); Battistoni, G [Sezione INFN di Milano (Italy); Bini, C; Zorzi, G De; Domenico, Adi; Gauzzi, P [Ubiversita degli Studi ' La Sapienza' e Sezine INFN di Roma (Italy); Branchini, P; Micco, B Di; Ngugen, F; Paseri, A [Universita degli di Studi ' Roma Tre' e Sezione INFN di Roma Tre (Italy); Ferrari, A [Fondazione CNAO, Milano (Italy); Prokfiev, A [Svedberg Laboratory, Uppsala University (Sweden); Fiore, S, E-mail: matteo.martino@inf.infn.i

    2009-04-01

    We have measured the overall detection efficiency of a small prototype of the KLOE PB-scintillation fiber calorimeter to neutrons with kinetic energy range [5,175] MeV. The measurement has been done in a dedicated test beam in the neutron beam facility of the Svedberg Laboratory, TSL Uppsala. The measurements of the neutron detection efficiency of a NE110 scintillator provided a reference calibration. At the lowest trigger threshold, the overall calorimeter efficiency ranges from 28% to 33%. This value largely exceeds the estimated {approx}8% expected if the response were proportional only to the scintillator equivalent thickness. A detailed simulation of the calorimeter and of the TSL beam line has been performed with the FLUKA Monte Carlo code. The simulated response of the detector to neutrons is presented together with the first data to Monte Carlo comparison. The results show an overall neutron efficiency of about 35%. The reasons for such an efficiency enhancement, in comparison with the typical scintillator-based neutron counters, are explained, opening the road to a novel neutron detector.

  15. Fission fragment simulation of fusion neutron radiation effects on bulk mechanical properties

    International Nuclear Information System (INIS)

    Van Konynenburg, R.A.; Mitchell, J.B.; Guinan, M.W.; Stuart, R.N.; Borg, R.J.

    1976-01-01

    This research demonstrates the feasibility of using homogeneously-generated fission fragments to simulate high-fluence fusion neutron damage in niobium tensile specimens. This technique makes it possible to measure radiation effects on bulk mechanical properties at high damage states, using conveniently short irradiation times. The primary knock-on spectrum for a fusion reactor is very similar to that produced by fission fragments, and nearly the same ratio of gas atoms to displaced atoms is produced in niobium. The damage from fission fragments is compared to that from fusion neutrons and fission reactor neutrons in terms of experimentally measured yield strength increase, transmission electron microscopy (TEM) observations, and calculated damage energies

  16. Coupled neutronics and thermal-hydraulics numerical simulations of a Molten Salt Fast Reactor (MSFR)

    International Nuclear Information System (INIS)

    Laureau, A.; Rubiolo, P.R.; Heuer, D.; Merle-Lucotte, E.; Brovchenko, M.

    2013-01-01

    Coupled neutronics and thermalhydraulic numerical analyses of a molten salt fast reactor (MSFR) are presented. These preliminary numerical simulations are carried-out using the Monte Carlo code MCNP and the Computation Fluid Dynamic code OpenFOAM. The main objectives of this analysis performed at steady-reactor conditions are to confirm the acceptability of the current neutronic and thermalhydraulic designs of the reactor, to study the effects of the reactor operating conditions on some of the key MSFR design parameters such as the temperature peaking factor. The effects of the precursor's motion on the reactor safety parameters such as the effective fraction of delayed neutrons have been evaluated. (authors)

  17. A new open-source pin power reconstruction capability in DRAGON5 and DONJON5 neutronic codes

    Energy Technology Data Exchange (ETDEWEB)

    Chambon, R., E-mail: richard-pierre.chambon@polymtl.ca; Hébert, A., E-mail: alain.hebert@polymtl.ca

    2015-08-15

    In order to better optimize the fuel energy efficiency in PWRs, the burnup distribution has to be known as accurately as possible, ideally in each pin. However, this level of detail is lost when core calculations are performed with homogenized cross-sections. The pin power reconstruction (PPR) method can be used to get back those levels of details as accurately as possible in a small additional computing time frame compared to classical core calculations. Such a de-homogenization technique for core calculations using arbitrarily homogenized fuel assembly geometries was presented originally by Fliscounakis et al. In our work, the same methodology was implemented in the open-source neutronic codes DRAGON5 and DONJON5. The new type of Selengut homogenization, called macro-calculation water gap, also proposed by Fliscounakis et al. was implemented. Some important details on the methodology were emphasized in order to get precise results. Validation tests were performed on 12 configurations of 3×3 clusters where simulations in transport theory and in diffusion theory followed by pin-power reconstruction were compared. The results shows that the pin power reconstruction and the Selengut macro-calculation water gap methods were correctly implemented. The accuracy of the simulations depends on the SPH method and on the homogenization geometry choices. Results show that the heterogeneous homogenization is highly recommended. SPH techniques were investigated with flux-volume and Selengut normalization, but the former leads to inaccurate results. Even though the new Selengut macro-calculation water gap method gives promising results regarding flux continuity at assembly interfaces, the classical Selengut approach is more reliable in terms of maximum and average errors in the whole range of configurations.

  18. Monte Carlo simulation of secondary neutron dose for scanning proton therapy using FLUKA.

    Directory of Open Access Journals (Sweden)

    Chaeyeong Lee

    Full Text Available Proton therapy is a rapidly progressing field for cancer treatment. Globally, many proton therapy facilities are being commissioned or under construction. Secondary neutrons are an important issue during the commissioning process of a proton therapy facility. The purpose of this study is to model and validate scanning nozzles of proton therapy at Samsung Medical Center (SMC by Monte Carlo simulation for beam commissioning. After the commissioning, a secondary neutron ambient dose from proton scanning nozzle (Gantry 1 was simulated and measured. This simulation was performed to evaluate beam properties such as percent depth dose curve, Bragg peak, and distal fall-off, so that they could be verified with measured data. Using the validated beam nozzle, the secondary neutron ambient dose was simulated and then compared with the measured ambient dose from Gantry 1. We calculated secondary neutron dose at several different points. We demonstrated the validity modeling a proton scanning nozzle system to evaluate various parameters using FLUKA. The measured secondary neutron ambient dose showed a similar tendency with the simulation result. This work will increase the knowledge necessary for the development of radiation safety technology in medical particle accelerators.

  19. Geant4 simulations of NIST beam neutron lifetime experiment

    Science.gov (United States)

    Valete, Daniel; Crawford, Bret; BL2 Collaboration Collaboration

    2017-09-01

    A free neutron is unstable and its decay is described by the Standard Model as the transformation of a down quark into an up quark through the weak interaction. Precise measurements of the neutron lifetime test the validity of the theory of the weak interaction and provide useful information for the predictions of the theory of Big Bang nucleosynthesis of the primordial helium abundance in the universe and the number of different types of light neutrinos Nν. The predominant experimental methods for determination of the neutron lifetime are commonly called `beam' and `bottle' methods, and the most recent uses of each method do not agree with each other within their stated uncertainties. An improved experiment of the beam technique, which uses magnetic and electric fields to trap and guide the decay protons of a beam of cold neutrons to a detector, is in progress at the National Institute of Standards and Technology, Gaithersburg, MD with a precision goal of 0.1. I acknowledge the support of the Cross-Diciplinary Institute at Gettysburg College.

  20. Advanced methodology to simulate boiling water reactor transient using coupled thermal-hydraulic/neutron-kinetic codes

    Energy Technology Data Exchange (ETDEWEB)

    Hartmann, Christoph Oliver

    2016-06-13

    -sets) predicted by SCALE6/TRITON and CASMO. Thereby the coupled TRACE/PARCS simulations reproduced the single fuel assembly depletion and stand-alone PARCS results. A turbine trip event, occurred at a BWR plant of type 72, has been investigated in detail using the cross-section libraries generated with SCALE/TRITON and CASMO. Thereby the evolution of the integral BWR parameters predicted by the coupled codes using cross-sections from SCALE/TRITON is very close to the global trends calculated using CASMO cross-sections. Further, to implement uncertainty quantifications, the PARCS reactor dynamic code was extended (uncertainty module) to facilitate the consideration of the uncertainty of neutron kinetic parameters in coupled TRACE/PARCS simulations. For a postulated pressure pertubation, an uncertainty and sensitivity study was performed using TRACE/PARCS and SUSA. The obtained results illustrated the capability of such methodologies which are still under development. Based on this analysis, the uncertainty band for key-parameters, e.g. reactivity, as well as the importance ranking of reactor kinetics parameters could be predicted and identified for this accident scenario.

  1. Coupled thermal-hydraulic and neutronic simulations of Phenix control rod withdrawal tests with SIMMER-IV

    International Nuclear Information System (INIS)

    Kriventsev, Vladimir; Gabrielli, Fabrizio; Rineiski, Andrei

    2014-01-01

    The “end-of-life” tests performed in the Phenix reactor before its final shutdown in 2009, in particular the Control Rod (CR) withdrawal experiments provide an excellent opportunity for the validation and verification of the reactor physics computer codes and modeling approaches. SIMMER-IV, a modern three-dimensional reactor safety code, has been recently employed at Karlsruhe Institute of Technology (KIT) for simulating Phenix experiments in the framework of a benchmark exercise organized under the IAEA project. In this paper, we report and discuss main results obtained with SIMMER-IV at KIT. Particular attention is devoted to the coupling features of thermal-hydraulics and neutronics and their mutual influences. The reactor reactivity, power and neutron flux distributions calculated with SIMMER-IV are in good agreement both with experimental results and with calculations with advanced neutronics codes, such as ERANOS, while the CR reactivity worth is overestimated due to neglecting heterogeneity effects. Because of its multi-physics capabilities SIMMER also calculates the temperature distributions which are in a good agreement with the experimental test results. In this work we describe the improvements in SIMMER neutronics model by employing a correction that is based on the results of cell calculations performed with ERANOS. The study confirms that the 3D SIMMER-IV code can accurately predict major fast reactor neutronics and thermal hydraulic parameters, provided that a special treatment is employed for CR modeling. The results of calculations are analyzed in frames of SIMMER-IV validation and verification assessment. (author)

  2. Status Report on Modelling and Simulation Capabilities for Nuclear-Renewable Hybrid Energy Systems

    Energy Technology Data Exchange (ETDEWEB)

    Rabiti, C. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Epiney, A. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Talbot, P. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Kim, J. S. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bragg-Sitton, S. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Alfonsi, A. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Yigitoglu, A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Greenwood, S. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Cetiner, S. M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Ganda, F. [Argonne National Lab. (ANL), Argonne, IL (United States); Maronati, G. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2017-09-01

    This report summarizes the current status of the modeling and simulation capabilities developed for the economic assessment of Nuclear-Renewable Hybrid Energy Systems (N-R HES). The increasing penetration of variable renewables is altering the profile of the net demand, with which the other generators on the grid have to cope. N-R HES analyses are being conducted to determine the potential feasibility of mitigating the resultant volatility in the net electricity demand by adding industrial processes that utilize either thermal or electrical energy as stabilizing loads. This coordination of energy generators and users is proposed to mitigate the increase in electricity cost and cost volatility through the production of a saleable commodity. Overall, the financial performance of a system that is comprised of peaking units (i.e. gas turbine), baseload supply (i.e. nuclear power plant), and an industrial process (e.g. hydrogen plant) should be optimized under the constraint of satisfying an electricity demand profile with a certain level of variable renewable (wind) penetration. The optimization should entail both the sizing of the components/subsystems that comprise the system and the optimal dispatch strategy (output at any given moment in time from the different subsystems). Some of the capabilities here described have been reported separately in [1, 2, 3]. The purpose of this report is to provide an update on the improvement and extension of those capabilities and to illustrate their integrated application in the economic assessment of N-R HES.

  3. Simulating urban-scale air pollutants and their predicting capabilities over the Seoul metropolitan area.

    Science.gov (United States)

    Park, Il-Soo; Lee, Suk-Jo; Kim, Cheol-Hee; Yoo, Chul; Lee, Yong-Hee

    2004-06-01

    Urban-scale air pollutants for sulfur dioxide, nitrogen dioxide, particulate matter with aerodynamic diameter > or = 10 microm, and ozone (O3) were simulated over the Seoul metropolitan area, Korea, during the period of July 2-11, 2002, and their predicting capabilities were discussed. The Air Pollution Model (TAPM) and the highly disaggregated anthropogenic and the biogenic gridded emissions (1 km x 1 km) recently prepared by the Korean Ministry of Environment were applied. Wind fields with observational nudging in the prognostic meteorological model TAPM are optionally adopted to comparatively examine the meteorological impact on the prediction capabilities of urban-scale air pollutants. The result shows that the simulated concentrations of secondary air pollutant largely agree with observed levels with an index of agreement (IOA) of >0.6, whereas IOAs of approximately 0.4 are found for most primary pollutants in the major cities, reflecting the quality of emission data in the urban area. The observationally nudged wind fields with higher IOAs have little effect on the prediction for both primary and secondary air pollutants, implying that the detailed wind field does not consistently improve the urban air pollution model performance if emissions are not well specified. However, the robust highest concentrations are better described toward observations by imposing observational nudging, suggesting the importance of wind fields for the predictions of extreme concentrations such as robust highest concentrations, maximum levels, and >90th percentiles of concentrations for both primary and secondary urban-scale air pollutants.

  4. New simulation capabilities of electron clouds in ion beams with large tune depression

    International Nuclear Information System (INIS)

    Vay, J.-L.; Furman, M.A.; Seidl, P.A.

    2007-01-01

    We have developed a new, comprehensive set of simulation tools aimed at modeling the interaction of intense ion beams and electron clouds (e-clouds). The set contains the 3-D accelerator PIC code WARP and the 2-D 'slice' e-cloud code POSINST [M. Furman, this workshop, paper TUAX05], as well as a merger of the two, augmented by new modules for impact ionization and neutral gas generation. The new capability runs on workstations or parallel supercomputers and contains advanced features such as mesh refinement, disparate adaptive time stepping, and a new 'drift-Lorentz' particle mover for tracking charged particles in magnetic fields using large time steps. It is being applied to the modeling of ion beams (1 MeV, 180 mA, K+) for heavy ion inertial fusion and warm dense matter studies, as they interact with electron clouds in the High-Current Experiment (HCX) [experimental results discussed by A. Molvik, this workshop, paper THAW02]. We describe the capabilities and present recent simulation results with detailed comparisons against the HCX experiment, as well as their application (in a different regime) to the modeling of e-clouds in the Large Hadron Collider (LHC). (author)

  5. New simulation capabilities of electron clouds in ion beams with large tune depression

    International Nuclear Information System (INIS)

    Lawrence Livermore National Laboratory

    2006-01-01

    We have developed a new, comprehensive set of simulation tools aimed at modeling the interaction of intense ion beams and electron clouds (e-clouds). The set contains the 3-D accelerator PIC code WARP and the 2-D ''slice'' e-cloud code POSINST [M. Furman, this workshop, paper TUAX05], as well as a merger of the two, augmented by new modules for impact ionization and neutral gas generation. The new capability runs on workstations or parallel supercomputers and contains advanced features such as mesh refinement, disparate adaptive time stepping, and a new ''drift-Lorentz'' particle mover for tracking charged particles in magnetic fields using large time steps. It is being applied to the modeling of ion beams (1 MeV, 180 mA, K+) for heavy ion inertial fusion and warm dense matter studies, as they interact with electron clouds in the High-Current Experiment (HCX) [experimental results discussed by A. Molvik, this workshop, paper THAW02]. We describe the capabilities and present recent simulation results with detailed comparisons against the HCX experiment, as well as their application (in a different regime) to the modeling of e-clouds in the Large Hadron Collider (LHC)

  6. New simulation capabilities of electron clouds in ion beams with large tune depression

    International Nuclear Information System (INIS)

    Vay, J.L.; Furman, M.A.; Seidl, P.A.; Cohen, R.H.; Friedman, A.; Grote, D.P.; Kireeff-Covo, M.; Molvik, A.W.; Stoltz, P.H.; Veitzer, S.; Verboncoeur, J.P.

    2006-01-01

    The authors have developed a new, comprehensive set of simulation tools aimed at modeling the interaction of intense ion beams and electron clouds (e-clouds). The set contains the 3-D accelerator PIC code WARP and the 2-D ''slice'' e-cloud code POSINST, as well as a merger of the two, augmented by new modules for impact ionization and neutral gas generation. The new capability runs on workstations or parallel supercomputers and contains advanced features such as mesh refinement, disparate adaptive time stepping, and a new ''drift-Lorentz'' particle mover for tracking charged particles in magnetic fields using large time steps. It is being applied to the modeling of ion beams (1 MeV, 180 mA, K+) for heavy ion inertial fusion and warm dense matter studies, as they interact with electron clouds in the High-Current Experiment (HCX). They describe the capabilities and present recent simulation results with detailed comparisons against the HCX experiment, as well as their application (in a different regime) to the modeling of e-clouds in the Large Hadron Collider (LHC)

  7. Large-Scale Testing and High-Fidelity Simulation Capabilities at Sandia National Laboratories to Support Space Power and Propulsion

    International Nuclear Information System (INIS)

    Dobranich, Dean; Blanchat, Thomas K.

    2008-01-01

    Sandia National Laboratories, as a Department of Energy, National Nuclear Security Agency, has major responsibility to ensure the safety and security needs of nuclear weapons. As such, with an experienced research staff, Sandia maintains a spectrum of modeling and simulation capabilities integrated with experimental and large-scale test capabilities. This expertise and these capabilities offer considerable resources for addressing issues of interest to the space power and propulsion communities. This paper presents Sandia's capability to perform thermal qualification (analysis, test, modeling and simulation) using a representative weapon system as an example demonstrating the potential to support NASA's Lunar Reactor System

  8. A large-eddy simulation based power estimation capability for wind farms over complex terrain

    Science.gov (United States)

    Senocak, I.; Sandusky, M.; Deleon, R.

    2017-12-01

    There has been an increasing interest in predicting wind fields over complex terrain at the micro-scale for resource assessment, turbine siting, and power forecasting. These capabilities are made possible by advancements in computational speed from a new generation of computing hardware, numerical methods and physics modelling. The micro-scale wind prediction model presented in this work is based on the large-eddy simulation paradigm with surface-stress parameterization. The complex terrain is represented using an immersed-boundary method that takes into account the parameterization of the surface stresses. Governing equations of incompressible fluid flow are solved using a projection method with second-order accurate schemes in space and time. We use actuator disk models with rotation to simulate the influence of turbines on the wind field. Data regarding power production from individual turbines are mostly restricted because of proprietary nature of the wind energy business. Most studies report percentage drop of power relative to power from the first row. There have been different approaches to predict power production. Some studies simply report available wind power in the upstream, some studies estimate power production using power curves available from turbine manufacturers, and some studies estimate power as torque multiplied by rotational speed. In the present work, we propose a black-box approach that considers a control volume around a turbine and estimate the power extracted from the turbine based on the conservation of energy principle. We applied our wind power prediction capability to wind farms over flat terrain such as the wind farm over Mower County, Minnesota and the Horns Rev offshore wind farm in Denmark. The results from these simulations are in good agreement with published data. We also estimate power production from a hypothetical wind farm in complex terrain region and identify potential zones suitable for wind power production.

  9. Response of a BGO detector to photon and neutron sources simulations and measurements

    CERN Document Server

    Vincke, H H; Fabjan, Christian Wolfgang; Otto, T

    2002-01-01

    In this paper Monte Carlo simulations (FLUKA) and measurements of the response of a BGO detector are reported. %For the measurements different radioactive sources were used to irradiate the BGO crystal. For the measurements three low-energy photon emitters $\\left({}^{60}\\rm{Co},\\right.$ ${}^{54}\\rm{Mn},$ $\\left. {}^{137}\\rm{Cs}\\right)$ were used to irradiate the BGO from various distances and angles. The neutron response was measured with an Am--Be neutron source. Simulations of the experimental irradiations were carried out. Our study can also be considered as a benchmark for FLUKA in terms of its reliability to predict the detector response of a BGO scintillator.

  10. Right Size Determining the Staff Necessary to Sustain Simulation and Computing Capabilities for Nuclear Security

    Energy Technology Data Exchange (ETDEWEB)

    Nikkel, Daniel J. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Meisner, Robert [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2010-09-10

    The Advanced Simulation and Computing Campaign, herein referred to as the ASC Program, is a core element of the science-based Stockpile Stewardship Program (SSP), which enables assessment, certification, and maintenance of the safety, security, and reliability of the U.S. nuclear stockpile without the need to resume nuclear testing. The use of advanced parallel computing has transitioned from proof-of-principle to become a critical element for assessing and certifying the stockpile. As the initiative phase of the ASC Program came to an end in the mid-2000s, the National Nuclear Security Administration redirected resources to other urgent priorities, and resulting staff reductions in ASC occurred without the benefit of analysis of the impact on modern stockpile stewardship that is dependent on these new simulation capabilities. Consequently, in mid-2008 the ASC Program management commissioned a study to estimate the essential size and balance needed to sustain advanced simulation as a core component of stockpile stewardship. The ASC Program requires a minimum base staff size of 930 (which includes the number of staff necessary to maintain critical technical disciplines as well as to execute required programmatic tasks) to sustain its essential ongoing role in stockpile stewardship.

  11. Neutron contamination of Varian Clinac iX 10 MV photon beam using Monte Carlo simulation

    International Nuclear Information System (INIS)

    Yani, S; Haryanto, F; Arif, I; Tursinah, R; Rhani, M F; Soh, R C X

    2016-01-01

    High energy medical accelerators are commonly used in radiotherapy to increase the effectiveness of treatments. As we know neutrons can be emitted from a medical accelerator if there is an incident of X-ray that hits any of its materials. This issue becomes a point of view of many researchers. The neutron contamination has caused many problems such as image resolution and radiation protection for patients and radio oncologists. This study concerns the simulation of neutron contamination emitted from Varian Clinac iX 10 MV using Monte Carlo code system. As neutron production process is very complex, Monte Carlo simulation with MCNPX code system was carried out to study this contamination. The design of this medical accelerator was modelled based on the actual materials and geometry. The maximum energy of photons and neutron in the scoring plane was 10.5 and 2.239 MeV, respectively. The number and energy of the particles produced depend on the depth and distance from beam axis. From these results, it is pointed out that the neutron produced by linac 10 MV photon beam in a typical treatment is not negligible. (paper)

  12. Monte Carlo simulation of grating-based neutron phase contrast imaging at CPHS

    International Nuclear Information System (INIS)

    Zhang Ran; Chen Zhiqiang; Huang Zhifeng; Xiao Yongshun; Wang Xuewu; Wie Jie; Loong, C.-K.

    2011-01-01

    Since the launching of the Compact Pulsed Hadron Source (CPHS) project of Tsinghua University in 2009, works have begun on the design and engineering of an imaging/radiography instrument for the neutron source provided by CPHS. The instrument will perform basic tasks such as transmission imaging and computerized tomography. Additionally, we include in the design the utilization of coded-aperture and grating-based phase contrast methodology, as well as the options of prompt gamma-ray analysis and neutron-energy selective imaging. Previously, we had implemented the hardware and data-analysis software for grating-based X-ray phase contrast imaging. Here, we investigate Geant4-based Monte Carlo simulations of neutron refraction phenomena and then model the grating-based neutron phase contrast imaging system according to the classic-optics-based method. The simulated experimental results of the retrieving phase shift gradient information by five-step phase-stepping approach indicate the feasibility of grating-based neutron phase contrast imaging as an option for the cold neutron imaging instrument at the CPHS.

  13. Simulation and optimisation of a position sensitive scintillation detector with wavelength shifting fibers for thermal neutrons

    Energy Technology Data Exchange (ETDEWEB)

    Herzkamp, Matthias; Engels, Ralf; Kemmerling, Guenter [ZEA-2, Forschungszentrum Juelich (Germany); Brueckel, Thomas [JCNS, Forschungszentrum Juelich (Germany); Stahl, Achim [III. Physikalisches Institut B, RWTH Aachen (Germany); Waasen, Stefan van [ZEA-2, Forschungszentrum Juelich (Germany); Faculty of Engineering, University of Duisburg-Essen (Germany)

    2015-07-01

    In neutron scattering experiments it is important to have position sensitive large scale detectors for thermal neutrons. A detector based on a neutron scintillator with wave length shifting fibers is a new kind of such a detector. We present the simulation of the detector based on the microscopic structure of the scintillation material of the mentioned detector. It consists of a converter and a scintillation powder bound in a matrix. The converter in our case is lithium fluoride with enriched lithium 6, to convert thermal neutrons into high energetic alpha and triton particles. The scintillation material is silver doped zinc sulfide. We show that pulse height spectra obtained by these scintillators can be be explained by the simple model of randomly distributed spheres of zinc sulfide and lithium fluoride. With this model, it is possible to optimise the mass ratio of zinc sulfide to lithium fluoride with respect to detection efficiency and/or energy deposition in zinc sulfide.

  14. Measurement and simulation of neutron detection efficiency in lead-scintillating fiber calorimeters

    Energy Technology Data Exchange (ETDEWEB)

    Anelli, M.; Bertolucci, S. [Laboratori Nazionali di Frascati, INFN (Italy); Bini, C. [Dipartimento di Fisica dell' Universita ' La Sapienza' , Roma (Italy); INFN Sezione di Roma, Roma (Italy); Branchini, P. [INFN Sezione di Roma Tre, Roma (Italy); Curceanu, C. [Laboratori Nazionali di Frascati, INFN (Italy); De Zorzi, G.; Di Domenico, A. [Dipartimento di Fisica dell' Universita ' La Sapienza' , Roma (Italy); INFN Sezione di Roma, Roma (Italy); Di Micco, B. [Dipartimento di Fisica dell' Universita ' Roma Tre' , Roma (Italy); INFN Sezione di Roma Tre, Roma (Italy); Ferrari, A. [Fondazione CNAO, Milano (Italy); Fiore, S.; Gauzzi, P. [Dipartimento di Fisica dell' Universita ' La Sapienza' , Roma (Italy); INFN Sezione di Roma, Roma (Italy); Giovannella, S., E-mail: simona.giovannella@lnf.infn.i [Laboratori Nazionali di Frascati, INFN (Italy); Happacher, F. [Laboratori Nazionali di Frascati, INFN (Italy); Iliescu, M. [Laboratori Nazionali di Frascati, INFN (Italy); IFIN-HH, Bucharest (Romania); Martini, M. [Laboratori Nazionali di Frascati, INFN (Italy); Dipartimento di Energetica dell' Universita ' La Sapienza' , Roma (Italy); Miscetti, S. [Laboratori Nazionali di Frascati, INFN (Italy); Nguyen, F. [Dipartimento di Fisica dell' Universita ' Roma Tre' , Roma (Italy); INFN Sezione di Roma Tre, Roma (Italy); Passeri, A. [INFN Sezione di Roma Tre, Roma (Italy); Prokofiev, A. [Svedberg Laboratory, Uppsala University (Sweden); Sciascia, B. [Laboratori Nazionali di Frascati, INFN (Italy)

    2009-12-15

    The overall detection efficiency to neutrons of a small prototype of the KLOE lead-scintillating fiber calorimeter has been measured at the neutron beam facility of The Svedberg Laboratory, TSL, Uppsala, in the kinetic energy range [5-175] MeV. The measurement of the neutron detection efficiency of a NE110 scintillator provided a reference calibration. At the lowest trigger threshold, the overall calorimeter efficiency ranges from 30% to 50%. This value largely exceeds the estimated 8-15% expected if the response were proportional only to the scintillator equivalent thickness. A detailed simulation of the calorimeter and of the TSL beam line has been performed with the FLUKA Monte Carlo code. First data-MC comparisons are encouraging and allow to disentangle a neutron halo component in the beam.

  15. Neutron tomography using projection data obtained by Monte Carlo simulation for nondestructive evaluation

    International Nuclear Information System (INIS)

    Silva, A.X. da; Crispim, V.R.

    2002-01-01

    This work present the application of a computer package for generating of projection data for neutron computerized tomography, and in second part, discusses an application of neutron tomography, using the projection data obtained by Monte Carlo technique, for the detection and localization of light materials such as those containing hydrogen, concealed by heavy materials such as iron and lead. For tomographic reconstructions of the samples simulated use was made of only six equal projection angles distributed between 0 deg C and 180 deg C, with reconstruction making use of an algorithm (ARIEM), based on the principle of maximum entropy. With the neutron tomography it was possible to detect and locate polyethylene and water hidden by lead and iron (with 1 cm-thick). Thus, it is demonstrated that thermal neutrons tomography is a viable test method which can provide important interior information about test components, so, extremely useful in routine industrial applications.(author)

  16. Measurement and simulation of neutron detection efficiency in lead-scintillating fiber calorimeters

    International Nuclear Information System (INIS)

    Anelli, M.; Bertolucci, S.; Bini, C.; Branchini, P.; Curceanu, C.; De Zorzi, G.; Di Domenico, A.; Di Micco, B.; Ferrari, A.; Fiore, S.; Gauzzi, P.; Giovannella, S.; Happacher, F.; Iliescu, M.; Martini, M.; Miscetti, S.; Nguyen, F.; Passeri, A.; Prokofiev, A.; Sciascia, B.

    2009-01-01

    The overall detection efficiency to neutrons of a small prototype of the KLOE lead-scintillating fiber calorimeter has been measured at the neutron beam facility of The Svedberg Laboratory, TSL, Uppsala, in the kinetic energy range [5-175] MeV. The measurement of the neutron detection efficiency of a NE110 scintillator provided a reference calibration. At the lowest trigger threshold, the overall calorimeter efficiency ranges from 30% to 50%. This value largely exceeds the estimated 8-15% expected if the response were proportional only to the scintillator equivalent thickness. A detailed simulation of the calorimeter and of the TSL beam line has been performed with the FLUKA Monte Carlo code. First data-MC comparisons are encouraging and allow to disentangle a neutron halo component in the beam.

  17. Development of Neutronics Model for ShinKori Unit 1 Simulator

    Energy Technology Data Exchange (ETDEWEB)

    Hong, JinHyuk; Lee, MyeongSoo; Lee, SeungHo; Suh, JungKwan; Hwang, DoHyun [KEPRI, Daejeon (Korea, Republic of)

    2008-05-15

    ShinKori-Unit 1 and 2 is being built in the Kori site which will be operated at 2815 MWt of thermal core power. The purpose of this paper is to report on the performance of the developed neutronics model of ShinKori Unit 1 and 2. Also this report includes the convenient tool (XS2R5) for processing the large quantity of information received from the DIT/ROCS model and generating cross-sections. The neutronics model is based on the NESTLE code inserted to RELAP5/MOD3 thermal-hydraulics analysis code which was funded as FY-93 LDRD Project 7201 and is running on the commercial simulator environment tool (the 3KeyMaster{sup TM} of the WSC). As some examples for the verification of the developed neutronics model, some figures are provided. The output of the developed neutronics model is in accord with the Preliminary Safety Analysis Report (PSAR) of the reference plant.

  18. Monte Carlo simulated dose to the human body due to neutrons emitted in laser-fusion

    International Nuclear Information System (INIS)

    Gileadi, A.E.; Cohen, M.O.

    1977-01-01

    Considering a point neutron source located at a given distance from the human body, modeled by a 'standard reference man' phantom, neutron doses to the whole body, as well as to selected organs thereof, are determined, using the SAM-CE system, a Monte Carlo computer code, written in Fortran and designed to solve time, space and energy dependent neutron and gamma ray transport equations in complex three-dimensional geometrice. Collision density, energy deposition and dose are treated in the SAM-CE system as flux functionals. A special feature of SAM-CE is its use of the 'Combinatorial Geometry' technique which affords the user geometric capabilities exceeding those available with other commonly used geometric packages. All neutron and gamma ray cross section data, as well as gamma ray production data, are derived from the ENDF libraries. Both resolved and unresolved resonance parameters from ENDF neutron data files are treated automatically and extremely precise and detailed descriptions of cross section behavior is permitted. Such treatment avoids the ambiguities usually associated with multi-group codes, which use flux-averaged cross sections based on assumed flux distributions which may or may not be appropriate. The 'standard reference man', a heterogeneous phantom, uses simple geometric forms to approximate the shape and dimensions of the human body. Materials composition of each subregion representing a certain 'organ' is given. Typical values of neutron doses to the whole body and to selected 'organs' of interest are presented

  19. Uncertainty and sensitivity analysis in the neutronic parameters generation for BWR and PWR coupled thermal-hydraulic–neutronic simulations

    International Nuclear Information System (INIS)

    Ánchel, F.; Barrachina, T.; Miró, R.; Verdú, G.; Juanas, J.; Macián-Juan, R.

    2012-01-01

    Highlights: ► Best-estimate codes are affected by the uncertainty in the methods and the models. ► Influence of the uncertainty in the macroscopic cross-sections in a BWR and PWR RIA accidents analysis. ► The fast diffusion coefficient, the scattering cross section and both fission cross sections are the most influential factors. ► The absorption cross sections very little influence. ► Using a normal pdf the results are more “conservative” comparing the power peak reached with uncertainty quantified with a uniform pdf. - Abstract: The Best Estimate analysis consists of a coupled thermal-hydraulic and neutronic description of the nuclear system's behavior; uncertainties from both aspects should be included and jointly propagated. This paper presents a study of the influence of the uncertainty in the macroscopic neutronic information that describes a three-dimensional core model on the most relevant results of the simulation of a Reactivity Induced Accident (RIA). The analyses of a BWR-RIA and a PWR-RIA have been carried out with a three-dimensional thermal-hydraulic and neutronic model for the coupled system TRACE-PARCS and RELAP-PARCS. The cross section information has been generated by the SIMTAB methodology based on the joint use of CASMO-SIMULATE. The statistically based methodology performs a Monte-Carlo kind of sampling of the uncertainty in the macroscopic cross sections. The size of the sampling is determined by the characteristics of the tolerance intervals by applying the Noether–Wilks formulas. A number of simulations equal to the sample size have been carried out in which the cross sections used by PARCS are directly modified with uncertainty, and non-parametric statistical methods are applied to the resulting sample of the values of the output variables to determine their intervals of tolerance.

  20. Monte-Carlo simulations of neutron shielding for the ATLAS forward region

    CERN Document Server

    Stekl, I; Kovalenko, V E; Vorobel, V; Leroy, C; Piquemal, F; Eschbach, R; Marquet, C

    2000-01-01

    The effectiveness of different types of neutron shielding for the ATLAS forward region has been studied by means of Monte-Carlo simulations and compared with the results of an experiment performed at the CERN PS. The simulation code is based on GEANT, FLUKA, MICAP and GAMLIB. GAMLIB is a new library including processes with gamma-rays produced in (n, gamma), (n, n'gamma) neutron reactions and is interfaced to the MICAP code. The effectiveness of different types of shielding against neutrons and gamma-rays, composed from different types of material, such as pure polyethylene, borated polyethylene, lithium-filled polyethylene, lead and iron, were compared. The results from Monte-Carlo simulations were compared to the results obtained from the experiment. The simulation results reproduce the experimental data well. This agreement supports the correctness of the simulation code used to describe the generation, spreading and absorption of neutrons (up to thermal energies) and gamma-rays in the shielding materials....

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-07-01

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

  2. A calibration method for realistic neutron dosimetry in radiobiological experiments assisted by MCNP simulation.

    Science.gov (United States)

    Shahmohammadi Beni, Mehrdad; Krstic, Dragana; Nikezic, Dragoslav; Yu, Kwan Ngok

    2016-09-01

    Many studies on biological effects of neutrons involve dose responses of neutrons, which rely on accurately determined absorbed doses in the irradiated cells or living organisms. Absorbed doses are difficult to measure, and are commonly surrogated with doses measured using separate detectors. The present work describes the determination of doses absorbed in the cell layer underneath a medium column (D A ) and the doses absorbed in an ionization chamber (D E ) from neutrons through computer simulations using the MCNP-5 code, and the subsequent determination of the conversion coefficients R (= D A /D E ). It was found that R in general decreased with increase in the medium thickness, which was due to elastic and inelastic scattering. For 2-MeV neutrons, conspicuous bulges in R values were observed at medium thicknesses of about 500, 1500, 2500 and 4000 μm, and these were attributed to carbon, oxygen and nitrogen nuclei, and were reflections of spikes in neutron interaction cross sections with these nuclei. For 0.1-MeV neutrons, no conspicuous bulges in R were observed (except one at ~2000 μm that was due to photon interactions), which was explained by the absence of prominent spikes in the interaction cross-sections with these nuclei for neutron energies <0.1 MeV. The ratio R could be increased by ~50% for small medium thickness if the incident neutron energy was reduced from 2 MeV to 0.1 MeV. As such, the absorbed doses in cells (D A ) would vary with the incident neutron energies, even when the absorbed doses shown on the detector were the same. © The Author 2016. Published by Oxford University Press on behalf of The Japan Radiation Research Society and Japanese Society for Radiation Oncology.

  3. Neutronic characteristics simulation of LMFBR of great size

    International Nuclear Information System (INIS)

    Kim, Y.C.

    1987-09-01

    The CONRAD experimental program to be executed on the critical mockup MASURCA in Cadarache and use all the european plutonium stock. The objectives of this program are to reduce the uncertainties on important project parameters such as the reactivity value of control rods, the flux distribution to valid calcul methods and data to use for new LMFBR conception (heterogeneous axial core by example) and to resolve the neutronic control problems for a LMFBR of great size. The present study has permitted to define this program and its physical characteristics [fr

  4. Research Capabilities for Oil-Free Turbomachinery Expanded by New Rotordynamic Simulator Facility

    Science.gov (United States)

    Howard, Samuel A.

