Interface requirements to couple thermal-hydraulic codes to 3D neutronic codes
Langenbuch, S.; Austregesilo, H.; Velkov, K.
1997-07-01
The present situation of thermalhydraulics codes and 3D neutronics codes is briefly described and general considerations for coupling of these codes are discussed. Two different basic approaches of coupling are identified and their relative advantages and disadvantages are discussed. The implementation of the coupling for 3D neutronics codes in the system ATHLET is presented. Meanwhile, this interface is used for coupling three different 3D neutronics codes.
3D neutronic codes coupled with thermal-hydraulic system codes for PWR, and BWR and VVER reactors
Langenbuch, S.; Velkov, K.; Lizorkin, M.
1997-07-01
This paper describes the objectives of code development for coupling 3D neutronics codes with thermal-hydraulic system codes. The present status of coupling ATHLET with three 3D neutronics codes for VVER- and LWR-reactors is presented. After describing the basic features of the 3D neutronic codes BIPR-8 from Kurchatov-Institute, DYN3D from Research Center Rossendorf and QUABOX/CUBBOX from GRS, first applications of coupled codes for different transient and accident scenarios are presented. The need of further investigations is discussed.
Parameterization of nuclear cross-sections for coupled neutronic- thermalhydraulic codes
Miro, R.; Verdu, G.; Barrachina, T.; Rosello, O.
2006-07-01
The present work consists of developing an in-house methodology, called SIMTAB, to characterize, in a simplified way, the reactor core of LWR Nuclear Power Plants. Specifically, a cross-sections and kinetic parameters set are obtained as a function of the prompt and control variables. So that, the core can be modeled using a limited number of neutronic regions, in such a way that the reactor kinetic behavior is properly characterized. This simplification of the reactor core permits, from an operative point of view, the use of few cross sections data sets in coupled 3D neutronic-thermalhydraulic codes. (authors)
2013-06-24
Version 07 TART2012 is a coupled neutron-photon Monte Carlo transport code designed to use three-dimensional (3-D) combinatorial geometry. Neutron and/or photon sources as well as neutron induced photon production can be tracked. It is a complete system to assist you with input preparation, running Monte Carlo calculations, and analysis of output results. TART2012 is also incredibly FAST; if you have used similar codes, you will be amazed at how fast this code is compared tomore » other similar codes. Use of the entire system can save you a great deal of time and energy. TART2012 extends the general utility of the code to even more areas of application than available in previous releases by concentrating on improving the physics, particularly with regard to improved treatment of neutron fission, resonance self-shielding, molecular binding, and extending input options used by the code. Several utilities are included for creating input files and displaying TART results and data. TART2012 uses the latest ENDF/B-VI, Release 8, data. New for TART2012 is the use of continuous energy neutron cross sections, in addition to its traditional multigroup cross sections. For neutron interaction, the data are derived using ENDF-ENDL2005 and include both continuous energy cross sections and 700 group neutron data derived using a combination of ENDF/B-VI, Release 8, and ENDL data. The 700 group structure extends from 10-5 eV up to 1 GeV. Presently nuclear data are only available up to 20 MeV, so that only 616 of the groups are currently used. For photon interaction, 701 point photon data were derived using the Livermore EPDL97 file. The new 701 point structure extends from 100 eV up to 1 GeV, and is currently used over this entire energy range. TART2012 completely supersedes all older versions of TART, and it is strongly recommended that one use only the most recent version of TART2012 and its data files. Check authors homepage for related information: http
A Coupled Neutron-Photon 3-D Combinatorial Geometry Monte Carlo Transport Code
1998-06-12
TART97 is a coupled neutron-photon, 3 dimensional, combinatorial geometry, time dependent Monte Carlo transport code. This code can run on any modern computer. It is a complete system to assist you with input preparation, running Monte Carlo calculations, and analysis of output results. TART97 is also incredibly fast: if you have used similar codes, you will be amazed at how fast this code is compared to other similar codes. Use of the entire system canmore » save you a great deal of time and energy. TART 97 is distributed on CD. This CD contains on-line documentation for all codes included in the system, the codes configured to run on a variety of computers, and many example problems that you can use to familiarize yourself with the system. TART97 completely supersedes all older versions of TART, and it is strongly recommended that users only use the most recent version of TART97 and ist data files.« less
Petruzzi, A.; D'Auria, F.; Giannotti, W.; Ivanov, K.
2005-02-15
The best-estimate calculation results from complex system codes are affected by approximations that are unpredictable without the use of computational tools that account for the various sources of uncertainty.The code with (the capability of) internal assessment of uncertainty (CIAU) has been previously proposed by the University of Pisa to realize the integration between a qualified system code and an uncertainty methodology and to supply proper uncertainty bands each time a nuclear power plant (NPP) transient scenario is calculated. The derivation of the methodology and the results achieved by the use of CIAU are discussed to demonstrate the main features and capabilities of the method.In a joint effort between the University of Pisa and The Pennsylvania State University, the CIAU method has been recently extended to evaluate the uncertainty of coupled three-dimensional neutronics/thermal-hydraulics calculations. The result is CIAU-TN. The feasibility of the approach has been demonstrated, and sample results related to the turbine trip transient in the Peach Bottom NPP are shown. Notwithstanding that the full implementation and use of the procedure requires a database of errors not available at the moment, the results give an idea of the errors expected from the present computational tools.
Code System for 2-Group, 3D Neutronic Kinetics Calculations Coupled to Core Thermal Hydraulics.
2000-05-12
Version 00 QUARK is a combined computer program comprising a revised version of the QUANDRY three-dimensional, two-group neutron kinetics code and an upgraded version of the COBRA transient core analysis code (COBRA-EN). Starting from either a critical steady-state (k-effective or critical dilute Boron problem) or a subcritical steady-state (fixed source problem) in a PWR plant, the code allows one to simulate the neutronic and thermal-hydraulic core transient response to reactivity accidents initiated both inside themore » vessel (such as a control rod ejection) and outside the vessel (such as the sudden change of the Boron concentration in the coolant). QUARK output can be used as input to PSR-470/NORMA-FP to perform a subchannel analysis from converged coarse-mesh nodal solutions.« less
Leap Frog and Time Step Sub-Cycle Scheme for Coupled Neutronics and Thermal-Hydraulic Codes
Lu, S.
2002-07-01
As the result of the advancing TCP/IP based inter-process communication technology, more and more legacy thermal-hydraulic codes have been coupled with neutronics codes to provide best-estimate capabilities for reactivity related reactor transient analysis. Most of the coupling schemes are based on closely coupled serial or parallel approaches. Therefore, the execution of the coupled codes usually requires significant CPU time, when a complicated system is analyzed. Leap Frog scheme has been used to reduce the run time. The extent of the decoupling is usually determined based on a trial and error process for a specific analysis. It is the intent of this paper to develop a set of general criteria, which can be used to invoke the automatic Leap Frog algorithm. The algorithm will not only provide the run time reduction but also preserve the accuracy. The criteria will also serve as the base of an automatic time step sub-cycle scheme when a sudden reactivity change is introduced and the thermal-hydraulic code is marching with a relatively large time step. (authors)
Bingham, Philip R; Santos-Villalobos, Hector J
2011-01-01
Coded aperture techniques have been applied to neutron radiography to address limitations in neutron flux and resolution of neutron detectors in a system labeled coded source imaging (CSI). By coding the neutron source, a magnified imaging system is designed with small spot size aperture holes (10 and 100 m) for improved resolution beyond the detector limits and with many holes in the aperture (50% open) to account for flux losses due to the small pinhole size. An introduction to neutron radiography and coded aperture imaging is presented. A system design is developed for a CSI system with a development of equations for limitations on the system based on the coded image requirements and the neutron source characteristics of size and divergence. Simulation has been applied to the design using McStas to provide qualitative measures of performance with simulations of pinhole array objects followed by a quantitative measure through simulation of a tilted edge and calculation of the modulation transfer function (MTF) from the line spread function. MTF results for both 100um and 10um aperture hole diameters show resolutions matching the hole diameters.
Curca-Tivig, Florin; Merk, Stephan; Pautz, Andreas; Thareau, Sebastien
2007-07-01
Anticipating future needs of our customers and willing to concentrate synergies and competences existing in the company for the benefit of our customers, AREVA NP decided in 2002 to develop the next generation of coupled neutronics/ core thermal-hydraulic (TH) code systems for fuel assembly and core design calculations for both, PWR and BWR applications. The global CONVERGENCE project was born: after a feasibility study of one year (2002) and a conceptual phase of another year (2003), development was started at the beginning of 2004. The present paper introduces the CONVERGENCE project, presents the main feature of the new code system ARCADIA{sup R} and concludes on customer benefits. ARCADIA{sup R} is designed to meet AREVA NP market and customers' requirements worldwide. Besides state-of-the-art physical modeling, numerical performance and industrial functionality, the ARCADIA{sup R} system is featuring state-of-the-art software engineering. The new code system will bring a series of benefits for our customers: e.g. improved accuracy for heterogeneous cores (MOX/ UOX, Gd...), better description of nuclide chains, and access to local neutronics/ thermal-hydraulics and possibly thermal-mechanical information (3D pin by pin full core modeling). ARCADIA is a registered trademark of AREVA NP. (authors)
Lee, Deokjung; Downar, Thomas J.; Ulses, Anthony; Akdeniz, Bedirhan; Ivanov, Kostadin N.
2004-10-15
An analysis of the Peach Bottom Unit 2 Turbine Trip 2 (TT2) experiment has been performed using the U.S. Nuclear Regulatory Commission coupled thermal-hydraulics and neutronics code TRAC-M/PARCS. The objective of the analysis was to assess the performance of TRAC-M/PARCS on a BWR transient with significance in two-phase flow and spatial variations of the neutron flux. TRAC-M/PARCS results are found to be in good agreement with measured plant data for both steady-state and transient phases of the benchmark. Additional analyses of four fictitious extreme scenarios are performed to provide a basis for code-to-code comparisons and comprehensive testing of the thermal-hydraulics/neutronics coupling. The obtained results of sensitivity studies on the effect of direct moderator heating on transient simulation indicate the importance of this modeling aspect.
Russell, G.J.; Pitcher, E.J.; Ferguson, P.D.
1995-12-01
Optimizing the neutronic performance of a coupled-moderator system for a Long-Pulse Spallation Source is a new and challenging area for the spallation target-system designer. For optimal performance of a neutron source, it is essential to have good communication with instrument scientists to obtain proper design criteria and continued interaction with mechanical, thermal-hydraulic, and materials engineers to attain a practical design. A good comprehension of the basics of coupled-moderator neutronics will aid in the proper design of a target system for a Long-Pulse Spallation Source.
Horiguchi, Hironori; Sato, Tatsuhiko; Kumada, Hiroaki; Yamamoto, Tetsuya; Sakae, Takeji
2015-01-01
The absorbed doses deposited by boron neutron capture therapy (BNCT) can be categorized into four components: α and 7Li particles from the 10B(n, α)7Li reaction, 0.54-MeV protons from the 14N(n, p)14C reaction, the recoiled protons from the 1H(n, n) 1H reaction, and photons from the neutron beam and 1H(n, γ)2H reaction. For evaluating the irradiation effect in tumors and the surrounding normal tissues in BNCT, it is of great importance to estimate the relative biological effectiveness (RBE) for each dose component in the same framework. We have, therefore, established a new method for estimating the RBE of all BNCT dose components on the basis of the microdosimetric kinetic model. This method employs the probability density of lineal energy, y, in a subcellular structure as the index for expressing RBE, which can be calculated using the microdosimetric function implemented in the particle transport simulation code (PHITS). The accuracy of this method was tested by comparing the calculated RBE values with corresponding measured data in a water phantom irradiated with an epithermal neutron beam. The calculation technique developed in this study will be useful for biological dose estimation in treatment planning for BNCT. PMID:25428243
NASA Astrophysics Data System (ADS)
Ichikawa, Ryoko; Masuhara, Yasuhiro; Kasahara, Fumio
The Best Estimate Plus Uncertainty (BEPU) method has been prepared for the regulatory cross-check analysis at Japan Nuclear Energy Safety Organization (JNES) on base of the three-dimensional neutron-kinetics/thermal- hydraulics coupled code SKETCH-INS/TRACE5.0. In the preparation, TRACE5.0 is verified against the large-scale thermal-hydraulic tests carried out with NUPEC facility. These tests were focused on the pressure drop of steam-liquid two phase flow and void fraction distribution. From the comparison of the experimental data with other codes (RELAP5/MOD3.3 and TRAC-BF1), TRACE5.0 was judged better than other codes. It was confirmed that TRACE5.0 has high reliability for thermal hydraulics behavior and are used as a best-estimate code for the statistical safety evaluation. Next, the coupled code SKETCH-INS/TRACE5.0 was applied to turbine trip tests performed at the Peach Bottom-2 BWR4 Plant. The turbine trip event shows the rapid power peak due to the voids collapse with the pressure increase. The analyzed peak value of core power is better simulated than the previous version SKETCH-INS/TRAC-BF1. And the statistical safety evaluation using SKETCH-INS/TRACE5.0 was applied to the loss of load transient for examining the influence of the choice of sampling method.
Coupled Neutron Transport for HZETRN
NASA Technical Reports Server (NTRS)
Slaba, Tony C.; Blattnig, Steve R.
2009-01-01
Exposure estimates inside space vehicles, surface habitats, and high altitude aircrafts exposed to space radiation are highly influenced by secondary neutron production. The deterministic transport code HZETRN has been identified as a reliable and efficient tool for such studies, but improvements to the underlying transport models and numerical methods are still necessary. In this paper, the forward-backward (FB) and directionally coupled forward-backward (DC) neutron transport models are derived, numerical methods for the FB model are reviewed, and a computationally efficient numerical solution is presented for the DC model. Both models are compared to the Monte Carlo codes HETC-HEDS, FLUKA, and MCNPX, and the DC model is shown to agree closely with the Monte Carlo results. Finally, it is found in the development of either model that the decoupling of low energy neutrons from the light particle transport procedure adversely affects low energy light ion fluence spectra and exposure quantities. A first order correction is presented to resolve the problem, and it is shown to be both accurate and efficient.
Morgan C. White
2000-07-01
The fundamental motivation for the research presented in this dissertation was the need to development a more accurate prediction method for characterization of mixed radiation fields around medical electron accelerators (MEAs). Specifically, a model is developed for simulation of neutron and other particle production from photonuclear reactions and incorporated in the Monte Carlo N-Particle (MCNP) radiation transport code. This extension of the capability within the MCNP code provides for the more accurate assessment of the mixed radiation fields. The Nuclear Theory and Applications group of the Los Alamos National Laboratory has recently provided first-of-a-kind evaluated photonuclear data for a select group of isotopes. These data provide the reaction probabilities as functions of incident photon energy with angular and energy distribution information for all reaction products. The availability of these data is the cornerstone of the new methodology for state-of-the-art mutually coupled photon-neutron transport simulations. The dissertation includes details of the model development and implementation necessary to use the new photonuclear data within MCNP simulations. A new data format has been developed to include tabular photonuclear data. Data are processed from the Evaluated Nuclear Data Format (ENDF) to the new class ''u'' A Compact ENDF (ACE) format using a standalone processing code. MCNP modifications have been completed to enable Monte Carlo sampling of photonuclear reactions. Note that both neutron and gamma production are included in the present model. The new capability has been subjected to extensive verification and validation (V&V) testing. Verification testing has established the expected basic functionality. Two validation projects were undertaken. First, comparisons were made to benchmark data from literature. These calculations demonstrate the accuracy of the new data and transport routines to better than 25 percent. Second, the ability to
NSCool: Neutron star cooling code
NASA Astrophysics Data System (ADS)
Page, Dany
2016-09-01
NSCool is a 1D (i.e., spherically symmetric) neutron star cooling code written in Fortran 77. The package also contains a series of EOSs (equation of state) to build stars, a series of pre-built stars, and a TOV (Tolman- Oppenheimer-Volkoff) integrator to build stars from an EOS. It can also handle “strange stars” that have a huge density discontinuity between the quark matter and the covering thin baryonic crust. NSCool solves the heat transport and energy balance equations in whole GR, resulting in a time sequence of temperature profiles (and, in particular, a Teff - age curve). Several heating processes are included, and more can easily be incorporated. In particular it can evolve a star undergoing accretion with the resulting deep crustal heating, under a steady or time-variable accretion rate. NSCool is robust, very fast, and highly modular, making it easy to add new subroutines for new processes.
Fast-neutron, coded-aperture imager
NASA Astrophysics Data System (ADS)
Woolf, Richard S.; Phlips, Bernard F.; Hutcheson, Anthony L.; Wulf, Eric A.
2015-06-01
This work discusses a large-scale, coded-aperture imager for fast neutrons, building off a proof-of concept instrument developed at the U.S. Naval Research Laboratory (NRL). The Space Science Division at the NRL has a heritage of developing large-scale, mobile systems, using coded-aperture imaging, for long-range γ-ray detection and localization. The fast-neutron, coded-aperture imaging instrument, designed for a mobile unit (20 ft. ISO container), consists of a 32-element array of 15 cm×15 cm×15 cm liquid scintillation detectors (EJ-309) mounted behind a 12×12 pseudorandom coded aperture. The elements of the aperture are composed of 15 cm×15 cm×10 cm blocks of high-density polyethylene (HDPE). The arrangement of the aperture elements produces a shadow pattern on the detector array behind the mask. By measuring of the number of neutron counts per masked and unmasked detector, and with knowledge of the mask pattern, a source image can be deconvolved to obtain a 2-d location. The number of neutrons per detector was obtained by processing the fast signal from each PMT in flash digitizing electronics. Digital pulse shape discrimination (PSD) was performed to filter out the fast-neutron signal from the γ background. The prototype instrument was tested at an indoor facility at the NRL with a 1.8-μCi and 13-μCi 252Cf neutron/γ source at three standoff distances of 9, 15 and 26 m (maximum allowed in the facility) over a 15-min integration time. The imaging and detection capabilities of the instrument were tested by moving the source in half- and one-pixel increments across the image plane. We show a representative sample of the results obtained at one-pixel increments for a standoff distance of 9 m. The 1.8-μCi source was not detected at the 26-m standoff. In order to increase the sensitivity of the instrument, we reduced the fastneutron background by shielding the top, sides and back of the detector array with 10-cm-thick HDPE. This shielding configuration led
Miro, R.; Maggini, F.; Barrachina, T.; Verdu, G.; Gomez, A.; Ortego, A.; Murillo, J. C.
2006-07-01
The Reactor Ejection Accident (REA) belongs to the Reactor Initiated Accidents (RIA) category of accidents and it is part of the licensing basis accident analyses required for pressure water reactors (PWR). The REA at hot zero power (HZP) is characterized by a single rod ejection from a core position with a very low power level. The evolution consists basically of a continuous reactivity insertion. The main feature limiting the consequences of the accident in a PWR is the Doppler Effect. To check the performance of the coupled code RELAP5/PARCS2.5 and RELAP5/VALKIN a REA in Trillo NPP is simulated. These analyses will allow knowing more accurately the PWR real plant phenomenology in the RIA most limiting conditions. (authors)
The CoCoNuT code: from neutron star oscillations to supernova explosions
NASA Astrophysics Data System (ADS)
Cerdá-Durán, P.; Gabler, M.; Müller, E.; Font, J. A.; Stergioulas, N.; Obergaulinger, M.; Aloy, M. A.; DeBrye, N.; Cordero-Carrión, I.; Ibáñez, J. M.
2013-05-01
CoCoNuT is a numerical code, that evolves the General relativistic magneto-hydrodynamics equations coupled to the Einstein equations in the CFC approximation. Its main purpose is to simulate astrophysical scenarios in which strong gravity is important such as the collapse of massive stars and the evolution of neutron stars. I review recent results of the numerical code regarding neutron star oscillations and core collapse supernova and its observational consequences.
1982-04-30
Version 00 As a part of the measurement and analysis plan for the Dosimetry Experiment at the "JOYO" experimental fast reactor, neutron flux spectral analysis is performed using the NEUPAC (Neutron Unfolding Code Package) code system. NEUPAC calculates the neutron flux spectra and other integral quantities from the activation data of the dosimeter foils.
Multigroup Complex Geometry Neutron Diffusion Code System.
2002-12-18
Version 01 SNAP-3D is based on SNAP2 and is a one- two- or three-dimensional multigroup diffusion code system. It is primarily intended for neutron diffusion calculations, but it can also carry out gamma-ray calculations if the diffusion approximation is accurate enough. It is suitable for fast and thermal reactor core calculations and for shield calculations. SNAP-3D can solve the multi-group neutron diffusion equations using finite difference methods in (x,y,z), (r,theta,z), (TRI,z), (HEX,z) or (spherical) coordinates.more » The one-dimensional slab and cylindrical geometries and the two-dimensional (x,y), (r,z), (r,theta), (HEX) and (TRI) are all treated as simple special cases of three-dimensional geometries. Numerous reflective and periodic symmetry options are available and may be used to reduce the number of mesh points necessary to represent the system. Extrapolation lengths can be specified at internal and external boundaries. The problem classes are: 1) eigenvalue search for critical k-effective, 2) eigenvalue search for critical buckling, 3) eigenvalue search for critical time-constant, 4) fixed source problems in which the sources are functions of regions, 5) fixed source problems in which the sources are provided, on disc, for every mesh point and group.« less
Multigroup Complex Geometry Neutron Diffusion Code System.
MCCALLIEN, C. W.J.
2002-12-18
Version 01 SNAP-3D is based on SNAP2 and is a one- two- or three-dimensional multigroup diffusion code system. It is primarily intended for neutron diffusion calculations, but it can also carry out gamma-ray calculations if the diffusion approximation is accurate enough. It is suitable for fast and thermal reactor core calculations and for shield calculations. SNAP-3D can solve the multi-group neutron diffusion equations using finite difference methods in (x,y,z), (r,theta,z), (TRI,z), (HEX,z) or (spherical) coordinates. The one-dimensional slab and cylindrical geometries and the two-dimensional (x,y), (r,z), (r,theta), (HEX) and (TRI) are all treated as simple special cases of three-dimensional geometries. Numerous reflective and periodic symmetry options are available and may be used to reduce the number of mesh points necessary to represent the system. Extrapolation lengths can be specified at internal and external boundaries. The problem classes are: 1) eigenvalue search for critical k-effective, 2) eigenvalue search for critical buckling, 3) eigenvalue search for critical time-constant, 4) fixed source problems in which the sources are functions of regions, 5) fixed source problems in which the sources are provided, on disc, for every mesh point and group.
ABAREX: A neutron spherical optical-statistical model code
Lawson, R.D.
1992-06-01
The spherical optical-statistical model is briefly reviewed and the capabilities of the neutron scattering code, ABAREX, are presented. Input files for ten examples, in which neutrons are scattered by various nuclei, are given and the output of each run is discussed in detail.
Monte Carlo code for neutron scattering instrumentation design and analysis
Daemen, L.; Fitzsimmons, M.; Hjelm, R.; Olah, G.; Roberts, J.; Seeger, P.; Smith, G.; Thelliez, T.
1996-09-01
This is the final report of a one-year, Laboratory-Directed Research and Development (LDRD) at the Los Alamos National Laboratory (LANL). The development of next generation, accelerator based neutron sources calls for the design of new instruments for neutron scattering studies of materials. It will be necessary, in the near future, to evaluate accurately and rapidly the performance of new and traditional neutron instruments at short- and long-pulse spallation neutron sources, as well as continuous sources. We have developed a code that is a design tool to assist the instrument designer model new or existing instruments, test their performance, and optimize their most important features.
Coupling procedure for TRANSURANUS and KTF codes
Jimenez, J.; Alglave, S.; Avramova, M.
2012-07-01
The nuclear industry aims to ensure safe and economic operation of each single fuel rod introduced in the reactor core. This goal is even more challenging nowadays due to the current strategy of going for higher burn-up (fuel cycles of 18 or 24 months) and longer residence time. In order to achieve that goal, fuel modeling is the key to predict the fuel rod behavior and lifetime under thermal and pressure loads, corrosion and irradiation. In this context, fuel performance codes, such as TRANSURANUS, are used to improve the fuel rod design. The modeling capabilities of the above mentioned tools can be significantly improved if they are coupled with a thermal-hydraulic code in order to have a better description of the flow conditions within the rod bundle. For LWR applications, a good representation of the two phase flow within the fuel assembly is necessary in order to have a best estimate calculation of the heat transfer inside the bundle. In this paper we present the coupling methodology of TRANSURANUS with KTF (Karlsruhe Two phase Flow subchannel code) as well as selected results of the coupling proof of principle. (authors)
Coupled neutronics and thermal-hydraulics numerical simulations of a Molten Fast Salt Reactor (MFSR)
NASA Astrophysics Data System (ADS)
Laureau, A.; Rubiolo, P. R.; Heuer, D.; Merle-Lucotte, E.; Brovchenko, M.
2014-06-01
Coupled neutronics and thermalhydraulic numerical analyses of a molten salt fast reactor 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.
User Guide for the R5EXEC Coupling Interface in the RELAP5-3D Code
Forsmann, J. Hope; Weaver, Walter L.
2015-04-01
This report describes the R5EXEC coupling interface in the RELAP5-3D computer code from the users perspective. The information in the report is intended for users who want to couple RELAP5-3D to other thermal-hydraulic, neutron kinetics, or control system simulation codes.
Modular Code and Data System for Fast Reactor Neutronics Analyses
2008-06-30
Version 00. The European Reactor ANalysis Optimized calculation System, ERANOS, has been developed and validated with the aim of providing a suitable basis for reliable neutronic calculations of current as well as advanced fast reactor cores. It consists of data libraries, deterministic codes and calculation procedures which have been developed within the European Collaboration on Fast Reactors over the past 20 years or so, in order to answer the needs of both industrial and R&Dmore » organisations. The whole system counts roughly 250 functions and 3000 subroutines totalling 450000 lines of FORTRAN-77 and ESOPE instructions. ERANOS is written using the ALOS software which requires only standard FORTRAN compilers and includes advanced programming features. A modular structure was adopted for easier evolution and incorporation of new functionalities. Blocks of data (SETs) can be created or used by the modules themselves or by the user via the LU control language. Programming, and dynamic memory allocation, are performed by means of the ESOPE language. External temporary storage and permanent storage capabilities are provided by the GEMAT and ARCHIVE functions, respectively. ESOPE, LU, GEMAT and ARCHIVE are all part of the ALOS software. This modular structure allows different modules to be linked together in procedures corresponding to recommended calculation routes ranging from fast-running and moderately-accurate 'routine' procedures to slow-running but highly-accurate 'reference' procedures. The main contents of the ERANOS-2.0 package are: nuclear data libraries (multigroup cross-sections from the JEF-2.2 evaluated nuclear data file, and other specific data files), a cell and lattice code (ECCO), reactor flux solvers (diffusion, Sn transport, nodal variational transport), a burn-up module, various processing modules (material and neutron balance, breeding gains,...), tools related to perturbation theory and sensitivity analysis, core follow-up modules (connected
Current status of the PSG Monte Carlo neutron transport code
Leppaenen, J.
2006-07-01
PSG is a new Monte Carlo neutron transport code, developed at the Technical Research Centre of Finland (VTT). The code is mainly intended for fuel assembly-level reactor physics calculations, such as group constant generation for deterministic reactor simulator codes. This paper presents the current status of the project and the essential capabilities of the code. Although the main application of PSG is in lattice calculations, the geometry is not restricted in two dimensions. This paper presents the validation of PSG against the experimental results of the three-dimensional MOX fuelled VENUS-2 reactor dosimetry benchmark. (authors)
IMPROVEMENTS IN CODED APERTURE THERMAL NEUTRON IMAGING.
VANIER,P.E.
2003-08-03
A new thermal neutron imaging system has been constructed, based on a 20-cm x 17-cm He-3 position-sensitive detector with spatial resolution better than 1 mm. New compact custom-designed position-decoding electronics are employed, as well as high-precision cadmium masks with Modified Uniformly Redundant Array patterns. Fast Fourier Transform algorithms are incorporated into the deconvolution software to provide rapid conversion of shadowgrams into real images. The system demonstrates the principles for locating sources of thermal neutrons by a stand-off technique, as well as visualizing the shapes of nearby sources. The data acquisition time could potentially be reduced two orders of magnitude by building larger detectors.
CALTRANS: A parallel, deterministic, 3D neutronics code
Carson, L.; Ferguson, J.; Rogers, J.
1994-04-01
Our efforts to parallelize the deterministic solution of the neutron transport equation has culminated in a new neutronics code CALTRANS, which has full 3D capability. In this article, we describe the layout and algorithms of CALTRANS and present performance measurements of the code on a variety of platforms. Explicit implementation of the parallel algorithms of CALTRANS using both the function calls of the Parallel Virtual Machine software package (PVM 3.2) and the Meiko CS-2 tagged message passing library (based on the Intel NX/2 interface) are provided in appendices.
Coded source neutron imaging at the PULSTAR reactor
Xiao, Ziyu; Mishra, Kaushal; Hawari, Ayman; Bingham, Philip R; Bilheux, Hassina Z; Tobin Jr, Kenneth William
2011-01-01
A neutron imaging facility is located on beam-tube No.5 of the 1-MW PULSTAR reactor at North Carolina State University. An investigation of high resolution imaging using the coded source imaging technique has been initiated at the facility. Coded imaging uses a mosaic of pinholes to encode an aperture, thus generating an encoded image of the object at the detector. To reconstruct the image data received by the detector, the corresponding decoding patterns are used. The optimized design of coded mask is critical for the performance of this technique and will depend on the characteristics of the imaging beam. In this work, a 34 x 38 uniformly redundant array (URA) coded aperture system is studied for application at the PULSTAR reactor neutron imaging facility. The URA pattern was fabricated on a 500 ?m gadolinium sheet. Simulations and experiments with a pinhole object have been conducted using the Gd URA and the optimized beam line.
Greene, N.M.; Ford, W.E. III; Petrie, L.M.; Arwood, J.W.
1992-10-01
AMPX-77 is a modular system of computer programs that pertain to nuclear analyses, with a primary emphasis on tasks associated with the production and use of multigroup cross sections. AH basic cross-section data are to be input in the formats used by the Evaluated Nuclear Data Files (ENDF/B), and output can be obtained in a variety of formats, including its own internal and very general formats, along with a variety of other useful formats used by major transport, diffusion theory, and Monte Carlo codes. Processing is provided for both neutron and gamma-my data. The present release contains codes all written in the FORTRAN-77 dialect of FORTRAN and wig process ENDF/B-V and earlier evaluations, though major modules are being upgraded in order to process ENDF/B-VI and will be released when a complete collection of usable routines is available.
A COUPLED TH/NEUTRONICS/CRUD FRAMEWORK IN PREDICTION OF CIPS PHENOMENON
Ling Zou; Hongbin Zhang; Jess Gehin; Brendan Kochunas
2012-04-01
A coupled TH/Neutronics/CRUD framework, which is able to simulate the CRUD deposits impact on CIPS phenomenon, was described in this paper. This framework includes the coupling among three essential physics, thermal-hydraulics, CRUD and neutronics. The overall framework was implemented by using the CFD software STAR-CCM+, developing CRUD codes, and using the neutronics code DeCART. The coupling was implemented by exchanging data between softwares using intermediate exchange files. A typical 3 by 3 PWR fuel pin problem was solved under this framework. The problem was solved in a 12 months length period of time. Time-dependent solutions were provided, including CRUD deposits inventory and their distributions on fuels, boron hideout amount inside CRUD deposits, as well as power shape changing over time. The results clearly showed the power shape suppression in regions where CRUD deposits exist, which is a strong indication of CIPS phenomenon.
An approach for coupled-code multiphysics core simulations from a common input
Schmidt, Rodney; Belcourt, Kenneth; Hooper, Russell; Pawlowski, Roger P.; Clarno, Kevin T.; Simunovic, Srdjan; Slattery, Stuart R.; Turner, John A.; Palmtag, Scott
2014-12-10
This study describes an approach for coupled-code multiphysics reactor core simulations that is being developed by the Virtual Environment for Reactor Applications (VERA) project in the Consortium for Advanced Simulation of Light-Water Reactors (CASL). In this approach a user creates a single problem description, called the “VERAIn” common input file, to define and setup the desired coupled-code reactor core simulation. A preprocessing step accepts the VERAIn file and generates a set of fully consistent input files for the different physics codes being coupled. The problem is then solved using a single-executable coupled-code simulation tool applicable to the problem, which is built using VERA infrastructure software tools and the set of physics codes required for the problem of interest. The approach is demonstrated by performing an eigenvalue and power distribution calculation of a typical three-dimensional 17 × 17 assembly with thermal–hydraulic and fuel temperature feedback. All neutronics aspects of the problem (cross-section calculation, neutron transport, power release) are solved using the Insilico code suite and are fully coupled to a thermal–hydraulic analysis calculated by the Cobra-TF (CTF) code. The single-executable coupled-code (Insilico-CTF) simulation tool is created using several VERA tools, including LIME (Lightweight Integrating Multiphysics Environment for coupling codes), DTK (Data Transfer Kit), Trilinos, and TriBITS. Parallel calculations are performed on the Titan supercomputer at Oak Ridge National Laboratory using 1156 cores, and a synopsis of the solution results and code performance is presented. Finally, ongoing development of this approach is also briefly described.
An approach for coupled-code multiphysics core simulations from a common input
Schmidt, Rodney; Belcourt, Kenneth; Hooper, Russell; Pawlowski, Roger P.; Clarno, Kevin T.; Simunovic, Srdjan; Slattery, Stuart R.; Turner, John A.; Palmtag, Scott
2014-12-10
This study describes an approach for coupled-code multiphysics reactor core simulations that is being developed by the Virtual Environment for Reactor Applications (VERA) project in the Consortium for Advanced Simulation of Light-Water Reactors (CASL). In this approach a user creates a single problem description, called the “VERAIn” common input file, to define and setup the desired coupled-code reactor core simulation. A preprocessing step accepts the VERAIn file and generates a set of fully consistent input files for the different physics codes being coupled. The problem is then solved using a single-executable coupled-code simulation tool applicable to the problem, which ismore » built using VERA infrastructure software tools and the set of physics codes required for the problem of interest. The approach is demonstrated by performing an eigenvalue and power distribution calculation of a typical three-dimensional 17 × 17 assembly with thermal–hydraulic and fuel temperature feedback. All neutronics aspects of the problem (cross-section calculation, neutron transport, power release) are solved using the Insilico code suite and are fully coupled to a thermal–hydraulic analysis calculated by the Cobra-TF (CTF) code. The single-executable coupled-code (Insilico-CTF) simulation tool is created using several VERA tools, including LIME (Lightweight Integrating Multiphysics Environment for coupling codes), DTK (Data Transfer Kit), Trilinos, and TriBITS. Parallel calculations are performed on the Titan supercomputer at Oak Ridge National Laboratory using 1156 cores, and a synopsis of the solution results and code performance is presented. Finally, ongoing development of this approach is also briefly described.« less
Wake coupling to full potential rotor analysis code
NASA Technical Reports Server (NTRS)
Torres, Francisco J.; Chang, I-Chung; Oh, Byung K.
1990-01-01
The wake information from a helicopter forward flight code is coupled with two transonic potential rotor codes. The induced velocities for the near-, mid-, and far-wake geometries are extracted from a nonlinear rigid wake of a standard performance and analysis code. These, together with the corresponding inflow angles, computation points, and azimuth angles, are then incorporated into the transonic potential codes. The coupled codes can then provide an improved prediction of rotor blade loading at transonic speeds.
3D Multigroup Sn Neutron Transport Code
2001-02-14
ATTILA is a 3D multigroup transport code with arbitrary order ansotropic scatter. The transport equation is solved in first order form using a tri-linear discontinuous spatial differencing on an arbitrary tetrahedral mesh. The overall solution technique is source iteration with DSA acceleration of the scattering source. Anisotropic boundary and internal sources may be entered in the form of spherical harmonics moments. Alpha and k eigenvalue problems are allowed, as well as fixed source problems. Forwardmore » and adjoint solutions are available. Reflective, vacumn, and source boundary conditions are available. ATTILA can perform charged particle transport calculations using slowing down (CSD) terms. ATTILA can also be used to peform infra-red steady-state calculations for radiative transfer purposes.« less
Coupled Monte Carlo neutronics and thermal hydraulics for power reactors
Bernnat, W.; Buck, M.; Mattes, M.; Zwermann, W.; Pasichnyk, I.; Velkov, K.
2012-07-01
The availability of high performance computing resources enables more and more the use of detailed Monte Carlo models even for full core power reactors. The detailed structure of the core can be described by lattices, modeled by so-called repeated structures e.g. in Monte Carlo codes such as MCNP5 or MCNPX. For cores with mainly uniform material compositions, fuel and moderator temperatures, there is no problem in constructing core models. However, when the material composition and the temperatures vary strongly a huge number of different material cells must be described which complicate the input and in many cases exceed code or memory limits. The second problem arises with the preparation of corresponding temperature dependent cross sections and thermal scattering laws. Only if these problems can be solved, a realistic coupling of Monte Carlo neutronics with an appropriate thermal-hydraulics model is possible. In this paper a method for the treatment of detailed material and temperature distributions in MCNP5 is described based on user-specified internal functions which assign distinct elements of the core cells to material specifications (e.g. water density) and temperatures from a thermal-hydraulics code. The core grid itself can be described with a uniform material specification. The temperature dependency of cross sections and thermal neutron scattering laws is taken into account by interpolation, requiring only a limited number of data sets generated for different temperatures. Applications will be shown for the stationary part of the Purdue PWR benchmark using ATHLET for thermal- hydraulics and for a generic Modular High Temperature reactor using THERMIX for thermal- hydraulics. (authors)
Interim report on fuel cycle neutronics code development.
Rabiti, C; Smith, M. A.; Kaushik, D.; Yang, W. S.
2008-05-13
As part of the Global Nuclear Energy Partnership (GNEP), a fast reactor simulation program was launched in April 2007 to develop a suite of modern simulation tools specifically for the analysis and design of sodium cooled fast reactors. The general goal of the new suite of codes is to reduce the uncertainties and biases in the various areas of reactor design activities by enhanced prediction capabilities. Under this fast reactor simulation program, a high-fidelity deterministic neutron transport code named UNIC is being developed. The final objective is to produce an integrated, advanced neutronics code that allows the high fidelity description of a nuclear reactor and simplifies the multi-step design process by direct coupling with thermal-hydraulics and structural mechanics calculations. Currently there are three solvers for the neutron transport code incorporated in UNIC: PN2ND, SN2ND, and MOCFE. PN2ND is based on a second-order even-parity spherical harmonics discretization of the transport equation and its primary target area of use is the existing homogenization approaches that are prevalent in reactor physics. MOCFE is based upon the method of characteristics applied to an unstructured finite element mesh and its primary target area of use is the fine grained nature of the explicit geometrical problems which is the long term goal of this project. SN2ND is based on a second-order, even-parity discrete ordinates discretization of the transport equation and its primary target area is the modeling transition region between the PN2ND and MOCFE solvers. The major development goal in fiscal year 2008 for the MOCFE solver was to include a two-dimensional capability that is scalable to hundreds of processors. The short term goal of this solver is to solve two-dimensional representations of reactor systems such that the energy and spatial self-shielding are accounted for and reliable cross sections can be generated for the homogeneous calculations. In this report we
A new pad-based neutron detector for stereo coded aperture thermal neutron imaging
NASA Astrophysics Data System (ADS)
Dioszegi, I.; Yu, B.; Smith, G.; Schaknowski, N.; Fried, J.; Vanier, P. E.; Salwen, C.; Forman, L.
2014-09-01
A new coded aperture thermal neutron imager system has been developed at Brookhaven National Laboratory. The cameras use a new type of position-sensitive 3He-filled ionization chamber, in which an anode plane is composed of an array of pads with independent acquisition channels. The charge is collected on each of the individual 5x5 mm2 anode pads, (48x48 in total, corresponding to 24x24 cm2 sensitive area) and read out by application specific integrated circuits (ASICs). The new design has several advantages for coded-aperture imaging applications in the field, compared to the previous generation of wire-grid based neutron detectors. Among these are its rugged design, lighter weight and use of non-flammable stopping gas. The pad-based readout occurs in parallel circuits, making it capable of high count rates, and also suitable to perform data analysis and imaging on an event-by-event basis. The spatial resolution of the detector can be better than the pixel size by using a charge sharing algorithm. In this paper we will report on the development and performance of the new pad-based neutron camera, describe a charge sharing algorithm to achieve sub-pixel spatial resolution and present the first stereoscopic coded aperture images of thermalized neutron sources using the new coded aperture thermal neutron imager system.
Calleja, M.; Stieglitz, R.; Sanchez, V.; Jimenez, J.; Imke, U.
2012-07-01
Multi-scale, multi-physics problems reveal significant challenges while dealing with coupled neutronic/thermal-hydraulic solutions. Current generation of codes applied to Light Water Reactors (LWR) are based on 3D neutronic nodal methods coupled with one or two phase flow thermal-hydraulic system or sub-channel codes. In addition, spatial meshing and temporal schemes are crucial for the proper description of the non-symmetrical core behavior in case of transient and accidents e.g. reactivity insertion accidents. This paper describes the coupling approach between the 3D neutron diffusion code COBAYA3 and the sub-channel code SUBCHANFLOW within SALOME. The coupling is done inside the SALOME open source platform that is characterized by a powerful pre- and post-processing capabilities and a novel functionality for mapping of the neutronic and thermal hydraulic domains. The peculiar functionalities of SALOME and the steps required for the code integration and coupling are presented. The validation of the coupled codes is done based on two benchmarks the PWR MOX/UO{sub 2} RIA and the TMI-1 MSLB benchmark. A discussion of the prediction capability of COBAYA3/SUBCHANFLOW compared to other coupled solutions will be provided too. (authors)
Bragg optics computer codes for neutron scattering instrument design
Popovici, M.; Yelon, W.B.; Berliner, R.R.; Stoica, A.D.
1997-09-01
Computer codes for neutron crystal spectrometer design, optimization and experiment planning are described. Phase space distributions, linewidths and absolute intensities are calculated by matrix methods in an extension of the Cooper-Nathans resolution function formalism. For modeling the Bragg reflection on bent crystals the lamellar approximation is used. Optimization is done by satisfying conditions of focusing in scattering and in real space, and by numerically maximizing figures of merit. Examples for three-axis and two-axis spectrometers are given.
A thermal neutron source imager using coded apertures
Vanier, P.E.; Forman, L.; Selcow, E.C.
1995-08-01
To facilitate the process of re-entry vehicle on-site inspections, it would be useful to have an imaging technique which would allow the counting of deployed multiple nuclear warheads without significant disassembly of a missile`s structure. Since neutrons cannot easily be shielded without massive amounts of materials, they offer a means of imaging the separate sources inside a sealed vehicle. Thermal neutrons carry no detailed spectral information, so their detection should not be as intrusive as gamma ray imaging. A prototype device for imaging at close range with thermal neutrons has been constructed using an array of {sup 3}He position-sensitive gas proportional counters combined with a uniformly redundant coded aperture array. A sealed {sup 252}Cf source surrounded by a polyethylene moderator is used as a test source. By means of slit and pinhole experiments, count rates of image-forming neutrons (those which cast a shadow of a Cd aperture on the detector) are compared with the count rates for background neutrons. The resulting ratio, which limits the available image contrast, is measured as a function of distance from the source. The envelope of performance of the instrument is defined by the contrast ratio, the angular resolution, and the total count rate as a function of distance from the source. These factors will determine whether such an instrument could be practical as a tool for treaty verification.
Development of a coupling code for PWR reactor cavity radiation streaming calculation
Zheng, Z.; Wu, H.; Cao, L.; Zheng, Y.; Zhang, H.; Wang, M.
2012-07-01
PWR reactor cavity radiation streaming is important for the safe of the personnel and equipment, thus calculation has to be performed to evaluate the neutron flux distribution around the reactor. For this calculation, the deterministic codes have difficulties in fine geometrical modeling and need huge computer resource; and the Monte Carlo codes require very long sampling time to obtain results with acceptable precision. Therefore, a coupling method has been developed to eliminate the two problems mentioned above in each code. In this study, we develop a coupling code named DORT2MCNP to link the Sn code DORT and Monte Carlo code MCNP. DORT2MCNP is used to produce a combined surface source containing top, bottom and side surface simultaneously. Because SDEF card is unsuitable for the combined surface source, we modify the SOURCE subroutine of MCNP and compile MCNP for this application. Numerical results demonstrate the correctness of the coupling code DORT2MCNP and show reasonable agreement between the coupling method and the other two codes (DORT and MCNP). (authors)
Interface requirements to couple thermal hydraulics codes to severe accident codes: ICARE/CATHARE
Camous, F.; Jacq, F.; Chatelard, P.
1997-07-01
In order to describe with the same code the whole sequence of severe LWR accidents, up to the vessel failure, the Institute of Protection and Nuclear Safety has performed a coupling of the severe accident code ICARE2 to the thermalhydraulics code CATHARE2. The resulting code, ICARE/CATHARE, is designed to be as pertinent as possible in all the phases of the accident. This paper is mainly devoted to the description of the ICARE2-CATHARE2 coupling.
Code Development in Coupled PARCS/RELAP5 for Supercritical Water Reactor
Hu, Po; Wilson, Paul
2014-01-01
The new capability is added to the existing coupled code package PARCS/RELAP5, in order to analyze SCWR design under supercritical pressure with the separated water coolant and moderator channels. This expansion is carried out on both codes. In PARCS, modification is focused on extending the water property tables to supercritical pressure, modifying the variable mapping input file and related code module for processing thermal-hydraulic information from separated coolant/moderator channels, and modifying neutronics feedback module to deal with the separated coolant/moderator channels. In RELAP5, modification is focused on incorporating more accurate water properties near SCWR operation/transient pressure and temperature in themore » code. Confirming tests of the modifications is presented and the major analyzing results from the extended codes package are summarized.« less
Valtavirta, V.
2013-07-01
This paper describes the unique way of simultaneously solving the power and temperature distributions of a nuclear system with the Monte Carlo neutron transport code Serpent 2. The coupled solution is achieved through the implementation of an internal temperature solver and material property correlations in the code. The program structure is reviewed concerning the temperature solver and the internal correlations as well as the internal coupling between these two and the neutron transport part. To estimate the reactivity differences resulting from correlation choices a simple pin-cell case has been calculated. It is established, that some correlation choices may result in difference in reactivity of approximately 100 pcm. (authors)
TART. Coupled Neutron & Photon MC Transport
Plechaty, E.F.
1988-10-06
TART is a three-dimensional, data-dependent Monte Carlo transport program. The program calculates the transport of neutrons, photons, and neutron-induced photons through zones described by algebraic functions. The zones and elements to be included are user-specified. Any one of 21 different output tallys (methods of calculating particle transport) may be selected for each zone. A spectral reflection tally, which calculates reflections from planes and quadratic surfaces, saves considerable time and effort for some classes of problems. The neutron and photon energy deposition output tally is included in all TART calculations. The neutron and gamma-ray production cross sections are specified from 10E-9 MeV to 20 MeV. The gamma-ray interaction cross sections are specified from 10E-4 MeV to 30 MeV. The three cross section libraries are provided in binary form. Variance reduction methods included are splitting and Russian roulette at zone boundaries. Each zone in the problem can be assigned a weight.
Coupled-Channel Computation of Direct Neutron Capture on Non-Spherical Nuclei
NASA Astrophysics Data System (ADS)
Arbanas, Goran; Thompson, Ian; Escher, Jutta; Nunes, Filomena; Elster, Charlotte; Zhang, Shi-Sheng
2014-09-01
Models of direct neutron capture of neutrons have so far accounted for the effects of non-spherical nuclei either in the incoming wave functions (via non-spherical optical model potentials), or in the final bound states (via non-spherical real potential wells), but not in both. Since it is known that spherical optical potentials do not give a good reproduction of low energy neutron-scattering observables of deformed nuclei, we have performed calculations in which the initial and final states are both treated in a self-consistent, non-spherical-nucleus picture. We have done this in the coupled-channels model of nuclear reactions implemented in the FRESCO code by using the same deformation-length for the couplings to the rotational-band states in the incoming and the final state configurations. We compute direct capture using this method for even-mass calcium isotopes 40 , 42 , 44 , 46 , 48Ca to study the effect across the two closed neutron shells, for neutron-rich even-mass tin isotopes relevant to models of astrophysical nucleosynthesis, and for 56Fe that is an important structural material used in nuclear applications. Models of direct neutron capture of neutrons have so far accounted for the effects of non-spherical nuclei either in the incoming wave functions (via non-spherical optical model potentials), or in the final bound states (via non-spherical real potential wells), but not in both. Since it is known that spherical optical potentials do not give a good reproduction of low energy neutron-scattering observables of deformed nuclei, we have performed calculations in which the initial and final states are both treated in a self-consistent, non-spherical-nucleus picture. We have done this in the coupled-channels model of nuclear reactions implemented in the FRESCO code by using the same deformation-length for the couplings to the rotational-band states in the incoming and the final state configurations. We compute direct capture using this method for even
Coupled Neutron Transport and Thermal Fluids Calculations for the VHTR
NASA Astrophysics Data System (ADS)
Connolly, Kevin John; Huning, Alexander J.; Rahnema, Farzad; Garimella, Srinivas
2014-06-01
A new multiphysics method is presented for coupled neutronics and thermal fluids calculations in the VHTR. This new method combines the capabilities of two existing solvers: the COMET neutronics solver, and a thermal fluids module designed specifically for the gas-cooled reactor lattice design featured in this next-generation reactor. This paper provides the necessary background on the neutronics and thermal fluids aspects of the new solution strategy and explains the mode of coupling. A test problem is presented in order to prove the efficacy of the new coupled method. Results are given for a whole-core VHTR, including detailed temperature information at the fuel pin level, explicit power calculations of individual pins, and a thermal map of the bulk graphite and coolant temperatures throughout the core. Solutions are determined quickly when compared to other methods which offer the same level of detail and accuracy, and thus justify further research and development of this method in the near future.
Bousbia-Salah, Anis; Vedovi, Juswald; D'Auria, Francesco; Ivanov, Kostadin; Galassi, Giorgio
2004-10-15
Thanks to continuous progress in computer technology, it is now possible to perform best-estimate simulations of complex scenarios in nuclear power plants. This method is carried out through the coupling of three-dimensional (3-D) neutron modeling of a reactor core into system codes. It is particularly appropriate for transients that involve strong interactions between core neutronics and reactor loop thermal hydraulics. For this purpose, the Peach Bottom boiling water reactor turbine trip test was selected to challenge the capability of such coupled codes. The test is characterized by a power excursion induced by rapid core pressurization and a self-limiting course behavior. In order to perform the closest simulation, the coupled thermal-hydraulic system code RELAP5 and 3-D neutron kinetic code PARCS were used. The obtained results are compared to those available from experimental data. Overall, the coupled code calculations globally predict the most significant observed aspects of the transient, such as the pressure wave amplitude across the core and the power course, with an acceptable agreement. However, sensitivity studies revealed that more-accurate code models should be considered in order to better match the void dynamic and the cross-section variations during transient conditions.
NASA Astrophysics Data System (ADS)
Kimura, T.; Iwakiri, W. B.; Enoto, T.; Wada, T.; Tao, C.
2015-12-01
In the binary neutron star system, angular momentum transfer from accretion disk to a star is essential process for spin-up/down of stars. The angular momentum transfer has been well formulated for the accretion disk strongly magnetized by the neutron star [e.g., Ghosh and Lamb, 1978, 1979a, b]. However, the electromagnetic (EM) coupling between the neutron star and accretion disk has not been self-consistently solved in the previous studies although the magnetic field lines from the star are strongly tied with the accretion disk. In this study, we applied the planet-magnetosphere coupling process established for Jupiter [Hill, 1979] to the binary neutron star system. Angular momentum distribution is solved based on the torque balance between the neutron star's surface and accretion disk coupled by the magnetic field tensions. We found the EM coupling can transfer significantly larger fraction of the angular momentum from the magnetized accretion disk to the star than the unmagnetized case. The resultant spin-up rate is estimated to ~10^-14 [sec/sec] for the nominal binary system parameters, which is comparable with or larger than the other common spin-down/up processes: e.g., the magnetic dipole radiation spin-down. The Joule heating energy dissipated in the EM coupling is estimated to be up to ~10^36 [erg/sec] for the nominal binary system parameters. The release is comparable to that of gravitation energy directly caused by the matters accreting onto the neutron star. This suggests the EM coupling at the neutron star can accompany the observable radiation as auroras with a similar manner to those at the rotating planetary magnetospheres like Jupiter, Saturn, and other gas giants.
Neutron Activation Analysis PRognosis and Optimization Code System.
2004-08-20
Version 00 NAAPRO predicts the results and main characteristics (detection limits, determination limits, measurement limits and relative precision of the analysis) of neutron activation analysis (instrumental and radiochemical). Gamma-ray dose rates for different points of time after sample irradiation and input count rate of the spectrometry system are also predicted. The code uses standard Windows user interface and extensive graphical tools for the visualization of the spectrometer characteristics (efficiency, response and background) and simulated spectrum.more » Optimization part is not included in the current version of the code. This release is designated NAAPRO, Version 01.beta. The MCNP code was used for generating detector responses. PREPRO-2000 and FCONV programs were used at the preparation of the program nuclear databases. A special program was developed for viewing, editing and updating of the program databases (not included into the present program package). The MCNP, PREPRO-2000 and FCONV software packages are not included in the NAAPRO package.« less
Neutron-induced background in charge-coupled device detectors
Jaanimagi, P. A.; Boni, R.; Keck, R. L.
2001-01-01
The inertial confinement fusion (ICF) community must become more cognizant of the neutron-induced background levels in charge-coupled device (CCD) detectors that are replacing film as the recording medium in many ICF diagnostics. This background degrades the signal-to-noise ratio (SNR) of the recorded signals and for the highest-yield shots comprises a substantial fraction of the pixel's full well capacity. CCD detectors located anywhere in the OMEGA Target Bay are precluded from recording high precision signals (SNR>30) for deuterium--tritium neutron yields greater than 10{sup 13}. CCDs make excellent calibrated neutron detectors. The average CCD background level is proportional to the neutron yield, and we have measured a linear response over four decades. The spectrum of deposited energy per pixel is heavily weighted to low energies, <50 keV, with a few isolated saturated pixels. Most of the background recorded by the CCDs is due to secondary radiation produced by interactions of the primary neutrons with all the materials in the Target Bay as well as the shield walls and the floor. Since the noise source comes from all directions it is very difficult to shield. The fallback position of using film instead of CCD cameras for high-neutron-yield target shots is flawed, as we have observed substantially increased fog levels on our x-ray recording film as a function of the neutron yield.