    2004-01-01

    A new test rig has been developed for simulating high-speed turbomachinery shafting using Oil-Free foil air bearing technology. Foil air journal bearings are self-acting hydrodynamic bearings with a flexible inner sleeve surface using air as the lubricant. These bearings have been used in turbomachinery, primarily air cycle machines, for the past four decades to eliminate the need for oil lubrication. More recently, interest has been growing in applying foil bearings to aircraft gas turbine engines. They offer potential improvements in efficiency and power density, decreased maintenance costs, and other secondary benefits. The goal of applying foil air bearings to aircraft gas turbine engines prompted the fabrication of this test rig. The facility enables bearing designers to test potential bearing designs with shafts that simulate the rotating components of a target engine without the high cost of building actual flight hardware. The data collected from this rig can be used to make changes to the shaft and bearings in subsequent design iterations. The rest of this article describes the new test rig and demonstrates some of its capabilities with an initial simulated shaft system. The test rig has two support structures, each housing a foil air journal bearing. The structures are designed to accept any size foil journal bearing smaller than 63 mm (2.5 in.) in diameter. The bearing support structures are mounted to a 91- by 152-cm (3- by 5-ft) table and can be separated by as much as 122 cm (4 ft) and as little as 20 cm (8 in.) to accommodate a wide range of shaft sizes. In the initial configuration, a 9.5-cm (3.75-in.) impulse air turbine drives the test shaft. The impulse turbine, as well as virtually any number of "dummy" compressor and turbine disks, can be mounted on the shaft inboard or outboard of the bearings. This flexibility allows researchers to simulate various engine shaft configurations. The bearing support structures include a unique bearing mounting

  5. Performance of a PADC personal neutron dosemeter at simulated and real workplace fields of the nuclear industry

    International Nuclear Information System (INIS)

    Fiechtner, A.; Boschung, M.; Wernli, C.

    2007-01-01

    In the framework of the EVIDOS (Evaluation of Individual Dosimetry in Mixed Neutron and Photon Radiation Fields) project, funded by the EC, measurements with PADC personal neutron dosemeters were carried out at several workplace fields of the nuclear industry and at simulated workplace fields. The measured personal neutron dose equivalents of the PADC personal neutron dosemeter are compared with values that were assessed within the EVIDOS project by other partners. The detection limits for different spectra types are given. In cases were the neutron dose was too low to be measured by the PADC personal neutron dosemeter, the response is estimated by convoluting the responses to monoenergetic neutrons with the dose energy distribution measured within EVIDOS. The advantages and limitations of the PADC personal neutron dosemeter are discussed. (authors)

  6. Prototype Stilbene Neutron Collar

    Energy Technology Data Exchange (ETDEWEB)

    Prasad, M. K. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Shumaker, D. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Snyderman, N. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Verbeke, J. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Wong, J. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2016-10-26

    A neutron collar using stilbene organic scintillator cells for fast neutron counting is described for the assay of fresh low enriched uranium (LEU) fuel assemblies. The prototype stilbene collar has a form factor similar to standard He-3 based collars and uses an AmLi interrogation neutron source. This report describes the simulation of list mode neutron correlation data on various fuel assemblies including some with neutron absorbers (burnable Gd poisons). Calibration curves (doubles vs 235U linear mass density) are presented for both thermal and fast (with Cd lining) modes of operation. It is shown that the stilbene collar meets or exceeds the current capabilities of He-3 based neutron collars. A self-consistent assay methodology, uniquely suited to the stilbene collar, using triples is described which complements traditional assay based on doubles calibration curves.

  7. A nuclear training simulator implementing a capability for multiple, concurrent-training sessions

    International Nuclear Information System (INIS)

    Groeneveld, B.J.; Nannister, D.G.; Estes, K.R.; Johnsen, M.R.

    1996-01-01

    The Advanced Test Reactor (ATR) Simulator at the Test Reactor Area of the Idaho National Engineering Laboratory (INEL) has recently been upgraded to reflect plant installation of a distributed control system (DCS). The ATR Simulator re-design implements traditional needs for software extensibility and plant installation prototyping, but the driving force behind its new design was an instruction requirement for multiple, concurrent-training sessions. Support is provided for up to three concurrent, independent or interacting, training sessions of reactor, balance of plant, and experiment loop operators. This capability has been achieved by modifying the existing design to consistently apply client-server, parent-child, and peer-to-peer processing technologies, and then to encapsulate concurrency software into all interfaces. When the resulting component-oriented design is linked with build and runtime flexibility in a distributed computing environment, traditional needs for extensibility and parallel software and scenario development are satisfied with minimal additional effort. Sensible configuration management practices coupled with the ability to perform piecewise system builds also greatly facilitate prototyping of plant changes prior to installation

  8. Standardization of accelerator irradiation procedures for simulation of neutron induced damage in reactor structural materials

    Science.gov (United States)

    Shao, Lin; Gigax, Jonathan; Chen, Di; Kim, Hyosim; Garner, Frank A.; Wang, Jing; Toloczko, Mychailo B.

    2017-10-01

    Self-ion irradiation is widely used as a method to simulate neutron damage in reactor structural materials. Accelerator-based simulation of void swelling, however, introduces a number of neutron-atypical features which require careful data extraction and, in some cases, introduction of innovative irradiation techniques to alleviate these issues. We briefly summarize three such atypical features: defect imbalance effects, pulsed beam effects, and carbon contamination. The latter issue has just been recently recognized as being relevant to simulation of void swelling and is discussed here in greater detail. It is shown that carbon ions are entrained in the ion beam by Coulomb force drag and accelerated toward the target surface. Beam-contaminant interactions are modeled using molecular dynamics simulation. By applying a multiple beam deflection technique, carbon and other contaminants can be effectively filtered out, as demonstrated in an irradiation of HT-9 alloy by 3.5 MeV Fe ions.

  9. Virtual experiments: the ultimate aim of neutron ray-tracing simulations

    DEFF Research Database (Denmark)

    Lefmann, Kim; Willendrup, Peter Kjær; Udby, Linda

    2008-01-01

    We define a virtual neutron experiment as a complete simulation of an experiment, from source over sample to detector. The virtual experiment (VE) will ideally interface with the instrument control software for the input and with standard data analysis packages for the virtual data output. Virtual...

  10. Monte Carlo simulations of neutron-scattering instruments using McStas

    DEFF Research Database (Denmark)

    Nielsen, K.; Lefmann, K.

    2000-01-01

    Monte Carlo simulations have become an essential tool for improving the performance of neutron-scattering instruments, since the level of sophistication in the design of instruments is defeating purely analytical methods. The program McStas, being developed at Rise National Laboratory, includes...

  11. Simulations of chopper jitter at the LET neutron spectrometer at the ISIS TS2

    DEFF Research Database (Denmark)

    Klenø, Kaspar Hewitt; Lefmann, Kim; Willendrup, Peter Kjær

    2014-01-01

    The effect of uncertainty in chopper phasing (jitter) has been investigated for the high-resolution time-of-flight spectrometer LET at the ISIS second target station. The investigation is carried out using virtual experiments, with the neutron simulation package McStas, where the chopper jitter i...

  12. Neutron star properties from an NJL model modified to simulate confinement

    Energy Technology Data Exchange (ETDEWEB)

    Lawley, S. [Special Research Centre for the Subatomic Structure of Matter, University of Adelaide, Adelaide SA 5005 (Australia); Bentz, W. [Department of Physics, School of Science, Tokai University, 117 Kita-Kaname, Hiratsuka 259-1207 (Japan); Thomas, A.W. [Jefferson Lab, 12000 Jefferson Avenue, Newport News, VA 23606 (United States)

    2005-04-15

    The NJL model has recently been extended with a method to simulate confinement. This leads in mean field approximation to a natural mechanism for the saturation of nuclear matter. We use the model to investigate the equation of state of asymmetric nuclear matter and then use it to compute the properties of neutron stars.

  13. Advances in thermal hydraulic and neutronic simulation for reactor analysis and safety

    International Nuclear Information System (INIS)

    Tentner, A.M.; Blomquist, R.N.; Canfield, T.R.; Ewing, T.F.; Garner, P.L.; Gelbard, E.M.; Gross, K.C.; Minkoff, M.; Valentin, R.A.

    1993-01-01

    This paper describes several large-scale computational models developed at Argonne National Laboratory for the simulation and analysis of thermal-hydraulic and neutronic events in nuclear reactors and nuclear power plants. The impact of advanced parallel computing technologies on these computational models is emphasized

  14. BRAND program complex for neutron-physical experiment simulation by the Monte-Carlo method

    International Nuclear Information System (INIS)

    Androsenko, A.A.; Androsenko, P.A.

    1984-01-01

    Possibilities of the BRAND program complex for neutron and γ-radiation transport simulation by the Monte-Carlo method are described in short. The complex includes the following modules: geometric module, source module, detector module, modules of simulation of a vector of particle motion direction after interaction and a free path. The complex is written in the FORTRAN langauage and realized by the BESM-6 computer

  15. Monte Carlo simulation of neutron detection efficiency for NE213 scintillation detector

    International Nuclear Information System (INIS)

    Xi Yinyin; Song Yushou; Chen Zhiqiang; Yang Kun; Zhangsu Yalatu; Liu Xingquan

    2013-01-01

    A NE213 liquid scintillation neutron detector was simulated by using the FLUKA code. The light output of the detector was obtained by transforming the secondary particles energy deposition using Birks formula. According to the measurement threshold, detection efficiencies can be calculated by integrating the light output. The light output, central efficiency and the average efficiency as a function of the front surface radius of the detector, were simulated and the results agreed well with experimental results. (authors)

  16. Simulation of Thermal Neutron Transport Processes Directly from the Evaluated Nuclear Data Files

    Science.gov (United States)

    Androsenko, P. A.; Malkov, M. R.

    The main idea of the method proposed in this paper is to directly extract thetrequired information for Monte-Carlo calculations from nuclear data files. The met od being developed allows to directly utilize the data obtained from libraries and seehs to be the most accurate technique. Direct simulation of neutron scattering in themmal energy range using file 7 ENDF-6 format in terms of code system BRAND has beer achieved. Simulation algorithms have been verified using the criterion x2

  17. M.C. simulation of GEM neutron beam monitor with 10B

    International Nuclear Information System (INIS)

    Wang Yanfeng; Sun Zhijia; Liu Ben; Zhou Jianrong; Yang Guian; Dong Jing; Xu Hong; Zhou Liang; Huang Guangming; Yang Lei; Li Yi

    2010-01-01

    The neutron beam monitor based on GEM detector has been carefully studied with the Monte-Carlo method in this article. The simulation framework is including the ANSYS and the Garfield, which was used to compute the electric field of GEM foils and simulate the movement of electrons in gas mixture respectively. The GEM foils' focus and extract coefficients have been obtained. According to the primary results, the performing of the monitor is improved. (authors)

  18. Investigating the response of Micromegas detector to low-energy neutrons using Monte Carlo simulation

    Science.gov (United States)

    Khezripour, S.; Negarestani, A.; Rezaie, M. R.

    2017-08-01

    Micromegas detector has recently been used for high-energy neutron (HEN) detection, but the aim of this research is to investigate the response of the Micromegas detector to low-energy neutron (LEN). For this purpose, a Micromegas detector (with air, P10, BF3, 3He and Ar/BF3 mixture) was optimized for the detection of 60 keV neutrons using the MCNP (Monte Carlo N Particle) code. The simulation results show that the optimum thickness of the cathode is 1 mm and the optimum of microgrid location is 100 μm above the anode. The output current of this detector for Ar (3%) + BF3 (97%) mixture is greater than the other ones. This mixture is considered as the appropriate gas for the Micromegas neutron detector providing the output current for 60 keV neutrons at the level of 97.8 nA per neutron. Consecuently, this detector can be introduced as LEN detector.

  19. Simulation of a silicon neutron detector coated with TiB2 absorber

    International Nuclear Information System (INIS)

    Krapohl, D; Nilsson, H-E; Petersson, S; Slavicek, T; Thungström, G; Pospisil, S

    2012-01-01

    Neutron radiation cannot be directly detected in semiconductor detectors and therefore needs converter layers. Planar clean-room processing can be used in the manufacturing process of semiconductor detectors with metal layers to produce a cost-effective device. We used the Geant4 Monte-Carlo toolkit to simulate the performance of a semiconductor neutron detector. A silicon photo-diode was coated with vapour deposited titanium, aluminium thin films and a titaniumdiboride (TiB 2 ) neutron absorber layer. The neutron capture reaction 10B(n, alpha)7Li is taken advantage of to create charged particles that can be counted. Boron-10 has a natural abundance of about SI 19.8%. The emitted alpha particles are absorbed in the underlying silicon detector. We varied the thickness of the converter layer and ran the simulation with a thermal neutron source in order to find the best efficiency of the TiB 2 converter layer and optimize the clean room process.

  20. An immersed body method for coupled neutron transport and thermal hydraulic simulations of PWR assemblies

    International Nuclear Information System (INIS)

    Jewer, S.; Buchan, A.G.; Pain, C.C.; Cacuci, D.G.

    2014-01-01

    Highlights: • A new method of coupled radiation transport, heat and momentum exchanges on fluids, and heat transfer simulations. • Simulation of the thermal hydraulics and radiative properties within whole PWR assemblies. • An immersed body method for modelling complex solid domains on practical computational meshes. - Abstract: A recently developed immersed body method is adapted and used to model a typical pressurised water reactor (PWR) fuel assembly. The approach is implemented with the numerical framework of the finite element, transient criticality code, FETCH which is composed of the neutron transport code, EVENT, and the CFD code, FLUIDITY. Within this framework the neutron transport equation, Navier–Stokes equations and a fluid energy conservation equation are solved in a coupled manner on a coincident structured or unstructured mesh. The immersed body method has been used to model the solid fuel pins. The key feature of this method is that the fluid/neutronic domain and the solid domain are represented by overlapping and non-conforming meshes. The main difficulty of this approach, for which a solution is proposed in this work, is the conservative mapping of the energy and momentum exchange between the fluid/neutronic mesh and the solid fuel pin mesh. Three numerical examples are presented which include a validation of the fuel pin submodel against an analytical solution; an uncoupled (no neutron transport solution) PWR fuel assembly model with a specified power distribution which was validated against the COBRA-EN subchannel analysis code; and finally a coupled model of a PWR fuel assembly with reflective neutron boundary conditions. Coupling between the fluid and neutron transport solutions is through the nuclear cross sections dependence on Doppler fuel temperature, coolant density and temperature, which was taken into account by using pre-calculated cross-section lookup tables generated using WIMS9a. The method was found to show good agreement

  1. High conduction neutron absorber to simulate fast reactor environment in an existing test reactor

    Energy Technology Data Exchange (ETDEWEB)

    Donna Post Guillen; Larry R. Greenwood; James R. Parry

    2014-06-22

    A new metal matrix composite material has been developed to serve as a thermal neutron absorber for testing fast reactor fuels and materials in an existing pressurized water reactor. The performance of this material was evaluated by placing neutron fluence monitors within shrouded and unshrouded holders and irradiating for up to four cycles. The monitor wires were analyzed by gamma and X-ray spectrometry to determine the activities of the activation products. Adjusted neutron fluences were calculated and grouped into three bins—thermal, epithermal, and fast—to evaluate the spectral shift created by the new material. A comparison of shrouded and unshrouded fluence monitors shows a thermal fluence decrease of ~11 % for the shielded monitors. Radioisotope activity and mass for each of the major activation products is given to provide insight into the evolution of thermal absorption cross-section during irradiation. The thermal neutron absorption capability of the composite material appears to diminish at total neutron fluence levels of ~8 × 1025 n/m2. Calculated values for dpa in excess of 2.0 were obtained for two common structural materials (iron and nickel) of interest for future fast flux experiments.

  2. Unfolding neutron spectra from simulated response of thermoluminescence dosimeters inside a polyethylene sphere using GRNN neural network

    Science.gov (United States)

    Lotfalizadeh, F.; Faghihi, R.; Bahadorzadeh, B.; Sina, S.

    2017-07-01

    Neutron spectrometry using a single-sphere containing dosimeters has been developed recently, as an effective replacement for Bonner sphere spectrometry. The aim of this study is unfolding the neutron energy spectra using GRNN artificial neural network, from the response of thermoluminescence dosimeters, TLDs, located inside a polyethylene sphere. The spectrometer was simulated using MCNP5. TLD-600 and TLD-700 dosimeters were simulated at different positions in all directions. Then the GRNN was used for neutron spectra prediction, using the TLDs' readings. Comparison of spectra predicted by the network with the real spectra, show that the single-sphere dosimeter is an effective instrument in unfolding neutron spectra.

  3. HYSPEC: A crystal time-of-flight hybrid spectrometer for the spallation neutron source with polarization capabilities

    International Nuclear Information System (INIS)

    Shapiro, S.M.; Zaliznyak, I.A.; Passell, L.; Ghosh, V.J.; Leonhardt, W.J.; Hagen, M.E.

    2006-01-01

    The hybrid spectrometer (HYSPEC) is a unique direct geometry inelastic scattering instrument under construction at the spallation neutron source (SNS). It combines the intensity enhancement features of focusing Bragg crystals with time-of-flight energy analysis. It will be located at beam-line 14B, which views a coupled liquid hydrogen moderator. A neutron beam from the moderator will travel along a curved guide, through a Fermi chopper and will then be focused onto a sample in an external building, 39 m from the source. In this configuration the intensity at the sample position is more than an order of magnitude larger than for other planned inelastic instrument. A movable detector bank 4.5 m from the sample will cover an angular range of 60 deg. in the horizontal plane and 15 deg. in the vertical direction. An important feature of HYSPEC is the ability to do neutron polarization analysis experiments. A Heusler crystal, which polarizes the neutron beam, can be used as the focusing crystal and a series of bender analyzers will analyze the polarization of the scattered beam

  4. Monte Carlo simulations of the pulsed thermal neutron flux in two-region hydrogenous systems (using standard MCNP data libraries)

    International Nuclear Information System (INIS)

    Wiacek, U.; Krynicka, E.

    2005-02-01

    Monte Carlo simulations of the pulsed neutron experiment in two- region systems (two concentric spheres and two coaxial finite cylinders) are presented. The MCNP code is used. Aqueous solutions of H 3 BO 3 or KCl are used in the inner region. The outer region is the moderator of Plexiglas. Standard data libraries of the thermal neutron scattering cross-sections of hydrogen in hydrogenous substances are used. The time-dependent thermal neutron transport is simulated when the inner region has a constant size and the external size of the surrounding outer region is variable. The time decay constant of the thermal neutron flux in the system is found in each simulation. The results of the simulations are compared with results of real pulsed neutron experiments on the corresponding systems. (author)

  5. Dynamics of crystalline acetanilide: Analysis using neutron scattering and computer simulation

    Science.gov (United States)

    Hayward, R. L.; Middendorf, H. D.; Wanderlingh, U.; Smith, J. C.

    1995-04-01

    The unusual temperature dependence of several optical spectroscopic vibrational bands in crystalline acetanilide has been interpreted as providing evidence for dynamic localization. Here we examine the vibrational dynamics of crystalline acetanilide over a spectral range of ˜20-4000 cm-1 using incoherent neutron scattering experiments, phonon normal mode calculations and molecular dynamics simulations. A molecular mechanics energy function is parametrized and used to perform the normal mode analyses in the full configurational space of the crystal i.e., including the intramolecular and intermolecular degrees of freedom. One- and multiphonon incoherent inelastic neutron scattering intensities are calculated from harmonic analyses in the first Brillouin zone and compared with the experimental data presented here. Phonon dispersion relations and mean-square atomic displacements are derived from the harmonic model and compared with data derived from coherent inelastic neutron scattering and neutron and x-ray diffraction. To examine the temperature effects on the vibrations the full, anharmonic potential function is used in molecular dynamics simulations of the crystal at 80, 140, and 300 K. Several, but not all, of the spectral features calculated from the molecular dynamics simulations exhibit temperature-dependent behavior in agreement with experiment. The significance of the results for the interpretation of the optical spectroscopic results and possible improvements to the model are discussed.

  6. X-ray and neutron diffraction and molecular dynamics simulation of molten lithium and rubidium nitrates

    International Nuclear Information System (INIS)

    Yamaguchi, Toshio; Okada, Isao; Ohtaki, Hitoshi; Mikami, Masuhiro; Kawamura, Kazutaka

    1986-01-01

    Molecular dynamics simulations have been performed for lithium and rubidium nitrate melts at 550 and 600K, respectively, together with X-ray and neutron diffraction experiments. Simple Coulomb pair potentials with Born-type repulsions have been adopted in the simulations with a rigid body model for the nitrate ion. Structure functions derived from the X-ray and neutron experiments are well reproduced by the simulations, from which the three-dimensional cation distribution around the nitrate ion has been revealed. The self-diffusion coefficients, the velocity autocorrelation functions and the self-exchange velocities of lithium, rubidium and nitrate ions have been calculated. Anisotropic motion of nitrate ions has been found and is discussed on the basis of the structure of the melts. (author)

  7. Simulations of Lithium-Based Neutron Coincidence Counter for Gd-Loaded Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Cowles, Christian C. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Kouzes, Richard T. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Siciliano, Edward R. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2014-10-01

    The Department of Energy Office of Nuclear Safeguards and Security (NA-241) is supporting the project Lithium-Based Alternative Neutron Detection Technology Coincidence Counting for Gd-loaded Fuels at Pacific Northwest National Laboratory for the development of a lithium-based neutron coincidence counter for nondestructively assaying Gd loaded nuclear fuel. This report provides results from MCNP simulations of a lithium-based coincidence counter for the possible measurement of Gd-loaded nuclear fuel. A comparison of lithium-based simulations and UNCL-II simulations with and without Gd loaded fuel is provided. A lithium-based model, referred to as PLNS3A-R1, showed strong promise for assaying Gd loaded fuel.

  8. Prediction of the number of 14 MeV neutron elastically scattered from large sample of aluminium using Monte Carlo simulation method

    International Nuclear Information System (INIS)

    Husin Wagiran; Wan Mohd Nasir Wan Kadir

    1997-01-01

    In neutron scattering processes, the effect of multiple scattering is to cause an effective increase in the measured cross-sections due to increase on the probability of neutron scattering interactions in the sample. Analysis of how the effective cross-section varies with thickness is very complicated due to complicated sample geometries and the variations of scattering cross-section with energy. Monte Carlo method is one of the possible method for treating the multiple scattering processes in the extended sample. In this method a lot of approximations have to be made and the accurate data of microscopic cross-sections are needed at various angles. In the present work, a Monte Carlo simulation programme suitable for a small computer was developed. The programme was capable to predict the number of neutrons scattered from various thickness of aluminium samples at all possible angles between 00 to 36011 with 100 increments. In order to make the the programme not too complicated and capable of being run on microcomputer with reasonable time, the calculations was done in two dimension coordinate system. The number of neutrons predicted from this model show in good agreement with previous experimental results

  9. Simulation software of 3-D two-neutron energy groups for ship reactor with hexagonal fuel subassembly

    International Nuclear Information System (INIS)

    Zhang Fan; Cai Zhangsheng; Yu Lei; Gui Xuewen

    2005-01-01

    Core simulation software for 3-D two-neutron energy groups is developed. This software is used to simulate the ship reactor with hexagonal fuel subassembly after 10, 150 and 200 burnup days, considering the hydraulic and thermal feedback. It accurately simulates the characteristics of the fast and thermal neutrons and the detailed power distribution in a reactor under normal and abnormal operation condition. (authors)

  10. SU-G-IeP4-04: DD-Neutron Source Collimation for Neutron Stimulated Emission Computed Tomography: A Monte Carlo Simulation Study

    Energy Technology Data Exchange (ETDEWEB)

    Fong, G; Kapadia, A [Carl E Ravin Advanced Imaging Laboratories, Durham, North Carolina (United States)

    2016-06-15

    Purpose: To optimize collimation and shielding for a deuterium-deuterium (DD) neutron generator for an inexpensive and compact clinical neutron imaging system. The envisioned application is cancer diagnosis through Neutron Stimulated Emission Computed Tomography (NSECT). Methods: Collimator designs were tested with an isotropic 2.5 MeV neutron source through GEANT4 simulations. The collimator is a 52×52×52 cm{sup 3} polyethylene block coupled with a 1 cm lead sheet in sequence. Composite opening was modeled into the collimator to permit passage of neutrons. The opening varied in shape (cylindrical vs. tapered), size (1–5 cm source-side and target-side openings) and aperture placements (13–39 cm from source-side). Spatial and energy distribution of neutrons and gammas were tracked from each collimator design. Parameters analyzed were primary beam width (FWHM), divergence, and efficiency (percent transmission) for different configurations of the collimator. Select resultant outputs were then used for simulated NSECT imaging of a virtual breast phantom containing a 2.5 cm diameter tumor to assess the effect of the collimator on spatial resolution, noise, and scan time. Finally, composite shielding enclosure made of polyethylene and lead was designed and evaluated to block 99.99% of neutron and gamma radiation generated in the system. Results: Analysis of primary beam indicated the beam-width is linear to the aperture size. Increasing source-side opening allowed at least 20% more neutron throughput for all designs relative to the cylindrical openings. Maximum throughput for all designs was 364% relative to cylindrical openings. Conclusion: The work indicates potential for collimating and shielding a DD neutron generator for use in a clinical NSECT system. The proposed collimator designs produced a well-defined collimated neutron beam that can be used to image samples of interest with millimeter resolution. Balance in output efficiency, noise reduction, and scan

  11. SU-G-IeP4-04: DD-Neutron Source Collimation for Neutron Stimulated Emission Computed Tomography: A Monte Carlo Simulation Study

    International Nuclear Information System (INIS)

    Fong, G; Kapadia, A

    2016-01-01

    Purpose: To optimize collimation and shielding for a deuterium-deuterium (DD) neutron generator for an inexpensive and compact clinical neutron imaging system. The envisioned application is cancer diagnosis through Neutron Stimulated Emission Computed Tomography (NSECT). Methods: Collimator designs were tested with an isotropic 2.5 MeV neutron source through GEANT4 simulations. The collimator is a 52×52×52 cm"3 polyethylene block coupled with a 1 cm lead sheet in sequence. Composite opening was modeled into the collimator to permit passage of neutrons. The opening varied in shape (cylindrical vs. tapered), size (1–5 cm source-side and target-side openings) and aperture placements (13–39 cm from source-side). Spatial and energy distribution of neutrons and gammas were tracked from each collimator design. Parameters analyzed were primary beam width (FWHM), divergence, and efficiency (percent transmission) for different configurations of the collimator. Select resultant outputs were then used for simulated NSECT imaging of a virtual breast phantom containing a 2.5 cm diameter tumor to assess the effect of the collimator on spatial resolution, noise, and scan time. Finally, composite shielding enclosure made of polyethylene and lead was designed and evaluated to block 99.99% of neutron and gamma radiation generated in the system. Results: Analysis of primary beam indicated the beam-width is linear to the aperture size. Increasing source-side opening allowed at least 20% more neutron throughput for all designs relative to the cylindrical openings. Maximum throughput for all designs was 364% relative to cylindrical openings. Conclusion: The work indicates potential for collimating and shielding a DD neutron generator for use in a clinical NSECT system. The proposed collimator designs produced a well-defined collimated neutron beam that can be used to image samples of interest with millimeter resolution. Balance in output efficiency, noise reduction, and scan time

  12. A stochastic model for neutron simulation considering the spectrum and nuclear properties with continuous dependence of energy

    International Nuclear Information System (INIS)

    Camargo, Dayana Q. de; Bodmann, Bardo E.J.; Vilhena, Marco T. de; Froehlich, Herberth B.

    2011-01-01

    In this work we developed a stochastic model to simulate neutron transport in a heterogeneous environment, considering continuous neutron spectra and the nuclear properties with its continuous dependence on energy. This model was implemented using the Monte Carlo method for the propagation of neutrons in different environments. Due to restrictions with respect to the number of neutrons that can be simulated in reasonable computational time we introduced a variable control volume together with (pseudo-) periodic boundary conditions in order to overcome this problem. This study allowed a detailed analysis of the influence of energy on the neutron population and its impact on the life cycle of neutrons. From the results, even for a simple geometrical arrangement, we can conclude that there is need to consider the energy dependence and hence defined a spectral effective multiplication factor per Monte Carlo step. (author)

  13. New developments in the McStas neutron instrument simulation package

    International Nuclear Information System (INIS)

    Willendrup, P K; Knudsen, E B; Klinkby, E; Nielsen, T; Farhi, E; Filges, U; Lefmann, K

    2014-01-01

    The McStas neutron ray-tracing software package is a versatile tool for building accurate simulators of neutron scattering instruments at reactors, short- and long-pulsed spallation sources such as the European Spallation Source. McStas is extensively used for design and optimization of instruments, virtual experiments, data analysis and user training. McStas was founded as a scientific, open-source collaborative code in 1997. This contribution presents the project at its current state and gives an overview of the main new developments in McStas 2.0 (December 2012) and McStas 2.1 (expected fall 2013), including many new components, component parameter uniformisation, partial loss of backward compatibility, updated source brilliance descriptions, developments toward new tools and user interfaces, web interfaces and a new method for estimating beam losses and background from neutron optics.

  14. Guideline of Monte Carlo calculation. Neutron/gamma ray transport simulation by Monte Carlo method

    CERN Document Server

    2002-01-01

    This report condenses basic theories and advanced applications of neutron/gamma ray transport calculations in many fields of nuclear energy research. Chapters 1 through 5 treat historical progress of Monte Carlo methods, general issues of variance reduction technique, cross section libraries used in continuous energy Monte Carlo codes. In chapter 6, the following issues are discussed: fusion benchmark experiments, design of ITER, experiment analyses of fast critical assembly, core analyses of JMTR, simulation of pulsed neutron experiment, core analyses of HTTR, duct streaming calculations, bulk shielding calculations, neutron/gamma ray transport calculations of the Hiroshima atomic bomb. Chapters 8 and 9 treat function enhancements of MCNP and MVP codes, and a parallel processing of Monte Carlo calculation, respectively. An important references are attached at the end of this report.

  15. Simulation code for the interaction of 14 MeV neutrons on cells

    Energy Technology Data Exchange (ETDEWEB)

    Nenot, M.L.; Alard, J.P.; Dionet, C.; Arnold, J.; Tchirkov, A.; Meunier, H.; Bodez, V.; Rapp, M.; Verrelle, P

    2002-07-01

    The structure of the survival curve of melanoma cells irradiated by 14 MeV neutrons displays unusual features at very low dose rate where a marked increase in cell killings at 0.05 Gy is followed by a plateau for survival from 0.1 to 0.32 Gy. In parallel a simulation code was constructed for the interaction of 14 MeV neutrons with cellular cultures. The code describes the interaction of the neutrons with the atomic nuclei of the cellular medium and of the external medium (flask culture and culture medium), and is used to compute the deposited energy into the cell volume. It was found that the large energy transfer events associated with heavy charged recoil can occur and that a large part of the energy deposition events are due to recoil protons emitted from the external medium. It is suggested that such events could partially explain the experimental results. (author)

  16. Event-by-event simulation of single-neutron experiments to test uncertainty relations

    International Nuclear Information System (INIS)

    Raedt, H De; Michielsen, K

    2014-01-01

    Results from a discrete-event simulation of a recent single-neutron experiment that tests Ozawa's generalization of Heisenberg's uncertainty relation are presented. The event-based simulation algorithm reproduces the results of the quantum theoretical description of the experiment but does not require the knowledge of the solution of a wave equation, nor does it rely on detailed concepts of quantum theory. In particular, the data from these non-quantum simulations satisfy uncertainty relations derived in the context of quantum theory. (paper)

  17. Synthetic neutron camera and spectrometer in JET based on AFSI-ASCOT simulations

    Science.gov (United States)

    Sirén, P.; Varje, J.; Weisen, H.; Koskela, T.; contributors, JET

    2017-09-01

    The ASCOT Fusion Source Integrator (AFSI) has been used to calculate neutron production rates and spectra corresponding to the JET 19-channel neutron camera (KN3) and the time-of-flight spectrometer (TOFOR) as ideal diagnostics, without detector-related effects. AFSI calculates fusion product distributions in 4D, based on Monte Carlo integration from arbitrary reactant distribution functions. The distribution functions were calculated by the ASCOT Monte Carlo particle orbit following code for thermal, NBI and ICRH particle reactions. Fusion cross-sections were defined based on the Bosch-Hale model and both DD and DT reactions have been included. Neutrons generated by AFSI-ASCOT simulations have already been applied as a neutron source of the Serpent neutron transport code in ITER studies. Additionally, AFSI has been selected to be a main tool as the fusion product generator in the complete analysis calculation chain: ASCOT - AFSI - SERPENT (neutron and gamma transport Monte Carlo code) - APROS (system and power plant modelling code), which encompasses the plasma as an energy source, heat deposition in plant structures as well as cooling and balance-of-plant in DEMO applications and other reactor relevant analyses. This conference paper presents the first results and validation of the AFSI DD fusion model for different auxiliary heating scenarios (NBI, ICRH) with very different fast particle distribution functions. Both calculated quantities (production rates and spectra) have been compared with experimental data from KN3 and synthetic spectrometer data from ControlRoom code. No unexplained differences have been observed. In future work, AFSI will be extended for synthetic gamma diagnostics and additionally, AFSI will be used as part of the neutron transport calculation chain to model real diagnostics instead of ideal synthetic diagnostics for quantitative benchmarking.

  18. Neutron sample cell suitable for the diffraction of aligned biomaterials and capable of exerting up to 370 MPa of hydrostatic pressure

    International Nuclear Information System (INIS)

    Watson, M.J.; Nieh, M.-P.; Harroun, T.A.; Katsaras, J.

    2003-01-01

    We describe a temperature controlled sample cell suitable for the study of biomimetic materials (e.g., lipid bilayers) using neutron diffraction, and capable of exerting hydrostatic pressures of up to 370 MPa. The advantage of this sample cell, compared to previous high-pressure cells of its type, is that it allows for the use of samples aligned on a solid support which, compared to 'powder' or so-called liposomal preparations, requires only small amounts of sample and allows for the clear differentiation between in-plane and out-of-plane structure

  19. Monte Carlo simulation of neutron irradiation facility developed for accelerator based in vivo neutron activation measurements in human hand bones

    International Nuclear Information System (INIS)

    Aslam; Prestwich, W.V.; McNeill, F.E.; Waker, A.J.

    2006-01-01

    The neutron irradiation facility developed at the McMaster University 3 MV Van de Graaff accelerator was employed to assess in vivo elemental content of aluminum and manganese in human hands. These measurements were carried out to monitor the long-term exposure of these potentially toxic trace elements through hand bone levels. The dose equivalent delivered to a patient during irradiation procedure is the limiting factor for IVNAA measurements. This article describes a method to estimate the average radiation dose equivalent delivered to the patient's hand during irradiation. The computational method described in this work augments the dose measurements carried out earlier [Arnold et al., 2002. Med. Phys. 29(11), 2718-2724]. This method employs the Monte Carlo simulation of hand irradiation facility using MCNP4B. Based on the estimated dose equivalents received by the patient hand, the proposed irradiation procedure for the IVNAA measurement of manganese in human hands [Arnold et al., 2002. Med. Phys. 29(11), 2718-2724] with normal (1 ppm) and elevated manganese content can be carried out with a reasonably low dose of 31 mSv to the hand. Sixty-three percent of the total dose equivalent is delivered by non-useful fast group (>10 keV); the filtration of this neutron group from the beam will further decrease the dose equivalent to the patient's hand

  20. D-T neutron streaming experiment simulating narrow gaps in ITER equatorial port

    International Nuclear Information System (INIS)

    Ochiai, K.; Sato, S.; Wada, M.; Iida, H.; Takakura, K.; Kutsukake, C.; Tanaka, S.; Abe, Y.; Konno, C.

    2008-01-01

    Under the ITER/ITA task, we have conducted the neutron streaming experiment simulating narrow and deep gaps at boundaries between ITER vacuum vessel and equatorial port plugs. Micro-fission chambers and some activation foils were used to measure fission rates and reaction rates to evaluate the relative fast and slow neutron fluences along the gap in the experimental assembly. The MCNP4C, TORT and Attila codes were used for the experimental analysis. From comparing our measurements and calculations, the following facts were found: (1) in case of a such narrow and deep gap structure, the calculation with MCNP, TORT and Attila codes and FENDL-2.1 is sufficient to predict fast neutron field inside the gap; (2) by scattering neutrons in the experimental room, experimental error considerably increased at the deeper region than 100 cm; (3) angular quadrature set of upward biased U315 and last collided source calculation on TORT and Attila were very important technique for accurate estimation of neutron transport

  1. SIMULATION OF CARGO CONTAINER INTERROGATION BY D-D NEUTRONS

    International Nuclear Information System (INIS)

    Lou, Tak Pui; Antolak, Arlyn

    2007-01-01

    High fidelity, three-dimensional computer models based on a CAD drawing of an intermodal cargo container, representative payload objects, and detector array panels were developed to simulate the underlying physical events taking place during active interrogation. These computer models are used to assess the performance of interrogation systems with different sources and detection schemes. In this presentation, we will show that the use oversimplified models, such as analyzing homogenized payloads only, can lead to errors in determining viable approaches for interrogation

  2. Automating Embedded Analysis Capabilities and Managing Software Complexity in Multiphysics Simulation, Part I: Template-Based Generic Programming

    Directory of Open Access Journals (Sweden)

    Roger P. Pawlowski

    2012-01-01

    Full Text Available An approach for incorporating embedded simulation and analysis capabilities in complex simulation codes through template-based generic programming is presented. This approach relies on templating and operator overloading within the C++ language to transform a given calculation into one that can compute a variety of additional quantities that are necessary for many state-of-the-art simulation and analysis algorithms. An approach for incorporating these ideas into complex simulation codes through general graph-based assembly is also presented. These ideas have been implemented within a set of packages in the Trilinos framework and are demonstrated on a simple problem from chemical engineering.