Coded aperture Fast Neutron Analysis: Latest design advances
NASA Astrophysics Data System (ADS)
Accorsi, Roberto; Lanza, Richard C.
2001-07-01
Past studies have showed that materials of concern like explosives or narcotics can be identified in bulk from their atomic composition. Fast Neutron Analysis (FNA) is a nuclear method capable of providing this information even when considerable penetration is needed. Unfortunately, the cross sections of the nuclear phenomena and the solid angles involved are typically small, so that it is difficult to obtain high signal-to-noise ratios in short inspection times. CAFNAaims at combining the compound specificity of FNA with the potentially high SNR of Coded Apertures, an imaging method successfully used in far-field 2D applications. The transition to a near-field, 3D and high-energy problem prevents a straightforward application of Coded Apertures and demands a thorough optimization of the system. In this paper, the considerations involved in the design of a practical CAFNA system for contraband inspection, its conclusions, and an estimate of the performance of such a system are presented as the evolution of the ideas presented in previous expositions of the CAFNA concept.
Liu Yueqiang
2008-07-15
In order to model a magnetohydrodynamic (MHD) instability that strongly couples to external conducting structures (walls and/or coils) in a fusion device, it is often necessary to combine a MHD code solving for the plasma response, with an eddy current code computing the fields and currents of conductors. We present a rigorous proof of the coupling schemes between these two types of codes. One of the coupling schemes has been introduced and implemented in the CARMA code [R. Albanese, Y. Q. Liu, A. Portone, G. Rubinacci, and F. Villone, IEEE Trans. Magn. 44, 1654 (2008); A. Portone, F. Villone, Y. Q. Liu, R. Albanese, and G. Rubinacci, Plasma Phys. Controlled Fusion 50, 085004 (2008)] that couples the MHD code MARS-F[Y. Q. Liu, A. Bondeson, C. M. Fransson, B. Lennartson, and C. Breitholtz, Phys. Plasmas 7, 3681 (2000)] and the eddy current code CARIDDI[R. Albanese and G. Rubinacci, Adv. Imaging Electron Phys. 102, 1 (1998)]. While the coupling schemes are described for a general toroidal geometry, we give the analytical proof for a cylindrical plasma.
Cheng, H.S.; Rohatgi, U.S.
1993-06-01
An investigation was made of the potential for thermal-hydraulic instabilities coupled to neutronic feedback in a BWR due to a two recirculation pump trip event using the RAMONA-4B computer code with 3D neutron kinetics. It is concluded that a high-power (100%) and low-flow (75%) initial condition would most likely lead to in-phase density wave oscillations after the tripping of both recirculation pumps, and that RAMONA-4B is capable of predicting such thermal-hydraulic instabilities coupled to neutronic feedback in BWR and in SBWR.
Coupled full core neutron transport/CFD simulations of pressurized water reactors
Kochunas, B.; Stimpson, S.; Collins, B.; Downar, T.; Brewster, R.; Baglietto, E.; Yan, J.
2012-07-01
Recently as part of the CASL project, a capability to perform 3D whole-core coupled neutron transport and computational fluid dynamics (CFD) calculations was demonstrated. This work uses the 2D/1D transport code DeCART and the commercial CFD code STAR-CCM+. It builds on previous CASL work demonstrating coupling for smaller spatial domains. The coupling methodology is described along with the problem simulated and results are presented for fresh hot full power conditions. An additional comparison is made to an equivalent model that uses lower order T/H feedback to assess the importance and cost of high fidelity feedback to the neutronics problem. A simulation of a quarter core Combustion Engineering (CE) PWR core was performed with the coupled codes using a Fixed Point Gauss-Seidel iteration technique. The total approximate calculation requirements are nearly 10,000 CPU hours and 1 TB of memory. The problem took 6 coupled iterations to converge. The CFD coupled model and low order T/H feedback model compared well for global solution parameters, with a difference in the critical boron concentration and average outlet temperature of 14 ppm B and 0.94 deg. C, respectively. Differences in the power distribution were more significant with maximum relative differences in the core-wide pin peaking factor (Fq) of 5.37% and average relative differences in flat flux region power of 11.54%. Future work will focus on analyzing problems more relevant to CASL using models with less approximations. (authors)
Strongly coupled chameleons and the neutronic quantum bouncer.
Brax, Philippe; Pignol, Guillaume
2011-09-01
We consider the potential detection of chameleons using bouncing ultracold neutrons. We show that the presence of a chameleon field over a planar plate would alter the energy levels of ultracold neutrons in the terrestrial gravitational field. When chameleons are strongly coupled to nuclear matter, β≳10(8), we find that the shift in energy levels would be detectable with the forthcoming GRANIT experiment, where a sensitivity of the order of 1% of a peV is expected. We also find that an extremely large coupling β≳10(11) would lead to new bound states at a distance of order 2 μm, which is already ruled out by previous Grenoble experiments. The resulting bound, β≲10(11), is already 3 orders of magnitude better than the upper bound, β≲10(14), from precision tests of atomic spectra.
Strongly Coupled Chameleons and the Neutronic Quantum Bouncer
NASA Astrophysics Data System (ADS)
Brax, Philippe; Pignol, Guillaume
2011-09-01
We consider the potential detection of chameleons using bouncing ultracold neutrons. We show that the presence of a chameleon field over a planar plate would alter the energy levels of ultracold neutrons in the terrestrial gravitational field. When chameleons are strongly coupled to nuclear matter, β≳108, we find that the shift in energy levels would be detectable with the forthcoming GRANIT experiment, where a sensitivity of the order of 1% of a peV is expected. We also find that an extremely large coupling β≳1011 would lead to new bound states at a distance of order 2μm, which is already ruled out by previous Grenoble experiments. The resulting bound, β≲1011, is already 3 orders of magnitude better than the upper bound, β≲1014, from precision tests of atomic spectra.
Measurement of the spin-rotation coupling in neutron polarimetry
NASA Astrophysics Data System (ADS)
Demirel, Bülent; Sponar, Stephan; Hasegawa, Yuji
2015-02-01
The effect of spin-rotation coupling is measured for the first time with neutrons. The coupling of spin with the angular velocity of a rotating spin turner can be observed as a phase shift in a neutron polarimeter set-up. After the neutron’s spin is rotated by 2π through a rotating magnetic field, different phase shifts are induced for ‘up’ and ‘down’ spin eigenstates. This phase difference results in the rotation of the neutron’s spin-vector, which turns out to depend solely on the frequency of the rotation of the magnetic field. The experimental results agree well with the solutions acquired by the Pauli-Schrödinger equation.
An example of neutronic penalizations in reactivity transient analysis using 3D coupled chain HEMERA
Dubois, F.; Normand, B.; Sargeni, A.
2012-07-01
HEMERA (Highly Evolutionary Methods for Extensive Reactor Analyses), is a fully coupled 3D computational chain developed jointly by IRSN and CEA. It is composed of CRONOS2 (core neutronics, cross sections library from APOLLO2), FLICA4 (core thermal-hydraulics) and the system code CATHARE. Multi-level and multi-dimensional models are developed to account for neutronics, core thermal-hydraulics, fuel thermal analysis and system thermal-hydraulics, dedicated to best-estimate, conservative simulations and sensitivity analysis. In IRSN, the HEMERA chain is widely used to study several types of reactivity accidents and for sensitivity studies. Just as an example of the HEMERA possibilities, we present here two types of neutronic penalizations and their impact on a power transient due to a REA (Rod Ejection Accident): in the first one, we studied a bum-up distribution modification and in the second one, a delayed-neutron fraction modification. Both modifications are applied to the whole core or localized in a few assemblies. Results show that it is possible to use global or local changes but 1) in case of bum-up modification, the total core power can increase when assembly peak power decrease so, care has to be taken if the goal is to maximize a local power peak and 2) for delayed-neutron fraction, a local modification can have the same effect as the one on the whole core, provided that it is large enough. (authors)
1991-08-01
Version: 00 The original MORSE code was a multipurpose neutron and gamma-ray transport Monte Carlo code. It was designed as a tool for solving most shielding problems. Through the use of multigroup cross sections, the solution of neutron, gamma-ray, or coupled neutron-gamma-ray problems could be obtained in either the forward or adjoint mode. Time dependence for both shielding and criticality problems is provided. General three-dimensional geometry could be used with an albedo option available atmore » any material surface. Isotropic or anisotropic scattering up to a P16 expansion of the angular distribution was allowed. MORSE-CG incorporated the Mathematical Applications, Inc. (MAGI) combinatorial geometry routines. MORSE-B modifies the Monte Carlo neutron and photon transport computer code MORSE-CG by adding routines which allow various flexible options.« less
A novel approach to correct the coded aperture misalignment for fast neutron imaging
Zhang, F. N.; Hu, H. S. Wang, D. M.; Jia, J.; Zhang, T. K.; Jia, Q. G.
2015-12-15
Aperture alignment is crucial for the diagnosis of neutron imaging because it has significant impact on the coding imaging and the understanding of the neutron source. In our previous studies on the neutron imaging system with coded aperture for large field of view, “residual watermark,” certain extra information that overlies reconstructed image and has nothing to do with the source is discovered if the peak normalization is employed in genetic algorithms (GA) to reconstruct the source image. Some studies on basic properties of residual watermark indicate that the residual watermark can characterize coded aperture and can thus be used to determine the location of coded aperture relative to the system axis. In this paper, we have further analyzed the essential conditions for the existence of residual watermark and the requirements of the reconstruction algorithm for the emergence of residual watermark. A gamma coded imaging experiment has been performed to verify the existence of residual watermark. Based on the residual watermark, a correction method for the aperture misalignment has been studied. A multiple linear regression model of the position of coded aperture axis, the position of residual watermark center, and the gray barycenter of neutron source with twenty training samples has been set up. Using the regression model and verification samples, we have found the position of the coded aperture axis relative to the system axis with an accuracy of approximately 20 μm. Conclusively, a novel approach has been established to correct the coded aperture misalignment for fast neutron coded imaging.
A neutron spectrum unfolding computer code based on artificial neural networks
NASA Astrophysics Data System (ADS)
Ortiz-Rodríguez, J. M.; Reyes Alfaro, A.; Reyes Haro, A.; Cervantes Viramontes, J. M.; Vega-Carrillo, H. R.
2014-02-01
The Bonner Spheres Spectrometer consists of a thermal neutron sensor placed at the center of a number of moderating polyethylene spheres of different diameters. From the measured readings, information can be derived about the spectrum of the neutron field where measurements were made. Disadvantages of the Bonner system are the weight associated with each sphere and the need to sequentially irradiate the spheres, requiring long exposure periods. Provided a well-established response matrix and adequate irradiation conditions, the most delicate part of neutron spectrometry, is the unfolding process. The derivation of the spectral information is not simple because the unknown is not given directly as a result of the measurements. The drawbacks associated with traditional unfolding procedures have motivated the need of complementary approaches. Novel methods based on Artificial Intelligence, mainly Artificial Neural Networks, have been widely investigated. In this work, a neutron spectrum unfolding code based on neural nets technology is presented. This code is called Neutron Spectrometry and Dosimetry with Artificial Neural networks unfolding code that was designed in a graphical interface. The core of the code is an embedded neural network architecture previously optimized using the robust design of artificial neural networks methodology. The main features of the code are: easy to use, friendly and intuitive to the user. This code was designed for a Bonner Sphere System based on a 6LiI(Eu) neutron detector and a response matrix expressed in 60 energy bins taken from an International Atomic Energy Agency compilation. The main feature of the code is that as entrance data, for unfolding the neutron spectrum, only seven rate counts measured with seven Bonner spheres are required; simultaneously the code calculates 15 dosimetric quantities as well as the total flux for radiation protection purposes. This code generates a full report with all information of the unfolding in
Optimal control of coupled PDE networks with automated code generation
NASA Astrophysics Data System (ADS)
Papadopoulos, D.
2012-09-01
The purpose of this work is to present a framework for the optimal control of coupled PDE networks. A coupled PDE network is a system of partial differential equations coupled together. Such systems can be represented as a directed graph. A domain specific language (DSL)—an extension of the DOT language—is used for the description of such a coupled PDE network. The adjoint equations and the gradient, required for its optimal control, are computed with the help of a computer algebra system (CAS). Automated code generation techniques have been used for the generation of the PDE systems of both the direct and the adjoint equations. Both the direct and adjoint equations are solved with the standard finite element method. Finally, for the numerical optimization of the system standard optimization techniques are used such as BFGS and Newton conjugate gradient.
Neutron Transport Models and Methods for HZETRN and Coupling to Low Energy Light Ion Transport
NASA Technical Reports Server (NTRS)
Blattnig, S.R.; Slaba, T.C.; Heinbockel, J.H.
2008-01-01
Exposure estimates inside space vehicles, surface habitats, and high altitude aircraft exposed to space radiation are highly influenced by secondary neutron production. The deterministic transport code HZETRN has been identified as a reliable and efficient tool for such studies, but improvements to the underlying transport models and numerical methods are still necessary. In this paper, the forward-backward (FB) and directionally coupled forward-backward (DC) neutron transport models are derived, numerical methods for the FB model are reviewed, and a computationally efficient numerical solution is presented for the DC model. Both models are compared to the Monte Carlo codes HETCHEDS and FLUKA, and the DC model is shown to agree closely with the Monte Carlo results. Finally, it is found in the development of either model that the decoupling of low energy neutrons from the light ion (A<4) transport procedure adversely affects low energy light ion fluence spectra and exposure quantities. A first order correction is presented to resolve the problem, and it is shown to be both accurate and efficient.
Validation of a coupled reactive transport code in porous media
NASA Astrophysics Data System (ADS)
Mugler, C.; Montarnal, P.; Dimier, A.
2003-04-01
The safety assessment of nuclear waste disposals needs to predict the migration of radionuclides and chemical species through a geological medium. It is therefore necessary to develop and assess qualified and validated tools which integrate both the transport mechanisms through the geological media and the chemical mechanisms governing the mobility of radionuclides. In this problem, both geochemical and hydrodynamic phenomena are tightly linked together. That is the reason why the French Nuclear Energy Agency (CEA) and the French Agency for the Management of Radioactive Wastes (ANDRA) are conjointly developping a coupled reactive transport code that solves simultaneously a geochemical model and a transport model. This code, which is part of the software project ALLIANCES, leans on the libraries of two geochemical codes solving the complex ensemble of reacting chemical species: CHESS and PHREEQC. Geochemical processes considered here include ion exchange, redox reactions, acid-base reactions, surface complexation and mineral dissolution and/or precipitation. Transport is simulated using the mixed-hybrid finite element scheme CAST3M or the finite volume scheme MT3D. All together solve Darcy's law and simulate several hydrological processes such as advection, diffusion and dispersion. The coupling algorithm is an iterative sequential algorithm. Several analytical test cases have been defined and used to validate the reactive transport code. Numerical results can be compared to analytical solutions.
Borehole parametric study for neutron induced capture gamma-ray spectrometry using the MCNP code.
Shahriari, M; Sohrabpour, M
2000-01-01
The MCNP Monte Carlo code has been used to simulate neutron transport from an Am-Be source into a granite formation surrounding a borehole. The effects of the moisture and the neutron poison on the thermal neutron flux distribution and the capture by the absorbing elements has been calculated. Thermal and nonthermal captures for certain absorbers having resonance structures in the epithermal and fast energy regions such as W and Si were performed. It is shown that for those absorbers having large resonances in the epithermal regions when they are present in dry formation or when accompanied by neutron poisons the resonance captures may be significant compared to the thermal captures.
Interface requirements to couple thermal-hydraulic codes to severe accident codes: ATHLET-CD
Trambauer, K.
1997-07-01
The system code ATHLET-CD is being developed by GRS in cooperation with IKE and IPSN. Its field of application comprises the whole spectrum of leaks and large breaks, as well as operational and abnormal transients for LWRs and VVERs. At present the analyses cover the in-vessel thermal-hydraulics, the early phases of core degradation, as well as fission products and aerosol release from the core and their transport in the Reactor Coolant System. The aim of the code development is to extend the simulation of core degradation up to failure of the reactor pressure vessel and to cover all physically reasonable accident sequences for western and eastern LWRs including RMBKs. The ATHLET-CD structure is highly modular in order to include a manifold spectrum of models and to offer an optimum basis for further development. The code consists of four general modules to describe the reactor coolant system thermal-hydraulics, the core degradation, the fission product core release, and fission product and aerosol transport. Each general module consists of some basic modules which correspond to the process to be simulated or to its specific purpose. Besides the code structure based on the physical modelling, the code follows four strictly separated steps during the course of a calculation: (1) input of structure, geometrical data, initial and boundary condition, (2) initialization of derived quantities, (3) steady state calculation or input of restart data, and (4) transient calculation. In this paper, the transient solution method is briefly presented and the coupling methods are discussed. Three aspects have to be considered for the coupling of different modules in one code system. First is the conservation of masses and energy in the different subsystems as there are fluid, structures, and fission products and aerosols. Second is the convergence of the numerical solution and stability of the calculation. The third aspect is related to the code performance, and running time.
Thermal neutron response of a boron-coated GEM detector via GEANT4 Monte Carlo code.
Jamil, M; Rhee, J T; Kim, H G; Ahmad, Farzana; Jeon, Y J
2014-10-22
In this work, we report the design configuration and the performance of the hybrid Gas Electron Multiplier (GEM) detector. In order to make the detector sensitive to thermal neutrons, the forward electrode of the GEM has been coated with the enriched boron-10 material, which works as a neutron converter. A total of 5×5cm(2) configuration of GEM has been used for thermal neutron studies. The response of the detector has been estimated via using GEANT4 MC code with two different physics lists. Using the QGSP_BIC_HP physics list, the neutron detection efficiency was determined to be about 3%, while with QGSP_BERT_HP physics list the efficiency was around 2.5%, at the incident thermal neutron energies of 25meV. The higher response of the detector proves that GEM-coated with boron converter improves the efficiency for thermal neutrons detection.
Current and anticipated uses of thermalhydraulic and neutronic codes at PSI
Aksan, S.N.; Zimmermann, M.A.; Yadigaroglu, G.
1997-07-01
The thermalhydraulic and/or neutronic codes in use at PSI mainly provide the capability to perform deterministic safety analysis for Swiss NPPs and also serve as analysis tools for experimental facilities for LWR and ALWR simulations. In relation to these applications, physical model development and improvements, and assessment of the codes are also essential components of the activities. In this paper, a brief overview is provided on the thermalhydraulic and/or neutronic codes used for safety analysis of LWRs, at PSI, and also of some experiences and applications with these codes. Based on these experiences, additional assessment needs are indicated, together with some model improvement needs. The future needs that could be used to specify both the development of a new code and also improvement of available codes are summarized.
1991-05-01
Version 00 MORSE-CGA was developed to add the capability of modelling rectangular lattices for nuclear reactor cores or for multipartitioned structures. It thus enhances the capability of the MORSE code system. The MORSE code is a multipurpose neutron and gamma-ray transport Monte Carlo code. It has been designed as a tool for solving most shielding problems. Through the use of multigroup cross sections, the solution of neutron, gamma-ray, or coupled neutron-gamma-ray problems may be obtainedmore » in either the forward or adjoint mode. Time dependence for both shielding and criticality problems is provided. General three-dimensional geometry may be used with an albedo option available at any material surface. Isotropic or anisotropic scattering up to a P16 expansion of the angular distribution is allowed.« less
Cooperative solutions coupling a geometry engine and adaptive solver codes
NASA Technical Reports Server (NTRS)
Dickens, Thomas P.
1995-01-01
Follow-on work has progressed in using Aero Grid and Paneling System (AGPS), a geometry and visualization system, as a dynamic real time geometry monitor, manipulator, and interrogator for other codes. In particular, AGPS has been successfully coupled with adaptive flow solvers which iterate, refining the grid in areas of interest, and continuing on to a solution. With the coupling to the geometry engine, the new grids represent the actual geometry much more accurately since they are derived directly from the geometry and do not use refits to the first-cut grids. Additional work has been done with design runs where the geometric shape is modified to achieve a desired result. Various constraints are used to point the solution in a reasonable direction which also more closely satisfies the desired results. Concepts and techniques are presented, as well as examples of sample case studies. Issues such as distributed operation of the cooperative codes versus running all codes locally and pre-calculation for performance are discussed. Future directions are considered which will build on these techniques in light of changing computer environments.
The Application Programming Interface for the PVMEXEC Program and Associated Code Coupling System
Walter L. Weaver III
2005-03-01
This report describes the Application Programming Interface for the PVMEXEC program and the code coupling systems that it implements. The information in the report is intended for programmers wanting to add a new code into the coupling system.
Spallation neutron production and the current intra-nuclear cascade and transport codes
NASA Astrophysics Data System (ADS)
Filges, D.; Goldenbaum, F.; Enke, M.; Galin, J.; Herbach, C.-M.; Hilscher, D.; Jahnke, U.; Letourneau, A.; Lott, B.; Neef, R.-D.; Nünighoff, K.; Paul, N.; Péghaire, A.; Pienkowski, L.; Schaal, H.; Schröder, U.; Sterzenbach, G.; Tietze, A.; Tishchenko, V.; Toke, J.; Wohlmuther, M.
A recent renascent interest in energetic proton-induced production of neutrons originates largely from the inception of projects for target stations of intense spallation neutron sources, like the planned European Spallation Source (ESS), accelerator-driven nuclear reactors, nuclear waste transmutation, and also from the application for radioactive beams. In the framework of such a neutron production, of major importance is the search for ways for the most efficient conversion of the primary beam energy into neutron production. Although the issue has been quite successfully addressed experimentally by varying the incident proton energy for various target materials and by covering a huge collection of different target geometries --providing an exhaustive matrix of benchmark data-- the ultimate challenge is to increase the predictive power of transport codes currently on the market. To scrutinize these codes, calculations of reaction cross-sections, hadronic interaction lengths, average neutron multiplicities, neutron multiplicity and energy distributions, and the development of hadronic showers are confronted with recent experimental data of the NESSI collaboration. Program packages like HERMES, LCS or MCNPX master the prevision of reaction cross-sections, hadronic interaction lengths, averaged neutron multiplicities and neutron multiplicity distributions in thick and thin targets for a wide spectrum of incident proton energies, geometrical shapes and materials of the target generally within less than 10% deviation, while production cross-section measurements for light charged particles on thin targets point out that appreciable distinctions exist within these models.
Interface requirements for coupling a containment code to a reactor system thermal hydraulic codes
Baratta, A.J.
1997-07-01
To perform a complete analysis of a reactor transient, not only the primary system response but the containment response must also be accounted for. Such transients and accidents as a loss of coolant accident in both pressurized water and boiling water reactors and inadvertent operation of safety relief valves all challenge the containment and may influence flows because of containment feedback. More recently, the advanced reactor designs put forth by General Electric and Westinghouse in the US and by Framatome and Seimens in Europe rely on the containment to act as the ultimate heat sink. Techniques used by analysts and engineers to analyze the interaction of the containment and the primary system were usually iterative in nature. Codes such as RELAP or RETRAN were used to analyze the primary system response and CONTAIN or CONTEMPT the containment response. The analysis was performed by first running the system code and representing the containment as a fixed pressure boundary condition. The flows were usually from the primary system to the containment initially and generally under choked conditions. Once the mass flows and timing are determined from the system codes, these conditions were input into the containment code. The resulting pressures and temperatures were then calculated and the containment performance analyzed. The disadvantage of this approach becomes evident when one performs an analysis of a rapid depressurization or a long term accident sequence in which feedback from the containment can occur. For example, in a BWR main steam line break transient, the containment heats up and becomes a source of energy for the primary system. Recent advances in programming and computer technology are available to provide an alternative approach. The author and other researchers have developed linkage codes capable of transferring data between codes at each time step allowing discrete codes to be coupled together.
MITOM: a new unfolding code based on a spectra model method applied to neutron spectrometry.
Tomás, M; Fernández, F; Bakali, M; Muller, H
2004-01-01
The MITOM code was developed at UAB (Universitat Autònoma de Barcelona) for unfolding neutron spectrometric measurements with a Bonner spheres system (BSS). One of the main characteristics of this code is that an initial parameterisation of the neutron energy components (thermal, intermediate and fast) is needed. This code uses the Monte Carlo method and the Bayesian theorem to obtain a set of solutions achieving different criteria and conditions between calculated and measured count rates. The final solution is an average of the acceptable solutions. The MITOM code was tested for ISO sources and a good agreement was observed between the reference values and the unfolded ones for global magnitudes. The code was applied recently to characterise both thermal SIGMA and CANEL/T400 sources of the IRSN facilities. The results of these applications were very satisfactory as well.
Coded moderator approach for fast neutron source detection and localization at standoff
NASA Astrophysics Data System (ADS)
Littell, Jennifer; Lukosi, Eric; Hayward, Jason; Milburn, Robert; Rowan, Allen
2015-06-01
Considering the need for directional sensing at standoff for some security applications and scenarios where a neutron source may be shielded by high Z material that nearly eliminates the source gamma flux, this work focuses on investigating the feasibility of using thermal neutron sensitive boron straw detectors for fast neutron source detection and localization. We utilized MCNPX simulations to demonstrate that, through surrounding the boron straw detectors by a HDPE coded moderator, a source-detector orientation-specific response enables potential 1D source localization in a high neutron detection efficiency design. An initial test algorithm has been developed in order to confirm the viability of this detector system's localization capabilities which resulted in identification of a 1 MeV neutron source with a strength equivalent to 8 kg WGPu at 50 m standoff within ±11°.
Image enhancement using MCNP5 code and MATLAB in neutron radiography.
Tharwat, Montaser; Mohamed, Nader; Mongy, T
2014-07-01
This work presents a method that can be used to enhance the neutron radiography (NR) image for objects with high scattering materials like hydrogen, carbon and other light materials. This method used Monte Carlo code, MCNP5, to simulate the NR process and get the flux distribution for each pixel of the image and determines the scattered neutron distribution that caused image blur, and then uses MATLAB to subtract this scattered neutron distribution from the initial image to improve its quality. This work was performed before the commissioning of digital NR system in Jan. 2013. The MATLAB enhancement method is quite a good technique in the case of static based film neutron radiography, while in neutron imaging (NI) technique, image enhancement and quantitative measurement were efficient by using ImageJ software. The enhanced image quality and quantitative measurements were presented in this work.
Rohatgi, U.S.; Cheng, H.S.; Khan, H.J.; Mallen, A.N.; Neymotin, L.Y.
1998-03-01
This document describes the major modifications and improvements made to the modeling of the RAMONA-3B/MOD0 code since 1981, when the code description and assessment report was completed. The new version of the code is RAMONA-4B. RAMONA-4B is a systems transient code for application to different versions of Boiling Water Reactors (BWR) such as the current BWR, the Advanced Boiling Water Reactor (ABWR), and the Simplified Boiling Water Reactor (SBWR). This code uses a three-dimensional neutron kinetics model coupled with a multichannel, non-equilibrium, drift-flux, two-phase flow formulation of the thermal hydraulics of the reactor vessel. The code is designed to analyze a wide spectrum of BWR core and system transients and instability issues. Chapter 1 is an overview of the code`s capabilities and limitations; Chapter 2 discusses the neutron kinetics modeling and the implementation of reactivity edits. Chapter 3 is an overview of the heat conduction calculations. Chapter 4 presents modifications to the thermal-hydraulics model of the vessel, recirculation loop, steam separators, boron transport, and SBWR specific components. Chapter 5 describes modeling of the plant control and safety systems. Chapter 6 presents and modeling of Balance of Plant (BOP). Chapter 7 describes the mechanistic containment model in the code. The content of this report is complementary to the RAMONA-3B code description and assessment document. 53 refs., 81 figs., 13 tabs.
Application of Monte Carlo codes to neutron dosimetry
Prevo, C.T.
1982-06-15
In neutron dosimetry, calculations enable one to predict the response of a proposed dosimeter before effort is expended to design and fabricate the neutron instrument or dosimeter. The nature of these calculations requires the use of computer programs that implement mathematical models representing the transport of radiation through attenuating media. Numerical, and in some cases analytical, solutions of these models can be obtained by one of several calculational techniques. All of these techniques are either approximate solutions to the well-known Boltzmann equation or are based on kernels obtained from solutions to the equation. The Boltzmann equation is a precise mathematical description of neutron behavior in terms of position, energy, direction, and time. The solution of the transport equation represents the average value of the particle flux density. Integral forms of the transport equation are generally regarded as the formal basis for the Monte Carlo method, the results of which can in principle be made to approach the exact solution. This paper focuses on the Monte Carlo technique.
NASA Astrophysics Data System (ADS)
Shi, Xue-Ming; Peng, Xian-Jue
2016-09-01
Fusion science and technology has made progress in the last decades. However, commercialization of fusion reactors still faces challenges relating to higher fusion energy gain, irradiation-resistant material, and tritium self-sufficiency. Fusion Fission Hybrid Reactors (FFHR) can be introduced to accelerate the early application of fusion energy. Traditionally, FFHRs have been classified as either breeders or transmuters. Both need partition of plutonium from spent fuel, which will pose nuclear proliferation risks. A conceptual design of a Fusion Fission Hybrid Reactor for Energy (FFHR-E), which can make full use of natural uranium with lower nuclear proliferation risk, is presented. The fusion core parameters are similar to those of the International Thermonuclear Experimental Reactor. An alloy of natural uranium and zirconium is adopted in the fission blanket, which is cooled by light water. In order to model blanket burnup problems, a linkage code MCORGS, which couples MCNP4B and ORIGEN-S, is developed and validated through several typical benchmarks. The average blanket energy Multiplication and Tritium Breeding Ratio can be maintained at 10 and 1.15 respectively over tens of years of continuous irradiation. If simple reprocessing without separation of plutonium from uranium is adopted every few years, FFHR-E can achieve better neutronic performance. MCORGS has also been used to analyze the ultra-deep burnup model of Laser Inertial Confinement Fusion Fission Energy (LIFE) from LLNL, and a new blanket design that uses Pb instead of Be as the neutron multiplier is proposed. In addition, MCORGS has been used to simulate the fluid transmuter model of the In-Zinerater from Sandia. A brief comparison of LIFE, In-Zinerater, and FFHR-E will be given.
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.
A 2D/1D coupling neutron transport method based on the matrix MOC and NEM methods
Zhang, H.; Zheng, Y.; Wu, H.; Cao, L.
2013-07-01
A new 2D/1D coupling method based on the matrix MOC method (MMOC) and nodal expansion method (NEM) is proposed for solving the three-dimensional heterogeneous neutron transport problem. The MMOC method, used for radial two-dimensional calculation, constructs a response matrix between source and flux with only one sweep and then solves the linear system by using the restarted GMRES algorithm instead of the traditional trajectory sweeping process during within-group iteration for angular flux update. Long characteristics are generated by using the customization of commercial software AutoCAD. A one-dimensional diffusion calculation is carried out in the axial direction by employing the NEM method. The 2D and ID solutions are coupled through the transverse leakage items. The 3D CMFD method is used to ensure the global neutron balance and adjust the different convergence properties of the radial and axial solvers. A computational code is developed based on these theories. Two benchmarks are calculated to verify the coupling method and the code. It is observed that the corresponding numerical results agree well with references, which indicates that the new method is capable of solving the 3D heterogeneous neutron transport problem directly. (authors)
1980-10-15
Version 00 PALLAS-2DCY-FX is a code for direct integration of the transport equation in two-dimensional (r,z) geometry. It solves the energy and angular-dependent Boltzmann transport equation with general anisotropic scattering in cylindrical geometry. Its principal applications are to neutron or gamma-ray transport problems in the forward mode. The code is particularly designed for and suited to the solution of deep penetration radiation transport problems with an external (fixed) source.
Systems guide to MCNP (Monte Carlo Neutron and Photon Transport Code)
Kirk, B.L.; West, J.T.
1984-06-01
The subject of this report is the implementation of the Los Alamos National Laboratory Monte Carlo Neutron and Photon Transport Code - Version 3 (MCNP) on the different types of computer systems, especially the IBM MVS system. The report supplements the documentation of the RSIC computer code package CCC-200/MCNP. Details of the procedure to follow in executing MCNP on the IBM computers, either in batch mode or interactive mode, are provided.
Neutron fluence calculations for the SDC detector and the results of codes comparison
Job, P.K.; Handler, T.; Gabriel, T.A.; Slater, C.O.; Waters, L.S.; Palounek, A.P.T.; Zeitnitz, C.
1994-04-01
CALOR89, A Monte Carlo particle physics code package in conjunction with ISAJET, a high energy particle collision code, has been successfully used to evaluate the radiation environment of a high energy physics collider detector. We found that for a collider luminosity of 10{sup 33} cm{sup {minus}2sec{minus}1}, the neutron fluences can be significant at certain detector locations.
Neutron Activation Analysis and Product Isotope Inventory Code System.
1990-10-31
Version 00 NAC was designed to predict the neutron-induced gamma-ray radioactivity for a wide variety of composite materials. The NAC output includes the input data, a list of all reactions for each constituent element, and the end-of-irradiation disintegration rates for each reaction. NAC also compiles a product isotope inventory containing the isotope name, the disintegration rate, the gamma-ray source strength, and the absorbed dose rate at 1 meter from an unshielded point source. The inducedmore » activity is calculated as a function of irradiation and decay times; the effect of cyclic irradiation can also be calculated.« less
2008-02-29
Version 00 (1) Problems to be solved: MVP/GMVP II can solve eigenvalue and fixed-source problems. The multigroup code GMVP can solve forward and adjoint problems for neutron, photon and neutron-photon coupled transport. The continuous-energy code MVP can solve only the forward problems. Both codes can also perform time-dependent calculations. (2) Geometry description: MVP/GMVP employs combinatorial geometry to describe the calculation geometry. It describes spatial regions by the combination of the 3-dimensional objects (BODIes). Currently, themore » following objects (BODIes) can be used. - BODIes with linear surfaces : half space, parallelepiped, right parallelepiped, wedge, right hexagonal prism - BODIes with quadratic surface and linear surfaces : cylinder, sphere, truncated right cone, truncated elliptic cone, ellipsoid by rotation, general ellipsoid - Arbitrary quadratic surface and torus The rectangular and hexagonal lattice geometry can be used to describe the repeated geometry. Furthermore, the statistical geometry model is available to treat coated fuel particles or pebbles for high temperature reactors. (3) Particle sources: The various forms of energy-, angle-, space- and time-dependent distribution functions can be specified. See Abstract for more detail.« less
Wilson, W B; Perry, R T; Charlton, W S; Parish, T A; Shores, E F
2005-01-01
SOURCES is a computer code that determines neutron production rates and spectra from (alpha,n) reactions, spontaneous fission and delayed neutron emission owing to the decay of radionuclides in homogeneous media, interface problems and three-region interface problems. The code is also capable of calculating the neutron production rates due to (alpha,n) reactions induced by a monoenergetic beam of alpha particles incident on a slab of target material. The (alpha,n) spectra are calculated using an assumed isotropic angular distribution in the centre-of-mass system with a library of 107 nuclide decay alpha-particle spectra, 24 sets of measured and/or evaluated (alpha,n) cross sections and product nuclide level branching fractions, and functional alpha particle stopping cross sections for Z < 106. Spontaneous fission sources and spectra are calculated with evaluated half-life, spontaneous fission branching and Watt spectrum parameters for 44 actinides. The delayed neutron spectra are taken from an evaluated library of 105 precursors. The code outputs the magnitude and spectra of the resultant neutron sources. It also provides an analysis of the contributions to that source by each nuclide in the problem. PMID:16381695
ACDOS2: an improved neutron-induced dose rate code
Lagache, J.C.
1981-06-01
To calculate the expected dose rate from fusion reactors as a function of geometry, composition, and time after shutdown a computer code, ACDOS2, was written, which utilizes up-to-date libraries of cross-sections and radioisotope decay data. ACDOS2 is in ANSI FORTRAN IV, in order to make it readily adaptable elsewhere.
Analyses of the MSLB benchmark V1000CT-2 by the coupled system code ATHLET-BIPR8KN
Nikonov, S. P.; Langenbuch, S.; Lizorkin, M. P.; Velkov, K.
2006-07-01
Within the activities of OECD/NEA is being initiated the second phase of the VVER-1000 Coolant Transient Benchmark (V1000CT-2). It considers the best estimate analyses of a Main Steam Line Break (MSLB) of a VVER-1000 NPP with two exercises. The analyses have been performed with the coupled system code ATHLET-BIPR8KN which enables to perform realistic simulation of three-dimensional neutron kinetics and thermal-hydraulic processes in VVER NPP. Results are presented and analysed for the two proposed scenarios. These results are supplemented by sensitivity studies varying the number of the thermo-hydraulic channels (THC) in the core and by comparisons with point kinetics calculations. This work is of considerable importance for the validation of the coupled system code ATHLET-BIPR8KN in case of asymmetric core inlet conditions. (authors)
ABAREX -- A neutron spherical optical-statistical-model code -- A user`s manual
Smith, A.B.; Lawson, R.D.
1998-06-01
The contemporary version of the neutron spherical optical-statistical-model code ABAREX is summarized with the objective of providing detailed operational guidance for the user. The physical concepts involved are very briefly outlined. The code is described in some detail and a number of explicit examples are given. With this document one should very quickly become fluent with the use of ABAREX. While the code has operated on a number of computing systems, this version is specifically tailored for the VAX/VMS work station and/or the IBM-compatible personal computer.
Verification of a neutronic code for transient analysis in reactors with Hex-z geometry
Gonzalez-Pintor, S.; Verdu, G.; Ginestar, D.
2012-07-01
Due to the geometry of the fuel bundles, to simulate reactors such as VVER reactors it is necessary to develop methods that can deal with hexagonal prisms as basic elements of the spatial discretization. The main features of a code based on a high order finite element method for the spatial discretization of the neutron diffusion equation and an implicit difference method for the time discretization of this equation are presented and the performance of the code is tested solving the first exercise of the AER transient benchmark. The obtained results are compared with the reference results of the benchmark and with the results provided by PARCS code. (authors)
A theory manual for multi-physics code coupling in LIME.
Belcourt, Noel; Bartlett, Roscoe Ainsworth; Pawlowski, Roger Patrick; Schmidt, Rodney Cannon; Hooper, Russell Warren
2011-03-01
The Lightweight Integrating Multi-physics Environment (LIME) is a software package for creating multi-physics simulation codes. Its primary application space is when computer codes are currently available to solve different parts of a multi-physics problem and now need to be coupled with other such codes. In this report we define a common domain language for discussing multi-physics coupling and describe the basic theory associated with multiphysics coupling algorithms that are to be supported in LIME. We provide an assessment of coupling techniques for both steady-state and time dependent coupled systems. Example couplings are also demonstrated.
Neutron Star Structure in the Presence of Conformally Coupled Scalar Fields
NASA Technical Reports Server (NTRS)
Sultana, Joseph; Bose, Benjamin; Kazanas, Demosthenes
2014-01-01
Neutron star models are studied in the context of scalar-tensor theories of gravity in the presence of a conformally coupled scalar field, using two different numerical equations of state (EoS) representing different degrees of stiffness. In both cases we obtain a complete solution by matching the interior numerical solution of the coupled Einstein-scalar field hydrostatic equations, with an exact metric on the surface of the star. These are then used to find the effect of the scalar field and its coupling to geometry, on the neutron star structure, particularly the maximum neutron star mass and radius. We show that in the presence of a conformally coupled scalar field, neutron stars are less dense and have smaller masses and radii than their counterparts in the minimally coupled case, and the effect increases with the magnitude of the scalar field at the center of the star.
Neutronics-thermalhydraulics coupling in a CANDU SCWR
NASA Astrophysics Data System (ADS)
Adouki, Pierre
In order to implement new nuclear technologies as a solution to the growing demand for energy, 10 countries agreed on a framework for international cooperation in 2002, to form the Generation IV International Forum (GIF). The goal of the GIF is to design the next generation of nuclear reactors that would be cost effective and would enhance safety. This forum has proposed several types of Generation IV reactors including the Supercritical Water-Cooled Reactor (SCWR). The SCWR comes in two main configurations: pressure vessel SCWR and pressure tube SCWR (PT-SCWR). In this study, the CANDU SCWR (a PT-SCWR) is considered. This reactor is oriented vertically and contains 336 channels with a length of 5 m. The target coolant inlet and outlet temperatures are 350 Celsius and 625 Celsius, respectively. The coolant flows downwards, and the reactor power is 2540 MWth. Various fuel designs have been considered in order not to exceed the linear element rating. However, the dependency between the core power and thermalhydraulics parameters results in the necessity to use a neutronics/thermalhydaulics coupling scheme to determine the core power and the thermalhydraulics parameters. The core power obtained has a power peaking factor of 1.4. The bundle power distribution for all channels has a peak at the third bundle from the inlet, but the value of this peak increases with the channel power. The heat-transfer coefficient and the specific-heat capacity have a peak at the same location in a channel, and this location shifts toward the inlet as the channel power increases. The exit coolant temperature increases with the channel power, while the exit coolant density and pressure decrease with the channel power. Also, higher channel powers lead to higher fuel and cladding temperatures. Moreover, as the coupling method is applied, the effective multiplication factor and the values of thermalhydaulics parameters oscillate as they converge.
GPU-accelerated 3D neutron diffusion code based on finite difference method
Xu, Q.; Yu, G.; Wang, K.
2012-07-01
Finite difference method, as a traditional numerical solution to neutron diffusion equation, although considered simpler and more precise than the coarse mesh nodal methods, has a bottle neck to be widely applied caused by the huge memory and unendurable computation time it requires. In recent years, the concept of General-Purpose computation on GPUs has provided us with a powerful computational engine for scientific research. In this study, a GPU-Accelerated multi-group 3D neutron diffusion code based on finite difference method was developed. First, a clean-sheet neutron diffusion code (3DFD-CPU) was written in C++ on the CPU architecture, and later ported to GPUs under NVIDIA's CUDA platform (3DFD-GPU). The IAEA 3D PWR benchmark problem was calculated in the numerical test, where three different codes, including the original CPU-based sequential code, the HYPRE (High Performance Pre-conditioners)-based diffusion code and CITATION, were used as counterpoints to test the efficiency and accuracy of the GPU-based program. The results demonstrate both high efficiency and adequate accuracy of the GPU implementation for neutron diffusion equation. A speedup factor of about 46 times was obtained, using NVIDIA's Geforce GTX470 GPU card against a 2.50 GHz Intel Quad Q9300 CPU processor. Compared with the HYPRE-based code performing in parallel on an 8-core tower server, the speedup of about 2 still could be observed. More encouragingly, without any mathematical acceleration technology, the GPU implementation ran about 5 times faster than CITATION which was speeded up by using the SOR method and Chebyshev extrapolation technique. (authors)
Ebert, D.
1997-07-01
This is a report on the CSNI Workshop on Transient Thermal-Hydraulic and Neutronic Codes Requirements held at Annapolis, Maryland, USA November 5-8, 1996. This experts` meeting consisted of 140 participants from 21 countries; 65 invited papers were presented. The meeting was divided into five areas: (1) current and prospective plans of thermal hydraulic codes development; (2) current and anticipated uses of thermal-hydraulic codes; (3) advances in modeling of thermal-hydraulic phenomena and associated additional experimental needs; (4) numerical methods in multi-phase flows; and (5) programming language, code architectures and user interfaces. The workshop consensus identified the following important action items to be addressed by the international community in order to maintain and improve the calculational capability: (a) preserve current code expertise and institutional memory, (b) preserve the ability to use the existing investment in plant transient analysis codes, (c) maintain essential experimental capabilities, (d) develop advanced measurement capabilities to support future code validation work, (e) integrate existing analytical capabilities so as to improve performance and reduce operating costs, (f) exploit the proven advances in code architecture, numerics, graphical user interfaces, and modularization in order to improve code performance and scrutibility, and (g) more effectively utilize user experience in modifying and improving the codes.
Non-uniform Neutron Source Approximation for Iterative Reconstruction of Coded Source Images
Gregor, Jens; Bingham, Philip R
2016-01-01
X-ray and neutron optics both lack ray focusing capabilities. An x-ray source can be made small and powerful enough to facilitate high-resolution imaging while providing adequate flux. This is not yet possible for neutrons. One remedy is to employ a computational imaging technique such as magnified coded source imaging. The greatest challenge associated with successful reconstruction of high-resolution images from such radiographs is to precisely model the flux distribution for complex non-uniform neutron sources. We have developed a framework based on Monte Carlo simulation and iterative reconstruction that facilitates high- resolution coded source neutron imaging. In this paper, we define a methodology to empirically measure and approximate the flux profile of a non-uniform neutron source, and we show how to incorporate the result within the forward model of an iterative reconstruction algorithm. We assess improvement in image quality by comparing reconstructions based respectively on the new empirical forward model and our previous analytic models.
Neutronics code VALE for two-dimensional triagonal (hexagonal) and three-dimensional geometries
Vondy, D.R.; Fowler, T.B.
1981-08-01
This report documents the computer code VALE designed to solve multigroup neutronics problems with the diffusion theory approximation to neutron transport for a triagonal arrangement of mesh points on planes in two- and three-dimensional geometry. This code parallels the VENTURE neutronics code in the local computation system, making exposure and fuel management capabilities available. It uses and generates interface data files adopted in the cooperative effort sponsored by Reactor Physics RRT Division of the US DOE. The programming in FORTRAN is straightforward, although data is transferred in blocks between auxiliary storage devices and main core, and direct access schemes are used. The size of problems which can be handled is essentially limited only by cost of calculation since the arrays are variably dimensioned. The memory requirement is held down while data transfer during iteration is increased only as necessary with problem size. There is provision for the more common boundary conditions including the repeating boundary, 180/sup 0/ rotational symmetry, and the rotational symmetry conditions for the 30/sup 0/, 60/sup 0/, and 120/sup 0/ triangular grids on planes. A variety of types of problems may be solved: the usual neutron flux eignevalue problem, or a direct criticality search on the buckling, on a reciprocal velocity absorber (prompt mode), or on nuclide concentrations. The adjoint problem and fixed source problem may be solved, as well as the dominating higher harmonic, or the importance problem for an arbitrary fixed source.
a New Method for Neutron Capture Therapy (nct) and Related Simulation by MCNP4C Code
NASA Astrophysics Data System (ADS)
Shirazi, Mousavi; Alireza, Seyed; Ali, Taheri
2010-01-01
Neutron capture therapy (NCT) is enumerated as one of the most important methods for treatment of some strong maladies among cancers in medical science thus is unavoidable controlling and protecting instances in use of this science. Among of treatment instances of this maladies with use of nuclear medical science is use of neutron therapy that is one of the most important and effective methods in treatment of cancers. But whereas fast neutrons have too destroyer effects and also sake of protection against additional absorbed energy (absorbed dose) by tissue during neutron therapy and also naught damaging to rest of healthy tissues, should be measured absorbed energy by tissue accurately, because destroyer effects of fast neutrons is almost quintuple more than gamma photons. In this article for neutron therapy act of male's liver has been simulated a system by the Monte Carlo method (MCNP4C code) and also with use of analytical method, thus absorbed dose by this tissue has been obtained for sources with different energies accurately and has been compared results of this two methods together.
Coupled Receiver/Decoders for Low-Rate Turbo Codes
NASA Technical Reports Server (NTRS)
Hamkins, Jon; Divsalar, Dariush
2005-01-01
been proposed for receiving weak single- channel phase-modulated radio signals bearing low-rate-turbo-coded binary data. Originally intended for use in receiving telemetry signals from distant spacecraft, the proposed receiver/ decoders may also provide enhanced reception in mobile radiotelephone systems. A radio signal of the type to which the proposal applies comprises a residual carrier signal and a phase-modulated data signal. The residual carrier signal is needed as a phase reference for demodulation as a prerequisite to decoding. Low-rate turbo codes afford high coding gains and thereby enable the extraction of data from arriving radio signals that might otherwise be too weak. In the case of a conventional receiver, if the signal-to-noise ratio (specifically, the symbol energy to one-sided noise power spectral density) of the arriving signal is below approximately 0 dB, then there may not be enough energy per symbol to enable the receiver to recover properly the carrier phase. One could solve the problem at the transmitter by diverting some power from the data signal to the residual carrier. A better solution . a coupled receiver/decoder according to the proposal . could reduce the needed amount of residual carrier power. In all that follows, it is to be understood that all processing would be digital and the incoming signals to be processed would be, more precisely, outputs of analog-to-digital converters that preprocess the residual carrier and data signals at a rate of multiple samples per symbol. The upper part of the figure depicts a conventional receiving system, in which the receiver and decoder are uncoupled, and which is also called a non-data-aided system because output data from the decoder are not used in the receiver to aid in recovering the carrier phase. The receiver tracks the carrier phase from the residual carrier signal and uses the carrier phase to wipe phase noise off the data signal. The receiver typically includes a phase-locked loop
Filippone, W.L.; Baker, R.S.
1990-12-31
The neutron transport equation is solved by a hybrid method that iteratively couples regions where deterministic (S{sub N}) and stochastic (Monte Carlo) methods are applied. Unlike previous hybrid methods, the Monte Carlo and S{sub N} regions are fully coupled in the sense that no assumption is made about geometrical separation or decoupling. The hybrid method provides a new means of solving problems involving both optically thick and optically thin regions that neither Monte Carlo nor S{sub N} is well suited for by themselves. The fully coupled Monte Carlo/S{sub N} technique consists of defining spatial and/or energy regions of a problem in which either a Monte Carlo calculation or an S{sub N} calculation is to be performed. The Monte Carlo region may comprise the entire spatial region for selected energy groups, or may consist of a rectangular area that is either completely or partially embedded in an arbitrary S{sub N} region. The Monte Carlo and S{sub N} regions are then connected through the common angular boundary fluxes, which are determined iteratively using the response matrix technique, and volumetric sources. The hybrid method has been implemented in the S{sub N} code TWODANT by adding special-purpose Monte Carlo subroutines to calculate the response matrices and volumetric sources, and linkage subrountines to carry out the interface flux iterations. The common angular boundary fluxes are included in the S{sub N} code as interior boundary sources, leaving the logic for the solution of the transport flux unchanged, while, with minor modifications, the diffusion synthetic accelerator remains effective in accelerating S{sub N} calculations. The special-purpose Monte Carlo routines used are essentially analog, with few variance reduction techniques employed. However, the routines have been successfully vectorized, with approximately a factor of five increase in speed over the non-vectorized version.
Ganapol, B.D.; Kornreich, D.E.
1997-07-01
Because of the requirement of accountability and quality control in the scientific world, a demand for high-quality analytical benchmark calculations has arisen in the neutron transport community. The intent of these benchmarks is to provide a numerical standard to which production neutron transport codes may be compared in order to verify proper operation. The overall investigation as modified in the second year renewal application includes the following three primary tasks. Task 1 on two dimensional neutron transport is divided into (a) single medium searchlight problem (SLP) and (b) two-adjacent half-space SLP. Task 2 on three-dimensional neutron transport covers (a) point source in arbitrary geometry, (b) single medium SLP, and (c) two-adjacent half-space SLP. Task 3 on code verification, includes deterministic and probabilistic codes. The primary aim of the proposed investigation was to provide a suite of comprehensive two- and three-dimensional analytical benchmarks for neutron transport theory applications. This objective has been achieved. The suite of benchmarks in infinite media and the three-dimensional SLP are a relatively comprehensive set of one-group benchmarks for isotropically scattering media. Because of time and resource limitations, the extensions of the benchmarks to include multi-group and anisotropic scattering are not included here. Presently, however, enormous advances in the solution for the planar Green`s function in an anisotropically scattering medium have been made and will eventually be implemented in the two- and three-dimensional solutions considered under this grant. Of particular note in this work are the numerical results for the three-dimensional SLP, which have never before been presented. The results presented were made possible only because of the tremendous advances in computing power that have occurred during the past decade.
DANDE: a linked code system for core neutronics/depletion analysis
LaBauve, R.J.; England, T.R.; George, D.C.; MacFarlane, R.E.; Wilson, W.B.
1985-06-01
This report describes DANDE - a modular neutronics, depletion code system for reactor analysis. It consists of nuclear data processing, core physics, and fuel depletion modules, and allows one to use diffusion and transport methods interchangeably in core neutronics calculations. This latter capability is especially important in the design of small modular cores. Additional unique features include the capability of updating the nuclear data file during a calculation; a detailed treatment of depletion, burnable poisons as well as fuel; and the ability to make geometric changes such as control rod repositioning and fuel relocation in the course of a calculation. The detailed treatment of reactor fuel burnup, fission-product creation and decay, as well as inventories of higher-order actinides is a necessity when predicting the behavior of reactor fuel under increased burn conditions. The operation of the code system is made clear in this report by following a sample problem.
VENTURE/PC manual: A multidimensional multigroup neutron diffusion code system. Version 3
Shapiro, A.; Huria, H.C.; Cho, K.W.
1991-12-01
VENTURE/PC is a recompilation of part of the Oak Ridge BOLD VENTURE code system, which will operate on an IBM PC or compatible computer. Neutron diffusion theory solutions are obtained for multidimensional, multigroup problems. This manual contains information associated with operating the code system. The purpose of the various modules used in the code system, and the input for these modules are discussed. The PC code structure is also given. Version 2 included several enhancements not given in the original version of the code. In particular, flux iterations can be done in core rather than by reading and writing to disk, for problems which allow sufficient memory for such in-core iterations. This speeds up the iteration process. Version 3 does not include any of the special processors used in the previous versions. These special processors utilized formatted input for various elements of the code system. All such input data is now entered through the Input Processor, which produces standard interface files for the various modules in the code system. In addition, a Standard Interface File Handbook is included in the documentation which is distributed with the code, to assist in developing the input for the Input Processor.
VENTURE/PC manual: A multidimensional multigroup neutron diffusion code system
Shapiro, A.; Huria, H.C.; Cho, K.W. )
1991-12-01
VENTURE/PC is a recompilation of part of the Oak Ridge BOLD VENTURE code system, which will operate on an IBM PC or compatible computer. Neutron diffusion theory solutions are obtained for multidimensional, multigroup problems. This manual contains information associated with operating the code system. The purpose of the various modules used in the code system, and the input for these modules are discussed. The PC code structure is also given. Version 2 included several enhancements not given in the original version of the code. In particular, flux iterations can be done in core rather than by reading and writing to disk, for problems which allow sufficient memory for such in-core iterations. This speeds up the iteration process. Version 3 does not include any of the special processors used in the previous versions. These special processors utilized formatted input for various elements of the code system. All such input data is now entered through the Input Processor, which produces standard interface files for the various modules in the code system. In addition, a Standard Interface File Handbook is included in the documentation which is distributed with the code, to assist in developing the input for the Input Processor.
Application of three-dimensional transport code to the analysis of the neutron streaming experiment
Chatani, K.; Slater, C.O.
1990-01-01
This paper summarized the calculational results of neutron streaming through a Clinch River Breeder Reactor (CRBR) Prototype coolant pipe chaseway. Particular emphasis is placed on results at bends in the chaseway. Calculations were performed with three three-dimensional codes: the discrete ordinates radiation transport code TORT and Monte Carlo radiation transport code MORSE, which were developed by Oak Ridge National Laboratory (ORNL), and the discrete ordinates code ENSEMBLE, which was developed in Japan. The purpose of the calculations is not only to compare the calculational results with the experimental results, but also to compare the results of TORT and MORSE with those of ENSEMBLE. In the TORT calculations, two types of difference methods, weighted-difference method was applied in ENSEMBLE calculation. Both TORT and ENSEMBLE produced nearly the same calculational results, but differed in the number of iterations required for converging each neutron group. Also, the two types of difference methods in the TORT calculations showed no appreciable variance in the number of iterations required. However, a noticeable disparity in the computer times and some variation in the calculational results did occur. The comparisons of the calculational results with the experimental results, showed for the epithermal neutron flux generally good agreement in the first and second legs and at the first bend where the two-dimensional modeling might be difficult. Results were fair to poor along the centerline of the first leg near the opening to the second leg because of discrete ordinates ray effects. Additionally, the agreement was good throughout the first and second legs for the thermal neutron region. Calculations with MORSE were made. These calculational results and comparisons are described also. 8 refs., 4 figs.
Simulations of inspiraling and merging double neutron stars using the Spectral Einstein Code
NASA Astrophysics Data System (ADS)
Haas, Roland; Ott, Christian D.; Szilagyi, Bela; Kaplan, Jeffrey D.; Lippuner, Jonas; Scheel, Mark A.; Barkett, Kevin; Muhlberger, Curran D.; Dietrich, Tim; Duez, Matthew D.; Foucart, Francois; Pfeiffer, Harald P.; Kidder, Lawrence E.; Teukolsky, Saul A.
2016-06-01
We present results on the inspiral, merger, and postmerger evolution of a neutron star-neutron star (NSNS) system. Our results are obtained using the hybrid pseudospectral-finite volume Spectral Einstein Code (SpEC). To test our numerical methods, we evolve an equal-mass system for ≈22 orbits before merger. This waveform is the longest waveform obtained from fully general-relativistic simulations for NSNSs to date. Such long (and accurate) numerical waveforms are required to further improve semianalytical models used in gravitational wave data analysis, for example, the effective one body models. We discuss in detail the improvements to SpEC's ability to simulate NSNS mergers, in particular mesh refined grids to better resolve the merger and postmerger phases. We provide a set of consistency checks and compare our results to NSNS merger simulations with the independent bam code. We find agreement between them, which increases confidence in results obtained with either code. This work paves the way for future studies using long waveforms and more complex microphysical descriptions of neutron star matter in SpEC.
NASA Astrophysics Data System (ADS)
Ortiz-Rodríguez, J. M.; Reyes Alfaro, A.; Reyes Haro, A.; Solís Sánches, L. O.; Miranda, R. Castañeda; Cervantes Viramontes, J. M.; Vega-Carrillo, H. R.
2013-07-01
In this work the performance of two neutron spectrum unfolding codes based on iterative procedures and artificial neural networks is evaluated. The first one code based on traditional iterative procedures and called Neutron spectrometry and dosimetry from the Universidad Autonoma de Zacatecas (NSDUAZ) use the SPUNIT iterative algorithm and was designed to unfold neutron spectrum and calculate 15 dosimetric quantities and 7 IAEA survey meters. The main feature of this code is the automated selection of the initial guess spectrum trough a compendium of neutron spectrum compiled by the IAEA. The second one code known as Neutron spectrometry and dosimetry with artificial neural networks (NDSann) is a code designed using neural nets technology. The artificial intelligence approach of neural net does not solve mathematical equations. By using the knowledge stored at synaptic weights on a neural net properly trained, the code is capable to unfold neutron spectrum and to simultaneously calculate 15 dosimetric quantities, needing as entrance data, only the rate counts measured with a Bonner spheres system. Similarities of both NSDUAZ and NSDann codes are: they follow the same easy and intuitive user's philosophy and were designed in a graphical interface under the LabVIEW programming environment. Both codes unfold the neutron spectrum expressed in 60 energy bins, calculate 15 dosimetric quantities and generate a full report in HTML format. Differences of these codes are: NSDUAZ code was designed using classical iterative approaches and needs an initial guess spectrum in order to initiate the iterative procedure. In NSDUAZ, a programming routine was designed to calculate 7 IAEA instrument survey meters using the fluence-dose conversion coefficients. NSDann code use artificial neural networks for solving the ill-conditioned equation system of neutron spectrometry problem through synaptic weights of a properly trained neural network. Contrary to iterative procedures, in neural
Ortiz-Rodriguez, J. M.; Reyes Alfaro, A.; Reyes Haro, A.; Solis Sanches, L. O.; Miranda, R. Castaneda; Cervantes Viramontes, J. M.; Vega-Carrillo, H. R.
2013-07-03
In this work the performance of two neutron spectrum unfolding codes based on iterative procedures and artificial neural networks is evaluated. The first one code based on traditional iterative procedures and called Neutron spectrometry and dosimetry from the Universidad Autonoma de Zacatecas (NSDUAZ) use the SPUNIT iterative algorithm and was designed to unfold neutron spectrum and calculate 15 dosimetric quantities and 7 IAEA survey meters. The main feature of this code is the automated selection of the initial guess spectrum trough a compendium of neutron spectrum compiled by the IAEA. The second one code known as Neutron spectrometry and dosimetry with artificial neural networks (NDSann) is a code designed using neural nets technology. The artificial intelligence approach of neural net does not solve mathematical equations. By using the knowledge stored at synaptic weights on a neural net properly trained, the code is capable to unfold neutron spectrum and to simultaneously calculate 15 dosimetric quantities, needing as entrance data, only the rate counts measured with a Bonner spheres system. Similarities of both NSDUAZ and NSDann codes are: they follow the same easy and intuitive user's philosophy and were designed in a graphical interface under the LabVIEW programming environment. Both codes unfold the neutron spectrum expressed in 60 energy bins, calculate 15 dosimetric quantities and generate a full report in HTML format. Differences of these codes are: NSDUAZ code was designed using classical iterative approaches and needs an initial guess spectrum in order to initiate the iterative procedure. In NSDUAZ, a programming routine was designed to calculate 7 IAEA instrument survey meters using the fluence-dose conversion coefficients. NSDann code use artificial neural networks for solving the ill-conditioned equation system of neutron spectrometry problem through synaptic weights of a properly trained neural network. Contrary to iterative procedures, in neural
Crust-core coupling in rotating neutron stars
Glampedakis, Kostas; Andersson, Nils
2006-08-15
Motivated by their gravitational wave driven instability, we investigate the influence of the crust on r-mode oscillations in a neutron star. Using a simplistic model of an elastic neutron star crust with constant shear modulus, we carry out an analytic calculation with the main objective of deriving an expression for the slippage between the core and the crust. Our analytic estimates support previous numerical results and provide useful insights into the details of the problem.
1985-02-01
Version 00 TP1 is a transport theory code, developed to determine reactivity effects and kinetic parameters such as effective delayed neutron fractions and mean generation time by applying the usual perturbation formalism for one-dimensional geometry.
PYESSENCE: Generalized Coupled Quintessence Linear Perturbation Python Code
NASA Astrophysics Data System (ADS)
Leithes, Alexander
2016-09-01
PYESSENCE evolves linearly perturbed coupled quintessence models with multiple (cold dark matter) CDM fluid species and multiple DE (dark energy) scalar fields, and can be used to generate quantities such as the growth factor of large scale structure for any coupled quintessence model with an arbitrary number of fields and fluids and arbitrary couplings.
Rohatgi, U.S.; Cheng, H.S.; Khan, H.J.; Mallen, A.N.; Neymotin, L.Y.
1998-03-01
This document is the User`s Manual for the Boiling Water Reactor (BWR), and Simplified Boiling Water Reactor (SBWR) systems transient code RAMONA-4B. The code uses a three-dimensional neutron-kinetics model coupled with a multichannel, nonequilibrium, drift-flux, phase-flow model of the thermal hydraulics of the reactor vessel. The code is designed to analyze a wide spectrum of BWR core and system transients. Chapter 1 gives an overview of the code`s capabilities and limitations; Chapter 2 describes the code`s structure, lists major subroutines, and discusses the computer requirements. Chapter 3 is on code, auxillary codes, and instructions for running RAMONA-4B on Sun SPARC and IBM Workstations. Chapter 4 contains component descriptions and detailed card-by-card input instructions. Chapter 5 provides samples of the tabulated output for the steady-state and transient calculations and discusses the plotting procedures for the steady-state and transient calculations. Three appendices contain important user and programmer information: lists of plot variables (Appendix A) listings of input deck for sample problem (Appendix B), and a description of the plotting program PAD (Appendix C). 24 refs., 18 figs., 11 tabs.
Neutron beam characterization measurements at the Manuel Lujan Jr. neutron scattering center
Mocko, Michal; Muhrer, Guenter; Daemen, Luke L; Kelsey, Charles T; Duran, Michael A; Tovesson, Fredrik K
2010-01-01
We have measured the neutron beam characteristics of neutron moderators at the Manuel Lujan Jr. Neutron Scattering Center at LANSCE. The absolute thermal neutron flux, energy spectra and time emission spectra were measured for the high resolution and high intensity decoupled water, partially coupled liquid hydrogen and partially coupled water moderators. The results of our experimental study will provide an insight into aging of different target-moderator-reflector-shield components as well as new experimental data for benchmarking of neutron transport codes.
Coupling External Radiation Transport Code Results to the GADRAS Detector Response Function
Mitchell, Dean J.; Thoreson, Gregory G.; Horne, Steven M.
2014-01-01
Simulating gamma spectra is useful for analyzing special nuclear materials. Gamma spectra are influenced not only by the source and the detector, but also by the external, and potentially complex scattering environment. The scattering environment can make accurate representations of gamma spectra difficult to obtain. By coupling the Monte Carlo Nuclear Particle (MCNP) code with the Gamma Detector Response and Analysis Software (GADRAS) detector response function, gamma spectrum simulations can be computed with a high degree of fidelity even in the presence of a complex scattering environment. Traditionally, GADRAS represents the external scattering environment with empirically derived scattering parameters. By modeling the external scattering environment in MCNP and using the results as input for the GADRAS detector response function, gamma spectra can be obtained with a high degree of fidelity. This method was verified with experimental data obtained in an environment with a significant amount of scattering material. The experiment used both gamma-emitting sources and moderated and bare neutron-emitting sources. The sources were modeled using GADRAS and MCNP in the presence of the external scattering environment, producing accurate representations of the experimental data.
Experimental validation of GADRAS's coupled neutron-photon inverse radiation transport solver.
Mattingly, John K.; Mitchell, Dean James; Harding, Lee T.
2010-08-01
Sandia National Laboratories has developed an inverse radiation transport solver that applies nonlinear regression to coupled neutron-photon deterministic transport models. The inverse solver uses nonlinear regression to fit a radiation transport model to gamma spectrometry and neutron multiplicity counting measurements. The subject of this paper is the experimental validation of that solver. This paper describes a series of experiments conducted with a 4.5 kg sphere of {alpha}-phase, weapons-grade plutonium. The source was measured bare and reflected by high-density polyethylene (HDPE) spherical shells with total thicknesses between 1.27 and 15.24 cm. Neutron and photon emissions from the source were measured using three instruments: a gross neutron counter, a portable neutron multiplicity counter, and a high-resolution gamma spectrometer. These measurements were used as input to the inverse radiation transport solver to evaluate the solver's ability to correctly infer the configuration of the source from its measured radiation signatures.
Alloggio, G.; Brega, E.; Basile, D.; Guandalini, R.; Pollachini, L.
1996-08-01
This paper is aimed at presenting a solution method for time-dependent problems coupling thermal-hydraulic behavior and neutronic changes during selected transients in Pressurized Water Reactors. A two-group three-dimensional reactor kinetics model, based on the Analytical Nodal Method with a detailed feedback model, has been implemented in TRAC-PF1 code replacing the original point-kinetics approximation. A geometry conversion was done to match, in the core discretization, the cylindrical geometry of the TRAC-PF1 with the Cartesian geometry of the three-dimensional neutronic model. In this version of TRAC-PFI, named TRAC-EN for IBM computers, a Boron transport model has been implemented. The transient Boron concentration is computed from a Boron mass balance after the coolant mass, energy and momentum balances have been completed. In order to evaluate the new code capabilities, a model of the two-loops 600 MW Westinghouse reactor was implemented. Some specific PWR transients that exhibit interesting nuclear and thermal-hydraulic responses, e.g., control rod ejection and pressurization transients, are presented. To check the Boron transport model, a local Boron dilution transient was analyzed. The results obtained by using the space and time dependent neutronic model can not be predicted by point kinetics approximation. Furthermore, some events that apparently concern only the core, are also involving the primary circuit the responses of which can not be neglected because they affect the neutronic behavior.
Solution of the neutronics code dynamic benchmark by finite element method
NASA Astrophysics Data System (ADS)
Avvakumov, A. V.; Vabishchevich, P. N.; Vasilev, A. O.; Strizhov, V. F.
2016-10-01
The objective is to analyze the dynamic benchmark developed by Atomic Energy Research for the verification of best-estimate neutronics codes. The benchmark scenario includes asymmetrical ejection of a control rod in a water-type hexagonal reactor at hot zero power. A simple Doppler feedback mechanism assuming adiabatic fuel temperature heating is proposed. The finite element method on triangular calculation grids is used to solve the three-dimensional neutron kinetics problem. The software has been developed using the engineering and scientific calculation library FEniCS. The matrix spectral problem is solved using the scalable and flexible toolkit SLEPc. The solution accuracy of the dynamic benchmark is analyzed by condensing calculation grid and varying degree of finite elements.
Gleicher, Frederick N.; Williamson, Richard L.; Ortensi, Javier; Wang, Yaqi; Spencer, Benjamin W.; Novascone, Stephen R.; Hales, Jason D.; Martineau, Richard C.
2014-10-01
The MOOSE neutron transport application RATTLESNAKE was coupled to the fuels performance application BISON to provide a higher fidelity tool for fuel performance simulation. This project is motivated by the desire to couple a high fidelity core analysis program (based on the self-adjoint angular flux equations) to a high fidelity fuel performance program, both of which can simulate on unstructured meshes. RATTLESNAKE solves self-adjoint angular flux transport equation and provides a sub-pin level resolution of the multigroup neutron flux with resonance treatment during burnup or a fast transient. BISON solves the coupled thermomechanical equations for the fuel on a sub-millimeter scale. Both applications are able to solve their respective systems on aligned and unaligned unstructured finite element meshes. The power density and local burnup was transferred from RATTLESNAKE to BISON with the MOOSE Multiapp transfer system. Multiple depletion cases were run with one-way data transfer from RATTLESNAKE to BISON. The eigenvalues are shown to agree well with values obtained from the lattice physics code DRAGON. The one-way data transfer of power density is shown to agree with the power density obtained from an internal Lassman-style model in BISON.
A new paradigm for local-global coupling in whole-core neutron transport.
Lewis, E.; Smith, M.; Palmiotti, G,; Nuclear Engineering Division; Northwestern Univ.; INL
2009-01-01
A new paradigm that increases the efficiency of whole-core neutron transport calculations without lattice homogenization is introduced. Quasi-reflected interface conditions are formulated to partially decouple periodic lattice effects from global flux gradients. The starting point is the finite subelement form of the variational nodal code VARIANT that eliminates fuel-coolant homogenization through the use of heterogeneous nodes. The interface spherical harmonics expansions that couple pin-cell-sized nodes are divided into low-order and high-order terms, and reflected interface conditions are applied to the high-order terms. Combined with an integral transport method within the node, the new approach dramatically reduces both the formation time and the dimensions of the nodal response matrices and leads to sharply reduced memory requirements and computational time. The method is applied to the two-dimensional C5G7 problem, an Organisation for Economic Co-operation and Development/Nuclear Energy Agency pressurized water reactor benchmark containing mixed oxide (MOX) and UO{sub 2} fuel assemblies, as well as to a three-dimensional MOX fuel assembly. Results indicate the new approach results in very little loss of accuracy relative to the corresponding full spherical harmonics expansions while reducing computational times by well over an order of magnitude.
Development of Monte Carlo code for coincidence prompt gamma-ray neutron activation analysis
NASA Astrophysics Data System (ADS)
Han, Xiaogang
Prompt Gamma-Ray Neutron Activation Analysis (PGNAA) offers a non-destructive, relatively rapid on-line method for determination of elemental composition of bulk and other samples. However, PGNAA has an inherently large background. These backgrounds are primarily due to the presence of the neutron excitation source. It also includes neutron activation of the detector and the prompt gamma rays from the structure materials of PGNAA devices. These large backgrounds limit the sensitivity and accuracy of PGNAA. Since most of the prompt gamma rays from the same element are emitted in coincidence, a possible approach for further improvement is to change the traditional PGNAA measurement technique and introduce the gamma-gamma coincidence technique. It is well known that the coincidence techniques can eliminate most of the interference backgrounds and improve the signal-to-noise ratio. A new Monte Carlo code, CEARCPG has been developed at CEAR to simulate gamma-gamma coincidence spectra in PGNAA experiment. Compared to the other existing Monte Carlo code CEARPGA I and CEARPGA II, a new algorithm of sampling the prompt gamma rays produced from neutron capture reaction and neutron inelastic scattering reaction, is developed in this work. All the prompt gamma rays are taken into account by using this new algorithm. Before this work, the commonly used method is to interpolate the prompt gamma rays from the pre-calculated gamma-ray table. This technique works fine for the single spectrum. However it limits the capability to simulate the coincidence spectrum. The new algorithm samples the prompt gamma rays from the nucleus excitation scheme. The primary nuclear data library used to sample the prompt gamma rays comes from ENSDF library. Three cases are simulated and the simulated results are benchmarked with experiments. The first case is the prototype for ETI PGNAA application. This case is designed to check the capability of CEARCPG for single spectrum simulation. The second
MCNP: a general Monte Carlo code for neutron and photon transport
Forster, R.A.; Godfrey, T.N.K.
1985-01-01
MCNP is a very general Monte Carlo neutron photon transport code system with approximately 250 person years of Group X-6 code development invested. It is extremely portable, user-oriented, and a true production code as it is used about 60 Cray hours per month by about 150 Los Alamos users. It has as its data base the best cross-section evaluations available. MCNP contains state-of-the-art traditional and adaptive Monte Carlo techniques to be applied to the solution of an ever-increasing number of problems. Excellent user-oriented documentation is available for all facets of the MCNP code system. Many useful and important variants of MCNP exist for special applications. The Radiation Shielding Information Center (RSIC) in Oak Ridge, Tennessee is the contact point for worldwide MCNP code and documentation distribution. A much improved MCNP Version 3A will be available in the fall of 1985, along with new and improved documentation. Future directions in MCNP development will change the meaning of MCNP to Monte Carlo N Particle where N particle varieties will be transported.
Verification of SMART Neutronics Design Methodology by the MCNAP Monte Carlo Code
Jong Sung Chung; Kyung Jin Shim; Chang Hyo Kim; Chungchan Lee; Sung Quun Zee
2000-11-12
SMART is a small advanced integral pressurized water reactor (PWR) of 330 MW(thermal) designed for both electricity generation and seawater desalinization. The CASMO-3/MASTER nuclear analysis system, a design-basis of Korean PWR plants, has been employed for the SMART core nuclear design and analysis because the fuel assembly (FA) characteristics and reactor operating conditions in temperature and pressure are similar to those of PWR plants. However, the SMART FAs are highly poisoned with more than 20 Al{sub 2}O{sub 3}-B{sub 4}C plus additional Gd{sub 2}O{sub 3}/UO{sub 2} BPRs each FA. The reactor is operated with control rods inserted. Therefore, the flux and power distribution may become more distorted than those of commercial PWR plants. In addition, SMART should produce power from room temperature to hot-power operating condition because it employs nuclear heating from room temperature. This demands reliable predictions of core criticality, shutdown margin, control rod worth, power distributions, and reactivity coefficients at both room temperature and hot operating condition, yet no such data are available to verify the CASMO-3/MASTER (hereafter MASTER) code system. In the absence of experimental verification data for the SMART neutronics design, the Monte Carlo depletion analysis program MCNAP is adopted as near-term alternatives for qualifying MASTER neutronics design calculations. The MCNAP is a personal computer-based continuous energy Monte Carlo neutronics analysis program written in C++ language. We established its qualification by presenting its prediction accuracy on measurements of Venus critical facilities and core neutronics analysis of a PWR plant in operation, and depletion characteristics of integral burnable absorber FAs of the current PWR. Here, we present a comparison of MASTER and MCNAP neutronics design calculations for SMART and establish the qualification of the MASTER system.
Methods and codes for neutronic calculations of the MARIA research reactor.
Andrzejewski, K.; Kulikowska, T.; Bretscher, M. M.; Hanan, N. A.; Matos, J. E.
2002-02-18
The core of the MARIA high flux multipurpose research reactor is highly heterogeneous. It consists of beryllium blocks arranged in 6 x 8 matrix, tubular fuel assemblies, control rods and irradiation channels. The reflector is also heterogeneous and consists of graphite blocks clad with aluminum. Its structure is perturbed by the experimental beam tubes. This paper presents methods and codes used to calculate the MARIA reactor neutronics characteristics and experience gained thus far at IAE and ANL. At ANL the methods of MARIA calculations were developed in connection with the RERTR program. At IAE the package of programs was developed to help its operator in optimization of fuel utilization.
VERIFICATION OF THE INL/COMBINE7 NEUTRON ENERGY SPECTRUM CODE
Barry D. Ganapol; Woo Y. Yoon; David W. Nigg
2008-09-01
We construct semi-analytic benchmarks for the neutron slowing down equations in the thermal, resonance and fast energy regimes through mathematical embedding. The method features a fictitious time-dependent slowing down equations solved via Taylor series expansion over discrete “time” intervals. Two classes of benchmarks are considered- the first treats methods of solution and the second the multigroup approximation itself. We present several meaningful benchmark methods comparisons with the COMBINE7 energy spectrum code and a simple demonstration of convergence of the multigroup approximation.
1996-12-19
Version 03 The NJOY nuclear data processing system is a comprehensive computer code system for producing pointwise and multigroup cross sections and related quantities from ENDF/B evaluated nuclear data in the ENDF format, including the latest US library, ENDF/B-VI. The NJOY code works with neutrons, photons, and charged particles and produces libraries for a wide variety of particle transport and reactor analysis codes.
Bruyères-le-Châtel Neutron Evaluations of Actinides with the TALYS Code: The Fission Channel
NASA Astrophysics Data System (ADS)
Romain, P.; Morillon, B.; Duarte, H.
2016-01-01
For several years, various neutron evaluations of plutonium and uranium isotopes have been performed at Bruyères-le-Châtel (BRC), from 1 keV up to 30 MeV. Since only nuclear reaction models have been used to produce these evaluations, our approach was named the "Full Model" approach. Total, shape elastic and direct inelastic cross sections were obtained from the coupled channels model using a dispersive optical potential developed for actinides, with a large enough coupling scheme including the lowest octupolar band. All other cross sections were calculated using the Hauser-Feshbach theory (TALYS code) with a pre-equilibrium component above 8-10 MeV. In this paper, we focus our attention on the fission channel. More precisely, we will present the BRC contribution to fission modeling and the philosophy adopted in our "Full Model" approach. Performing evaluations with the "Full Model" approach implies the optimization of a large number of model parameters. With increasing neutron incident energy, many residual nuclei produced by nucleon emission also lead to fission. All available experimental data assigned to various fission mechanisms of the same nucleus were used to determine fission barrier parameters. For uranium isotopes, triple-humped fission barriers were required in order to reproduce accurately variations of the experimental fission cross sections. Our BRC fission modeling has shown that the effects of the class II or class III states located in the wells of the fission barrier sometimes provide an anti-resonant transmission rather than a resonant one. Consistent evaluations were produced for a large series of U and Pu isotopes. Resulting files were tested against integral data.
Psychometric properties of the system for coding couples' interactions in therapy--alcohol.
Owens, Mandy D; McCrady, Barbara S; Borders, Adrienne Z; Brovko, Julie M; Pearson, Matthew R
2014-12-01
Few systems are available for coding in-session behaviors for couples in therapy. Alcohol Behavioral Couple Therapy (ABCT) is an empirically supported treatment, but little is known about its mechanisms of behavior change. In the current study, an adapted version of the Motivational Interviewing for Significant Others coding system was developed into the System for Coding Couples' Interactions in Therapy-Alcohol (SCCIT-A), which was used to code couples' interactions and behaviors during ABCT. Results showed good interrater reliability of the SCCIT-A and provided evidence that the SCCIT-A may be a promising measure for understanding couples in therapy. A 3-factor model of the SCCIT-A (Positive, Negative, and Change Talk/Counter-Change Talk) was examined using a confirmatory factor analysis, but model fit was poor. Because model fit was poor, ratios were computed for Positive/Negative ratings and for Change Talk/Counter-Change Talk codes based on previous research in the couples and Motivational Interviewing literature. Post hoc analyses examined correlations between specific SCCIT-A codes and baseline characteristics, and indicated some concurrent validity. Correlations were run between ratios and baseline characteristics; ratios may be an alternative to using the factors from the SCCIT-A. Reliability and validity analyses suggest that the SCCIT-A has the potential to be a useful measure for coding in-session behaviors of both partners in couples therapy and could be used to identify mechanisms of behavior change for ABCT. Additional research is needed to improve the reliability of some codes and to further develop the SCCIT-A and other measures of couples' interactions in therapy.
A portable, parallel, object-oriented Monte Carlo neutron transport code in C++
Lee, S.R.; Cummings, J.C.; Nolen, S.D. |
1997-05-01
We have developed a multi-group Monte Carlo neutron transport code using C++ and the Parallel Object-Oriented Methods and Applications (POOMA) class library. This transport code, called MC++, currently computes k and {alpha}-eigenvalues and is portable to and runs parallel on a wide variety of platforms, including MPPs, clustered SMPs, and individual workstations. It contains appropriate classes and abstractions for particle transport and, through the use of POOMA, for portable parallelism. Current capabilities of MC++ are discussed, along with physics and performance results on a variety of hardware, including all Accelerated Strategic Computing Initiative (ASCI) hardware. Current parallel performance indicates the ability to compute {alpha}-eigenvalues in seconds to minutes rather than hours to days. Future plans and the implementation of a general transport physics framework are also discussed.
Zhang, Tiankui; Hu, Huasi; Jia, Qinggang; Zhang, Fengna; Chen, Da; Li, Zhenghong; Wu, Yuelei; Liu, Zhihua; Hu, Guang; Guo, Wei
2012-11-01
Monte-Carlo simulation of neutron coded imaging based on encoding aperture for Z-pinch of large field-of-view with 5 mm radius has been investigated, and then the coded image has been obtained. Reconstruction method of source image based on genetic algorithms (GA) has been established. "Residual watermark," which emerges unavoidably in reconstructed image, while the peak normalization is employed in GA fitness calculation because of its statistical fluctuation amplification, has been discovered and studied. Residual watermark is primarily related to the shape and other parameters of the encoding aperture cross section. The properties and essential causes of the residual watermark were analyzed, while the identification on equivalent radius of aperture was provided. By using the equivalent radius, the reconstruction can also be accomplished without knowing the point spread function (PSF) of actual aperture. The reconstruction result is close to that by using PSF of the actual aperture.
Zhang Tiankui; Hu Huasi; Jia Qinggang; Zhang Fengna; Liu Zhihua; Hu Guang; Guo Wei; Chen Da; Li Zhenghong; Wu Yuelei
2012-11-15
Monte-Carlo simulation of neutron coded imaging based on encoding aperture for Z-pinch of large field-of-view with 5 mm radius has been investigated, and then the coded image has been obtained. Reconstruction method of source image based on genetic algorithms (GA) has been established. 'Residual watermark,' which emerges unavoidably in reconstructed image, while the peak normalization is employed in GA fitness calculation because of its statistical fluctuation amplification, has been discovered and studied. Residual watermark is primarily related to the shape and other parameters of the encoding aperture cross section. The properties and essential causes of the residual watermark were analyzed, while the identification on equivalent radius of aperture was provided. By using the equivalent radius, the reconstruction can also be accomplished without knowing the point spread function (PSF) of actual aperture. The reconstruction result is close to that by using PSF of the actual aperture.
M. J. Russell
2006-06-01
This is an assessment of codes and standards applicable to a hydrogen production plant to be coupled to a nuclear reactor. The result of the assessment is a list of codes and standards that are expected to be applicable to the plant during its design and construction.
Santos-Villalobos, Hector J; Gregor, Jens; Bingham, Philip R
2014-01-01
At the present, neutron sources cannot be fabricated small and powerful enough in order to achieve high resolution radiography while maintaining an adequate flux. One solution is to employ computational imaging techniques such as a Magnified Coded Source Imaging (CSI) system. A coded-mask is placed between the neutron source and the object. The system resolution is increased by reducing the size of the mask holes and the flux is increased by increasing the size of the coded-mask and/or the number of holes. One limitation of such system is that the resolution of current state-of-the-art scintillator-based detectors caps around 50um. To overcome this challenge, the coded-mask and object are magnified by making the distance from the coded-mask to the object much smaller than the distance from object to detector. In previous work, we have shown via synthetic experiments that our least squares method outperforms other methods in image quality and reconstruction precision because of the modeling of the CSI system components. However, the validation experiments were limited to simplistic neutron sources. In this work, we aim to model the flux distribution of a real neutron source and incorporate such a model in our least squares computational system. We provide a full description of the methodology used to characterize the neutron source and validate the method with synthetic experiments.
RSAP - A Code for Display of Neutron Cross Section Data and SAMMY Fit Results
Sayer, R.O.
2001-02-02
RSAP is a computer code for display of neutron cross section data and selected SAMMY output. SAMMY is a multilevel R-matrix code for fitting neutron time-of-flight cross-section data using Bayes' method. RSAP, which runs on the Digital Unix Alpha platform, reads ORELA Data Files (ODF) created by SAMMY and uses graphics routines from the PLPLOT package. In addition, RSAP can read data and/or computed values from ASCII files with a format specified by the user. Plot output may be displayed in an X window, sent to a postscript file (rsap.ps), or sent to a color postscript file (rsap.psc). Thirteen plot types are supported, allowing the user to display cross section data, transmission data, errors, theory, Bayes fits, and residuals in various combinations. In this document the designations theory and Bayes refer to the initial and final theoretical cross sections, respectively, as evaluated by SAMMY. Special plot types include Bayes/Data, Theory--Data, and Bayes--Data. Output from two SAMMY runs may be compared by plotting the ratios Theory2/Theory1 and Bayes2/Bayes1 or by plotting the differences (Theory2-Theory1) and (Bayes2-Bayes1).
Yang, W. S.; Lee, C. H.
2008-05-16
Under the fast reactor simulation program launched in April 2007, development of an advanced multigroup cross section generation code was initiated in July 2007, in conjunction with the development of the high-fidelity deterministic neutron transport code UNIC. The general objectives are to simplify the existing multi-step schemes and to improve the resolved and unresolved resonance treatments. Based on the review results of current methods and the fact that they have been applied successfully to fast critical experiment analyses and fast reactor designs for last three decades, the methodologies of the ETOE-2/MC{sup 2}-2/SDX code system were selected as the starting set of methodologies for multigroup cross section generation for fast reactor analysis. As the first step for coupling with the UNIC code and use in a parallel computing environment, the MC{sup 2}-2 code was updated by modernizing the memory structure and replacing old data management package subroutines and functions with FORTRAN 90 based routines. Various modifications were also made in the ETOE-2 and MC{sup 2}-2 codes to process the ENDF/B-VII.0 data properly. Using the updated ETOE-2/MC{sup 2}-2 code system, the ENDF/B-VII.0 data was successfully processed for major heavy and intermediate nuclides employed in sodium-cooled fast reactors. Initial verification tests of the MC{sup 2}-2 libraries generated from ENDF/B-VII.0 data were performed by inter-comparison of twenty-one group infinite dilute total cross sections obtained from MC{sup 2}-2, VIM, and NJOY. For almost all nuclides considered, MC{sup 2}-2 cross sections agreed very well with those from VIM and NJOY. Preliminary validation tests of the ENDF/B-VII.0 libraries of MC{sup 2}-2 were also performed using a set of sixteen fast critical benchmark problems. The deterministic results based on MC{sup 2}-2/TWODANT calculations were in good agreement with MCNP solutions within {approx}0.25% {Delta}{rho}, except a few small LANL fast assemblies
NASA Astrophysics Data System (ADS)
Ortiz-Rodríguez, J. M.; Reyes Alfaro, A.; Reyes Haro, A.; Solís Sánches, L. O.; Miranda, R. Castañeda; Cervantes Viramontes, J. M.; Vega-Carrillo, H. R.
2013-07-01
In this work a neutron spectrum unfolding code, based on artificial intelligence technology is presented. The code called "Neutron Spectrometry and Dosimetry with Artificial Neural Networks and two Bonner spheres", (NSDann2BS), was designed in a graphical user interface under the LabVIEW programming environment. The main features of this code are to use an embedded artificial neural network architecture optimized with the "Robust design of artificial neural networks methodology" and to use two Bonner spheres as the only piece of information. In order to build the code here presented, once the net topology was optimized and properly trained, knowledge stored at synaptic weights was extracted and using a graphical framework build on the LabVIEW programming environment, the NSDann2BS code was designed. This code is friendly, intuitive and easy to use for the end user. The code is freely available upon request to authors. To demonstrate the use of the neural net embedded in the NSDann2BS code, the rate counts of 252Cf, 241AmBe and 239PuBe neutron sources measured with a Bonner spheres system.
Ortiz-Rodriguez, J. M.; Reyes Alfaro, A.; Reyes Haro, A.; Solis Sanches, L. O.; Miranda, R. Castaneda; Cervantes Viramontes, J. M.; Vega-Carrillo, H. R.
2013-07-03
In this work a neutron spectrum unfolding code, based on artificial intelligence technology is presented. The code called ''Neutron Spectrometry and Dosimetry with Artificial Neural Networks and two Bonner spheres'', (NSDann2BS), was designed in a graphical user interface under the LabVIEW programming environment. The main features of this code are to use an embedded artificial neural network architecture optimized with the ''Robust design of artificial neural networks methodology'' and to use two Bonner spheres as the only piece of information. In order to build the code here presented, once the net topology was optimized and properly trained, knowledge stored at synaptic weights was extracted and using a graphical framework build on the LabVIEW programming environment, the NSDann2BS code was designed. This code is friendly, intuitive and easy to use for the end user. The code is freely available upon request to authors. To demonstrate the use of the neural net embedded in the NSDann2BS code, the rate counts of {sup 252}Cf, {sup 241}AmBe and {sup 239}PuBe neutron sources measured with a Bonner spheres system.
Alotaibi, Ghazi S; Wu, Cynthia; Senthilselvan, Ambikaipakan; McMurtry, M Sean
2015-08-01
The purpose of this study was to evaluate the accuracy of using a combination of International Classification of Diseases (ICD) diagnostic codes and imaging procedure codes for identifying deep vein thrombosis (DVT) and pulmonary embolism (PE) within administrative databases. Information from the Alberta Health (AH) inpatients and ambulatory care administrative databases in Alberta, Canada was obtained for subjects with a documented imaging study result performed at a large teaching hospital in Alberta to exclude venous thromboembolism (VTE) between 2000 and 2010. In 1361 randomly-selected patients, the proportion of patients correctly classified by AH administrative data, using both ICD diagnostic codes and procedure codes, was determined for DVT and PE using diagnoses documented in patient charts as the gold standard. Of the 1361 patients, 712 had suspected PE and 649 had suspected DVT. The sensitivities for identifying patients with PE or DVT using administrative data were 74.83% (95% confidence interval [CI]: 67.01-81.62) and 75.24% (95% CI: 65.86-83.14), respectively. The specificities for PE or DVT were 91.86% (95% CI: 89.29-93.98) and 95.77% (95% CI: 93.72-97.30), respectively. In conclusion, when coupled with relevant imaging codes, VTE diagnostic codes obtained from administrative data provide a relatively sensitive and very specific method to ascertain acute VTE. PMID:25834115
Khattab, K; Sulieman, I
2009-04-01
The MCNP-4C code, based on the probabilistic approach, was used to model the 3D configuration of the core of the Syrian miniature neutron source reactor (MNSR). The continuous energy neutron cross sections from the ENDF/B-VI library were used to calculate the thermal and fast neutron fluxes in the inner and outer irradiation sites of MNSR. The thermal fluxes in the MNSR inner irradiation sites were also measured experimentally by the multiple foil activation method ((197)Au (n, gamma) (198)Au and (59)Co (n, gamma) (60)Co). The foils were irradiated simultaneously in each of the five MNSR inner irradiation sites to measure the thermal neutron flux and the epithermal index in each site. The calculated and measured results agree well.
NASA Astrophysics Data System (ADS)
Chen, Y.; Fischer, U.
2005-10-01
A program system for three-dimensional coupled Monte Carlo-deterministic shielding analysis has been developed to solve problems with complex geometry and bulk shield by integrating the Monte Carlo transport code MCNP, the three-dimensional discrete ordinates code TORT and a coupling interface program. A newly-proposed mapping approach is implemented in the interface program to calculate the angular flux distribution from the scored Monte Carlo particle tracks and generate the boundary source file for the use of TORT. Test calculations were performed with comparison to MCNP solutions. Satisfactory agreements were obtained between the results calculated by these two approaches. The program system has been chosen to treat the complicated shielding problem of the accelerator-based IFMIF neutron source. The successful application demonstrates that coupling scheme with the program system is a useful computational tool for the shielding analysis of complex and large nuclear facilities.
A Complex-Geometry Validation Experiment for Advanced Neutron Transport Codes
David W. Nigg; Anthony W. LaPorta; Joseph W. Nielsen; James Parry; Mark D. DeHart; Samuel E. Bays; William F. Skerjanc
2013-11-01
The Idaho National Laboratory (INL) has initiated a focused effort to upgrade legacy computational reactor physics software tools and protocols used for support of core fuel management and experiment management in the Advanced Test Reactor (ATR) and its companion critical facility (ATRC) at the INL.. This will be accomplished through the introduction of modern high-fidelity computational software and protocols, with appropriate new Verification and Validation (V&V) protocols, over the next 12-18 months. Stochastic and deterministic transport theory based reactor physics codes and nuclear data packages that support this effort include MCNP5[1], SCALE/KENO6[2], HELIOS[3], SCALE/NEWT[2], and ATTILA[4]. Furthermore, a capability for sensitivity analysis and uncertainty quantification based on the TSUNAMI[5] system has also been implemented. Finally, we are also evaluating the Serpent[6] and MC21[7] codes, as additional verification tools in the near term as well as for possible applications to full three-dimensional Monte Carlo based fuel management modeling in the longer term. On the experimental side, several new benchmark-quality code validation measurements based on neutron activation spectrometry have been conducted using the ATRC. Results for the first four experiments, focused on neutron spectrum measurements within the Northwest Large In-Pile Tube (NW LIPT) and in the core fuel elements surrounding the NW LIPT and the diametrically opposite Southeast IPT have been reported [8,9]. A fifth, very recent, experiment focused on detailed measurements of the element-to-element core power distribution is summarized here and examples of the use of the measured data for validation of corresponding MCNP5, HELIOS, NEWT, and Serpent computational models using modern least-square adjustment methods are provided.
Modeling equal and unequal mass binary neutron star mergers using public codes
NASA Astrophysics Data System (ADS)
De Pietri, Roberto; Feo, Alessandra; Maione, Francesco; Löffler, Frank
2016-03-01
We present three-dimensional simulations of the dynamics of binary neutron star mergers from the late stage of the inspiral process up to ˜20 ms after the system has merged, either to form a hypermassive neutron star or a rotating black hole. We investigate five equal mass models of total gravitational mass 2.207, 2.373, 2.537, 2.697, and 2.854 M⊙, respectively, and four unequal mass models with MADM≃2.53 M⊙ and q ≃0.94 , 0.88, 0.83, and 0.77 (where q =M(1 )/M(2 ) is the mass ratio). We use a semirealistic equation of state, namely, the seven-segment piecewise polytropic SLyPP with a thermal component given by Γth=1.8 . We have also compared the resulting dynamics (for one model) using both the BSSN-NOK and CCZ4 methods for the evolution of the gravitational sector and also different reconstruction methods for the matter sector, namely, PPM, WENO, and MP5. Our results show agreement at high resolution, but superiority of BSSN-NOK supplemented by WENO reconstruction at lower resolutions. One of the important characteristics of the present investigation is that for the first time it has been done using only publicly available open source software: the Einstein Toolkit code, deployed for the dynamical evolution, and the LORENE code, for the generation of the initial models. All of the source code and parameters used for the simulations have been made publicly available. This not only makes it possible to rerun and reanalyze our data but also enables others to directly build upon this work for future research.
NASA Astrophysics Data System (ADS)
Monticello, D. A.; Reiman, A. H.; Watanabe, K. Y.; Nakajima, N.; Okamoto, M.
1997-11-01
The existence of bootstrap currents in both tokamaks and stellarators was confirmed, experimentally, more than ten years ago. Such currents can have significant effects on the equilibrium and stability of these MHD devices. In addition, stellarators, with the notable exception of W7-X, are predicted to have such large bootstrap currents that reliable equilibrium calculations require the self-consistent evaluation of bootstrap currents. Modeling of discharges which contain islands requires an algorithm that does not assume good surfaces. Only one of the two 3-D equilibrium codes that exist, PIES( Reiman, A. H., Greenside, H. S., Compt. Phys. Commun. 43), (1986)., can easily be modified to handle bootstrap current. Here we report on the coupling of the PIES 3-D equilibrium code and NIFS bootstrap code(Watanabe, K., et al., Nuclear Fusion 35) (1995), 335.
The Transient 3-D Transport Coupled Code TORT-TD/ATTICA3D for High-Fidelity Pebble-Bed HTGR Analyses
NASA Astrophysics Data System (ADS)
Seubert, Armin; Sureda, Antonio; Lapins, Janis; Bader, Johannes; Laurien, Eckart
2012-01-01
This article describes the 3D discrete ordinates-based coupled code system TORT-TD/ATTICA3D that aims at steady state and transient analyses of pebble-bed high-temperature gas cooled reactors. In view of increasing computing power, the application of time-dependent neutron transport methods becomes feasible for best estimate evaluations of safety margins. The calculation capabilities of TORT-TD/ATTICA3D are presented along with the coupling approach, with focus on the time-dependent neutron transport features of TORT-TD. Results obtained for the OECD/NEA/NSC PBMR-400 benchmark demonstrate the transient capabilities of TORT-TD/ATTICA3D.
Samarin, V. V.
2015-10-15
The fusion cross sections for the {sup 17,18}O+{sup 27}Al, {sup 18}O+{sup 58}Ni, and {sup 6}He+{sup 197}Au reactions were calculated by the coupled-channel method. The radial dependence of matrices that describe coupling to valence-neutron-rearrangement channels was determined with the aid of two-center wave functions. The coupling-strength parameters were evaluated on the basis of numerically solving the time-dependent Schrödinger equation. Satisfactory agreement with experimental data was obtained.
Sayer, R.O.
2003-07-29
RSAP [1] is a computer code for display and manipulation of neutron cross section data and selected SAMMY output. SAMMY [2] is a multilevel R-matrix code for fitting neutron time-of-flight cross-section data using Bayes' method. This users' guide provides documentation for the recently updated RSAP code (version 6). The code has been ported to the Linux platform, and several new features have been added, including the capability to read cross section data from ASCII pointwise ENDF files as well as double-precision PLT output from SAMMY. A number of bugs have been found and corrected, and the input formats have been improved. Input items are parsed so that items may be separated by spaces or commas.
MC++: A parallel, portable, Monte Carlo neutron transport code in C++
Lee, S.R.; Cummings, J.C.; Nolen, S.D.
1997-03-01
MC++ is an implicit multi-group Monte Carlo neutron transport code written in C++ and based on the Parallel Object-Oriented Methods and Applications (POOMA) class library. MC++ runs in parallel on and is portable to a wide variety of platforms, including MPPs, SMPs, and clusters of UNIX workstations. MC++ is being developed to provide transport capabilities to the Accelerated Strategic Computing Initiative (ASCI). It is also intended to form the basis of the first transport physics framework (TPF), which is a C++ class library containing appropriate abstractions, objects, and methods for the particle transport problem. The transport problem is briefly described, as well as the current status and algorithms in MC++ for solving the transport equation. The alpha version of the POOMA class library is also discussed, along with the implementation of the transport solution algorithms using POOMA. Finally, a simple test problem is defined and performance and physics results from this problem are discussed on a variety of platforms.
Validation of the new code package APOLLO2.8 for accurate PWR neutronics calculations
Santamarina, A.; Bernard, D.; Blaise, P.; Leconte, P.; Palau, J. M.; Roque, B.; Vaglio, C.; Vidal, J. F.
2013-07-01
This paper summarizes the Qualification work performed to demonstrate the accuracy of the new APOLLO2.S/SHEM-MOC package based on JEFF3.1.1 nuclear data file for the prediction of PWR neutronics parameters. This experimental validation is based on PWR mock-up critical experiments performed in the EOLE/MINERVE zero-power reactors and on P.I. Es on spent fuel assemblies from the French PWRs. The Calculation-Experiment comparison for the main design parameters is presented: reactivity of UOX and MOX lattices, depletion calculation and fuel inventory, reactivity loss with burnup, pin-by-pin power maps, Doppler coefficient, Moderator Temperature Coefficient, Void coefficient, UO{sub 2}-Gd{sub 2}O{sub 3} poisoning worth, Efficiency of Ag-In-Cd and B4C control rods, Reflector Saving for both standard 2-cm baffle and GEN3 advanced thick SS reflector. From this qualification process, calculation biases and associated uncertainties are derived. This code package APOLLO2.8 is already implemented in the ARCADIA new AREVA calculation chain for core physics and is currently under implementation in the future neutronics package of the French utility Electricite de France. (authors)
Code System for Calculating Alpha, N; Spontaneous Fission; and Delayed Neutron Sources and Spectra.
2002-07-18
Version: 04 SOURCES4C is a code system that determines neutron production rates and spectra from (alpha,n) reactions, spontaneous fission, and delayed neutron emission due to radionuclide decay. In this release the three-region problem was modified to correct several bugs, and new documentation was added to the package. Details are available in the included LA-UR-02-1617 (2002) report. The code is capable of calculating (alpha,n) source rates and spectra in four types of problems: homogeneous media (i.e.,more » an intimate mixture of alpha-emitting source material and low-Z target material), two-region interface problems (i.e., a slab of alpha-emitting source material in contact with a slab of low-Z target material), three-region interface problems (i.e., a thin slab of low-Z target material sandwiched between alpha-emitting source material and low-Z target material), and (alpha,n) reactions induced by a monoenergetic beam of alpha-particles incident on a slab of target material. The process of creating a SOURCES input file (tape1) is streamlined with the Los Alamos SOURCES Tape1 Creator and Library Link (LASTCALL) Version 1. Intended to supplement the SOURCES manual, LASTCALL is a simple graphical user interface designed to minimize common errors made during input. The optional application, LASTCALL, consists of a single dialog window launched from an executable (lastcall.exe) on Windows-based personal computers.« less
Prettyman, T.H.; Gardner, R.P.; Verghese, K. . Center for Engineering Applications and Radioisotopes)
1993-08-01
A new specific purpose Monte Carlo code called McENL for modeling the time response of epithermal neutron lifetime tools is described. The code was developed so that the Monte Carlo neophyte can easily use it. A minimum amount of input preparation is required and specified fixed values of the parameters used to control the code operation can be used. The weight windows technique, employing splitting and Russian Roulette, is used with an automated importance function based on the solution of an adjoint diffusion model to improve the code efficiency. Complete composition and density correlated sampling is also included in the code and can be used to study the effect on tool response of small variations in the formation, borehole, or logging tool composition and density. An illustration of the latter application is given here for the density of a thermal neutron filter. McENL was benchmarked against test-pit data for the Mobil pulsed neutron porosity (PNP) tool and found to be very accurate. Results of the experimental validation and details of code performance are presented.
Osiris: A Modern, High-Performance, Coupled, Multi-Physics Code For Nuclear Reactor Core Analysis
Procassini, R J; Chand, K K; Clouse, C J; Ferencz, R M; Grandy, J M; Henshaw, W D; Kramer, K J; Parsons, I D
2007-02-26
To meet the simulation needs of the GNEP program, LLNL is leveraging a suite of high-performance codes to be used in the development of a multi-physics tool for modeling nuclear reactor cores. The Osiris code project, which began last summer, is employing modern computational science techniques in the development of the individual physics modules and the coupling framework. Initial development is focused on coupling thermal-hydraulics and neutral-particle transport, while later phases of the project will add thermal-structural mechanics and isotope depletion. Osiris will be applicable to the design of existing and future reactor systems through the use of first-principles, coupled physics models with fine-scale spatial resolution in three dimensions and fine-scale particle-energy resolution. Our intent is to replace an existing set of legacy, serial codes which require significant approximations and assumptions, with an integrated, coupled code that permits the design of a reactor core using a first-principles physics approach on a wide range of computing platforms, including the world's most powerful parallel computers. A key research activity of this effort deals with the efficient and scalable coupling of physics modules which utilize rather disparate mesh topologies. Our approach allows each code module to use a mesh topology and resolution that is optimal for the physics being solved, and employs a mesh-mapping and data-transfer module to effect the coupling. Additional research is planned in the area of scalable, parallel thermal-hydraulics, high-spatial-accuracy depletion and coupled-physics simulation using Monte Carlo transport.
Wang, G. B.; Wang, K.; Liu, H. G.; Li, R. D.
2013-07-01
A Monte Carlo tool RSMC (Reaction Sequence Monte Carlo) was developed to simulate deuteron/triton transportation and reaction coupled problem. The 'Forced particle production' variance reduction technique was used to improve the simulation speed, which made the secondary product play a major role. The mono-energy 14 MeV fusion neutron source was employed as a validation. Then the thermal-to-fusion neutron convertor was studied with our tool. Moreover, an in-core conversion efficiency measurement experiment was performed with {sup 6}LiD and {sup 6}LiH converters. Threshold activation foils was used to indicate the fast and fusion neutron flux. Besides, two other pivotal parameters were calculated theoretically. Finally, the conversion efficiency of {sup 6}LiD is obtained as 1.97x10{sup -4}, which matches well with the theoretical result. (authors)
Core-coupled states and split proton-neutron quasiparticle multiplets in 122-126Ag
NASA Astrophysics Data System (ADS)
Lalkovski, S.; Bruce, A. M.; Jungclaus, A.; Górska, M.; Pfützner, M.; Cáceres, L.; Naqvi, F.; Pietri, S.; Podolyák, Zs.; Simpson, G. S.; Andgren, K.; Bednarczyk, P.; Beck, T.; Benlliure, J.; Benzoni, G.; Casarejos, E.; Cederwall, B.; Crespi, F. C. L.; Cuenca-García, J. J.; Cullen, I. J.; Denis Bacelar, A. M.; Detistov, P.; Doornenbal, P.; Farrelly, G. F.; Garnsworthy, A. B.; Geissel, H.; Gelletly, W.; Gerl, J.; Grebosz, J.; Hadinia, B.; Hellström, M.; Hinke, C.; Hoischen, R.; Ilie, G.; Jaworski, G.; Jolie, J.; Khaplanov, A.; Kisyov, S.; Kmiecik, M.; Kojouharov, I.; Kumar, R.; Kurz, N.; Maj, A.; Mandal, S.; Modamio, V.; Montes, F.; Myalski, S.; Palacz, M.; Prokopowicz, W.; Reiter, P.; Regan, P. H.; Rudolph, D.; Schaffner, H.; Sohler, D.; Steer, S. J.; Tashenov, S.; Walker, J.; Walker, P. M.; Weick, H.; Werner-Malento, E.; Wieland, O.; Wollersheim, H. J.; Zhekova, M.
2013-03-01
Neutron-rich silver isotopes were populated in the fragmentation of a 136Xe beam and the relativistic fission of 238U. The fragments were mass analyzed with the GSI Fragment Separator and subsequently implanted into a passive stopper. Isomeric transitions were detected by 105 high-purity germanium detectors. Eight isomeric states were observed in 122-126Ag nuclei. The level schemes of 122,123,125Ag were revised and extended with isomeric transitions being observed for the first time. The excited states in the odd-mass silver isotopes are interpreted as core-coupled states. The isomeric states in the even-mass silver isotopes are discussed in the framework of the proton-neutron split multiplets. The results of shell-model calculations, performed for the most neutron-rich silver nuclei are compared to the experimental data.
NASA Technical Reports Server (NTRS)
Armstrong, T. W.
1972-01-01
Several Monte Carlo radiation transport computer codes are used to predict quantities of interest in the fields of radiotherapy and radiobiology. The calculational methods are described and comparisions of calculated and experimental results are presented for dose distributions produced by protons, neutrons, and negatively charged pions. Comparisons of calculated and experimental cell survival probabilities are also presented.
Incorporation of coupled nonequilibrium chemistry into a two-dimensional nozzle code (SEAGULL)
NASA Technical Reports Server (NTRS)
Ratliff, A. W.
1979-01-01
A two-dimensional multiple shock nozzle code (SEAGULL) was extended to include the effects of finite rate chemistry. The basic code that treats multiple shocks and contact surfaces was fully coupled with a generalized finite rate chemistry and vibrational energy exchange package. The modified code retains all of the original SEAGULL features plus the capability to treat chemical and vibrational nonequilibrium reactions. Any chemical and/or vibrational energy exchange mechanism can be handled as long as thermodynamic data and rate constants are available for all participating species.
The TORT three-dimensional discrete ordinates neutron/photon transport code (TORT version 3)
Rhoades, W.A.; Simpson, D.B.