  3. General relativistic simulations of binary neutron star mergers

    Energy Technology Data Exchange (ETDEWEB)

    Giacomazzo, Bruno [Trento Univ. (Italy)

    2016-11-01

    Currently, we are running additional simulations to investigate additional models for the properties of dense matter. Furthermore, we are using remote visualization resources provided by LRZ to produce movies showing 3D visualizations of our simulations, which will be available soon on the web page of our group. One of the main challenges for our simulations is the fact that some important effects leading to magnetic field amplification happen on small length scales. This makes it very difficult to resolve them numerically. In order to further improve the accuracy, we proposed a follow- up study in which we will evolve one or more models with very high resolution and then use the results to calibrate a so-called sub-grid model, which is designed to capture the field amplification on scales not resolved with the lower, more affordable resolutions. Once calibrated, the sub-grid approach will allow to investigate a large number of models without the need for very high resolutions.

  4. General relativistic simulations of binary neutron star mergers

    International Nuclear Information System (INIS)

    Giacomazzo, Bruno

    2016-01-01

    Currently, we are running additional simulations to investigate additional models for the properties of dense matter. Furthermore, we are using remote visualization resources provided by LRZ to produce movies showing 3D visualizations of our simulations, which will be available soon on the web page of our group. One of the main challenges for our simulations is the fact that some important effects leading to magnetic field amplification happen on small length scales. This makes it very difficult to resolve them numerically. In order to further improve the accuracy, we proposed a follow- up study in which we will evolve one or more models with very high resolution and then use the results to calibrate a so-called sub-grid model, which is designed to capture the field amplification on scales not resolved with the lower, more affordable resolutions. Once calibrated, the sub-grid approach will allow to investigate a large number of models without the need for very high resolutions.

  5. MCNP efficiency calculations of INEEL passive active neutron assay system for simulated TRU waste assays

    International Nuclear Information System (INIS)

    Yoon, W.Y.; Meachum, T.R.; Blackwood, L.G.; Harker, Y.D.

    2000-01-01

    The Idaho National Engineering and Environmental Laboratory Stored Waste Examination Pilot Plant (SWEPP) passive active neutron (PAN) radioassay system is used to certify transuranic (TRU) waste drums in terms of quantifying plutonium and other TRU element activities. Depending on the waste form involved, significant systematic and random errors need quantification in addition to the counting statistics. To determine the total uncertainty of the radioassay results, a statistical sampling and verification approach has been developed. In this approach, the total performance of the PAN nondestructive assay system is simulated using the computer models of the assay system, and the resultant output is compared with the known input to assess the total uncertainty. The supporting steps in performing the uncertainty analysis for the passive assay measurements in particular are as follows: (1) Create simulated waste drums and associated conditions; (2) Simulate measurements to determine the basic counting data that would be produced by the PAN assay system under the conditions specified; and (3) Apply the PAN assay system analysis algorithm to the set of counting data produced by simulating measurements to determine the measured plutonium mass. The validity of this simulation approach was verified by comparing simulated output against results from actual measurements using known plutonium sources and surrogate waste drums. The computer simulation of the PAN system performance uses the Monte Carlo N-Particle (MCNP) Code System to produce a neutron transport calculation for a simulated waste drum. Specifically, the passive system uses the neutron coincidence counting technique, utilizing the spontaneous fission of 240 Pu. MCNP application to the SWEPP PAN assay system uncertainty analysis has been very useful for a variety of waste types contained in 208-ell drums measured by a passive radioassay system. The application of MCNP to the active radioassay system is also feasible

  6. Molecular simulation study to examine the possibility of detecting collective motion in protein by inelastic neutron scattering

    International Nuclear Information System (INIS)

    Yasumasa, Joti; Nobuhiro, Go; Akio, Kitao; Nobuhiro, Go

    2003-01-01

    Inelastic and quasielastic neutron scattering gives the information on the dynamics of biological macromolecules. The combination of computer simulation with neutron scattering experiments allows us to characterize a wide range of dynamical phenomena in condensed phase bio-molecular systems. In this work, the dynamic structure factors in (Q,ω)-space were calculated by using the results of bio-molecular simulations. From the simulated inelastic neutron scattering spectra, we discuss the (Q,ω)-range and the resolution of a detector needed to observe function-related protein dynamics. (authors)

  7. Gravitational waveforms for neutron star binaries from binary black hole simulations

    Science.gov (United States)

    Barkett, Kevin; Scheel, Mark; Haas, Roland; Ott, Christian; Bernuzzi, Sebastiano; Brown, Duncan; Szilagyi, Bela; Kaplan, Jeffrey; Lippuner, Jonas; Muhlberger, Curran; Foucart, Francois; Duez, Matthew

    2016-03-01

    Gravitational waves from binary neutron star (BNS) and black-hole/neutron star (BHNS) inspirals are primary sources for detection by the Advanced Laser Interferometer Gravitational-Wave Observatory. The tidal forces acting on the neutron stars induce changes in the phase evolution of the gravitational waveform, and these changes can be used to constrain the nuclear equation of state. Current methods of generating BNS and BHNS waveforms rely on either computationally challenging full 3D hydrodynamical simulations or approximate analytic solutions. We introduce a new method for computing inspiral waveforms for BNS/BHNS systems by adding the post-Newtonian (PN) tidal effects to full numerical simulations of binary black holes (BBHs), effectively replacing the non-tidal terms in the PN expansion with BBH results. Comparing a waveform generated with this method against a full hydrodynamical simulation of a BNS inspiral yields a phase difference of < 1 radian over ~ 15 orbits. The numerical phase accuracy required of BNS simulations to measure the accuracy of the method we present here is estimated as a function of the tidal deformability parameter λ.

  8. Application of Nonnegative Tensor Factorization for neutron-gamma discrimination of Monte Carlo simulated fission chamber’s output signals

    Directory of Open Access Journals (Sweden)

    Mounia Laassiri

    Full Text Available For efficient exploitation of research reactors, it is important to discern neutron flux distribution inside the reactor with the best possible precision. For this reason, fission and ionization chambers are used to measure the neutron field. In these arrays, the sequences of the neutron interaction points in the fission chamber can correctly be identified in order to obtain true neutron energies emitted by nuclei of interest. However, together with the neutrons, gamma-rays are also emitted from nuclei and thereby affect neutron spectra. The originality of this study consists in the application of tensor based blind source separation methods to extract independent components from signals recorded at the fission chamber preamplifier’s output. The objective is to achieve software neutron-gamma discrimination using Nonnegative Tensor Factorization tools. For reasons of nuclear safety, we first simulate the neutron flux inside the TRIGA Mark II Reactor using Monte Carlo methods under Geant4 platform linked to Garfield++. Geant4 simulations allow the fission chamber construction whereas linking the model to Garfield++ permits to simulate drift parameters from the ionization of the filling gas, which is not possible otherwise. Keywords: Fission chamber (FC, Geant4, Garfield++, Neutron-gamma discrimination, Nonnegative Tensor Factorization (NTF

  9. Multiscale simulation of neutron induced damage in tritium breeding blankets with different spectral shifters

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Yong Hee; Joo, Han Gyu, E-mail: joohan@snu.ac.kr

    2013-10-15

    Highlights: • A multiscale defect simulation system tailored for neutron damage estimation is introduced. • The new recoil spectrum code can use the most recent ENDF-B/VII nuclear data. • The high energy cascades are broken into subcascades using the INCAS model. • OKMC simulation provides data for shear stress estimation using dislocation dynamics formula. • Demonstration is made with a fusion blanket design having different spectral shifters. -- Abstract: A multiscale material defect simulation established to evaluate neutron induced damages on metals is applied to an estimation of material degradation in helium cooled molten lithium blankets in which four different spectral shifter materials are examined as a means of maximizing the tritium breeding ratio through proper shaping of the neutron spectrum. The multiscale system consists of a Monte Carlo neutron transport code, a recoil spectrum generation code, a molecular dynamics code, a high energy cascade breakup model, an object kinetic Monte Carlo code, and a simple formula as the shear stress estimator. The average recoil energy of the primary knock-on atoms, the total concentration of the defects, average defect sizes, and the increase in shear stress after a certain irradiation time are calculated for each spectral shifter. Among the four proposed materials of B4C, Be, Graphite and TiC, B4C reveals the best shielding performance in terms of neutron radiation hardening. The result for the increase in shear stress after 100 days of irradiation indicates that the increased shear stress is 1.5 GPa for B4C which is about 40% less than that of the worst one, the graphite spectral shifter. The other damage indicators show consistent trends.

  10. GNES-R: Global nuclear energy simulator for reactors task 1: High-fidelity neutron transport

    International Nuclear Information System (INIS)

    Clarno, K.; De Almeida, V.; D'Azevedo, E.; De Oliveira, C.; Hamilton, S.

    2006-01-01

    A multi-laboratory, multi-university collaboration has formed to advance the state-of-the-art in high-fidelity, coupled-physics simulation of nuclear energy systems. We are embarking on the first-phase in the development of a new suite of simulation tools dedicated to the advancement of nuclear science and engineering technologies. We seek to develop and demonstrate a new generation of multi-physics simulation tools that will explore the scientific phenomena of tightly coupled physics parameters within nuclear systems, support the design and licensing of advanced nuclear reactors, and provide benchmark quality solutions for code validation. In this paper, we have presented the general scope of the collaborative project and discuss the specific challenges of high-fidelity neutronics for nuclear reactor simulation and the inroads we have made along this path. The high-performance computing neutronics code system utilizes the latest version of SCALE to generate accurate, problem-dependent cross sections, which are used in NEWTRNX - a new 3-D, general-geometry, discrete-ordinates solver based on the Slice-Balance Approach. The Global Nuclear Energy Simulator for Reactors (GNES-R) team is embarking on a long-term simulation development project that encompasses multiple laboratories and universities for the expansion of high-fidelity coupled-physics simulation of nuclear energy systems. (authors)

  11. Simulation of Neutron-Induced Prompt Gamma-ray Spectra Emitted from Fake Tungsten Gold Bar

    Energy Technology Data Exchange (ETDEWEB)

    Lee, K. M.; Sum, G. M. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    Fake gold bars on the market cannot be identified easily without testing because they have the same appearance as a pure gold bar. A non-destructive monitoring method is needed to avoid the trading of fake gold bars on the market. The ultimate goal of this study is to find a fake gold bar detection method using a PGAA (Prompt Gamma Activation Analysis). Using existing data, the number of neutron capture for gold and tungsten in fake tungsten gold bar was calculated and a Monte Carlo simulation for the prompt neutron-induced gamma-ray spectra was conducted. A simulation for neutron-induced prompt gamma-rays spectra when a neutron beam is irradiated onto pure and fake gold bars was successfully conducted. Through a comparison between the prompt gamma-ray spectra of the pure gold bar and those of the fake gold bar, it was concluded that the observation of prompt high-energy gamma-rays from tungsten or a reduction of prompt gamma-rays from gold can be evidence of a fake gold bar. The possibility for detecting a fake gold bar using a PGAA facility was verified.

  12. Simulation of Neutron-Induced Prompt Gamma-ray Spectra Emitted from Fake Tungsten Gold Bar

    International Nuclear Information System (INIS)

    Lee, K. M.; Sum, G. M.

    2016-01-01

    Fake gold bars on the market cannot be identified easily without testing because they have the same appearance as a pure gold bar. A non-destructive monitoring method is needed to avoid the trading of fake gold bars on the market. The ultimate goal of this study is to find a fake gold bar detection method using a PGAA (Prompt Gamma Activation Analysis). Using existing data, the number of neutron capture for gold and tungsten in fake tungsten gold bar was calculated and a Monte Carlo simulation for the prompt neutron-induced gamma-ray spectra was conducted. A simulation for neutron-induced prompt gamma-rays spectra when a neutron beam is irradiated onto pure and fake gold bars was successfully conducted. Through a comparison between the prompt gamma-ray spectra of the pure gold bar and those of the fake gold bar, it was concluded that the observation of prompt high-energy gamma-rays from tungsten or a reduction of prompt gamma-rays from gold can be evidence of a fake gold bar. The possibility for detecting a fake gold bar using a PGAA facility was verified

  13. A New Coupled CFD/Neutron Kinetics System for High Fidelity Simulations of LWR Core Phenomena: Proof of Concept

    Directory of Open Access Journals (Sweden)

    Jorge Pérez Mañes

    2014-01-01

    Full Text Available The Institute for Neutron Physics and Reactor Technology (INR at the Karlsruhe Institute of Technology (KIT is investigating the application of the meso- and microscale analysis for the prediction of local safety parameters for light water reactors (LWR. By applying codes like CFD (computational fluid dynamics and SP3 (simplified transport reactor dynamics it is possible to describe the underlying phenomena in a more accurate manner than by the nodal/coarse 1D thermal hydraulic coupled codes. By coupling the transport (SP3 based neutron kinetics (NK code DYN3D with NEPTUNE-CFD, within a parallel MPI-environment, the NHESDYN platform is created. The newly developed system will allow high fidelity simulations of LWR fuel assemblies and cores. In NHESDYN, a heat conduction solver, SYRTHES, is coupled to NEPTUNE-CFD. The driver module of NHESDYN controls the sequence of execution of the solvers as well as the communication between the solvers based on MPI. In this paper, the main features of NHESDYN are discussed and the proof of the concept is done by solving a single pin problem. The prediction capability of NHESDYN is demonstrated by a code-to-code comparison with the DYNSUB code. Finally, the future developments and validation efforts are highlighted.

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

    Energy Technology Data Exchange (ETDEWEB)

    Burns, Kimberly A. [Georgia Inst. of Technology, Atlanta, GA (United States)

    2009-08-01

    The accurate and efficient simulation of coupled neutron-photon problems is necessary for several important radiation detection applications. Examples include the detection of nuclear threats concealed in cargo containers and prompt gamma neutron activation analysis for nondestructive determination of elemental composition of unknown samples.

  15. Advanced Simulation Capability for Turbopump Cavitation Dynamics Guided by Experimental Validation, Phase I

    Data.gov (United States)

    National Aeronautics and Space Administration — Numerical cavitation modeling capability is critical in the design of liquid rocket engine turbopumps, feed lines, injector manifolds and engine test facilities....

  16. Simulation of the spherical experimental assemblies for the mixed neutron-gamma reference fields implementation

    International Nuclear Information System (INIS)

    Kluson, J.; Jansky, B.

    2009-01-01

    Reference mixed neutron-gamma fields are used for test and calibration of dosimetric and spectrometric systems, intercomparison measurements, and benchmark tests and represent experimental base for reactor studies. Set of the spherical experimental assemblies for the mixed neutron-gamma reference fields implementation was build in the NRI Rez. Extended sets of measurements and simulation calculations were done to describe the reference mixed field dosimetry and spectral characteristics with best achievable precision. The Monte Carlo technique was used for different experimental setups models description, comparison and verification and field characteristics simulation. Effects (hardly distinguishable experimentally) were also studied ( contributions from individual parts of experimental setup, field individual components and next effects as shadow shield cones transparency, etc.). Some results and main conclusions of these studies and calculations are presented and discussed. (authors)

  17. Peculiarities of approximation for reactor neutron energy spectra during computerized simulation of radiation defects

    International Nuclear Information System (INIS)

    Kupchishin, A.A.; Kupchishin, A.I.; Stusik, G.; Omarbekova, Zh.

    2001-01-01

    Peculiarities of approximation for reactor neutron energy spectra during radiation defects computerized simulation were discussed. Approximation of neutron spectra N(E) was carried out by N(E)=α·exp(-β·E)·sh(γ·E) formula (1), where α, β, γ - approximation coefficients. In the capacity of operating reactor data experimental data on 235 U and 239 Pu were applied. The algorithm was designed, and acting soft ware for spectra parameters calculation was developed. The following values of approximation parameters were obtained: α=80.8; β=0.935;γ=2.04 (for uranium and plutonium these coefficients are less distinguishing). Then with use of formula 1 and α, β, γ coefficients the approximation curves were constructed. These curves satisfactorily describe existing experimental data and allowing to use its for radiation defects simulation in the reactor materials

  18. Simulation of neutron multiplicity measurements using Geant4. Open source software for nuclear arms control

    Energy Technology Data Exchange (ETDEWEB)

    Kuett, Moritz

    2016-07-07

    Nuclear arms control, including nuclear safeguards and verification technologies for nuclear disarmament typically use software as part of many different technological applications. This thesis proposes to use three open source criteria for such software, allowing users and developers to have free access to a program, have access to the full source code and be able to publish modifications for the program. This proposition is presented and analyzed in detail, together with the description of the development of ''Open Neutron Multiplicity Simulation'', an open source software tool to simulate neutron multiplicity measurements. The description includes physical background of the method, details of the developed program and a comprehensive set of validation calculations.

  19. Simulation of the spherical experimental assemblies for the mixed neutron-gamma reference fields implementation

    International Nuclear Information System (INIS)

    Kluson, J.; Jansky, B.

    2008-01-01

    Reference mixed neutron-gamma fields are used for test and calibration of dosimetric and spectrometric systems, intercomparison measurements, and benchmark tests and represent experimental base for reactor studies. Set of the spherical experimental assemblies for the mixed neutron-gamma reference fields implementation was build in the NRI Rez. Extended sets of measurements and simulation calculations were done to describe the reference mixed field dosimetry and spectral characteristics with best achievable precision. The Monte Carlo technique was used for different experimental setups models description, comparison and verification and field characteristics simulation. Effects (hardly distinguishable experimentally) were also studied ( contributions from individual parts of experimental setup, field individual components and next effects as shadow shield cones transparency, etc.). Some results and main conclusions of these studies and calculations are presented and discussed. (authors)

  20. Monte Carlo simulation for neutron yield produced by bombarding thick targets with high energy heavy ions

    Energy Technology Data Exchange (ETDEWEB)

    Oranj, Leila Mokhtari; Oh, Joo Hee; Yoon, Moo Hyun; Lee, Hee Seock [POSTECH, Pohang (Korea, Republic of)

    2013-04-15

    One of radiation shielding issues at heavy-ion accelerator facilities is to estimate neutron production by primary heavy ions. A few Monte Carlo transport codes such as FLUKA and PHITS can work with primary heavy ions. Recently IBS/RISP((Rare Isotope Science Project) started to design a high-energy, high-power rare isotope accelerator complex for nuclear physics, medical and material science and applications. There is a lack of experimental and simulated data about the interaction of major beam, {sup 238}U with materials. For the shielding design of the end of first accelerating section section, we calculate a differential neutron yield using the FLUKA code for the interaction of 18.5 MeV/u uranium ion beam with thin carbon stripper of 1.3 μm). The benchmarking studies were also done to prove the yield calculation for 400 MeV/n {sup 131}Xe and other heavy ions. In this study, the benchmarking for Xe-C, Xe-Cu, Xe-Al, Xe-Pb and U-C, other interactions were performed using the FLUKA code. All of results show that the FLUKA can evaluate the heavy ion induced reaction with good uncertainty. For the evaluation of neutron source term, the calculated neutron yields are shown in Fig. 2. The energy of Uranium ion beam is only 18.5 MeV/u, but the energy of produced secondary neutrons was extended over 100 MeV. So the neutron shielding and the damage by those neutrons is expected to be serious. Because of thin stripper, the neutron intensity at forward direction was high. But the the intensity of produced secondary photons was relatively low and mostly the angular property was isotropic. For the detail shielding design of stripper section of RISP rare istope accelerator, the benchmarking study and preliminary evaluation of neutron source term from uranium beam have been carried out using the FLUKA code. This study is also compared with the evaluation results using the PHITS code performed coincidently. Both studies shows that two monte carlo codes can give a good results for

  1. Modeling and Simulation Optimization and Feasibility Studies for the Neutron Detection without Helium-3 Project

    Energy Technology Data Exchange (ETDEWEB)

    Ely, James H.; Siciliano, Edward R.; Swinhoe, Martyn T.; Lintereur, Azaree T.

    2013-01-01

    This report details the results of the modeling and simulation work accomplished for the ‘Neutron Detection without Helium-3’ project during the 2011 and 2012 fiscal years. The primary focus of the project is to investigate commercially available technologies that might be used in safeguards applications in the relatively near term. Other technologies that are being developed may be more applicable in the future, but are outside the scope of this study.

  2. Validation of Monte Carlo simulation of neutron production in a spallation experiment

    Czech Academy of Sciences Publication Activity Database

    Zavorka, L.; Adam, Jindřich; Artiushenko, M.; Baldin, A. A.; Brudanin, V. B.; Katovsky, K.; Suchopár, M.; Svoboda, Ondřej; Vrzalová, Jitka; Wagner, Vladimír

    2015-01-01

    Roč. 80, JUN (2015), s. 178-187 ISSN 0306-4549 R&D Projects: GA MŠk LA08002; GA MŠk LG14004 Institutional support: RVO:61389005 Keywords : accelerator-driven systems * uranium spallation target * neutron emission * activation measurement * Monte Carlo simulation Subject RIV: BG - Nuclear, Atomic and Molecular Physics, Colliders Impact factor: 1.174, year: 2015

  3. Hydrodynamic simulations of a combined hydrogen, helium thermonuclear runaway on a 10-km neutron star

    International Nuclear Information System (INIS)

    Starrfield, S.; Kenyon, S.; Truran, J.W.; Sparks, W.M.

    1983-01-01

    We have used a Lagrangian, hydrodynamic stellar-evolution computer code to evolve a thermonuclear runaway in the accreted hydrogen rich envelope of a 1.0M, 10-km neutron star. Our simulation produced an outburst which lasted about 2000 sec and peak effective temperature was 3 keV. The peak luminosity exceeded 2 x 10 5 L. A shock wave caused a precursor in the light curve which lasted 10 -5 sec

  4. Statistical analysis for discrimination of prompt gamma ray peak induced by high energy neutron: Monte Carlo simulation study

    International Nuclear Information System (INIS)

    Do-Kun Yoon; Joo-Young Jung; Tae Suk Suh; Seong-Min Han

    2015-01-01

    The purpose of this research is a statistical analysis for discrimination of prompt gamma ray peak induced by the 14.1 MeV neutron particles from spectra using Monte Carlo simulation. For the simulation, the information of 18 detector materials was used to simulate spectra by the neutron capture reaction. The discrimination of nine prompt gamma ray peaks from the simulation of each detector material was performed. We presented the several comparison indexes of energy resolution performance depending on the detector material using the simulation and statistics for the prompt gamma activation analysis. (author)

  5. Design of a graphite-moderated {sup 241}Am-Li neutron field to simulate reactor spectra

    Energy Technology Data Exchange (ETDEWEB)

    Tsujimura, N., E-mail: tsujimura.norio@jaea.go.j [Nuclear Fuel Cycle Engineering Laboratories, Japan Atomic Energy Agency, 4-33, Tokai-mura, Ibaraki-ken, 319-1194 (Japan); Yoshida, T. [Nuclear Fuel Cycle Engineering Laboratories, Japan Atomic Energy Agency, 4-33, Tokai-mura, Ibaraki-ken, 319-1194 (Japan)

    2010-12-15

    A neutron calibration field using {sup 241}Am-Li sources and a moderator was designed to simulate the neutron fields found outside a reactor. The moderating assembly selected for the design calculation consists of a cube of graphite blocks with dimensions of 50 cm by 50 cm by 50 cm, in which the {sup 241}Am-Li sources are placed. Monte Carlo calculations revealed the optimal depth of the source to be 15 cm. This moderated neutron source can be used to provide a test field that has a large number of intermediate energy neutrons with a small portion of MeV component.

  6. Neutron spectrometry around the VENUS reactor using Monte Carlo simulations and Bonner spheres measurements

    International Nuclear Information System (INIS)

    Coeck, M.; Lacoste, V.; Muller, H.

    2005-01-01

    Full text: Reliable determination of neutron doses in workplaces is still an issue in the field of radiation protection. The EVIDOS project ('evaluation of individual dosimetry in mixed neutron and photon radiation fields', 5FP supported by the EC) aims to evaluate different methods for individual dosimetry in mixed neutron-photon workplaces in nuclear industry, and focuses on the neutron component. This objective cannot be reached on the basis of investigations in calibration fields only, but requires studies in representative workplaces of the nuclear industry. The VENUS reactor, a zero-power research reactor established by the SCK·CEN, was chosen as one of these workplaces. This paper presents the assessment of the neutron field near the VENUS reactor, particularly in areas near the reactor shielding and in the control room where operators are frequently present during a reactor run. From the neutron spectrum, an evaluation of H*(10) can be made. MCNPX simulations were performed to obtain a reference spectrum at the two areas of interest. Using a k eff calculation the source term was acquired which was subsequently used in a fixed source MCNPX model of the complete shielding geometry of the reactor hall. Reference spectrometry was also performed using a Bonner spheres system. The unfolding spectra were obtained using the NUBAY and GRAVEL codes. The NUBAY program, based on Bayesian parameter estimation methods, assumes a parameterized spectrum and provides posterior probability distributions for both the set of parameters and a set of integral quantities. The code GRAVEL, an iterative algorithm based on SAND-II, was used with various default spectra, among them the NUBAY solution. Bonner spheres data GRAVEL unfolding was also performed using the MCNPX spectra as an initial guess. In this paper the outcome of both calculations and measurements is compared. (author)

  7. Simulating irradiation hardening in tungsten under fast neutron irradiation including Re production by transmutation

    Science.gov (United States)

    Huang, Chen-Hsi; Gilbert, Mark R.; Marian, Jaime

    2018-02-01

    Simulations of neutron damage under fusion energy conditions must capture the effects of transmutation, both in terms of accurate chemical inventory buildup as well as the physics of the interactions between transmutation elements and irradiation defect clusters. In this work, we integrate neutronics, primary damage calculations, molecular dynamics results, Re transmutation calculations, and stochastic cluster dynamics simulations to study neutron damage in single-crystal tungsten to mimic divertor materials. To gauge the accuracy and validity of the simulations, we first study the material response under experimental conditions at the JOYO fast reactor in Japan and the High Flux Isotope Reactor at Oak Ridge National Laboratory, for which measurements of cluster densities and hardening levels up to 2 dpa exist. We then provide calculations under expected DEMO fusion conditions. Several key mechanisms involving Re atoms and defect clusters are found to govern the accumulation of irradiation damage in each case. We use established correlations to translate damage accumulation into hardening increases and compare our results to the experimental measurements. We find hardening increases in excess of 5000 MPa in all cases, which casts doubts about the integrity of W-based materials under long-term fusion exposure.

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

    Science.gov (United States)

    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.

  9. IFMIF [International Fusion Materials Irradiation Facility], an accelerator-based neutron source for fusion components irradiation testing: Materials testing capabilities

    International Nuclear Information System (INIS)

    Mann, F.M.

    1988-08-01

    The International Fusion Materials Irradiation Facility (IFMIF) is proposed as an advanced accelerator-based neutron source for high-flux irradiation testing of large-sized fusion reactor components. The facility would require only small extensions to existing accelerator and target technology originally developed for the Fusion Materials Irradiation Test (FMIT) facility. At the extended facility, neutrons would be produced by a 0.1-A beam of 35-MeV deuterons incident upon a liquid lithium target. The volume available for high-flux (>10/sup 15/ n/cm/sup 2/-s) testing in IFMITF would be over a liter, a factor of about three larger than in the FMIT facility. This is because the effective beam current of 35-MeV deuterons on target can be increased by a factor of ten to 1A or more. Such an increase can be accomplished by funneling beams of deuterium ions from the radio-frequency quadruple into a linear accelerator and by taking advantage of recent developments in accelerator technology. Multiple beams and large total current allow great variety in available testing. For example, multiple simultaneous experiments, and great flexibility in tailoring spatial distributions of flux and spectra can be achieved. 5 refs., 2 figs., 1 tab

  10. Monte Carlo simulation of moderator and reflector in coal analyzer based on a D-T neutron generator.

    Science.gov (United States)

    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. Copyright © 2015 Elsevier Ltd. All rights reserved.

  11. Neutron activation analysis of maltenes recovered from EUROBITUM simulates

    International Nuclear Information System (INIS)

    Impens, N

    2006-01-01

    the bitumen. Moreover, the removal of maltenes and soluble salts results in an important volume reduction of the primary waste. The room temperature method is quite safe, as liquids with high flash points are used, and because volatilisation of radionuclides is avoided. The aim is to study the elemental composition of the non-aqueous secondary waste streams of the room temperature re-treatment method for Eurobitum. It is necessary to assess the purity of maltene solutions recovered from Eurobitum simulates in order to define experimental conditions leading to the free release of the recovered maltene solutions. As the detection limit should be very low and as we only used self-made non-radioactive Eurobitum simulates, NAA is the most suitable technique for this study

  12. Simulation of neutron transport process, photons and charged particles within the Monte Carlo method

    International Nuclear Information System (INIS)

    Androsenko, A.A.; Androsenko, P.A.; Artamonov, S.N.; Bolonkina, G.V.; Lomtev, V.L.; Pupko, S.V.

    1991-01-01

    Description is given to the program system BRAND designed for the accurate solution of non-stationary transport equation of neutrons, photons and charged particles in the conditions of real three-dimensional geometry. An extensive set of local and non-local estimates provides an opportunity of calculating a great set of linear functionals normally being of interest in the calculation of reactors, radiation protection and experiment simulation. The process of particle interaction with substance is simulated on the basis of individual non-group data on each isotope of the composition. 24 refs

  13. Nodal deterministic simulation for problems of neutron shielding in multigroup formulation

    International Nuclear Information System (INIS)

    Baptista, Josue Costa; Heringer, Juan Diego dos Santos; Santos, Luiz Fernando Trindade; Alves Filho, Hermes

    2013-01-01

    In this paper, we propose the use of some computational tools, with the implementation of numerical methods SGF (Spectral Green's Function), making use of a deterministic model of transport of neutral particles in the study and analysis of a known and simplified problem of nuclear engineering, known in the literature as a problem of neutron shielding, considering the model with two energy groups. These simulations are performed in MatLab platform, version 7.0, and are presented and developed with the help of a Computer Simulator providing a friendly computer application for their utilities

  14. Fast neutron spectroscopy by gas proton-recoil methods at the light water reactor pressure vessel simulator

    International Nuclear Information System (INIS)

    Rogers, J.W.

    1980-10-01

    Fast neutron spectrum measurements were made in a Light Water Reactor (LWR) Pressure Vessel Simulator (PVS) to provide neutron spectral definition required to appropriately perform and interpret neutron dosimetry measurements related to fast neutron damage in LWR-PV steels. Proton-recoil proportional counter methods using hydrogen and methane gas-filled detectors were applied to obtain the proton spectra from which the neutron spectra were derived. Cylindrical and spherical geometry detectors were used to cover the neutron energy range between 50 keV and 2 MeV. Results show that the neutron spectra shift in energy distribution toward lower energy between the front and back of a PVS. The relative neutron flux densities increase in this energy range with increasing thickness of the steel. Neutron spectrum fine structure shapes and changes are observed. These results should assist in the generation of more accurate effective cross sections and fluences for use in LWR-PV fast neutron dosimetry and materials damage analyses

  15. Neutrons Flux Distributions of the Pu-Be Source and its Simulation by the MCNP-4B Code

    Science.gov (United States)

    Faghihi, F.; Mehdizadeh, S.; Hadad, K.

    Neutron Fluence rate of a low intense Pu-Be source is measured by Neutron Activation Analysis (NAA) of 197Au foils. Also, the neutron fluence rate distribution versus energy is calculated using the MCNP-4B code based on ENDF/B-V library. Theoretical simulation as well as our experimental performance are a new experience for Iranians to make reliability with the code for further researches. In our theoretical investigation, an isotropic Pu-Be source with cylindrical volume distribution is simulated and relative neutron fluence rate versus energy is calculated using MCNP-4B code. Variation of the fast and also thermal neutrons fluence rate, which are measured by NAA method and MCNP code, are compared.

  16. Simulation for developing new pulse neutron spectrometers I. Creation of new McStas components of moderators of JSNS

    CERN Document Server

    Tamura, I; Arai, M; Harada, M; Maekawa, F; Shibata, K; Soyama, K

    2003-01-01

    Moderators components of the McStas code have been created for the design of JSNS instruments. Three cryogenic moderators are adopted in JSNS, one is coupled H sub 2 moderators for high intensity experiments and other two are decoupled H sub 2 with poisoned or unpoisoned for high resolution moderators. Since the characteristics of neutron beams generated from moderators make influence on the performance of pulse neutron spectrometers, it is important to perform the Monte Carlo simulation with neutron source component written precisely. The neutron spectrum and time structure were calculated using NMTC/JAERI97 and MCNP4a codes. The simulation parameters, which describe the pulse shape over entire spectrum as a function of time, are optimized. In this paper, the creation of neutron source components for port No.16 viewed to coupled H sub 2 moderator and for port No.11 viewed to decoupled H sub 2 moderator of JSNS are reported.

  17. Dynamics of biopolymers on nanomaterials studied by quasielastic neutron scattering and MD simulations

    Science.gov (United States)

    Dhindsa, Gurpreet K.

    Neutron scattering has been proved to be a powerful tool to study the dynamics of biological systems under various conditions. This thesis intends to utilize neutron scattering techniques, combining with MD simulations, to develop fundamental understanding of several biologically interesting systems. Our systems include a drug delivery system containing Nanodiamonds with nucleic acid (RNA), and two specific model proteins, beta-Casein and Inorganic Pyrophosphatase (IPPase). RNA and nanodiamond (ND) both are suitable for drug-delivery applications in nano-biotechnology. The architecturally flexible RNA with catalytic functionality forms nanocomposites that can treat life-threatening diseases. The non-toxic ND has excellent mechanical and optical properties and functionalizable high surface area, and thus actively considered for biomedical applications. In this thesis, we utilized two tools, quasielastic neutron scattering (QENS) and Molecular Dynamics Simulations to probe the effect of ND on RNA dynamics. Our work provides fundamental understanding of how hydrated RNA motions are affected in the RNA-ND nanocomposites. From the experimental and Molecular Dynamics Simulation (MD), we found that hydrated RNA motion is faster on ND surface than a freestanding one. MD Simulation results showed that the failure of Stokes Einstein relation results the presence of dynamic heterogeneities in the biomacromolecules. Radial pair distribution function from MD Simulation confirmed that the hydrophilic nature of ND attracts more water than RNA results the de-confinement of RNA on ND. Therefore, RNA exhibits faster motion in the presence of ND than freestanding RNA. In the second project, we studied the dynamics of a natively disordered protein beta-Casein which lacks secondary structures. In this study, the temperature and hydration effects on the dynamics of beta-Casein are explored by Quasielastic Neutron Scattering (QENS). We investigated the mean square displacement (MSD) of

  18. Development of a portable micro-environmental cell for the testing of neutron bubble detectors in a simulated jet-aircraft environment

    International Nuclear Information System (INIS)

    Tume, P.; Bennett, L.G.I.; Lewis, B.J.; Wieland, H.K.; Reid, M.K.; Cousins, T.

    1998-01-01

    Neutron-sensitive bubble detectors were chosen as a primary detection tool to survey the dose equivalent received by aircrew exposed to natural radiation. As part of the qualification criterion, a novel micro-environmental cell was designed, assembled and tested. This apparatus is capable of simulating the climate, i.e., pressure, temperature and relative humidity, inside a jet aircraft while irradiating bubble detectors in-situ. The cell environment was manipulated using an on-line control and data acquisition system developed using LabView software. (author)

  19. An Evaluation of Neutron Energy Spectrum Effects in Iron Based on Molecular Dynamics Displacement Cascade Simulations

    International Nuclear Information System (INIS)

    Greenwood, L.R.; Stoller, R.E.

    1998-01-01

    The results of molecular dynamics (MD) displacement cascade simulations in bcc iron have been used to obtain effective cross sections for two measures of primary damage production: (1) the number of surviving point defects expressed as a fraction of the displacements calculated using the standard secondary displacement model of Norgett, Robinson, and Torrens (NRT), and (2) the fraction of the surviving interstitials contained in clusters that formed during the cascade event. Primary knockon atom spectra for iron obtained from the SPECTER code have been used to weight these MD-based damage production cross sections in order to obtain spectrally-averaged values for several locations in commercial fission reactors and materials test reactors. An evaluation of these results indicates that neutron energy spectrum differences between the various enviromnents do not lead to significant differences between the average primary damage formation parameters. In particular, the defect production cross sections obtained for PWR and BWR neutron spectra were not significantly different. The variation of the defect production cross sections as a function of depth into the reactor pressure vessel wall is used as a sample application of the cross sections. A slight difference between the attenuation behavior of the PWR and BWR was noted; this difference could be explained by a subtle difference in the energy dependence of the neutron spectra. Overall, the simulations support the continued use of dpa as a damage correlation parameter

  20. Beam shaping assembly of a D-T neutron source for BNCT and its dosimetry simulation in deeply-seated tumor

    Science.gov (United States)

    Faghihi, F.; Khalili, S.

    2013-08-01

    This article involves two aims for BNCT. First case includes a beam shaping assembly estimation for a D-T neutron source to find epi-thermal neutrons which are the goal in the BNCT. Second issue is the percent depth dose calculation in the adult Snyder head phantom. Monte-Carlo simulations and verification of a suggested beam shaping assembly (including internal neutron multiplier, moderator, filter, external neutron multiplier, collimator, and reflector dimensions) for thermalizing a D-T neutron source as well as increasing neutron flux are carried out and our results are given herein. Finally, we have simulated its corresponding doses for treatment planning of a deeply-seated tumor.