1997-10-01
TORT calculates the flux or fluence of neutrons and/or photons throughout three-dimensional systems due to particles incident upon the system`s external boundaries, due to fixed internal sources, or due to sources generated by interaction with the system materials. The transport process is represented by the Boltzman transport equation. The method of discrete ordinates is used to treat the directional variable, and a multigroup formulation treats the energy dependence. Anisotropic scattering is treated using a Legendre expansion. Various methods are used to treat spatial dependence, including nodal and characteristic procedures that have been especially adapted to resist numerical distortion. A method of body overlay assists in material zone specification, or the specification can be generated by an external code supplied by the user. Several special features are designed to concentrate machine resources where they are most needed. The directional quadrature and Legendre expansion can vary with energy group. A discontinuous mesh capability has been shown to reduce the size of large problems by a factor of roughly three in some cases. The emphasis in this code is a robust, adaptable application of time-tested methods, together with a few well-tested extensions.
A Visualization Code System for Gamma and Neutron Shielding Calculations, Version 2.0
2008-08-01
EASYQAD, Version 2.0, is a standalone Windows XP or Windows 7 code system which facilitates gamma and neutron shielding calculations with user friendly graphical interfaces. It is used to analyze radiation shielding problems and includes: - 8 kinds of geometry types - Various flexible source options - Common material library - Various detector types The update contents of EASYQAD Version 2.0 are below: - Addition of starting option with P-code files - Addition of multi-sourcemore » calculation function - Expansion of source geometries - Addition of warning message - Modifications of EASYQAD program errors a. Coordination application problem in source division b. Source position error c. Rotation problem of source geometry d. Program running error in using more than six gamma energy distribution e. EASYQAD display problem of the right elliptic cylinder, ellipsoid and truncated right cone geometries Through intuitive windows and their interactions inside EASYQAD, the user can specify the dimensions of 3D-shapes, their material compositions, their densities, the type of radioactive sources, the locations of the sources, the type and positions of detectors. With the ease of using these sequences, shielding problems will become simpler and more clearly understandable to the analyzer. Furthermore, the error checking system can prevent users from making mistakes by automatically debugging the user inputs and giving modal dialog windows. The included AECL implementation of QAD-CGGP-A, Version 95.2 (C00645MNYCP00), is run from the user interface.« less
Burns, Kimberly A.
2009-08-01
The accurate and efficient simulation of coupled neutron-photon problems is necessary for several important radiation detection applications. Examples include the detection of nuclear threats concealed in cargo containers and prompt gamma neutron activation analysis for nondestructive determination of elemental composition of unknown samples.
Inlet-Compressor Analysis Performed Using Coupled Computational Fluid Dynamics Codes
NASA Technical Reports Server (NTRS)
Cole, Gary L.; Suresh, Ambady; Townsend, Scott
1999-01-01
A thorough understanding of dynamic interactions between inlets and compressors is extremely important to the design and development of propulsion control systems, particularly for supersonic aircraft such as the High-Speed Civil Transport (HSCT). Computational fluid dynamics (CFD) codes are routinely used to analyze individual propulsion components. By coupling the appropriate CFD component codes, it is possible to investigate inlet-compressor interactions. The objectives of this work were to gain a better understanding of inlet-compressor interaction physics, formulate a more realistic compressor-face boundary condition for time-accurate CFD simulations of inlets, and to take a first step toward the CFD simulation of an entire engine by coupling multidimensional component codes. This work was conducted at the NASA Lewis Research Center by a team of civil servants and support service contractors as part of the High Performance Computing and Communications Program (HPCCP).
An introduction to LIME 1.0 and its use in coupling codes for multiphysics simulations.
Belcourt, Noel; Pawlowski, Roger Patrick; Schmidt, Rodney Cannon; Hooper, Russell Warren
2011-11-01
LIME is a small software package for creating multiphysics simulation codes. The name was formed as an acronym denoting 'Lightweight Integrating Multiphysics Environment for coupling codes.' LIME is intended to be especially useful when separate computer codes (which may be written in any standard computer language) already exist to solve different parts of a multiphysics problem. LIME provides the key high-level software (written in C++), a well defined approach (with example templates), and interface requirements to enable the assembly of multiple physics codes into a single coupled-multiphysics simulation code. In this report we introduce important software design characteristics of LIME, describe key components of a typical multiphysics application that might be created using LIME, and provide basic examples of its use - including the customized software that must be written by a user. We also describe the types of modifications that may be needed to individual physics codes in order for them to be incorporated into a LIME-based multiphysics application.
NASA Astrophysics Data System (ADS)
Barnes, Michael
2009-11-01
To faithfully simulate ITER and other modern fusion devices, one must resolve electron and ion fluctuation scales in a five-dimensional phase space and time. Simultaneously, one must account for the interaction of this turbulence with the slow evolution of the large-scale plasma profiles. Because of the enormous range of scales involved and the high dimensionality of the problem, resolved first-principles simulations of the full core volume over the confinement time are very challenging using conventional (brute force) techniques. In order to address this problem, we have developed a new approach in which turbulence calculations from multiple gyrokinetic flux tube simulations are coupled together using gyrokinetic transport equations to obtain self-consistent equilibrium profiles and corresponding turbulent fluxes. This multi-scale approach is embodied in a new code, Trinity, which is capable of evolving equilibrium profiles for multiple species, including electromagnetic effects and realistic magnetic geometry, at a fraction of the cost of conventional direct numerical simulations. Key components in the cost reduction are the extreme parallelism enabled by the use of coupled flux tubes and the use of a nonlinear implicit algorithm to take large time steps when evolving the equilibrium. In this talk, we describe the multi-scale model employed in Trinity and present simulation results using nonlinear fluxes calculated with the gyrokinetic turbulence codes GS2 and GENE. We compare the numerical predictions from Trinity simulations with experimental results from a number of fusion devices, including JET and MAST.
A mono-dimensional nuclear fuel performance analysis code, PUMA, development from a coupled approach
Cheon, J. S.; Lee, B. O.; Lee, C. B.; Yacout, A. M.
2013-07-01
Multidimensional-multi-physical phenomena in nuclear fuels are treated as a set of mono-dimensional-coupled problems which encompass heat, displacement, fuel constituent redistribution, and fission gas release. Rather than uncoupling these coupled equations as in conventional fuel performance analysis codes, efforts are put into to obtain fully coupled solutions by relying on the recent advances of numerical analysis. Through this approach, a new SFR metal fuel performance analysis code, called PUMA (Performance of Uranium Metal fuel rod Analysis code) is under development. Although coupling between temperature and fuel constituent was made easily, the coupling between the mechanical equilibrium equation and a set of stiff kinetics equations for fission gas release is accomplished by introducing one-level Newton scheme through backward differentiation formula. Displacement equations from 1D finite element formulation of the mechanical equilibrium equation are solved simultaneously with stress equation, creep equation, swelling equation, and FGR equations. Calculations was made successfully such that the swelling and the hydrostatic pressure are interrelated each other. (authors)
Scrape-off layer modeling using coupled plasma and neutral transport codes
Stotler, D.P.; Coster, D.P.; Ehrdardt, A.B.; Karney, C.F.F.; Petravic, M.; Braams, B.J.
1992-05-01
An effort is made to refine the neutral transport model used in the B2 edge plasma code by coupling it to the DEGAS Monte Carlo code. Results are discussed for a simulation of a high recycling divertor. It appears that on the order of 100 iterations between the two codes are required to achieve a converged solution. However, the amount of computer time used in the DEGAS simulations is large, making complete runs impractical for design purposes. On the other hand, the differences in the resulting plasma parameters when compared to the B2 analytic neutrals model indicate that it would be worthwhile to explore techniques for speeding up the control system of codes.
Experimental validation of a coupled neutron-photon inverse radiation transport solver.
Mattingly, John K.; Harding, Lee; Mitchell, Dean James
2010-03-01
Forward radiation transport is the problem of calculating the radiation field given a description of the radiation source and transport medium. In contrast, inverse transport is the problem of inferring the configuration of the radiation source and transport medium from measurements of the radiation field. As such, the identification and characterization of special nuclear materials (SNM) is a problem of inverse radiation transport, and numerous techniques to solve this problem have been previously developed. The authors have developed a solver based on nonlinear regression applied to deterministic coupled neutron-photon transport calculations. The subject of this paper is the experimental validation of that solver. This paper describes a series of experiments conducted with a 4.5-kg sphere of alpha-phase, weapons-grade plutonium. The source was measured in six different configurations: bare, and reflected by high-density polyethylene (HDPE) spherical shells with total thicknesses of 1.27, 2.54, 3.81, 7.62, and 15.24 cm. Neutron and photon emissions from the source were measured using three instruments: a gross neutron counter, a portable neutron multiplicity counter, and a high-resolution gamma spectrometer. These measurements were used as input to the inverse radiation transport solver to characterize the solver's ability to correctly infer the configuration of the source from its measured signatures.
Monte Carlo Code System for Calculation of Multiple Scattering of Neutrons in the Resonance Region.
1983-01-25
Version 00 MCRTOF systematically calculates capture and scattering probabilities for neutrons incident on a material disk, with neutron cross sections calculated from the resonance parameters. Capture, front and rear face scattering, transmission, etc., probabilities are obtained from the average destinations of the incident neutrons.
The stability of tidally deformed neutron stars to three- and four-mode coupling
Venumadhav, Tejaswi; Zimmerman, Aaron; Hirata, Christopher M.
2014-01-20
It has recently been suggested that the tidal deformation of a neutron star excites daughter p- and g-modes to large amplitudes via a quasi-static instability. This would remove energy from the tidal bulge, resulting in dissipation and possibly affecting the phase evolution of inspiralling binary neutron stars and hence the extraction of binary parameters from gravitational wave observations. This instability appears to arise because of a large three-mode interaction among the tidal mode and high-order p- and g-modes of similar radial wavenumber. We show that additional four-mode interactions enter into the analysis at the same order as the three-mode terms previously considered. We compute these four-mode couplings by finding a volume-preserving coordinate transformation that relates the energy of a tidally deformed star to that of a radially perturbed spherical star. Using this method, we relate the four-mode coupling to three-mode couplings and show that there is a near-exact cancellation between the destabilizing effect of the three-mode interactions and the stabilizing effect of the four-mode interaction. We then show that the equilibrium tide is stable against the quasi-static decay into daughter p- and g-modes to leading order. The leading deviation from the quasi-static approximation due to orbital motion of the binary is considered; while it may slightly spoil the near-cancellation, any resulting instability timescale is at least of order the gravitational wave inspiral time. We conclude that the p-/g-mode coupling does not lead to a quasi-static instability, and does not impact the phase evolution of gravitational waves from binary neutron stars.
Keney, G.S.
1981-08-01
A computer code has been written to calculate neutron induced activation of neutral-beam injector components and the corresponding dose rates as a function of geometry, component composition, and time after shutdown. The code, ACDOS1, was written in FORTRAN IV to calculate both activity and dose rates for up to 30 target nuclides and 50 neutron groups. Sufficient versatility has also been incorporated into the code to make it applicable to a variety of general activation problems due to neutrons of energy less than 20 MeV.
Neutron-Gamma-ray-coupled albedo Monte Carlo method streaming analysis
Yamauchi, M.; Kawai, M.; Seki, Y.
1986-11-01
The neutron-gamma-ray-coupled albedo Monte Carlo (AMC) method has been developed and implemented in MORSE-I. The energy- and angle-dependent differential albedo data, which include secondary gamma rays, are calculated for a slab layer with one dimensional transport theory. Fundamental formulas for this method are described. The applicability to shielding design of fusion reactors is confirmed by analyzing the radiation streaming experiment conducted at the Fusion Neutronics Source facility, Japan Atomic Energy Research Institute. The AMC method has reproduced well the experimental data of radiation dose rates and spectra with an accuracy of approximately 10%. It is shown that the AMC method is several times more efficient than the ordinary Monte Carlo calculation in obtaining data necessary for the design with expected accuracy.
NASA Astrophysics Data System (ADS)
Williams, J. G.; Ribaric, A. P.; Schnauber, T.
2009-08-01
A new spectrum adjustment code, NSVA-3, has been developed and is being made available to the community. The name refers to Neutron Spectrum Validation and Adjustment. The designation NSVA-3 is a version of the code that simultaneously adjusts spectra for multiple environments. The code is written in MATLAB®, a high-level script language. The main advantage of the NSVA code is its use of graphic user interfaces (GUIs) to assist the user with the data input and in interactive execution of adjustment cases. Items of data may be easily swapped in or out of the calculation. As with previous least-squares adjustment codes, the data input requires the preparation of files for fluence spectra, dosimetry measurements, the standard deviations of each of these, and correlation matrices of each. In the case of multiple environments, the cross correlations between environments of the input fluence and dosimetry measurements can also be included. The GUI assists the user in keeping track of all of these files. An 89-group cross section library including covariance matrices is incorporated in the code package. The paper presents the basic theory used in the code, the limitations and assumptions that are built into this implementation, and will describe the operation of the code by means of an example problem.
Chatani, K. )
1992-08-01
This report summarizes the calculational results from analyses of a Clinch River Breeder Reactor (CRBR) prototypic coolant pipe chaseway neutron streaming experiment Comparisons of calculated and measured results are presented, major emphasis being placed on results at bends in the chaseway. Calculations were performed with three three-dimensional radiation transport codes: the discrete ordinates code TORT and the Monte Carlo code MORSE, both developed by the Oak Ridge National Laboratory (ORNL), and the discrete ordinates code ENSEMBLE, developed by Japan. The calculated results from the three codes are compared (1) with previously-calculated DOT3.5 two-dimensional results, (2) among themselves, and (3) with measured results. Calculations with TORT used both the weighted-difference and nodal methods. Only the weighted-difference method was used in ENSEMBLE. When the calculated results were compared to measured results, it was found that calculation-to-experiment (C/E) ratios were good in the regions of the chaseway where two-dimensional modeling might be difficult and where there were no significant discrete ordinates ray effects. Excellent agreement was observed for responses dominated by thermal neutron contributions. MORSE-calculated results and comparisons are described also, and detailed results are presented in an appendix.
ITS Version 6 : the integrated TIGER series of coupled electron/photon Monte Carlo transport codes.
Franke, Brian Claude; Kensek, Ronald Patrick; Laub, Thomas William
2008-04-01
ITS is a powerful and user-friendly software package permitting state-of-the-art Monte Carlo solution of lineartime-independent coupled electron/photon radiation transport problems, with or without the presence of macroscopic electric and magnetic fields of arbitrary spatial dependence. Our goal has been to simultaneously maximize operational simplicity and physical accuracy. Through a set of preprocessor directives, the user selects one of the many ITS codes. The ease with which the makefile system is applied combines with an input scheme based on order-independent descriptive keywords that makes maximum use of defaults and internal error checking to provide experimentalists and theorists alike with a method for the routine but rigorous solution of sophisticated radiation transport problems. Physical rigor is provided by employing accurate cross sections, sampling distributions, and physical models for describing the production and transport of the electron/photon cascade from 1.0 GeV down to 1.0 keV. The availability of source code permits the more sophisticated user to tailor the codes to specific applications and to extend the capabilities of the codes to more complex applications. Version 6, the latest version of ITS, contains (1) improvements to the ITS 5.0 codes, and (2) conversion to Fortran 90. The general user friendliness of the software has been enhanced through memory allocation to reduce the need for users to modify and recompile the code.
Wilson, W. B.; Perry, R. T.; Shores, E. F.; Charlton, W. S.; Parish, Theodore A.; Estes, G. P.; Brown, T. H.; Arthur, Edward D. ,; Bozoian, Michael; England, T. R.; Madland, D. G.; Stewart, J. E.
2002-01-01
SOURCES 4C is a computer code that determines neutron production rates and spectra from ({alpha},n) reactions, spontaneous fission, and delayed neutron emission due to radionuclide decay. The code is capable of calculating ({alpha},n) source rates and spectra in four types of problems: homogeneous media (i.e., an intimate mixture of a-emitting source material and low-Z target material), two-region interface problems (i.e., a slab of {alpha}-emitting source material in contact with a slab of low-Z target material), three-region interface problems (i.e., a thin slab of low-Z target material sandwiched between {alpha}-emitting source material and low-Z target material), and ({alpha},n) reactions induced by a monoenergetic beam of {alpha}-particles incident on a slab of target material. Spontaneous fission spectra are calculated with evaluated half-life, spontaneous fission branching, and Watt spectrum parameters for 44 actinides. The ({alpha},n) spectra are calculated using an assumed isotropic angular distribution in the center-of-mass system with a library of 107 nuclide decay {alpha}-particle spectra, 24 sets of measured and/or evaluated ({alpha},n) cross sections and product nuclide level branching fractions, and functional {alpha}-particle stopping cross sections for Z < 106. The delayed neutron spectra are taken from an evaluated library of 105 precursors. The code provides the magnitude and spectra, if desired, of the resultant neutron source in addition to an analysis of the'contributions by each nuclide in the problem. LASTCALL, a graphical user interface, is included in the code package.
Chang H. Oh; Eung S. Kim; Mike Patterson
2010-06-01
Abstract – A tritium permeation analyses code (TPAC) was developed by Idaho National Laboratory for the purpose of analyzing tritium distributions in very high temperature reactor (VHTR) systems, including integrated hydrogen production systems. A MATLAB SIMULINK software package was used in developing the code. The TPAC is based on the mass balance equations of tritium-containing species and various forms of hydrogen coupled with a variety of tritium sources, sinks, and permeation models. In the TPAC, ternary fission and neutron reactions with 6Li, 7Li 10B, and 3He were taken into considerations as tritium sources. Purification and leakage models were implemented as main tritium sinks. Permeation of tritium and H2 through pipes, vessels, and heat exchangers were considered as main tritium transport paths. In addition, electroyzer and isotope exchange models were developed for analyzing hydrogen production systems, including high temperature electrolysis and sulfur-iodine processes.
Coupling the beam tracing code TORBEAM and the Fokker-Planck solver RELAX for fast electrons
NASA Astrophysics Data System (ADS)
Maj, O.; Poli, E.; Westerhof, E.
2012-12-01
In this paper the interface between the beam tracing code TORBEAM [Poli, Peeters and Pereverzev, Comp. Phys. Comm. 136, 90 (2001)] and the quasi-linear Fokker-Planck solver RELAX [Westerhof, Peeters and Schippers, Rijnhuizen Report No. RR 92-211 CA, 1992] is presented together with preliminary testing results for electron cyclotron waves in ITER plasmas and their effects on the electron distribution function. The resulting numerical package allows us to account for diffraction effects in the construction of the quasi-linear wave-particle diffusion operator. The coupling of the paraxial-WKB code TORBEAM to the ray-based code RELAX requires a reinterpretation of the paraxial wave field in terms of extended rays, which are addressed in details.
Use of SUSA in Uncertainty and Sensitivity Analysis for INL VHTR Coupled Codes
Gerhard Strydom
2010-06-01
The need for a defendable and systematic Uncertainty and Sensitivity approach that conforms to the Code Scaling, Applicability, and Uncertainty (CSAU) process, and that could be used for a wide variety of software codes, was defined in 2008.The GRS (Gesellschaft für Anlagen und Reaktorsicherheit) company of Germany has developed one type of CSAU approach that is particularly well suited for legacy coupled core analysis codes, and a trial version of their commercial software product SUSA (Software for Uncertainty and Sensitivity Analyses) was acquired on May 12, 2010. This interim milestone report provides an overview of the current status of the implementation and testing of SUSA at the INL VHTR Project Office.
Large Wind Turbine Rotor Design using an Aero-Elastic / Free-Wake Panel Coupling Code
NASA Astrophysics Data System (ADS)
Sessarego, Matias; Ramos-García, Néstor; Shen, Wen Zhong; Nørkær Sørensen, Jens
2016-09-01
Despite the advances in computing resources in the recent years, the majority of large wind-turbine rotor design problems still rely on aero-elastic codes that use blade element momentum (BEM) approaches to model the rotor aerodynamics. The present work describes an approach to wind-turbine rotor design by incorporating a higher-fidelity free-wake panel aero-elastic coupling code called MIRAS-FLEX. The optimization procedure includes a series of design load cases and a simple structural design code. Due to the heavy MIRAS-FLEX computations, a surrogate-modeling approach is applied to mitigate the overall computational cost of the optimization. Improvements in cost of energy, annual energy production, maximum flap-wise root bending moment, and blade mass were obtained for the NREL 5MW baseline wind turbine.
Samet Y. Kadioglu; Dana A. Knoll; Cassiano de Oliveira
2009-05-01
Coupling neutronics to thermomechanics is important for the analysis of fast burst reactors, because the criticality and safety study of fast burst reactors heavily depends on the thermomechanical behavior of fuel materials. For instance, the shut down mechanism or the transition between super and sub-critical states are driven by the fuel material expansion or contraction. The material expansion or contraction is due to temperature gradient which results from fission power. In this paper, we introduce a numerical model for coupling of neutron diffusion and thermomechanics in fast burst reactors. We also provide some analysis of the coupled system. We studied material behaviors corresponding to different levels of power pulses.
Characterisation of the TRIUMF neutron facility using a Monte Carlo simulation code.
Monk, S D; Abram, T; Joyce, M J
2015-04-01
Here, the characterisation of the high-energy neutron field at TRIUMF (The Tri Universities Meson Facility, Vancouver, British Columbia) with Monte Carlo simulation software is described. The package used is MCNPX version 2.6.0, with the neutron fluence rate determined at three locations within the TRIUMF Thermal Neutron Facility (TNF), including the exit of the neutron channel where users of the facility can test devices that may be susceptible to the effects of this form of radiation. The facility is often used to roughly emulate the field likely to be encountered at high altitudes due to radiation of galactic origin and thus the simulated information is compared with the energy spectrum calculated to be due to neutron radiation of cosmic origin at typical aircraft altitudes. The calculated values were also compared with neutron flux measurements that were estimated using the activation of various foils by the staff of the facility, showing agreement within an order of magnitude. PMID:25342608
Characterisation of the TRIUMF neutron facility using a Monte Carlo simulation code.
Monk, S D; Abram, T; Joyce, M J
2015-04-01
Here, the characterisation of the high-energy neutron field at TRIUMF (The Tri Universities Meson Facility, Vancouver, British Columbia) with Monte Carlo simulation software is described. The package used is MCNPX version 2.6.0, with the neutron fluence rate determined at three locations within the TRIUMF Thermal Neutron Facility (TNF), including the exit of the neutron channel where users of the facility can test devices that may be susceptible to the effects of this form of radiation. The facility is often used to roughly emulate the field likely to be encountered at high altitudes due to radiation of galactic origin and thus the simulated information is compared with the energy spectrum calculated to be due to neutron radiation of cosmic origin at typical aircraft altitudes. The calculated values were also compared with neutron flux measurements that were estimated using the activation of various foils by the staff of the facility, showing agreement within an order of magnitude.
MULTIDIMENSIONAL COUPLED PHOTON-ELECTRON TRANSPORT SIMULATIONS USING NEUTRAL PARTICLE SN CODES
Ilas, Dan; Williams, Mark L; Peplow, Douglas E.; Kirk, Bernadette Lugue
2008-01-01
During the past two years a study was underway at ORNL to assess the suitability of the popular SN neutral particle codes ANISN, DORT and TORT for coupled photon-electron calculations specific to external beam therapy of medical physics applications. The CEPXS-BFP code was used to generate the cross sections. The computational tests were performed on phantoms typical of those used in medical physics for external beam therapy, with materials simulated by water at different densities and the comparisons were made against Monte Carlo simulations that served as benchmarks. Although the results for one-dimensional calculations were encouraging, it appeared that the higher dimensional transport codes had fundamental difficulties in handling the electron transport. The results of two-dimensional simulations using the code DORT with an S16 fully symmetric quadrature set agree fairly with the reference Monte Carlo results but not well enough for clinical applications. While the photon fluxes are in better agreement (generally, within less than 5% from the reference), the discrepancy increases, sometimes very significantly, for the electron fluxes. The paper, however, focuses on the results obtained with the three-dimensional code TORT which had convergence difficulties for the electron groups. Numerical instabilities occurred in these groups. These instabilities were more pronounced with the degree of anisotropy of the problem.
Neutron Star EOS and Symmetry Energy in RMF model with three-body couplings
NASA Astrophysics Data System (ADS)
Tsubakihara, Kohsuke; Ohnishi, Akira; Harada, Toru
2014-09-01
Neutron Star EOS(NS-EOS) is one of most interesting topics not only in astrophysics but also in nuclear physics. Symmetry energy in nuclear system and the emergence of hyperons in dense matter are key ingredients to investigate NS-EOS theoretically. We introduced n = 3 three-body couplings to RMF model and examine how valid they are to give reasonable descriptions of nuclear/hypernuclear properties. We have been able to obtain the quantitatively enough fit of both the bulk properties of finite nuclear systems and consistent symmetry energy with the one deduced from recent observations simultaneously. In this presentation, we present the results of hadronic star matter EOS, M-R relation, possibility of appearance of Σ- in NS-EOS providing we fix isovector-vector couplings by fitting Σ- atomic shift data, and so on.
Madland, D.G.; Arthur, E.D.; Estes, G.P.; Stewart, J.E.; Bozoian, M.; Perry, R.T.; Parish, T.A.; Brown, T.H.; England, T.R.; Wilson, W.B.; Charlton, W.S.
1999-09-01
SOURCES 4A is a computer code that determines neutron production rates and spectra from ({alpha},n) reactions, spontaneous fission, and delayed neutron emission due to the decay of radionuclides. The code is capable of calculating ({alpha},n) source rates and spectra in four types of problems: homogeneous media (i.e., a mixture of {alpha}-emitting source material and low-Z target material), two-region interface problems (i.e., a slab of {alpha}-emitting source material in contact with a slab of low-Z target material), three-region interface problems (i.e., a thin slab of low-Z target material sandwiched between {alpha}-emitting source material and low-Z target material), and ({alpha},n) reactions induced by a monoenergetic beam of {alpha}-particles incident on a slab of target material. Spontaneous fission spectra are calculated with evaluated half-life, spontaneous fission branching, and Watt spectrum parameters for 43 actinides. The ({alpha},n) spectra are calculated using an assumed isotropic angular distribution in the center-of-mass system with a library of 89 nuclide decay {alpha}-particle spectra, 24 sets of measured and/or evaluated ({alpha},n) cross sections and product nuclide level branching fractions, and functional {alpha}-particle stopping cross sections for Z < 106. The delayed neutron spectra are taken from an evaluated library of 105 precursors. The code outputs the magnitude and spectra of the resultant neutron source. It also provides an analysis of the contributions to that source by each nuclide in the problem.
NASA Astrophysics Data System (ADS)
Glushkov, Y. S.; Ponomarov-Stepnoy, N. N.; Kompaniets, G. V.; Gomin, Y. A.; Mayorov, L. V.; Lobyntsev, V. A.; Polyakov, D. N.; Sapir, Joe; Pelowitz, Denise; Streetman, J. Robert
1994-07-01
The TOPAZ-2 reactor system is a heterogeneous epithermal system fueled with highly-enriched fuel based on uranium oxide, cooled by a sodium-potassium liquid metal (NaK), using a zirconium hydride moderator, with 37 thermionic fuel elements (TFEs) built into the core. The core is surrounded by a radial beryllium reflector which contains rotating regulating drums with moderating segments. An important problem is the guaranteeing of nuclear safety upon the accidental falling of the TOPAZ-2 reactor into water, which leads to the growth of the reactivity of the reactor. It has turned out that it is necessary to use the Monte-Carlo method for the conduct of neutronics calculations of such a complex reactor. In the United States (U.S.) and Russia, different codes based on the Monte-Carlo method are used for calculations - the MCNP code in the U.S., and the MCU-2 code in Russia. The goal of this work is the comparison of the codes and neutronics data used in the U.S. and Russia for the basis of the TOPAZ-2 nuclear safety. With this goal, a joint computer model benchmark of the TOPAZ-2 reactor was developed and the calculations of a series of variants, differing by the presence and absence of water in the reactor cavities and behind the radial reflector, in the position of the regulating drums, in the presence of the radial reflector, etc. were done independently by specialists in both the U.S. and Russia. Along with the reactor calculations, calculations were also done of the nuclei of the core using the MCNP code (U.S.) and the MCU-2 code (Russia). The work done allowed one to obtain results comparing the MCNP code to the MCU-2 code which gave somewhat different results both for the absolute values of Keff and for reactivity effects. In the future it remains to conduct a detailed analysis of the reasons for the discrepancies. For this it is necessary to exchange neutronics data used for TOPAZ-2 reactor calculations in the U.S. and Russia.
Dawahra, S; Khattab, K; Saba, G
2015-10-01
A comparative study for fuel conversion from the HEU to LEU in the Miniature Neutron Source Reactor (MNSR) has been performed in this paper using the MCNP4C code. The neutron energy and lethargy flux spectra in the first inner and outer irradiation sites of the MNSR reactor for the existing HEU fuel (UAl4-Al, 90% enriched) and the potential LEU fuels (U3Si2-Al, U3Si-Al, U9Mo-Al, 19.75% enriched and UO2, 12.6% enriched) were investigated using the MCNP4C code. The neutron energy flux spectra for each group was calculated by dividing the neutron flux by the width of each energy group. The neutron flux spectra per unit lethargy was calculated by multiplying the neutron energy flux spectra for each energy group by the average energy of each group. The thermal neutron flux was calculated by summing the neutron fluxes from 0.0 to 0.625 eV, the fast neutron flux was calculated by summing the neutron fluxes from 0.5 MeV to 10 MeV for the existing HEU and potential LEU fuels. Good agreements have been noticed between the flux spectra for the potential LEU fuels and the existing HEU fuels with maximum relative differences less than 10% and 8% in the inner and outer irradiation sites.
Cold Uniform Matter and Neutron Stars in the Quark-Meson-Coupling Model
J.R. Stone; P.A.M. Guichon; H.H. Matevosyan; A.W. Thomas
2007-08-01
A new density dependent effective baryon-baryon interaction has been recently derived from the quark-meson-coupling (QMC) model, offering impressive results in application to finite nuclei and dense baryon matter. This self-consistent, relativistic quark-level approach is used to construct the Equation of State (EoS) and to calculate key properties of high density matter and cold, slowly rotating neutron stars. The results include predictions for the maximum mass of neutron star models, together with the corresponding radius and central density, as well the properties of neutron stars with mass of order 1.4 M{sub {circle_dot}}. The cooling mechanism allowed by the QMC EoS is explored and the parameters relevant to slow rotation, namely the moment of inertia and the period of rotation investigated. The results of the calculation, which are found to be in good agreement with available observational data, are compared with the predictions of more traditional EoS, based on the A18+{delta}v+UIX* and modified Reid soft core potentials, the Skyrme SkM* interaction and two relativistic mean field (RMF) models for a hybrid stars including quark matter. The QMC EoS provides cold neutron star models with maximum mass 1.9-2.1 M{sub {circle_dot}}, with central density less than 6 times nuclear saturation density (n{sub 0} = 0.16 fm{sup -3}) and offers a consistent description of the stellar mass up to this density limit. In contrast with other models, QMC predicts no hyperon contribution at densities lower than 3n{sub 0}, for matter in {beta}-equilibrium. At higher densities, {Xi}{sup -,0} and {Lambda} hyperons are present. The absence of lighter {Sigma}{sup {+-},0} hyperons is understood as a consequence of antisymmetrization, together with the implementation of the color hyperfine interaction in the response of the quark bag to the nuclear scalar field.
Multigroup Time-Independent Neutron Transport Code System for Plane or Spherical Geometry.
1986-12-01
Version 00 PALLAS-PL/SP solves multigroup time-independent one-dimensional neutron transport problems in plane or spherical geometry. The problems solved are subject to a variety of boundary conditions or a distributed source. General anisotropic scattering problems are treated for solving deep-penetration problems in which angle-dependent neutron spectra are calculated in detail.
Code System to Calculate Neutron and Gamma-Ray Skyshine Doses Using the Integral Line-Beam Method.
2000-11-16
Version 03 This package includes the SKYNEUT 1.1, SKYDOSE 2.3, MCSKY 2.3 and SKYCONES 1.1 codes plus the DLC-188/SKYDATA library to form a comprehensive system for calculating skyshine doses. See the author's web site for related information: http://athena.mne.ksu.edu/~jks/ SKYNEUT evaluates the neutron and neutron-induced secondary gamma-ray skyshine doses from an isotropic, point, neutron source collimated by three simple geometries: an open silo, a vertical black (perfectly absorbing) wall, and a rectangular building. The source maymore » emit monoenergetic neutrons or neutrons with an arbitrary multigroup spectrum of energies. SKYDOSE evaluates the gamma-ray skyshine dose from an isotropic, monoenergetic, point gamma-photon source collimated by three simple geometries: (1) a source in a silo, (2) a source behind an infinitely long, vertical, black wall, and (3) a source in a rectangular building. In all three geometries an optional overhead slab shield may be specified. MCSKY evaluates the gamma-ray skyshine dose from an isotropic, monoenergetic, point gamma-photon source collimated into either a vertical cone (i.e., silo geometry) or into a vertically oriented structure with an N-sided polygon cross section. An overhead laminate shield composed of two different materials is assumed, although shield thicknesses of zero may be specified to model an unshielded SKYSHINE source. SKYCONES evaluates the skyshine doses produced by a point neutron or gamma-photon source emitting, into the atmosphere, radiation that is collimated into an upward conical annulus between two arbitrary polar angles. The source is assumed to be axially (azimuthally) symmetric about a vertical axis through the source and can have an arbitrary polyenergetic spectrum. Nested contiguous annular cones can thus be used to represent the energy and polar-angle dependence of a skyshine source emitting radiation into the atmosphere.« less
Integrated TIGER Series of Coupled Electron/Photon Monte Carlo Transport Codes System.
VALDEZ, GREG D.
2012-11-30
Version: 00 Distribution is restricted to US Government Agencies and Their Contractors Only. The Integrated Tiger Series (ITS) is a powerful and user-friendly software package permitting state-of-the-art Monte Carlo solution of linear time-independent coupled electron/photon radiation transport problems, with or without the presence of macroscopic electric and magnetic fields of arbitrary spatial dependence. The goal has been to simultaneously maximize operational simplicity and physical accuracy. Through a set of preprocessor directives, the user selects one of the many ITS codes. The ease with which the makefile system is applied combines with an input scheme based on order-independent descriptive keywords that makes maximum use of defaults and internal error checking to provide experimentalists and theorists alike with a method for the routine but rigorous solution of sophisticated radiation transport problems. Physical rigor is provided by employing accurate cross sections, sampling distributions, and physical models for describing the production and transport of the electron/photon cascade from 1.0 GeV down to 1.0 keV. The availability of source code permits the more sophisticated user to tailor the codes to specific applications and to extend the capabilities of the codes to more complex applications. Version 6, the latest version of ITS, contains (1) improvements to the ITS 5.0 codes, and (2) conversion to Fortran 95. The general user friendliness of the software has been enhanced through memory allocation to reduce the need for users to modify and recompile the code.
Integrated TIGER Series of Coupled Electron/Photon Monte Carlo Transport Codes System.
2012-11-30
Version: 00 Distribution is restricted to US Government Agencies and Their Contractors Only. The Integrated Tiger Series (ITS) is a powerful and user-friendly software package permitting state-of-the-art Monte Carlo solution of linear time-independent coupled electron/photon radiation transport problems, with or without the presence of macroscopic electric and magnetic fields of arbitrary spatial dependence. The goal has been to simultaneously maximize operational simplicity and physical accuracy. Through a set of preprocessor directives, the user selects onemore » of the many ITS codes. The ease with which the makefile system is applied combines with an input scheme based on order-independent descriptive keywords that makes maximum use of defaults and internal error checking to provide experimentalists and theorists alike with a method for the routine but rigorous solution of sophisticated radiation transport problems. Physical rigor is provided by employing accurate cross sections, sampling distributions, and physical models for describing the production and transport of the electron/photon cascade from 1.0 GeV down to 1.0 keV. The availability of source code permits the more sophisticated user to tailor the codes to specific applications and to extend the capabilities of the codes to more complex applications. Version 6, the latest version of ITS, contains (1) improvements to the ITS 5.0 codes, and (2) conversion to Fortran 95. The general user friendliness of the software has been enhanced through memory allocation to reduce the need for users to modify and recompile the code.« less
Radiation Coupling with the FUN3D Unstructured-Grid CFD Code
NASA Technical Reports Server (NTRS)
Wood, William A.
2012-01-01
The HARA radiation code is fully-coupled to the FUN3D unstructured-grid CFD code for the purpose of simulating high-energy hypersonic flows. The radiation energy source terms and surface heat transfer, under the tangent slab approximation, are included within the fluid dynamic ow solver. The Fire II flight test, at the Mach-31 1643-second trajectory point, is used as a demonstration case. Comparisons are made with an existing structured-grid capability, the LAURA/HARA coupling. The radiative surface heat transfer rates from the present approach match the benchmark values within 6%. Although radiation coupling is the focus of the present work, convective surface heat transfer rates are also reported, and are seen to vary depending upon the choice of mesh connectivity and FUN3D ux reconstruction algorithm. On a tetrahedral-element mesh the convective heating matches the benchmark at the stagnation point, but under-predicts by 15% on the Fire II shoulder. Conversely, on a mixed-element mesh the convective heating over-predicts at the stagnation point by 20%, but matches the benchmark away from the stagnation region.
NASA Astrophysics Data System (ADS)
Schimeczek, C.; Engel, D.; Wunner, G.
2014-05-01
Our previously published code for calculating energies and bound-bound transitions of medium-Z elements at neutron star magnetic field strengths [D. Engel, M. Klews, G. Wunner, Comp. Phys. Comm. 180, 3-2-311 (2009)] was based on the adiabatic approximation. It assumes a complete decoupling of the (fast) gyration of the electrons under the action of the magnetic field and the (slow) bound motion along the field under the action of the Coulomb forces. For the single-particle orbitals this implied that each is a product of a Landau state and an (unknown) longitudinal wave function whose B-spline coefficients were determined self-consistently by solving the Hartree-Fock equations for the many-electron problem on a finite-element grid. In the present code we go beyond the adiabatic approximation, by allowing the transverse part of each orbital to be a superposition of Landau states, while assuming that the longitudinal part can be approximated by the same wave function in each Landau level. Inserting this ansatz into the energy variational principle leads to a system of coupled equations in which the B-spline coefficients depend on the weights of the individual Landau states, and vice versa, and which therefore has to be solved in a doubly self-consistent manner. The extended ansatz takes into account the back-reaction of the Coulomb motion of the electrons along the field direction on their motion in the plane perpendicular to the field, an effect which cannot be captured by the adiabatic approximation. The new code allows for the inclusion of up to 8 Landau levels. This reduces the relative error of energy values as compared to the adiabatic approximation results by typically a factor of three (1/3 of the original error) and yields accurate results also in regions of lower neutron star magnetic field strengths where the adiabatic approximation fails. Further improvements in the code are a more sophisticated choice of the initial wave functions, which takes into
NASA Technical Reports Server (NTRS)
Ghorai, S. K.
1983-01-01
The purpose of this project was to use a one-dimensional discrete coordinates transport code called ANISN in order to determine the energy-angle-spatial distribution of neutrons in a 6-feet cube rock box which houses a D-T neutron generator at its center. The project was two-fold. The first phase of the project involved adaptation of the ANISN code written for an IBM 360/75/91 computer to the UNIVAC system at JSC. The second phase of the project was to use the code with proper geometry, source function and rock material composition in order to determine the neutron flux distribution around the rock box when a 14.1 MeV neutron generator placed at its center is activated.
Glushkov, Y.S.; Ponomarev-Stepnoi, N.N.; Kompanietz, G.V.; Gomin, Y.A.; Maiorov, L.V.; Lobynstev, V.A.; Polyakov, D.N.; Sapir, J.; Streetman, J.R.
1993-11-01
Topaz-II is a heterogeneous, epithermal reactor, fueled with highly enriched uranium-dioxide, cooled with NaK, and moderated with zirconium-hydride. The reactor core contains 37 single-cell thermionic fuel elements, and is surrounded by a radial beryllium reflector that contains 12 rotatable control drums with poison segments. For the physics analysis of TOPAZ II it is necessary to use the Monte Carlo method. The United States (US) and Russia used two different Monte Carlo codes, namely MCNP and MCU-2, respectively. The work described in this paper was aimed at comparing the codes and neutronic data used in the US and Russia for verification of Topaz-II nuclear safety. For this purpose, the US and Russia developed a joint benchmark model of the Topaz-II reactor. The American and Russian teams performed independent computations for a series of variants representing potential water immersion accidents. Comparison of the MCNP and MCU-2 codes showed somewhat different results both for the absolute values of k{sub eff} and for reactivity effects. Future calculations will be performed to obtain a detailed understanding of the reasons for such discrepancies. For these analyses it will be necessary for the US and Russian teams to exchange neutronic data on Topaz-II physics calculations.
Pedestal Fueling Simulations with a Coupled Kinetic-kinetic Plasma-neutral Transport Code
D.P. Stotler, C.S. Chang, S.H. Ku, J. Lang and G.Y. Park
2012-08-29
A Monte Carlo neutral transport routine, based on DEGAS2, has been coupled to the guiding center ion-electron-neutral neoclassical PIC code XGC0 to provide a realistic treatment of neutral atoms and molecules in the tokamak edge plasma. The DEGAS2 routine allows detailed atomic physics and plasma-material interaction processes to be incorporated into these simulations. The spatial pro le of the neutral particle source used in the DEGAS2 routine is determined from the uxes of XGC0 ions to the material surfaces. The kinetic-kinetic plasma-neutral transport capability is demonstrated with example pedestal fueling simulations.
Energy Conservation Tests of a Coupled Kinetic-kinetic Plasma-neutral Transport Code
Stotler, D. P.; Chang, C. S.; Ku, S. H.; Lang, J.; Park, G.
2012-08-29
A Monte Carlo neutral transport routine, based on DEGAS2, has been coupled to the guiding center ion-electron-neutral neoclassical PIC code XGC0 to provide a realistic treatment of neutral atoms and molecules in the tokamak edge plasma. The DEGAS2 routine allows detailed atomic physics and plasma-material interaction processes to be incorporated into these simulations. The spatial pro le of the neutral particle source used in the DEGAS2 routine is determined from the uxes of XGC0 ions to the material surfaces. The kinetic-kinetic plasma-neutral transport capability is demonstrated with example pedestal fueling simulations.
NASA Astrophysics Data System (ADS)
Pica, G.; Lovett, B. W.; Bhatt, R. N.; Schenkel, T.; Lyon, S. A.
2016-01-01
A scaled quantum computer with donor spins in silicon would benefit from a viable semiconductor framework and a strong inherent decoupling of the qubits from the noisy environment. Coupling neighboring spins via the natural exchange interaction according to current designs requires gate control structures with extremely small length scales. We present a silicon architecture where bismuth donors with long coherence times are coupled to electrons that can shuttle between adjacent quantum dots, thus relaxing the pitch requirements and allowing space between donors for classical control devices. An adiabatic SWAP operation within each donor/dot pair solves the scalability issues intrinsic to exchange-based two-qubit gates, as it does not rely on subnanometer precision in donor placement and is robust against noise in the control fields. We use this SWAP together with well established global microwave Rabi pulses and parallel electron shuttling to construct a surface code that needs minimal, feasible local control.
Shen, G.; Li, J.; Hillhouse, G.C.; Meng, J.
2005-01-01
We study the sensitivity of the neutron skin thickness S in a {sup 208}Pb nucleus to the addition of nucleon-{sigma}-{rho} coupling corrections to a selection (PK1, NL3, S271, and Z271) of interactions in a relativistic mean field model. The PK1 and NL3 effective interactions lead to a minimum value of S= 0.16 fm in comparison with the original value of S= 0.28 fm. The S271 and Z271 effective interactions yield even smaller values of S= 0.11 fm, which are similar to those for nonrelativistic mean field models. A precise measurement of the neutron radius, and therefore S, in {sup 208}Pb will place an important constraint on both relativistic and nonrelativistic mean field models. We also study the correlation between the radius of a 1.4-solar-mass neutron star and S.
MAC/GMC Code Enhanced for Coupled Electromagnetothermoelastic Analysis of Smart Composites
NASA Technical Reports Server (NTRS)
Bednarcyk, Brett A.; Arnold, Steven M.; Aboudi, Jacob
2002-01-01
Intelligent materials are those that exhibit coupling between their electromagnetic response and their thermomechanical response. This coupling allows smart materials to react mechanically (e.g., an induced displacement) to applied electrical or magnetic fields (for instance). These materials find many important applications in sensors, actuators, and transducers. Recently interest has arisen in the development of smart composites that are formed via the combination of two or more phases, one or more of which is a smart material. To design with and utilize smart composites, designers need theories that predict the coupled smart behavior of these materials from the electromagnetothermoelastic properties of the individual phases. The micromechanics model known as the generalized method of cells (GMC) has recently been extended to provide this important capability. This coupled electromagnetothermoelastic theory has recently been incorporated within NASA Glenn Research Center's Micromechanics Analysis Code with Generalized Method of Cells (MAC/GMC). This software package is user friendly and has many additional features that render it useful as a design and analysis tool for composite materials in general, and with its new capabilities, for smart composites as well.
2000-06-14
Version 00 NORMA-FP is an auxiliary program which can perform a neutronic and thermal-hydraulic subchannel analysis, starting from global core calculations carried out by both PSR-471/NORMA or PSR-492/QUARK codes. Detailed flux and power distributions inside homogenized nodes are computed by a two-stage bivariate interpolation method, upon separation of the axial variable for which an analytical solution is adopted. The actual heterogeneous structure of a node is accounted for by fuel rod power factors computed asmore » functions of burnup, burnup-weighted coolant density, and instantaneous coolant density.« less
NASA Astrophysics Data System (ADS)
Péron, A.; Malouch, F.; Zoia, A.; Diop, C. M.
2014-06-01
Nuclear heating evaluation by Monte-Carlo simulation requires coupled neutron-photon calculation so as to take into account the contribution of secondary photons. Nuclear data are essential for a good calculation of neutron and photon energy deposition and for secondary photon generation. However, a number of isotopes of the most common nuclear data libraries happen to be affected by energy and/or momentum conservation errors concerning the photon production or inaccurate thresholds for photon emission sections. In this paper, we perform a comprehensive survey of the three evaluations JEFF3.1.1, JEFF3.2T2 (beta version) and ENDF/B-VII.1, over 142 isotopes. The aim of this survey is, on the one hand, to check the existence of photon production data by neutron reaction and, on the other hand, to verify the consistency of these data using the kinematic limits method recently implemented in the TRIPOLI-4 Monte-Carlo code, developed by CEA (Saclay center). Then, the impact of these inconsistencies affecting energy deposition scores has been estimated for two materials using a specific nuclear heating calculation scheme in the context of the OSIRIS Material Testing Reactor (CEA/Saclay).
Wulff, W; Cheng, H S; Diamond, D J; Khatib-Rahbar, M
1984-01-01
This report documents the physical models and the numerical methods employed in the BWR systems code RAMONA-3B. The RAMONA-3B code simulates three-dimensional neutron kinetics and multichannel core hydraulics of nonhomogeneous, nonequilibrium two-phase flows. RAMONA-3B is programmed to calculate the steady and transient conditions in the main steam supply system for normal and abnormal operational transients, including the performances of plant control and protection systems. Presented are code capabilities and limitations, models and solution techniques, the results of development code assessment and suggestions for improving the code in the future.
A Coupling Methodology for Mesoscale-informed Nuclear Fuel Performance Codes
Michael Tonks; Derek Gaston; Cody Permann; Paul Millett; Glen Hansen; Dieter Wolf
2010-10-01
This study proposes an approach for capturing the effect of microstructural evolution on reactor fuel performance by coupling a mesoscale irradiated microstructure model with a finite element fuel performance code. To achieve this, the macroscale system is solved in a parallel, fully coupled, fully-implicit manner using the preconditioned Jacobian-free Newton Krylov (JFNK) method. Within the JFNK solution algorithm, microstructure-influenced material parameters are calculated by the mesoscale model and passed back to the macroscale calculation. Due to the stochastic nature of the mesoscale model, a dynamic fitting technique is implemented to smooth roughness in the calculated material parameters. The proposed methodology is demonstrated on a simple model of a reactor fuel pellet. In the model, INL’s BISON fuel performance code calculates the steady-state temperature profile in a fuel pellet and the microstructure-influenced thermal conductivity is determined with a phase field model of irradiated microstructures. This simple multiscale model demonstrates good nonlinear convergence and near ideal parallel scalability. By capturing the formation of large mesoscale voids in the pellet interior, the multiscale model predicted the irradiation-induced reduction in the thermal conductivity commonly observed in reactors.
Physics guide to CEPXS: A multigroup coupled electron-photon cross-section generating code
Lorence, L.J. Jr.; Morel, J.E.; Valdez, G.D.; Los Alamos National Lab., NM; Applied Methods, Inc., Albuquerque, NM )
1989-10-01
CEPXS is a multigroup-Legendre cross-section generating code. The multigroup-Legendre cross sections produced by CEPXS enable coupled electron-photon transport calculations to be performed with the one-dimensional discrete ordinates code, ONEDANT. We recommend that the 1989 version of ONEDANT that contains linear-discontinuous spatial differencing and S2 synthetic acceleration be used for such calculations. CEPXS/ONEDANT effectively solves the Boltzmann-CSD transport equation for electrons and the Boltzmann transport equation for photons over the energy range from 100 MeV to 1.0 keV. The continuous slowing-down approximation is used for those electron interactions that result in small-energy losses. The extended transport correction is applied to the forward-peaked elastic scattering cross section for electrons. A standard multigroup-Legendre treatment is used for the other coupled electron-photon cross sections. CEPXS extracts electron cross-section information from the DATAPAC data set and photon cross-section information from Biggs-Lighthill data. The model that is used for ionization/relaxation in CEPXS is essentially the same as that employed in ITS. 43 refs., 8 figs.
Role of nuclear couplings in the inelastic excitation of weakly-bound neutron-rich nuclei
Dasso, C.H.; Lenzi, S.M.; Vitturi, A.
1996-12-31
Much effort is presently devoted to the study of nuclear systems far from the stability line. Particular emphasis has been placed in light systems such as {sup 11}Li, {sup 8}B and others, where the very small binding energy of the last particles causes their density distribution to extend considerably outside of the remaining nuclear core. Some of the properties associated with this feature are expected to characterize also heavier systems in the vicinity of the proton or neutron drip lines. It is by now well established that low-lying concentrations of multipole strength arise from pure configurations in which a peculiar matching between the wavelength of the continuum wavefunction of the particles and the range of the weakly-bound hole states occurs. To this end the authors consider the break-up of a weakly-bound system in a heavy-ion collision and focus attention in the inelastic excitation of the low-lying part of the continuum. They make use of the fact that previous investigations have shown that the multipole response in this region is not of a collective nature and describe their excited states as pure particle-hole configurations. Since the relevant parameter determining the strength distributions is the binding energy of the last bound orbital they find it most convenient to use single-particle wavefunctions generated by a sperical square-well potential with characteristic nuclear dimensions and whose depth has been adjusted to give rise to a situation in which the last occupied neutron orbital is loosely-bound. Spin-orbit couplings are, for the present purpose, ignored. The results of this investigation clearly indicate that nuclear couplings have the predominant role in causing projectile dissociation in many circumstances, even at bombarding energies remarkably below the Coulomb barrier.
Sato, Tatsuhiko; Endo, Akira; Yamaguchi, Yasuhiro; Takahashi, Fumiaki
2004-01-01
A phoswitch-type detector has been developed for monitoring neutron doses in high-energy accelerator facilities. The detector is composed of a liquid organic scintillator (BC501A) coupled with ZnS(Ag) sheets doped with 6Li. The dose from neutrons with energies above 1 MeV is evaluated from the light output spectrum of the BC501A by applying the G-function, which relates the spectrum to the neutron dose directly. The dose from lower energy neutrons, on the other hand, is estimated from the number of scintillations emitted from the ZnS(Ag) sheets. Characteristics of the phoswitch-type detector were studied experimentally in some neutron fields. It was found from the experiments that the detector has an excellent property of pulse-shape discrimination between the scintillations of BC501A and the ZnS(Ag) sheets. The experimental results also indicate that the detector is capable of reproducing doses from thermal neutrons as well as neutrons with energies from one to several tens of megaelectronvolts (MeV).
Coupling of Sph and Finite Element Codes for Multi-Layer Orbital Debris Shield Design
NASA Technical Reports Server (NTRS)
Fahrenthold, Eric P.
1997-01-01
Particle-based hydrodynamics models offer distinct advantages over Eulerian and Lagrangian hydrocodes in particular shock physics applications. Particle models are designed to avoid the mesh distortion and state variable diffusion problems which can hinder the effective use of Lagrangian and Eulerian codes respectively. However conventional particle-in-cell and smooth particle hydrodynamics methods employ particles which are actually moving interpolation points. A new particle-based modeling methodology, termed Hamiltonian particle hydrodynamics, was developed by Fahrenthold and Koo (1997) to provide an alternative, fully Lagrangian, energy-based approach to shock physics simulations. This alternative formulation avoids the tensile and boundary instabilities associated with standard smooth particle hydrodynamics formulations and the diffusive grid- to-particle mapping schemes characteristic of particle-in-cell methods. In the work described herein, the method of Fahrenthold and Koo has been extended, by coupling the aforementioned hydrodynamic particle model to a hexahedral finite element based description of the continuum dynamics. The resulting continuum model retains all of the features (including general contact-impact effects) of Hamiltonian particle hydrodynamics, while in addition accounting for tensile strength, plasticity, and damage effects important in the simulation of hypervelocity impact on orbital debris shielding. A three dimensional, vectorized, and autotasked implementation of the extended particle method described here has been coded for application to orbital debris shielding design. Source code for the pre-processor (PREP), analysis code (EXOS), post-processor (POST), and rezoner (ZONE), have been delivered separately, along with a User's Guide describing installation and application of the software.
Trading Speed and Accuracy by Coding Time: A Coupled-circuit Cortical Model
Standage, Dominic; You, Hongzhi; Wang, Da-Hui; Dorris, Michael C.
2013-01-01
Our actions take place in space and time, but despite the role of time in decision theory and the growing acknowledgement that the encoding of time is crucial to behaviour, few studies have considered the interactions between neural codes for objects in space and for elapsed time during perceptual decisions. The speed-accuracy trade-off (SAT) provides a window into spatiotemporal interactions. Our hypothesis is that temporal coding determines the rate at which spatial evidence is integrated, controlling the SAT by gain modulation. Here, we propose that local cortical circuits are inherently suited to the relevant spatial and temporal coding. In simulations of an interval estimation task, we use a generic local-circuit model to encode time by ‘climbing’ activity, seen in cortex during tasks with a timing requirement. The model is a network of simulated pyramidal cells and inhibitory interneurons, connected by conductance synapses. A simple learning rule enables the network to quickly produce new interval estimates, which show signature characteristics of estimates by experimental subjects. Analysis of network dynamics formally characterizes this generic, local-circuit timing mechanism. In simulations of a perceptual decision task, we couple two such networks. Network function is determined only by spatial selectivity and NMDA receptor conductance strength; all other parameters are identical. To trade speed and accuracy, the timing network simply learns longer or shorter intervals, driving the rate of downstream decision processing by spatially non-selective input, an established form of gain modulation. Like the timing network's interval estimates, decision times show signature characteristics of those by experimental subjects. Overall, we propose, demonstrate and analyse a generic mechanism for timing, a generic mechanism for modulation of decision processing by temporal codes, and we make predictions for experimental verification. PMID:23592967
1998-05-13
Version 00 The NJOY nuclear data processing system is a modular computer code used for converting evaluated nuclear data in the ENDF format into libraries useful for applications calculations. Because the Evaluated Nuclear Data File (ENDF) format is used all around the world (e.g., ENDF/B-VI in the US, JEF-2.2 in Europe, JENDL-3.2 in Japan, BROND-2.2 in Russia), NJOY gives its users access to a wide variety of the most up-to-date nuclear data. NJOY provides comprehensivemore » capabilities for processing evaluated data, and it can serve applications ranging from continuous-energy Monte Carlo (MCNP), through deterministic transport codes (DANT, ANISN, DORT), to reactor lattice codes (WIMS, EPRI). NJOY handles a wide variety of nuclear effects, including resonances, Doppler broadening, heating (KERMA), radiation damage, thermal scattering (even cold moderators), gas production, neutrons and charged particles, photoatomic interactions, self shielding, probability tables, photon production, and high-energy interactions (to 150 MeV). Output can include printed listings, special library files for applications, and Postscript graphics (plus color). More information on NJOY is available from the developer's home page at http://t2.lanl.gov. Follow the Tourbus section of the Tour area to find notes from the ICTP lectures held at Trieste in March 1998 on the ENDF format and on the NJOY code.« less
Chen, N.C.J.; Ibn-Khayat, M.; March-Leuba, J.A.; Wendel, M.W.
1993-12-01
As part of verification and validation, the Advanced Neutron Source reactor RELAP5 system model was benchmarked by the Advanced Neutron Source dynamic model (ANSDM) and PRSDYN models. RELAP5 is a one-dimensional, two-phase transient code, developed by the Idaho National Engineering Laboratory for reactor safety analysis. Both the ANSDM and PRSDYN models use a simplified single-phase equation set to predict transient thermal-hydraulic performance. Brief descriptions of each of the codes, models, and model limitations were included. Even though comparisons were limited to single-phase conditions, a broad spectrum of accidents was benchmarked: a small loss-of-coolant-accident (LOCA), a large LOCA, a station blackout, and a reactivity insertion accident. The overall conclusion is that the three models yield similar results if the input parameters are the same. However, ANSDM does not capture pressure wave propagation through the coolant system. This difference is significant in very rapid pipe break events. Recommendations are provided for further model improvements.
Cummings, F.M.
1984-06-01
The TEPC analysis code (TACI) is a computer program designed to analyze pulse height data generated by a tissue equivalent proportional counter (TEPC). It is written in HP BASIC and is for use on an HP-87XM personal computer. The theory of TEPC analysis upon which this code is based is summarized.
Ui, Atsushi; Miyaji, Takamasa
2004-10-15
The best-estimate coupled three-dimensional (3-D) core and thermal-hydraulic code system TRAC-BF1/COS3D has been developed. COS3D, based on a modified one-group neutronic model, is a 3-D core simulator used for licensing analyses and core management of commercial boiling water reactor (BWR) plants in Japan. TRAC-BF1 is a plant simulator based on a two-fluid model. TRAC-BF1/COS3D is a coupled system of both codes, which are connected using a parallel computing tool. This code system was applied to the OECD/NRC BWR Turbine Trip Benchmark. Since the two-group cross-section tables are provided by the benchmark team, COS3D was modified to apply to this specification. Three best-estimate scenarios and four hypothetical scenarios were calculated using this code system. In the best-estimate scenario, the predicted core power with TRAC-BF1/COS3D is slightly underestimated compared with the measured data. The reason seems to be a slight difference in the core boundary conditions, that is, pressure changes and the core inlet flow distribution, because the peak in this analysis is sensitive to them. However, the results of this benchmark analysis show that TRAC-BF1/COS3D gives good precision for the prediction of the actual BWR transient behavior on the whole. Furthermore, the results with the modified one-group model and the two-group model were compared to verify the application of the modified one-group model to this benchmark. This comparison shows that the results of the modified one-group model are appropriate and sufficiently precise.
1991-01-25
Version 00 TPHEX calculates the multigroup neutron flux distribution in an assembly of hexagonal cells using a transmission probability (interface current) method. It is primarily intended for calculations on hexagonal LWR fuel assemblies but can be used for other purposes subject to the qualifications mentioned in Restrictions/Limitations.
Multigroup 3-Dimensional Neutron Diffusion Nodal Code System with Thermohydraulic Feedbacks.
1994-02-07
Version 01 GNOMER is a program which solves the multigroup neutron diffusion equation on coarse mesh in 1D, 2D, and 3D Cartesian geometry. The program is designed to calculate the global core power distributions (with thermohydraulic feedbacks) as well as power distributions and homogenized cross sections over a fuel assembly.
Testing universal relations of neutron stars with a nonlinear matter-gravity coupling theory
Sham, Y.-H.; Lin, L.-M.; Leung, P. T. E-mail: lmlin@phy.cuhk.edu.hk
2014-02-01
Due to our ignorance of the equation of state (EOS) beyond nuclear density, there is still no unique theoretical model for neutron stars (NSs). It is therefore surprising that universal EOS-independent relations connecting different physical quantities of NSs can exist. Lau et al. found that the frequency of the f-mode oscillation, the mass, and the moment of inertia are connected by universal relations. More recently, Yagi and Yunes discovered the I-Love-Q universal relations among the mass, the moment of inertia, the Love number, and the quadrupole moment. In this paper, we study these universal relations in the Eddington-inspired Born-Infeld (EiBI) gravity. This theory differs from general relativity (GR) significantly only at high densities due to the nonlinear coupling between matter and gravity. It thus provides us an ideal case to test how robust the universal relations of NSs are with respect to the change of the gravity theory. Due to the apparent EOS formulation of EiBI gravity developed recently by Delsate and Steinhoff, we are able to study the universal relations in EiBI gravity using the same techniques as those in GR. We find that the universal relations in EiBI gravity are essentially the same as those in GR. Our work shows that, within the currently viable coupling constant, there exists at least one modified gravity theory that is indistinguishable from GR in view of the unexpected universal relations.
Wicaksono, D.; Zerkak, O.; Nikitin, K.; Ferroukhi, H.; Chawla, R.
2013-07-01
This paper reports refinement studies on the temporal coupling scheme and time-stepping management of TRACE/S3K, a dynamically coupled code version of the thermal-hydraulics system code TRACE and the 3D core simulator Simulate-3K. The studies were carried out for two test cases, namely a PWR rod ejection accident and the Peach Bottom 2 Turbine Trip Test 2. The solution of the coupled calculation, especially the power peak, proves to be very sensitive to the time-step size with the currently employed conventional operator-splitting. Furthermore, a very small time-step size is necessary to achieve decent accuracy. This degrades the trade-off between accuracy and performance. A simple and computationally cheap implementation of time-projection of power has been shown to be able to improve the convergence of the coupled calculation. This scheme is able to achieve a prescribed accuracy with a larger time-step size. (authors)
Surkov, A. V. Kochkin, V. N.; Pesnya, Yu. E.; Nasonov, V. A.; Vihrov, V. I.; Erak, D. Yu.
2015-12-15
A comparison of measured and calculated neutronic characteristics (fast neutron flux and fission rate of {sup 235}U) in the core and reflector of the IR-8 reactor is presented. The irradiation devices equipped with neutron activation detectors were prepared. The determination of fast neutron flux was performed using the {sup 54}Fe (n, p) and {sup 58}Ni (n, p) reactions. The {sup 235}U fission rate was measured using uranium dioxide with 10% enrichment in {sup 235}U. The determination of specific activities of detectors was carried out by measuring the intensity of characteristic gamma peaks using the ORTEC gamma spectrometer. Neutron fields in the core and reflector of the IR-8 reactor were calculated using the MCU-PTR code.
Warthog: A MOOSE-Based Application for the Direct Code Coupling of BISON and PROTEUS
McCaskey, Alexander J.; Slattery, Stuart; Billings, Jay Jay
2015-09-01
The Nuclear Energy Advanced Modeling and Simulation (NEAMS) program from the Department of Energy s Office of Nuclear Energy provides a robust toolkit for the modeling and simulation of current and future advanced nuclear reactor designs. This toolkit provides these technologies organized across product lines: two divisions targeted at fuels and end-to-end reactor modeling, and a third for integration, coupling, and high-level workflow management. The Fuels Product Line and the Reactor Product line provide advanced computational technologies that serve each respective field well, however, their current lack of integration presents a major impediment to future improvements of simulation solution fidelity. There is a desire for the capability to mix and match tools across Product Lines in an effort to utilize the best from both to improve NEAMS modeling and simulation technologies. This report will detail a new effort to provide this Product Line interoperability through the development of a new application called Warthog. This application couples the BISON Fuel Performance application from the Fuels Product Line and the PROTEUS Core Neutronics application from the Reactors Product Line in an effort to utilize the best from all parts of the NEAMS toolkit and improve overall solution fidelity of nuclear fuel simulations. To acheive this, Warthog leverages as much prior work from the NEAMS program as possible, and in doing so, enables interoperability between the disparate MOOSE and SHARP frameworks, and the libMesh and MOAB mesh data formats. The remainder of this report will describe this work in full. We will begin with a detailed look at the individual NEAMS framework technologies used and developed in the various Product Lines, and the current status of their interoperability. We will then introduce the Warthog application: its overall architecture and the ways it leverages the best existing tools from accross the NEAMS toolkit to enable BISON-PROTEUS integration
Simulations of the Dynamics of the Coupled Energetic and Relativistic Electrons Using VERB Code
NASA Astrophysics Data System (ADS)
Shprits, Y.; Kellerman, A. C.; Drozdov, A.
2015-12-01
Modeling and understanding of ring current and radiation belt coupled system has been a grand challenge since the beginning of the space age. In this study we show long term simulations with a 3D VERB code of modeling the radiation belts with boundary conditions derived from observations around geosynchronous orbit. We also present 4D VERB simulations that include convective transport, radial diffusion, pitch angle scattering and local acceleration. VERB simulations show that the lower energy inward transport is dominated by the convection and higher energy transport is dominated by the diffusive radial transport. We also show that at energies of 100s of keV a number of processes work simultaneously including convective transport, radial diffusion, local acceleration, loss to the loss cone and loss to the magnetopause. The results of the simulaiton of March 2013 storm are compared with Van Allen Probes observations.
Simulation of high-energy radiation belt electron fluxes using NARMAX-VERB coupled codes
Pakhotin, I P; Drozdov, A Y; Shprits, Y Y; Boynton, R J; Subbotin, D A; Balikhin, M A
2014-01-01
This study presents a fusion of data-driven and physics-driven methodologies of energetic electron flux forecasting in the outer radiation belt. Data-driven NARMAX (Nonlinear AutoRegressive Moving Averages with eXogenous inputs) model predictions for geosynchronous orbit fluxes have been used as an outer boundary condition to drive the physics-based Versatile Electron Radiation Belt (VERB) code, to simulate energetic electron fluxes in the outer radiation belt environment. The coupled system has been tested for three extended time periods totalling several weeks of observations. The time periods involved periods of quiet, moderate, and strong geomagnetic activity and captured a range of dynamics typical of the radiation belts. The model has successfully simulated energetic electron fluxes for various magnetospheric conditions. Physical mechanisms that may be responsible for the discrepancies between the model results and observations are discussed. PMID:26167432
NASA Astrophysics Data System (ADS)
Pescarini, Massimo; Orsi, Roberto; Frisoni, Manuela
2016-02-01
The PCA-Replica 12/13 (H2O/Fe) neutron shielding benchmark experiment was analysed using the ORNL TORT-3.2 3D SN code. PCA-Replica, specifically conceived to test the accuracy of nuclear data and transport codes employed in LWR shielding and radiation damage calculations, reproduces a PWR ex-core radial geometry with alternate layers of water and steel including a PWR pressure vessel simulator. Three broad-group coupled neutron/photon working cross section libraries in FIDO-ANISN format with the same energy group structure (47 n + 20 γ) and based on different nuclear data were alternatively used: the ENEA BUGJEFF311.BOLIB (JEFF-3.1.1) and BUGENDF70.BOLIB (ENDF/B-VII.0) libraries and the ORNL BUGLE-96 (ENDF/B-VI.3) library. Dosimeter cross sections derived from the IAEA IRDF-2002 dosimetry file were employed. The calculated reaction rates for the Rh-103(n,n')Rh-103 m, In-115(n,n')In-115m and S-32(n,p)P-32 threshold activation dosimeters and the calculated neutron spectra are compared with the corresponding experimental results.
NASA Astrophysics Data System (ADS)
Tanure, L. P. A. R.; Sousa, R. V.; Costa, D. F.; Cardoso, F.; Veloso, M. A. F.; Pereira, C.
2014-02-01
Neutronic parameters of some fourth generation nuclear reactors have been investigated at the Departamento de Engenharia Nuclear/UFMG. Previous studies show the possibility to increase the transmutation capabilities of these fourth generation systems to achieve significant reduction concerning transuranic elements in spent fuel. To validate the studies, a benchmark on core physics analysis, related to initial testing of the High Temperature Engineering Test Reactor and provided by International Atomic Energy Agency (IAEA) was simulated using the Standardized Computer Analysis for Licensing Evaluation (SCALE). The CSAS6/KENO-VI control sequence and the 44-group ENDF/B-V 0 cross-section neutron library were used to evaluate the keff (effective multiplication factor) and the result presents good agreement with experimental value.
NASA Astrophysics Data System (ADS)
Tani, K.; Shinohara, K.; Oikawa, T.; Tsutsui, H.; McClements, K. G.; Akers, R. J.; Liu, Y. Q.; Suzuki, M.; Ide, S.; Kusama, Y.; Tsuji-Iio, S.
2016-11-01
As part of the verification and validation of a newly developed non-steady-state orbit-following Monte-Carlo code, application studies of time dependent neutron rates have been made for a specific shot in the Mega Amp Spherical Tokamak (MAST) using 3D fields representing vacuum resonant magnetic perturbations (RMPs) and toroidal field (TF) ripples. The time evolution of density, temperature and rotation rate in the application of the code to MAST are taken directly from experiment. The calculation results approximately agree with the experimental data. It is also found that a full orbit-following scheme is essential to reproduce the neutron rates in MAST.
2000-03-28
Version 00 The NJOY nuclear data processing system is a modular computer code used for converting evaluated nuclear data in the ENDF format into libraries useful for applications calculations. Because the Evaluated Nuclear Data File (ENDF) format is used all around the world (e.g., ENDF/B‑VI in the US, JEF‑2.2 in Europe, JENDL‑3.2 in Japan, BROND‑2.2 in Russia), NJOY gives its users access to a wide variety of the most up‑to‑date nuclear data. NJOY provides comprehensivemore » capabilities for processing evaluated data, and it can serve applications ranging from continuous‑energy Monte Carlo (MCNP), through deterministic transport codes (DANT, ANISN, DORT), to reactor lattice codes (WIMS, EPRI). NJOY handles a wide variety of nuclear effects, including resonances, Doppler broadening, heating (KERMA), radiation damage, thermal scattering (even cold moderators), gas production, neutrons and charged particles, photoatomic interactions, self shielding, probability tables, photon production, and high‑energy interactions (to 150 MeV). Output can include printed listings, special library files for applications, and Postscript graphics (plus color).« less
Joseph, Ilon
2014-05-27
Jacobian-free Newton-Krylov (JFNK) algorithms are a potentially powerful class of methods for solving the problem of coupling codes that address dfferent physics models. As communication capability between individual submodules varies, different choices of coupling algorithms are required. The more communication that is available, the more possible it becomes to exploit the simple sparsity pattern of the Jacobian, albeit of a large system. The less communication that is available, the more dense the Jacobian matrices become and new types of preconditioners must be sought to efficiently take large time steps. In general, methods that use constrained or reduced subsystems can offer a compromise in complexity. The specific problem of coupling a fluid plasma code to a kinetic neutrals code is discussed as an example.
1995-06-01
Version 04 The NJOY nuclear data processing system is a comprehensive computer code package for producing pointwise and multigroup neutron and photon cross sections from ENDF/B evaluated nuclear data. This is the last NJOY-91 series. It uses the same module structure as the earlier versions and its graphics options depend on DISSPLA. This new release, designated NJOY91.119, includes bug fixes, improvements in several modules, and some new capabilities. Information on the changes is included inmore » the README file. A new test problem was added to test some ENDF-6 features, including Reich-Moore resonance reconstruction, energy-angle matrices in GROUPR, and energy-angle distributions in ACER. The 91.119 release is basically configured for UNIX.« less
Saadi, Slami; Touiza, Maamar; Kharfi, Fayçal; Guessoum, Abderrezak
2013-12-01
In this work, we present a mixed software/hardware implementation of 2-D signals encoder/decoder using dyadic discrete wavelet transform (DWT) based on quadrature mirror filters (QMF); using fast wavelet Mallat's algorithm. This work is designed and compiled on the embedded development kit EDK6.3i, and the synthesis software, ISE6.3i, which is available with Xilinx Virtex-IIV2MB1000 FPGA. Huffman coding scheme is used to encode the wavelet coefficients so that they can be transmitted progressively through an Ethernet TCP/IP based connection. The possible reconfiguration can be exploited to attain higher performance. The design will be integrated with the neutron radiography system that is used with the Es-Salem research reactor. PMID:24041807
Saadi, Slami; Touiza, Maamar; Kharfi, Fayçal; Guessoum, Abderrezak
2013-12-01
In this work, we present a mixed software/hardware implementation of 2-D signals encoder/decoder using dyadic discrete wavelet transform (DWT) based on quadrature mirror filters (QMF); using fast wavelet Mallat's algorithm. This work is designed and compiled on the embedded development kit EDK6.3i, and the synthesis software, ISE6.3i, which is available with Xilinx Virtex-IIV2MB1000 FPGA. Huffman coding scheme is used to encode the wavelet coefficients so that they can be transmitted progressively through an Ethernet TCP/IP based connection. The possible reconfiguration can be exploited to attain higher performance. The design will be integrated with the neutron radiography system that is used with the Es-Salem research reactor.
Advanced Neutron Source Dynamic Model (ANSDM) code description and user guide
March-Leuba, J.
1995-08-01
A mathematical model is designed that simulates the dynamic behavior of the Advanced Neutron Source (ANS) reactor. Its main objective is to model important characteristics of the ANS systems as they are being designed, updated, and employed; its primary design goal, to aid in the development of safety and control features. During the simulations the model is also found to aid in making design decisions for thermal-hydraulic systems. Model components, empirical correlations, and model parameters are discussed; sample procedures are also given. Modifications are cited, and significant development and application efforts are noted focusing on examination of instrumentation required during and after accidents to ensure adequate monitoring during transient conditions.
Code System for Two-Dinensional Sn-Neutronics and Fluid Dynamics.
2003-07-28
Version 00 SIMMERII is designed to predict the neutronic and fluid-dynamic behavior of an LMFBR during a hypothetical core-disruptive accident. Cross sections depend on temperature and background cross sections. The structure, liquid, and vapor fields are modeled to predict the fluid-dynamic behavior of the reactor. Each field consists of density components to follow the material motion and energy components to predict the material temperatures. For typical accident calculations, the materials are fertile fuel, fissile fuel,more » stainless steel, sodium, control material, and fission gas. Heat, mass, and momentum transfer among the three fields and their components are calculated.« less
A Two-Dimensional Monte Carlo Code System for Linear Neutron Transport Calculations.
1980-04-24
Version 00 KIM (k-infinite-Monte Carlo) solves the steady-state linear neutron transport equation for a fixed source problem or, by successive fixed-source runs, for the eigenvalue problem, in a two-dimensional infinite thermal reactor lattice using the Monte Carlo method. In addition to the combinatorial description of domains, the program allows complex configurations to be represented by a discrete set of points whereby the calculation speed is greatly improved. Configurations are described as the result of overlaysmore » of elementary figures over a basic domain.« less
Ensemble Activation of G-Protein -Coupled Receptors Revealed by Small-Angle Neutron Scattering
NASA Astrophysics Data System (ADS)
Chu, Xiang-Qiang; Perera, Suchithranga; Shrestha, Utsab; Chawla, Udeep; Struts, Andrey; Qian, Shuo; Brown, Michael
2014-03-01
Rhodopsin is a G-protein -coupled receptor (GPCR) involved in visual light perception and occurs naturally in a membrane lipid environment. Rhodopsin photoactivation yields cis-trans isomerization of retinal giving equilibrium between inactive Meta-I and active Meta-II states. Does photoactivation lead to a single Meta-II conformation, or do substates exist as described by an ensemble-activation mechanism (EAM)? We use small-angle neutron scattering (SANS) to investigate conformational changes in rhodopsin-detergent and rhodopsin-lipid complexes upon photoactivation. Meta-I state is stabilized in CHAPS-solubilized rhodopsin, while Meta-II is trapped in DDM-solubilized rhodopsin. SANS data are acquired from 80% D2O solutions and at contrast-matching points for both DDM and CHAPS samples. Our experiments demonstrate that for detergent-solubilized rhodopsin, SANS with contrast variation can detect structural differences between the rhodopsin dark-state, Meta-I, Meta-II, and ligand-free opsin states. Dark-state rhodopsin has more conformational flexibility in DDM micelles compared to CHAPS, which is consistent with an ensemble of activated Meta-II states. Furthermore, time-resolved SANS enables study of the time-dependent structural transitions between Meta-I and Meta-II, which is crucial to understanding the ensemble-based activation.
Imaging spin structures of exchange-coupled materials with neutron reflectometry
NASA Astrophysics Data System (ADS)
O'Donovan, K. V.; Borchers, J. A.; Majkrzak, C. F.; Fullerton, E. E.; Hellwig, O.
2001-03-01
We used polarized neutron reflectometry to examine the spin structure of an Fe_55Pt_45(200 ÅNi_80Fe_20(500 Åexchange-coupled magnetic film in fields up to 630 mT (near saturation) at room temperature. By measuring both the two spin-flip and the two non-spin-flip cross-sections, we can determine the spin structure as a function of depth throughout the entire sample. Magnetometry of the film indicates spring-magnet behavior.(Physical Review B 62), 11 694 (2000) Our measurements in a field of 16 mT (which ostensibly is in the reversible regime) indicate a twist, which begins in the magnetically soft permalloy, extends throughout a significant fraction of the magnetically hard FePt, contrary to theoretical expectations. At larger fields the permalloy is aligned with the external field and the Bloch wall is localized within the FePt. In remanence we see no twist. Instead, the spins are aligned in one direction, canted with respect to the saturation magnetization, but with disorder at the interface. We will discuss fine details of the spin structure near 16 mT and the evolution of the spin structure from 5 mT through 630 mT and in remanence.
Sato, Tatsuhiko; Endo, Akira; Zankl, Maria; Petoussi-Henss, Nina; Niita, Koji
2009-04-01
The fluence to organ-dose and effective-dose conversion coefficients for neutrons and protons with energies up to 100 GeV was calculated using the PHITS code coupled to male and female adult reference computational phantoms, which are to be released as a common ICRP/ICRU publication. For the calculation, the radiation and tissue weighting factors, w(R) and w(T), respectively, as revised in ICRP Publication 103 were employed. The conversion coefficients for effective dose equivalents derived using the radiation quality factors of both Q(L) and Q(y) relationships were also estimated, utilizing the functions for calculating the probability densities of the absorbed dose in terms of LET (L) and lineal energy (y), respectively, implemented in PHITS. By comparing these data with the corresponding data for the effective dose, we found that the numerical compatibilities of the revised w(R) with the Q(L) and Q(y) relationships are fairly established. The calculated data of these dose conversion coefficients are indispensable for constructing the radiation protection systems based on the new recommendations given in ICRP103 for aircrews and astronauts, as well as for workers in accelerators and nuclear facilities.
Three-Dimensional, Nodal, Neutron Diffusion Criticality Code System in Hex-Z Geometry.
1992-07-27
Version: 00 SIXTUS-3 is a 3D extention of SIXTUS-2 and is based on a response matrix nodal model. The code offers a fast and accurate analysis of critical systems in the regular hex-z geometry with the multigroup cross section representation including arbitrary upscattering.
Blostein, Juan Jerónimo; Estrada, Juan; Tartaglione, Aureliano; Sofo haro, Miguel; Fernández Moroni, Guillermo; Cancelo, Gustavo
2015-01-19
This article describes the design features and the first test measurements obtained during the installation of a novel high resolution 2D neutron detection technique. The technique proposed in this work consists of a boron layer (enriched in ${^{10}}$B) placed on a scientific Charge Coupled Device (CCD). After the nuclear reaction ${^{10}}$B(n,$\\alpha$)${^{7}}$Li, the CCD detects the emitted charge particles thus obtaining information on the neutron absorption position. The above mentioned ionizing particles, with energies in the range 0.5-5.5 MeV, produce a plasma effect in the CCD which is recorded as a circular spot. This characteristic circular shape, as well as the relationship observed between the spot diameter and the charge collected, is used for the event recognition, allowing the discrimination of undesirable gamma events. We present the first results recently obtained with this technique, which has the potential to perform neutron tomography investigations with a spatial resolution better than that previously achieved. Numerical simulations indicate that the spatial resolution of this technique will be about 15 $\\mu$m, and the intrinsic detection efficiency for thermal neutrons will be about 3 %. We compare the proposed technique with other neutron detection techniques and analyze its advantages and disadvantages.
Tandem Mirror Reactor Systems Code (Version I)
Reid, R.L.; Finn, P.A.; Gohar, M.Y.; Barrett, R.J.; Gorker, G.E.; Spampinaton, P.T.; Bulmer, R.H.; Dorn, D.W.; Perkins, L.J.; Ghose, S.
1985-09-01
A computer code was developed to model a Tandem Mirror Reactor. Ths is the first Tandem Mirror Reactor model to couple, in detail, the highly linked physics, magnetics, and neutronic analysis into a single code. This report describes the code architecture, provides a summary description of the modules comprising the code, and includes an example execution of the Tandem Mirror Reactor Systems Code. Results from this code for two sensitivity studies are also included. These studies are: (1) to determine the impact of center cell plasma radius, length, and ion temperature on reactor cost and performance at constant fusion power; and (2) to determine the impact of reactor power level on cost.
Ozaki, Hiromasa; Ebihara, Mitsuru
2007-02-01
Radiochemical neutron activation analysis coupled with the k0-standardization method (k0-RNAA method) was applied to silicate rock samples for the simultaneous determination of trace halogens (Cl, Br and I). Analytical results obtained by the k0-RNAA method for geological standard rocks and meteorite samples agreed with those determined by the conventional comparison method conducted in the same set of experiments, suggesting that the k0-RNAA method is as reliable as the conventional method. Our data for these samples are in good agreement with their literature values except for rare cases. Detection limits calculated under the present experimental condition are sufficiently low for Cl and Br but not for I for typical geologic and meteoritic samples. The k0-RNAA method coupled with longer neutron-irradiation is expected to yield satisfactorily low detection limits for halogens including I in these samples.
NASA Astrophysics Data System (ADS)
Yamamoto, H.; Nakajima, K.; Zhang, K.; Nanai, S.
2015-12-01
Powerful numerical codes that are capable of modeling complex coupled processes of physics and chemistry have been developed for predicting the fate of CO2 in reservoirs as well as its potential impacts on groundwater and subsurface environments. However, they are often computationally demanding for solving highly non-linear models in sufficient spatial and temporal resolutions. Geological heterogeneity and uncertainties further increase the challenges in modeling works. Two-phase flow simulations in heterogeneous media usually require much longer computational time than that in homogeneous media. Uncertainties in reservoir properties may necessitate stochastic simulations with multiple realizations. Recently, massively parallel supercomputers with more than thousands of processors become available in scientific and engineering communities. Such supercomputers may attract attentions from geoscientist and reservoir engineers for solving the large and non-linear models in higher resolutions within a reasonable time. However, for making it a useful tool, it is essential to tackle several practical obstacles to utilize large number of processors effectively for general-purpose reservoir simulators. We have implemented massively-parallel versions of two TOUGH2 family codes (a multi-phase flow simulator TOUGH2 and a chemically reactive transport simulator TOUGHREACT) on two different types (vector- and scalar-type) of supercomputers with a thousand to tens of thousands of processors. After completing implementation and extensive tune-up on the supercomputers, the computational performance was measured for three simulations with multi-million grid models, including a simulation of the dissolution-diffusion-convection process that requires high spatial and temporal resolutions to simulate the growth of small convective fingers of CO2-dissolved water to larger ones in a reservoir scale. The performance measurement confirmed that the both simulators exhibit excellent
NASA Astrophysics Data System (ADS)
Ballarini, F.; Alloni, D.; Battistoni, G.; Cerutti, F.; Ferrari, A.; Gadioli, E.; Garzelli, M. V.; Liotta, M.; Mairani, A.; Ottolenghi, A.; Paretzke, H. G.; Parini, V.; Pelliccioni, M.; Pinsky, L.; Sala, P.; Scannicchio, D.; Trovati, S.; Zankl, M.
2006-05-01
Astronauts' exposure to the various components of the space radiation field is of great concern for long-term missions, especially for those in deep space such as a possible travel to Mars. Simulations based on radiation transport/interaction codes coupled with anthropomorphic model phantoms can be of great help in view of risk evaluation and shielding optimisation, which is therefore a crucial issue. The FLUKA Monte Carlo code can be coupled with two types of anthropomorphic phantom (a mathematical model and a ''voxel'' model) to calculate organ-averaged absorbed dose, dose equivalent and ''biological'' dose under different shielding conditions. Herein the ''biological dose'' is represented by the average number of ''Complex Lesions'' (CLs) per cell in a given organ. CLs are clustered DNA breaks previously calculated by means of event-by-event track structure simulations at the nm level and integrated on-line into FLUKA, which adopts a condensed-history approach; such lesions have been shown to play a fundamental role in chromosome aberration induction, which in turn can be correlated with carcinogenesis. Examples of calculation results will be presented relative to Galactic Cosmic Rays, as well as to the August 1972 Solar Particle Event. The contributions from primary ions and secondary particles will be shown separately, thus allowing quantification of the role played by nuclear reactions occurring in the shield and in the human body itself. As expected, the SPE doses decrease dramatically with increasing the Al shielding thickness; nuclear reaction products, essentially due to target fragmentation, are of minor importance. A 10 g/cm2 Al shelter resulted to be sufficient to respect the 30-day limits for deterministic effects recommended for missions in Low Earth Orbit. In contrast with the results obtained for SPEs, the calculated GCR doses are almost independent of the Al shield thickness, and the GCR doses to internal organs are not significantly lower than
Hartini, Entin Andiwijayakusuma, Dinan
2014-09-30
This research was carried out on the development of code for uncertainty analysis is based on a statistical approach for assessing the uncertainty input parameters. In the butn-up calculation of fuel, uncertainty analysis performed for input parameters fuel density, coolant density and fuel temperature. This calculation is performed during irradiation using Monte Carlo N-Particle Transport. The Uncertainty method based on the probabilities density function. Development code is made in python script to do coupling with MCNPX for criticality and burn-up calculations. Simulation is done by modeling the geometry of PWR terrace, with MCNPX on the power 54 MW with fuel type UO2 pellets. The calculation is done by using the data library continuous energy cross-sections ENDF / B-VI. MCNPX requires nuclear data in ACE format. Development of interfaces for obtaining nuclear data in the form of ACE format of ENDF through special process NJOY calculation to temperature changes in a certain range.
NASA Technical Reports Server (NTRS)
Meyer, H. D.
1993-01-01
The Acoustic Radiation Code (ARC) is a finite element program used on the IBM mainframe to predict far-field acoustic radiation from a turbofan engine inlet. In this report, requirements for developers of internal aerodynamic codes regarding use of their program output an input for the ARC are discussed. More specifically, the particular input needed from the Bolt, Beranek and Newman/Pratt and Whitney (turbofan source noise generation) Code (BBN/PWC) is described. In a separate analysis, a method of coupling the source and radiation models, that recognizes waves crossing the interface in both directions, has been derived. A preliminary version of the coupled code has been developed and used for initial evaluation of coupling issues. Results thus far have shown that reflection from the inlet is sufficient to indicate that full coupling of the source and radiation fields is needed for accurate noise predictions ' Also, for this contract, the ARC has been modified for use on the Sun and Silicon Graphics Iris UNIX workstations. Changes and additions involved in this effort are described in an appendix.
Franke, Brian Claude; Kensek, Ronald Patrick; Laub, Thomas William
2004-06-01
ITS is a powerful and user-friendly software package permitting state of the art Monte Carlo solution of linear time-independent couple electron/photon radiation transport problems, with or without the presence of macroscopic electric and magnetic fields of arbitrary spatial dependence. Our goal has been to simultaneously maximize operational simplicity and physical accuracy. Through a set of preprocessor directives, the user selects one of the many ITS codes. The ease with which the makefile system is applied combines with an input scheme based on order-independent descriptive keywords that makes maximum use of defaults and internal error checking to provide experimentalists and theorists alike with a method for the routine but rigorous solution of sophisticated radiation transport problems. Physical rigor is provided by employing accurate cross sections, sampling distributions, and physical models for describing the production and transport of the electron/photon cascade from 1.0 GeV down to 1.0 keV. The availability of source code permits the more sophisticated user to tailor the codes to specific applications and to extend the capabilities of the codes to more complex applications. Version 5.0, the latest version of ITS, contains (1) improvements to the ITS 3.0 continuous-energy codes, (2)multigroup codes with adjoint transport capabilities, and (3) parallel implementations of all ITS codes. Moreover the general user friendliness of the software has been enhanced through increased internal error checking and improved code portability.
An overview of the ENEA activities in the field of coupled codes NPP simulation
Parisi, C.; Negrenti, E.; Sepielli, M.; Del Nevo, A.
2012-07-01
In the framework of the nuclear research activities in the fields of safety, training and education, ENEA (the Italian National Agency for New Technologies, Energy and the Sustainable Development) is in charge of defining and pursuing all the necessary steps for the development of a NPP engineering simulator at the 'Casaccia' Research Center near Rome. A summary of the activities in the field of the nuclear power plants simulation by coupled codes is here presented with the long term strategy for the engineering simulator development. Specifically, results from the participation in international benchmarking activities like the OECD/NEA 'Kalinin-3' benchmark and the 'AER-DYN-002' benchmark, together with simulations of relevant events like the Fukushima accident, are here reported. The ultimate goal of such activities performed using state-of-the-art technology is the re-establishment of top level competencies in the NPP simulation field in order to facilitate the development of Enhanced Engineering Simulators and to upgrade competencies for supporting national energy strategy decisions, the nuclear national safety authority, and the R and D activities on NPP designs. (authors)
NASA Astrophysics Data System (ADS)
Ghoos, K.; Dekeyser, W.; Samaey, G.; Börner, P.; Baelmans, M.
2016-10-01
The plasma and neutral transport in the plasma edge of a nuclear fusion reactor is usually simulated using coupled finite volume (FV)/Monte Carlo (MC) codes. However, under conditions of future reactors like ITER and DEMO, convergence issues become apparent. This paper examines the convergence behaviour and the numerical error contributions with a simplified FV/MC model for three coupling techniques: Correlated Sampling, Random Noise and Robbins Monro. Also, practical procedures to estimate the errors in complex codes are proposed. Moreover, first results with more complex models show that an order of magnitude speedup can be achieved without any loss in accuracy by making use of averaging in the Random Noise coupling technique.
Gravitational effects on planetary neutron flux spectra
NASA Astrophysics Data System (ADS)
Feldman, W. C.; Drake, D. M.; O'dell, R. D.; Brinkley, F. W.; Anderson, R. C.
1989-01-01
The effects of gravity on the planetary neutron flux spectra for planet Mars, and the lifetime of the neutron, were investigated using a modified one-dimensional diffusion accelerated neutral-particle transport code, coupled with a multigroup cross-section library tailored specifically for Mars. The results showed the presence of a qualitatively new feature in planetary neutron leakage spectra in the form of a component of returning neutrons with kinetic energies less than the gravitational binding energy (0.132 eV for Mars). The net effect is an enhancement in flux at the lowest energies that is largest at and above the outermost layer of planetary matter.
Cramer, S.N.
1984-01-01
The MCNP code is the major Monte Carlo coupled neutron-photon transport research tool at the Los Alamos National Laboratory, and it represents the most extensive Monte Carlo development program in the United States which is available in the public domain. The present code is the direct descendent of the original Monte Carlo work of Fermi, von Neumaum, and Ulam at Los Alamos in the 1940s. Development has continued uninterrupted since that time, and the current version of MCNP (or its predecessors) has always included state-of-the-art methods in the Monte Carlo simulation of radiation transport, basic cross section data, geometry capability, variance reduction, and estimation procedures. The authors of the present code have oriented its development toward general user application. The documentation, though extensive, is presented in a clear and simple manner with many examples, illustrations, and sample problems. In addition to providing the desired results, the output listings give a a wealth of detailed information (some optional) concerning each state of the calculation. The code system is continually updated to take advantage of advances in computer hardware and software, including interactive modes of operation, diagnostic interrupts and restarts, and a variety of graphical and video aids.
Unsteady Analysis of Inlet-Compressor Acoustic Interactions Using Coupled 3-D and 1-D CFD Codes
NASA Technical Reports Server (NTRS)
Suresh, A.; Cole, G. L.
2000-01-01
It is well known that the dynamic response of a mixed compression supersonic inlet is very sensitive to the boundary condition imposed at the subsonic exit (engine face) of the inlet. In previous work, a 3-D computational fluid dynamics (CFD) inlet code (NPARC) was coupled at the engine face to a 3-D turbomachinery code (ADPAC) simulating an isolated rotor and the coupled simulation used to study the unsteady response of the inlet. The main problem with this approach is that the high fidelity turbomachinery simulation becomes prohibitively expensive as more stages are included in the simulation. In this paper, an alternative approach is explored, wherein the inlet code is coupled to a lesser fidelity 1-D transient compressor code (DYNTECC) which simulates the whole compressor. The specific application chosen for this evaluation is the collapsing bump experiment performed at the University of Cincinnati, wherein reflections of a large-amplitude acoustic pulse from a compressor were measured. The metrics for comparison are the pulse strength (time integral of the pulse amplitude) and wave form (shape). When the compressor is modeled by stage characteristics the computed strength is about ten percent greater than that for the experiment, but the wave shapes are in poor agreement. An alternate approach that uses a fixed rise in duct total pressure and temperature (so-called 'lossy' duct) to simulate a compressor gives good pulse shapes but the strength is about 30 percent low.
NASA Astrophysics Data System (ADS)
Schimeczek, C.; Engel, D.; Wunner, G.
2012-07-01
Our previously published code for calculating energies and bound-bound transitions of medium-Z elements at neutron star magnetic field strengths [D. Engel, M. Klews, G. Wunner, Comput. Phys. Comm. 180 (2009) 302-311] was based on the adiabatic approximation. It assumes a complete decoupling of the (fast) gyration of the electrons under the action of the magnetic field and the (slow) bound motion along the field under the action of the Coulomb forces. For the single-particle orbitals this implied that each is a product of a Landau state and an (unknown) longitudinal wave function whose B-spline coefficients were determined self-consistently by solving the Hartree-Fock equations for the many-electron problem on a finite-element grid. In the present code we go beyond the adiabatic approximation, by allowing the transverse part of each orbital to be a superposition of Landau states, while assuming that the longitudinal part can be approximated by the same wave function in each Landau level. Inserting this ansatz into the energy variational principle leads to a system of coupled equations in which the B-spline coefficients depend on the weights of the individual Landau states, and vice versa, and which therefore has to be solved in a doubly self-consistent manner. The extended ansatz takes into account the back-reaction of the Coulomb motion of the electrons along the field direction on their motion in the plane perpendicular to the field, an effect which cannot be captured by the adiabatic approximation. The new code allows for the inclusion of up to 8 Landau levels. This reduces the relative error of energy values as compared to the adiabatic approximation results by typically a factor of three (1/3 of the original error), and yields accurate results also in regions of lower neutron star magnetic field strengths where the adiabatic approximation fails. Further improvements in the code are a more sophisticated choice of the initial wave functions, which takes into
NASA Astrophysics Data System (ADS)
María Gómez Castro, Berta; De Simone, Silvia; Rossi, Riccardo; Larese De Tetto, Antonia; Carrera Ramírez, Jesús
2015-04-01
Coupled thermo-hydro-mechanical modeling is essential for CO2 storage because of (1) large amounts of CO2 will be injected, which will cause large pressure buildups and might compromise the mechanical stability of the caprock seal, (2) the most efficient technique to inject CO2 is the cold injection, which induces thermal stress changes in the reservoir and seal. These stress variations can cause mechanical failure in the caprock and can also trigger induced earthquakes. To properly assess these effects, numerical models that take into account the short and long-term thermo-hydro-mechanical coupling are an important tool. For this purpose, there is a growing need of codes that couple these processes efficiently and accurately. This work involves the development of an open-source, finite element code written in C ++ for correctly modeling the effects of thermo-hydro-mechanical coupling in the field of CO2 storage and in others fields related to these processes (geothermal energy systems, fracking, nuclear waste disposal, etc.), and capable to simulate induced seismicity. In order to be able to simulate earthquakes, a new lower dimensional interface element will be implemented in the code to represent preexisting fractures, where pressure continuity will be imposed across the fractures.
Testing and Modeling of a 3-MW Wind Turbine Using Fully Coupled Simulation Codes (Poster)
LaCava, W.; Guo, Y.; Van Dam, J.; Bergua, R.; Casanovas, C.; Cugat, C.
2012-06-01
This poster describes the NREL/Alstom Wind testing and model verification of the Alstom 3-MW wind turbine located at NREL's National Wind Technology Center. NREL,in collaboration with ALSTOM Wind, is studying a 3-MW wind turbine installed at the National Wind Technology Center(NWTC). The project analyzes the turbine design using a state-of-the-art simulation code validated with detailed test data. This poster describes the testing and the model validation effort, and provides conclusions about the performance of the unique drive train configuration used in this wind turbine. The 3-MW machine has been operating at the NWTC since March 2011, and drive train measurements will be collected through the spring of 2012. The NWTC testing site has particularly turbulent wind patterns that allow for the measurement of large transient loads and the resulting turbine response. This poster describes the 3-MW turbine test project, the instrumentation installed, and the load cases captured. The design of a reliable wind turbine drive train increasingly relies on the use of advanced simulation to predict structural responses in a varying wind field. This poster presents a fully coupled, aero-elastic and dynamic model of the wind turbine. It also shows the methodology used to validate the model, including the use of measured tower modes, model-to-model comparisons of the power curve, and mainshaft bending predictions for various load cases. The drivetrain is designed to only transmit torque to the gearbox, eliminating non-torque moments that are known to cause gear misalignment. Preliminary results show that the drivetrain is able to divert bending loads in extreme loading cases, and that a significantly smaller bending moment is induced on the mainshaft compared to a three-point mounting design.
NASA Astrophysics Data System (ADS)
Dzhalandinov, A.; Tsofin, V.; Kochkin, V.; Panferov, P.; Timofeev, A.; Reshetnikov, A.; Makhotin, D.; Erak, D.; Voloschenko, A.
2016-02-01
Usually the synthesis of two-dimensional and one-dimensional discrete ordinate calculations is used to evaluate neutron fluence on VVER-1000 reactor pressure vessel (RPV) for prognosis of radiation embrittlement. But there are some cases when this approach is not applicable. For example the latest projects of VVER-1000 have upgraded surveillance program. Containers with surveillance specimens are located on the inner surface of RPV with fast neutron flux maximum. Therefore, the synthesis approach is not suitable enough for calculation of local disturbance of neutron field in RPV inner surface behind the surveillance specimens because of their complicated and heterogeneous structure. In some cases the VVER-1000 core loading consists of fuel assemblies with different fuel height and the applicability of synthesis approach is also ambiguous for these fuel cycles. Also, the synthesis approach is not enough correct for the neutron fluence estimation at the RPV area above core top. Because of these reasons only the 3D neutron transport codes seem to be satisfactory for calculation of neutron fluence on the VVER-1000 RPV. The direct 3D calculations are also recommended by modern regulations.
NASA Astrophysics Data System (ADS)
Frisoni, Manuela
2016-03-01
ANITA-2000 is a code package for the activation characterization of materials exposed to neutron irradiation released by ENEA to OECD-NEADB and ORNL-RSICC. The main component of the package is the activation code ANITA-4M that computes the radioactive inventory of a material exposed to neutron irradiation. The code requires the decay data library (file fl1) containing the quantities describing the decay properties of the unstable nuclides and the library (file fl2) containing the gamma ray spectra emitted by the radioactive nuclei. The fl1 and fl2 files of the ANITA-2000 code package, originally based on the evaluated nuclear data library FENDL/D-2.0, were recently updated on the basis of the JEFF-3.1.1 Radioactive Decay Data Library. This paper presents the results of the validation of the new fl1 decay data library through the comparison of the ANITA-4M calculated values with the measured electron and photon decay heats and activities of fusion material samples irradiated at the 14 MeV Frascati Neutron Generator (FNG) of the NEA-Frascati Research Centre. Twelve material samples were considered, namely: Mo, Cu, Hf, Mg, Ni, Cd, Sn, Re, Ti, W, Ag and Al. The ratios between calculated and experimental values (C/E) are shown and discussed in this paper.
Specification of the Advanced Burner Test Reactor Multi-Physics Coupling Demonstration Problem
Shemon, E. R.; Grudzinski, J. J.; Lee, C. H.; Thomas, J. W.; Yu, Y. Q.
2015-12-21
This document specifies the multi-physics nuclear reactor demonstration problem using the SHARP software package developed by NEAMS. The SHARP toolset simulates the key coupled physics phenomena inside a nuclear reactor. The PROTEUS neutronics code models the neutron transport within the system, the Nek5000 computational fluid dynamics code models the fluid flow and heat transfer, and the DIABLO structural mechanics code models structural and mechanical deformation. The three codes are coupled to the MOAB mesh framework which allows feedback from neutronics, fluid mechanics, and mechanical deformation in a compatible format.
NASA Astrophysics Data System (ADS)
Jin, Xin; Nie, Rencan; Zhou, Dongming; Yao, Shaowen; Chen, Yanyan; Yu, Jiefu; Wang, Quan
2016-11-01
A novel method for the calculation of DNA sequence similarity is proposed based on simplified pulse-coupled neural network (S-PCNN) and Huffman coding. In this study, we propose a coding method based on Huffman coding, where the triplet code was used as a code bit to transform DNA sequence into numerical sequence. The proposed method uses the firing characters of S-PCNN neurons in DNA sequence to extract features. Besides, the proposed method can deal with different lengths of DNA sequences. First, according to the characteristics of S-PCNN and the DNA primary sequence, the latter is encoded using Huffman coding method, and then using the former, the oscillation time sequence (OTS) of the encoded DNA sequence is extracted. Simultaneously, relevant features are obtained, and finally the similarities or dissimilarities of the DNA sequences are determined by Euclidean distance. In order to verify the accuracy of this method, different data sets were used for testing. The experimental results show that the proposed method is effective.
SUBBAIAH, K. V.
2001-10-01
Version 01 GUI2QAD is an aid in preparation of input for the included QAD-CGPIC program, which is based on CCC-493/QAD-CGGP and PICTURE. QAD-CGPIC is a Fortran code for fast neutron and gamma-ray shielding calculations through various shield configurations defined by combinatorial geometry specifications. Provision is available to interactively input the geometry and view the same in three dimensions with arbitrary rotations along x,y,z axis. The salient features of the present package include: a) Handles off centered multiple identical sources b) Axis of cylindrical sources can be parallel to any of the axes. c) Provides plots of buildup factors (ANSI-1990) and material cross sections d) Estimates dose rate for point source-slab shield situations e) Interactive input of CG geometry with 3D view and rotation f) Fission product decay power computation and plots for source term calculations. g) Provision to read and graphical 1y display picture input file.
2001-10-01
Version 01 GUI2QAD is an aid in preparation of input for the included QAD-CGPIC program, which is based on CCC-493/QAD-CGGP and PICTURE. QAD-CGPIC is a Fortran code for fast neutron and gamma-ray shielding calculations through various shield configurations defined by combinatorial geometry specifications. Provision is available to interactively input the geometry and view the same in three dimensions with arbitrary rotations along x,y,z axis. The salient features of the present package include: a) Handles offmore » centered multiple identical sources b) Axis of cylindrical sources can be parallel to any of the axes. c) Provides plots of buildup factors (ANSI-1990) and material cross sections d) Estimates dose rate for point source-slab shield situations e) Interactive input of CG geometry with 3D view and rotation f) Fission product decay power computation and plots for source term calculations. g) Provision to read and graphical 1y display picture input file.« less
NASA Technical Reports Server (NTRS)
Kumar, A.; Graeves, R. A.
1980-01-01
A user's guide for a computer code 'COLTS' (Coupled Laminar and Turbulent Solutions) is provided which calculates the laminar and turbulent hypersonic flows with radiation and coupled ablation injection past a Jovian entry probe. Time-dependent viscous-shock-layer equations are used to describe the flow field. These equations are solved by an explicit, two-step, time-asymptotic finite-difference method. Eddy viscosity in the turbulent flow is approximated by a two-layer model. In all, 19 chemical species are used to describe the injection of carbon-phenolic ablator in the hydrogen-helium gas mixture. The equilibrium composition of the mixture is determined by a free-energy minimization technique. A detailed frequency dependence of the absorption coefficient for various species is considered to obtain the radiative flux. The code is written for a CDC-CYBER-203 computer and is capable of providing solutions for ablated probe shapes also.
Mar, M.H.
1995-07-01
Based on the vulnerability Lethality (V/L) taxonomy developed by the Ballistic Vulnerability Lethality Division (BVLD) of the Survivability Lethality Analysis Directorate (SLAD), a nuclear electromagnetic pulse (EMP) coupling V/L analysis taxonomy has been developed. A nuclear EMP threat to a military system can be divided into two levels: (1) coupling to a system level through a cable, antenna, or aperture; and (2) the component level. This report will focus on the initial condition, which includes threat definition and target description, as well as the mapping process from the initial condition to damaged components state. EMP coupling analysis at a system level is used to accomplish this. This report introduces the nature of EMP threat, interaction between the threat and target, and how the output of EMP coupling analysis at a system level becomes the input to the component level analysis. Many different tools (EMP coupling codes) will be discussed for the mapping process, which correponds to the physics of phenomenology. This EMP coupling V/L taxonomy and the models identified in this report will provide the tools necessary to conduct basic V/L analysis of EMP coupling.
Gagner, Renata; Lafitte, Helene; Dormeau, Pascal; Stoudt, Roger H.
2004-07-01
Anticipated Transients Without Scram (ATWS) accident analyses make part of the Safety Analysis Report of the European Pressurized water Reactor (EPR), covering Risk Reduction Category A (Core Melt Prevention) events. This paper deals with three of the most penalizing RRC-A sequences of ATWS caused by mechanical blockage of the control/shutdown rods, regarding their consequences on the Reactor Coolant System (RCS) and core integrity. A new 3D code internal coupling calculation method has been introduced. (authors)
Fishman, Randy Scott; Campo, Javier; Vos, Thomas E.; Miller, Joel S.
2012-01-01
The diruthenium compound [Ru2(O2CMe)4]3[Cr(CN)6] contains two weakly coupled, ferrimag- netically ordered sublattices occupying the same volume. The magnetic field Hc 800 Oe required to align the two sublattice moments is proportional to the antiferromagnetic dipolar interaction Kc B Hc 5 10 3 meV between sublattices. Powder neutron-diffraction measurements on a deuterated sample reveal that the sublattice moments are restricted by the anisotropy of the diruthenium paddle-wheel complexes to the cubic diagonals. Those measurements also suggest that the quantum corrections to the ground state are significant.
Michael Podowski
2009-11-30
The major research objectives of this project included the formulation of flow and heat transfer modeling framework for the analysis of flow-induced instabilities in advanced light water nuclear reactors such as boiling water reactors. General multifield model of two-phase flow, including the necessary closure laws. Development of neurton kinetics models compatible with the proposed models of heated channel dynamics. Formulation and encoding of complete coupled neutronics/thermal-hydraulics models for the analysis of spatially-dependent local core instabilities. Computer simulations aimed at testing and validating the new models of reactor dynamics.
D. Scott Lucas; D. S. Lucas
2005-09-01
An LDRD (Laboratory Directed Research and Development) project is underway at the Idaho National Laboratory (INL) to apply the three-dimensional multi-group deterministic neutron transport code (Attila®) to criticality, flux and depletion calculations of the Advanced Test Reactor (ATR). This paper discusses the development of Attila models for ATR, capabilities of Attila, the generation and use of different cross-section libraries, and comparisons to ATR data, MCNP, MCNPX and future applications.
Neutron production using a pyroelectric driven target coupled with a gated field ionization source
Ellsworth, J. L.; Tang, V.; Falabella, S.; Naranjo, B.; Putterman, S.
2013-04-19
A palm sized, portable neutron source would be useful for widespread implementation of detection systems for shielded, special nuclear material. We present progress towards the development of the components for an ultracompact neutron generator using a pulsed, meso-scale field ionization source, a deuterated (or tritiated) titanium target driven by a negative high voltage lithium tantalate crystal. Neutron production from integrated tests using an ion source with a single, biased tungsten tip and a 3 Multiplication-Sign 1 cm, vacuum insulated crystal with a plastic deuterated target are presented. Component testing of the ion source with a single tip produces up to 3 nA of current. Dielectric insulation of the lithium tantalate crystals appears to reduce flashover, which should improve the robustness. The field emission losses from a 3 cm diameter crystal with a plastic target and 6 cm diameter crystal with a metal target are compared.
Monte Carlo calculation for the development of a BNCT neutron source (1eV-10KeV) using MCNP code.