  1. FLUKA simulations of a moderated reduced weight high energy neutron detection system

    Energy Technology Data Exchange (ETDEWEB)

    Biju, K., E-mail: bijusivolli@gmail.com [Health Physics Division, Bhabha Atomic Research Centre, Mumbai 400085 (India); Tripathy, S.P.; Sunil, C.; Sarkar, P.K. [Health Physics Division, Bhabha Atomic Research Centre, Mumbai 400085 (India)

    2012-08-01

    Neutron response of the systems containing high density polyethylene (HDPE) spheres coupled with different external metallic converters has been studied using the FLUKA Monte Carlo simulation code. A diameter of 17.8 cm (7 in.) of the moderating sphere is found to be optimum to obtain the maximum response when used with the neutron converter shells like W, Pb and Zr. Enhancement ratios of the neutron response due to the induced (n, xn) reactions in the outer converters made of W, Pb and Zr are analyzed. It is observed that the enhancement in the response by 1 cm thick Zr shell is comparable to that of 1 cm thick Pb in the energy region of 10-50 MeV. An appreciable enhancement is observed in the case of Zr converter for the higher energy neutrons. Thus, by reducing the dimension of the moderating sphere and using a Zr converter shell, the weight of the system reduces to 10 kg which is less compared to the presently available extended high energy neutron rem meters. The normalized energy dependent ambient dose equivalent response of the zirconium based rem counter (ZReC) at high energies is found to be in good agreement with the energy differential H{sup Low-Asterisk }(10) values suggested by the International Commission on Radiological Protection (ICRP). Based on this study, it is proposed that a rem meter made of 17.8 cm diameter HDPE sphere with 1 cm thick Zr can be used effectively and conveniently for routine monitoring in the accelerator environment.

  2. Simulation of silicon microdosimetry spectra in fast neutron therapy using the GEANT4 Monte Carlo toolkit

    International Nuclear Information System (INIS)

    Cornelius, I.M.; Rosenfeld, A.B.

    2003-01-01

    Microdosimetry is used to predict the biological effects of the densely ionizing radiation environments of hadron therapy and space. The creation of a solid state microdosimeter to replace the conventional Tissue Equivalent Proportional Counter (TEPC) is a topic of ongoing research. The Centre for Medical Radiation Physics has been investigating a technique using microscopic arrays of reverse biased PN junctions. A prototype silicon-on-insulator (SOI) microdosimeter was developed and preliminary measurements have been conducted at several hadron therapy facilities. Several factors impede the application of silicon microdosimeters to hadron therapy. One of the major limitations is that of tissue equivalence, ideally the silicon microdosimeter should provide a microdosimetry distribution identical to that of a microscopic volume of tissue. For microdosimetry in neutron fields, such as Fast Neutron Therapy, it is important that products resulting from neutron interactions in the non tissue equivalent sensitive volume do not contribute significantly to the spectrum. Experimental measurements have been conducted at the Gershenson Radiation Oncology Center, Harper Hospital, Detroit by Bradley et al. The aim was to provide a comparison with measurements performed with a TEPC under identical experimental conditions. Monte Carlo based calculations of these measurements were made using the GEANT4 Monte Carlo toolkit. Agreement between experimental and theoretical results was observed. The model illustrated the importance of neutron interactions in the non tissue equivalent sensitive volume and showed this effect to decrease with sensitive volume size as expected. Simulations were also performed for 1 micron cubic silicon sensitive volumes embedded in tissue equivalent material to predict the best case scenario for silicon microdosimetry in Fast Neutron Therapy

  3. Analysis of MCNP simulated gamma spectra of CdTe detectors for boron neutron capture therapy.

    Science.gov (United States)

    Winkler, Alexander; Koivunoro, Hanna; Savolainen, Sauli

    2017-06-01

    The next step in the boron neutron capture therapy (BNCT) is the real time imaging of the boron concentration in healthy and tumor tissue. Monte Carlo simulations are employed to predict the detector response required to realize single-photon emission computed tomography in BNCT, but have failed to correctly resemble measured data for cadmium telluride detectors. In this study we have tested the gamma production cross-section data tables of commonly used libraries in the Monte Carlo code MCNP in comparison to measurements. The cross section data table TENDL-2008-ACE is reproducing measured data best, whilst the commonly used ENDL92 and other studied libraries do not include correct tables for the gamma production from the cadmium neutron capture reaction that is occurring inside the detector. Furthermore, we have discussed the size of the annihilation peaks of spectra obtained by cadmium telluride and germanium detectors. Copyright © 2017 Elsevier Ltd. All rights reserved.

  4. Neutronic simulation calculations to assess the proliferation resistance of nuclear technologies

    International Nuclear Information System (INIS)

    Englert, Matthias

    2009-01-01

    This thesis investigates the proliferation resistance of nuclear technologies on the basis of three case studies. After a brief description of the concept of proliferation resistance the utilized computer codes and methods are presented. The first case study investigates the potential of monolithic fuel for the conversion of one-fuel-element high-flux research reactors from highly enriched to low enriched uranium using the example of the german research reactor FRM-II. The second case study assesses the proliferation potential of future tokamak based fusion reactors by using neutronic simulations of a possible plutonium production. The third example investigates the proliferation potential of spallation neutron sources to produce nuclear weapon relevant material and the proliferation resistance of such facilities. (orig.)

  5. Visualization of bed material movement in a simulated fluidized bed heat exchanger by neutron radiography

    International Nuclear Information System (INIS)

    Umekawa, Hisashi; Ozawa, Mamoru; Takenaka, Nobuyuki; Matsubayashi, Masahito

    1999-01-01

    The bulk movement of fluidized bed material was visualized by neutron radiography by introducing tracers into the bed materials. The simulated fluidized bed consisted of aluminum plates, and the bed material was sand of 99.7% SiO 2 (mean diameter: 0.218 mm, density: 2555 kg/m 3 ). Both materials were almost transparent to neutrons. Then the sand was colored by the contamination of the sand coated by CdSO 4 . Tracer particles of about 2 mm diameter were made by the B 4 C, bonded by the vinyl resin. The tracer was about ten times as large as the particle of fluidized bed material, but the traceability was enough to observe the bed-material bulk movement owing to the large effective viscosity of the fluidized bed. The visualized images indicated that the bubbles and/or wakes were important mechanism of the behavior of the fluidized bed movement

  6. Simulation of complete neutron scattering experiments: from model systems to liquid germanium; Simulation complete d'une experience de diffusion de neutrons: des systemes modeles au germanium liquide

    Energy Technology Data Exchange (ETDEWEB)

    Hugouvieux, V

    2004-11-15

    In this thesis, both theoretical and experimental studies of liquids are done. Neutron scattering enables structural and dynamical properties of liquids to be investigated. On the theoretical side, molecular dynamics simulations are of great interest since they give positions and velocities of the atoms and the forces acting on each of them. They also enable spatial and temporal correlations to be computed and these quantities are also available from neutron scattering experiments. Consequently, the comparison can be made between results from molecular dynamics simulations and from neutron scattering experiments, in order to improve our understanding of the structure and dynamics of liquids. However, since extracting reliable data from a neutron scattering experiment is difficult, we propose to simulate the experiment as a whole, including both instrument and sample, in order to gain understanding and to evaluate the impact of the different parasitic contributions (absorption, multiple scattering associated with elastic and inelastic scattering, instrument resolution). This approach, in which the sample is described by its structure and dynamics as computed from molecular dynamics simulations, is presented and tested on isotropic model systems. Then liquid germanium is investigated by inelastic neutron scattering and both classical and ab initio molecular dynamics simulations. This enables us to simulate the experiment we performed and to evaluate the influence of the contributions from the instrument and from the sample on the detected signal. (author)

  7. Simulation of e-{gamma}-n targets by FLUKA and measurement of neutron flux at various angles for accelerator based neutron source

    Energy Technology Data Exchange (ETDEWEB)

    Patil, B.J., E-mail: bjp@physics.unipune.ernet.i [Department of Physics, University of Pune, Pune 411 007 (India); Chavan, S.T.; Pethe, S.N.; Krishnan, R. [SAMEER, IIT Powai Campus, Mumbai 400 076 (India); Bhoraskar, V.N. [Department of Physics, University of Pune, Pune 411 007 (India); Dhole, S.D., E-mail: sanjay@physics.unipune.ernet.i [Department of Physics, University of Pune, Pune 411 007 (India)

    2010-10-15

    A 6 MeV Race track Microtron (an electron accelerator) based pulsed neutron source has been designed specifically for the elemental analysis of short lived activation products where the low neutron flux requirement is desirable. The bremsstrahlung radiation emitted by impinging 6 MeV electron on the e-{gamma} primary target, was made to fall on the {gamma}-n secondary target to produce neutrons. The optimisation of bremsstrahlung and neutron producing target along with their spectra were estimated using FLUKA code. The measurement of neutron flux was carried out by activation of vanadium and the measured fluxes were 1.1878 x 10{sup 5}, 0.9403 x 10{sup 5}, 0.7428 x 10{sup 5}, 0.6274 x 10{sup 5}, 0.5659 x 10{sup 5}, 0.5210 x 10{sup 5} n/cm{sup 2}/s at 0{sup o}, 30{sup o}, 60{sup o}, 90{sup o}, 115{sup o}, 140{sup o} respectively. The results indicate that the neutron flux was found to be decreased as increase in the angle and in good agreement with the FLUKA simulation.

  8. Responses of commercially available neutron electronic personal dosemeters in neutron fields simulating workplaces at MOX fuel fabrication facilities

    International Nuclear Information System (INIS)

    Tsujimura, N.; Yoshida, T.; Takada, C.

    2011-01-01

    The authors investigated the performance of three commercially available electronic personal dosemeters (EPDs) in evaluating neutron dose equivalents and discussed their suitability to work environments in MOX fuel fabrication facilities. The EPDs selected for this study were NRY21 (Fuji Electric Systems), PDM-313 (Aloka) and DMC 2000 GN (MGP Instruments). All tests were conducted in moderated 252 Cf neutron fields with neutron spectral and dosimetric characteristics similar to those found in MOX fuel facilities. The test results revealed trends and the magnitude of response variations in relation to neutron spectral changes expected in work environments.

  9. Modeling Taylor series approximations for prompt neutron kinetics with lab view simulations

    International Nuclear Information System (INIS)

    Adzri, E. P.

    2012-09-01

    The reactor point kinetics equations have been subjected to intense research in an effort to find simple yet accurate numerical solutions methods. The equations are very stiff numerically, meaning that there is a wide variation in the decay constants, so that using a particular time step in the numerical solution may provide sufficient accuracy for the group, but not for another. Several solutions techniques have been presented on the point kinetics equations with varying degrees of complexity. These include Power Series Solutions, CORE, PCA, Genapol and Taylor series methods. In this research, algorithms were developed based on the first and second order Taylor series expansion and simulated in LabVIEW to solve the Reactor Point Kinetics equations using block diagram nodes implemented within stacked sequences. The algorithms developed were fast,accurate and simple to code. Several reactivity insertions were used to simulate the change in neutron population with time. The LabVIEW- Taylor series solutions were compared with other solution techniques such as Power Series Solutions, CORE, PCA, Genapol and McMahon and Pierson's Taylor series approximation. The results of LabVIEW-Taylor series technique used by McMahon and Pearson The LabVIEW-implemented techniques were found to agree very well with these other methods. At 1x10 -8 s the neutron population was 1.000220 neutrons / cm 3 , at 1 x 10 -2 s it was 2.007681 neutrons / cm 3 and at 1x10 -1 s it was 2.075317 neutrons / cm 3 ; same results reported by Genapol for a fast reactor, it produced good and accurate results and compared very favorably with other methods found in the literature. Using much smaller time steps to the order or 10 -8 s commensurate with fast reactor parameters also produced very satisfactory results, indicating that the LabVIEW-based Taylor series technique is suitable for simulating the kinetics of fast reactors as well as thermal reactors. Algorithms developed that included second order terms

  10. Development of capability for microtopography-resolving simulations of hydrologic processes in permafrost affected regions

    Science.gov (United States)

    Painter, S.; Moulton, J. D.; Berndt, M.; Coon, E.; Garimella, R.; Lewis, K. C.; Manzini, G.; Mishra, P.; Travis, B. J.; Wilson, C. J.

    2012-12-01

    The frozen soils of the Arctic and subarctic regions contain vast amounts of stored organic carbon. This carbon is vulnerable to release to the atmosphere as temperatures warm and permafrost degrades. Understanding the response of the subsurface and surface hydrologic system to degrading permafrost is key to understanding the rate, timing, and chemical form of potential carbon releases to the atmosphere. Simulating the hydrologic system in degrading permafrost regions is challenging because of the potential for topographic evolution and associated drainage network reorganization as permafrost thaws and massive ground ice melts. The critical process models required for simulating hydrology include subsurface thermal hydrology of freezing/thawing soils, thermal processes within ice wedges, mechanical deformation processes, overland flow, and surface energy balances including snow dynamics. A new simulation tool, the Arctic Terrestrial Simulator (ATS), is being developed to simulate these coupled processes. The computational infrastructure must accommodate fully unstructured grids that track evolving topography, allow accurate solutions on distorted grids, provide robust and efficient solutions on highly parallel computer architectures, and enable flexibility in the strategies for coupling among the various processes. The ATS is based on Amanzi (Moulton et al. 2012), an object-oriented multi-process simulator written in C++ that provides much of the necessary computational infrastructure. Status and plans for the ATS including major hydrologic process models and validation strategies will be presented. Highly parallel simulations of overland flow using high-resolution digital elevation maps of polygonal patterned ground landscapes demonstrate the feasibility of the approach. Simulations coupling three-phase subsurface thermal hydrology with a simple thaw-induced subsidence model illustrate the strong feedbacks among the processes. D. Moulton, M. Berndt, M. Day, J

  11. Numerical simulations on efficiency and measurement of capabilities of BGO detectors for high energy gamma ray

    CERN Document Server

    Wen Wan Xin

    2002-01-01

    The energy resolution and time resolution of two phi 75 x 100 BGO detectors for high energy gamma ray newly made were measured with sup 1 sup 3 sup 7 Cs and sup 6 sup 0 Co resources. The two characteristic gamma rays of high energy emitted from the thermal neutron capture of germanium in BGO crystal were used for the energy calibration of gamma spectra. The intrinsic photopeak efficiency, single escape probability and double escape probabilities of BGO detectors in photon energy range of 4-30 MeV are numerically calculated with GEANT code. The real count response and count ratio of the uniformly distributed incident photons in energy range of 0-30 MeV are also calculated. The distortion of gamma spectra caused by the photon energy loss extension to lower energy in detection medium is discussed

  12. Prototype Development Capabilities of 3D Spatial Interactions and Failures During Scenario Simulation

    Energy Technology Data Exchange (ETDEWEB)

    Steven Prescott; Ramprasad Sampath; Curtis Smith; Tony Koonce

    2014-09-01

    Computers have been used for 3D modeling and simulation, but only recently have computational resources been able to give realistic results in a reasonable time frame for large complex models. This report addressed the methods, techniques, and resources used to develop a prototype for using 3D modeling and simulation engine to improve risk analysis and evaluate reactor structures and components for a given scenario. The simulations done for this evaluation were focused on external events, specifically tsunami floods, for a hypothetical nuclear power facility on a coastline.

  13. Dynamics and denaturation of a protein. Simulations and neutron scattering on staphylococcus nuclease

    International Nuclear Information System (INIS)

    Goupil-Lamy, Anne

    1997-01-01

    This research thesis reports simulations and experiments of inelastic scattering on the whole frequency spectrum to analyse the vibrations of the staphylococcus nuclease and its fragment, in order to study protein folding. Based on these experiments, information on eigenvectors which describe vibration modes can be directly obtained. Inelastic intensities are indeed fully determined by nuclear cross sections and the mean square displacement of each atom. Some experimentally noticed peaks are then explained by calculating a theoretical spectrum from an analysis of normal modes. The studied fragment is made of 136 c-terminal residues. The fragment structure obtained by molecular dynamics simulation is compared with available experimental data. Then, experiments of neutron scattering on the nuclease of staphylococcus and its fragment have been performed. Quasi elastic scattering spectra have been measured. The author then used simulations to try to reproduce the quasi-elastic spectrum. Experiments of inelastic scattering have then been performed [fr

  14. National Research Council Dialogue to Assess Progress on NASA's Advanced Modeling, Simulation and Analysis Capability and Systems Engineering Capability Roadmap Development

    Science.gov (United States)

    Aikins, Jan

    2005-01-01

    Contents include the following: General Background and Introduction of Capability Roadmaps. Agency Objective. Strategic Planning Transformation. Advanced Planning Organizational Roles. Public Involvement in Strategic Planning. Strategic Roadmaps and Schedule. Capability Roadmaps and Schedule. Purpose of NRC Review. Capability Roadmap Development (Progress to Date).

  15. Joint Command and Control (JC2) capability development utilising a Modelling and Simulation Framework

    CSIR Research Space (South Africa)

    Ramadeen, P

    2010-09-01

    Full Text Available : situational picture management; data and sensor fusion; user interaction; tactical simulation; incident management; and system interoperability. Applications developed with the framework can be executed and distributed over multiple hosts through a proprietary...

  16. Improving the fidelity of electrically heated nuclear systems testing using simulated neutronic feedback

    International Nuclear Information System (INIS)

    Bragg-Sitton, Shannon M.; Godfroy, Thomas J.; Webster, Kenny

    2010-01-01

    Nonnuclear test platforms and methodologies can be employed to reduce the overall cost, risk and complexity of testing nuclear systems while allowing one to evaluate the operation of an integrated nuclear system within a reasonable timeframe, providing valuable input to the overall system design. In a nonnuclear test bed, electric heaters are used to simulate the heat from nuclear fuel. Standard electric test techniques allow one to fully assess thermal, heat transfer, and stress related attributes of a given system, but these approaches fail to demonstrate the dynamic response that would be present in an integrated, fueled reactor system. The integration of thermal hydraulic hardware tests with simulated neutronic response provides a bridge between electrically heated testing and testing with nuclear fuel elements installed. By implementing a neutronic response model to simulate the dynamic response that would be expected in a fueled reactor system, one can better understand system integration issues, characterize integrated system response times and response characteristics, and assess potential design improvements at a relatively small fiscal investment. This paper summarizes the results of initial system dynamic response testing for two electrically heated reactor concepts: a heat pipe-cooled reactor simulator with integrated heat exchanger and a gas-cooled reactor simulator with integrated Brayton power conversion system. Initial applications apply a simplified reactor kinetics model with either a single or an averaged measured state point. Preliminary results demonstrate the applicability of the dynamic test methodology to any reactor type, elucidating the variation in system response characteristics in different reactor concepts. These results suggest a need to further enhance the dynamic test approach by incorporating a more accurate model of the reactor dynamics and improved hardware instrumentation for better state estimation in application of the

  17. CFD simulation of the pulsed neutron activation technique for water flow measurements

    International Nuclear Information System (INIS)

    Mattsson, H.; Nordlund, A.

    2005-01-01

    A pulse neutron activation (PNA) flowmeter uses a radioactive substance to measure water flow in pipes. The water in the pipe is bombarded with neutron pulses, thus introducing activity into the pipe. The activity is then transported and mixed with the flow. Gamma radiation emitted from the activity is measured with one or two detectors downstream from the activation point. The average velocity of the water is calculated using the time-resolved signal from the detector. The CFD program FLUENT was used to simulate the transport and mixing of the activity induced in the pipe. The turbulence of the flow is described with the k-ε model. Some parameters affecting a PNA measurement have been investigated. From the calculations it was possible to quantify how much the average initial velocity of the activity differs from the average velocity of the water. Results also show that activity initially produced far away from the wall has a substantial effect on the detector signal. To accurately simulate the detector signal it is necessary to include activity produced in a large part of the pipe. The results also indicate that the collimation of the detectors have a significant impact on the data and should be included when evaluating simulated data. Three different response functions were also tested. (authors)

  18. Conceptual design and neutronics analyses of a fusion reactor blanket simulation facility

    International Nuclear Information System (INIS)

    Beller, D.E.; Ott, K.O.; Terry, W.K.

    1987-01-01

    A new conceptual design of a fusion reactor blanket simulation facility has been developed. This design follows the principles that have been successfully employed in the Purdue Fast Breeder Blanket Facility (FBBF), where experiments have resulted in the discovery of substantial deficiencies in neutronics predictions. With this design, discrepancies between calculation and experimental data can be nearly fully attributed to calculation methods because design deficiencies that could affect results are insignificant. The conceptual design of this FBBF analog, the Fusion Reactor Blanket Facility, is presented

  19. Uncertainty quantification's role in modeling and simulation planning, and credibility assessment through the predictive capability maturity model

    Energy Technology Data Exchange (ETDEWEB)

    Rider, William J. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Witkowski, Walter R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Mousseau, Vincent Andrew [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2016-04-13

    The importance of credible, trustworthy numerical simulations is obvious especially when using the results for making high-consequence decisions. Determining the credibility of such numerical predictions is much more difficult and requires a systematic approach to assessing predictive capability, associated uncertainties and overall confidence in the computational simulation process for the intended use of the model. This process begins with an evaluation of the computational modeling of the identified, important physics of the simulation for its intended use. This is commonly done through a Phenomena Identification Ranking Table (PIRT). Then an assessment of the evidence basis supporting the ability to computationally simulate these physics can be performed using various frameworks such as the Predictive Capability Maturity Model (PCMM). There were several critical activities that follow in the areas of code and solution verification, validation and uncertainty quantification, which will be described in detail in the following sections. Here, we introduce the subject matter for general applications but specifics are given for the failure prediction project. In addition, the first task that must be completed in the verification & validation procedure is to perform a credibility assessment to fully understand the requirements and limitations of the current computational simulation capability for the specific application intended use. The PIRT and PCMM are tools used at Sandia National Laboratories (SNL) to provide a consistent manner to perform such an assessment. Ideally, all stakeholders should be represented and contribute to perform an accurate credibility assessment. PIRTs and PCMMs are both described in brief detail below and the resulting assessments for an example project are given.

  20. SITHA program for simulating hadron transport in the 100-1010eV energy range. Simulation of neutron transport with E6 eV

    International Nuclear Information System (INIS)

    Daniehl', A.V.; Dushin, V.N.

    1987-01-01

    The methods for simulation of neutron transport with Z<20 MeV used in the SITHA (simulation transport hadron) program, the original library of group microconstants (175 groups) with subgroup description of resonance range and a set of programs for its creation are described. The results of a number of integral experiments are discussed

  1. Conceptual design and neutronics analyses of a fusion reactor blanket simulation facility

    International Nuclear Information System (INIS)

    Beller, D.E.

    1986-01-01

    A new conceptual design of a fusion reactor blanket simulation facility was developed. This design follows the principles that have been successfully employed in the Purdue Fast Breeder Blanket Facility (FBBR), because experiments conducted in it have resulted in the discovery of deficiencies in neutronics prediction methods. With this design, discrepancies between calculation and experimental data can be fully attributed to calculation methods because design deficiencies that could affect results are insignificant. Inelastic scattering cross sections are identified as a major source of these discrepancies. The conceptual design of this FBBR analog, the fusion reactor blanket facility (FRBF), is presented. Essential features are a cylindrical geometry and a distributed, cosine-shaped line source of 14-MeV neutrons. This source can be created by sweeping a deuteron beam over an elongated titanium-tritide target. To demonstrate that the design of the FRBF will not contribute significant deviations in experimental results, neutronics analyses were performed: results of comparisons of 2-dimensional to 1-dimensional predictions are reported for two blanket compositions. Expected deviations from 1-D predictions which are due to source anisotropy and blanket asymmetry are minimal. Thus, design of the FRBF allows simple and straightforward interpretation of the experimental results, without a need for coarse 3-D calculations

  2. High energy proton simulation of 14-MeV neutron damage in Al2O3

    International Nuclear Information System (INIS)

    Muir, D.W.; Bunch, J.M.

    1975-01-01

    High-energy protons are a potentially useful tool for simulating the radiation damage produced by 14-MeV neutrons in CTR materials. A comparison is given of calculations and measurements of the relative damage effectiveness of these two types of radiation in single-crystal Al 2 O 3 . The experiments make use of the prominent absorption band at 206 nm as an index to lattice damage, on the assumption that peak absorption is proportional to the concentration of lattice vacancies. The induced absorption is measured for incident proton energies ranging from 5 to 15 MeV and for 14-MeV neutrons. Recoil-energy spectra are calculated for elastic and inelastic scattering using published angular distributions. Recoil-energy spectra also are calculated for the secondary alpha particles and 12 C nuclei produced by (p,p'α) reactions on 16 O. The recoil spectra are converted to damage-energy spectra and then integrated to yield the damage-energy cross section at each proton energy and for 14 MeV neutrons. A comparison of the calculations with experimental results suggests that damage energy, at least at high energies, is a reasonable criterion for estimating this type of radiation damage. (auth)

  3. Simulation of a parallel processor on a serial processor: The neutron diffusion equation

    International Nuclear Information System (INIS)

    Honeck, H.C.

    1981-01-01

    Parallel processors could provide the nuclear industry with very high computing power at a very moderate cost. Will we be able to make effective use of this power. This paper explores the use of a very simple parallel processor for solving the neutron diffusion equation to predict power distributions in a nuclear reactor. We first describe a simple parallel processor and estimate its theoretical performance based on the current hardware technology. Next, we show how the parallel processor could be used to solve the neutron diffusion equation. We then present the results of some simulations of a parallel processor run on a serial processor and measure some of the expected inefficiencies. Finally we extrapolate the results to estimate how actual design codes would perform. We find that the standard numerical methods for solving the neutron diffusion equation are still applicable when used on a parallel processor. However, some simple modifications to these methods will be necessary if we are to achieve the full power of these new computers. (orig.) [de

  4. An investigation of the neutron flux in bone-fluorine phantoms comparing accelerator based in vivo neutron activation analysis and FLUKA simulation data

    International Nuclear Information System (INIS)

    Mostafaei, F.; McNeill, F.E.; Chettle, D.R.; Matysiak, W.; Bhatia, C.; Prestwich, W.V.

    2015-01-01

    We have tested the Monte Carlo code FLUKA for its ability to assist in the development of a better system for the in vivo measurement of fluorine. We used it to create a neutron flux map of the inside of the in vivo neutron activation analysis irradiation cavity at the McMaster Accelerator Laboratory. The cavity is used in a system that has been developed for assessment of fluorine levels in the human hand. This study was undertaken to (i) assess the FLUKA code, (ii) find the optimal hand position inside the cavity and assess the effects on precision of a hand being in a non-optimal position and (iii) to determine the best location for our γ-ray detection system within the accelerator beam hall. Simulation estimates were performed using FLUKA. Experimental measurements of the neutron flux were performed using Mn wires. The activation of the wires was measured inside (1) an empty bottle, (2) a bottle containing water, (3) a bottle covered with cadmium and (4) a dry powder-based fluorine phantom. FLUKA was used to simulate the irradiation cavity, and used to estimate the neutron flux in different positions both inside, and external to, the cavity. The experimental results were found to be consistent with the Monte Carlo simulated neutron flux. Both experiment and simulation showed that there is an optimal position in the cavity, but that the effect on the thermal flux of a hand being in a non-optimal position is less than 20%, which will result in a less than 10% effect on the measurement precision. FLUKA appears to be a code that can be useful for modeling of this type of experimental system

  5. Improvement of the matrix effect compensation in active neutron measurement by simulated annealing algorithm (June 2009)

    International Nuclear Information System (INIS)

    Raoux, A. C.; Loridon, J.; Mariani, A.; Passard, C.

    2009-01-01

    Active neutron measurements such as the Differential Die-Away (DDA) technique involving pulsed neutron generator, are widely applied to determine the fissile content of waste packages. Unfortunately, the main drawback of such techniques is coming from the lack of knowledge of the waste matrix composition. Thus, the matrix effect correction for the DDA measurement is an essential improvement in the field of fissile material content determination. Different solutions have been developed to compensate the effect of the matrix on the neutron measurement interpretation. In this context, this paper describes an innovative matrix correction method we have developed with the goal of increasing the accuracy of the matrix effect correction and reducing the measurement time. The implementation of this method is based on the analysis of the raw signal with an optimisation algorithm called the simulated annealing algorithm. This algorithm needs a reference data base of Multi-Channel Scaling (MCS) spectra, to fit the raw signal. The construction of the MCS library involves a learning phase to define and acquire the DDA signals. This database has been provided by a set of active signals from experimental matrices (mock-up waste drums of 118 litres) recorded in a specific device dedicated to neutron measurement research and development of the Nuclear Measurement Laboratory of CEA-Cadarache, called PROMETHEE 6. The simulated annealing algorithm is applied to make use of the effect of the matrices on the total active signal of DDA measurement. Furthermore, as this algorithm is directly applied to the raw active signal, it is very useful when active background contributions can not be easily estimated and removed. Most of the cases tested during this work which represents the feasibility phase of the method, are within a 4% agreement interval with the expected experimental value. Moreover, one can notice that without any compensation of the matrix effect, the classical DDA prompt

  6. Improvement of the matrix effect compensation in active neutron measurement by simulated annealing algorithm (June 2009)

    Energy Technology Data Exchange (ETDEWEB)

    Raoux, A. C.; Loridon, J.; Mariani, A.; Passard, C. [French Atomic Energy Commission, DEN, Cadarache, F-3108 Saint-Paul-Lez-Durance (France)

    2009-07-01

    Active neutron measurements such as the Differential Die-Away (DDA) technique involving pulsed neutron generator, are widely applied to determine the fissile content of waste packages. Unfortunately, the main drawback of such techniques is coming from the lack of knowledge of the waste matrix composition. Thus, the matrix effect correction for the DDA measurement is an essential improvement in the field of fissile material content determination. Different solutions have been developed to compensate the effect of the matrix on the neutron measurement interpretation. In this context, this paper describes an innovative matrix correction method we have developed with the goal of increasing the accuracy of the matrix effect correction and reducing the measurement time. The implementation of this method is based on the analysis of the raw signal with an optimisation algorithm called the simulated annealing algorithm. This algorithm needs a reference data base of Multi-Channel Scaling (MCS) spectra, to fit the raw signal. The construction of the MCS library involves a learning phase to define and acquire the DDA signals. This database has been provided by a set of active signals from experimental matrices (mock-up waste drums of 118 litres) recorded in a specific device dedicated to neutron measurement research and development of the Nuclear Measurement Laboratory of CEA-Cadarache, called PROMETHEE 6. The simulated annealing algorithm is applied to make use of the effect of the matrices on the total active signal of DDA measurement. Furthermore, as this algorithm is directly applied to the raw active signal, it is very useful when active background contributions can not be easily estimated and removed. Most of the cases tested during this work which represents the feasibility phase of the method, are within a 4% agreement interval with the expected experimental value. Moreover, one can notice that without any compensation of the matrix effect, the classical DDA prompt

  7. Monte Carlo simulation of explosive detection system based on a Deuterium-Deuterium (D-D) neutron generator.

    Science.gov (United States)

    Bergaoui, K; Reguigui, N; Gary, C K; Brown, C; Cremer, J T; Vainionpaa, J H; Piestrup, M A

    2014-12-01

    An explosive detection system based on a Deuterium-Deuterium (D-D) neutron generator has been simulated using the Monte Carlo N-Particle Transport Code (MCNP5). 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 emission (10.82MeV) following radiative neutron capture by (14)N nuclei. The explosive detection system was built based on a fully high-voltage-shielded, axial D-D neutron generator with a radio frequency (RF) driven ion source and nominal yield of about 10(10) fast neutrons per second (E=2.5MeV). Polyethylene and paraffin were used as moderators with borated polyethylene and lead as neutron and gamma ray shielding, respectively. The shape and the thickness of the moderators and shields are optimized to produce the highest thermal neutron flux at the position of the explosive and the minimum total dose at the outer surfaces of the explosive detection system walls. In addition, simulation of the response functions of NaI, BGO, and LaBr3-based γ-ray detectors to different explosives is described. Copyright © 2014 Elsevier Ltd. All rights reserved.

  8. Concatenating algorithms for parallel numerical simulations coupling radiation hydrodynamics with neutron transport

    International Nuclear Information System (INIS)

    Mo Zeyao

    2004-11-01

    Multiphysics parallel numerical simulations are usually essential to simplify researches on complex physical phenomena in which several physics are tightly coupled. It is very important on how to concatenate those coupled physics for fully scalable parallel simulation. Meanwhile, three objectives should be balanced, the first is efficient data transfer among simulations, the second and the third are efficient parallel executions and simultaneously developments of those simulation codes. Two concatenating algorithms for multiphysics parallel numerical simulations coupling radiation hydrodynamics with neutron transport on unstructured grid are presented. The first algorithm, Fully Loosely Concatenation (FLC), focuses on the independence of code development and the independence running with optimal performance of code. The second algorithm. Two Level Tightly Concatenation (TLTC), focuses on the optimal tradeoffs among above three objectives. Theoretical analyses for communicational complexity and parallel numerical experiments on hundreds of processors on two parallel machines have showed that these two algorithms are efficient and can be generalized to other multiphysics parallel numerical simulations. In especial, algorithm TLTC is linearly scalable and has achieved the optimal parallel performance. (authors)

  9. Ab initio simulations and neutron scattering studies of structure and dynamics in PdH

    International Nuclear Information System (INIS)

    Totolici, I.E.

    2001-07-01

    The work presented in this PhD thesis is concerned with the interpretation of the neutron scattering measurements from the palladium hydrogen system by means of ab initio electronic structure calculations. The motivation of performing such calculations was due to recent neutron scattering studies on this system that showed a strong directional dependence to the dynamical structure factor together with a complex dependence on energy. Here we attempt to describe the origin of these features by ab initio simulations of the dynamical structure factor. The method assumes an adiabatic separation of the motion of the proton and palladium atoms. The proton wave functions are calculated by a direct solution of the associated single-particle Schroedinger equation using a plane wave basis set method and a mapping of the adiabatic surface. The Fourier components of the adiabatic potential are obtained from LDA pseudopotential calculations. Using Fermi's golden rule within the Born approximation we were then able to calculate the dynamical structure factor, S(Q,ω), for exciting the proton from its ground state to various excited states as a function of the magnitude and direction of the scattering wave vector. The results are in agreement with the inelastic neutron scattering spectra and allow us to identify the origin of previous inexplicable features, in particular the strong directional dependence to the experimental data. The method was extended to investigate the expansion of the equilibrium lattice constant as a function of the H isotope when the zero-point energy of the proton/deuterium is explicitly taken into account in the relaxation process. The results we obtained predicted a bigger lattice constant for the hydride, as expected. Furthermore, other complex ab initio calculations were carried out in order to describe the origin of the large optic dispersion, seen previously in the coherent neutron scattering data. Our calculated dispersion proved to be in good

  10. Monte Carlo simulation of the neutron-induced prompt γ ray spectroscopy of the CW abandoned by Japan

    International Nuclear Information System (INIS)

    Wang Bairong; Yang Zhongping; Zhang Wenzhong

    2005-01-01

    This paper introduced the principle of identifying the chemical weapon by neutron-induced prompt γ ray, simulated and analyzed the neutron-induced prompt γ ray spectroscopy of chemical weapon abandoned by Japan in the different condition, using the MCNP-4C Monte Carlo program, whereby supply important datum and reference for the aftertime deeper research and disposal of Japan-abandoned chemical weapon. (authors)

  11. The Monte Carlo simulation of the neutron-induced prompt gamma ray spectroscopy of the CW abandoned by Japan

    International Nuclear Information System (INIS)

    Wang Bairong; Yang Zhongping; Zhan Wenzhong

    2003-01-01

    This paper introduced the principle of identifying the chemical weapon abandoned by Japan by neutron-induced prompt gamma ray. Using the MCNP-4C Monte Carlo program, this paper simulated and analyzed the neutron-induced prompt gamma ray spectroscopy of chemical weapon abandoned by Japan, whereby supply important datum and reference for the aftertime deeper research and disposal of Japan-abandoned chemical weapon. (authors)

  12. The diffusional pulsed cooling of the thermal neutron flux in small two-region systems. Monte Carlo Simulation

    International Nuclear Information System (INIS)

    Wiacek, U.

    2006-06-01

    The thermal neutron transport in small unhomogeneous system and namely in two- layers where the first one -outer moderator is of hydride type (polyethylene or plexiglas) and the second one - inner is made with other materials is investigated. The diffusional cooling of neutrons has been calculated by means of monte Carlo simulations using MCPN code. Because of un consistency of calculated and measured data the MCPN code library has been modified for polyethylene and plexiglas

  13. Advanced multi-physics simulation capability for very high temperature reactors

    International Nuclear Information System (INIS)

    Lee, Hyun Chul; Tak, Nam Il; Jo Chang Keun; Noh, Jae Man; Cho, Bong Hyun; Cho, Jin Woung; Hong, Ser Gi

    2012-01-01

    The purpose of this research is to develop methodologies and computer code for high-fidelity multi-physics analysis of very high temperature gas-cooled reactors(VHTRs). The research project was performed through Korea-US I-NERI program. The main research topic was development of methodologies for high-fidelity 3-D whole core transport calculation, development of DeCART code for VHTR reactor physics analysis, generation of VHTR specific 190-group cross-section library for DeCART code, development of DeCART/CORONA coupled code system for neutronics/thermo-fluid multi-physics analysis, and benchmark analysis against various benchmark problems derived from PMR200 reactor. The methodologies and the code systems will be utilized a key technologies in the Nuclear Hydrogen Development and Demonstration program. Export of code system is expected in the near future and the code systems developed in this project are expected to contribute to development and export of nuclear hydrogen production system

  14. Thick-foils activation technique for neutron spectrum unfolding with the MINUIT routine-Comparison with GEANT4 simulations

    Science.gov (United States)

    Vagena, E.; Theodorou, K.; Stoulos, S.

    2018-04-01

    Neutron activation technique has been applied using a proposed set of twelve thick metal foils (Au, As, Cd, In, Ir, Er, Mn, Ni, Se, Sm, W, Zn) for off-site measurements to obtain the neutron spectrum over a wide energy range (from thermal up to a few MeV) in intense neutron-gamma mixed fields such as around medical Linacs. The unfolding procedure takes into account the activation rates measured using thirteen (n , γ) and two (n , p) reactions without imposing a guess solution-spectrum. The MINUIT minimization routine unfolds a neutron spectrum that is dominated by fast neutrons (70%) peaking at 0.3 MeV, while the thermal peak corresponds to the 15% of the total neutron fluence equal to the epithermal-resonances area. The comparison of the unfolded neutron spectrum against the simulated one with the GEANT4 Monte-Carlo code shows a reasonable agreement within the measurement uncertainties. Therefore, the proposed set of activation thick-foils could be a useful tool in order to determine low flux neutrons spectrum in intense mixed field.

  15. Health Worker Focused Distributed Simulation for Improving Capability of Health Systems in Liberia.

    Science.gov (United States)

    Gale, Thomas C E; Chatterjee, Arunangsu; Mellor, Nicholas E; Allan, Richard J

    2016-04-01

    The main goal of this study was to produce an adaptable learning platform using virtual learning and distributed simulation, which can be used to train health care workers, across a wide geographical area, key safety messages regarding infection prevention control (IPC). A situationally responsive agile methodology, Scrum, was used to develop a distributed simulation module using short 1-week iterations and continuous synchronous plus asynchronous communication including end users and IPC experts. The module contained content related to standard IPC precautions (including handwashing techniques) and was structured into 3 distinct sections related to donning, doffing, and hazard perception training. Using Scrum methodology, we were able to link concepts applied to best practices in simulation-based medical education (deliberate practice, continuous feedback, self-assessment, and exposure to uncommon events), pedagogic principles related to adult learning (clear goals, contextual awareness, motivational features), and key learning outcomes regarding IPC, as a rapid response initiative to the Ebola outbreak in West Africa. Gamification approach has been used to map learning mechanics to enhance user engagement. The developed IPC module demonstrates how high-frequency, low-fidelity simulations can be rapidly designed using scrum-based agile methodology. Analytics incorporated into the tool can help demonstrate improved confidence and competence of health care workers who are treating patients within an Ebola virus disease outbreak region. These concepts could be used in a range of evolving disasters where rapid development and communication of key learning messages are required.

  16. Evolution of magnetized, differentially rotating neutron stars: Simulations in full general relativity

    International Nuclear Information System (INIS)

    Duez, Matthew D.; Liu, Yuk Tung; Shapiro, Stuart L.; Stephens, Branson C.; Shibata, Masaru

    2006-01-01

    We study the effects of magnetic fields on the evolution of differentially rotating neutron stars, which can be formed in stellar core collapse or binary neutron star coalescence. Magnetic braking and the magnetorotational instability (MRI) both act on differentially rotating stars to redistribute angular momentum. Simulations of these stars are carried out in axisymmetry using our recently developed codes which integrate the coupled Einstein-Maxwell-MHD equations. We consider stars with two different equations of state (EOS), a gamma-law EOS with Γ=2, and a more realistic hybrid EOS, and we evolve them adiabatically. Our simulations show that the fate of the star depends on its mass and spin. For initial data, we consider three categories of differentially rotating, equilibrium configurations, which we label normal, hypermassive and ultraspinning. Normal configurations have rest masses below the maximum achievable with uniform rotation, and angular momentum below the maximum for uniform rotation at the same rest mass. Hypermassive stars have rest masses exceeding the mass limit for uniform rotation. Ultraspinning stars are not hypermassive, but have angular momentum exceeding the maximum for uniform rotation at the same rest mass. We show that a normal star will evolve to a uniformly rotating equilibrium configuration. An ultraspinning star evolves to an equilibrium state consisting of a nearly uniformly rotating central core, surrounded by a differentially rotating torus with constant angular velocity along magnetic field lines, so that differential rotation ceases to wind the magnetic field. In addition, the final state is stable against the MRI, although it has differential rotation. For a hypermassive neutron star, the MHD-driven angular momentum transport leads to catastrophic collapse of the core. The resulting rotating black hole is surrounded by a hot, massive, magnetized torus undergoing quasistationary accretion, and a magnetic field collimated along

  17. Measurements and calculations of neutron fluxes through a simulation of the CRBR upper axial shielding

    International Nuclear Information System (INIS)

    Maerker, R.E.; Muckenthaler, F.J.

    1976-01-01

    Measurements, using a 4-in. Bonner Ball, have been made of the neutron fluxes penetrating a simulation of CRBR upper axial biological shielding at the Tower Shielding Facility. The simulation consisted of a 45.7 cm thick slab of SS-304 followed by a series of sodium tanks having a total thickness of 457 cm followed by slabs of carbon steel up to 61.0 cm thick. Measurements were made behind the stainless steel, behind intermediate thicknesses of 152 cm, 305 cm, and 457 cm of sodium (with the stainless steel in place), and behind various thicknesses of the carbon steel following both 305 cm and 457 cm of sodium (also with the stainless steel in place). Calculated and measured data are presented and compared

  18. Simulation of space protons influence on silicon semiconductor devices using gamma-neutron irradiation

    International Nuclear Information System (INIS)

    Zhukov, Y.N.; Zinchenko, V.F.; Ulimov, V.N.

    1999-01-01

    In this study the authors focus on the problems of simulating the space proton energy spectra under laboratory gamma-neutron radiation tests of semiconductor devices (SD). A correct simulation of radiation effects implies to take into account and evaluate substantial differences in the processes of formation of primary defects in SD in space environment and under laboratory testing. These differences concern: 1) displacement defects, 2) ionization defects and 3) intensity of radiation. The study shows that: - the energy dependence of nonionizing energy loss (NIEL) is quite universal to predict the degradation of SD parameters associated to displacement defects, and - MOS devices that are sensitive to ionization defects indicated the same variation of parameters under conditions of equality of ionization density generated by protons and gamma radiations. (A.C.)

  19. Simulating neutron star mergers as r-process sources in ultrafaint dwarf galaxies

    Science.gov (United States)

    Safarzadeh, Mohammadtaher; Scannapieco, Evan

    2017-10-01

    To explain the high observed abundances of r-process elements in local ultrafaint dwarf (UFD) galaxies, we perform cosmological zoom simulations that include r-process production from neutron star mergers (NSMs). We model star formation stochastically and simulate two different haloes with total masses ≈108 M⊙ at z = 6. We find that the final distribution of [Eu/H] versus [Fe/H] is relatively insensitive to the energy by which the r-process material is ejected into the interstellar medium, but strongly sensitive to the environment in which the NSM event occurs. In one halo, the NSM event takes place at the centre of the stellar distribution, leading to high levels of r-process enrichment such as seen in a local UFD, Reticulum II (Ret II). In a second halo, the NSM event takes place outside of the densest part of the galaxy, leading to a more extended r-process distribution. The subsequent star formation occurs in an interstellar medium with shallow levels of r-process enrichment that results in stars with low levels of [Eu/H] compared to Ret II stars even when the maximum possible r-process mass is assumed to be ejected. This suggests that the natal kicks of neutron stars may also play an important role in determining the r-process abundances in UFD galaxies, a topic that warrants further theoretical investigation.

  20. General Relativistic Simulations of Low-Mass Magnetized Binary Neutron Star Mergers

    Science.gov (United States)

    Giacomazzo, Bruno

    2017-01-01

    We will present general relativistic magnetohydrodynamic (GRMHD) simulations of binary neutron star (BNS) systems that produce long-lived neutron stars (NSs) after merger. While the standard scenario for short gamma-ray bursts (SGRBs) requires the formation after merger of a spinning black hole surrounded by an accretion disk, other theoretical models, such as the time-reversal scenario, predict the formation of a long-lived magnetar. The formation of a long-lived magnetar could in particular explain the X-ray plateaus that have been observed in some SGRBs. Moreover, observations of NSs with masses of 2 solar masses indicate that the equation of state of NS matter should support masses larger than that. Therefore a significant fraction of BNS mergers will produce long-lived NSs. This has important consequences both on the emission of gravitational wave signals and on their electromagnetic counterparts. We will discuss GRMHD simulations of ``low-mass'' magnetized BNS systems with different equations of state and mass ratios. We will describe the properties of their post-merger remnants and of their gravitational and electromagnetic emission.

  1. Assessing the capability of CORDEX models in simulating onset of rainfall in West Africa

    Science.gov (United States)

    Mounkaila, Moussa S.; Abiodun, Babatunde J.; `Bayo Omotosho, J.

    2015-01-01

    Reliable forecasts of rainfall-onset dates (RODs) are crucial for agricultural planning and food security in West Africa. This study evaluates the ability of nine CORDEX regional climate models (RCMs: ARPEGE, CRCM5, RACMO, RCA35, REMO, RegCM3, PRECIS, CCLM and WRF) in simulating RODs over the region. Four definitions are used to compute RODs, and two observation datasets (GPCP and TRMM) are used in the model evaluation. The evaluation considers how well the RCMs, driven by ERA-Interim reanalysis (ERAIN), simulate the observed mean, standard deviation and inter-annual variability of RODs over West Africa. It also investigates how well the models link RODs with the northward movement of the monsoon system over the region. The model performances are compared to that of the driving reanalysis—ERAIN. Observations show that the mean RODs in West Africa have a zonal distribution, and the dates increase from the Guinea coast northward. ERAIN fails to reproduce the spatial distribution of the RODs as observed. The performance of some RCMs in simulating the RODs depends on the ROD definition used. For instance, ARPEGE, RACMO, PRECIS and CCLM produce a better ROD distribution than that of ERAIN when three of the ROD definitions are used, but give a worse ROD distribution than that of ERAIN when the fourth definition is used. However, regardless of the definition used, CCRM5, RCA35, REMO, RegCM3 and WRF show a remarkable improvement over ERAIN. The study shows that the ability of the RCMs in simulating RODs over West Africa strongly depends on how well the models reproduce the northward movement of the monsoon system and the associated features. The results show that there are some differences in the RODs obtained between the two observation datasets and RCMs, and the differences are magnified by differences in the ROD definitions. However, the study shows that most CORDEX RCMs have remarkable skills in predicting the RODs in West Africa.

  2. Monte Carlo simulation of fission yields, kinetic energy, fission neutron spectrum and decay γ-ray spectrum for 232Th(n,f) reaction induced by 3H(d,n) 4He neutron source

    International Nuclear Information System (INIS)

    Zheng Wei; Zeen Yao; Changlin Lan; Yan Yan; Yunjian Shi; Siqi Yan; Jie Wang; Junrun Wang; Jingen Chen; Chinese Academy of Sciences, Shanghai

    2015-01-01

    Monte Carlo transport code Geant4 has been successfully utilised to study of neutron-induced fission reaction for 232 Th in the transport neutrons generated from 3 H(d,n) 4 He neutron source. The purpose of this work is to examine the applicability of Monte Carlo simulations for the computation of fission reaction process. For this, Monte Carlo simulates and calculates the characteristics of fission reaction process of 232 Th(n,f), such as the fission yields distribution, kinetic energy distribution, fission neutron spectrum and decay γ-ray spectrum. This is the first time to simulate the process of neutron-induced fission reaction using Geant4 code. Typical computational results of neutron-induced fission reaction of 232 Th(n,f) reaction are presented. The computational results are compared with the previous experimental data and evaluated nuclear data to confirm the certain physical process model in Geant4 of scientific rationality. (author)

  3. Monte Carlo simulation of the scattered component of neutron capture prompt gamma-ray analyzer responses

    International Nuclear Information System (INIS)

    Jin, Y.; Verghese, K.; Gardner, R.P.

    1986-01-01

    This paper describes a major part of our efforts to simulate the entire spectral response of the neutron capture prompt gamma-ray analyzer for bulk media (or conveyor belt) samples by the Monte Carlo method. This would allow one to use such a model to augment or, in most cases, essentially replace experiments in the calibration and optimum design of these analyzers. In previous work, we simulated the unscattered gamma-ray intensities, but would like to simulate the entire spectral response as we did with the energy-dispersive x-ray fluorescence analyzers. To accomplish this, one must account for the scattered gamma rays as well as the unscattered and one must have available the detector response function to translate the incident gamma-ray spectrum calculated by the Monte Carlo simulation into the detected pulse-height spectrum. We recently completed our work on the germanium detector response function, and the present paper describes our efforts to simulate the entire spectral response by using it with Monte Carlo predicted unscattered and scattered gamma rays

  4. Quasielastic neutron scattering and molecular dynamics simulation studies of the melting transition in butane and hexane monolayers adsorbed on graphite

    DEFF Research Database (Denmark)

    Hervig, K.W.; Wu, Z.; Dai, P.

    1997-01-01

    Quasielastic neutron scattering experiments and molecular dynamics (MD) simulations have been used to investigate molecular diffusive motion near the melting transition of monolayers of flexible rod-shaped molecules. The experiments were conducted on butane and hexane monolayers adsorbed...... comparison with experiment, quasielastic spectra calculated from the MD simulations were analyzed using the same models and fitting algorithms as for the neutron spectra. This combination of techniques gives a microscopic picture of the melting process in these two monolayers which is consistent with earlier...... neutron diffraction experiments. Butane melts abruptly to a liquid phase where the molecules in the trans conformation translationally diffuse while rotating about their center of mass. In the case of the hexane monolayer, the MD simulations show that the appearance of quasielastic scattering below T...

  5. Dynamics of water and ions in clays of type montmorillonite by microscopic simulation and quasi-elastic neutron scattering

    International Nuclear Information System (INIS)

    Malikova, N.

    2005-09-01

    Montmorillonite clays in low hydration states, with Na + and Cs + compensating counter ions, are investigated by a combination of microscopic simulation and quasi-elastic neutron scattering to obtain information on the local structure and dynamics of water and ions in the interlayer. At first predictions of simulation into the dynamics of water and ions at elevate temperatures are shown (0 deg C 80 deg C, pertinent for the radioactive waste disposal scenario) Marked difference is observed between the modes of diffusion of the Na + and C + counter ions. In water dynamics, a significant step towards bulk water behaviour is seen on transition from the mono- to bilayer states. Secondly, a detailed comparison between simulation and quasi-elastic neutron scattering (Neutron Spin Echo and Time-of-Flight) regarding ambient temperature water dynamics is presented. Overall, the approaches are found to be in good agreement with each other and limitations of each of the methods are clearly shown. (author)

  6. The effects of nuclear data library processing on Geant4 and MCNP simulations of the thermal neutron scattering law

    Science.gov (United States)

    Hartling, K.; Ciungu, B.; Li, G.; Bentoumi, G.; Sur, B.

    2018-05-01

    Monte Carlo codes such as MCNP and Geant4 rely on a combination of physics models and evaluated nuclear data files (ENDF) to simulate the transport of neutrons through various materials and geometries. The grid representation used to represent the final-state scattering energies and angles associated with neutron scattering interactions can significantly affect the predictions of these codes. In particular, the default thermal scattering libraries used by MCNP6.1 and Geant4.10.3 do not accurately reproduce the ENDF/B-VII.1 model in simulations of the double-differential cross section for thermal neutrons interacting with hydrogen nuclei in a thin layer of water. However, agreement between model and simulation can be achieved within the statistical error by re-processing ENDF/B-VII.I thermal scattering libraries with the NJOY code. The structure of the thermal scattering libraries and sampling algorithms in MCNP and Geant4 are also reviewed.

  7. Decay of the pulsed thermal neutron flux in two-zone hydrogenous systems - Monte Carlo simulations using MCNP standard data libraries

    International Nuclear Information System (INIS)

    Wiacek, Urszula; Krynicka, Ewa

    2006-01-01

    Pulsed neutron experiments in two-zone spherical and cylindrical geometry has been simulated using the MCNP code. The systems are built of hydrogenous materials. The inner zone is filled with aqueous solutions of absorbers (H 3 BO 3 or KCl). It is surrounded by the outer zone built of Plexiglas. The system is irradiated with the pulsed thermal neutron flux and the thermal neutron decay in time is observed. Standard data libraries of the thermal neutron scattering cross-sections of hydrogen in hydrogenous substances have been used to simulate the neutron transport. The time decay constant of the fundamental mode of the thermal neutron flux determined in each simulation has been compared with the corresponding result of the real pulsed neutron experiment

  8. Production and use of Li(d,n) neutrons for simulation of radiation effects in fusion reactors

    International Nuclear Information System (INIS)

    Goland, A.N.; Gurinsky, D.H.; Hendrie, J.; Kukkonen, J.; Sheehan, T.; Snead, C.L. Jr.

    1975-01-01

    In the Brookhaven Accelerator-Based Neutron Generator 1.5-cm thick x 12-cm wide films of lithium flowing at the velocity of approximately 10 m sec -1 will be the targets for 30-MeV D + and D - beams 1-cm high and 10-cm wide. At this energy a beam of energetic neutrons is emitted mainly in the forward direction (theta less than or equal to 20 0 ) as a result of the Li(d,n) breakup reaction. Measurements of the neutron flux and spectrum as a function of incident deuteron energy and emission angle theta(theta less than or equal to 20 0 ) indicate that the yield increases approximately linearly with increasing deuteron energy from 25 MeV to at least 35 MeV, and that the mean energy of the neutrons (theta = 0 0 ) is about 0.4 of the incident deuteron energies between 25 and 35 MeV. The most probable neutron energy in the forward-directed (theta = 0 0 ) spectrum is also about 0.4 of the deuteron energy over this range. For a 30-MeV beam, the full width at half maximum of the neutron spectrum is 11.8 MeV (theta = 0 0 ), and the mean neutron energy is 13 MeV. Pertinent radiation-damage parameters were calculated for various materials exposed to this neutron spectrum. In Nb, for example, the helium production rate and the displacement rate simulate the values anticipated in a D-T fusion reactor spectrum of comparable flux. Furthermore, the primary-recoil-atom energy distributions produced by Li(d,n) neutrons in Al, Nb, and Au are similar to those produced by 14-MeV neutrons. (U.S.)

  9. 3CE Methodology for Conducting a Modeling, Simulation, and Instrumentation Tool Capability Analysis

    Science.gov (United States)

    2010-05-01

    flRmurn I F )T:Ir,tir)l! MCr)lto.-lng DHin nttbli..’"Ollc:~ E,;m:a..liut .!,)’l’lt’Mn:l’lll.ll~ t Managemen t F unction a l Arem 1 .5 Toola na...a modeling, simulation, and instrumentation (MS&I) environment. This methodology uses the DoDAF product set to document operational and systems...engineering process were identified and resolved, such as duplication of data elements derived from DoDAF operational and system views used to

  10. Expand the Modeling Capabilities of DOE's EnergyPlus Building Energy Simulation Program

    Energy Technology Data Exchange (ETDEWEB)

    Don Shirey

    2008-02-28

    EnergyPlus{trademark} is a new generation computer software analysis tool that has been developed, tested, and commercialized to support DOE's Building Technologies (BT) Program in terms of whole-building, component, and systems R&D (http://www.energyplus.gov). It is also being used to support evaluation and decision making of zero energy building (ZEB) energy efficiency and supply technologies during new building design and existing building retrofits. Version 1.0 of EnergyPlus was released in April 2001, followed by semiannual updated versions over the ensuing seven-year period. This report summarizes work performed by the University of Central Florida's Florida Solar Energy Center (UCF/FSEC) to expand the modeling capabilities of EnergyPlus. The project tasks involved implementing, testing, and documenting the following new features or enhancement of existing features: (1) A model for packaged terminal heat pumps; (2) A model for gas engine-driven heat pumps with waste heat recovery; (3) Proper modeling of window screens; (4) Integrating and streamlining EnergyPlus air flow modeling capabilities; (5) Comfort-based controls for cooling and heating systems; and (6) An improved model for microturbine power generation with heat recovery. UCF/FSEC located existing mathematical models or generated new model for these features and incorporated them into EnergyPlus. The existing or new models were (re)written using Fortran 90/95 programming language and were integrated within EnergyPlus in accordance with the EnergyPlus Programming Standard and Module Developer's Guide. Each model/feature was thoroughly tested and identified errors were repaired. Upon completion of each model implementation, the existing EnergyPlus documentation (e.g., Input Output Reference and Engineering Document) was updated with information describing the new or enhanced feature. Reference data sets were generated for several of the features to aid program users in selecting proper

  11. A Simplified Ab Initio Cosmic-ray Modulation Model with Simulated Time Dependence and Predictive Capability

    Science.gov (United States)

    Moloto, K. D.; Engelbrecht, N. E.; Burger, R. A.

    2018-06-01

    A simplified ab initio approach is followed to model cosmic-ray proton modulation, using a steady-state three-dimensional stochastic solver of the Parker transport equation that simulates some effects of time dependence. Standard diffusion coefficients based on Quasilinear Theory and Nonlinear Guiding Center Theory are employed. The spatial and temporal dependences of the various turbulence quantities required as inputs for the diffusion, as well as the turbulence-reduced drift coefficients, follow from parametric fits to results from a turbulence transport model as well as from spacecraft observations of these turbulence quantities. Effective values are used for the solar wind speed, magnetic field magnitude, and tilt angle in the modulation model to simulate temporal effects due to changes in the large-scale heliospheric plasma. The unusually high cosmic-ray intensities observed during the 2009 solar minimum follow naturally from the current model for most of the energies considered. This demonstrates that changes in turbulence contribute significantly to the high intensities during that solar minimum. We also discuss and illustrate how this model can be used to predict future cosmic-ray intensities, and comment on the reliability of such predictions.

  12. A contribution to the development of the modular neutron detector (DEMON): performance evaluation through measurements and simulations; Contribution a la realisation du detecteur modulaire de neutrons (DEMON): etudes des performances par mesures et simulations

    Energy Technology Data Exchange (ETDEWEB)

    Mouatassim, S

    1994-07-01

    The modular neutron detector is dedicated to the study of heavy ion reaction mechanisms. Monte Carlo simulations are performed for the optimization of the NE213 scintillator cell size and the general geometrical setup for the DEMON multidetector of neutrons with a minimum of cross-talk. Tests are performed with various types of photomultiplier tubes and scintillators. Using high energy neutron beams, more than six different reaction processes were identified with pulse shape discrimination by the charge comparison method. Cross sections were estimated. Light yields of charged particles p, d, t and alpha in the NE213 organic scintillator were analyzed using different theoretical approaches, and the intrinsic efficiency of the DEMON`s modules was measured and compared to Monte Carlo calculations. The DEMON experimental filter was simulated and has been associated with the Gemini physical events generator to study the performance of such a multidetector. Thus, the DEMON response for neutron evaporation of excited nuclei and its influence on energy measurement and temperature determination were studied. The same filter was used to simulate pre- and post-fission emission of neutrons for the fission process of the composite {sup 126}Ba system formed in the {sup 19}F + {sup 107}Ag entrance channel. (from author) 70 figs., 99 refs.

  13. A fast and flexible reactor physics model for simulating neutron spectra and depletion in fast reactors - 202

    International Nuclear Information System (INIS)

    Recktenwald, G.D.; Bronk, L.A.; Deinert, M.R.

    2010-01-01

    Determining the time dependent concentration of isotopes within a nuclear reactor core is central to the analysis of nuclear fuel cycles. We present a fast, flexible tool for determining the time dependent neutron spectrum within fast reactors. The code (VBUDS: visualization, burnup, depletion and spectra) uses a two region, multigroup collision probability model to simulate the energy dependent neutron flux and tracks the buildup and burnout of 24 actinides, as well as fission products. While originally developed for LWR simulations, the model is shown to produce fast reactor spectra that show high degree of fidelity to available fast reactor benchmarks. (authors)

  14. Monte Carlo simulation for fragment mass and kinetic energy distributions from the neutron-induced fission of 235U

    International Nuclear Information System (INIS)

    Montoya, M.; Rojas, J.; Saettone, E.

    2007-01-01

    The mass and kinetic energy distribution of nuclear fragments from the thermal neutron-induced fission of 235 U have been studied using a Monte Carlo simulation. Besides reproducing the pronounced broadening on the standard deviation of the final fragment kinetic energy distribution (σ e (m)) around the mass number m = 109, our simulation also produces a second broadening around m = 125 that is in agreement with the experimental data obtained by Belhafaf et al. These results are a consequence of the characteristics of the neutron emission, the variation in the primary fragment mean kinetic energy, and the yield as a function of the mass. (Author)

  15. Monte Carlo simulation for fragment mass and kinetic energy distributions from the neutron-induced fission of {sup 235}U

    Energy Technology Data Exchange (ETDEWEB)

    Montoya, M.; Rojas, J. [Instituto Peruano de Energia Nuclear, Av. Canada 1470, Lima 41 (Peru); Saettone, E. [Facultad de Ciencias, Universidad Nacional de lngenieria, Av. Tupac Amaru 210, Apartado 31-139, Lima (Peru)

    2007-07-01

    The mass and kinetic energy distribution of nuclear fragments from the thermal neutron-induced fission of {sup 235}U have been studied using a Monte Carlo simulation. Besides reproducing the pronounced broadening on the standard deviation of the final fragment kinetic energy distribution ({sigma}{sub e}(m)) around the mass number m = 109, our simulation also produces a second broadening around m = 125 that is in agreement with the experimental data obtained by Belhafaf et al. These results are a consequence of the characteristics of the neutron emission, the variation in the primary fragment mean kinetic energy, and the yield as a function of the mass. (Author)

  16. Stabilized FE simulation of prototype thermal-hydraulics problems with integrated adjoint-based capabilities

    International Nuclear Information System (INIS)

    Shadid, J.N.; Smith, T.M.; Cyr, E.C.; Wildey, T.M.; Pawlowski, R.P.

    2016-01-01

    A critical aspect of applying modern computational solution methods to complex multiphysics systems of relevance to nuclear reactor modeling, is the assessment of the predictive capability of specific proposed mathematical models. In this respect the understanding of numerical error, the sensitivity of the solution to parameters associated with input data, boundary condition uncertainty, and mathematical models is critical. Additionally, the ability to evaluate and or approximate the model efficiently, to allow development of a reasonable level of statistical diagnostics of the mathematical model and the physical system, is of central importance. In this study we report on initial efforts to apply integrated adjoint-based computational analysis and automatic differentiation tools to begin to address these issues. The study is carried out in the context of a Reynolds averaged Navier–Stokes approximation to turbulent fluid flow and heat transfer using a particular spatial discretization based on implicit fully-coupled stabilized FE methods. Initial results are presented that show the promise of these computational techniques in the context of nuclear reactor relevant prototype thermal-hydraulics problems.

  17. Stabilized FE simulation of prototype thermal-hydraulics problems with integrated adjoint-based capabilities

    Energy Technology Data Exchange (ETDEWEB)

    Shadid, J.N., E-mail: jnshadi@sandia.gov [Sandia National Laboratories, Computational Mathematics Department (United States); Department of Mathematics and Statistics, University of New Mexico (United States); Smith, T.M. [Sandia National Laboratories, Multiphysics Applications Department (United States); Cyr, E.C. [Sandia National Laboratories, Computational Mathematics Department (United States); Wildey, T.M. [Sandia National Laboratories, Optimization and UQ Department (United States); Pawlowski, R.P. [Sandia National Laboratories, Multiphysics Applications Department (United States)

    2016-09-15

    A critical aspect of applying modern computational solution methods to complex multiphysics systems of relevance to nuclear reactor modeling, is the assessment of the predictive capability of specific proposed mathematical models. In this respect the understanding of numerical error, the sensitivity of the solution to parameters associated with input data, boundary condition uncertainty, and mathematical models is critical. Additionally, the ability to evaluate and or approximate the model efficiently, to allow development of a reasonable level of statistical diagnostics of the mathematical model and the physical system, is of central importance. In this study we report on initial efforts to apply integrated adjoint-based computational analysis and automatic differentiation tools to begin to address these issues. The study is carried out in the context of a Reynolds averaged Navier–Stokes approximation to turbulent fluid flow and heat transfer using a particular spatial discretization based on implicit fully-coupled stabilized FE methods. Initial results are presented that show the promise of these computational techniques in the context of nuclear reactor relevant prototype thermal-hydraulics problems.

  18. The sensitivity of LaBr{sub 3}:Ce scintillation detectors to low energy neutrons: Measurement and Monte Carlo simulation

    Energy Technology Data Exchange (ETDEWEB)

    Tain, J.L., E-mail: tain@ific.uv.es [Instituto de Física Corpuscular, CSIC–Universidad de Valencia, Apdo. Correos 22085, E-46071 Valencia (Spain); Agramunt, J.; Algora, A. [Instituto de Física Corpuscular, CSIC–Universidad de Valencia, Apdo. Correos 22085, E-46071 Valencia (Spain); Aprahamian, A. [University of Notre Dame, Department of Physics, IN 46556, Notre Dame (United States); Cano-Ott, D. [Centro de Investigaciones Energéticas Medioambientales y Tecnológicas, E-28040 Madrid (Spain); Fraile, L.M. [Universidad Complutense, Grupo de Fisica Nuclear, CEI Moncloa, E-28040 Madrid (Spain); Guerrero, C. [CERN, Geneva (Switzerland); Jordan, M.D. [Instituto de Física Corpuscular, CSIC–Universidad de Valencia, Apdo. Correos 22085, E-46071 Valencia (Spain); Mach, H. [University of Notre Dame, Department of Physics, IN 46556, Notre Dame (United States); Universidad Complutense, Grupo de Fisica Nuclear, CEI Moncloa, E-28040 Madrid (Spain); Martinez, T.; Mendoza, E. [Centro de Investigaciones Energéticas Medioambientales y Tecnológicas, E-28040 Madrid (Spain); Mosconi, M.; Nolte, R. [Physikalisch-Technische Bundesanstalt, D-38116 Braunschweig (Germany)

    2015-02-21

    The neutron sensitivity of a cylindrical ⊘1.5 in.×1.5 in. LaBr{sub 3}:Ce scintillation detector was measured using quasi-monoenergetic neutron beams in the energy range from 40 keV to 2.5 MeV. In this energy range the detector is sensitive to γ-rays generated in neutron inelastic and capture processes. The experimental energy response was compared with Monte Carlo simulations performed with the Geant4 simulation toolkit using the so-called High Precision Neutron Models. These models rely on relevant information stored in evaluated nuclear data libraries. The performance of the Geant4 Neutron Data Library as well as several standard nuclear data libraries was investigated. In the latter case this was made possible by the use of a conversion tool that allowed the direct use of the data from other libraries in Geant4. Overall it was found that there was good agreement with experiment for some of the neutron data bases like ENDF/B-VII.0 or JENDL-3.3 but not with the others such as ENDF/B-VI.8 or JEFF-3.1.

  19. A stochastic model for neutron simulation considering the spectrum and nuclear properties with continuous dependence of energy

    International Nuclear Information System (INIS)

    Camargo, Dayana Queiroz de

    2011-01-01

    This thesis has developed a stochastic model to simulate the neutrons transport in a heterogeneous environment, considering continuous neutron spectra and the nuclear properties with its continuous dependence on energy. This model was implemented using Monte Carlo method for the propagation of neutrons in different environment. Due to restrictions with respect to the number of neutrons that can be simulated in reasonable computational processing time introduced the variable control volume along the (pseudo-) periodic boundary conditions in order to overcome this problem. The choice of class physical Monte Carlo is due to the fact that it can decompose into simpler constituents the problem of solve a transport equation. The components may be treated separately, these are the propagation and interaction while respecting the laws of energy conservation and momentum, and the relationships that determine the probability of their interaction. We are aware of the fact that the problem approached in this thesis is far from being comparable to building a nuclear reactor, but this discussion the main target was to develop the Monte Carlo model, implement the code in a computer language that allows extensions of modular way. This study allowed a detailed analysis of the influence of energy on the neutron population and its impact on the life cycle of neutrons. From the results, even for a simple geometrical arrangement, we can conclude the need to consider the energy dependence, i.e. an spectral effective multiplication factor should be introduced each energy group separately. (author)

  20. Neutronic design of pulse operation simulating device for in-pile functional test of fusion blanket by MCNP

    Energy Technology Data Exchange (ETDEWEB)

    Nagao, Yoshiharu; Nakamichi, Masaru; Kawamura, Hiroshi [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan)

    2000-03-01

    The pulse operation of a fusion reactor can be simulated in a fission reactor by controlling the neutron flux entering a test section by using a rotating 'hollow cylinder with window' made of hafnium. The rotating cylinder is installed between the test section and the fixed outer neutron absorber cylinder and is also made of hafnium with an opening in the direction to the core center. For gathering engineering data for the tritium breeding blanket such as characteristics of temperature change, tritium release and recovery, etc., it is desirable that the ratio of minimum to maximum thermal neutron fluxes is greater than 1:10. Design calculations were performed for the test assembly which considered local neutronic effects and the mechanical constraints of the device. From the results of these calculations, the ratio of minimum to maximum thermal neutron flux under irradiation would be about 1:10 using a pulse operation simulating device which has a thickness of 6.5 mm and a 150deg window angle for the rotating hollow cylinder and 5.0 mm in thickness of fixed neutron absorber. (author)

  1. Sci-Sat AM: Brachy - 04: Neutron production around a radiation therapy linac bunker - monte carlo simulations and physical measurements.

    Science.gov (United States)

    Khatchadourian, R; Davis, S; Evans, M; Licea, A; Seuntjens, J; Kildea, J

    2012-07-01

    Photoneutrons are a major component of the equivalent dose in the maze and near the door of linac bunkers. Physical measurements and Monte Carlo (MC) calculations of neutron dose are key for validating bunker design with respect to health regulations. We attempted to use bubble detectors and a 3 He neutron spectrometer to measure neutron equivalent dose and neutron spectra in the maze and near the door of one of our bunkers. We also ran MC simulations with MCNP5 to measure the neutron fluence in the same region. Using a point source of neutrons, a Clinac 1800 linac operating at 10 MV was simulated and the fluence measured at various locations of interest. We describe the challenges faced when measuring dose with bubble detectors in the maze and the complexity of photoneutron spectrometry with linacs operating in pulsed mode. Finally, we report on the development of a userfriendly GUI for shielding calculations based on the NCRP 151 formalism. © 2012 American Association of Physicists in Medicine.

  2. General Relativistic Radiation MHD Simulations of Supercritical Accretion onto a Magnetized Neutron Star: Modeling of Ultraluminous X-Ray Pulsars

    Energy Technology Data Exchange (ETDEWEB)

    Takahashi, Hiroyuki R. [Center for Computational Astrophysics, National Astronomical Observatory of Japan, National Institutes of Natural Sciences, Mitaka, Tokyo 181-8588 (Japan); Ohsuga, Ken, E-mail: takahashi@cfca.jp, E-mail: ken.ohsuga@nao.ac.jp [Division of Theoretical Astronomy, National Astronomical Observatory of Japan, National Institutes of Natural Sciences, Mitaka, Tokyo 181-8588 (Japan)

    2017-08-10

    By performing 2.5-dimensional general relativistic radiation magnetohydrodynamic simulations, we demonstrate supercritical accretion onto a non-rotating, magnetized neutron star, where the magnetic field strength of dipole fields is 10{sup 10} G on the star surface. We found the supercritical accretion flow consists of two parts: the accretion columns and the truncated accretion disk. The supercritical accretion disk, which appears far from the neutron star, is truncated at around ≃3 R {sub *} ( R {sub *} = 10{sup 6} cm is the neutron star radius), where the magnetic pressure via the dipole magnetic fields balances with the radiation pressure of the disks. The angular momentum of the disk around the truncation radius is effectively transported inward through magnetic torque by dipole fields, inducing the spin up of a neutron star. The evaluated spin-up rate, ∼−10{sup −11} s s{sup −1}, is consistent with the recent observations of the ultraluminous X-ray pulsars. Within the truncation radius, the gas falls onto a neutron star along the dipole fields, which results in a formation of accretion columns onto the northern and southern hemispheres. The net accretion rate and the luminosity of the column are ≃66 L {sub Edd}/ c {sup 2} and ≲10 L {sub Edd}, where L {sub Edd} is the Eddington luminosity and c is the light speed. Our simulations support a hypothesis whereby the ultraluminous X-ray pulsars are powered by the supercritical accretion onto the magnetized neutron stars.

  3. Simulation and high performance computing-Building a predictive capability for fusion

    International Nuclear Information System (INIS)

    Strand, P.I.; Coelho, R.; Coster, D.; Eriksson, L.-G.; Imbeaux, F.; Guillerminet, Bernard

    2010-01-01

    The Integrated Tokamak Modelling Task Force (ITM-TF) is developing an infrastructure where the validation needs, as being formulated in terms of multi-device data access and detailed physics comparisons aiming for inclusion of synthetic diagnostics in the simulation chain, are key components. As the activity and the modelling tools are aimed for general use, although focused on ITER plasmas, a device independent approach to data transport and a standardized approach to data management (data structures, naming, and access) is being developed in order to allow cross-validation between different fusion devices using a single toolset. Extensive work has already gone into, and is continuing to go into, the development of standardized descriptions of the data (Consistent Physical Objects). The longer term aim is a complete simulation platform which is expected to last and be extended in different ways for the coming 30 years. The technical underpinning is therefore of vital importance. In particular the platform needs to be extensible and open-ended to be able to take full advantage of not only today's most advanced technologies but also be able to marshal future developments. As a full level comprehensive prediction of ITER physics rapidly becomes expensive in terms of computing resources, the simulation framework needs to be able to use both grid and HPC computing facilities. Hence data access and code coupling technologies are required to be available for a heterogeneous, possibly distributed, environment. The developments in this area are pursued in a separate project-EUFORIA (EU Fusion for ITER Applications) which is providing about 15 professional person year (ppy) per annum from 14 different institutes. The range and size of the activity is not only technically challenging but is providing some unique management challenges in that a large and geographically distributed team (a truly pan-European set of researchers) need to be coordinated on a fairly detailed

  4. Capability of the RELAP5 code to simulate natural circulation behaviour in test facilities

    International Nuclear Information System (INIS)

    Mangal, Amit; Jain, Vikas; Nayak, A.K.