El Moussaoui, F; El Bardouni, T; Azahra, M; Kamili, A; Boukhal, H
2008-09-01
Different materials have been studied in order to produce the epithermal neutron beam between 1eV and 10KeV, which are extensively used to irradiate patients with brain tumors such as GBM. For this purpose, we have studied three different neutrons moderators (H(2)O, D(2)O and BeO) and their combinations, four reflectors (Al(2)O(3), C, Bi, and Pb) and two filters (Cd and Bi). Results of calculation showed that the best obtained assembly configuration corresponds to the combination of the three moderators H(2)O, BeO and D(2)O jointly to Al(2)O(3) reflector and two filter Cd+Bi optimize the spectrum of the epithermal neutron at 72%, and minimize the thermal neutron to 4% and thus it can be used to treat the deep tumor brain. The calculations have been performed by means of the Monte Carlo N (particle code MCNP 5C). Our results strongly encourage further studying of irradiation of the head with epithermal neutron fields.
Vasconcellos, C. A. Zen
2015-12-17
Nuclear science has developed many excellent theoretical models for many-body systems in the domain of the baryon-meson strong interaction for the nucleus and nuclear matter at low, medium and high densities. However, a full microscopic understanding of nuclear systems in the extreme density domain of compact stars is still lacking. The aim of this contribution is to shed some light on open questions facing the nuclear many-body problem at the very high density domain. Here we focus our attention on the conceptual issue of naturalness and its role in shaping the baryon-meson phase space dynamics in the description of the equation of state (EoS) of nuclear matter and neutrons stars. In particular, in order to stimulate possible new directions of research, we discuss relevant aspects of a recently developed relativistic effective theory for nuclear matter within Quantum Hadrodynamics (QHD) with genuine many-body forces and derivative natural parametric couplings. Among other topics we discuss in this work the connection of this theory with other known effective QHD models of the literature and its potentiality in describing a new physics for dense matter. The model with parameterized couplings exhausts the whole fundamental baryon octet (n, p, Σ{sup −}, Σ{sup 0}, Σ{sup +}, Λ, Ξ{sup −}, Ξ{sup 0}) and simulates n-order corrections to the minimal Yukawa baryon couplings by considering nonlinear self-couplings of meson fields and meson-meson interaction terms coupled to the baryon fields involving scalar-isoscalar (σ, σ∗), vector-isoscalar (ω, Φ), vector-isovector (ϱ) and scalar-isovector (δ) virtual sectors. Following recent experimental results, we consider in our calculations the extreme case where the Σ{sup −} experiences such a strong repulsion that its influence in the nuclear structure of a neutron star is excluded at all. A few examples of calculations of properties of neutron stars are shown and prospects for the future are discussed.
Kim, E; Endo, A; Yamaguchi, Y; Yoshizawa, M; Nakamura, T
2002-01-01
A dose evaluation method for neutrons in the energy range of a few MeV to 100 MeV has been developed using a spectrum weight function (G-function), which is applied to an organic liquid scintillator of 12.7 cm in diameter and 12.7 cm in length. The G-function that converts the pulse height spectrum of the scintillator into the ambient dose equivalent, H*(10), was calculated by an unfolding method using successive approximation of the response function of the scintillator and the ambient dose equivalent per unit neutron fluence (H*(10) conversion coefficients) of ICRP 74. To verify the response function of the scintillator and the value of H*(10) evaluated by the G-function. pulse height spectra of the scintillator were measured in some different neutron fields, which have continuous energy, monoenergetic and quasi-monoenergetic spectra. Values of H*(10) estimated using the G-function and pulse height spectra of the scintillator were compared with those calculated using neutron energy spectra. These doses agreed with each other. From the results, it was concluded that H*(10) can be evaluated directly from the pulse height spectrum of the scintillator by applying the G-function proposed in this study. PMID:12212900
1988-09-15
Version 00 Group averaged neutron cross sections, energy resonance self shielding factors, elastic transfer elements up to P5 approximation, the inelastic, (n,2n) and (n,3n) transfer elements, fission spectra, etc., for coarse groups (26 groups in the standard case) in the fast neutron energy range are calculated.
Stimulation at Desert Peak -modeling with the coupled THM code FEHM
kelkar, sharad
2013-04-30
Numerical modeling of the 2011 shear stimulation at the Desert Peak well 27-15. This submission contains the FEHM executable code for a 64-bit PC Windows-7 machine, and the input and output files for the results presented in the included paper from ARMA-213 meeting.
Lee, Y K
2005-01-01
TRIPOLI-4.3 Monte Carlo transport code has been used to evaluate the QUADOS (Quality Assurance of Computational Tools for Dosimetry) problem P4, neutron and photon response of an albedo-type thermoluminescence personal dosemeter (TLD) located on an ISO slab phantom. Two enriched 6LiF and two 7LiF TLD chips were used and they were protected, in front or behind, with a boron-loaded dosemeter-holder. Neutron response of the four chips was determined by counting 6Li(n,t)4He events using ENDF/B-VI.4 library and photon response by estimating absorbed dose (MeV g(-1)). Ten neutron energies from thermal to 20 MeV and six photon energies from 33 keV to 1.25 MeV were used to study the energy dependence. The fraction of the neutron and photon response owing to phantom backscatter has also been investigated. Detailed TRIPOLI-4.3 solutions are presented and compared with MCNP-4C calculations. PMID:16381740
ICECO-CEL: a coupled Eulerian-Lagrangian code for analyzing primary system response in fast reactors
Wang, C.Y.
1981-02-01
This report describes a coupled Eulerian-Lagrangian code, ICECO-CEL, for analyzing the response of the primary system during hypothetical core disruptive accidents. The implicit Eulerian method is used to calculate the fluid motion so that large fluid distortion, two-dimensional sliding interface, flow around corners, flow through coolant passageways, and out-flow boundary conditions can be treated. The explicit Lagrangian formulation is employed to compute the response of the containment vessel and other elastic-plastic solids inside the reactor containment. Large displacements, as well as geometrical and material nonlinearities are considered in the analysis. Marker particles are utilized to define the free surface or the material interface and to visualize the fluid motion. The basic equations and numerical techniques used in the Eulerian hydrodynamics and Lagrangian structural dynamics are described. Treatment of the above-core hydrodynamics, sodium spillage, fluid cavitation, free-surface boundary conditions and heat transfer are also presented. Examples are given to illustrate the capabilities of the computer code. Comparisons of the code predictions with available experimental data are also made.
Coupling an ICRF core spectral solver to an edge FEM code
NASA Astrophysics Data System (ADS)
Wright, J. C.; Shiraiwa, S.
2015-12-01
The finite element method (FEM) and the spectral approaches to simulation of ion cyclotron (IC) waves in toroidal plasmas each have strengths and weaknesses. For example, the spectral approach (eg TORIC) has a natural algebraic representation of the parallel wavenumber and hence the wave dispersion but does not easily represent complex geometries outside the last closed flux surface, whereas the FEM approach (eg LHEAF) naturally represents arbitrary geometries but does not easily represent thermal corrections to the plasma dispersion. The two domains: thermal core with flux surfaces and cold edge plasma with open field lines may be combined in such as way that each approach is used where it works naturally. Among the possible ways of doing this, we demonstrate the method of mode matching. This method provides an easy way of combining the two linear systems without significant modifications to the separate codes. We will present proof of principal cases and initial applications to minority heating.
Coupling an ICRF core spectral solver to an edge FEM code
NASA Astrophysics Data System (ADS)
Wright, John; Shirwaiwa, Syunichi; RF SciDAC Team
2015-11-01
The finite element method (FEM) and the spectral approaches to simulation of ion cyclotron (IC) waves in toroidal plasmas each have strengths and weaknesses. For example, the spectral approach (eg TORIC) has a natural algebraic representation of the parallel wavenumber and hence the wave dispersion but does not easily represent complex geometries outside the last closed flux surface, whereas the FEM approach (eg LHEAF) naturally represents arbitrary geometries but does not easily represent thermal corrections to the plasma dispersion. The two domains: thermal core with flux surfaces and cold edge plasma with open field lines may be combined in such as way that each approach is used where it works naturally. Among the possible ways of doing this, we demonstrate the method of mode matching. This method provides an easy way of combining the two linear systems without significant modifications to the separate codes. We will present proof of principal cases and initial applications to minority heating.
ICANT, a code for the self-consistent computation of ICRH antenna coupling
Pecoul, S.; Heuraux, S.; Koch, R.; Leclert, G.
1996-02-01
The code deals with 3D antenna structures (finite length antennae) that are used to launch electromagnetic waves into tokamak plasmas. The antenna radiation problem is solved using a finite boundary element technique combined with a spectral solution of the interior problem. The slab approximation is used, and periodicity in {ital y} and {ital z} directions is introduced to account for toroidal geometry. We present results for various types of antennae radiating in vacuum: antenna with a finite Faraday screen and ideal Faraday screen, antenna with side limiters and phased antenna arrays. The results (radiated power, current profile) obtained are very close to analytical solutions when available. {copyright} {ital 1996 American Institute of Physics.}
Chang Oh
2008-02-01
The very high-temperature gas-cooled reactor (VHTR) is envisioned as a single- or dual-purpose reactor for electricity and hydrogen generation. The concept has average coolant temperatures above 9000C and operational fuel temperatures above 12500C. The concept provides the potential for increased energy conversion efficiency and for high-temperature process heat application in addition to power generation. While all the High Temperature Gas Cooled Reactor (HTGR) concepts have sufficiently high temperature to support process heat applications, such as coal gasification, desalination or cogenerative processes, the VHTR’s higher temperatures allow broader applications, including thermochemical hydrogen production. However, the very high temperatures of this reactor concept can be detrimental to safety if a loss-ofcoolant accident (LOCA) occurs. Following the loss of coolant through the break and coolant depressurization, air will enter the core through the break by molecular diffusion and ultimately by natural convection, leading to oxidation of the in-core graphite structure and fuel. The oxidation will accelerate heatup of the reactor core and the release of a toxic gas, CO, and fission products. Thus, without any effective countermeasures, a pipe break may lead to significant fuel damage and fission product release. Prior to the start of this Korean/United States collaboration, no computer codes were available that had been sufficiently developed and validated to reliably simulate a LOCA in the VHTR. Therefore, we have worked for the past three years on developing and validating advanced computational methods for simulating LOCAs in a VHTR. GAMMA code is being developed to implement turbomachinery models in the power conversion unit (PCU) and ultimately models associated with the hydrogen plant. Some preliminary results will be described in this paper.
Hursin, M.; Shanjie, X.; Burns, A.; Hopkins, J.; Satvat, N.; Gert, G.; Tsoukalas, L. H.; Jevremovic, T.
2006-07-01
Introduction. This paper summarizes salient aspects of the 'virtual' reactor system developed at Purdue Univ. emphasizing efficient neutronic modeling through AGENT (Arbitrary Geometry Neutron Transport) a deterministic neutron transport code. DOE's Big-10 Innovations in Nuclear Infrastructure and Education (INIE) Consortium was launched in 2002 to enhance scholarship activities pertaining to university research and training reactors (URTRs). Existing and next generation URTRs are powerful campus tools for nuclear engineering as well as a number of disciplines that include, but are not limited to, medicine, biology, material science, and food science. Advancing new computational environments for the analysis and configuration of URTRs is an important Big-10 INIE aim. Specifically, Big-10 INIE has pursued development of a 'virtual' reactor, an advanced computational environment to serve as a platform on which to build operations, utilization (research and education), and systemic analysis of URTRs physics. The 'virtual' reactor computational system will integrate computational tools addressing the URTR core and near core physics (transport, dynamics, fuel management and fuel configuration); thermal-hydraulics; beam line, in-core and near-core experiments; instrumentation and controls; confinement/containment and security issues. Such integrated computational environment does not currently exist. The 'virtual' reactor is designed to allow researchers and educators to configure and analyze their systems to optimize experiments, fuel locations for flux shaping, as well as detector selection and configuration. (authors)
Brink, Jeandrew; Teukolsky, Saul A; Wasserman, Ira
2005-03-15
Two mechanisms for nonlinear mode saturation of the r mode in neutron stars have been suggested: the parametric instability mechanism involving a small number of modes and the formation of a nearly continuous Kolmogorov-type cascade. Using a network of oscillators constructed from the eigenmodes of a perfect fluid incompressible star, we investigate the transition between the two regimes numerically. Our network includes the 4995 inertial modes up to n{<=}30 with 146 998 direct couplings to the r mode and 1 306 999 couplings with detuning <0.002 (out of a total of approximately 10{sup 9} possible couplings). The lowest parametric instability thresholds for a range of temperatures are calculated and it is found that the r mode becomes unstable to modes with 13
Deciphering the Code to Aminergic G-Protein Coupled Receptor Drug Design
Tan, Edwin S.; Groban, Eli S.; Jacobson, Matthew P.; Scanlan, Thomas S.
2009-01-01
Summary The trace amine-associated receptor 1 (TAAR1) is a biogenic amine G-protein coupled receptor (GPCR) that is potently activated by 3-iodothyronamine (1, T1AM) in vitro. Compound 1 is an endogenous derivative of the thyroid hormone thyroxine that rapidly induces hypothermia, anergia, and bradycardia when administered to mice. To explore the role of TAAR1 in mediating the effects of 1, we rationally designed and synthesized rat TAAR1 superagonists and lead antagonists using the rotamer toggle switch model of aminergic GPCR activation. The functional activity of a ligand was found to be correlated to the nature of its interactions with the rotamer switch residues. Allowing the rotamer switch residues to toggle to their active conformation was associated with agonism while interfering with this conformational transition resulted in antagonism. These agonist and antagonist design principles provide a conceptual model for understanding the relationship between the molecular structure of a drug and its pharmacological properties. PMID:18420141
Neutron metrology laboratory facility simulation.
Pereira, Mariana; Salgado, Ana P; Filho, Aidano S; Pereira, Walsan W; Patrão, Karla C S; Fonseca, Evaldo S
2014-10-01
The Neutron Low Scattering Laboratory in Brazil has been completely rebuilt. Evaluation of air attenuation parameters and neutron component scattering in the room was done using Monte Carlo simulation code. Neutron fields produced by referenced neutron source were used to calculate neutron scattering and air attenuation.
Neutron metrology laboratory facility simulation.
Pereira, Mariana; Salgado, Ana P; Filho, Aidano S; Pereira, Walsan W; Patrão, Karla C S; Fonseca, Evaldo S
2014-10-01
The Neutron Low Scattering Laboratory in Brazil has been completely rebuilt. Evaluation of air attenuation parameters and neutron component scattering in the room was done using Monte Carlo simulation code. Neutron fields produced by referenced neutron source were used to calculate neutron scattering and air attenuation. PMID:24864318
NASA Astrophysics Data System (ADS)
Tajik, M.; Ghal-Eh, N.
2015-08-01
The NE102 plastic scintillator response to 137Cs gamma rays and NE213 liquid scintillator response to both mono-energetic and 241Am-Be neutrons have been modeled using FLUKA's EVENTBIN and MCNPX's PTRAC cards. The comparison made in different energy regions confirms that the overall difference is less than 6%.
1987-09-30
Version 00 The REFERDOU system can be used to calculate the response function of a NE-213 scintillation detector for energies up to 100 MeV, to interpolate and spread (Gaussian) the response function, and unfold the measured spectrum of neutrons while propagating errors from the response functions to the unfolded spectrum.
Fujiwara, Satoru; Plazanet, Marie; Oda, Toshiro
2013-02-15
Highlights: ► Quasielastic neutron scattering spectra of F-actin and G-actin were measured. ► Analysis of the samples in D{sub 2}O and H{sub 2}O provided the spectra of hydration water. ► The first layer hydration water around F-actin is less mobile than around G-actin. ► This difference in hydration water is in concert with the internal dynamics of actin. ► Water outside the first layer behaves bulk-like but influenced by the first layer. -- Abstract: In order to characterize dynamics of water molecules around F-actin and G-actin, quasielastic neutron scattering experiments were performed on powder samples of F-actin and G-actin, hydrated either with D{sub 2}O or H{sub 2}O, at hydration ratios of 0.4 and 1.0. By combined analysis of the quasielastic neutron scattering spectra, the parameter values characterizing the dynamics of the water molecules in the first hydration layer and those of the water molecules outside of the first layer were obtained. The translational diffusion coefficients (D{sub T}) of the hydration water in the first layer were found to be 1.2 × 10{sup −5} cm{sup 2}/s and 1.7 × 10{sup −5} cm{sup 2}/s for F-actin and G-actin, respectively, while that for bulk water was 2.8 × 10{sup −5} cm{sup 2}/s. The residence times were 6.6 ps and 5.0 ps for F-actin and G-actin, respectively, while that for bulk water was 0.62 ps. These differences between F-actin and G-actin, indicating that the hydration water around G-actin is more mobile than that around F-actin, are in concert with the results of the internal dynamics of F-actin and G-actin, showing that G-actin fluctuates more rapidly than F-actin. This implies that the dynamics of the hydration water is coupled to the internal dynamics of the actin molecules. The D{sub T} values of the water molecules outside of the first hydration layer were found to be similar to that of bulk water though the residence times are strongly affected by the first hydration layer. This supports the
NASA Astrophysics Data System (ADS)
Zhou, Zhenggan; Ma, Baoquan; Jiang, Jingtao; Yu, Guang; Liu, Kui; Zhang, Dongmei; Liu, Weiping
2014-10-01
Air-coupled ultrasonic testing (ACUT) technique has been viewed as a viable solution in defect detection of advanced composites used in aerospace and aviation industries. However, the giant mismatch of acoustic impedance in air-solid interface makes the transmission efficiency of ultrasound low, and leads to poor signal-to-noise (SNR) ratio of received signal. The utilisation of signal-processing techniques in non-destructive testing is highly appreciated. This paper presents a wavelet filtering and phase-coded pulse compression hybrid method to improve the SNR and output power of received signal. The wavelet transform is utilised to filter insignificant components from noisy ultrasonic signal, and pulse compression process is used to improve the power of correlated signal based on cross-correction algorithm. For the purpose of reasonable parameter selection, different families of wavelets (Daubechies, Symlet and Coiflet) and decomposition level in discrete wavelet transform are analysed, different Barker codes (5-13 bits) are also analysed to acquire higher main-to-side lobe ratio. The performance of the hybrid method was verified in a honeycomb composite sample. Experimental results demonstrated that the proposed method is very efficient in improving the SNR and signal strength. The applicability of the proposed method seems to be a very promising tool to evaluate the integrity of high ultrasound attenuation composite materials using the ACUT.
Neutronic/Thermalhydraulic Coupling Technigues for Sodium Cooled Fast Reactor Simulations
Jean Ragusa; Andrew Siegel; Jean-Michel Ruggieri
2010-09-28
The objective of this project was to test new coupling algorithms and enable efficient and scalable multi-physics simulations of advanced nuclear reactors, with considerations regarding the implementation of such algorithms in massively parallel environments. Numerical tests were carried out to verify the proposed approach and the examples included some reactor transients. The project was directly related to the Sodium Fast Reactor program element of the Generation IV Nuclear Energy Systems Initiative and the Advanced Fuel cycle Initiative, and, supported the requirement of high-fidelity simulation as a mean of achieving the goals of the presidential Global Nuclear Energy Partnership (GNEP) vision.
Coupled high fidelity thermal hydraulics and neutronics for reactor safety simulations
Vincent A. Mousseau; Hongbin Zhang; Haihua Zhao
2008-09-01
This work is a continuation of previous work on the importance of accuracy in the simulation of nuclear reactor safety transients. This work is qualitative in nature and future work will be more quantitative. The focus of this work will be on a simplified single phase nuclear reactor primary. The transient of interest investigates the importance of accuracy related to passive (inherent) safety systems. The transient run here will be an Unprotected Loss of Flow (ULOF) transient. Here the coolant pump is turned off and the un’SCRAM’ed reactor transitions from forced to free convection (Natural circulation). Results will be presented that show the difference that the first order in time truncation physics makes on the transient. The purpose of this document is to illuminate a possible problem in traditional reactor simulation approaches. Detailed studies need to be done on each simulation code for each transient analyzed to determine if the first order truncation physics plays an important role.
Wang, R. C.; Xu, Y.; Downar, T.; Hudson, N.
2012-07-01
The Special Power Excursion Reactor Test III (SPERT III) was a series of reactivity insertion experiments conducted in the 1950's. This paper describes the validation of the U.S. NRC Coupled Code system TRITON/PARCS/TRACE to simulate reactivity insertion accidents (RIA) by using several of the SPERT III tests. The work here used the SPERT III E-core configuration tests in which the RIA was initiated by ejecting a control rod. The resulting super-prompt reactivity excursion and negative reactivity feedback produced the familiar bell shaped power increase and decrease. The energy deposition during such a power peak has important safety consequences and provides validation basis for core coupled multi-physics codes. The transients of five separate tests are used to benchmark the PARCS/TRACE coupled code. The models were thoroughly validated using the original experiment documentation. (authors)
A system for environmental model coupling and code reuse: The Great Rivers Project
NASA Astrophysics Data System (ADS)
Eckman, B.; Rice, J.; Treinish, L.; Barford, C.
2008-12-01
As part of the Great Rivers Project, IBM is collaborating with The Nature Conservancy and the Center for Sustainability and the Global Environment (SAGE) at the University of Wisconsin, Madison to build a Modeling Framework and Decision Support System (DSS) designed to help policy makers and a variety of stakeholders (farmers, fish & wildlife managers, hydropower operators, et al.) to assess, come to consensus, and act on land use decisions representing effective compromises between human use and ecosystem preservation/restoration. Initially focused on Brazil's Paraguay-Parana, China's Yangtze, and the Mississippi Basin in the US, the DSS integrates data and models from a wide variety of environmental sectors, including water balance, water quality, carbon balance, crop production, hydropower, and biodiversity. In this presentation we focus on the modeling framework aspect of this project. In our approach to these and other environmental modeling projects, we see a flexible, extensible modeling framework infrastructure for defining and running multi-step analytic simulations as critical. In this framework, we divide monolithic models into atomic components with clearly defined semantics encoded via rich metadata representation. Once models and their semantics and composition rules have been registered with the system by their authors or other experts, non-expert users may construct simulations as workflows of these atomic model components. A model composition engine enforces rules/constraints for composing model components into simulations, to avoid the creation of Frankenmodels, models that execute but produce scientifically invalid results. A common software environment and common representations of data and models are required, as well as an adapter strategy for code written in e.g., Fortran or python, that still enables efficient simulation runs, including parallelization. Since each new simulation, as a new composition of model components, requires calibration
NASA Astrophysics Data System (ADS)
Choudhury, Sayantan; Dasgupta, Arnab
2014-05-01
In this paper we introduce an idea of leptogenesis scenario in higher derivative gravity induced DBI Galileon framework aka Galileogenesis in presence of one-loop R-parity violating couplings in the background of a low energy effective supergravity setup derived from higher dimensional string theory framework. We have studied extensively the detailed feature of reheating constraints and the cosmophenomenological consequences of thermal gravitino dark matter in light of PLANCK and PDG data. Finally, we have also established a direct cosmophenomenological connection among dark matter relic abundance, reheating temperature and tensor-to-scalar ratio in the context of DBI Galileon inflation. Higher order correction terms in the gravity sector are introduced in the effective action as a perturbative correction to the Einstein-Hilbert counterpart coming from the computation of Conformal Field Theory disk amplitude at the two loop level [34-36]. The matter sector encounters the effect of N=1, D=4 supergravity motivated DBI Galileon interaction which is embedded in the D3 brane. Additionally, we have considered the effect of R-parity violating interactions [37-40] in the matter sector which provide a convenient framework for quantifying quark and lepton-flavor violating effects. The low energy UV protective effective action for the proposed cosmophenomenological model is described by [31,32]: S=∫d4x √{-g}[K(Φ,X)-G(Φ,X)□Φ+B1R+(B2RRαβγδ-4B3RRαβ+B4R2)+B5] where the model dependent characteristic functions K(Φ,X) and G(Φ,X) are the implicit functions of Galileon and its kinetic counterpart is X=-1/2 >g∂μΦ∂νΦ. Additionally, Bi∀i are the self-coupling constants of graviton degrees of freedom appearing via dimensional reduction from higher dimensional string theory. Specifically B5 be the effective four dimensional cosmological constant. In general, B2≠B3≠B4 which implies that the quadratic curvature terms originated from two loop correction to the
Callori, S. J. Bertinshaw, J.; Cortie, D. L.; Cai, J. W. Zhu, T.; Le Brun, A. P.; Klose, F.
2014-07-21
We have observed 90° magnetic coupling in a NiFe/FeMn/biased NiFe multilayer system using polarized neutron reflectometry. Magnetometry results show magnetic switching for both the biased and free NiFe layers, the latter of which reverses at low applied fields. As these measurements are only capable of providing information about the total magnetization within a sample, polarized neutron reflectometry was used to investigate the reversal behavior of the NiFe layers individually. Both the non-spin-flip and spin-flip neutron reflectometry signals were tracked around the free NiFe layer hysteresis loop and were used to detail the evolution of the magnetization during reversal. At low magnetic fields near the free NiFe coercive field, a large spin-flip signal was observed, indicating magnetization aligned perpendicular to both the applied field and pinned layer.
Disc-jet coupling in the Terzan 5 neutron star X-ray binary EXO 1745-248
NASA Astrophysics Data System (ADS)
Tetarenko, A. J.; Bahramian, A.; Sivakoff, G. R.; Tremou, E.; Linares, M.; Tudor, V.; Miller-Jones, J. C. A.; Heinke, C. O.; Chomiuk, L.; Strader, J.; Altamirano, D.; Degenaar, N.; Maccarone, T.; Patruno, A.; Sanna, A.; Wijnands, R.
2016-07-01
We present the results of Very Large Array, Australia Telescope Compact Array, and Swift X-ray Telescope observations of the 2015 outburst of the transient neutron star X-ray binary (NSXB), EXO 1745-248, located in the globular cluster Terzan 5. Combining (near-) simultaneous radio and X-ray measurements, we measure a correlation between the radio and X-ray luminosities of L_R∝ L_X^β with β =1.68^{+0.10}_{-0.09}, linking the accretion flow (probed by X-ray luminosity) and the compact jet (probed by radio luminosity). While such a relationship has been studied in multiple black hole X-ray binaries (BHXBs), this work marks only the third NSXB with such a measurement. Constraints on this relationship in NSXBs are strongly needed, as comparing this correlation between different classes of XB systems is key in understanding the properties that affect the jet production process in accreting objects. Our best-fitting disc-jet coupling index for EXO 1745-248 is consistent with the measured correlation in NSXB 4U 1728-34 (β = 1.5 ± 0.2) but inconsistent with the correlation we fit using the most recent measurements from the literature of NSXB Aql X-1 (β =0.76^{+0.14}_{-0.15}). While a similar disc-jet coupling index appears to hold across multiple BHXBs in the hard accretion state, this does not appear to be the case with the three NSXBs measured so far. Additionally, the normalization of the EXO 1745-248 correlation is lower than the other two NSXBs, making it one of the most radio faint XBs ever detected in the hard state. We also report the detection of a type-I X-ray burst during this outburst, where the decay time-scale is consistent with hydrogen burning.
Shrestha, Utsab R.; Perera, Suchithranga M. D. C.; Bhowmik, Debsindhu; Chawla, Udeep; Mamontov, Eugene; Brown, Michael F.; Chu, Xiang -Qiang
2016-09-15
Light activation of the visual G-protein-coupled receptor (GPCR) rhodopsin leads to significant structural fluctuations of the protein embedded within the membrane yielding the activation of cognate G-protein (transducin), which initiates biological signaling. Here, we report a quasi-elastic neutron scattering study of the activation of rhodopsin as a GPCR prototype. Our results reveal a broadly distributed relaxation of hydrogen atom dynamics of rhodopsin on a picosecond–nanosecond time scale, crucial for protein function, as only observed for globular proteins previously. Interestingly, the results suggest significant differences in the intrinsic protein dynamics of the dark-state rhodopsin versus the ligand-free apoprotein, opsin. These differencesmore » can be attributed to the influence of the covalently bound retinal ligand. Moreover, an idea of the generic free-energy landscape is used to explain the GPCR dynamics of ligand-binding and ligand-free protein conformations, which can be further applied to other GPCR systems.« less
March-Leuba, J. ); Rey, J.M. )
1992-01-01
This paper provides a review of the current state of the art on the topic of coupled neutronic-thermohydraulic instabilities in boiling water nuclear reactors (BWRs). The topic of BWR instabilities is of great current relevance since it affects the operation of a large number of commercial nuclear reactors. The recent trends towards introduction of high efficiency fuels that permit reactor operation at higher power densities with increased void reactivity feedback and decreased response times, has resulted in a decrease of the stability margin in the low-flow, high-power region of the operating map. This trend has resulted in a number of unexpected'' instability events. For instance, United States plants have experienced two instability events recently, one of them resulted in an automatic reactor scram; in Spain, two BWR plants have experienced unstable limit cycle oscillations that required operator action to suppress. Similar events have been experienced in other European countries. In recent years, BWR instabilities have been one of the more exciting topics of work in the area of transient thermohydraulics. As a result, significant advances in understanding the physics behind these events have occurred, and a new and improved'' state of the art has emerged recently.
March-Leuba, J.; Rey, J.M.
1992-05-01
This paper provides a review of the current state of the art on the topic of coupled neutronic-thermohydraulic instabilities in boiling water nuclear reactors (BWRs). The topic of BWR instabilities is of great current relevance since it affects the operation of a large number of commercial nuclear reactors. The recent trends towards introduction of high efficiency fuels that permit reactor operation at higher power densities with increased void reactivity feedback and decreased response times, has resulted in a decrease of the stability margin in the low-flow, high-power region of the operating map. This trend has resulted in a number of ``unexpected`` instability events. For instance, United States plants have experienced two instability events recently, one of them resulted in an automatic reactor scram; in Spain, two BWR plants have experienced unstable limit cycle oscillations that required operator action to suppress. Similar events have been experienced in other European countries. In recent years, BWR instabilities have been one of the more exciting topics of work in the area of transient thermohydraulics. As a result, significant advances in understanding the physics behind these events have occurred, and a ``new and improved`` state of the art has emerged recently.
Russell Feder and Mahmoud Z. Yousef
2009-05-29
Neutronics analysis to find nuclear heating rates and personnel dose rates were conducted in support of the integration of diagnostics in to the ITER Upper Port Plugs. Simplified shielding models of the Visible-Infrared diagnostic and of the ECH heating system were incorporated in to the ITER global CAD model. Results for these systems are representative of typical designs with maximum shielding and a small aperture (Vis-IR) and minimal shielding with a large aperture (ECH). The neutronics discrete-ordinates code ATTILA® and SEVERIAN® (the ATTILA parallel processing version) was used. Material properties and the 500 MW D-T volume source were taken from the ITER “Brand Model” MCNP benchmark model. A biased quadrature set equivelant to Sn=32 and a scattering degree of Pn=3 were used along with a 46-neutron and 21-gamma FENDL energy subgrouping. Total nuclear heating (neutron plug gamma heating) in the upper port plugs ranged between 380 and 350 kW for the Vis-IR and ECH cases. The ECH or Large Aperture model exhibited lower total heating but much higher peak volumetric heating on the upper port plug structure. Personnel dose rates are calculated in a three step process involving a neutron-only transport calculation, the generation of activation volume sources at pre-defined time steps and finally gamma transport analyses are run for selected time steps. ANSI-ANS 6.1.1 1977 Flux-to-Dose conversion factors were used. Dose rates were evaluated for 1 full year of 500 MW DT operation which is comprised of 3000 1800-second pulses. After one year the machine is shut down for maintenance and personnel are permitted to access the diagnostic interspace after 2-weeks if dose rates are below 100 μSv/hr. Dose rates in the Visible-IR diagnostic model after one day of shutdown were 130 μSv/hr but fell below the limit to 90 μSv/hr 2-weeks later. The Large Aperture or ECH style shielding model exhibited higher and more persistent dose rates. After 1-day the dose rate was 230
Rahnema, Farzad; Garimeela, Srinivas; Ougouag, Abderrafi; Zhang, Dingkang
2013-11-29
This project will develop a 3D, advanced coarse mesh transport method (COMET-Hex) for steady- state and transient analyses in advanced very high-temperature reactors (VHTRs). The project will lead to a coupled neutronics and thermal hydraulic (T/H) core simulation tool with fuel depletion capability. The computational tool will be developed in hexagonal geometry, based solely on transport theory without (spatial) homogenization in complicated 3D geometries. In addition to the hexagonal geometry extension, collaborators will concurrently develop three additional capabilities to increase the code’s versatility as an advanced and robust core simulator for VHTRs. First, the project team will develop and implement a depletion method within the core simulator. Second, the team will develop an elementary (proof-of-concept) 1D time-dependent transport method for efficient transient analyses. The third capability will be a thermal hydraulic method coupled to the neutronics transport module for VHTRs. Current advancements in reactor core design are pushing VHTRs toward greater core and fuel heterogeneity to pursue higher burn-ups, efficiently transmute used fuel, maximize energy production, and improve plant economics and safety. As a result, an accurate and efficient neutron transport, with capabilities to treat heterogeneous burnable poison effects, is highly desirable for predicting VHTR neutronics performance. This research project’s primary objective is to advance the state of the art for reactor analysis.
Rapidly rotating neutron star progenitors
NASA Astrophysics Data System (ADS)
Postnov, K. A.; Kuranov, A. G.; Kolesnikov, D. A.; Popov, S. B.; Porayko, N. K.
2016-08-01
Rotating proto-neutron stars can be important sources of gravitational waves to be searched for by present-day and future interferometric detectors. It was demonstrated by Imshennik that in extreme cases the rapid rotation of a collapsing stellar core may lead to fission and formation of a binary proto-neutron star which subsequently merges due to gravitational wave emission. In the present paper, we show that such dynamically unstable collapsing stellar cores may be the product of a former merger process of two stellar cores in a common envelope. We applied population synthesis calculations to assess the expected fraction of such rapidly rotating stellar cores which may lead to fission and formation of a pair of proto-neutron stars. We have used the BSE population synthesis code supplemented with a new treatment of stellar core rotation during the evolution via effective core-envelope coupling, characterized by the coupling time, τc. The validity of this approach is checked by direct MESA calculations of the evolution of a rotating 15 M⊙ star. From comparison of the calculated spin distribution of young neutron stars with the observed one, reported by Popov and Turolla, we infer the value τc ≃ 5 × 105 years. We show that merging of stellar cores in common envelopes can lead to collapses with dynamically unstable proto-neutron stars, with their formation rate being ˜0.1 - 1% of the total core collapses, depending on the common envelope efficiency.
Spectrometry and dosimetry of fast neutrons using pin diode detectors
NASA Astrophysics Data System (ADS)
Zaki Dizaji, H.; Kakavand, T.; Abbasi Davani, F.
2014-03-01
Elastic scattering of light nuclei, especially hydrogen, is widely used for detection of fast neutrons. Semiconductor devices based on silicon detectors are frequently used for different radiation detections. In this work, a neutron spectrometer consisting of a pin diode coupled with a polyethylene converter and aluminum degrader layers has been developed. Aluminum layers are used as discriminators of different neutron energies for detectors. The response of the converter-degrader-pin diode configuration, the optimum thickness of the converter and the degrader layers have been extracted using MCNP and SRIM simulation codes. The possibility of using this type of detector for fast neutron spectrometry and dosimetry has been investigated. A fairly good agreement was seen between neutron energy spectrum and dose obtained from our configurations and these specifications from an 241Am-Be neutron source.
NASA Astrophysics Data System (ADS)
Yettou, Leila; Belgaid, Mohamed
2015-07-01
In this study, a new version EMPIRE 3.2 code was used in the cross section calculations of (n,p) reactions and in the calculation of proton emission spectra produced by (n,xp) reactions. Exciton model predictions combined with the Kalbach angular distribution systematics were used and some parameters such as those of mean free path, cluster emission in terms of Iwamoto-Harada model, optical model potentials of Morillon for neutrons and protons in the energy range up to 20 MeV, level density for spherical nuclei of Gilbert-Cameron model and width fluctuation correction in terms of compound nucleus have been investigated our calculations. The excitation functions and the proton emission spectra for 32,34S nuclei were calculated, discussed and found in good agreement with available experimental data.
2004-04-21
Version 04 NESTLE solves the few-group neutron diffusion equation utilizing the NEM. The NESTLE code can solve the eigenvalue (criticality), eigenvalue adjoint, external fixed-source steady-state, and external fixed-source or eigenvalue initiated transient problems. The eigenvalue problem allows criticality searches to be completed, and the external fixed-source steady-state problem can search to achieve a specified power level. Transient problems model delayed neutrons via precursor groups. Several core properties can be input as time dependent. Two- ormore » four-energy groups can be utilized, with all energy groups being thermal groups (i.e., upscatter exits) if desired. Core geometries modeled include Cartesian and hexagonal. Three-, two-, and one-dimensional models can be utilized with various symmetries. The thermal conditions predicted by the thermal-hydraulic model of the core are used to correct cross sections for temperature and density effects. Cross sections are parameterized by color, control rod state (i.e., in or out), and burnup, allowing fuel depletion to be modeled. Either a macroscopic or microscopic model may be employed.« less
Technology Transfer Automated Retrieval System (TEKTRAN)
One possible way of integrating subsurface flow and transport processes with (bio)geochemical reactions is to couple by means of an operator-splitting approach two completely separate codes, one for variably-saturated flow and solute transport and one for equilibrium and kinetic biogeochemical react...
Liu, T.; Ding, A.; Ji, W.; Xu, X. G.; Carothers, C. D.; Brown, F. B.
2012-07-01
Monte Carlo (MC) method is able to accurately calculate eigenvalues in reactor analysis. Its lengthy computation time can be reduced by general-purpose computing on Graphics Processing Units (GPU), one of the latest parallel computing techniques under development. The method of porting a regular transport code to GPU is usually very straightforward due to the 'embarrassingly parallel' nature of MC code. However, the situation becomes different for eigenvalue calculation in that it will be performed on a generation-by-generation basis and the thread coordination should be explicitly taken care of. This paper presents our effort to develop such a GPU-based MC code in Compute Unified Device Architecture (CUDA) environment. The code is able to perform eigenvalue calculation under simple geometries on a multi-GPU system. The specifics of algorithm design, including thread organization and memory management were described in detail. The original CPU version of the code was tested on an Intel Xeon X5660 2.8 GHz CPU, and the adapted GPU version was tested on NVIDIA Tesla M2090 GPUs. Double-precision floating point format was used throughout the calculation. The result showed that a speedup of 7.0 and 33.3 were obtained for a bare spherical core and a binary slab system respectively. The speedup factor was further increased by a factor of {approx}2 on a dual GPU system. The upper limit of device-level parallelism was analyzed, and a possible method to enhance the thread-level parallelism was proposed. (authors)
Status report on SHARP coupling framework.
Caceres, A.; Tautges, T. J.; Lottes, J.; Fischer, P.; Rabiti, C.; Smith, M. A.; Siegel, A.; Yang, W. S.; Palmiotti, G.
2008-05-30
This report presents the software engineering effort under way at ANL towards a comprehensive integrated computational framework (SHARP) for high fidelity simulations of sodium cooled fast reactors. The primary objective of this framework is to provide accurate and flexible analysis tools to nuclear reactor designers by simulating multiphysics phenomena happening in complex reactor geometries. Ideally, the coupling among different physics modules (such as neutronics, thermal-hydraulics, and structural mechanics) needs to be tight to preserve the accuracy achieved in each module. However, fast reactor cores in steady state mode represent a special case where weak coupling between neutronics and thermal-hydraulics is usually adequate. Our framework design allows for both options. Another requirement for SHARP framework has been to implement various coupling algorithms that are parallel and scalable to large scale since nuclear reactor core simulations are among the most memory and computationally intensive, requiring the use of leadership-class petascale platforms. This report details our progress toward achieving these goals. Specifically, we demonstrate coupling independently developed parallel codes in a manner that does not compromise performance or portability, while minimizing the impact on individual developers. This year, our focus has been on developing a lightweight and loosely coupled framework targeted at UNIC (our neutronics code) and Nek (our thermal hydraulics code). However, the framework design is not limited to just using these two codes.
NASA Astrophysics Data System (ADS)
Pavlou, Andrew Theodore
The Monte Carlo simulation of full-core neutron transport requires high fidelity data to represent not only the various types of possible interactions that can occur, but also the temperature and energy regimes for which these data are relevant. For isothermal conditions, nuclear cross section data are processed in advance of running a simulation. In reality, the temperatures in a neutronics simulation are not fixed, but change with respect to the temperatures computed from an associated heat transfer or thermal hydraulic (TH) code. To account for the temperature change, a code user must either 1) compute new data at the problem temperature inline during the Monte Carlo simulation or 2) pre-compute data at a variety of temperatures over the range of possible values. Inline data processing is computationally inefficient while pre-computing data at many temperatures can be memory expensive. An alternative on-the-fly approach to handle the temperature component of nuclear data is desired. By on-the-fly we mean a procedure that adjusts cross section data to the correct temperature adaptively during the Monte Carlo random walk instead of before the running of a simulation. The on-the-fly procedure should also preserve simulation runtime efficiency. While on-the-fly methods have recently been developed for higher energy regimes, the double differential scattering of thermal neutrons has not been examined in detail until now. In this dissertation, an on-the-fly sampling method is developed by investigating the temperature dependence of the thermal double differential scattering distributions. The temperature dependence is analyzed with a linear least squares regression test to develop fit coefficients that are used to sample thermal scattering data at any temperature. The amount of pre-stored thermal scattering data has been drastically reduced from around 25 megabytes per temperature per nuclide to only a few megabytes per nuclide by eliminating the need to compute data
Conlin, Jeremy Lloyd; Tobin, Stephen J
2010-10-13
There is a great need in the safeguards community to be able to nondestructively quantify the mass of plutonium of a spent nuclear fuel assembly. As part of the Next Generation of Safeguards Initiative, we are investigating several techniques, or detector systems, which, when integrated, will be capable of quantifying the plutonium mass of a spent fuel assembly without dismantling the assembly. This paper reports on the simulation of one of these techniques, the Passive Neutron Albedo Reactivity with Fission Chambers (PNAR-FC) system. The response of this system over a wide range of spent fuel assemblies with different burnup, initial enrichment, and cooling time characteristics is shown. A Monte Carlo method of using these modeled results to estimate the fissile content of a spent fuel assembly has been developed. A few numerical simulations of using this method are shown. Finally, additional developments still needed and being worked on are discussed.
Black Hole - Neutron Star Binary Simulations at Georgia Tech
NASA Astrophysics Data System (ADS)
Haas, Roland
2009-05-01
Mixed compact object binaries consisting of a black hole and a neutron star are expected to be not only one of the primary sources of gravitational radiation to be observed by interferometric detectors but also the central engine of short gamma-ray bursts. We report on the status of our effort at Georgia Tech to model these mixed binary systems using the moving puncture method. The results are obtained with an enhanced version our vacuum MayaKranc code coupled to the hydrodynamics Whisky code. We present preliminary results of gravitational waveforms and the disruption of the neutron star for simple polytropic equations of state.
Frisch, E.; Johnson, C.G.
1962-05-15
A detachable coupling arrangement is described which provides for varying the length of the handle of a tool used in relatively narrow channels. The arrangement consists of mating the key and keyhole formations in the cooperating handle sections. (AEC)
Schütz, C L; Brochhausen, C; Hampel, G; Iffland, D; Kuczewski, B; Otto, G; Schmitz, T; Stieghorst, C; Kratz, J V
2012-10-01
Boron determination in blood and tissue samples is a crucial task especially for treatment planning, preclinical research, and clinical application of boron neutron capture therapy (BNCT). Comparison of clinical findings remains difficult due to a variety of analytical methods, protocols, and standard reference materials in use. This paper addresses the comparability of inductively coupled plasma mass spectrometry, quantitative neutron capture radiography, and prompt gamma activation analysis for the determination of boron in biological samples. It was possible to demonstrate that three different methods relying on three different principles of sample preparation and boron detection can be validated against each other and yield consistent results for both blood and tissue samples. The samples were obtained during a clinical study for the application of BNCT for liver malignancies and therefore represent a realistic situation for boron analysis.
NASA Technical Reports Server (NTRS)
Hanson, Donald B.
1994-01-01
A two dimensional linear aeroacoustic theory for rotor/stator interaction with unsteady coupling was derived and explored in Volume 1 of this report. Computer program CUP2D has been written in FORTRAN embodying the theoretical equations. This volume (Volume 2) describes the structure of the code, installation and running, preparation of the input file, and interpretation of the output. A sample case is provided with printouts of the input and output. The source code is included with comments linking it closely to the theoretical equations in Volume 1.
Advanced neutron absorber materials
Branagan, Daniel J.; Smolik, Galen R.
2000-01-01
A neutron absorbing material and method utilizing rare earth elements such as gadolinium, europium and samarium to form metallic glasses and/or noble base nano/microcrystalline materials, the neutron absorbing material having a combination of superior neutron capture cross sections coupled with enhanced resistance to corrosion, oxidation and leaching.
Simulation of the full-core pin-model by JMCT Monte Carlo neutron-photon transport code
Li, D.; Li, G.; Zhang, B.; Shu, L.; Shangguan, D.; Ma, Y.; Hu, Z.
2013-07-01
Since the large numbers of cells over a million, the tallies over a hundred million and the particle histories over ten billion, the simulation of the full-core pin-by-pin model has become a real challenge for the computers and the computational methods. On the other hand, the basic memory of the model has exceeded the limit of a single CPU, so the spatial domain and data decomposition must be considered. JMCT (J Monte Carlo Transport code) has successful fulfilled the simulation of the full-core pin-by-pin model by the domain decomposition and the nested parallel computation. The k{sub eff} and flux of each cell are obtained. (authors)
Franke, Brian Claude; Kensek, Ronald Patrick; Laub, Thomas William
2005-09-01
ITS is a powerful and user-friendly software package permitting state-of-the-art Monte Carlo solution of linear time-independent coupled electron/photon radiation transport problems, with or without the presence of macroscopic electric and magnetic fields of arbitrary spatial dependence. Our goal has been to simultaneously maximize operational simplicity and physical accuracy. Through a set of preprocessor directives, the user selects one of the many ITS codes. The ease with which the makefile system is applied combines with an input scheme based on order-independent descriptive keywords that makes maximum use of defaults and internal error checking to provide experimentalists and theorists alike with a method for the routine but rigorous solution of sophisticated radiation transport problems. Physical rigor is provided by employing accurate cross sections, sampling distributions, and physical models for describing the production and transport of the electron/photon cascade from 1.0 GeV down to 1.0 keV. The availability of source code permits the more sophisticated user to tailor the codes to specific applications and to extend the capabilities of the codes to more complex applications. Version 5.0, the latest version of ITS, contains (1) improvements to the ITS 3.0 continuous-energy codes, (2) multigroup codes with adjoint transport capabilities, (3) parallel implementations of all ITS codes, (4) a general purpose geometry engine for linking with CAD or other geometry formats, and (5) the Cholla facet geometry library. Moreover, the general user friendliness of the software has been enhanced through increased internal error checking and improved code portability.
Cowan, C.L.; Protsik, R.; Lewellen, J.W.
1984-01-01
The Large Core Code Evaluation Working Group Benchmark Problem Four was specified to provide a stringent test of the current methods which are used in the nuclear design and analyses process. The benchmark specifications provided a base for performing detailed burnup calculations over the first two irradiation cycles for a large heterogeneous fast reactor. Particular emphasis was placed on the techniques for modeling the three-dimensional benchmark geometry, and sensitivity studies were carried out to determine the performance parameter sensitivities to changes in the neutronics and burnup specifications. The results of the Benchmark Four calculations indicated that a linked RZ-XY (Hex) two-dimensional representation of the benchmark model geometry can be used to predict mass balance data, power distributions, regionwise fuel exposure data and burnup reactivities with good accuracy when compared with the results of direct three-dimensional computations. Most of the small differences in the results of the benchmark analyses by the different participants were attributed to ambiguities in carrying out the regionwise flux renormalization calculations throughout the burnup step.
Demaziere, C.; Larsson, V.
2012-07-01
This paper investigates the reliability of different noise estimators aimed at determining the Moderator Temperature Coefficient (MTC) of reactivity in Pressurized Water Reactors. By monitoring the inherent fluctuations in the neutron flux and moderator temperature, an on-line monitoring of the MTC without perturbing reactor operation is possible. In order to get an accurate estimation of the MTC by noise analysis, the point-kinetic component of the neutron noise and the core-averaged moderator temperature noise have to be used. Because of the scarcity of the in-core instrumentation, the determination of these quantities is difficult, and several possibilities thus exist for estimating the MTC by noise analysis. Furthermore, the effect of feedback has to be negligible at the frequency chosen for estimating the MTC in order to get a proper determination of the MTC. By using an integrated neutronic/thermal- hydraulic model specifically developed for estimating the three-dimensional distributions of the fluctuations in neutron flux, moderator properties, and fuel temperature, different approaches for estimating the MTC by noise analysis can be tested individually. It is demonstrated that a reliable MTC estimation can only be provided if the core is equipped with a sufficient number of both neutron detectors and temperature sensors, i.e. if the core contain in-core detectors monitoring both the axial and radial distributions of the fluctuations in neutron flux and moderator temperature. It is further proven that the effect of feedback is negligible for frequencies higher than 0.1 Hz, and thus the MTC noise estimations have to be performed at higher frequencies. (authors)
D. KATONAK; J. BERNARDIN; S. HOPKINS
2001-06-01
The Spallation Neutron Source (SNS) is a facility being designed for scientific and industrial research and development. SNS will generate and use neutrons as a diagnostic tool for medical purposes, material science, etc. The neutrons will be produced by bombarding a heavy metal target with a high-energy beam of protons, generated and accelerated with a linear particle accelerator, or linac. The low energy end of the linac consists of two room temperature copper structures, the drift tube linac (DTL), and the coupled cavity linac (CCL). Both of these accelerating structures use large amounts of electrical energy to accelerate the proton beam. Approximately 60-80% of the electrical energy is dissipated in the copper structure and must be removed. This is done using specifically designed water cooling passages within the linac's copper structure. Cooling water is supplied to these cooling passages by specially designed resonance control and water cooling systems. One of the primary components in the DTL and CCL water cooling systems, is a water purification system that is responsible for minimizing erosion, corrosion, scaling, biological growth, and hardware activation. The water purification system consists of filters, ion exchange resins, carbon beds, an oxygen scavenger, a UV source, and diagnostic instrumentation. This paper reviews related issues associated with water purification and describes the mechanical design of the SNS Linac water purification system.
Matthew Ellis; Derek Gaston; Benoit Forget; Kord Smith
2011-07-01
In recent years the use of Monte Carlo methods for modeling reactors has become feasible due to the increasing availability of massively parallel computer systems. One of the primary challenges yet to be fully resolved, however, is the efficient and accurate inclusion of multiphysics feedback in Monte Carlo simulations. The research in this paper presents a preliminary coupling of the open source Monte Carlo code OpenMC with the open source Multiphysics Object-Oriented Simulation Environment (MOOSE). The coupling of OpenMC and MOOSE will be used to investigate efficient and accurate numerical methods needed to include multiphysics feedback in Monte Carlo codes. An investigation into the sensitivity of Doppler feedback to fuel temperature approximations using a two dimensional 17x17 PWR fuel assembly is presented in this paper. The results show a functioning multiphysics coupling between OpenMC and MOOSE. The coupling utilizes Functional Expansion Tallies to accurately and efficiently transfer pin power distributions tallied in OpenMC to unstructured finite element meshes used in MOOSE. The two dimensional PWR fuel assembly case also demonstrates that for a simplified model the pin-by-pin doppler feedback can be adequately replicated by scaling a representative pin based on pin relative powers.
Weinberg, Nevin N.; Arras, Phil; Burkart, Joshua
2013-06-01
A weakly nonlinear fluid wave propagating within a star can be unstable to three-wave interactions. The resonant parametric instability is a well-known form of three-wave interaction in which a primary wave of frequency ω {sub a} excites a pair of secondary waves of frequency ω {sub b} + ω {sub c} ≅ ω {sub a}. Here we consider a nonresonant form of three-wave interaction in which a low-frequency primary wave excites a high-frequency p-mode and a low-frequency g-mode such that ω {sub b} + ω {sub c} >> ω {sub a}. We show that a p-mode can couple so strongly to a g-mode of similar radial wavelength that this type of nonresonant interaction is unstable even if the primary wave amplitude is small. As an application, we analyze the stability of the tide in coalescing neutron star binaries to p-g mode coupling. We find that the equilibrium tide and dynamical tide are both p-g unstable at gravitational wave frequencies f {sub gw} ≳ 20 Hz and drive short wavelength p-g mode pairs to significant energies on very short timescales (much less than the orbital decay time due to gravitational radiation). Resonant parametric coupling to the tide is, by contrast, either stable or drives modes at a much smaller rate. We do not solve for the saturation of the p-g instability and therefore we cannot say precisely how it influences the evolution of neutron star binaries. However, we show that if even a single daughter mode saturates near its wave breaking amplitude, the p-g instability of the equilibrium tide will (1) induce significant orbital phase errors (Δφ ≳ 1 radian) that accumulate primarily at low frequencies (f {sub gw} ≲ 50 Hz) and (2) heat the neutron star core to a temperature of T ∼ 10{sup 10} K. Since there are at least ∼100 unstable p-g daughter pairs, Δφ and T are potentially much larger than these values. Tides might therefore significantly influence the gravitational wave signal and electromagnetic emission from coalescing neutron star binaries
NASA Astrophysics Data System (ADS)
Weinberg, Nevin N.; Arras, Phil; Burkart, Joshua
2013-06-01
A weakly nonlinear fluid wave propagating within a star can be unstable to three-wave interactions. The resonant parametric instability is a well-known form of three-wave interaction in which a primary wave of frequency ω a excites a pair of secondary waves of frequency ω b + ω c ~= ω a . Here we consider a nonresonant form of three-wave interaction in which a low-frequency primary wave excites a high-frequency p-mode and a low-frequency g-mode such that ω b + ω c Gt ω a . We show that a p-mode can couple so strongly to a g-mode of similar radial wavelength that this type of nonresonant interaction is unstable even if the primary wave amplitude is small. As an application, we analyze the stability of the tide in coalescing neutron star binaries to p-g mode coupling. We find that the equilibrium tide and dynamical tide are both p-g unstable at gravitational wave frequencies f gw >~ 20 Hz and drive short wavelength p-g mode pairs to significant energies on very short timescales (much less than the orbital decay time due to gravitational radiation). Resonant parametric coupling to the tide is, by contrast, either stable or drives modes at a much smaller rate. We do not solve for the saturation of the p-g instability and therefore we cannot say precisely how it influences the evolution of neutron star binaries. However, we show that if even a single daughter mode saturates near its wave breaking amplitude, the p-g instability of the equilibrium tide will (1) induce significant orbital phase errors (Δphi >~ 1 radian) that accumulate primarily at low frequencies (f gw <~ 50 Hz) and (2) heat the neutron star core to a temperature of T ~ 1010 K. Since there are at least ~100 unstable p-g daughter pairs, Δphi and T are potentially much larger than these values. Tides might therefore significantly influence the gravitational wave signal and electromagnetic emission from coalescing neutron star binaries at much larger orbital separations than previously thought.