    2011-01-01

    In the present study, one of the extensively used best estimate code RELAP5 has been used for simulation of steady state, transient and stability behavior of natural circulation based experimental facilities, such as the High-Pressure Natural Circulation Loop (HPNCL) and the Parallel Channel Loop (PCL) installed and operating at BARC. The test data have been generated for a range of pressure, power and subcooling conditions. The computer code RELAP5/MOD3.2 was applied to predict the transient natural circulation characteristics under single-phase and two-phase conditions, thresholds of flow instability, amplitude and frequency of flow oscillations for different operating conditions of the loops. This paper presents the effect of nodalisation in prediction of natural circulation behavior in test facilities and a comparison of experimental data in with that of code predictions. The errors associated with the predictions are also characterized

  5. The GEANT4 toolkit capability in the hadron therapy field: simulation of a transport beam line

    International Nuclear Information System (INIS)

    Cirrone, G.A.P.; Cuttone, G.; Di Rosa, F.; Raffaele, L.; Russo, G.; Guatelli, S.; Pia, M.G.

    2006-01-01

    At Laboratori Nazionali del Sud of the Instituto Nazionale di Fisica Nucleare of Catania (Sicily, Italy), the first Italian hadron therapy facility named CATANA (Centro di AdroTerapia ed Applicazioni Nucleari Avanzate) has been realized. Inside CATANA 62 MeV proton beams, accelerated by a superconducting cyclotron, are used for the radiotherapeutic treatments of some types of ocular tumours. Therapy with hadron beams still represents a pioneer technique, and only a few centers worldwide can provide this advanced specialized cancer treatment. On the basis of the experience so far gained, and considering the future hadron-therapy facilities to be developed (Rinecker, Munich Germany, Heidelberg/GSI, Darmstadt, Germany, PSI Villigen, Switzerland, CNAO, Pavia, Italy, Centro di Adroterapia, Catania, Italy) we decided to develop a Monte Carlo application based on the GEANT4 toolkit, for the design, the realization and the optimization of a proton-therapy beam line. Another feature of our project is to provide a general tool able to study the interactions of hadrons with the human tissue and to test the analytical-based treatment planning systems actually used in the routine practice. All the typical elements of a hadron-therapy line, such as diffusers, range shifters, collimators and detectors were modelled. In particular, we simulated the Markus type ionization chamber and a Gaf Chromic film as dosimeters to reconstruct the depth (Bragg peak and Spread Out Bragg Peak) and lateral dose distributions, respectively. We validated our simulated detectors comparing the results with the experimental data available in our facility

  6. The GEANT4 toolkit capability in the hadron therapy field: simulation of a transport beam line

    Science.gov (United States)

    Cirrone, G. A. P.; Cuttone, G.; Di Rosa, F.; Raffaele, L.; Russo, G.; Guatelli, S.; Pia, M. G.

    2006-01-01

    At Laboratori Nazionali del Sud of the Instituto Nazionale di Fisica Nucleare of Catania (Sicily, Italy), the first Italian hadron therapy facility named CATANA (Centro di AdroTerapia ed Applicazioni Nucleari Avanzate) has been realized. Inside CATANA 62 MeV proton beams, accelerated by a superconducting cyclotron, are used for the radiotherapeutic treatments of some types of ocular tumours. Therapy with hadron beams still represents a pioneer technique, and only a few centers worldwide can provide this advanced specialized cancer treatment. On the basis of the experience so far gained, and considering the future hadron-therapy facilities to be developed (Rinecker, Munich Germany, Heidelberg/GSI, Darmstadt, Germany, PSI Villigen, Switzerland, CNAO, Pavia, Italy, Centro di Adroterapia, Catania, Italy) we decided to develop a Monte Carlo application based on the GEANT4 toolkit, for the design, the realization and the optimization of a proton-therapy beam line. Another feature of our project is to provide a general tool able to study the interactions of hadrons with the human tissue and to test the analytical-based treatment planning systems actually used in the routine practice. All the typical elements of a hadron-therapy line, such as diffusers, range shifters, collimators and detectors were modelled. In particular, we simulated the Markus type ionization chamber and a Gaf Chromic film as dosimeters to reconstruct the depth (Bragg peak and Spread Out Bragg Peak) and lateral dose distributions, respectively. We validated our simulated detectors comparing the results with the experimental data available in our facility.

  7. The GEANT4 toolkit capability in the hadron therapy field: simulation of a transport beam line

    Energy Technology Data Exchange (ETDEWEB)

    Cirrone, G.A.P. [Laboratori Nazionali del Sud, Istituto Nazionale di Fisica Nucleare, Via S. Sofia 62, Catania (Italy); Cuttone, G. [Laboratori Nazionali del Sud, Istituto Nazionale di Fisica Nucleare, Via S. Sofia 62, Catania (Italy); Di Rosa, F. [Laboratori Nazionali del Sud, Istituto Nazionale di Fisica Nucleare, Via S. Sofia 62, Catania (Italy); Raffaele, L. [Laboratori Nazionali del Sud, Istituto Nazionale di Fisica Nucleare, Via S. Sofia 62, Catania (Italy); Russo, G. [Laboratori Nazionali del Sud, Istituto Nazionale di Fisica Nucleare, Via S. Sofia 62, Catania (Italy); Guatelli, S. [Istituto Nazionale di Fisica Nucleare, Sezione di Genova, Via Dodecaneso 33, Genova (Italy); Pia, M.G. [Istituto Nazionale di Fisica Nucleare, Sezione di Genova, Via Dodecaneso 33, Genova (Italy)

    2006-01-15

    At Laboratori Nazionali del Sud of the Instituto Nazionale di Fisica Nucleare of Catania (Sicily, Italy), the first Italian hadron therapy facility named CATANA (Centro di AdroTerapia ed Applicazioni Nucleari Avanzate) has been realized. Inside CATANA 62 MeV proton beams, accelerated by a superconducting cyclotron, are used for the radiotherapeutic treatments of some types of ocular tumours. Therapy with hadron beams still represents a pioneer technique, and only a few centers worldwide can provide this advanced specialized cancer treatment. On the basis of the experience so far gained, and considering the future hadron-therapy facilities to be developed (Rinecker, Munich Germany, Heidelberg/GSI, Darmstadt, Germany, PSI Villigen, Switzerland, CNAO, Pavia, Italy, Centro di Adroterapia, Catania, Italy) we decided to develop a Monte Carlo application based on the GEANT4 toolkit, for the design, the realization and the optimization of a proton-therapy beam line. Another feature of our project is to provide a general tool able to study the interactions of hadrons with the human tissue and to test the analytical-based treatment planning systems actually used in the routine practice. All the typical elements of a hadron-therapy line, such as diffusers, range shifters, collimators and detectors were modelled. In particular, we simulated the Markus type ionization chamber and a Gaf Chromic film as dosimeters to reconstruct the depth (Bragg peak and Spread Out Bragg Peak) and lateral dose distributions, respectively. We validated our simulated detectors comparing the results with the experimental data available in our facility.

  8. PENTrack-a simulation tool for ultracold neutrons, protons, and electrons in complex electromagnetic fields and geometries

    Science.gov (United States)

    Schreyer, W.; Kikawa, T.; Losekamm, M. J.; Paul, S.; Picker, R.

    2017-06-01

    Modern precision experiments trapping low-energy particles require detailed simulations of particle trajectories and spin precession to determine systematic measurement limitations and apparatus deficiencies. We developed PENTrack, a tool that allows to simulate trajectories of ultracold neutrons and their decay products-protons and electrons-and the precession of their spins in complex geometries and electromagnetic fields. The interaction of ultracold neutrons with matter is implemented with the Fermi-potential formalism and diffuse scattering using Lambert and microroughness models. The results of several benchmark simulations agree with STARucn v1.2, uncovered several flaws in Geant4 v10.2.2, and agree with experimental data. Experiment geometry and electromagnetic fields can be imported from commercial computer-aided-design and finite-element software. All simulation parameters are defined in simple text files allowing quick changes. The simulation code is written in C++ and is freely available at github.com/wschreyer/PENTrack.git.

  9. PENTrack—a simulation tool for ultracold neutrons, protons, and electrons in complex electromagnetic fields and geometries

    Energy Technology Data Exchange (ETDEWEB)

    Schreyer, W., E-mail: w.schreyer@tum.de [Technical University of Munich, James-Franck-Str. 1, 85748 Garching (Germany); Kikawa, T. [TRIUMF, 4004 Wesbrook Mall, Vancouver (Canada); Losekamm, M.J.; Paul, S. [Technical University of Munich, James-Franck-Str. 1, 85748 Garching (Germany); Picker, R. [TRIUMF, 4004 Wesbrook Mall, Vancouver (Canada); Simon Fraser University, 8888 University Drive, Burnaby (Canada)

    2017-06-21

    Modern precision experiments trapping low-energy particles require detailed simulations of particle trajectories and spin precession to determine systematic measurement limitations and apparatus deficiencies. We developed PENTrack, a tool that allows to simulate trajectories of ultracold neutrons and their decay products—protons and electrons—and the precession of their spins in complex geometries and electromagnetic fields. The interaction of ultracold neutrons with matter is implemented with the Fermi-potential formalism and diffuse scattering using Lambert and microroughness models. The results of several benchmark simulations agree with STARucn v1.2, uncovered several flaws in Geant4 v10.2.2, and agree with experimental data. Experiment geometry and electromagnetic fields can be imported from commercial computer-aided-design and finite-element software. All simulation parameters are defined in simple text files allowing quick changes. The simulation code is written in C++ and is freely available at (github.com/wschreyer/PENTrack.git).

  10. General-relativistic Large-eddy Simulations of Binary Neutron Star Mergers

    Energy Technology Data Exchange (ETDEWEB)

    Radice, David, E-mail: dradice@astro.princeton.edu [Institute for Advanced Study, 1 Einstein Drive, Princeton, NJ 08540 (United States)

    2017-03-20

    The flow inside remnants of binary neutron star (NS) mergers is expected to be turbulent, because of magnetohydrodynamics instability activated at scales too small to be resolved in simulations. To study the large-scale impact of these instabilities, we develop a new formalism, based on the large-eddy simulation technique, for the modeling of subgrid-scale turbulent transport in general relativity. We apply it, for the first time, to the simulation of the late-inspiral and merger of two NSs. We find that turbulence can significantly affect the structure and survival time of the merger remnant, as well as its gravitational-wave (GW) and neutrino emissions. The former will be relevant for GW observation of merging NSs. The latter will affect the composition of the outflow driven by the merger and might influence its nucleosynthetic yields. The accretion rate after black hole formation is also affected. Nevertheless, we find that, for the most likely values of the turbulence mixing efficiency, these effects are relatively small and the GW signal will be affected only weakly by the turbulence. Thus, our simulations provide a first validation of all existing post-merger GW models.

  11. Irradiations of human melanoma cells by 14 MeV neutrons; survival curves interpretation; physical simulation of neutrons interactions in the cellular medium

    International Nuclear Information System (INIS)

    Bodez, Veronique

    2000-01-01

    14 MeV neutrons are used to irradiate human melanoma cells in order to study survival curves at low dose and low dose rate. We have simulated with the MCNP code, transport of neutrons through the experimental setup to evaluate the contamination of the primary beam by gamma and electrons, for the feasibility of our experiments. We have shown a rapid decrease of the survival curve in the first cGy followed by a plateau for doses up to 30 cGy; after we observed an exponential decrease. This results are observed for the first time, for neutrons at low dose rate (5 cGy/h). In parallel with this experimental point, we have developed a simulation code which permitted the study of neutrons interactions with the cellular medium for individual cells defined as in our experimental conditions. We show that most of the energy is deposited by protons from neutron interactions with external medium, and by heavy ions for interactions into the cell. On the other hand the code gives a good order of magnitude of the dose rate, compared to the experimental values given by silicon diodes. The first results show that we can, using a theory based on induced repair of cells, give an interpretation of the observed experimental plateau. We can give an estimation of the radial distribution of dose for the tracks of charged ions, we show the possibility of calculate interaction cross sections with cellular organelles. Such a work gives interesting perspectives for the future in radiobiology, radiotherapy or radioprotection. (author) [fr

  12. A contribution to the development of the modular neutron detector (DEMON): performance evaluation through measurements and simulations

    International Nuclear Information System (INIS)

    Mouatassim, S.

    1994-07-01

    The modular neutron detector is dedicated to the study of heavy ion reaction mechanisms. Monte Carlo simulations are performed for the optimization of the NE213 scintillator cell size and the general geometrical setup for the DEMON multidetector of neutrons with a minimum of cross-talk. Tests are performed with various types of photomultiplier tubes and scintillators. Using high energy neutron beams, more than six different reaction processes were identified with pulse shape discrimination by the charge comparison method. Cross sections were estimated. Light yields of charged particles p, d, t and alpha in the NE213 organic scintillator were analyzed using different theoretical approaches, and the intrinsic efficiency of the DEMON's modules was measured and compared to Monte Carlo calculations. The DEMON experimental filter was simulated and has been associated with the Gemini physical events generator to study the performance of such a multidetector. Thus, the DEMON response for neutron evaporation of excited nuclei and its influence on energy measurement and temperature determination were studied. The same filter was used to simulate pre- and post-fission emission of neutrons for the fission process of the composite 126 Ba system formed in the 19 F + 107 Ag entrance channel. (from author) 70 figs., 99 refs

  13. Ion irradiation to simulate neutron irradiation in model graphites: Consequences for nuclear graphite

    Science.gov (United States)

    Galy, N.; Toulhoat, N.; Moncoffre, N.; Pipon, Y.; Bérerd, N.; Ammar, M. R.; Simon, P.; Deldicque, D.; Sainsot, P.

    2017-10-01

    Due to its excellent moderator and reflector qualities, graphite was used in CO2-cooled nuclear reactors such as UNGG (Uranium Naturel-Graphite-Gaz). Neutron irradiation of graphite resulted in the production of 14C which is a key issue radionuclide for the management of the irradiated graphite waste. In order to elucidate the impact of neutron irradiation on 14C behavior, we carried out a systematic investigation of irradiation and its synergistic effects with temperature in Highly Oriented Pyrolitic Graphite (HOPG) model graphite used to simulate the coke grains of nuclear graphite. We used 13C implantation in order to simulate 14C displaced from its original structural site through recoil. The collision of the impinging neutrons with the graphite matrix carbon atoms induces mainly ballistic damage. However, a part of the recoil carbon atom energy is also transferred to the graphite lattice through electronic excitation. The effects of the different irradiation regimes in synergy with temperature were simulated using ion irradiation by varying Sn(nuclear)/Se(electronic) stopping power. Thus, the samples were irradiated with different ions of different energies. The structure modifications were followed by High Resolution Transmission Electron Microscopy (HRTEM) and Raman microspectrometry. The results show that temperature generally counteracts the disordering effects of irradiation but the achieved reordering level strongly depends on the initial structural state of the graphite matrix. Thus, extrapolating to reactor conditions, for an initially highly disordered structure, irradiation at reactor temperatures (200 - 500 °C) should induce almost no change of the initial structure. On the contrary, when the structure is initially less disordered, there should be a "zoning" of the reordering: In "cold" high flux irradiated zones where the ballistic damage is important, the structure should be poorly reordered; In "hot" low flux irradiated zones where the ballistic

  14. Flexural Capability of Patterned Transparent Conductive Substrate by Performing Electrical Measurements and Stress Simulations

    Directory of Open Access Journals (Sweden)

    Chang-Chun Lee

    2016-10-01

    Full Text Available The suitability of stacked thin films for next-generation display technology was analyzed based on their properties and geometrical designs to evaluate the mechanical reliability of transparent conducting thin films utilized in flexural displays. In general, the high bending stress induced by various operation conditions is a major concern regarding the mechanical reliability of indium–tin–oxide (ITO films deposited on polyethylene terephthalate (PET substrates; mechanical reliability is commonly used to estimate the flexibility of displays. However, the pattern effect is rarely investigated to estimate the mechanical reliability of ITO/PET films. Thus, this study examined the flexible content of patterned ITO/PET films with two different line widths by conducting bending tests and sheet resistance measurements. Moreover, a stress–strain simulation enabled by finite element analysis was performed on the patterned ITO/PET to explore the stress impact of stacked film structures under various levels of flexural load. Results show that the design of the ITO/PET film can be applied in developing mechanically reliable flexible electronics.

  15. Simulating pasta phases by molecular dynamics and cold atoms. Formation in supernovae and superfluid neutrons in neutron stars

    International Nuclear Information System (INIS)

    Watanabe, Gentaro

    2010-01-01

    In dense stars such as collapsing cores of supernovae and neutron stars, nuclear 'pasta' such as rod-like and slab-like nuclei are speculated to exist. However, whether or not they are actually formed in supernova cores is still unclear. Here we solve this problem by demonstrating that a lattice of rod-like nuclei is formed from a bcc lattice by compression. We also find that the formation process is triggered by an attractive force between nearest neighbor nuclei, which starts to act when their density profile overlaps, rather than the fission instability. We also discuss the connection between pasta phases in neutron star crusts and ultracold Fermi gases. (author)

  16. Merger of white dwarf-neutron star binaries: Prelude to hydrodynamic simulations in general relativity

    International Nuclear Information System (INIS)

    Paschalidis, Vasileios; MacLeod, Morgan; Baumgarte, Thomas W.; Shapiro, Stuart L.

    2009-01-01

    White dwarf-neutron star binaries generate detectable gravitational radiation. We construct Newtonian equilibrium models of corotational white dwarf-neutron star (WDNS) binaries in circular orbit and find that these models terminate at the Roche limit. At this point the binary will undergo either stable mass transfer (SMT) and evolve on a secular time scale, or unstable mass transfer (UMT), which results in the tidal disruption of the WD. The path a given binary will follow depends primarily on its mass ratio. We analyze the fate of known WDNS binaries and use population synthesis results to estimate the number of LISA-resolved galactic binaries that will undergo either SMT or UMT. We model the quasistationary SMT epoch by solving a set of simple ordinary differential equations and compute the corresponding gravitational waveforms. Finally, we discuss in general terms the possible fate of binaries that undergo UMT and construct approximate Newtonian equilibrium configurations of merged WDNS remnants. We use these configurations to assess plausible outcomes of our future, fully relativistic simulations of these systems. If sufficient WD debris lands on the NS, the remnant may collapse, whereby the gravitational waves from the inspiral, merger, and collapse phases will sweep from LISA through LIGO frequency bands. If the debris forms a disk about the NS, it may fragment and form planets.

  17. Monte Carlo simulation of neutron transport in a homogeneous reactor with a partially inserted control rod

    International Nuclear Information System (INIS)

    Karlsson, J.K.H.; Linden, P.

    1997-01-01

    The neutron transport in a bare, cylindrical and homogeneous reactor, with and without the presence of a central partially inserted control rod, has been simulated by using a Monte Carlo transport code. The behaviour of both the flux and current in this system have been investigated. We have found that the flux and especially the current are strongly affected by the presence of the control rod in its close vicinity. The results indicate the feasibility to identify the position and especially the tip of the rod from the flux and current. Further, the direction to the rod can be found from the current vector. The information content regarding the position of the rod, in both the neutron flux and the current, decays strongly as a function of distance and it is dependent on the size of the rod. In our model, the practical range over which the flux or current can be a useful indicator of the position of the tip of the rod is about 10-15 cm for a rod with a diameter of 2 cm. The practical range for identification of the position of the rod is greater for a rod of larger diameter

  18. PERIGEE computer codes for reactor simulation in 3 dimensions, using 1 or 2 neutron velocity groups

    International Nuclear Information System (INIS)

    Olson, A.P.

    1964-02-01

    PERIGEE is a code written in SNAP for the G-20 computer. It solves the one- or two-group neutron diffusion equations by finite-difference methods on a three-dimensional, uniform mesh having a common spacing in the two directions normal to the fuel channels. The positions of mesh points along a fuel channel, relative to points in adjacent channels, may correspond to either NPD or CANDU fuel bundle positions. The extrapolated flux boundary may be specified in sufficient detail to represent a tapered or stepped circumferential reflector, a variable axial length and, for a reactor with axis horizontal, a variable moderator level and a variable plane bottom surface equivalent to the CANDU dump structure. The neutron flux may be normalized to give a specified power output from the hottest fuel bundle or hottest channel, or to give a total thermal power limited by the turbine and generator. Reactor operation may be simulated in finite time steps, taking into account any fuel shifts, any changes in moderator level and the change in nuclear properties of the fuel with increasing irradiation. The appropriate properties are obtained by interpolation from tables supplied for as many as 8 types of fuel bundle. The mean fuel exit burnup can be calculated at equilibrium for a reactor in which the exit burnups for two zones may be adjusted to give radial power flattening and the fuelling schedules may be designed to give axial power flattening in one or both zones. (author)

  19. DIANE, a simulation code for the interaction of neutrons with living tissues. Application to low doses of fast neutrons on human tumoral cells

    International Nuclear Information System (INIS)

    Nenot, M.L.

    2003-07-01

    Our work deals with the irradiation of cells and living tissues by 14 MeV neutrons at very low doses (a few 10 -2 Gy). Such experiments require an accurate knowledge of the values of neutron dose rates and fluences at the level of cell cultures. We have performed measurements of fluence rates through an activation method applied to gold and copper foils. The fluence rate is deduced from the gamma rays emitted by the irradiated foils. Neutron doses and dose rates have been measured through varied methods: PIN diodes, ionization tissue equivalent chambers, and Geiger-Mueller counters. We have designed the DIANE code to simulate the impact of energetic neutrons on cells. This code can be used with isolated cells or macroscopic tissues, it takes into account the roles of the ionisation electrons produced by recoil nuclei entering the cell. This point is all the more important since recent works have highlighted the impact of very low energy electrons on DNA. (A.C.)

  20. Analysis of ASTEC-Na capabilities for simulating a loss of flow CABRI experiment

    International Nuclear Information System (INIS)

    Flores y Flores, A.; Matuzas, V.; Perez-Martin, S.; Bandini, G.; Ederli, S.; Ammirabile, L.; Pfrang, W.

    2016-01-01

    Highlights: • ASTEC-Na results for CABRI BI1 test have been compared with experimental data. • The ASTEC-Na calculations reached the boiling onset within the error bar of the test. • The coolant axial profile in ASTEC-Na fit almost perfectly with the experimental data. • All the calculations have a good agreement with the two-phase front downwards. • All the calculations have a worse agreement with the two-phase front upwards. - Abstract: This paper presents simulation results of the CABRI BI1 test using the code ASTEC-Na, currently under development, as well as a comparison of the results with available experimental data. The EU-JASMIN project (7th FP of EURATOM) centres on the development and validation of the new severe accident analysis code ASTEC-Na (Accident Source Term Evaluation Code) for sodium-cooled fast reactors whose owner and developer is IRSN. A series of experiments performed in the past (CABRI/SCARABEE experiments) and new experiments to be conducted in the new experimental sodium facility KASOLA have been chosen to validate the developed ASTEC-Na code. One of the in-pile experiments considered for the validation of ASTEC-Na thermal–hydraulic models is the CABRI BI1 test, a pure loss-of-flow transient using a low burnup MOX fuel pin. The experiment resulted in a channel voiding as a result of the flow coast-down leading to clad melting. Only some fuel melting took place. Results from the analysis of this test using SIMMER and SAS-SFR codes are also presented in this work to check their suitability for further code benchmarking purposes.

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

    International Nuclear Information System (INIS)

    Collins, Benjamin; Stimpson, Shane; Kelley, Blake W.; Young, Mitchell T.H.; Kochunas, Brendan; Graham, Aaron; Larsen, Edward W.; Downar, Thomas; Godfrey, Andrew

    2016-01-01

    A consistent “2D/1D” neutron transport method is derived from the 3D Boltzmann transport equation, to calculate fuel-pin-resolved neutron fluxes for realistic full-core Pressurized Water Reactor (PWR) problems. The 2D/1D method employs the Method of Characteristics to discretize the radial variables and a lower order transport solution to discretize the axial variable. This paper describes the theory of the 2D/1D method and its implementation in the MPACT code, which has become the whole-core deterministic neutron transport solver for the Consortium for Advanced Simulations of Light Water Reactors (CASL) core simulator VERA-CS. Several applications have been performed on both leadership-class and industry-class computing clusters. Results are presented for whole-core solutions of the Watts Bar Nuclear Power Station Unit 1 and compared to both continuous-energy Monte Carlo results and plant data.

  2. MCNP simulation of the influence of the external moisture on low calorific value in the coal quality analysis by neutron

    International Nuclear Information System (INIS)

    Liu Dekun; Zhang Hongyu; Zhang Lihong; Dong Huan; Gu Deshan

    2012-01-01

    An important index in assessment of coal quality is low calorific value. Using neutron to analysis coal quality, the more the coal moisture content, especially the increasing of external moisture will reduce the low calorific value. The principle of coal quality analysis by neutron prompt Gamma-ray is introduced. The influence of the gamma count of the carbon element peak with increasing external moisture in coal samples was simulated using MCNP code. And discussed the reasons how external moisture content influence the calorific value. Simulation results indicate that with the increasing of external moisture in the coal samples, the gamma count of the carbon element peak dwindling, and the low calorific value reducing. The conclusion is : using neutrons method to analysis coal quality, the more external moisture content, the larger error of the measurement results of the carbon element, and will influence the calculation accuracy of the low calorific value. (authors)

  3. Experimental and Simulated Characterization of a Beam Shaping Assembly for Accelerator- Based Boron Neutron Capture Therapy (AB-BNCT)

    International Nuclear Information System (INIS)

    Burlon, Alejandro A.; Valda, Alejandro A.; Girola, Santiago; Minsky, Daniel M.; Kreiner, Andres J.

    2010-01-01

    In the frame of the construction of a Tandem Electrostatic Quadrupole Accelerator facility devoted to the Accelerator-Based Boron Neutron Capture Therapy, a Beam Shaping Assembly has been characterized by means of Monte-Carlo simulations and measurements. The neutrons were generated via the 7 Li(p, n) 7 Be reaction by irradiating a thick LiF target with a 2.3 MeV proton beam delivered by the TANDAR accelerator at CNEA. The emerging neutron flux was measured by means of activation foils while the beam quality and directionality was evaluated by means of Monte Carlo simulations. The parameters show compliance with those suggested by IAEA. Finally, an improvement adding a beam collimator has been evaluated.

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

    Energy Technology Data Exchange (ETDEWEB)

    Collins, Benjamin, E-mail: collinsbs@ornl.gov [Oak Ridge National Laboratory, One Bethel Valley Rd., Oak Ridge, TN 37831 (United States); Stimpson, Shane, E-mail: stimpsonsg@ornl.gov [Oak Ridge National Laboratory, One Bethel Valley Rd., Oak Ridge, TN 37831 (United States); Kelley, Blake W., E-mail: kelleybl@umich.edu [Department of Nuclear Engineering and Radiological Sciences, University of Michigan, Ann Arbor, MI 48109 (United States); Young, Mitchell T.H., E-mail: youngmit@umich.edu [Department of Nuclear Engineering and Radiological Sciences, University of Michigan, Ann Arbor, MI 48109 (United States); Kochunas, Brendan, E-mail: bkochuna@umich.edu [Department of Nuclear Engineering and Radiological Sciences, University of Michigan, Ann Arbor, MI 48109 (United States); Graham, Aaron, E-mail: aarograh@umich.edu [Department of Nuclear Engineering and Radiological Sciences, University of Michigan, Ann Arbor, MI 48109 (United States); Larsen, Edward W., E-mail: edlarsen@umich.edu [Department of Nuclear Engineering and Radiological Sciences, University of Michigan, Ann Arbor, MI 48109 (United States); Downar, Thomas, E-mail: downar@umich.edu [Department of Nuclear Engineering and Radiological Sciences, University of Michigan, Ann Arbor, MI 48109 (United States); Godfrey, Andrew, E-mail: godfreyat@ornl.gov [Oak Ridge National Laboratory, One Bethel Valley Rd., Oak Ridge, TN 37831 (United States)

    2016-12-01

    A consistent “2D/1D” neutron transport method is derived from the 3D Boltzmann transport equation, to calculate fuel-pin-resolved neutron fluxes for realistic full-core Pressurized Water Reactor (PWR) problems. The 2D/1D method employs the Method of Characteristics to discretize the radial variables and a lower order transport solution to discretize the axial variable. This paper describes the theory of the 2D/1D method and its implementation in the MPACT code, which has become the whole-core deterministic neutron transport solver for the Consortium for Advanced Simulations of Light Water Reactors (CASL) core simulator VERA-CS. Several applications have been performed on both leadership-class and industry-class computing clusters. Results are presented for whole-core solutions of the Watts Bar Nuclear Power Station Unit 1 and compared to both continuous-energy Monte Carlo results and plant data.

  5. Calculations to support JET neutron yield calibration: Modelling of neutron emission from a compact DT neutron generator

    Science.gov (United States)

    Čufar, Aljaž; Batistoni, Paola; Conroy, Sean; Ghani, Zamir; Lengar, Igor; Milocco, Alberto; Packer, Lee; Pillon, Mario; Popovichev, Sergey; Snoj, Luka; JET Contributors

    2017-03-01

    At the Joint European Torus (JET) the ex-vessel fission chambers and in-vessel activation detectors are used as the neutron production rate and neutron yield monitors respectively. In order to ensure that these detectors produce accurate measurements they need to be experimentally calibrated. A new calibration of neutron detectors to 14 MeV neutrons, resulting from deuterium-tritium (DT) plasmas, is planned at JET using a compact accelerator based neutron generator (NG) in which a D/T beam impinges on a solid target containing T/D, producing neutrons by DT fusion reactions. This paper presents the analysis that was performed to model the neutron source characteristics in terms of energy spectrum, angle-energy distribution and the effect of the neutron generator geometry. Different codes capable of simulating the accelerator based DT neutron sources are compared and sensitivities to uncertainties in the generator's internal structure analysed. The analysis was performed to support preparation to the experimental measurements performed to characterize the NG as a calibration source. Further extensive neutronics analyses, performed with this model of the NG, will be needed to support the neutron calibration experiments and take into account various differences between the calibration experiment and experiments using the plasma as a source of neutrons.

  6. Calculations to support JET neutron yield calibration: Modelling of neutron emission from a compact DT neutron generator

    Energy Technology Data Exchange (ETDEWEB)

    Čufar, Aljaž, E-mail: aljaz.cufar@ijs.si [Reactor Physics Department, Jožef Stefan Institute, Jamova cesta 39, SI-1000 Ljubljana (Slovenia); EUROfusion Consortium, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Batistoni, Paola [ENEA, Department of Fusion and Nuclear Safety Technology, I-00044 Frascati, Rome (Italy); EUROfusion Consortium, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Conroy, Sean [Uppsala University, Department of Physics and Astronomy, PO Box 516, SE-75120 Uppsala (Sweden); EUROfusion Consortium, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Ghani, Zamir [Culham Centre for Fusion Energy, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); EUROfusion Consortium, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Lengar, Igor [Reactor Physics Department, Jožef Stefan Institute, Jamova cesta 39, SI-1000 Ljubljana (Slovenia); EUROfusion Consortium, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Milocco, Alberto; Packer, Lee [Culham Centre for Fusion Energy, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); EUROfusion Consortium, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Pillon, Mario [ENEA, Department of Fusion and Nuclear Safety Technology, I-00044 Frascati, Rome (Italy); EUROfusion Consortium, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Popovichev, Sergey [Culham Centre for Fusion Energy, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); EUROfusion Consortium, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Snoj, Luka [Reactor Physics Department, Jožef Stefan Institute, Jamova cesta 39, SI-1000 Ljubljana (Slovenia); EUROfusion Consortium, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom)

    2017-03-01

    At the Joint European Torus (JET) the ex-vessel fission chambers and in-vessel activation detectors are used as the neutron production rate and neutron yield monitors respectively. In order to ensure that these detectors produce accurate measurements they need to be experimentally calibrated. A new calibration of neutron detectors to 14 MeV neutrons, resulting from deuterium–tritium (DT) plasmas, is planned at JET using a compact accelerator based neutron generator (NG) in which a D/T beam impinges on a solid target containing T/D, producing neutrons by DT fusion reactions. This paper presents the analysis that was performed to model the neutron source characteristics in terms of energy spectrum, angle–energy distribution and the effect of the neutron generator geometry. Different codes capable of simulating the accelerator based DT neutron sources are compared and sensitivities to uncertainties in the generator's internal structure analysed. The analysis was performed to support preparation to the experimental measurements performed to characterize the NG as a calibration source. Further extensive neutronics analyses, performed with this model of the NG, will be needed to support the neutron calibration experiments and take into account various differences between the calibration experiment and experiments using the plasma as a source of neutrons.

  7. Hydration of Caffeine at High Temperature by Neutron Scattering and Simulation Studies.

    Science.gov (United States)

    Tavagnacco, L; Brady, J W; Bruni, F; Callear, S; Ricci, M A; Saboungi, M L; Cesàro, A

    2015-10-22

    The solvation of caffeine in water is examined with neutron diffraction experiments at 353 K. The experimental data, obtained by taking advantage of isotopic H/D substitution in water, were analyzed by empirical potential structure refinement (EPSR) in order to extract partial structure factors and site-site radial distribution functions. In parallel, molecular dynamics (MD) simulations were carried out to interpret the data and gain insight into the intermolecular interactions in the solutions and the solvation process. The results obtained with the two approaches evidence differences in the individual radial distribution functions, although both confirm the presence of caffeine stacks at this temperature. The two approaches point to different accessibility of water to the caffeine sites due to different stacking configurations.

  8. Measurements and simulations of electromagnetic perturbations on neutronic and thermodynamic control systems of nuclear reactors

    International Nuclear Information System (INIS)

    Lecompte, J.-C.; Buisson, Jacques.

    1981-01-01

    Neutronics and thermodynamics measuring devices realize the control and nuclear safety of the reactor and the core. They cause the emergency stop of the reactor if the informations given by the sensors are out of the normal operation tolerances. Defining and assuring a protection level sufficiently high against electromagnetic disturbances is essential for avoiding those measuring devices of giving erroneous indications in the environment where they are located. To obtain this result, one procceds according to the following three steps: a) In situ measurements of the ambient level of disturbances (oscilloscope analysis, statistical measurements). b) Measurement in laboratory of the immunity against the disturbances caused by every constituent of the device (electronic cables, connectors) and then the complete device. c) In situ simulation on the devices of the perturbations caused by the measured ambient level [fr

  9. Verification of MCNP simulation of neutron flux parameters at TRIGA MK II reactor of Malaysia.

    Science.gov (United States)

    Yavar, A R; Khalafi, H; Kasesaz, Y; Sarmani, S; Yahaya, R; Wood, A K; Khoo, K S

    2012-10-01

    A 3-D model for 1 MW TRIGA Mark II research reactor was simulated. Neutron flux parameters were calculated using MCNP-4C code and were compared with experimental results obtained by k(0)-INAA and absolute method. The average values of φ(th),φ(epi), and φ(fast) by MCNP code were (2.19±0.03)×10(12) cm(-2)s(-1), (1.26±0.02)×10(11) cm(-2)s(-1) and (3.33±0.02)×10(10) cm(-2)s(-1), respectively. These average values were consistent with the experimental results obtained by k(0)-INAA. The findings show a good agreement between MCNP code results and experimental results. Copyright © 2012 Elsevier Ltd. All rights reserved.

  10. GPU-based high performance Monte Carlo simulation in neutron transport

    Energy Technology Data Exchange (ETDEWEB)

    Heimlich, Adino; Mol, Antonio C.A.; Pereira, Claudio M.N.A. [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil). Lab. de Inteligencia Artificial Aplicada], e-mail: cmnap@ien.gov.br

    2009-07-01

    Graphics Processing Units (GPU) are high performance co-processors intended, originally, to improve the use and quality of computer graphics applications. Since researchers and practitioners realized the potential of using GPU for general purpose, their application has been extended to other fields out of computer graphics scope. The main objective of this work is to evaluate the impact of using GPU in neutron transport simulation by Monte Carlo method. To accomplish that, GPU- and CPU-based (single and multicore) approaches were developed and applied to a simple, but time-consuming problem. Comparisons demonstrated that the GPU-based approach is about 15 times faster than a parallel 8-core CPU-based approach also developed in this work. (author)

  11. GPU-based high performance Monte Carlo simulation in neutron transport

    International Nuclear Information System (INIS)

    Heimlich, Adino; Mol, Antonio C.A.; Pereira, Claudio M.N.A.

    2009-01-01

    Graphics Processing Units (GPU) are high performance co-processors intended, originally, to improve the use and quality of computer graphics applications. Since researchers and practitioners realized the potential of using GPU for general purpose, their application has been extended to other fields out of computer graphics scope. The main objective of this work is to evaluate the impact of using GPU in neutron transport simulation by Monte Carlo method. To accomplish that, GPU- and CPU-based (single and multicore) approaches were developed and applied to a simple, but time-consuming problem. Comparisons demonstrated that the GPU-based approach is about 15 times faster than a parallel 8-core CPU-based approach also developed in this work. (author)

  12. Simulation of pulsed neutron activation for determination of water flow in pipes

    International Nuclear Information System (INIS)

    Mattsson, H.; Owrang, F.; Nordlund, A.

    2002-01-01

    The effect of the asymmetric distribution of activated water in PNA (pulsed neutron activation) measurements has been investigated experimentally by depositing a small amount of colour, simulating the activated water, in a transparent Plexiglas pipe. Based on the colour experiments, a semi-empirical model has been developed that describes the distribution of the activated water at different distances from the activation point. The model shows that the combination of inhomogeneous activation and a radial velocity profile makes the mean velocity of the activity lower than the mean velocity of the water. It can also be seen that the velocity of the activity increases as the distance from the activation point increases. The model has been compared with experimental values from PNA measurements and the measured mean velocity shows a similar dependence on the distance form the activation point. (orig.) [de

  13. Nanocrystallite characterization of milled simulated dry process fuel powders by neutron diffraction

    International Nuclear Information System (INIS)

    Ryu, Ho Jin; Kang, Kwon Ho; Moon, Je Sun; Song, Kee Chan; Choi, Yong Nam

    2003-01-01

    The nano-scale crystallite sizes of simulated spent fuel powders were measured by the neutron diffraction line broadening method in order to analyze the sintering behavior of the dry process fuel. The mixed U0 2 and fission product oxide powders were dry-milled in an attritor for 30, 60, and 120 min. The diffraction patterns of the powders were obtained by using the high resolution powder diffractometer in the HANARO research reactor. Diffraction line broadening due to crystallite size was measured using various techniques such as the Stokes' deconvolution, profile fitting methods using Cauchy function, Gaussian function, and Voigt function, and the Warren-Averbach method. The r.m.s. strain, stacking fault, twin and dislocation density were measured using the information from the diffraction pattern. The realistic crystallite size can be obtained after separation of the contribution from the non-uniform strain, stacking fault and twin

  14. DUST, Albedo Monte-Carlo Simulation of Neutron Streaming in Multi-legged Square Concrete Ducts

    International Nuclear Information System (INIS)

    1993-01-01

    1 - Description of program or function: DUST simulates the thermal neutron streaming through multi-legged square concrete ducts. 2 - Method of solution: DUST uses the albedo Monte Carlo method. The albedo data used are in the form of empirical formulae based on the measured doubly differential albedo data. Sampling of the reflected polar and azimuthal angles is done by the rejection method. Variance reduction devices such as Russian Roulette are used. 3 - Restrictions on the complexity of the problem: - The albedo data and the subroutines for sampling the reflected polar and azimuthal angles are specific for concrete ducts. The maximum number of legs (as specified by dimension statements) is 5 and the maximum number of dose points is 50. The dose points considered are only in the last leg of the multi-legged duct

  15. Three-dimensional GRMHD Simulations of Neutrino-cooled Accretion Disks from Neutron Star Mergers

    Science.gov (United States)

    Siegel, Daniel M.; Metzger, Brian D.

    2018-05-01

    Merging binaries consisting of two neutron stars (NSs) or an NS and a stellar-mass black hole typically form a massive accretion torus around the remnant black hole or long-lived NS. Outflows from these neutrino-cooled accretion disks represent an important site for r-process nucleosynthesis and the generation of kilonovae. We present the first three-dimensional, general-relativistic magnetohydrodynamic (GRMHD) simulations including weak interactions and a realistic equation of state of such accretion disks over viscous timescales (380 ms). We witness the emergence of steady-state MHD turbulence, a magnetic dynamo with an ∼20 ms cycle, and the generation of a “hot” disk corona that launches powerful thermal outflows aided by the energy released as free nucleons recombine into α-particles. We identify a self-regulation mechanism that keeps the midplane electron fraction low (Y e ∼ 0.1) over viscous timescales. This neutron-rich reservoir, in turn, feeds outflows that retain a sufficiently low value of Y e ≈ 0.2 to robustly synthesize third-peak r-process elements. The quasi-spherical outflows are projected to unbind 40% of the initial disk mass with typical asymptotic escape velocities of 0.1c and may thus represent the dominant mass ejection mechanism in NS–NS mergers. Including neutrino absorption, our findings agree with previous hydrodynamical α-disk simulations that the entire range of r-process nuclei from the first to the third r-process peak can be synthesized in the outflows, in good agreement with observed solar system abundances. The asymptotic escape velocities and quantity of ejecta, when extrapolated to moderately higher disk masses, are consistent with those needed to explain the red kilonova emission following the NS merger GW170817.

  16. Analysis of neutronic parameters of AP1000 core for 18 month and 16/20 month cycle schemes using CASMO4E and SIMULATE-3 codes

    International Nuclear Information System (INIS)

    Nawaz Amjad; Yoshikawa, Hidekazu; Ming Yang

    2015-01-01

    AP1000 reactor is designed for 18 month of operating cycle. The core can also be used for 16/20 months of operating cycle. This study is performed to analyze and compare the neutronic parameters of typical AP1000 reactor core for 18 month and 16/20 month alternate cycle lengths. CASMO4E and SIMULATE-3 code package is used for the analysis of initial and equilibrium cores. The key reactor physics safety parameters were analyzed including power peaking factors, core radial and axial power distribution and core reactivity feedback coefficients. Moreover, the analysis of fuel depletion, fission product buildup and burnable poison behaviour with burnup is also analyzed. Full 2-D fuel assembly model in CASMO4E and full 3-D core model in SIMULATE-3 is employed to examine core performance and safety parameters. In order to evaluate the equilibrium core neutronic parameters, the equilibrium core model is attained by performing burnup analysis from initial to equilibrium cycle, where optimized transition core design is obtained so that the power peaking factors remain within designed limits. The MTC for higher concentration of critical boron concentrations is slightly positive at lower moderator temperatures. However, it remains negative at operating temperature ranges. The radial core relative power distribution indicates that low leakage capability of initial and equilibrium cores is reduced at EOC. (author)

  17. CORTAP: a coupled neutron kinetics-heat transfer digital computer program for the dynamic simulation of the high temperature gas cooled reactor core

    International Nuclear Information System (INIS)

    Cleveland, J.C.

    1977-01-01

    CORTAP (Core Transient Analysis Program) was developed to predict the dynamic behavior of the High Temperature Gas Cooled Reactor (HTGR) core under normal operational transients and postulated accident conditions. CORTAP is used both as a stand-alone component simulation and as part of the HTGR nuclear steam supply (NSS) system simulation code ORTAP. The core thermal neutronic response is determined by solving the heat transfer equations for the fuel, moderator and coolant in an average powered region of the reactor core. The space independent neutron kinetics equations are coupled to the heat transfer equations through a rapidly converging iterative technique. The code has the capability to determine conservative fuel, moderator, and coolant temperatures in the ''hot'' fuel region. For transients involving a reactor trip, the core heat generation rate is determined from an expression for decay heat following a scram. Nonlinear effects introduced by temperature dependent fuel, moderator, and coolant properties are included in the model. CORTAP predictions will be compared with dynamic test results obtained from the Fort St. Vrain reactor owned by Public Service of Colorado, and, based on these comparisons, appropriate improvements will be made in CORTAP

  18. Simulation of thermal-neutron-induced single-event upset using particle and heavy-ion transport code system

    International Nuclear Information System (INIS)

    Arita, Yutaka; Kihara, Yuji; Mitsuhasi, Junichi; Niita, Koji; Takai, Mikio; Ogawa, Izumi; Kishimoto, Tadafumi; Yoshihara, Tsutomu

    2007-01-01

    The simulation of a thermal-neutron-induced single-event upset (SEU) was performed on a 0.4-μm-design-rule 4 Mbit static random access memory (SRAM) using particle and heavy-ion transport code system (PHITS): The SEU rates obtained by the simulation were in very good agreement with the result of experiments. PHITS is a useful tool for simulating SEUs in semiconductor devices. To further improve the accuracy of the simulation, additional methods for tallying the energy deposition are required for PHITS. (author)

  19. A Preliminary Study on Detecting Fake Gold Bars Using Prompt Gamma Activation Analysis: Simulation of Neutron Transmission in Gold Bar

    International Nuclear Information System (INIS)

    Lee, K. M.; Sun, G. M.

    2016-01-01

    The purpose of this study is to develop fake gold bar detecting method by using Prompt-gamma activation analysis (PGAA) facility at the Korea Atomic Energy Research Institute (KAERI). PGAA is an established nuclear analytical technique for non-destructive determination of elemental and isotopic compositions. For a preliminary study on detecting fake gold bar, Monte Carlo simulation of neutron transmission in gold bar was conducted and the possibility for detecting fake gold bar was confirmed. Under the gold bullion standard, it guaranteed the government would redeem any amount of currency for its value in gold. After the gold bullion standard ended, gold bars have been the target for investment as ever. But it is well known that fake gold bar exist in the gold market. This cannot be identified easily without performing a testing as it has the same appearance as the pure gold bar. In order to avoid the trading of fake gold bar in the market, they should be monitored thoroughly. Although the transmissivity of cold neutrons are low comparing that of thermal neutrons, the slower neutrons are more apt to be absorbed in a target, and can increase the prompt gamma emission rate. Also the flux of both thermal and cold neutron beam is high enough to activate thick target. If the neutron beam is irradiated on the front and the reverse side of gold bar, all insides of it can be detected

  20. A Preliminary Study on Detecting Fake Gold Bars Using Prompt Gamma Activation Analysis: Simulation of Neutron Transmission in Gold Bar

    Energy Technology Data Exchange (ETDEWEB)

    Lee, K. M.; Sun, G. M. [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    The purpose of this study is to develop fake gold bar detecting method by using Prompt-gamma activation analysis (PGAA) facility at the Korea Atomic Energy Research Institute (KAERI). PGAA is an established nuclear analytical technique for non-destructive determination of elemental and isotopic compositions. For a preliminary study on detecting fake gold bar, Monte Carlo simulation of neutron transmission in gold bar was conducted and the possibility for detecting fake gold bar was confirmed. Under the gold bullion standard, it guaranteed the government would redeem any amount of currency for its value in gold. After the gold bullion standard ended, gold bars have been the target for investment as ever. But it is well known that fake gold bar exist in the gold market. This cannot be identified easily without performing a testing as it has the same appearance as the pure gold bar. In order to avoid the trading of fake gold bar in the market, they should be monitored thoroughly. Although the transmissivity of cold neutrons are low comparing that of thermal neutrons, the slower neutrons are more apt to be absorbed in a target, and can increase the prompt gamma emission rate. Also the flux of both thermal and cold neutron beam is high enough to activate thick target. If the neutron beam is irradiated on the front and the reverse side of gold bar, all insides of it can be detected.

  1. Development and verification of a high performance multi-group SP3 transport capability in the ARTEMIS core simulator

    International Nuclear Information System (INIS)

    Van Geemert, Rene

    2008-01-01

    For satisfaction of future global customer needs, dedicated efforts are being coordinated internationally and pursued continuously at AREVA NP. The currently ongoing CONVERGENCE project is committed to the development of the ARCADIA R next generation core simulation software package. ARCADIA R will be put to global use by all AREVA NP business regions, for the entire spectrum of core design processes, licensing computations and safety studies. As part of the currently ongoing trend towards more sophisticated neutronics methodologies, an SP 3 nodal transport concept has been developed for ARTEMIS which is the steady-state and transient core simulation part of ARCADIA R . For enabling a high computational performance, the SP N calculations are accelerated by applying multi-level coarse mesh re-balancing. In the current implementation, SP 3 is about 1.4 times as expensive computationally as SP 1 (diffusion). The developed SP 3 solution concept is foreseen as the future computational workhorse for many-group 3D pin-by-pin full core computations by ARCADIA R . With the entire numerical workload being highly parallelizable through domain decomposition techniques, associated CPU-time requirements that adhere to the efficiency needs in the nuclear industry can be expected to become feasible in the near future. The accuracy enhancement obtainable by using SP 3 instead of SP 1 has been verified by a detailed comparison of ARTEMIS 16-group pin-by-pin SP N results with KAERI's DeCart reference results for the 2D pin-by-pin Purdue UO 2 /MOX benchmark. This article presents the accuracy enhancement verification and quantifies the achieved ARTEMIS-SP 3 computational performance for a number of 2D and 3D multi-group and multi-box (up to pin-by-pin) core computations. (authors)

  2. Head-on collisions of binary white dwarf-neutron stars: Simulations in full general relativity

    International Nuclear Information System (INIS)

    Paschalidis, Vasileios; Etienne, Zachariah; Liu, Yuk Tung; Shapiro, Stuart L.

    2011-01-01

    We simulate head-on collisions from rest at large separation of binary white dwarf-neutron stars (WDNSs) in full general relativity. Our study serves as a prelude to our analysis of the circular binary WDNS problem. We focus on compact binaries whose total mass exceeds the maximum mass that a cold-degenerate star can support, and our goal is to determine the fate of such systems. A fully general relativistic hydrodynamic computation of a realistic WDNS head-on collision is prohibitive due to the large range of dynamical time scales and length scales involved. For this reason, we construct an equation of state (EOS) which captures the main physical features of neutron stars (NSs) while, at the same time, scales down the size of white dwarfs (WDs). We call these scaled-down WD models 'pseudo-WDs (pWDs)'. Using pWDs, we can study these systems via a sequence of simulations where the size of the pWD gradually increases toward the realistic case. We perform two sets of simulations; One set studies the effects of the NS mass on the final outcome, when the pWD is kept fixed. The other set studies the effect of the pWD compaction on the final outcome, when the pWD mass and the NS are kept fixed. All simulations show that after the collision, 14%-18% of the initial total rest mass escapes to infinity. All remnant masses still exceed the maximum rest mass that our cold EOS can support (1.92M · ), but no case leads to prompt collapse to a black hole. This outcome arises because the final configurations are hot. All cases settle into spherical, quasiequilibrium configurations consisting of a cold NS core surrounded by a hot mantle, resembling Thorne-Zytkow objects. Extrapolating our results to realistic WD compactions, we predict that the likely outcome of a head-on collision of a realistic, massive WDNS system will be the formation of a quasiequilibrium Thorne-Zytkow-like object.

  3. Quantitative comparison between experimental and simulated gamma-ray spectra induced by 14 MeV tagged neutrons

    Energy Technology Data Exchange (ETDEWEB)

    Perot, B., E-mail: bertrand.perot@cea.fr [CEA, DEN, Cadarache, Nuclear Measurement Laboratory, F-13108 Saint-Paul-lez-Durance (France); El Kanawati, W.; Carasco, C.; Eleon, C. [CEA, DEN, Cadarache, Nuclear Measurement Laboratory, F-13108 Saint-Paul-lez-Durance (France); Valkovic, V. [A.C.T.d.o.o., Prilesje 4, 10000 Zagreb (Croatia); Sudac, D.; Obhodas, J. [Ruder Boskovic Institute, Bijenicka c. 54, 10000 Zagreb (Croatia); Sannie, G. [CEA, LIST, Saclay, F-91191 Gif-sur-Yvette (France)

    2012-07-15

    Fast neutron interrogation with the associated particle technique can be used to identify explosives in cargo containers (EURITRACK FP6 project) and unexploded ordnance on the seabed (UNCOSS FP7 project), by detecting gamma radiations induced by 14 MeV neutrons produced in the {sup 2}H({sup 3}H,{alpha})n reaction. The origin of the gamma rays can be determined in 3D by the detection of the alpha particle, which provides the direction of the opposite neutron and its time-of-flight. Gamma spectroscopy provides the relative counts of carbon, nitrogen, and oxygen, which are converted to chemical fractions to differentiate explosives from other organic substances. To this aim, Monte Carlo calculations are used to take into account neutron moderation and gamma attenuation in cargo materials or seawater. This paper presents an experimental verification that C, N, and O counts are correctly reproduced by numerical simulation. A quantitative comparison is also reported for silicon, iron, lead, and aluminium. - Highlights: Black-Right-Pointing-Pointer Gamma-ray spectra produced by 14 MeV neutrons in C, N, O, Si, Al, Fe, and Pb elements. Black-Right-Pointing-Pointer Quantitative comparison with MCNPX simulations using the ENDF/B-VII.0 library. Black-Right-Pointing-Pointer C, N, and O counts correctly reproduced and chemical proportions recovered using calculation. Black-Right-Pointing-Pointer Application to the detection of explosives or illicit drugs in cargo containers.

  4. The structure of molecular liquids. Neutron diffraction and molecular dynamics simulations

    International Nuclear Information System (INIS)

    Bianchi, L.

    2000-05-01

    Neutron diffraction (ND) measurements on liquid methanol (CD 3 OD, CD 3 O(H/D), CD 3 OH) under ambient conditions were performed to obtain the distinct (intra- + inter-molecular), G dist (r) and inter-molecular, G inter (r) radial distribution functions (rdfs) for the three samples. The H/D substitution on hydroxyl-hydrogen (Ho) has been used to extract the partial distribution functions, G XHo (r) (X=C, O, and H - a methyl hydrogen) and G XX (r) at both the distinct and inter-molecular levels from the difference techniques of ND. The O-Ho bond length, which has been the subject of controversy in the past, is found purely from the distinct partial distribution function, G XHo (r) to be 0.98 ± 0.01 A. The C-H distance obtained from the distinct G XX (r) partial is 1.08 ± 0.01 A. These distances determined by fitting an intra-molecular model to the total distinct structure functions are 0.961 ± 0.001 A and 1.096 ± 0.001 A, respectively. The inter-molecular G XX (r) function, dominated by contributions from the methyl groups, apart from showing broad oscillations extending up to ∼14 A is featureless, mainly because of cancellation effects from six contributing pairs. The Ho-Ho partial pair distribution function (pdf), g HoHo (r), determined from the second order difference, shows that only one other Ho atom can be found within a mean Ho-Ho separation of 2.36 A. The average position of the O-Ho hydrogen bond determined for the first time purely from experimental inter-molecular G XHo (r) partial distribution function is found to be at 1.75 ± 0.03 A. The experimental structural results at the partial distribution level are compared with those obtained from molecular dynamics (MD) simulations performed in NVE ensemble by using both 3- and 6-site force field models for the first time in this study. The MD simulations with both the models reproduce the ND rdfs rather well. However, discrepancies begin to appear between the simulated and the experimental partial

  5. Neutronics simulations on hypothetical power excursion and possible core melt scenarios in CANDU6

    International Nuclear Information System (INIS)

    Kim, Yonghee

    2015-01-01

    LOCA (Loss of coolant accident) is an outstanding safety issue in the CANDU reactor system since the coolant void reactivity is strongly positive. To deal with the LOCA, the CANDU systems are equipped with specially designed quickly-acting secondary shutdown system. Nevertheless, the so-called design-extended conditions are requested to be taken into account in the safety analysis for nuclear reactor systems after the Fukushima accident. As a DEC scenario, the worst accident situation in a CANDU reactor system is a unprotected LOCA, which is supposed to lead to a power excursion and possibly a core melt-down. In this work, the hypothetical unprotected LOCA scenario is simulated in view of the power excursion and fuel temperature changes by using a simplified point-kinetics (PK) model accounting for the fuel temperature change. In the PK model, the core reactivity is assumed to be affected by a large break LOCA and the fuel temperature is simulated to account for the Doppler effect. In addition, unlike the conventional PK simulation, we have also considered the Xe-I model to evaluate the impact of Xe during the LOCA. Also, we tried to simulate the fuel and core melt-down scenario in terms of the reactivity through a series of neutronics calculations for hypothetical core conditions. In case of a power excursion and possible fuel melt-down situation, the reactor system behavior is very uncertain. In this work, we tried to understand the impacts of fuel melt and relocation within the pressure vessel on the core reactivity and failure of pressure and calandria tubes. (author)

  6. CATHARE2 V1.4 capability to simulate the performance of isolation condenser systems with thermal valve

    International Nuclear Information System (INIS)

    Meloni, P.

    2001-01-01

    ENEA (Italy) in co-operation with CEA (France) has carried out an R and D activity aimed at increasing the reliability of Decay Heat Removal (DHR) passive systems that implement in-pool heat exchangers. The main outcome reached was the definition of a device, called Thermal Valve (TV), able to avoid the installation of mechanical valve on the primary circuit, thus reducing thermalmechanical constrains and thermal-hydraulic instabilities. This paper presents a preliminary assessment performed with CATHARE of this innovative device. In the first part the code capability to simulate in-pool heat exchangers is verified against experimental data of the PANDA facility, that are available within the frame of the ISP 42. In the second part a CATHARE calculation showing the performances of the PANDA passive condenser with TV (start-up and shutdown) is described.(author)

  7. Development of Experimental Icing Simulation Capability for Full-Scale Swept Wings: Hybrid Design Process, Years 1 and 2

    Science.gov (United States)

    Fujiwara, Gustavo; Bragg, Mike; Triphahn, Chris; Wiberg, Brock; Woodard, Brian; Loth, Eric; Malone, Adam; Paul, Bernard; Pitera, David; Wilcox, Pete; hide

    2017-01-01

    This report presents the key results from the first two years of a program to develop experimental icing simulation capabilities for full-scale swept wings. This investigation was undertaken as a part of a larger collaborative research effort on ice accretion and aerodynamics for large-scale swept wings. Ice accretion and the resulting aerodynamic effect on large-scale swept wings presents a significant airplane design and certification challenge to air frame manufacturers, certification authorities, and research organizations alike. While the effect of ice accretion on straight wings has been studied in detail for many years, the available data on swept-wing icing are much more limited, especially for larger scales.

  8. CHIPS_TPT models for exclusive Geant4 simulation of neutron-nuclear reactions at low energies

    Directory of Open Access Journals (Sweden)

    Kosov Mikhail V.

    2014-03-01

    Full Text Available A novel TPT code (Toolkit for Particle Transport, which is included in CHIPS_TPT physics list for Geant4 simulations, is briefly overviewed. Underlying concept of exclusive modelling is introduced and its beneficial features are illustrated with several examples. Widely used neutron Monte Carlo codes, MCNP and Geant4/HP, are based on inclusive algorithms that independently model neutron state change and secondary particles production while tracking. The exclusive approach implemented in TPT overcomes this unphysical separation and makes it possible to allow for kinematic restrictions as well as correlated emission of gamma-rays and secondaries.

  9. Predictive capability of average Stokes polarimetry for simulation of phase multilevel elements onto LCoS devices.

    Science.gov (United States)

    Martínez, Francisco J; Márquez, Andrés; Gallego, Sergi; Ortuño, Manuel; Francés, Jorge; Pascual, Inmaculada; Beléndez, Augusto

    2015-02-20

    Parallel-aligned (PA) liquid-crystal on silicon (LCoS) microdisplays are especially appealing in a wide range of spatial light modulation applications since they enable phase-only operation. Recently we proposed a novel polarimetric method, based on Stokes polarimetry, enabling the characterization of their linear retardance and the magnitude of their associated phase fluctuations or flicker, exhibited by many LCoS devices. In this work we apply the calibrated values obtained with this technique to show their capability to predict the performance of spatially varying phase multilevel elements displayed onto the PA-LCoS device. Specifically we address a series of multilevel phase blazed gratings. We analyze both their average diffraction efficiency ("static" analysis) and its associated time fluctuation ("dynamic" analysis). Two different electrical configuration files with different degrees of flicker are applied in order to evaluate the actual influence of flicker on the expected performance of the diffractive optical elements addressed. We obtain a good agreement between simulation and experiment, thus demonstrating the predictive capability of the calibration provided by the average Stokes polarimetric technique. Additionally, it is obtained that for electrical configurations with less than 30° amplitude for the flicker retardance, they may not influence the performance of the blazed gratings. In general, we demonstrate that the influence of flicker greatly diminishes when the number of quantization levels in the optical element increases.

  10. Use of a Boron Doped Spherical Phantom for the Investigation of Neutron Directional Properties: Comparison Between Experiment and MCNP Simulation

    Energy Technology Data Exchange (ETDEWEB)

    Drake, P.; Kierkegaard, J

    1999-07-01

    A boron doped 19 cm diameter spherical phantom was constructed to give information on the direction of neutrons inside the Ringhals 4 containment. The phantom was made of 40% paraffin and 60% boric acid. 10B contributes 2% of the total phantom weight. The phantom was tested for its angular sensitivity to neutrons. The response was tested with a {sup 252}Cf source and with a Monte Carlo calculation (MCNP) simulating a {sup 252}Cf source. In these investigations the phantom showed a strong directional response. However, there was only a fair correspondence between the experiment and the simulation. The discrepancies are, at least in part, due to the difference in energy and angular response of the dosemeters as compared with the idealised response characteristics in the MCNP calculation. In the MCNP calculation the experimental conditions were not fully simulated. The investigations also showed that the addition of boron to the phantom reduces the leakage of thermalised neutrons from the phantom, and the production of neutron induced photons in the phantom to insignificant levels. (author)

  11. Use of a Boron Doped Spherical Phantom for the Investigation of Neutron Directional Properties: Comparison Between Experiment and MCNP Simulation

    International Nuclear Information System (INIS)

    Drake, P.; Kierkegaard, J.

    1999-01-01

    A boron doped 19 cm diameter spherical phantom was constructed to give information on the direction of neutrons inside the Ringhals 4 containment. The phantom was made of 40% paraffin and 60% boric acid. 10B contributes 2% of the total phantom weight. The phantom was tested for its angular sensitivity to neutrons. The response was tested with a 252 Cf source and with a Monte Carlo calculation (MCNP) simulating a 252 Cf source. In these investigations the phantom showed a strong directional response. However, there was only a fair correspondence between the experiment and the simulation. The discrepancies are, at least in part, due to the difference in energy and angular response of the dosemeters as compared with the idealised response characteristics in the MCNP calculation. In the MCNP calculation the experimental conditions were not fully simulated. The investigations also showed that the addition of boron to the phantom reduces the leakage of thermalised neutrons from the phantom, and the production of neutron induced photons in the phantom to insignificant levels. (author)

  12. Neutron spectrum unfolding using genetic algorithm in a Monte Carlo simulation

    Energy Technology Data Exchange (ETDEWEB)

    Suman, Vitisha [Health Physics Division, Bhabha Atomic Research Centre, Mumbai 400085 (India); Sarkar, P.K., E-mail: pksarkar02@gmail.com [Manipal Centre for Natural Sciences, Manipal University, Manipal 576104 (India)

    2014-02-11

    A spectrum unfolding technique GAMCD (Genetic Algorithm and Monte Carlo based spectrum Deconvolution) has been developed using the genetic algorithm methodology within the framework of Monte Carlo simulations. Each Monte Carlo history starts with initial solution vectors (population) as randomly generated points in the hyper dimensional solution space that are related to the measured data by the response matrix of the detection system. The transition of the solution points in the solution space from one generation to another are governed by the genetic algorithm methodology using the techniques of cross-over (mating) and mutation in a probabilistic manner adding new solution points to the population. The population size is kept constant by discarding solutions having lesser fitness values (larger differences between measured and calculated results). Solutions having the highest fitness value at the end of each Monte Carlo history are averaged over all histories to obtain the final spectral solution. The present method shows promising results in neutron spectrum unfolding for both under-determined and over-determined problems with simulated test data as well as measured data when compared with some existing unfolding codes. An attractive advantage of the present method is the independence of the final spectra from the initial guess spectra.

  13. Neutron spectrum unfolding using genetic algorithm in a Monte Carlo simulation

    International Nuclear Information System (INIS)

    Suman, Vitisha; Sarkar, P.K.

    2014-01-01

    A spectrum unfolding technique GAMCD (Genetic Algorithm and Monte Carlo based spectrum Deconvolution) has been developed using the genetic algorithm methodology within the framework of Monte Carlo simulations. Each Monte Carlo history starts with initial solution vectors (population) as randomly generated points in the hyper dimensional solution space that are related to the measured data by the response matrix of the detection system. The transition of the solution points in the solution space from one generation to another are governed by the genetic algorithm methodology using the techniques of cross-over (mating) and mutation in a probabilistic manner adding new solution points to the population. The population size is kept constant by discarding solutions having lesser fitness values (larger differences between measured and calculated results). Solutions having the highest fitness value at the end of each Monte Carlo history are averaged over all histories to obtain the final spectral solution. The present method shows promising results in neutron spectrum unfolding for both under-determined and over-determined problems with simulated test data as well as measured data when compared with some existing unfolding codes. An attractive advantage of the present method is the independence of the final spectra from the initial guess spectra

  14. Aggregation in concentrated protein solutions: Insights from rheology, neutron scattering and molecular simulations

    Science.gov (United States)

    Castellanos, Maria Monica

    Aggregation of therapeutic proteins is currently one of the major challenges in the bio-pharmaceutical industry, because aggregates could induce immunogenic responses and compromise the quality of the product. Current scientific efforts, both in industry and academia, are focused on developing rational approaches to screen different drug candidates and predict their stability under different conditions. Moreover, aggregation is promoted in highly concentrated protein solutions, which are typically required for subcutaneous injection. In order to gain further understanding about the mechanisms that lead to aggregation, an approach that combined rheology, neutron scattering, and molecular simulations was undertaken. Two model systems were studied in this work: Bovine Serum Albumin in surfactant-free Phosphate Buffered Saline at pH = 7.4 at concentrations from 11 mg/mL up to ˜519 mg/mL, and a monoclonal antibody in 20 mM Histidine/Histidine Hydrochloride at pH = 6.0 with 60 mg/mL trehalose and 0.2 mg/mL polysorbate-80 at concentrations from 53 mg/mL up to ˜220 mg/mL. The antibody used here has three mutations in the CH2 domain, which result in lower stability upon incubation at 40 °C with respect to the wild-type protein, based on size-exclusion chromatography assays. This temperature is below 49 °C, where unfolding of the least stable, CH2 domain occurs, according to differential scanning calorimetry. This dissertation focuses on identifying the role of aggregation on the viscosity of protein solutions. The protein solutions of this work show an increase in the low shear viscosity in the absence of surfactants, because proteins adsorb at the air/water interface forming a viscoelastic film that affects the measured rheology. Stable surfactant-laden protein solutions behave as simple Newtonian fluids. However, the surfactant-laden antibody solution also shows an increase in the low shear viscosity from bulk aggregation, after prolonged incubation at 40 °C. Small

  15. ATHENA2D, Simulation Hypothetical Recriticality Accident in a Thermal Neutron Spectrum

    International Nuclear Information System (INIS)

    1999-01-01

    1 - Description of program or function: ATHENA 2 D was written to simulate a hypothetical water reflood of a highly-damaged light water reactor (such as the Three-Mile-Island Unit-2 after meltdown, with a packed debris bed near the center of the core), but with insufficiently-borated reflood water. A recriticality transient may result because of the potentially more reactive debris bed. ATHENA-2D solves the transient multigroup neutron diffusion equations in (r,z) geometry. Executing in parallel with the transient neutronics, is a single-phase computational fluid dynamics (CFD) model, driven by multichannel thermal hydraulics based on detailed pin models. Numerous PV-Wave procedure files are included on the distribution media, useful for those who already have PV-Wave from Visual Numerics. These procedures are documented in the 'README' files included on the distribution CD. Some reactor lattice computer code such as WIMS-E, CCC-576/WIMSD4, or CCC-656/WIMSD5B is required for the creation of macroscopic cross section libraries, given pin-cell geometries. WIMS-E is a commercial product available from AEA Technologies, England, WIMS is not included on the ATHENA 2 D distribution CD. Several auxiliary routines are included in the package. TFMAX: Utility that searches through ATHENA 2 D binary output to find the maximum fuel temperature over space and time. POST V EL: Utility that searches through ATHENA 2 D binary output to find maximum scalar and flow field values (over space) and outputs normalization factors as a function of time. These results are used to correctly scale animations. CONVT: If executing ATHENA 2 D on a PC under Windows, this utility converts one form of binary output (directly from ATHENA 2 D) to another, which is readable by PV-Wave for Windows (PV-Wave is data animation and visualization software from Visual Numerics, Inc.) CALC M TX: Post-processing utility for calculating the model coefficients for the calculation matrix. 2 - Methods: Both the

  16. Experimental and simulation studies of neutron-induced single-event burnout in SiC power diodes

    Science.gov (United States)

    Shoji, Tomoyuki; Nishida, Shuichi; Hamada, Kimimori; Tadano, Hiroshi

    2014-01-01

    Neutron-induced single-event burnouts (SEBs) of silicon carbide (SiC) power diodes have been investigated by white neutron irradiation experiments and transient device simulations. It was confirmed that a rapid increase in lattice temperature leads to formation of crown-shaped aluminum and cracks inside the device owing to expansion stress when the maximum lattice temperature reaches the sublimation temperature. SEB device simulation indicated that the peak lattice temperature is located in the vicinity of the n-/n+ interface and anode contact, and that the positions correspond to a hammock-like electric field distribution caused by the space charge effect. Moreover, the locations of the simulated peak lattice temperature agree closely with the positions of the observed destruction traces. Furthermore, it was theoretically demonstrated that the period of temperature increase of a SiC power device is two orders of magnitude less than that of a Si power device, using a thermal diffusion equation.

  17. Development of an anthropomorfic simulator for simulation and measurements of neutron dose and flux the facility for BNCT studies

    International Nuclear Information System (INIS)

    Muniz, Rafael Oliveira Rondon

    2010-01-01

    IPEN facility for researches in BNCT (Boron Neutron Capture Therapy) uses IEA-R1 reactor's irradiation channel number 3, where there is a mixed radiation field - neutrons and gamma. The researches in progress require the radiation fields, in the position of the irradiation of sample, to have in its composition maximized thermal neutrons component and minimized, fast and epithermal neutron flux and gamma radiation. This work was developed with the objective of evaluating whether the present radiation field in the facility is suitable for BNCT researches. In order to achieve this objective, a methodology for the dosimetry of thermal neutrons and gamma radiation in mixed fields of high doses, which was not available in IPEN, was implemented in the Center of Nuclear Engineering of IPEN, by using thermoluminescent dosimeters - TLDs 400, 600 and 700. For the measurements of thermal and epithermal neutron flux, activation detectors of gold were used applying the cadmium ratio technique. A cylindrical phantom composed by acrylic discs was developed and tested in the facility and the DOT 3.5. computational code was used in order to obtain theoretical values of neutron flux and the dose along phantom. In the position corresponding to about half the length of the cylinder of the phantom, the following values were obtained: thermal neutron flux (2,52 ± 0,06).10 8 n/cm 2 s, epithermal neutron flux (6,17 ± 0,26).10 7 .10 6 n/cm 2 s, absorbed dose due to thermal neutrons (4,2 ± 1,8)Gy and (10,1 ± 1,3)Gy due to gamma radiation. The obtained values show that the fluxes of thermal and epithermal neutrons flux are appropriate for studies in BNCT, however, the dose due to gamma radiation is high, indicating that the facility should be improved. (author)

  18. BWR modeling capability and Scale/Triton lattice-to-core integration of the Nestle nodal simulator - 331

    International Nuclear Information System (INIS)

    Galloway, J.; Hernandez, H.; Maldonado, G.I.; Jessee, M.; Popov, E.; Clarno, K.

    2010-01-01

    This article reports the status of recent and substantial enhancements made to the NESTLE nodal core simulator, a code originally developed at North Carolina State University (NCSU) of which version 5.2.1 has been available for several years through the Oak Ridge National Laboratory (ORNL) Radiation Safety Information Computational Center (RSICC) software repository. In its released and available form, NESTLE is a seasoned, well-developed and extensively tested code system particularly useful to model PWRs. In collaboration with NCSU, University of Tennessee (UT) and ORNL researchers have recently developed new enhancements for the NESTLE code, including the implementation of a two-phase drift-flux thermal hydraulic and flow redistribution model to facilitate modeling of Boiling Water Reactors (BWRs) as well as the development of an integrated coupling of SCALE/TRITON lattice physics to NESTLE so to produce an end-to-end capability for reactor simulations. These latest advancements implemented into NESTLE as well as an update of other ongoing efforts of this project are herein reported. (authors)

  19. Simulated production rates of exotic nuclei from the ion guide for neutron-induced fission at IGISOL

    Energy Technology Data Exchange (ETDEWEB)

    Jansson, Kaj; Al-Adili, Ali; Nilsson, Nicklas; Norlin, Martin; Solders, Andreas [Uppsala University, Department of Physics and Astronomy, Uppsala (Sweden)

    2017-12-15

    An investigation of the stopping efficiency of fission products, in the new ion guide designed for ion production through neutron-induced fission at IGISOL in Jyvaeskylae, Finland, has been conducted. Our simulations take into account the new neutron converter, enabling measurements of neutron-induced fission yields, and thereby provide estimates of the obtained yields as a function of primary proton beam current. Different geometries, targets, and pressures, as well as models for the effective charge of the stopped ions were tested, and optimisations to the setup for higher yields are suggested. The predicted number of ions stopped in the gas lets us estimate the survival probability of the ions reaching the downstream measurements stations. (orig.)

  20. Activation neutron detector

    International Nuclear Information System (INIS)

    Ambardanishvili, T.S.; Kolomiitsev, M.A.; Zakharina, T.Y.; Dundua, V.J.; Chikhladze, N.V.