NASA Astrophysics Data System (ADS)
Noack, Lena; Rivoldini, Attilio; Van Hoolst, Tim
2015-04-01
We present a new numerical code (CHIC) for the simulation of the thermal evolution of terrestrial planets. The code consists of both a 1d parameterised model to evaluate the temperature profile in the planet's interior and a 2d/3d convection model for the silicate mantle - the latter uses either a Cartesian box, a 2d cylindrical sphere or a 2d spherical annulus. The code is modular and can be easily extended (for example to include an atmosphere module). In the convection model next to the energy equation the conservation equations of mass and momentum are solved, as well. We apply either a Boussinesq approximation or an extended Boussinesq approximation for mantle convection; compressible treatment is planned for the future. The code provides information on the temperature field in the mantle, convective velocities and convective stresses. Simulations can be run under steady-state or thermal evolution conditions. The CHIC code handles surface volcanism, crustal development, and different regimes of surface mobilization like plate tectonics. It is therefore well suited for studying scenarios related to the habitability of terrestrial planets. The code provides a user updatable library of thermodynamic properties of iron and common mantle silicates as well as associated equations of state that allow to compute material properties at high pressure and temperature. Furthermore, the interior structure of a planet for given composition and mass can be determined, yielding the core and planet radius that can then be automatically used for the thermal evolution simulation. CHIC does also accommodate a module for computing a simple parameterised thermal evolution model of a planet's core that includes the formation of an inner core. This module can be combined with either the 1d parameterised thermal evolution model or the 2d/3d mantle convection model. The code has been benchmarked with different convection codes, and compared to published interior-structure models and 1d
Smolik, Marek; Polkowska-Motrenko, Halina; Hubicki, Zbigniew; Jakóbik-Kolon, Agata; Danko, Bożena
2014-01-01
Hafnium at the very low level of 1-8 ppm (in relation to zirconium) was determined in zirconium sulfate solutions (originating from investigations of the separation of ca. 44 ppm Hf from zirconium by means of the ion exchange method) by using three independent methods: inductively coupled plasma mass spectrometry (ICP MS), neutron activation analysis (NAA) and inductively coupled plasma atomic emission spectrometry (ICP-AES). The results of NAA and ICP MS determinations were consistent with each other across the entire investigated range (the RSD of both methods did not exceed 38%). The results of ICP-AES determination were more diverse, particularly at less than 5 ppm Hf (RSD was significantly higher: 29-253%). The ion exchange method exploiting Diphonix(®) resin proved sufficient efficiency in Zr-Hf separation when the initial concentration ratio of the elements ([Zr]0/[Hf]0) ranged from 1200 to ca. 143,000.
Hirai, S; Yoshino, Y; Suzuki, S; Horiuchi, N
1982-05-01
Neutron flux and irradiation time should be accurately known in neutron activation analysis using very short lived nuclides in which conventional monitoring methods i.e., a comparator method, flux monitor method and so on cannot be used satisfactorily. Especially, fluctuation of neutron flux has not been corrected. We noted a change of reactor power at a pneumatic operation, and found out a new correction method for its correction in activation analysis. In our small nuclear reactor, TRIGA-II, the reactor power increased rapidly a few % when a pneumatic-operated capsule arrived at a core of the reactor, and decreased when the capsule left from the core. If the duration between these two changes of the reactor power is equal to the irradiation time, and that the reactor power is proportional to the neutron flux, we can regard an activity formation as a time integration of the reactor power. Then, the correction system was made of a reactor power meter, a V-F converter (voltage to frequency converter), a clock time, a counter, a microcomputer, electric circuits and so on. The signal of the reactor power during the irradiation was counted through the V-F converter, and was accumulated in a memory of the microcomputer. The neutron fluence was calculated in this microcomputer. This method was examined by means of activation of copper and selenium standard samples by 9-11 sec irradiations. The observed activity involved +/- 10% error. However, the error in the corrected activity was decreased to a few % using this correction method. As a result, we found that this method can be used to obtain accurate value for radionuclide formation.
An integrated radiation physics computer code system.
NASA Technical Reports Server (NTRS)
Steyn, J. J.; Harris, D. W.
1972-01-01
An integrated computer code system for the semi-automatic and rapid analysis of experimental and analytic problems in gamma photon and fast neutron radiation physics is presented. Such problems as the design of optimum radiation shields and radioisotope power source configurations may be studied. The system codes allow for the unfolding of complex neutron and gamma photon experimental spectra. Monte Carlo and analytic techniques are used for the theoretical prediction of radiation transport. The system includes a multichannel pulse-height analyzer scintillation and semiconductor spectrometer coupled to an on-line digital computer with appropriate peripheral equipment. The system is geometry generalized as well as self-contained with respect to material nuclear cross sections and the determination of the spectrometer response functions. Input data may be either analytic or experimental.
Multigroup neutron dose calculations for proton therapy
Kelsey Iv, Charles T; Prinja, Anil K
2009-01-01
We have developed tools for the preparation of coupled multigroup proton/neutron cross section libraries. Our method is to use NJOY to process evaluated nuclear data files for incident particles below 150 MeV and MCNPX to produce data for higher energies. We modified the XSEX3 program of the MCNPX code system to produce Legendre expansions of scattering matrices generated by sampling the physics models that are comparable to the output of the GROUPR routine of NJOY. Our code combines the low and high energy scattering data with user input stopping powers and energy deposition cross sections that we also calculated using MCNPX. Our code also calculates momentum transfer coefficients for the library and optionally applies an energy straggling model to the scattering cross sections and stopping powers. The motivation was initially for deterministic solution of space radiation shielding calculations using Attila, but noting that proton therapy treatment planning may neglect secondary neutron dose assessments because of difficulty and expense, we have also investigated the feasibility of multi group methods for this application. We have shown that multigroup MCNPX solutions for secondary neutron dose compare well with continuous energy solutions and are obtainable with less than half computational cost. This efficiency comparison neglects the cost of preparing the library data, but this becomes negligible when distributed over many multi group calculations. Our deterministic calculations illustrate recognized obstacles that may have to be overcome before discrete ordinates methods can be efficient alternatives for proton therapy neutron dose calculations.
Gamma scintillator system using boron carbide for neutron detection
NASA Astrophysics Data System (ADS)
Ben-Galim, Y.; Wengrowicz, U.; Raveh, A.; Orion, I.
2014-08-01
A new approach for neutron detection enhancement to scintillator gamma-ray detectors is suggested. By using a scintillator coupled with a boron carbide (B4C) disc, the 478 keV gamma-photon emitted from the excited Li in 94% of the 10B(n,α)7Li interactions was detected. This suggests that the performance of existing gamma detection systems in Homeland security applications can be improved. In this study, a B4C disc (2 in. diameter, 0.125 in. thick) with ~19.8% 10B was used and coupled with a scintillator gamma-ray detector. In addition, the neutron thermalization moderator was studied in order to be able to increase the neutron sensitivity. An improvement in the detector which is easy to assemble, affordable and efficient was demonstrated. Furthermore, a tailored Monte-Carlo code written in MATLAB was developed for validation of the proposed application through efficiency estimation for thermal neutrons. Validation of the code was accomplished by showing that the MATLAB code results were well correlated to a Monte-Carlo MCNP code results. The measured efficiency of the assembled experimental model was observed to be in agreement with both models calculations.
NASA Astrophysics Data System (ADS)
Mattie, P. D.; Knowlton, R. G.; Arnold, B. W.; Tien, N.; Kuo, M.
2006-12-01
Sandia National Laboratories (Sandia), a U.S. Department of Energy National Laboratory, has over 30 years experience in radioactive waste disposal and is providing assistance internationally in a number of areas relevant to the safety assessment of radioactive waste disposal systems. International technology transfer efforts are often hampered by small budgets, time schedule constraints, and a lack of experienced personnel in countries with small radioactive waste disposal programs. In an effort to surmount these difficulties, Sandia has developed a system that utilizes a combination of commercially available codes and existing legacy codes for probabilistic safety assessment modeling that facilitates the technology transfer and maximizes limited available funding. Numerous codes developed and endorsed by the United States Nuclear Regulatory Commission and codes developed and maintained by United States Department of Energy are generally available to foreign countries after addressing import/export control and copyright requirements. From a programmatic view, it is easier to utilize existing codes than to develop new codes. From an economic perspective, it is not possible for most countries with small radioactive waste disposal programs to maintain complex software, which meets the rigors of both domestic regulatory requirements and international peer review. Therefore, re-vitalization of deterministic legacy codes, as well as an adaptation of contemporary deterministic codes, provides a creditable and solid computational platform for constructing probabilistic safety assessment models. External model linkage capabilities in Goldsim and the techniques applied to facilitate this process will be presented using example applications, including Breach, Leach, and Transport-Multiple Species (BLT-MS), a U.S. NRC sponsored code simulating release and transport of contaminants from a subsurface low-level waste disposal facility used in a cooperative technology transfer
NASA Astrophysics Data System (ADS)
Engel, D.; Klews, M.; Wunner, G.
2009-02-01
We have developed a new method for the fast computation of wavelengths and oscillator strengths for medium-Z atoms and ions, up to iron, at neutron star magnetic field strengths. The method is a parallelized Hartree-Fock approach in adiabatic approximation based on finite-element and B-spline techniques. It turns out that typically 15-20 finite elements are sufficient to calculate energies to within a relative accuracy of 10-5 in 4 or 5 iteration steps using B-splines of 6th order, with parallelization speed-ups of 20 on a 26-processor machine. Results have been obtained for the energies of the ground states and excited levels and for the transition strengths of astrophysically relevant atoms and ions in the range Z=2…26 in different ionization stages. Catalogue identifier: AECC_v1_0 Program summary URL:http://cpc.cs.qub.ac.uk/summaries/AECC_v1_0.html Program obtainable from: CPC Program Library, Queen's University, Belfast, N. Ireland Licensing provisions: Standard CPC licence, http://cpc.cs.qub.ac.uk/licence/licence.html No. of lines in distributed program, including test data, etc.: 3845 No. of bytes in distributed program, including test data, etc.: 27 989 Distribution format: tar.gz Programming language: MPI/Fortran 95 and Python Computer: Cluster of 1-26 HP Compaq dc5750 Operating system: Fedora 7 Has the code been vectorised or parallelized?: Yes RAM: 1 GByte Classification: 2.1 External routines: MPI/GFortran, LAPACK, PyLab/Matplotlib Nature of problem: Calculations of synthetic spectra [1] of strongly magnetized neutron stars are bedevilled by the lack of data for atoms in intense magnetic fields. While the behaviour of hydrogen and helium has been investigated in detail (see, e.g., [2]), complete and reliable data for heavier elements, in particular iron, are still missing. Since neutron stars are formed by the collapse of the iron cores of massive stars, it may be assumed that their atmospheres contain an iron plasma. Our objective is to fill the gap
Ueda, Tatsuki; Fujioka, Shinsuke; Nishimura, Hiroaki; Nozaki, Shinya; Azuma, Rumiko; Chen, Yen-Wei
2010-07-15
A coded imaging and decoding (image reconstruction) scheme was developed for diagnosing a hot and dense region emitting hard x-rays and neutrons in laser-fusion plasmas. Because the imager was a uniformly redundant array of penumbral aperture (URPA) arranged in an M-matrix, URPA leads to N times (N: the total number of apertures) enhancement of signal intensity in comparison with a single penumbral aperture. A recorded penumbral image was reconstructed by a computer-based heuristic method to reduce artifacts caused by noises contained in a penumbral image. Applicability of this technique was investigated by imaging x-rays emitted from laser-produced plasmas, demonstrating a spatial resolution of 16 {mu}m. Under the present conditions, the spatial resolution was determined dominantly by a detector resolution (10.5 {mu}m) and a signal-to-noise ratio of the obtained penumbral image.
FOREWORD: Neutron metrology Neutron metrology
NASA Astrophysics Data System (ADS)
Thomas, David J.; Nolte, Ralf; Gressier, Vincent
2011-12-01
industry, from the initial fuel enrichment and fabrication processes right through to storage or reprocessing, and neutron metrology is clearly important in this area. Neutron fields do, however, occur in other areas, for example where neutron sources are used in oil well logging and moisture measurements. They also occur around high energy accelerators, including photon linear accelerators used for cancer therapy, and are expected to be a more serious problem around the new hadron radiation therapy facilities. Roughly 50% of the cosmic ray doses experienced by fliers at the flight altitudes of commercial aircraft are due to neutrons. Current research on fusion presents neutron metrology with a whole new range of challenges because of the very high fluences expected. One of the most significant features of neutron fields is the very wide range of possible neutron energies. In the nuclear industry, for example, neutrons occur with energies from those of thermal neutrons at a few meV to the upper end of the fission spectrum at perhaps 10 MeV. For cosmic ray dosimetry the energy range extends into the GeV region. This enormous range sets a challenge for designing measuring devices and a parallel challenge of developing measurement standards for characterizing these devices. One of the major considerations when deciding on topics for this special issue was agreeing on what not to include. Modelling, i.e. the use of radiation transport codes, is now a very important aspect of neutron measurements. These calculations are vital for shielding and for instrument design; nevertheless, the topic has only been included here where it has a direct bearing on metrology and the development of standards. Neutron spectrometry is an increasingly important technique for unravelling some of the problems of dose equivalent measurements and for plasma diagnostics in fusion research. However, this topic is at least one step removed from primary metrology and so it was felt that it should not be
Code System for the Analysis of Material Test Reactor (MTR) Cores.
1995-03-24
Version 00 The RETRAC code uses a set of coupled neutron point-kinetics equations and thermal-hydraulic conservation laws to simulate nuclear reactor core behavior under transient or accident conditions. The reactor core is represented by a single equivalent unit cell composed of three regions: fuel, clad, and moderator (coolant).
Mook, H.A. Jr.
1984-01-01
In one aspect, the invention is an improved pulsed-neutron monochromator of the vibrated-crystal type. The monochromator is designed to provide neutron pulses which are characterized both by short duration and high density. A row of neutron-reflecting crystals is disposed in a neutron beam to reflect neutrons onto a common target. The crystals in the row define progressively larger neutron-scattering angles and are vibrated sequentially in descending order with respect to the size of their scattering angles, thus generating neutron pulses which arrive simultaneously at the target. Transducers are coupled to one end of the crystals to vibrate them in an essentially non-resonant mode. The transducers propagate transverse waves in the crystal which progress longitudinally therein. The waves are absorbed at the undriven ends of the crystals by damping material mounted thereon. In another aspect, the invention is a method for generating neutron pulses characterized by high intensity and short duration.
Mook, Jr., Herbert A.
1985-01-01
In one aspect, the invention is an improved pulsed-neutron monochromator of the vibrated-crystal type. The monochromator is designed to provide neutron pulses which are characterized both by short duration and high density. A row of neutron-reflecting crystals is disposed in a neutron beam to reflect neutrons onto a common target. The crystals in the row define progressively larger neutron-scattering angles and are vibrated sequentially in descending order with respect to the size of their scattering angles, thus generating neutron pulses which arrive simultaneously at the target. Transducers are coupled to one end of the crystals to vibrate them in an essentially non-resonant mode. The transducers propagate transverse waves in the crystal which progress longitudinally therein. The wave are absorbed at the undriven ends of the crystals by damping material mounted thereon. In another aspect, the invention is a method for generating neutron pulses characterized by high intensity and short duration.
MCNP: Neutron benchmark problems
Whalen, D.J.; Cardon, D.A.; Uhle, J.L.; Hendricks, J.S.
1991-11-01
The recent widespread and increased use of radiation transport codes has produced greater user and institutional demand for assurances that such codes give correct results. Responding to these requirements for code validation, the general purpose Monte Carlo transport code MCNP has been tested on criticality, pulsed sphere, and shielding neutron problem families. Results for each were compared to experimental data. MCNP successfully predicted the experimental results of all three families within the expected data and statistical uncertainties. These successful predictions demonstrate that MCNP can successfully model a broad spectrum of neutron transport problems. 18 refs., 27 figs., 4 tabs.
A user's manual for MASH 1. 0: A Monte Carlo Adjoint Shielding Code System
Johnson, J.O.
1992-03-01
The Monte Carlo Adjoint Shielding Code System, MASH, calculates neutron and gamma-ray environments and radiation protection factors for armored military vehicles, structures, trenches, and other shielding configurations by coupling a forward discrete ordinates air-over-ground transport calculation with an adjoint Monte Carlo treatment of the shielding geometry. Efficiency and optimum use of computer time are emphasized. The code system include the GRTUNCL and DORT codes for air-over-ground transport calculations, the MORSE code with the GIFT5 combinatorial geometry package for adjoint shielding calculations, and several peripheral codes that perform the required data preparations, transformations, and coupling functions. MASH is the successor to the Vehicle Code System (VCS) initially developed at Oak Ridge National Laboratory (ORNL). The discrete ordinates calculation determines the fluence on a coupling surface surrounding the shielding geometry due to an external neutron/gamma-ray source. The Monte Carlo calculation determines the effectiveness of the fluence at that surface in causing a response in a detector within the shielding geometry, i.e., the dose importance'' of the coupling surface fluence. A coupling code folds the fluence together with the dose importance, giving the desired dose response. The coupling code can determine the dose response a a function of the shielding geometry orientation relative to the source, distance from the source, and energy response of the detector. This user's manual includes a short description of each code, the input required to execute the code along with some helpful input data notes, and a representative sample problem (input data and selected output edits) for each code.
Freeze, G.A.; Larson, K.W.; Davies, P.B.; Webb, S.W.
1995-10-01
Long-term repository assessment must consider the processes of (1) gas generation, (2) room closure and expansions due to salt creep, and (3) multiphase (brine and gas) fluid flow, as well as the complex coupling between these three processes. The mechanical creep closure code SANCHO was used to simulate the closure of a single, perfectly sealed disposal room filled with water and backfill. SANCHO uses constitutive models to describe salt creep, waste consolidation, and backfill consolidation, Five different gas-generation rate histories were simulated, differentiated by a rate multiplier, f, which ranged from 0.0 (no gas generation) to 1.0 (expected gas generation under brine-dominated conditions). The results of the SANCHO f-series simulations provide a relationship between gas generation, room closure, and room pressure for a perfectly sealed room. Several methods for coupling this relationship with multiphase fluid flow into and out of a room were examined. Two of the methods are described.
FOREWORD: Neutron metrology Neutron metrology
NASA Astrophysics Data System (ADS)
Thomas, David J.; Nolte, Ralf; Gressier, Vincent
2011-12-01
industry, from the initial fuel enrichment and fabrication processes right through to storage or reprocessing, and neutron metrology is clearly important in this area. Neutron fields do, however, occur in other areas, for example where neutron sources are used in oil well logging and moisture measurements. They also occur around high energy accelerators, including photon linear accelerators used for cancer therapy, and are expected to be a more serious problem around the new hadron radiation therapy facilities. Roughly 50% of the cosmic ray doses experienced by fliers at the flight altitudes of commercial aircraft are due to neutrons. Current research on fusion presents neutron metrology with a whole new range of challenges because of the very high fluences expected. One of the most significant features of neutron fields is the very wide range of possible neutron energies. In the nuclear industry, for example, neutrons occur with energies from those of thermal neutrons at a few meV to the upper end of the fission spectrum at perhaps 10 MeV. For cosmic ray dosimetry the energy range extends into the GeV region. This enormous range sets a challenge for designing measuring devices and a parallel challenge of developing measurement standards for characterizing these devices. One of the major considerations when deciding on topics for this special issue was agreeing on what not to include. Modelling, i.e. the use of radiation transport codes, is now a very important aspect of neutron measurements. These calculations are vital for shielding and for instrument design; nevertheless, the topic has only been included here where it has a direct bearing on metrology and the development of standards. Neutron spectrometry is an increasingly important technique for unravelling some of the problems of dose equivalent measurements and for plasma diagnostics in fusion research. However, this topic is at least one step removed from primary metrology and so it was felt that it should not be
NASA Astrophysics Data System (ADS)
Engel, D.; Klews, M.; Wunner, G.
2009-02-01
We have developed a new method for the fast computation of wavelengths and oscillator strengths for medium-Z atoms and ions, up to iron, at neutron star magnetic field strengths. The method is a parallelized Hartree-Fock approach in adiabatic approximation based on finite-element and B-spline techniques. It turns out that typically 15-20 finite elements are sufficient to calculate energies to within a relative accuracy of 10-5 in 4 or 5 iteration steps using B-splines of 6th order, with parallelization speed-ups of 20 on a 26-processor machine. Results have been obtained for the energies of the ground states and excited levels and for the transition strengths of astrophysically relevant atoms and ions in the range Z=2…26 in different ionization stages. Catalogue identifier: AECC_v1_0 Program summary URL:http://cpc.cs.qub.ac.uk/summaries/AECC_v1_0.html Program obtainable from: CPC Program Library, Queen's University, Belfast, N. Ireland Licensing provisions: Standard CPC licence, http://cpc.cs.qub.ac.uk/licence/licence.html No. of lines in distributed program, including test data, etc.: 3845 No. of bytes in distributed program, including test data, etc.: 27 989 Distribution format: tar.gz Programming language: MPI/Fortran 95 and Python Computer: Cluster of 1-26 HP Compaq dc5750 Operating system: Fedora 7 Has the code been vectorised or parallelized?: Yes RAM: 1 GByte Classification: 2.1 External routines: MPI/GFortran, LAPACK, PyLab/Matplotlib Nature of problem: Calculations of synthetic spectra [1] of strongly magnetized neutron stars are bedevilled by the lack of data for atoms in intense magnetic fields. While the behaviour of hydrogen and helium has been investigated in detail (see, e.g., [2]), complete and reliable data for heavier elements, in particular iron, are still missing. Since neutron stars are formed by the collapse of the iron cores of massive stars, it may be assumed that their atmospheres contain an iron plasma. Our objective is to fill the gap
Chen, Z.
2001-01-01
The Spallation Neutron Source (SNS) is an accelerator-based neutron scattering research facility. The linear accelerator (linac) is the principal accelerating structure and divided into a room-temperature linac and a superconducting linac. The normal conducting linac system that consists of a Drift Tube Linac (DTL) and a Coupled Cavity Linac (CCL) is to be built by Los Alamos National Laboratory. The CCL structure is 55.36-meters long. It accelerates H- beam from 86.8 Mev to 185.6 Mev at operating frequency of 805 MHz. This side coupled cavity structure has 8 cells per segment, 12 segments and 11 bridge couplers per module, and 4 modules total. A 5-MW klystron powers each module. The number 3 and number 9 bridge coupler of each module are connected to the 5-MW RF power supply. The bridge coupler with length of 2.5 {beta}{gamma} is a three-cell structure and located between the segments and allows power flow through the module. The center cell of each bridge coupler is excited during normal operation. To obtain a uniform electromagnetic filed and meet the resonant frequency shift, the RF induced heat must be removed. Thus, the thermal deformation and frequency shift studies are performed via numerical simulations in order to have an appropriate cooling design and predict the frequency shift under operation. The center cell of the bridge coupler also contains a large 4-inch slug tuner and a tuning post that used to provide bulk frequency adjustment and field intensity adjustment, so that produce the proper total field distribution in the module assembly.
Kasinathan, N.; Rajakumar, A.; Vaidyanathan, G.; Chetal, S.C.
1995-09-01
Post shutdown decay heat removal is an important safety requirement in any nuclear system. In order to improve the reliability of this function, Liquid metal (sodium) cooled fast breeder reactors (LMFBR) are equipped with redundant hot pool dipped immersion coolers connected to natural draught air cooled heat exchangers through intermediate sodium circuits. During decay heat removal, flow through the core, immersion cooler primary side and in the intermediate sodium circuits are also through natural convection. In order to establish the viability and validate computer codes used in making predictions, a 1:20 scale experimental model called RAMONA with water as coolant has been built and experimental simulation of decay heat removal situation has been performed at KfK Karlsruhe. Results of two such experiments have been compiled and published as benchmarks. This paper brings out the results of the numerical simulation of one of the benchmark case through a 1D/2D coupled code system, DHDYN-1D/THYC-2D and the salient features of the comparisons. Brief description of the formulations of the codes are also included.
A User's Manual for MASH V1.5 - A Monte Carlo Adjoint Shielding Code System
C. O. Slater; J. M. Barnes; J. O. Johnson; J.D. Drischler
1998-10-01
The Monte Carlo ~djoint ~ielding Code System, MASH, calculates neutron and gamma- ray environments and radiation protection factors for armored military vehicles, structures, trenches, and other shielding configurations by coupling a forward discrete ordinates air- over-ground transport calculation with an adjoint Monte Carlo treatment of the shielding geometry. Efficiency and optimum use of computer time are emphasized. The code system includes the GRTUNCL and DORT codes for air-over-ground transport calculations, the MORSE code with the GIFT5 combinatorial geometry package for adjoint shielding calculations, and several peripheral codes that perform the required data preparations, transformations, and coupling functions. The current version, MASH v 1.5, is the successor to the original MASH v 1.0 code system initially developed at Oak Ridge National Laboratory (ORNL). The discrete ordinates calculation determines the fluence on a coupling surface surrounding the shielding geometry due to an external neutron/gamma-ray source. The Monte Carlo calculation determines the effectiveness of the fluence at that surface in causing a response in a detector within the shielding geometry, i.e., the "dose importance" of the coupling surface fluence. A coupling code folds the fluence together with the dose importance, giving the desired dose response. The coupling code can determine the dose response as a function of the shielding geometry orientation relative to the source, distance from the source, and energy response of the detector. This user's manual includes a short description of each code, the input required to execute the code along with some helpful input data notes, and a representative sample problem.
Polyethylene-reflected plutonium metal sphere : subcritical neutron and gamma measurements.
Mattingly, John K.
2009-11-01
Numerous benchmark measurements have been performed to enable developers of neutron transport models and codes to evaluate the accuracy of their calculations. In particular, for criticality safety applications, the International Criticality Safety Benchmark Experiment Program (ICSBEP) annually publishes a handbook of critical and subcritical benchmarks. Relatively fewer benchmark measurements have been performed to validate photon transport models and codes, and unlike the ICSBEP, there is no program dedicated to the evaluation and publication of photon benchmarks. Even fewer coupled neutron-photon benchmarks have been performed. This report documents a coupled neutron-photon benchmark for plutonium metal reflected by polyethylene. A 4.5-kg sphere of ?-phase, weapons-grade plutonium metal was measured in six reflected configurations: (1) Bare; (2) Reflected by 0.5 inch of high density polyethylene (HDPE); (3) Reflected by 1.0 inch of HDPE; (4) Reflected by 1.5 inches of HDPE; (5) Reflected by 3.0 inches of HDPE; and (6) Reflected by 6.0 inches of HDPE. Neutron and photon emissions from the plutonium sphere were measured using three instruments: (1) A gross neutron counter; (2) A neutron multiplicity counter; and (3) A high-resolution gamma spectrometer. This report documents the experimental conditions and results in detail sufficient to permit developers of radiation transport models and codes to construct models of the experiments and to compare their calculations to the measurements. All of the data acquired during this series of experiments are available upon request.
Neutron spectrometry with a monolithic silicon telescope.
Agosteo, S; D'Angelo, G; Fazzi, A; Para, A Foglio; Pola, A; Zotto, P
2007-01-01
A neutron spectrometer was set-up by coupling a polyethylene converter with a monolithic silicon telescope, consisting of a DeltaE and an E stage-detector (about 2 and 500 microm thick, respectively). The detection system was irradiated with monoenergetic neutrons at INFN-Laboratori Nazionali di Legnaro (Legnaro, Italy). The maximum detectable energy, imposed by the thickness of the E stage, is about 8 MeV for the present detector. The scatter plots of the energy deposited in the two stages were acquired using two independent electronic chains. The distributions of the recoil-protons are well-discriminated from those due to secondary electrons for energies above 0.350 MeV. The experimental spectra of the recoil-protons were compared with the results of Monte Carlo simulations using the FLUKA code. An analytical model that takes into account the geometrical structure of the silicon telescope was developed, validated and implemented in an unfolding code. The capability of reproducing continuous neutron spectra was investigated by irradiating the detector with neutrons from a thick beryllium target bombarded with protons. The measured spectra were compared with data taken from the literature. Satisfactory agreement was found. PMID:17522037
Francis, Brian R.
2015-01-01
Although analysis of the genetic code has allowed explanations for its evolution to be proposed, little evidence exists in biochemistry and molecular biology to offer an explanation for the origin of the genetic code. In particular, two features of biology make the origin of the genetic code difficult to understand. First, nucleic acids are highly complicated polymers requiring numerous enzymes for biosynthesis. Secondly, proteins have a simple backbone with a set of 20 different amino acid side chains synthesized by a highly complicated ribosomal process in which mRNA sequences are read in triplets. Apparently, both nucleic acid and protein syntheses have extensive evolutionary histories. Supporting these processes is a complex metabolism and at the hub of metabolism are the carboxylic acid cycles. This paper advances the hypothesis that the earliest predecessor of the nucleic acids was a β-linked polyester made from malic acid, a highly conserved metabolite in the carboxylic acid cycles. In the β-linked polyester, the side chains are carboxylic acid groups capable of forming interstrand double hydrogen bonds. Evolution of the nucleic acids involved changes to the backbone and side chain of poly(β-d-malic acid). Conversion of the side chain carboxylic acid into a carboxamide or a longer side chain bearing a carboxamide group, allowed information polymers to form amide pairs between polyester chains. Aminoacylation of the hydroxyl groups of malic acid and its derivatives with simple amino acids such as glycine and alanine allowed coupling of polyester synthesis and protein synthesis. Use of polypeptides containing glycine and l-alanine for activation of two different monomers with either glycine or l-alanine allowed simple coded autocatalytic synthesis of polyesters and polypeptides and established the first genetic code. A primitive cell capable of supporting electron transport, thioester synthesis, reduction reactions, and synthesis of polyesters and
Neutronics Modeling of the High Flux Isotope Reactor using COMSOL
Chandler, David; Primm, Trent; Freels, James D; Maldonado, G Ivan
2011-01-01
The High Flux Isotope Reactor located at the Oak Ridge National Laboratory is a versatile 85 MWth research reactor with cold and thermal neutron scattering, materials irradiation, isotope production, and neutron activation analysis capabilities. HFIR staff members are currently in the process of updating the thermal hydraulic and reactor transient modeling methodologies. COMSOL Multiphysics has been adopted for the thermal hydraulic analyses and has proven to be a powerful finite-element-based simulation tool for solving multiple physics-based systems of partial and ordinary differential equations. Modeling reactor transients is a challenging task because of the coupling of neutronics, heat transfer, and hydrodynamics. This paper presents a preliminary COMSOL-based neutronics study performed by creating a two-dimensional, two-group, diffusion neutronics model of HFIR to study the spatially-dependent, beginning-of-cycle fast and thermal neutron fluxes. The 238-group ENDF/B-VII neutron cross section library and NEWT, a two-dimensional, discrete-ordinates neutron transport code within the SCALE 6 code package, were used to calculate the two-group neutron cross sections required to solve the diffusion equations. The two-group diffusion equations were implemented in the COMSOL coefficient form PDE application mode and were solved via eigenvalue analysis using a direct (PARDISO) linear system solver. A COMSOL-provided adaptive mesh refinement algorithm was used to increase the number of elements in areas of largest numerical error to increase the accuracy of the solution. The flux distributions calculated by means of COMSOL/SCALE compare well with those calculated with benchmarked three-dimensional MCNP and KENO models, a necessary first step along the path to implementing two- and three-dimensional models of HFIR in COMSOL for the purpose of studying the spatial dependence of transient-induced behavior in the reactor core.
Rutqvist, J.; Feng, X-T.; Hudson, J.; Jing, L.; Kobayashi, A.; Koyama, T.; Pan, P-Z.; Lee, H-S.; Rinne, M.; Sonnenthal, E.; Yamamoto, Y.
2006-05-10
An international, multiple-code benchmark test (BMT) studyis being conducted within the international DECOVALEX project to analysecoupled thermal, hydrological, mechanical and chemical (THMC) processesin the excavation disturbed zone (EDZ) around emplacement drifts of anuclear waste repository. This BMT focuses on mechanical responses andlong-term chemo-mechanical effects that may lead to changes in mechanicaland hydrological properties in the EDZ. This includes time-de-pendentprocesses such as creep, and subcritical crack, or healing of fracturesthat might cause "weakening" or "hardening" of the rock over the longterm. Five research teams are studying this BMT using a wide range ofmodel approaches, including boundary element, finite element, and finitedifference, particle mechanics, and elasto-plastic cellular automatamethods. This paper describes the definition of the problem andpreliminary simulation results for the initial model inception part, inwhich time dependent effects are not yet included.
Initial Coupling of the RELAP-7 and PRONGHORN Applications
J. Ortensi; D. Andrs; A.A. Bingham; R.C. Martineau; J.W. Peterson
2012-10-01
Modern nuclear reactor safety codes require the ability to solve detailed coupled neutronic- thermal fluids problems. For larger cores, this implies fully coupled higher dimensionality spatial dynamics with appropriate feedback models that can provide enough resolution to accurately compute core heat generation and removal during steady and unsteady conditions. The reactor analysis code PRONGHORN is being coupled to RELAP-7 as a first step to extend RELAP’s current capabilities. This report details the mathematical models, the type of coupling, and the testing results from the integrated system. RELAP-7 is a MOOSE-based application that solves the continuity, momentum, and energy equations in 1-D for a compressible fluid. The pipe and joint capabilities enable it to model parts of the power conversion unit. The PRONGHORN application, also developed on the MOOSE infrastructure, solves the coupled equations that define the neutron diffusion, fluid flow, and heat transfer in a full core model. The two systems are loosely coupled to simplify the transition towards a more complex infrastructure. The integration is tested on a simplified version of the OECD/NEA MHTGR-350 Coupled Neutronics-Thermal Fluids benchmark model.
Biondo, Elliott D; Ibrahim, Ahmad M; Mosher, Scott W; Grove, Robert E
2015-01-01
Detailed radiation transport calculations are necessary for many aspects of the design of fusion energy systems (FES) such as ensuring occupational safety, assessing the activation of system components for waste disposal, and maintaining cryogenic temperatures within superconducting magnets. Hybrid Monte Carlo (MC)/deterministic techniques are necessary for this analysis because FES are large, heavily shielded, and contain streaming paths that can only be resolved with MC. The tremendous complexity of FES necessitates the use of CAD geometry for design and analysis. Previous ITER analysis has required the translation of CAD geometry to MCNP5 form in order to use the AutomateD VAriaNce reducTion Generator (ADVANTG) for hybrid MC/deterministic transport. In this work, ADVANTG was modified to support CAD geometry, allowing hybrid (MC)/deterministic transport to be done automatically and eliminating the need for this translation step. This was done by adding a new ray tracing routine to ADVANTG for CAD geometries using the Direct Accelerated Geometry Monte Carlo (DAGMC) software library. This new capability is demonstrated with a prompt dose rate calculation for an ITER computational benchmark problem using both the Consistent Adjoint Driven Importance Sampling (CADIS) method an the Forward Weighted (FW)-CADIS method. The variance reduction parameters produced by ADVANTG are shown to be the same using CAD geometry and standard MCNP5 geometry. Significant speedups were observed for both neutrons (as high as a factor of 7.1) and photons (as high as a factor of 59.6).
Paul, Rick L.; Davis, W. Clay; Yu, Lee; Murphy, Karen E.; Guthrie, William F.; Leber, Dennis D.; Bryan, Colleen E.; Vetter, Thomas W.; Shakirova, Gulchekhra; Mitchell, Graylin; Kyle, David J.; Jarrett, Jeffery M.; Caldwell, Kathleen L.; Jones, Robert L.; Eckdahl, Steven; Wermers, Michelle; Maras, Melissa; Palmer, C. D.; Verostek, M.F.; Geraghty, C. M.; Steuerwald, Amy J.; Parsons, Patrick J.
2015-01-01
A newly developed procedure for determination of arsenic by radiochemical neutron activation analysis (RNAA) was used to measure arsenic at four levels in SRM 955c Toxic Elements in Caprine Blood and at two levels in SRM 2668 Toxic Elements in Frozen Human Urine for the purpose of providing mass concentration values for certification. Samples were freeze-dried prior to analysis followed by neutron irradiation for 3 h at a fluence rate of 1×1014cm−2s−1. After sample dissolution in perchloric and nitric acids, arsenic was separated from the matrix by extraction into zinc diethyldithiocarbamate in chloroform, and 76As quantified by gamma-ray spectroscopy. Differences in chemical yield and counting geometry between samples and standards were monitored by measuring the count rate of a 77As tracer added before sample dissolution. RNAA results were combined with inductively coupled plasma – mass spectrometry (ICP-MS) values from NIST and collaborating laboratories to provide certified values of (10.81 ± 0.54) μg/kg and (213.1 ± 0.73) μg/kg for SRM 2668 Levels I and II, and certified values of (21.66 ± 0.73) μg/kg, (52.7 ± 1.1) μg/kg, and (78.8 ± 4.9) μg/kg for SRM 955c Levels 2, 3, and 4 respectively. Because of discrepancies between values obtained by different methods for SRM 955c Level 1, an information value of < 5 μg/kg was assigned for this material. PMID:26300575
Sun, K.; Chenu, A.; Mikityuk, K.; Krepel, J.; Chawla, R.
2012-07-01
The core behaviour of a large (3600 MWth) sodium-cooled fast reactor (SFR) is investigated in this paper with the use of a coupled TRACE/PARCS model. The SFR neutron spectrum is characterized by several performance advantages, but also leads to one dominating neutronics drawback - a positive sodium void reactivity. This implies a positive reactivity effect when sodium coolant is removed from the core. In order to evaluate such feedback in terms of the dynamics, a representative unprotected loss-of-flow (ULOF) transient, i.e. flow run-down without SCRAM in which sodium boiling occurs, is analyzed. Although analysis of a single transient cannot allow general conclusions to be drawn, it does allow better understanding of the underlying physics and can lead to proposals for improving the core response during such an accident. The starting point of this study is the reference core design considered in the framework of the Collaborative Project on the European Sodium Fast Reactor (CP-ESFR). To reduce the void effect, the core has been modified by introducing an upper sodium plenum (along with a boron layer) and by reducing the core height-to-diameter ratio. For the ULOF considered, a sharp increase in core power results in melting of the fuel in the case of the reference core. In the modified core, a large dryout leads to melting of the clad. It seems that, for the hypothetical event considered, fuel failure cannot be avoided with just improvement of the neutronics design; therefore, thermal-hydraulics optimization has been considered. An innovative assembly design is proposed to prevent sodium vapour blocking the fuel channel. This results in preventing a downward propagation of the sodium boiling to the core center, thus limiting it to the upper region. Such a void map introduces a negative coolant density reactivity feedback, which dominates the total reactivity change. As a result, the power level and the fuel temperature are effectively reduced, and a large dryout
J. Rutqvist; X. Feng; J. Hudson; L. Jing; A. Kobayashi; T. Koyama; P.Pan; H. Lee; M. Rinne; E. Sonnenthal; Y. Yamamoto
2006-05-08
An international, multiple-code benchmark test (BMT) study is being conducted within the international DECOVALEX project to analyze coupled thermal, hydrological, mechanical and chemical (THMC) processes in the excavation disturbed zone (EDZ) around emplacement drifts of a nuclear waste repository. This BMT focuses on mechanical responses and long-term chemo-mechanical effects that may lead to changes in mechanical and hydrological properties in the EDZ. This includes time-dependent processes such as creep, and subcritical crack, or healing of fractures that might cause ''weakening'' or ''hardening'' of the rock over the long term. Five research teams are studying this BMT using a wide range of model approaches, including boundary element, finite element, and finite difference, particle mechanics, and elasto-plastic cellular automata methods. This paper describes the definition of the problem and preliminary simulation results for the initial model inception part, in which time dependent effects are not yet included.
Vogt, Andreas; Fuerholzner, Bettina; Kinkl, Norbert; Boldt, Karsten; Ueffing, Marius
2013-05-01
High confidence definition of protein interactions is an important objective toward the understanding of biological systems. Isotope labeling in combination with affinity-based isolation of protein complexes has increased in accuracy and reproducibility, yet, larger organisms--including humans--are hardly accessible to metabolic labeling and thus, a major limitation has been its restriction to small animals, cell lines, and yeast. As composition as well as the stoichiometry of protein complexes can significantly differ in primary tissues, there is a great demand for methods capable to combine the selectivity of affinity-based isolation as well as the accuracy and reproducibility of isotope-based labeling with its application toward analysis of protein interactions from intact tissue. Toward this goal, we combined isotope coded protein labeling (ICPL)(1) with immunoprecipitation (IP) and quantitative mass spectrometry (MS). ICPL-IP allows sensitive and accurate analysis of protein interactions from primary tissue. We applied ICPL-IP to immuno-isolate protein complexes from bovine retinal tissue. Protein complexes of immunoprecipitated β-tubulin, a highly abundant protein with known interactors as well as the lowly expressed small GTPase RhoA were analyzed. The results of both analyses demonstrate sensitive and selective identification of known as well as new protein interactions by our method.
Jacques, D; Simůnek, J; Mallants, D; van Genuchten, M Th
2006-12-15
One possible way of integrating subsurface flow and transport processes with (bio)geochemical reactions is to couple by means of an operator-splitting approach two completely separate codes, one for variably-saturated flow and solute transport and one for equilibrium and kinetic biogeochemical reactions. This paper evaluates the accuracy of the operator-splitting approach for multicomponent systems for typical soil environmental problems involving transient atmospheric boundary conditions (precipitation, evapotranspiration) and layered soil profiles. The recently developed HP1 code was used to solve the coupled transport and chemical equations. For steady-state flow conditions, the accuracy was found to be mainly a function of the adopted spatial discretization and to a lesser extent of the temporal discretization. For transient flow situations, the accuracy depended in a complex manner on grid discretization, time stepping and the main flow conditions (infiltration versus evaporation). Whereas a finer grid size reduced the numerical errors during steady-state flow or the main infiltration periods, the errors sometimes slightly increased (generally less than 50%) when a finer grid size was used during periods with a high evapotranspiration demand (leading to high pressure head gradients near the soil surface). This indicates that operator-splitting errors are most significant during periods with high evaporative boundary conditions. The operator-splitting errors could be decreased by constraining the time step using the performance index (the product of the grid Peclet and Courant numbers) during infiltration, or the maximum time step during evapotranspiration. Several test problems were used to provide guidance for optimal spatial and temporal discretization. PMID:16919364
NASA Astrophysics Data System (ADS)
María Gómez Castro, Berta; De Simone, Silvia; Carrera, Jesús
2016-04-01
Nowadays, there are still some unsolved relevant questions which must be faced if we want to proceed to the hydraulic fracturing in a safe way. How much will the fracture propagate? This is one of the most important questions that have to be solved in order to avoid the formation of pathways leading to aquifer targets and atmospheric release. Will the fracture failure provoke a microseismic event? Probably this is the biggest fear that people have in fracking. The aim of this work (developed as a part of the EU - FracRisk project) is to understand the hydro-mechanical coupling that controls the shear of existing fractures and their propagation during a hydraulic fracturing operation, in order to identify the key parameters that dominate these processes and answer the mentioned questions. This investigation focuses on the development of a new C++ code which simulates hydro-mechanical coupling, shear movement and propagation of a fracture. The framework employed, called Kratos, uses the Finite Element Method and the fractures are represented with an interface element which is zero thickness. This means that both sides of the element lie together in the initial configuration (it seems a 1D element in a 2D domain, and a 2D element in a 3D domain) and separate as the adjacent matrix elements deform. Since we are working in hard, fragile rocks, we can assume an elastic matrix and impose irreversible displacements in fractures when rock failure occurs. The formulation used to simulate shear and tensile failures is based on the analytical solution proposed by Okada, 1992 and it is part of an iterative process. In conclusion, the objective of this work is to employ the new code developed to analyze the main uncertainties related with the hydro-mechanical behavior of fractures derived from the hydraulic fracturing operations.
A user`s manual for MASH 1.0: A Monte Carlo Adjoint Shielding Code System
Johnson, J.O.
1992-03-01
The Monte Carlo Adjoint Shielding Code System, MASH, calculates neutron and gamma-ray environments and radiation protection factors for armored military vehicles, structures, trenches, and other shielding configurations by coupling a forward discrete ordinates air-over-ground transport calculation with an adjoint Monte Carlo treatment of the shielding geometry. Efficiency and optimum use of computer time are emphasized. The code system include the GRTUNCL and DORT codes for air-over-ground transport calculations, the MORSE code with the GIFT5 combinatorial geometry package for adjoint shielding calculations, and several peripheral codes that perform the required data preparations, transformations, and coupling functions. MASH is the successor to the Vehicle Code System (VCS) initially developed at Oak Ridge National Laboratory (ORNL). The discrete ordinates calculation determines the fluence on a coupling surface surrounding the shielding geometry due to an external neutron/gamma-ray source. The Monte Carlo calculation determines the effectiveness of the fluence at that surface in causing a response in a detector within the shielding geometry, i.e., the ``dose importance`` of the coupling surface fluence. A coupling code folds the fluence together with the dose importance, giving the desired dose response. The coupling code can determine the dose response a a function of the shielding geometry orientation relative to the source, distance from the source, and energy response of the detector. This user`s manual includes a short description of each code, the input required to execute the code along with some helpful input data notes, and a representative sample problem (input data and selected output edits) for each code.
NASA Astrophysics Data System (ADS)
Samarin, V. V.
2016-05-01
The time-dependent Schrödinger equation and the coupled channel approach based on the method of perturbed stationary two-center states are used to describe nucleon transfers and fusion in low-energy nuclear reactions. Results of the cross sections calculation for the formation of the 198Au and fusion in the 6He+197Au reaction and for the formation of the 65Zn in 6He+64Zn reaction agree satisfactorily with the experimental data near the barrier. The Feynman's continual integrals calculations for a few-body systems were used for the proposal of the new form of the shell model mean field for helium isotopes.
Chen, Jin; Coakley, Arthur; O'Connor, Michelle; Petrov, Alexey; O'Leary, Seán E; Atkins, John F; Puglisi, Joseph D
2015-11-19
Nearly half of the ribosomes translating a particular bacteriophage T4 mRNA bypass a region of 50 nt, resuming translation 3' of this gap. How this large-scale, specific hop occurs and what determines whether a ribosome bypasses remain unclear. We apply single-molecule fluorescence with zero-mode waveguides to track individual Escherichia coli ribosomes during translation of T4's gene 60 mRNA. Ribosomes that bypass are characterized by a 10- to 20-fold longer pause in a non-canonical rotated state at the take-off codon. During the pause, mRNA secondary structure rearrangements are coupled to ribosome forward movement, facilitated by nascent peptide interactions that disengage the ribosome anticodon-codon interactions for slippage. Close to the landing site, the ribosome then scans mRNA in search of optimal base-pairing interactions. Our results provide a mechanistic and conformational framework for bypassing, highlighting a non-canonical ribosomal state to allow for mRNA structure refolding to drive large-scale ribosome movements. PMID:26590426
Neutron detector and fabrication method thereof
Bhandari, Harish B.; Nagarkar, Vivek V.; Ovechkina, Olena E.
2016-08-16
A neutron detector and a method for fabricating a neutron detector. The neutron detector includes a photodetector, and a solid-state scintillator operatively coupled to the photodetector. In one aspect, the method for fabricating a neutron detector includes providing a photodetector, and depositing a solid-state scintillator on the photodetector to form a detector structure.
NASA Astrophysics Data System (ADS)
Baiotti, Luca; Shibata, Masaru; Yamamoto, Tetsuro
2010-09-01
We present the first quantitative comparison of two independent general-relativistic hydrodynamics codes, the whisky code and the sacra code. We compare the output of simulations starting from the same initial data and carried out with the configuration (numerical methods, grid setup, resolution, gauges) which for each code has been found to give consistent and sufficiently accurate results, in particular, in terms of cleanness of gravitational waveforms. We focus on the quantities that should be conserved during the evolution (rest mass, total mass energy, and total angular momentum) and on the gravitational-wave amplitude and frequency. We find that the results produced by the two codes agree at a reasonable level, with variations in the different quantities but always at better than about 10%.
Interfaces MATXS Cross-Section Libraries to Nuclear Transport Codes for Fusion Systems Analysis.
1985-04-10
Version: 00 TRANSX-CTR is a computer code that reads nuclear data from a library in MATXS format and produces transport tables with many discrete-ordinates (Sn) and diffusion codes. Tables can be produced for neutron, photon, or coupled transport. Options include adjoint tables, mixtures, self-shielding, group collapse, homogenization, thermal upscatter, prompt or steady-state fission, transport corrections, elastic removal corrections, and flexible response-function edits. The ability to prepare coupled tables and response edits for heating, damage, gasmore » production, and delayed activity makes TRANSX-CTR especially useful for fusion reactor studies.« less
NASA Astrophysics Data System (ADS)
Wilson, Stephen Christian
Small, highly enriched reactors designed for weapons effects simulations undergo extreme thermal transients during pulsed operations. The primary shutdown mechanism of these reactors---thermal expansion of fuel material---experiences an inertial delay resulting in a different value for the fuel temperature coefficient of reactivity during pulse operation as compared to the value appropriate for steady-state operation. The value appropriate for pulsed operation may further vary as a function of initial reactivity addition. Here we design and implement a finite element numerical method to predict the pulse operation behavior of Sandia Pulsed Reactor (SPR) II, SPR III, and a hypothetical spherical assembly with identical fuel properties without using operationally observed data in our model. These numerical results are compared to available SPR II and SPR III operational data. The numerical methods employed herein may be modified and expanded in functionality to provide both accurate characterization of the behavior of fast burst reactors of any common geometry or isotropic fuel material in the design phase, as well as a computational tool for general coupled thermomechanical-neutronics behavior in the solid state for any reactor type.
Maxey, L Curt; Ally, Tanya R; Brunson, Aly; Garcia, Frances; Goetz, Kathleen C; Hasse, Katelyn E; McManamy, Thomas J; Shea, Thomas J; Simpson, Marc Livingstone
2011-01-01
A fiber-coupled imaging system for monitoring the proton beam profile on the target of the Spallation Neutron Source was developed using reflective, refractive and diffractive optics to focus an image onto a fiber optic imaging bundle. The imaging system monitors the light output from a chromium-doped aluminum oxide (Al{sub 2}0{sub 3}:Cr) scintillator on the nose of the target. Metal optics are used to relay the image to the lenses that focus the image onto the fiber. The material choices for the lenses and fiber were limited to high-purity fused silica, due to the anticipated radiation dose of 10{sup 8} R. In the first generation system (which had no diffractive elements), radiation damage to the scintillator on the nose of the target significantly broadened the normally monochromatic (694 nm) spectrum. This created the need for an achromatic design in the second generation system. This was achieved through the addition of a diffractive optic for chromatic correction. An overview of the target imaging system and its performance, with particular emphasis on the design and testing of a hybrid refractive/diffractive high-purity fused silica imaging triplet, is presented.
Verbeke, J. M.; Petit, O.
2016-06-01
From nuclear safeguards to homeland security applications, the need for the better modeling of nuclear interactions has grown over the past decades. Current Monte Carlo radiation transport codes compute average quantities with great accuracy and performance; however, performance and averaging come at the price of limited interaction-by-interaction modeling. These codes often lack the capability of modeling interactions exactly: for a given collision, energy is not conserved, energies of emitted particles are uncorrelated, and multiplicities of prompt fission neutrons and photons are uncorrelated. Many modern applications require more exclusive quantities than averages, such as the fluctuations in certain observables (e.g., themore » neutron multiplicity) and correlations between neutrons and photons. In an effort to meet this need, the radiation transport Monte Carlo code TRIPOLI-4® was modified to provide a specific mode that models nuclear interactions in a full analog way, replicating as much as possible the underlying physical process. Furthermore, the computational model FREYA (Fission Reaction Event Yield Algorithm) was coupled with TRIPOLI-4 to model complete fission events. As a result, FREYA automatically includes fluctuations as well as correlations resulting from conservation of energy and momentum.« less
Methodology, status and plans for development and assessment of Cathare code
Bestion, D.; Barre, F.; Faydide, B.
1997-07-01
This paper presents the methodology, status and plans for the development, assessment and uncertainty evaluation of the Cathare code. Cathare is a thermalhydraulic code developed by CEA (DRN), IPSN, EDF and FRAMATOME for PWR safety analysis. First, the status of the code development and assessment is presented. The general strategy used for the development and the assessment of the code is presented. Analytical experiments with separate effect tests, and component tests are used for the development and the validation of closure laws. Successive Revisions of constitutive laws are implemented in successive Versions of the code and assessed. System tests or integral tests are used to validate the general consistency of the Revision. Each delivery of a code Version + Revision is fully assessed and documented. A methodology is being developed to determine the uncertainty on all constitutive laws of the code using calculations of many analytical tests and applying the Discrete Adjoint Sensitivity Method (DASM). At last, the plans for the future developments of the code are presented. They concern the optimization of the code performance through parallel computing - the code will be used for real time full scope plant simulators - the coupling with many other codes (neutronic codes, severe accident codes), the application of the code for containment thermalhydraulics. Also, physical improvements are required in the field of low pressure transients and in the modeling for the 3-D model.
Direct Neutron Capture Calculations with Covariant Density Functional Theory Inputs
NASA Astrophysics Data System (ADS)
Zhang, Shi-Sheng; Peng, Jin-Peng; Smith, Michael S.; Arbanas, Goran; Kozub, Ray L.
2014-09-01
Predictions of direct neutron capture are of vital importance for simulations of nucleosynthesis in supernovae, merging neutron stars, and other astrophysical environments. We calculate the direct capture cross sections for E1 transitions using nuclear structure information from a covariant density functional theory as input for the FRESCO coupled-channels reaction code. We find good agreement of our predictions with experimental cross section data on the double closed-shell targets 16O, 48Ca, and 90Zr, and the exotic nucleus 36S. Extensions of the technique for unstable nuclei and for large-scale calculations will be discussed. Predictions of direct neutron capture are of vital importance for simulations of nucleosynthesis in supernovae, merging neutron stars, and other astrophysical environments. We calculate the direct capture cross sections for E1 transitions using nuclear structure information from a covariant density functional theory as input for the FRESCO coupled-channels reaction code. We find good agreement of our predictions with experimental cross section data on the double closed-shell targets 16O, 48Ca, and 90Zr, and the exotic nucleus 36S. Extensions of the technique for unstable nuclei and for large-scale calculations will be discussed. Supported by the U.S. Dept. of Energy, Office of Nuclear Physics.