    1976-01-01

    An activation neutron detector made as a moulded and cured composition of a material capable of being neutron-activated is described. The material is selected from a group consisting of at least two chemical elements, a compound of at least two chemical elements and their mixture, each of the chemical elements and their mixture, each of the chemical elements being capable of interacting with neutrons to form radioactive isotopes having different radiation energies when disintegrating. The material capable of being neutron-activated is distributed throughout the volume of a polycondensation resin inert with respect to neutrons and capable of curing. 17 Claims, No Drawings

  1. Validation of MCNP NPP Activation Simulations for Decommissioning Studies by Analysis of NPP Neutron Activation Foil Measurement Campaigns

    Directory of Open Access Journals (Sweden)

    Volmert Ben

    2016-01-01

    Full Text Available In this paper, an overview of the Swiss Nuclear Power Plant (NPP activation methodology is presented and the work towards its validation by in-situ NPP foil irradiation campaigns is outlined. Nuclear Research and consultancy Group (NRG in The Netherlands has been given the task of performing the corresponding neutron metrology. For this purpose, small Aluminium boxes containing a set of circular-shaped neutron activation foils have been prepared. After being irradiated for one complete reactor cycle, the sets have been successfully retrieved, followed by gamma-spectrometric measurements of the individual foils at NRG. Along with the individual activities of the foils, the reaction rates and thermal, intermediate and fast neutron fluence rates at the foil locations have been determined. These determinations include appropriate corrections for gamma self-absorption and neutron self-shielding as well as corresponding measurement uncertainties. The comparison of the NPP Monte Carlo calculations with the results of the foil measurements is done by using an individual generic MCNP model functioning as an interface and allowing the simulation of individual foil activation by predetermined neutron spectra. To summarize, the comparison between calculation and measurement serve as a sound validation of the Swiss NPP activation methodology by demonstrating a satisfying agreement between measurement and calculation. Finally, the validation offers a chance for further improvements of the existing NPP models by ensuing calibration and/or modelling optimizations for key components and structures.

  2. Signatures of asymmetry in neutron spectra and images predicted by three-dimensional radiation hydrodynamics simulations of indirect drive implosions

    Energy Technology Data Exchange (ETDEWEB)

    Chittenden, J. P., E-mail: j.chittenden@imperial.ac.uk; Appelbe, B. D.; Manke, F.; McGlinchey, K.; Niasse, N. P. L. [Centre for Inertial Fusion Studies, The Blackett Laboratory, Imperial College, London SW7 2AZ (United Kingdom)

    2016-05-15

    We present the results of 3D simulations of indirect drive inertial confinement fusion capsules driven by the “high-foot” radiation pulse on the National Ignition Facility. The results are post-processed using a semi-deterministic ray tracing model to generate synthetic deuterium-tritium (DT) and deuterium-deuterium (DD) neutron spectra as well as primary and down scattered neutron images. Results with low-mode asymmetries are used to estimate the magnitude of anisotropy in the neutron spectra shift, width, and shape. Comparisons of primary and down scattered images highlight the lack of alignment between the neutron sources, scatter sites, and detector plane, which limits the ability to infer the ρr of the fuel from a down scattered ratio. Further calculations use high bandwidth multi-mode perturbations to induce multiple short scale length flows in the hotspot. The results indicate that the effect of fluid velocity is to produce a DT neutron spectrum with an apparently higher temperature than that inferred from the DD spectrum and which is also higher than the temperature implied by the DT to DD yield ratio.

  3. The sensitivity studies of a landmine explosive detection system based on neutron backscattering using Monte Carlo simulation

    Directory of Open Access Journals (Sweden)

    Khan Hamda

    2017-01-01

    Full Text Available This paper carries out a Monte Carlo simulation of a landmine detection system, using the MCNP5 code, for the detection of concealed explosives such as trinitrotoluene and cyclonite. In portable field detectors, the signal strength of backscattered neutrons and gamma rays from thermal neutron activation is sensitive to a number of parameters such as the mass of explosive, depth of concealment, neutron moderation, background soil composition, soil porosity, soil moisture, multiple scattering in the background material, and configuration of the detection system. In this work, a detection system, with BF3 detectors for neutrons and sodium iodide scintillator for g-rays, is modeled to investigate the neutron signal-to-noise ratio and to obtain an empirical formula for the photon production rate Ri(n,γ= SfGfMf(d,m from radiative capture reactions in constituent nuclides of trinitrotoluene. This formula can be used for the efficient landmine detection of explosives in quantities as small as ~200 g of trinitrotoluene concealed at depths down to about 15 cm. The empirical formula can be embedded in a field programmable gate array on a field-portable explosives' sensor for efficient online detection.

  4. Quasielastic neutron scattering measurements and ab initio MD-simulations on single ion motions in molten NaF

    Energy Technology Data Exchange (ETDEWEB)

    Demmel, F. [ISIS Facility, Rutherford Appleton Laboratory, Didcot OX11 0QX (United Kingdom); Mukhopadhyay, S. [ISIS Facility, Rutherford Appleton Laboratory, Didcot OX11 0QX (United Kingdom); Department of Materials, Imperial College London, Exhibition Road, London SW7 2AZ (United Kingdom)

    2016-01-07

    The ionic stochastic motions in the molten alkali halide NaF are investigated by quasielastic neutron scattering and first principles molecular dynamics simulation. Quasielastic neutron scattering was employed to extract the diffusion behavior of the sodium ions in the melt. An extensive first principles based simulation on a box of up to 512 particles has been performed to complement the experimental data. From that large box, a smaller 64-particle box has then been simulated over a runtime of 60 ps. A good agreement between calculated and neutron data on the level of spectral shape has been obtained. The obtained sodium diffusion coefficients agree very well. The simulation predicts a fluorine diffusion coefficient similar to the sodium one. Applying the Nernst-Einstein equation, a remarkable large cross correlation between both ions can be deduced. The velocity cross correlations demonstrate a positive correlation between the ions over a period of 0.1 ps. That strong correlation is evidence that the unlike ions do not move completely statistically independent and have a strong association over a short period of time.

  5. Simulation and optimization of a new focusing polarizing bender for the diffuse neutrons scattering spectrometer DNS at MLZ

    International Nuclear Information System (INIS)

    Nemkovski, K; Ioffe, A; Su, Y; Babcock, E; Schweika, W; Brückel, Th

    2017-01-01

    We present the concept and the results of the simulations of a new polarizer for the diffuse neutron scattering spectrometer DNS at MLZ. The concept of the polarizer is based on the idea of a bender made from the stack of the silicon wafers with a double-side supermirror polarizing coating and absorbing spacers in between. Owing to its compact design, such a system provides more free space for the arrangement of other instrument components. To reduce activation of the polarizer in the high intensity neutron beam of the DNS spectrometer we plan to use the Fe/Si supermirrors instead of currently used FeCoV/Ti:N ones. Using the VITESS simulation package we have performed simulations for horizontally focusing polarizing benders with different geometries in the combination with the double-focusing crystal monochromator of DNS. Neutron transmission and polarization efficiency as well as the effects of the focusing for convergent conventional C-benders and S-benders have been analyzed both for wedge-like and plane-parallel convergent geometries of the channels. The results of these simulations and the advantages/disadvantages of the various configurations are discussed. (paper)

  6. Simulation and optimization of a new focusing polarizing bender for the diffuse neutrons scattering spectrometer DNS at MLZ

    Science.gov (United States)

    Nemkovski, K.; Ioffe, A.; Su, Y.; Babcock, E.; Schweika, W.; Brückel, Th

    2017-06-01

    We present the concept and the results of the simulations of a new polarizer for the diffuse neutron scattering spectrometer DNS at MLZ. The concept of the polarizer is based on the idea of a bender made from the stack of the silicon wafers with a double-side supermirror polarizing coating and absorbing spacers in between. Owing to its compact design, such a system provides more free space for the arrangement of other instrument components. To reduce activation of the polarizer in the high intensity neutron beam of the DNS spectrometer we plan to use the Fe/Si supermirrors instead of currently used FeCoV/Ti:N ones. Using the VITESS simulation package we have performed simulations for horizontally focusing polarizing benders with different geometries in the combination with the double-focusing crystal monochromator of DNS. Neutron transmission and polarization efficiency as well as the effects of the focusing for convergent conventional C-benders and S-benders have been analyzed both for wedge-like and plane-parallel convergent geometries of the channels. The results of these simulations and the advantages/disadvantages of the various configurations are discussed.

  7. Cellular neural networks (CNN) simulation for the TN approximation of the time dependent neutron transport equation in slab geometry

    International Nuclear Information System (INIS)

    Hadad, Kamal; Pirouzmand, Ahmad; Ayoobian, Navid

    2008-01-01

    This paper describes the application of a multilayer cellular neural network (CNN) to model and solve the time dependent one-speed neutron transport equation in slab geometry. We use a neutron angular flux in terms of the Chebyshev polynomials (T N ) of the first kind and then we attempt to implement the equations in an equivalent electrical circuit. We apply this equivalent circuit to analyze the T N moments equation in a uniform finite slab using Marshak type vacuum boundary condition. The validity of the CNN results is evaluated with numerical solution of the steady state T N moments equations by MATLAB. Steady state, as well as transient simulations, shows a very good comparison between the two methods. We used our CNN model to simulate space-time response of total flux and its moments for various c (where c is the mean number of secondary neutrons per collision). The complete algorithm could be implemented using very large-scale integrated circuit (VLSI) circuitry. The efficiency of the calculation method makes it useful for neutron transport calculations

  8. Static feedback model for neutronic and thermodynamic simulation of fast reactors

    International Nuclear Information System (INIS)

    Waintraub, M.; Jachic, J.

    1985-01-01

    It is analysed the variation of the microscopic cross sections with neutronic spectra and temperature of materials for reactors such as SUPER-PHENIX. It was realized a parametric study of each spectral component, where the influence of each isotope was analysed separately. To include the Doppler effect and other important effects, neutronic and thermodynamic calculations in an iterative form were done allowing to determine neutron temperatures for fuel, structural material and coolant. (M.C.K.) [pt

  9. Three dimensional neutronic/thermal-hydraulic coupled simulation of MSR in transient state condition

    International Nuclear Information System (INIS)

    Zhou, Jianjun; Zhang, Daling; Qiu, Suizheng; Su, Guanghui; Tian, Wenxi; Wu, Yingwei

    2015-01-01

    Highlights: • Developed a three dimensional neutronic/thermal-hydraulic coupled transient analysis code for MSR. • Investigated the neutron distribution and thermal-hydraulic characters of the core under transient condition. • Analyzed three different transient conditions of inlet temperature drop, reactivity jump and pump coastdown. - Abstract: MSR (molten salt reactor) use liquid molten salt as coolant and fuel solvent, which was the only one liquid reactor of six Generation IV reactor types. As a liquid reactor the physical property of reactor was significantly influenced by fuel salt flow and the conventional analysis methods applied in solid fuel reactors are not applicable for this type of reactors. The present work developed a three dimensional neutronic/thermal-hydraulic coupled code investigated the neutronics and thermo-hydraulics characteristics of the core in transient condition based on neutron diffusion theory and numerical heat transfer. The code consists of two group neutron diffusion equations for fast and thermal neutron fluxes and six group balance equations for delayed neutron precursors. The code was separately validated by neutron benchmark and flow and heat transfer benchmark. Three different transient conditions was analyzed with inlet temperature drop, reactivity jump and pump coastdown. The results provide some valuable information in design and research this kind of reactor

  10. Three dimensional neutronic/thermal-hydraulic coupled simulation of MSR in transient state condition

    Energy Technology Data Exchange (ETDEWEB)

    Zhou, Jianjun [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xianning Road, 28, Xi’an 710049, Shaanxi (China); College of Mechanical and Power Engineering, China Three Gorges University, No 8, Daxue road, Yichang, Hubei 443002 (China); Zhang, Daling, E-mail: dlzhang@mail.xjtu.edu.cn [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xianning Road, 28, Xi’an 710049, Shaanxi (China); Qiu, Suizheng; Su, Guanghui; Tian, Wenxi; Wu, Yingwei [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xianning Road, 28, Xi’an 710049, Shaanxi (China)

    2015-02-15

    Highlights: • Developed a three dimensional neutronic/thermal-hydraulic coupled transient analysis code for MSR. • Investigated the neutron distribution and thermal-hydraulic characters of the core under transient condition. • Analyzed three different transient conditions of inlet temperature drop, reactivity jump and pump coastdown. - Abstract: MSR (molten salt reactor) use liquid molten salt as coolant and fuel solvent, which was the only one liquid reactor of six Generation IV reactor types. As a liquid reactor the physical property of reactor was significantly influenced by fuel salt flow and the conventional analysis methods applied in solid fuel reactors are not applicable for this type of reactors. The present work developed a three dimensional neutronic/thermal-hydraulic coupled code investigated the neutronics and thermo-hydraulics characteristics of the core in transient condition based on neutron diffusion theory and numerical heat transfer. The code consists of two group neutron diffusion equations for fast and thermal neutron fluxes and six group balance equations for delayed neutron precursors. The code was separately validated by neutron benchmark and flow and heat transfer benchmark. Three different transient conditions was analyzed with inlet temperature drop, reactivity jump and pump coastdown. The results provide some valuable information in design and research this kind of reactor.

  11. Simulation and analysis of the transmission properties of curved-straight neutron guide systems

    International Nuclear Information System (INIS)

    Copley, J.R.D.; Mildner, D.F.R.

    1992-01-01

    This paper reports that the spatial intensity distribution of neutrons emerging from a curved guide is far from uniform, particularly at short wavelengths, and curved guides are sometimes followed by a straight section of guide to make the intensity distribution more uniform. The behavior of neutrons within curved-straight neutron guide systems is examined using both ray-tracing and analytical approaches to the problem. The intensity distribution within the straight guide tends to wash from one side of the guide to the other. The amplitude of this transverse wave decreases with increasing guide length, and the characteristic length of the wave decreases with increasing neutron wavelength

  12. Simulations of the neutron energy-spectra at the Olympus Gate Environmental Monitoring Station due to historical Bevatron operations

    International Nuclear Information System (INIS)

    Donahue, R.J.; Thomas, R.H.; Zeman, G.H.

    2001-01-01

    Offsite neutron fluences resulting from Bevatron operations reached a maximum in 1959, prior to the addition of a permanent concrete roof shield, which was constructed in 1962. From the first operation of the Bevatron measurements of neutron fluence were made at locations around the perimeter of the Lawrence Berkeley National Laboratory (LBNL) campus. Since the late 1950's measurements made at several locations, and particularly at the site of what is now called the Olympus Gate Environmental Monitoring Station, have been routinely reported and published. Early measurements were used to establish the shape of the neutron-energy spectrum from which an energy-averaged fluence-to-dose equivalent conversion coefficient could be derived. This conversion coefficient was then applied to a measured total neutron fluence to obtain the appropriate dose equivalent quantity required by regulation. Recent work by Thomas et al. (2000) have compared the early conversion coefficients used in the sixties with those accepted today and suggest suggested that ''the dose equivalents reported in the late fifties and early sixties were conservative by factors between two and four. In any current review of the historical data, therefore it would be prudent to reduce the reported dose equivalents by at least a factor of two.'' However, that analysis was based on the ''state of the art'' neutron energy-spectra of the '60s. This paper provides a detailed knowledge of the neutron energy spectrum at the site boundary paper thus removing any uncertainty in the analysis of Thomas et al., which might be caused by the use of the early neutron energy-spectra. Detailed Monte Carlo analyses of the interactions of 6.2 GeV protons in thick, medium-A targets are described. In the computer simulations, neutrons produced were allowed to scatter in the atmosphere. Detailed neutron energy spectra were calculated at a distance and elevation corresponding to the location of the Olympus Gate EMS. Both older

  13. Neutronic Analysis of the 3 MW TRIGA MARK II Research Reactor, Part I: Monte Carlo Simulation

    International Nuclear Information System (INIS)

    Huda, M.Q.; Chakrobortty, T.K.; Rahman, M.; Sarker, M.M.; Mahmood, M.S.

    2003-05-01

    This study deals with the neutronic analysis of the current core configuration of a 3 MW TRIGA MARK II research reactor at Atomic Energy Research Establishment (AERE), Savar, Dhaka, Bangladesh and validation of the results by benchmarking with the experimental, operational and available Final Safety Analysis Report (FSAR) values. The three-dimensional continuous-energy Monte Carlo code MCNP4C was used to develop a versatile and accurate full-core model of the TRIGA core. The model represents in detail all components of the core with literally no physical approximation. All fresh fuel and control elements as well as the vicinity of the core were precisely described. Continuous energy cross-section data from ENDF/B-VI and S(α, β) scattering functions from the ENDF/B-V library were used. The validation of the model against benchmark experimental results is presented. The MCNP predictions and the experimentally determined values are found to be in very good agreement, which indicates that the Monte Carlo model is correctly simulating the TRIGA reactor. (author)

  14. Neutronics Phenomena Important in Modeling and Simulation of Liquid-Fuel Molten Salt Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Diamond, David J.

    2018-11-11

    This paper discusses liquid-fuel molten salt reactors, how they will operate under normal, transient, and accident conditions, and the results of an expert elicitation to determine the corresponding neutronic phenomena important to understanding their behavior. Identifying these phenomena will enable the U.S. Nuclear Regulatory Commission (NRC) to develop or identify modeling functionalities and tools required to carry out confirmatory analyses that examine the validity and accuracy of applicants’ calculations and help determine the margin of safety in plant design. NRC frequently does an expert elicitation using a Phenomena Identification and Ranking Table (PIRT) to identify and evaluate the state of knowledge of important modeling phenomena. However, few details about the design of these reactors and the sequence of events during accidents are known, so the process used was considered a preliminary PIRT. A panel met to define phenomena that would need to be modeled and considered the impact/importance of each phenomenon with respect to specific figures-of-merit (FoMs) (e.g., power distribution, fluence, kinetics parameters and reactivity). Each FoM reflected a potential impact on radionuclide release or loss of a barrier to release. The panel considered what the path forward might be with respect to being able to model the phenomenon in a simulation code. Results are explained for both thermal and fast spectrum designs.

  15. Dynamic neutron scattering from conformational dynamics. II. Application using molecular dynamics simulation and Markov modeling.

    Science.gov (United States)

    Yi, Zheng; Lindner, Benjamin; Prinz, Jan-Hendrik; Noé, Frank; Smith, Jeremy C

    2013-11-07

    Neutron scattering experiments directly probe the dynamics of complex molecules on the sub pico- to microsecond time scales. However, the assignment of the relaxations seen experimentally to specific structural rearrangements is difficult, since many of the underlying dynamical processes may exist on similar timescales. In an accompanying article, we present a theoretical approach to the analysis of molecular dynamics simulations with a Markov State Model (MSM) that permits the direct identification of structural transitions leading to each contributing relaxation process. Here, we demonstrate the use of the method by applying it to the configurational dynamics of the well-characterized alanine dipeptide. A practical procedure for deriving the MSM from an MD is introduced. The result is a 9-state MSM in the space of the backbone dihedral angles and the side-chain methyl group. The agreement between the quasielastic spectrum calculated directly from the atomic trajectories and that derived from the Markov state model is excellent. The dependence on the wavevector of the individual Markov processes is described. The procedure means that it is now practicable to interpret quasielastic scattering spectra in terms of well-defined intramolecular transitions with minimal a priori assumptions as to the nature of the dynamics taking place.

  16. Development of a 3D consistent 1D neutronics model for reactor core simulation

    International Nuclear Information System (INIS)

    Lee, Ki Bog; Joo, Han Gyu; Cho, Byung Oh; Zee, Sung Quun

    2001-02-01

    In this report a 3D consistent 1D model based on nonlinear analytic nodal method is developed to reproduce the 3D results. During the derivation, the current conservation factor (CCF) is introduced which guarantees the same axial neutron currents obtained from the 1D equation as the 3D reference values. Furthermore in order to properly use 1D group constants, a new 1D group constants representation scheme employing tables for the fuel temperature, moderator density and boron concentration is developed and functionalized for the control rod tip position. To test the 1D kinetics model with CCF, several steady state and transient calculations were performed and compared with 3D reference values. The errors of K-eff values were reduced about one tenth when using CCF without significant computational overhead. And the errors of power distribution were decreased to the range of one fifth or tenth at steady state calculation. The 1D kinetics model with CCF and the 1D group constant functionalization employing tables as a function of control rod tip position can provide preciser results at the steady state and transient calculation. Thus it is expected that the 1D kinetics model derived in this report can be used in the safety analysis, reactor real time simulation coupled with system analysis code, operator support system etc.

  17. Identification of Native Bacteria of the Candelaria and Tatacoa Semiarid Zone, Capable of Withstanding a Mars UV Radiation Simulation

    Science.gov (United States)

    Mendez, Y.; Vives, M.

    2017-07-01

    This work is the first study to describe native bacteria from the semi-arid areas in Candelaria and Tatacoa in Colombia, able to withstand a simulation of UV radiation, in order to draw an analogy with microbial growth on the surface of Mars. Sampling was carried out in the areas mentioned taking 50 samples of sediment divided into 25 samples of surface and 25 deep samples. As soon as the samples were transferred, they were subjected to a test of UV radiation in an atmospheric simulation chamber designed for the experiment, for periods of 1, 6 and 12 hours of exposure. Microbiological analysis as a method of plate dilution and isolation were performed using the modified AIS growth medium, macroscopic and microscopic description of morphotypes, biochemical identification of the morphotypes found, extraction of the feasible mycelium, DNA extraction and amplification of the gene 16 S by PCR. 13 morphotypes of bacteria resistant to UV radiation were found, mostly compatible with the gender of Streptomyces. One of the morphotypes found resisted 12 hours exposure. Molecular analyzes did not produce any results, because it was not possible to amplify the 16S by PCR, this may be due to that the exposure to UV radiation can degrade the DNA in existence, a affecting the results. The finding of native bacteria capable of withstanding conditions UV radiation can give us an approximation of microbial growth, mechanisms of resistance and survival under extreme conditions such as those found on Mars, in order to develop biotechnological applications and establish planetary analogues to understand the origin and evolution of the universe.

  18. GEANT4 simulation of the neutron background of the C$_6$D$_6$ set-up for capture studies at n_TOF

    CERN Document Server

    Žugec, P.; Bosnar, D.; Altstadt, S.; Andrzejewski, J.; Audouin, L.; Barbagallo, M.; Bécares, V.; Bečvář, F.; Belloni, F.; Berthoumieux, E.; Billowes, J.; Boccone, V.; Brugger, M.; Calviani, M.; Calviño, F.; Cano-Ott, D.; Carrapiço, C.; Cerutti, F.; Chiaveri, E.; Chin, M.; Cortés, G.; Cortés-Giraldo, M.A.; Diakaki, M.; Domingo-Pardo, C.; Dressler, R.; Duran, I.; Dzysiuk, N.; Eleftheriadis, C.; Ferrari, A.; Fraval, K.; Ganesan, S.; García, A.R.; Giubrone, G.; Gómez-Hornillos, M.B.; Gonçalves, I.F.; González-Romero, E.; Griesmayer, E.; Guerrero, C.; Gunsing, F.; Gurusamy, P.; Heinitz, S.; Jenkins, D.G.; Jericha, E.; Kadi, Y.; Käppeler, F.; Karadimos, D.; Kivel, N.; Koehler, P.; Kokkoris, M.; Krtička, M.; Kroll, J.; Langer, C.; Lederer, C.; Leeb, H.; Leong, L.S.; Lo Meo, S.; Losito, R.; Manousos, A.; Marganiec, J.; Martìnez, T.; Massimi, C.; Mastinu, P.F.; Mastromarco, M.; Meaze, M.; Mendoza, E.; Mengoni, A.; Milazzo, P.M.; Mingrone, F.; Mirea, M.; Mondalaers, W.; Paradela, C.; Pavlik, A.; Perkowski, J.; Plompen, A.; Praena, J.; Quesada, J.M.; Rauscher, T.; Reifarth, R.; Riego, A.; Roman, F.; Rubbia, C.; Sarmento, R.; Saxena, A.; Schillebeeckx, P.; Schmidt, S.; Schumann, D.; Tagliente, G.; Tain, J.L.; Tarrío, D.; Tassan-Got, L.; Tsinganis, A.; Valenta, S.; Vannini, G.; Variale, V.; Vaz, P.; Ventura, A.; Versaci, R.; Vermeulen, M.J.; Vlachoudis, V.; Vlastou, R.; Wallner, A.; Ware, T.; Weigand, M.; Weiß, C.; Wright, T.

    2014-05-09

    The neutron sensitivity of the C$_6$D$_6$ detector setup used at n_TOF for capture measurements has been studied by means of detailed GEANT4 simulations. A realistic software replica of the entire n_TOF experimental hall, including the neutron beam line, sample, detector supports and the walls of the experimental area has been implemented in the simulations. The simulations have been analyzed in the same manner as experimental data, in particular by applying the Pulse Height Weighting Technique. The simulations have been validated against a measurement of the neutron background performed with a $^\\mathrm{nat}$C sample, showing an excellent agreement above 1 keV. At lower energies, an additional component in the measured $^\\mathrm{nat}$C yield has been discovered, which prevents the use of $^\\mathrm{nat}$C data for neutron background estimates at neutron energies below a few hundred eV. The origin and time structure of the neutron background have been derived from the simulations. Examples of the neutron backg...

  19. MCNPX simulations of fast neutron diagnostics for accelerator-driven systems

    International Nuclear Information System (INIS)

    Habib, Moinul

    2005-12-01

    In accelerator-driven systems, the neutron spectrum will extend all the way up to the incident beam energy, i.e., several hundred MeV or even up to GeV energies. The high neutron energy allows novel diagnostics with a set of measurement techniques that can be used in a sub-critical reactor environment. Such measurements are primarily connected to system safety and validation. This report shows that in-core fast-neutron diagnostics can be employed to monitor changes in the position of incidence of the primary proton beam onto the neutron production target. It has also been shown that fast neutrons can be used to detect temperature-dependent density changes in a liquid lead-bismuth target. Fast neutrons can escape the system via the beam pipe for the incident proton beam. Out-of-core monitoring of these so called back-streaming neutrons could potentially be used to monitor beam changes if the target has a suitable shape. Moreover, diagnostics of back-streaming neutrons might be used for validation of the system design

  20. MCNPX simulations of fast neutron diagnostics for accelerator-driven systems

    Energy Technology Data Exchange (ETDEWEB)

    Habob, Moinul

    2005-12-15

    In accelerator-driven systems, the neutron spectrum will extend all the way up to the incident beam energy, i.e., several hundred MeV or even up to GeV energies. The high neutron energy allows novel diagnostics with a set of measurement techniques that can be used in a sub-critical reactor environment. Such measurements are primarily connected to system safety and validation. This report shows that in-core fast-neutron diagnostics can be employed to monitor changes in the position of incidence of the primary proton beam onto the neutron production target. It has also been shown that fast neutrons can be used to detect temperature-dependent density changes in a liquid lead-bismuth target. Fast neutrons can escape the system via the beam pipe for the incident proton beam. Out-of-core monitoring of these so called back-streaming neutrons could potentially be used to monitor beam changes if the target has a suitable shape. Moreover, diagnostics of back-streaming neutrons might be used for validation of the system design.

  1. DIANE, a simulation code for the interaction of neutrons with living tissues. Application to low doses of fast neutrons on human tumoral cells; DIANE, un code de simulation de l'interaction des neutrons avec la matiere vivante. Applications aux faibles doses de neutrons rapides sur des cellules tumorales humaines

    Energy Technology Data Exchange (ETDEWEB)

    Nenot, M.L

    2003-07-15

    Our work deals with the irradiation of cells and living tissues by 14 MeV neutrons at very low doses (a few 10{sup -2} Gy). Such experiments require an accurate knowledge of the values of neutron dose rates and fluences at the level of cell cultures. We have performed measurements of fluence rates through an activation method applied to gold and copper foils. The fluence rate is deduced from the gamma rays emitted by the irradiated foils. Neutron doses and dose rates have been measured through varied methods: PIN diodes, ionization tissue equivalent chambers, and Geiger-Mueller counters. We have designed the DIANE code to simulate the impact of energetic neutrons on cells. This code can be used with isolated cells or macroscopic tissues, it takes into account the roles of the ionisation electrons produced by recoil nuclei entering the cell. This point is all the more important since recent works have highlighted the impact of very low energy electrons on DNA. (A.C.)

  2. Modeling and Simulation Monte Carlo by the MCNP code for determining neutron parameters of the nuclear reactor-subcritical assembly in CNSTN

    International Nuclear Information System (INIS)

    Romdhani, Ibtissem

    2014-01-01

    As part of developing its nuclear infrastructure base, the National Science and Technology Center Nuclear (CNSTN) examines the technical feasibility of setting up a new installation of subcritical assembly. Our study focuses on determining the neutron parameters of a nuclear zero power reactor based on Monte Carlo simulation MCNP. The objective of the simulation is to model the installation, determine the effective multiplication factor, and spatial distribution of neutron flux.

  3. Progress in the development of the neutron flux monitoring system of the French GEN-IV SFR: simulations and experimental validations [ANIMMA--2015-IO-392

    Energy Technology Data Exchange (ETDEWEB)

    Jammes, C.; Filliatre, P.; Izarra, G. de [CEA, DEN, Cadarache, Reactor Studies Department, 13108 Saint-Paul-lez-Durance (France); Elter, Zs. [CEA, DEN, Cadarache, Reactor Studies Department, 13108 Saint-Paul-lez-Durance (France); Chalmers University of Technology, Department of Applied Physics, Division of Nuclear Engineering, SE-412 96 Goeteborg (Sweden); Verma, V. [CEA, DEN, Cadarache, Reactor Studies Department, 13108 Saint-Paul-lez-Durance (France); Uppsala University, Division of Applied Nuclear Physics, Box 516, SE-75120 Uppsala (Sweden); Hamrita, H.; Bakkali, M. [CEA, DRT, LIST, Metrology, Instrumentation and Information Department, Saclay, 91191 Gif-sur-Yvette (France); Chapoutier, N.; Scholer, A.C.; Verrier, D. [AREVA NP, 10 rue Juliette Recamier F-69456 Lyon (France); Hellesen, C.; Jacobsson, S. [Uppsala University, Division of Applied Nuclear Physics, Box 516, SE-75120 Uppsala (Sweden); Pazsit, I. [Chalmers University of Technology, Department of Applied Physics, Division of Nuclear Engineering, SE-412 96 Goeteborg (Sweden); Cantonnet, B.; Nappe, J.C. [PHOTONIS France, Nuclear Instrumentation, 19100 Brive-la-Gaillarde (France); Molinie, P.; Dessante, P.; Hanna, R.; Kirkpatrick, M.; Odic, E. [Supelec, Energy Department, 3 rue Joliot-Curie, 91191 Gif-sur-Yvette (France)

    2015-07-01

    France has a long experience of about 50 years in designing, building and operating sodium-cooled fast reactors (SFR) such as RAPSODIE, PHENIX and SUPER PHENIX. Fast reactors feature the double capability of reducing nuclear waste and saving nuclear energy resources by burning actinides. Since this reactor type is one of those selected by the Generation IV International Forum, the French government asked, in the year 2006, CEA, namely the French Alternative Energies and Atomic Energy Commission, to lead the development of an innovative GEN-IV nuclear- fission power demonstrator. The major objective is to improve the safety and availability of an SFR. The neutron flux monitoring (NFM) system of any reactor must, in any situation, permit both reactivity control and power level monitoring from startup to full power. It also has to monitor possible changes in neutron flux distribution within the core region in order to prevent any local melting accident. The neutron detectors will have to be installed inside the reactor vessel because locations outside the vessel will suffer from severe disadvantages; radially the neutron shield that is also contained in the reactor vessel will cause unacceptable losses in neutron flux; below the core the presence of a core-catcher prevents from inserting neutron guides; and above the core the distance is too large to obtain decent neutron signals outside the vessel. Another important point is to limit the number of detectors placed in the vessel in order to alleviate their installation into the vessel. In this paper, we show that the architecture of the NFM system will rely on high-temperature fission chambers (HTFC) featuring wide-range flux monitoring capability. The definition of such a system is presented and the justifications of technological options are brought with the use of simulation and experimental results. Firstly, neutron-transport calculations allow us to propose two in-vessel regions, namely the above-core and under

  4. McStas 1.1. A freeware package for neutron Monte Carlo ray-tracing simulations

    International Nuclear Information System (INIS)

    Lefmann, K.; Nielsen, K.

    1999-01-01

    Neutron simulation is becoming an indispensable tool for neutron instrument design. At Risoe National Laboratory, a user-friendly, versatile, and fast simulation package, McStas has been developed, which may be freely downloaded from our website. An instrument is described in the McStas meta-language and is composed of elements from the McStas component library, which is under constant development and debugging by both the users and us. The McStas front- and back-ends take care of performing the simulations and displaying their results, respectively. McStas 1.1 facilities detailed simulations of complicated triple-axis instruments like the Riso RITA spectrometer, and it is equally well equipped for time-of flight spectrometers. At ECNS'99, a brief tutorial of McStas including a few on-line demonstrations is presented. Further, results from the latest simulation work in the growing McStas user group are presented and the future of this project is discussed. (author)

  5. The GEANT4 simulation study of the characteristic γ-ray spectrum of TNT under soil induced by DT neutron

    International Nuclear Information System (INIS)

    Qin Xue; Han Jifeng; Yang Chaowen

    2014-01-01

    The characteristic γ-ray spectrum of TNT under soil induced by DT neutron is measured based on the PFTNA demining system. GEANT4 Monte Carlo simulation toolkit is used to simulate the whole experimental procedure. The simulative spectrum is compared with the experimental spectrum. The result shows that they are mainly consistent. It is for the first time to analyze the spectrum by Monte Carlo simulation, the share of the background sources such as neutron, gamma are obtained, the contribution that the experimental apparatus such as shielding, detector sleeve, moderator make to the background is analysed. The study found that the effective gamma signal (from soil and TNT) is only 29% of the full-spectrum signal, and the background signal is more than 68% of the full-spectrum signal, which is mainly produced in the shielding and the detector sleeve. The simulation result shows that by gradually improving the shielding and the cadmium of the detector sleeve, the share of the effective gamma signal can increase to 36% and the background signal can fell 7% eventually. (authors)

  6. Simulations of the muon-induced neutron background of the EDELWEISS-II experiment for Dark Matter search

    International Nuclear Information System (INIS)

    Horn, O.M.

    2007-01-01

    In modern astroparticle physics and cosmology, the nature of Dark Matter is one of the central problems. Particle Dark Matter in form of WIMPs is favoured among many proposed candidates. The EDELWEISS direct Dark Matter search uses Germanium bolometers to detect these particles by nuclear recoils. Here, the use of two signal channels on an event-by-event basis, namely the heat and ionisation signal, enables the detectors to discriminate between electron and nuclear recoils. This technique leaves neutrons in the underground laboratory as the main background for the experiment. Besides (α,n) reactions of natural radioactivity, neutrons are produced in electromagnetic and hadronic showers induced by cosmic ray muons in the surrounding rock and shielding material of the Germanium crystals. To reach high sensitivities, the EDELWEISS-II experiment, as well as other direct Dark Matter searches, has to efficiently suppress this neutron background. The present work is devoted to study the muon-induced neutron flux in the underground laboratory LSM and the interaction rate within the Germanium crystals by using the Monte Carlo simulation toolkit Geant4. To ensure reliable results, the implemented physics in the toolkit regarding neutron production is tested in a benchmark geometry and results are compared to experimental data and other simulation codes. Also, the specific energy and angular distribution of the muon flux in the underground laboratory as a consequence of the asymmetric mountain overburden is implemented. A good agreement of the simulated muon flux is shown in a comparison to preliminary experimental data obtained with the EDELWEISS-II muon veto system. Furthermore, within a detailed geometry of the experimental setup, the muon-induced background rate of nuclear recoils in the bolometers is simulated. Coincidences of recoil events in the Germanium with an energy deposit of the muoninduced shower in the plastic scintillators of the veto system are studied to

  7. Simulations of the muon-induced neutron background of the EDELWEISS-II experiment for Dark Matter search

    Energy Technology Data Exchange (ETDEWEB)

    Horn, O M

    2007-12-21

    In modern astroparticle physics and cosmology, the nature of Dark Matter is one of the central problems. Particle Dark Matter in form of WIMPs is favoured among many proposed candidates. The EDELWEISS direct Dark Matter search uses Germanium bolometers to detect these particles by nuclear recoils. Here, the use of two signal channels on an event-by-event basis, namely the heat and ionisation signal, enables the detectors to discriminate between electron and nuclear recoils. This technique leaves neutrons in the underground laboratory as the main background for the experiment. Besides ({alpha},n) reactions of natural radioactivity, neutrons are produced in electromagnetic and hadronic showers induced by cosmic ray muons in the surrounding rock and shielding material of the Germanium crystals. To reach high sensitivities, the EDELWEISS-II experiment, as well as other direct Dark Matter searches, has to efficiently suppress this neutron background. The present work is devoted to study the muon-induced neutron flux in the underground laboratory LSM and the interaction rate within the Germanium crystals by using the Monte Carlo simulation toolkit Geant4. To ensure reliable results, the implemented physics in the toolkit regarding neutron production is tested in a benchmark geometry and results are compared to experimental data and other simulation codes. Also, the specific energy and angular distribution of the muon flux in the underground laboratory as a consequence of the asymmetric mountain overburden is implemented. A good agreement of the simulated muon flux is shown in a comparison to preliminary experimental data obtained with the EDELWEISS-II muon veto system. Furthermore, within a detailed geometry of the experimental setup, the muon-induced background rate of nuclear recoils in the bolometers is simulated. Coincidences of recoil events in the Germanium with an energy deposit of the muoninduced shower in the plastic scintillators of the veto system are studied

  8. Energy and direction distribution of neutrons in workplace fields: Implication of the results from the EVIDOS project for the set-up of simulated workplace fields

    International Nuclear Information System (INIS)

    Luszik-Bhadra, M.; Lacoste, V.; Reginatto, M.; Zimbal, A.

    2007-01-01

    Workplace neutron spectra from nuclear facilities obtained within the European project EVIDOS are compared with those of the simulated workplace fields CANEL and SIGMA and fields set-up with radionuclide sources at the PTB. Contributions of neutrons to ambient dose equivalent and personal dose equivalent are given in three energy intervals (for thermal, intermediate and fast neutrons) together with the corresponding direction distribution, characterised by three different types of distributions (isotropic, weakly directed and directed). The comparison shows that none of the simulated workplace fields investigated here can model all the characteristics of the fields observed at power reactors. (authors)

  9. Reference rad