THERMAL NEUTRON BACKSCATTER IMAGING.
VANIER,P.; FORMAN,L.; HUNTER,S.; HARRIS,E.; SMITH,G.
2004-10-16
Objects of various shapes, with some appreciable hydrogen content, were exposed to fast neutrons from a pulsed D-T generator, resulting in a partially-moderated spectrum of backscattered neutrons. The thermal component of the backscatter was used to form images of the objects by means of a coded aperture thermal neutron imaging system. Timing signals from the neutron generator were used to gate the detection system so as to record only events consistent with thermal neutrons traveling the distance between the target and the detector. It was shown that this time-of-flight method provided a significant improvement in image contrast compared to counting all events detected by the position-sensitive {sup 3}He proportional chamber used in the imager. The technique may have application in the detection and shape-determination of land mines, particularly non-metallic types.
Moisseytsev, A.; Sienicki, J. J.
2012-05-10
Significant progress has been made on the development of a control strategy for the supercritical carbon dioxide (S-CO{sub 2}) Brayton cycle enabling removal of power from an autonomous load following Sodium-Cooled Fast Reactor (SFR) down to decay heat levels such that the S-CO{sub 2} cycle can be used to cool the reactor until decay heat can be removed by the normal shutdown heat removal system or a passive decay heat removal system such as Direct Reactor Auxiliary Cooling System (DRACS) loops with DRACS in-vessel heat exchangers. This capability of the new control strategy eliminates the need for use of a separate shutdown heat removal system which might also use supercritical CO{sub 2}. It has been found that this capability can be achieved by introducing a new control mechanism involving shaft speed control for the common shaft joining the turbine and two compressors following reduction of the load demand from the electrical grid to zero. Following disconnection of the generator from the electrical grid, heat is removed from the intermediate sodium circuit through the sodium-to-CO{sub 2} heat exchanger, the turbine solely drives the two compressors, and heat is rejected from the cycle through the CO{sub 2}-to-water cooler. To investigate the effectiveness of shaft speed control, calculations are carried out using the coupled Plant Dynamics Code-SAS4A/SASSYS-1 code for a linear load reduction transient for a 1000 MWt metallic-fueled SFR with autonomous load following. No deliberate motion of control rods or adjustment of sodium pump speeds is assumed to take place. It is assumed that the S-CO{sub 2} turbomachinery shaft speed linearly decreases from 100 to 20% nominal following reduction of grid load to zero. The reactor power is calculated to autonomously decrease down to 3% nominal providing a lengthy window in time for the switchover to the normal shutdown heat removal system or for a passive decay heat removal system to become effective. However, the
NASA Astrophysics Data System (ADS)
Ballarini, F.; Fluka-Phantoms Team
Astronauts' exposure to space radiation is of major concern for long-term missions, especially for those in deep space such as a possible mission to Mars. Shielding optimization is therefore a crucial issue, and simulations based on radiation transport codes coupled with anthropomorphic model phantoms can be of great help. In this work, carried out with the FLUKA MC code and two anthropomorphic phantoms (a mathematical model and a "voxel" model), distributions of physical (i.e. absorbed), equivalent and "biological" dose in the various tissues and organs were calculated in different shielding conditions for solar minimum and solar maximum GCR spectra, as well as for the August 1972 Solar Particle Event. The biological dose was modeled as the average number of "Complex Lesions" (CL) per cell in a given organ. CLs are clustered DNA breaks previously calculated with "event-by-event" track structure simulations and integrated in the condensed-history FLUKA code. This approach is peculiar in that it is an example of a mechanistically-based quantification of the ionizing radiation action in biological targets; indeed CLs have been shown to play a fundamental role in chromosome aberration induction. The contributions of primary particles and secondary hadrons were calculated separately, thus allowing quantification of the role of nuclear reactions in the shield and in the human body. As expected, the doses calculated for the 1972 SPE decrease dramatically with increasing the Al shielding; nuclear reactions were found to be of minor importance, although their role is higher for internal organs and large shielding. An Al shield thickness of 10 g/cm2 appears sufficient to respect the 30-day deterministic limits recommended by NCRP for missions in Low Earth Orbit. In contrast with the results obtained for SPE, GCR doses to internal organs are not significantly lower than skin doses. However, the relative contribution of secondary hadrons was found to be more important for
Axion cooling of neutron stars
NASA Astrophysics Data System (ADS)
Sedrakian, Armen
2016-03-01
Cooling simulations of neutron stars and their comparison with the data from thermally emitting x-ray sources put constraints on the properties of axions, and by extension, of any light pseudoscalar dark matter particles, whose existence has been postulated to solve the strong-C P problem of QCD. We incorporate the axion emission by pair-breaking and formation processes by S - and P -wave nucleonic condensates in a benchmark code for cooling simulations, as well as provide fit formulas for the rates of these processes. Axion cooling of neutron stars has been simulated for 24 models covering the mass range 1 to 1.8 solar masses, featuring nonaccreted iron and accreted light-element envelopes, and a range of nucleon-axion couplings. The models are based on an equation state predicting conservative physics of superdense nuclear matter that does not allow for the onset of fast cooling processes induced by phase transitions to non-nucleonic forms of matter or high proton concentration. The cooling tracks in the temperature vs age plane were confronted with the (time-averaged) measured surface temperature of the central compact object in the Cas A supernova remnant as well as surface temperatures of three nearby middle-aged thermally emitting pulsars. We find that the axion coupling is limited to fa/107 GeV ≥(5 - 10 ) , which translates into an upper bound on axion mass ma≤(0.06 - 0.12 ) eV for Peccei-Quinn charges of the neutron |Cn|˜0.04 and proton |Cp|˜0.4 characteristic for hadronic models of axions.
Buck, R M; Hall, J M
1999-06-01
COG is a major multiparticle simulation code in the LLNL Monte Carlo radiation transport toolkit. It was designed to solve deep-penetration radiation shielding problems in arbitrarily complex 3D geometries, involving coupled transport of photons, neutrons, and electrons. COG was written to provide as much accuracy as the underlying cross-sections will allow, and has a number of variance-reduction features to speed computations. Recently COG has been applied to the simulation of high- resolution radiographs of complex objects and the evaluation of contraband detection schemes. In this paper we will give a brief description of the capabilities of the COG transport code and show several examples of neutron and gamma-ray imaging simulations. Keywords: Monte Carlo, radiation transport, simulated radiography, nonintrusive inspection, neutron imaging.
Monte Carlo simulation of moderator and reflector in coal analyzer based on a D-T neutron generator.
Shan, Qing; Chu, Shengnan; Jia, Wenbao
2015-11-01
Coal is one of the most popular fuels in the world. The use of coal not only produces carbon dioxide, but also contributes to the environmental pollution by heavy metals. In prompt gamma-ray neutron activation analysis (PGNAA)-based coal analyzer, the characteristic gamma rays of C and O are mainly induced by fast neutrons, whereas thermal neutrons can be used to induce the characteristic gamma rays of H, Si, and heavy metals. Therefore, appropriate thermal and fast neutrons are beneficial in improving the measurement accuracy of heavy metals, and ensure that the measurement accuracy of main elements meets the requirements of the industry. Once the required yield of the deuterium-tritium (d-T) neutron generator is determined, appropriate thermal and fast neutrons can be obtained by optimizing the neutron source term. In this article, the Monte Carlo N-Particle (MCNP) Transport Code and Evaluated Nuclear Data File (ENDF) database are used to optimize the neutron source term in PGNAA-based coal analyzer, including the material and shape of the moderator and neutron reflector. The optimized targets include two points: (1) the ratio of the thermal to fast neutron is 1:1 and (2) the total neutron flux from the optimized neutron source in the sample increases at least 100% when compared with the initial one. The simulation results show that, the total neutron flux in the sample increases 102%, 102%, 85%, 72%, and 62% with Pb, Bi, Nb, W, and Be reflectors, respectively. Maximum optimization of the targets is achieved when the moderator is a 3-cm-thick lead layer coupled with a 3-cm-thick high-density polyethylene (HDPE) layer, and the neutron reflector is a 27-cm-thick hemispherical lead layer. PMID:26325583
Monte Carlo simulation of moderator and reflector in coal analyzer based on a D-T neutron generator.
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.
1990-11-20
Version 00 REX2-87 is a computer code developed for the calculation of self-shielded multigroup average cross sections, and self-shielding factors for total, elastic, fission and capture processes from an ENDF/B formatted nuclear data file in which the tabulated cross sections follow linear interpolation throughout.
Metcalf, H.E.
1958-10-14
Methods of controlling reactors are presented. Specifically, a plurality of neutron absorber members are adjustably disposed in the reactor core at different distances from the center thereof. The absorber members extend into the core from opposite faces thereof and are operated by motive means coupled in a manner to simultaneously withdraw at least one of the absorber members while inserting one of the other absorber members. This feature effects fine control of the neutron reproduction ratio by varying the total volume of the reactor effective in developing the neutronic reaction.
Exploration of direct neutron capture with covariant density functional theory inputs
NASA Astrophysics Data System (ADS)
Zhang, Shi-Sheng; Peng, Jin-Peng; Smith, M. S.; Arbanas, G.; Kozub, R. L.
2015-04-01
Predictions of direct neutron capture are of vital importance for simulations of nucleosynthesis in supernovae, merging neutron stars, and other astrophysical environments. We calculated direct capture cross sections using nuclear structure information obtained from a covariant density functional theory as input for the fresco coupled reaction channels code. We investigated the impact of pairing, spectroscopic factors, and optical potentials on our results to determine a robust method to calculate cross sections of direct neutron capture on exotic nuclei. Our predictions agree reasonably well with experimental cross section data for the closed shell nuclei 16O and 48Ca, and for the exotic nucleus 36S . We then used this approach to calculate the direct neutron capture cross section on the doubly magic unstable nucleus 132Sn which is of interest for the astrophysical r-process.
Current and anticipated uses of thermal hydraulic codes in Korea
Kim, Kyung-Doo; Chang, Won-Pyo
1997-07-01
In Korea, the current uses of thermal hydraulic codes are categorized into 3 areas. The first application is in designing both nuclear fuel and NSSS. The codes have usually been introduced based on the technology transfer programs agreed between KAERI and the foreign vendors. Another area is in the supporting of the plant operations and licensing by the utility. The third category is research purposes. In this area assessments and some applications to the safety issue resolutions are major activities using the best estimate thermal hydraulic codes such as RELAP5/MOD3 and CATHARE2. Recently KEPCO plans to couple thermal hydraulic codes with a neutronics code for the design of the evolutionary type reactor by 2004. KAERI also plans to develop its own best estimate thermal hydraulic code, however, application range is different from KEPCO developing code. Considering these activities, it is anticipated that use of the best estimate hydraulic analysis code developed in Korea may be possible in the area of safety evaluation within 10 years.
Radiation transport phenomena and modeling - part A: Codes
Lorence, L.J.
1997-06-01
The need to understand how particle radiation (high-energy photons and electrons) from a variety of sources affects materials and electronics has motivated the development of sophisticated computer codes that describe how radiation with energies from 1.0 keV to 100.0 GeV propagates through matter. Predicting radiation transport is the necessary first step in predicting radiation effects. The radiation transport codes that are described here are general-purpose codes capable of analyzing a variety of radiation environments including those produced by nuclear weapons (x-rays, gamma rays, and neutrons), by sources in space (electrons and ions) and by accelerators (x-rays, gamma rays, and electrons). Applications of these codes include the study of radiation effects on electronics, nuclear medicine (imaging and cancer treatment), and industrial processes (food disinfestation, waste sterilization, manufacturing.) The primary focus will be on coupled electron-photon transport codes, with some brief discussion of proton transport. These codes model a radiation cascade in which electrons produce photons and vice versa. This coupling between particles of different types is important for radiation effects. For instance, in an x-ray environment, electrons are produced that drive the response in electronics. In an electron environment, dose due to bremsstrahlung photons can be significant once the source electrons have been stopped.
NASA Astrophysics Data System (ADS)
Zheng, Jingjing; Meana-Pañeda, Rubén; Truhlar, Donald G.
2013-08-01
We present an improved version of the MSTor program package, which calculates partition functions and thermodynamic functions of complex molecules involving multiple torsions; the method is based on either a coupled torsional potential or an uncoupled torsional potential. The program can also carry out calculations in the multiple-structure local harmonic approximation. The program package also includes seven utility codes that can be used as stand-alone programs to calculate reduced moment of inertia matrices by the method of Kilpatrick and Pitzer, to generate conformational structures, to calculate, either analytically or by Monte Carlo sampling, volumes for torsional subdomains defined by Voronoi tessellation of the conformational subspace, to generate template input files for the MSTor calculation and Voronoi calculation, and to calculate one-dimensional torsional partition functions using the torsional eigenvalue summation method. Restrictions: There is no limit on the number of torsions that can be included in either the Voronoi calculation or the full MS-T calculation. In practice, the range of problems that can be addressed with the present method consists of all multitorsional problems for which one can afford to calculate all the conformational structures and their frequencies. Unusual features: The method can be applied to transition states as well as stable molecules. The program package also includes the hull program for the calculation of Voronoi volumes, the symmetry program for determining point group symmetry of a molecule, and seven utility codes that can be used as stand-alone programs to calculate reduced moment-of-inertia matrices by the method of Kilpatrick and Pitzer, to generate conformational structures, to calculate, either analytically or by Monte Carlo sampling, volumes of the torsional subdomains defined by Voronoi tessellation of the conformational subspace, to generate template input files, and to calculate one-dimensional torsional
EXTENSION OF THE NUCLEAR REACTION MODEL CODE EMPIRE TO ACTINIDES NUCLEAR DATA EVALUATION.
CAPOTE,R.; SIN, M.; TRKOV, A.; HERMAN, M.; CARLSON, B.V.; OBLOZINSKY, P.
2007-04-22
Recent extensions and improvements of the EMPIRE code system are outlined. They add new capabilities to the code, such as prompt fission neutron spectra calculations using Hauser-Feshbach plus pre-equilibrium pre-fission spectra, cross section covariance matrix calculations by Monte Carlo method, fitting of optical model parameters, extended set of optical model potentials including new dispersive coupled channel potentials, parity-dependent level densities and transmission through numerically defined fission barriers. These features, along with improved and validated ENDF formatting, exclusive/inclusive spectra, and recoils make the current EMPIRE release a complete and well validated tool for evaluation of nuclear data at incident energies above the resonance region. The current EMPIRE release has been used in evaluations of neutron induced reaction files for {sup 232}Th and {sup 231,233}Pa nuclei in the fast neutron region at IAEA. Triple-humped fission barriers and exclusive pre-fission neutron spectra were considered for the fission data evaluation. Total, fission, capture and neutron emission cross section, average resonance parameters and angular distributions of neutron scattering are in excellent agreement with the available experimental data.
Neutrons in the moon. [neutron flux and production rate calculations
NASA Technical Reports Server (NTRS)
Kornblum, J. J.; Fireman, E. L.; Levine, M.; Aronson, A.
1973-01-01
Neutron fluxes for energies between 15 MeV and thermal at depths of 0 to 300 g/sq cm in the moon are calculated by the discrete ordinate mathod with the ANISN code. With the energy spectrum of Lingenfelter et al. (1972). A total neutron-production rate for the moon of 26 plus or minus neutrons/sq cm sec is determined from the Ar-37 activity measurements in the Apollo 16 drill string, which are found to have a depth dependence in accordance with a neutron source function that decreases exponentially with an attenuation length of 155 g/sq cm.
Compressible Astrophysics Simulation Code
Howell, L.; Singer, M.
2007-07-18
This is an astrophysics simulation code involving a radiation diffusion module developed at LLNL coupled to compressible hydrodynamics and adaptive mesh infrastructure developed at LBNL. One intended application is to neutrino diffusion in core collapse supernovae.
Shifting scintillator neutron detector
Clonts, Lloyd G; Cooper, Ronald G; Crow, Jr., Morris Lowell; Hannah, Bruce W; Hodges, Jason P; Richards, John D; Riedel, Richard A
2014-03-04
Provided are sensors and methods for detecting thermal neutrons. Provided is an apparatus having a scintillator for absorbing a neutron, the scintillator having a back side for discharging a scintillation light of a first wavelength in response to the absorbed neutron, an array of wavelength-shifting fibers proximate to the back side of the scintillator for shifting the scintillation light of the first wavelength to light of a second wavelength, the wavelength-shifting fibers being disposed in a two-dimensional pattern and defining a plurality of scattering plane pixels where the wavelength-shifting fibers overlap, a plurality of photomultiplier tubes, in coded optical communication with the wavelength-shifting fibers, for converting the light of the second wavelength to an electronic signal, and a processor for processing the electronic signal to identify one of the plurality of scattering plane pixels as indicative of a position within the scintillator where the neutron was absorbed.
Ford, W.E. III; Arwood, J.W.; Greene, N.M.; Moses, D.L.; Petrie, L.M.; Primm, R.T. III; Slater, C.O.; Westfall, R.M.; Wright, R.Q.
1990-09-01
Pseudo-problem-independent, multigroup cross-section libraries were generated to support Advanced Neutron Source (ANS) Reactor design studies. The ANS is a proposed reactor which would be fueled with highly enriched uranium and cooled with heavy water. The libraries, designated ANSL-V (Advanced Neutron Source Cross Section Libraries based on ENDF/B-V), are data bases in AMPX master format for subsequent generation of problem-dependent cross-sections for use with codes such as KENO, ANISN, XSDRNPM, VENTURE, DOT, DORT, TORT, and MORSE. Included in ANSL-V are 99-group and 39-group neutron, 39-neutron-group 44-gamma-ray-group secondary gamma-ray production (SGRP), 44-group gamma-ray interaction (GRI), and coupled, 39-neutron group 44-gamma-ray group (CNG) cross-section libraries. The neutron and SGRP libraries were generated primarily from ENDF/B-V data; the GRI library was generated from DLC-99/HUGO data, which is recognized as the ENDF/B-V photon interaction data. Modules from the AMPX and NJOY systems were used to process the multigroup data. Validity of selected data from the fine- and broad-group neutron libraries was satisfactorily tested in performance parameter calculations.
Simple Coupling of Reactor Physics Effects and Uncertain Nuances
2012-08-27
The "Simple Coupling of Reactor Physics Effects and Uncertain Nuances" (SCORPEUN) code is a simple r-z 1-group neutron diffusion code where each r-mesh is coupled to a single-flow-channel model that represents all flow-channels in that r-mesh. This 1-D model assesses q=m*Cp*deletaT for each z-mesh in that channel. This flow channel model is then coupled to a simple 1-D heat conduction model for ascertaining the peak center-line fuel temperature in a hypothetical pin assigned to thatmore » flow channel. The code has property lookup capability for water, Na, Zirc, HT9, metalic fuel, oxide fuel, etc. It has linear interpolation features for micro-scopic cross-sections with respect to coolant density and fuel temperature. ***This last feature has not been fully tested and may need development***. The interpolated microscopic cross-sections are then combined (using the water density from the T/H calculation) to generate macroscopic diffusion coefficient, removal cross-section and nu-sigmaF for each r-z mesh of the neutron diffusion code.« less
Slater, C.O.
1992-01-01
The DRC2 code, which couples MASH or MASHX adjoint leakages with DORT 2-D discrete ordinates forward directional fluences, is described. The forward fluences are allowed to vary both axially and radially over the coupling surface, as opposed to the strictly axial variation allowed by the predecessor DRC code. Input instructions are presented along with descriptions and results from several sample problems. Results from the sample problems are used to compare DRC2 with DRC, DRC2 with DORT, and DRC2 with itself for the case of x-y dependence versus no x-y dependence of the forward fluence. The test problems demonstrate that for small systems DRC and DRC2 give essentially the same results. Some significant differences are noted for larger systems. Additionally, DRC2 results with no x-y dependence of the forward directional fluences are practically the same as those calculated by DRC.
ALEPH2 - A general purpose Monte Carlo depletion code
Stankovskiy, A.; Van Den Eynde, G.; Baeten, P.; Trakas, C.; Demy, P. M.; Villatte, L.
2012-07-01
The Monte-Carlo burn-up code ALEPH is being developed at SCK-CEN since 2004. A previous version of the code implemented the coupling between the Monte Carlo transport (any version of MCNP or MCNPX) and the ' deterministic' depletion code ORIGEN-2.2 but had important deficiencies in nuclear data treatment and limitations inherent to ORIGEN-2.2. A new version of the code, ALEPH2, has several unique features making it outstanding among other depletion codes. The most important feature is full data consistency between steady-state Monte Carlo and time-dependent depletion calculations. The last generation general-purpose nuclear data libraries (JEFF-3.1.1, ENDF/B-VII and JENDL-4) are fully implemented, including special purpose activation, spontaneous fission, fission product yield and radioactive decay data. The built-in depletion algorithm allows to eliminate the uncertainties associated with obtaining the time-dependent nuclide concentrations. A predictor-corrector mechanism, calculation of nuclear heating, calculation of decay heat, decay neutron sources are available as well. The validation of the code on the results of REBUS experimental program has been performed. The ALEPH2 has shown better agreement with measured data than other depletion codes. (authors)
Computational radiology and imaging with the MCNP Monte Carlo code
Estes, G.P.; Taylor, W.M.
1995-05-01
MCNP, a 3D coupled neutron/photon/electron Monte Carlo radiation transport code, is currently used in medical applications such as cancer radiation treatment planning, interpretation of diagnostic radiation images, and treatment beam optimization. This paper will discuss MCNP`s current uses and capabilities, as well as envisioned improvements that would further enhance MCNP role in computational medicine. It will be demonstrated that the methodology exists to simulate medical images (e.g. SPECT). Techniques will be discussed that would enable the construction of 3D computational geometry models of individual patients for use in patient-specific studies that would improve the quality of care for patients.
Sekimoto, Shun; Ebihara, Mitsuru
2013-07-01
Trace amounts of three halogens (chlorine, bromine, and iodine) were determined using radiochemical neutron activation analysis (RNAA) for nine sedimentary rocks and three rhyolite samples. To obtain high-quality analytical data, the radiochemical procedure of RNAA was improved by lowering the background in gamma-ray spectrometry and completing the chemical procedure more rapidly than in conventional procedures. A comparison of the RNAA data of Br and I with corresponding inductively coupled plasma mass spectrometry (ICPMS) literature data revealed that the values obtained by ICPMS coupled with pyrohydrolysis preconcentration were systematically lower than the RNAA data for some reference samples, suggesting that the quantitative collection of Br and I cannot always be achieved by the pyrohydrolysis for some solid samples. The RNAA data of three halogens can classify sedimentary rock reference samples into two groups (the samples from inland water and those from seawater), implying the geochemical significance of halogen data.
Sekimoto, Shun; Ebihara, Mitsuru
2013-07-01
Trace amounts of three halogens (chlorine, bromine, and iodine) were determined using radiochemical neutron activation analysis (RNAA) for nine sedimentary rocks and three rhyolite samples. To obtain high-quality analytical data, the radiochemical procedure of RNAA was improved by lowering the background in gamma-ray spectrometry and completing the chemical procedure more rapidly than in conventional procedures. A comparison of the RNAA data of Br and I with corresponding inductively coupled plasma mass spectrometry (ICPMS) literature data revealed that the values obtained by ICPMS coupled with pyrohydrolysis preconcentration were systematically lower than the RNAA data for some reference samples, suggesting that the quantitative collection of Br and I cannot always be achieved by the pyrohydrolysis for some solid samples. The RNAA data of three halogens can classify sedimentary rock reference samples into two groups (the samples from inland water and those from seawater), implying the geochemical significance of halogen data. PMID:23710630
Recent development and applications of the MORSE Code
Cramer, S.N.
1993-06-01
Several recent analyses using the multigroup MORSE Monte Carlo code are presented. In the calculation of a highly directional-dependent neutron streaming experiment it is shown that P{sub 7} cross section representation produces results virtually identical with those from an analog code. Use has been made here of a recently released ENDF/B-VI data set. In the analysis of neutron distributions inside the water-cooled ORELA accelerator target and positron source, an analytic hydrogen scattering model is incorporated into the otherwise multigroup treatment. The radiation from a nuclear weapon is analyzed in a large concrete building in Nagasaki by coupling MORSE and the DOT discrete ordinates code. The spatial variation of the DOT-generated free-field radiation is utilized, and the building is modeled with the array feature of the MORSE geometry package. An analytic directional biasing, applicable to the discrete scattering angle procedure in MORSE, is combined with the exponential transform. As in more general studies, it is shown that the combined biasing is more efficient than either biasing used separately. Other tracking improvements are included in a difficult streaming and penetration radiation analysis through a concrete structure. Proposals are given for the code generation of the required biasing parameters.
Slow neutron leakage spectra from spallation neutron sources
Das, S G; Carpenter, J M; Prael, R E
1980-02-01
An efficient technique is described for Monte Carlo simulation of neutron beam spectra from target-moderator-reflector assemblies typical of pulsed spallation neutron sources. The technique involves the scoring of the transport-theoretical probability that a neutron will emerge from the moderator surface in the direction of interest, at each collision. An angle-biasing probability is also introduced which further enhances efficiency in simple problems. These modifications were introduced into the VIM low energy neutron transport code, representing the spatial and energy distributions of the source neutrons approximately as those of evaporation neutrons generated through the spallation process by protons of various energies. The intensity of slow neutrons leaking from various reflected moderators was studied for various neutron source arrangements. These include computations relating to early measurements on a mockup-assembly, a brief survey of moderator materials and sizes, and a survey of the effects of varying source and moderator configurations with a practical, liquid metal cooled uranium source Wing and slab, i.e., tangential and radial moderator arrangements, and Be vs CH/sub 2/ reflectors are compared. Results are also presented for several complicated geometries which more closely represent realistic arrangements for a practical source, and for a subcritical fission multiplier such as might be driven by an electron linac. An adaptation of the code was developed to enable time dependent calculations, and investigated the effects of the reflector, decoupling and void liner materials on the pulse shape.
NASA Technical Reports Server (NTRS)
Berger, H.
1970-01-01
Electronic intensifier tube with a demagnification ratio of 9-1 enhances the usefulness of neutron-radiographic techniques. A television signal can be obtained by optical coupling of a small-output phosphor-light image to a television camera.
Advanced multiphysics coupling for LWR fuel performance analysis
Hales, J. D.; Tonks, M. R.; Gleicher, F. N.; Spencer, B. W.; Novascone, S. R.; Williamson, R. L.; Pastore, G.; Perez, D. M.
2015-10-01
Even the most basic nuclear fuel analysis is a multiphysics undertaking, as a credible simulation must consider at a minimum coupled heat conduction and mechanical deformation. The need for more realistic fuel modeling under a variety of conditions invariably leads to a desire to include coupling between a more complete set of the physical phenomena influencing fuel behavior, including neutronics, thermal hydraulics, and mechanisms occurring at lower length scales. This paper covers current efforts toward coupled multiphysics LWR fuel modeling in three main areas. The first area covered in this paper concerns thermomechanical coupling. The interaction of these two physics, particularly related to the feedback effect associated with heat transfer and mechanical contact at the fuel/clad gap, provides numerous computational challenges. An outline is provided of an effective approach used to manage the nonlinearities associated with an evolving gap in BISON, a nuclear fuel performance application. A second type of multiphysics coupling described here is that of coupling neutronics with thermomechanical LWR fuel performance. DeCART, a high-fidelity core analysis program based on the method of characteristics, has been coupled to BISON. DeCART provides sub-pin level resolution of the multigroup neutron flux, with resonance treatment, during a depletion or a fast transient simulation. Two-way coupling between these codes was achieved by mapping fission rate density and fast neutron flux fields from DeCART to BISON and the temperature field from BISON to DeCART while employing a Picard iterative algorithm. Finally, the need for multiscale coupling is considered. Fission gas production and evolution significantly impact fuel performance by causing swelling, a reduction in the thermal conductivity, and fission gas release. The mechanisms involved occur at the atomistic and grain scale and are therefore not the domain of a fuel performance code. However, it is possible to use
Advanced multiphysics coupling for LWR fuel performance analysis
Hales, J. D.; Tonks, M. R.; Gleicher, F. N.; Spencer, B. W.; Novascone, S. R.; Williamson, R. L.; Pastore, G.; Perez, D. M.
2015-10-01
Even the most basic nuclear fuel analysis is a multiphysics undertaking, as a credible simulation must consider at a minimum coupled heat conduction and mechanical deformation. The need for more realistic fuel modeling under a variety of conditions invariably leads to a desire to include coupling between a more complete set of the physical phenomena influencing fuel behavior, including neutronics, thermal hydraulics, and mechanisms occurring at lower length scales. This paper covers current efforts toward coupled multiphysics LWR fuel modeling in three main areas. The first area covered in this paper concerns thermomechanical coupling. The interaction of these two physics,more » particularly related to the feedback effect associated with heat transfer and mechanical contact at the fuel/clad gap, provides numerous computational challenges. An outline is provided of an effective approach used to manage the nonlinearities associated with an evolving gap in BISON, a nuclear fuel performance application. A second type of multiphysics coupling described here is that of coupling neutronics with thermomechanical LWR fuel performance. DeCART, a high-fidelity core analysis program based on the method of characteristics, has been coupled to BISON. DeCART provides sub-pin level resolution of the multigroup neutron flux, with resonance treatment, during a depletion or a fast transient simulation. Two-way coupling between these codes was achieved by mapping fission rate density and fast neutron flux fields from DeCART to BISON and the temperature field from BISON to DeCART while employing a Picard iterative algorithm. Finally, the need for multiscale coupling is considered. Fission gas production and evolution significantly impact fuel performance by causing swelling, a reduction in the thermal conductivity, and fission gas release. The mechanisms involved occur at the atomistic and grain scale and are therefore not the domain of a fuel performance code. However, it is
Domino, Stefan; Luketa-Hanlin, Anay; Gallegos, Carlos
2006-10-27
FAA Smoke Transport Code, a physics-based Computational Fluid Dynamics tool, which couples heat, mass, and momentum transfer, has been developed to provide information on smoke transport in cargo compartments with various geometries and flight conditions. The software package contains a graphical user interface for specification of geometry and boundary conditions, analysis module for solving the governing equations, and a post-processing tool. The current code was produced by making substantial improvements and additions to a code obtained from a university. The original code was able to compute steady, uniform, isothermal turbulent pressurization. In addition, a preprocessor and postprocessor were added to arrive at the current software package.
Artificial Neural Networks in Spectrometry and Neutron Dosimetry
Vega-Carrillo, H. R.; Martinez-Blanco, M. R.; Ortiz-Rodriguez, J. M.; Hernandez-Davila, V. M.
2010-12-07
The ANN technology has been applied to unfold the neutron spectra of three neutron sources and to estimate their dosimetric features. To compare these results, neutron spectra were also unfolded with the BUNKIUT code. Both unfolding procedures were carried out using the count rates of a Bonner sphere spectrometer. The spectra unfolded with ANN result similar to those unfolded with the BUNKIUT code. The H*(10) values obtained with ANN agrees well with H*(10) values calculated with the BUNKIUT code.
Applications guide to the MORSE Monte Carlo code
Cramer, S.N.
1985-08-01
A practical guide for the implementation of the MORESE-CG Monte Carlo radiation transport computer code system is presented. The various versions of the MORSE code are compared and contrasted, and the many references dealing explicitly with the MORSE-CG code are reviewed. The treatment of angular scattering is discussed, and procedures for obtaining increased differentiality of results in terms of reaction types and nuclides from a multigroup Monte Carlo code are explained in terms of cross-section and geometry data manipulation. Examples of standard cross-section data input and output are shown. Many other features of the code system are also reviewed, including (1) the concept of primary and secondary particles, (2) fission neutron generation, (3) albedo data capability, (4) DOMINO coupling, (5) history file use for post-processing of results, (6) adjoint mode operation, (7) variance reduction, and (8) input/output. In addition, examples of the combinatorial geometry are given, and the new array of arrays geometry feature (MARS) and its three-dimensional plotting code (JUNEBUG) are presented. Realistic examples of user routines for source, estimation, path-length stretching, and cross-section data manipulation are given. A deatiled explanation of the coupling between the random walk and estimation procedure is given in terms of both code parameters and physical analogies. The operation of the code in the adjoint mode is covered extensively. The basic concepts of adjoint theory and dimensionality are discussed and examples of adjoint source and estimator user routines are given for all common situations. Adjoint source normalization is explained, a few sample problems are given, and the concept of obtaining forward differential results from adjoint calculations is covered. Finally, the documentation of the standard MORSE-CG sample problem package is reviewed and on-going and future work is discussed.
Development of a dynamic simulation mode in Serpent 2 Monte Carlo code
Leppaenen, J.
2013-07-01
This paper presents a dynamic neutron transport mode, currently being implemented in the Serpent 2 Monte Carlo code for the purpose of simulating short reactivity transients with temperature feedback. The transport routine is introduced and validated by comparison to MCNP5 calculations. The method is also tested in combination with an internal temperature feedback module, which forms the inner part of a multi-physics coupling scheme in Serpent 2. The demo case for the coupled calculation is a reactivity-initiated accident (RIA) in PWR fuel. (authors)
NASA Astrophysics Data System (ADS)
Cao, X. G.; Cai, X. Z.; Ma, Y. G.; Fang, D. Q.; Zhang, G. Q.; Guo, W.; Chen, J. G.; Wang, J. S.
2012-10-01
Proton-neutron, neutron-neutron, and proton-proton momentum-correlation functions (Cpn,Cnn, and Cpp) are systematically investigated for 15C and other C-isotope-induced collisions at different entrance channel conditions within the framework of the isospin-dependent quantum-molecular-dynamics model complemented by the correlation after burner (crab) computation code. 15C is a prime exotic nucleus candidate due to the weakly bound valence neutron coupling with closed-neutron-shell nucleus 14C. To study density dependence of the correlation function by removing the isospin effect, the initialized 15C projectiles are sampled from two kinds of density distribution from the relativistic mean-field (RMF) model in which the valence neutron of 15C is populated in both 1d5/2 and 2s1/2 states, respectively. The results show that the density distributions of the valence neutron significantly influence the nucleon-nucleon momentum-correlation function at large impact parameters and high incident energies. The extended density distribution of the valence neutron largely weakens the strength of the correlation function. The size of the emission source is extracted by fitting the correlation function by using the Gaussian source method. The emission source size as well as the size of the final-state phase space are larger for projectile samplings from more extended density distributions of the valence neutron, which corresponds to the 2s1/2 state in the RMF model. Therefore, the nucleon-nucleon momentum-correlation function can be considered as a potentially valuable tool to diagnose exotic nuclear structures, such as the skin and halo.
NASA Technical Reports Server (NTRS)
Korff, S. A.; Mendell, R. B.; Merker, M.; Light, E. S.; Verschell, H. J.; Sandie, W. S.
1979-01-01
Contributions to fast neutron measurements in the atmosphere are outlined. The results of a calculation to determine the production, distribution and final disappearance of atmospheric neutrons over the entire spectrum are presented. An attempt is made to answer questions that relate to processes such as neutron escape from the atmosphere and C-14 production. In addition, since variations of secondary neutrons can be related to variations in the primary radiation, comment on the modulation of both radiation components is made.
Hardening neutron spectrum for advanced actinide transmutation experiments in the ATR.
Chang, G S; Ambrosek, R G
2005-01-01
The most effective method for transmuting long-lived isotopes contained in spent nuclear fuel into shorter-lived fission products is in a fast neutron spectrum reactor. In the absence of a fast test reactor in the United States, initial irradiation testing of candidate fuels can be performed in a thermal test reactor that has been modified to produce a test region with a hardened neutron spectrum. Such a test facility, with a spectrum similar but somewhat softer than that of the liquid-metal fast breeder reactor (LMFBR), has been constructed in the INEEL's Advanced Test Reactor (ATR). The radial fission power distribution of the actinide fuel pin, which is an important parameter in fission gas release modelling, needs to be accurately predicted and the hardened neutron spectrum in the ATR and the LMFBR fast neutron spectrum is compared. The comparison analyses in this study are performed using MCWO, a well-developed tool that couples the Monte Carlo transport code MCNP with the isotope depletion and build-up code ORIGEN-2. MCWO analysis yields time-dependent and neutron-spectrum-dependent minor actinide and Pu concentrations and detailed radial fission power profile calculations for a typical fast reactor (LMFBR) neutron spectrum and the hardened neutron spectrum test region in the ATR. The MCWO-calculated results indicate that the cadmium basket used in the advanced fuel test assembly in the ATR can effectively depress the linear heat generation rate in the experimental fuels and harden the neutron spectrum in the test region.
Greene, Geoffrey L.
1999-01-01
A neutron guide in which lengths of cylindrical glass tubing have rectangular glass plates properly dimensioned to allow insertion into the cylindrical glass tubing so that a sealed geometrically precise polygonal cross-section is formed in the cylindrical glass tubing. The neutron guide provides easier alignment between adjacent sections than do the neutron guides of the prior art.
Quinby, Thomas C.
1976-07-27
A method of measuring neutron radiation within a nuclear reactor is provided. A sintered oxide wire is disposed within the reactor and exposed to neutron radiation. The induced radioactivity is measured to provide an indication of the neutron energy and flux within the reactor.
1993-08-09
Version 00 This data base was developed for use in Monte Carlo or discrete ordinate transport codes, for example, the general Monte Carlo code MCNP. Various modules of the NJOY processing code system have been enhanced to permit processing of the ENDF/B-VI formatted evaluations into both continuous-energy and multi-group format. The transport data files for all 18 projectile-plus-target systems have been processed through NJOY, and coupled multi-particle, multi-group transport libraries for MCNP now exist. Inmore » addition, pointwise MCNP libraries to 100 MeV for incident neutrons have been prepared for the nine targets. The production version of the MCNP code is being modified to handle the new pointwise libraries. The production version of MCNP already supports the use of coupled multi-group libraries.« less
NASA Technical Reports Server (NTRS)
Ballarini, F.; Battistoni, G.; Campanella, M.; Carboni, M.; Cerutti, F.; Empl, A.; Fasso, A.; Ferrari, A.; Gadioli, E.; Garzelli, M. V.; Lantz, M.; Liotta, M.; Mairani, A.; Mostacci, A.; Muraro, S.; Ottolenghi, A.; Pelliccioni, M.; Pinsky, L.; Ranft, J.; Roesler, S.; Sala, P. R.; Scannicchio, D.; Trovati, S.; Villari, R.; Wilson, T.
2006-01-01
FLUKA is a multipurpose Monte Carlo code which can transport a variety of particles over a wide energy range in complex geometries. The code is a joint project of INFN and CERN: part of its development is also supported by the University of Houston and NASA. FLUKA is successfully applied in several fields, including but not only, particle physics, cosmic ray physics, dosimetry, radioprotection, hadron therapy, space radiation, accelerator design and neutronics. The code is the standard tool used at CERN for dosimetry, radioprotection and beam-machine interaction studies. Here we give a glimpse into the code physics models with a particular emphasis to the hadronic and nuclear sector.
Ballarini, F.; Battistoni, G.; Campanella, M.; Carboni, M.; Cerutti, F.; Empl, A.; Fasso, A.; Ferrari, A.; Gadioli, E.; Garzelli, M.V.; Lantz, M.; Liotta, M.; Mairani, A.; Mostacci, A.; Muraro, S.; Ottolenghi, A.; Pelliccioni, M.; Pinsky, L.; Ranft, J.; Roesler, S.; Sala, P.R.; /Milan U. /INFN, Milan /Pavia U. /INFN, Pavia /CERN /Siegen U. /Houston U. /SLAC /Frascati /NASA, Houston /ENEA, Frascati
2005-11-09
FLUKA is a multipurpose Monte Carlo code which can transport a variety of particles over a wide energy range in complex geometries. The code is a joint project of INFN and CERN: part of its development is also supported by the University of Houston and NASA. FLUKA is successfully applied in several fields, including but not only, particle physics, cosmic ray physics, dosimetry, radioprotection, hadron therapy, space radiation, accelerator design and neutronics. The code is the standard tool used at CERN for dosimetry, radioprotection and beam-machine interaction studies. Here we give a glimpse into the code physics models with a particular emphasis to the hadronic and nuclear sector.
Boltzmann-Fokker-Planck calculations using standard discrete-ordinates codes
Morel, J.E.
1987-01-01
The Boltzmann-Fokker-Planck (BFP) equation can be used to describe both neutral and charged-particle transport. Over the past several years, the author and several collaborators have developed methods for representing Fokker-Planck operators with standard multigroup-Legendre cross-section data. When these data are input to a standard S/sub n/ code such as ONETRAN, the code actually solves the Boltzmann-Fokker-Planck equation rather than the Boltzmann equation. This is achieved wihout any modification to the S/sub n/ codes. Because BFP calculations can be more demanding from a numerical viewpoint than standard neutronics calculations, we have found it useful to implement new quadrature methods ad convergence acceleration methods in the standard discrete-ordinates code, ONETRAN. We discuss our BFP cross-section representation techniques, our improved quadrature and acceleration techniques, and present results from BFP coupled electron-photon transport calculations performed with ONETRAN. 19 refs., 7 figs.
Accelerator shield design of KIPT neutron source facility
Zhong, Z.; Gohar, Y.
2013-07-01
Argonne National Laboratory (ANL) of the United States and Kharkov Institute of Physics and Technology (KIPT) of Ukraine have been collaborating on the design development of a neutron source facility at KIPT utilizing an electron-accelerator-driven subcritical assembly. Electron beam power is 100 kW, using 100 MeV electrons. The facility is designed to perform basic and applied nuclear research, produce medical isotopes, and train young nuclear specialists. The biological shield of the accelerator building is designed to reduce the biological dose to less than 0.5-mrem/hr during operation. The main source of the biological dose is the photons and the neutrons generated by interactions of leaked electrons from the electron gun and accelerator sections with the surrounding concrete and accelerator materials. The Monte Carlo code MCNPX serves as the calculation tool for the shield design, due to its capability to transport electrons, photons, and neutrons coupled problems. The direct photon dose can be tallied by MCNPX calculation, starting with the leaked electrons. However, it is difficult to accurately tally the neutron dose directly from the leaked electrons. The neutron yield per electron from the interactions with the surrounding components is less than 0.01 neutron per electron. This causes difficulties for Monte Carlo analyses and consumes tremendous computation time for tallying with acceptable statistics the neutron dose outside the shield boundary. To avoid these difficulties, the SOURCE and TALLYX user subroutines of MCNPX were developed for the study. The generated neutrons are banked, together with all related parameters, for a subsequent MCNPX calculation to obtain the neutron and secondary photon doses. The weight windows variance reduction technique is utilized for both neutron and photon dose calculations. Two shielding materials, i.e., heavy concrete and ordinary concrete, were considered for the shield design. The main goal is to maintain the total
European Neutron Activation System.
2013-01-11
Version 03 EASY-2010 (European Activation System) consists of a wide range of codes, data and documentation all aimed at satisfying the objective of calculating the response of materials irradiated in a neutron flux. The main difference from the previous version is the upper energy limit, which has increased from 20 to 60 MeV. It is designed to investigate both fusion devices and accelerator based materials test facilities that will act as intense sources of high-energymore » neutrons causing significant activation of the surrounding materials. The very general nature of the calculational method and the data libraries means that it is applicable (with some reservations) to all situations (e.g. fission reactors or neutron sources) where materials are exposed to neutrons below 60 MeV. EASY can be divided into two parts: data and code development tools and user tools and data. The former are required to develop the latter, but EASY users only need to be able to use the inventory code FISPACT and be aware of the contents of the EAF library (the data source). The complete EASY package contains the FISPACT-2007 inventory code, the EAF-2003, EAF-2005, EAF-2007 and EAF-2010 libraries, and the EASY User Interface for the Window version. The activation package EASY-2010 is the result of significant development to extend the upper energy range from 20 to 60 MeV so that it is capable of being used for IFMIF calculations. The EAF-2010 library contains 66,256 reactions, almost five times more than in EAF-2003 (12,617). Deuteron-induced and proton-induced cross section libraries are also included, and can be used with EASY to enable calculations of the activation due to deuterons and proton [2].« less
An Improved Neutron Transport Algorithm for Space Radiation
NASA Technical Reports Server (NTRS)
Heinbockel, John H.; Clowdsley, Martha S.; Wilson, John W.
2000-01-01
A low-energy neutron transport algorithm for use in space radiation protection is developed. The algorithm is based upon a multigroup analysis of the straight-ahead Boltzmann equation by using a mean value theorem for integrals. This analysis is accomplished by solving a realistic but simplified neutron transport test problem. The test problem is analyzed by using numerical and analytical procedures to obtain an accurate solution within specified error bounds. Results from the test problem are then used for determining mean values associated with rescattering terms that are associated with a multigroup solution of the straight-ahead Boltzmann equation. The algorithm is then coupled to the Langley HZETRN code through the evaporation source term. Evaluation of the neutron fluence generated by the solar particle event of February 23, 1956, for a water and an aluminum-water shield-target configuration is then compared with LAHET and MCNPX Monte Carlo code calculations for the same shield-target configuration. The algorithm developed showed a great improvement in results over the unmodified HZETRN solution. In addition, a two-directional solution of the evaporation source showed even further improvement of the fluence near the front of the water target where diffusion from the front surface is important.
New precision measurements of free neutron beta decay with cold neutrons
Baeßler, Stefan; Bowman, James David; Penttilä, Seppo I.; Počanić, Dinko
2014-10-14
Precision measurements in free neutron beta decay serve to determine the coupling constants of beta decay, and offer several stringent tests of the standard model. This study describes the free neutron beta decay program planned for the Fundamental Physics Beamline at the Spallation Neutron Source at Oak Ridge National Laboratory, and finally puts it into the context of other recent and planned measurements of neutron beta decay observables.
Neutron Transport Characteristics of a Nuclear Reactor Based Dynamic Neutron Imaging System
Khaial, Anas M.; Harvel, Glenn D.; Chang, Jen-Shih
2006-07-01
An advanced dynamic neutron imaging system has been constructed in the McMaster Nuclear Reactor (MNR) for nondestructive testing and multi-phase flow studies in energy and environmental applications. A high quality neutron beam is required with a thermal neutron flux greater than 5.0 x 10{sup 6} n/cm{sup 2}-s and a collimation ratio of 120 at image plane to promote high-speed neutron imaging up to 2000 frames per second. Neutron source strength and neutron transport have been experimentally and numerically investigated. Neutron source strength at the beam tube entrance was evaluated experimentally by measuring the thermal and fast neutron fluxes, and simple analytical neutron transport calculations were performed based upon these measured neutron fluxes to predict facility components in accordance with high-speed dynamic neutron imaging and operation safety requirements. Monte-Carlo simulations (using MCNP-4B code) with multiple neutron energy groups have also been used to validate neutron beam parameters and to ensure shielding capabilities of facility shutter and cave walls. Neutron flux distributions at the image plane and the neutron beam characteristics were experimentally measured by irradiating a two-dimensional array of Copper foils and using a real-time neutron radiography system. The neutron image characteristics -- such as neutron flux, image size, beam quality -- measured experimentally and predicted numerically for beam tube, beam shutter and radiography cave are compared and discussed in detail in this paper. The experimental results show that thermal neutron flux at image plane is nearly uniform over an imaging area of 20.0-cm diameter and its magnitude ranges from 8.0 x 10{sup 6} - 1.0 x 10{sup 7} n/cm{sup 2}-sec while the neutron-to-gamma ratio is 6.0 x 10{sup 5} n/cm{sup 2}-{mu}Sv. (authors)
Neutron dosimetry in solid water phantom
Benites-Rengifo, Jorge Luis; Vega-Carrillo, Hector Rene
2014-11-07
The neutron spectra, the Kerma and the absorbed dose due to neutrons were estimated along the incoming beam in a solid water phantom. Calculations were carried out with the MCNP5 code, where the bunker, the phantom and the model of the15 MV LINAC head were modeled. As the incoming beam goes into the phantom the neutron spectrum is modified and the dosimetric values are reduced.
Neutron dosimetry in solid water phantom
NASA Astrophysics Data System (ADS)
Benites-Rengifo, Jorge Luis; Vega-Carrillo, Hector Rene
2014-11-01
The neutron spectra, the Kerma and the absorbed dose due to neutrons were estimated along the incoming beam in a solid water phantom. Calculations were carried out with the MCNP5 code, where the bunker, the phantom and the model of the15 MV LINAC head were modeled. As the incoming beam goes into the phantom the neutron spectrum is modified and the dosimetric values are reduced.
Lahl, H; Henke, G
1997-11-01
Neutron activation analysis (NAA) and inductively coupled plasma emission spectrometry (ICP-AES) were used to quantify the relative contents of Fe, Sc, Ce, Pa, Cr, Co, respectively the absolute contents of Cr, Zn, Mn, Fe, Mg, Al, Cu, Ti, Ca, Sr in hashish samples, seized in different countries. The samples were processed after dry ashing by means of instrumental NAA and after wet mineralization by means of ICP-AES. For determination of the sampling and measurement errors, one of the samples was analyzed repeatedly with both methods. Classifying hashish samples with regard to concentration of certain elements could be done by artificial neural networks with a modified backpropagation algorithm. By this way, identity and non identity of one unknown sample with one of many different samples as data pool can be ascertained, on principle. PMID:9446107
Bower, W.
1999-01-08
The National Electrical Code@ (NEC@) focuses primarily on electrical system installation requirements in the U.S. The NEC addresses both fire and personnel safety. This paper will describe recent efforts of the PV industry in the U.S. and the resulting requirements in the 1999 National Electrical Code-- Article 690 --Solar Photovoltaic Systems. The Article 690 requirements spell out the PV-unique requirements for safe installations of PV systems in the U.S.A. This paper provides an overview of the most significant changes that appear in Article 690 of the 1999 edition of the NEC. The related and coordinated efforts of the other standards- making groups will also be briefly reviewed.
Stephan, Andrew C.; Jardret; Vincent D.
2011-04-05
A neutron detector has a volume of neutron moderating material and a plurality of individual neutron sensing elements dispersed at selected locations throughout the moderator, and particularly arranged so that some of the detecting elements are closer to the surface of the moderator assembly and others are more deeply embedded. The arrangement captures some thermalized neutrons that might otherwise be scattered away from a single, centrally located detector element. Different geometrical arrangements may be used while preserving its fundamental characteristics. Different types of neutron sensing elements may be used, which may operate on any of a number of physical principles to perform the function of sensing a neutron, either by a capture or a scattering reaction, and converting that reaction to a detectable signal. High detection efficiency, an ability to acquire spectral information, and directional sensitivity may be obtained.
Stephan, Andrew C; Jardret, Vincent D
2009-04-07
A neutron detector has a volume of neutron moderating material and a plurality of individual neutron sensing elements dispersed at selected locations throughout the moderator, and particularly arranged so that some of the detecting elements are closer to the surface of the moderator assembly and others are more deeply embedded. The arrangement captures some thermalized neutrons that might otherwise be scattered away from a single, centrally located detector element. Different geometrical arrangements may be used while preserving its fundamental characteristics. Different types of neutron sensing elements may be used, which may operate on any of a number of physical principles to perform the function of sensing a neutron, either by a capture or a scattering reaction, and converting that reaction to a detectable signal. High detection efficiency, an ability to acquire spectral information, and directional sensitivity may be obtained.
Cramer, S.N.
1984-01-01
The KENO-V code is the current release of the Oak Ridge multigroup Monte Carlo criticality code development. The original KENO, with 16 group Hansen-Roach cross sections and P/sub 1/ scattering, was one ot the first multigroup Monte Carlo codes and it and its successors have always been a much-used research tool for criticality studies. KENO-V is able to accept large neutron cross section libraries (a 218 group set is distributed with the code) and has a general P/sub N/ scattering capability. A supergroup feature allows execution of large problems on small computers, but at the expense of increased calculation time and system input/output operations. This supergroup feature is activated automatically by the code in a manner which utilizes as much computer memory as is available. The primary purpose of KENO-V is to calculate the system k/sub eff/, from small bare critical assemblies to large reflected arrays of differing fissile and moderator elements. In this respect KENO-V neither has nor requires the many options and sophisticated biasing techniques of general Monte Carlo codes.
Research on Fast-Doppler-Broadening of neutron cross sections
Li, S.; Wang, K.; Yu, G.
2012-07-01
A Fast-Doppler-Broadening method is developed in this work to broaden Continuous Energy neutron cross-sections for Monte Carlo calculations. Gauss integration algorithm and parallel computing are implemented in this method, which is unprecedented in the history of cross section processing. Compared to the traditional code (NJOY, SIGMA1, etc.), the new Fast-Doppler-Broadening method shows a remarkable speedup with keeping accuracy. The purpose of using Gauss integration is to avoid complex derivation of traditional broadening formula and heavy load of computing complementary error function that slows down the Doppler broadening process. The OpenMP environment is utilized in parallel computing which can take full advantage of modern multi-processor computers. Combination of the two can reduce processing time of main actinides (such as {sup 238}U, {sup 235}U) to an order of magnitude of 1{approx}2 seconds. This new method is fast enough to be applied to Online Doppler broadening. It can be combined or coupled with Monte Carlo transport code to solve temperature dependent problems and neutronics-thermal hydraulics coupled scheme which is a big challenge for the conventional NJOY-MCNP system. Examples are shown to determine the efficiency and relative errors compared with the NJOY results. A Godiva Benchmark is also used in order to test the ACE libraries produced by the new method. (authors)
Neutron producing reactions in PuBe neutron sources
NASA Astrophysics Data System (ADS)
Bagi, János; Lakosi, László; Nguyen, Cong Tam
2016-01-01
There are a plenty of out-of-use plutonium-beryllium neutron sources in Eastern Europe presenting both nuclear safeguards and security issues. Typically, their actual Pu content is not known. In the last couple of years different non-destructive methods were developed for their characterization. For such methods detailed knowledge of the nuclear reactions taking place within the source is necessary. In this paper we investigate the role of the neutron producing reactions, their contribution to the neutron yield and their dependence on the properties of the source.
NASA Technical Reports Server (NTRS)
Preszler, A. M.; Moon, S.; White, R. S.
1976-01-01
Additional calibrations of the University of California double-scatter neutron detector and additional analysis corrections lead to slightly changed neutron fluxes. The theoretical angular distributions of Merker (1975) are in general agreement with the reported experimental fluxes but do not give the peaks for vertical upward and downward moving neutrons. The theoretical neutron escape current is in agreement with the experimental values from 10 to 100 MeV. The experimental fluxes obtained agree with those of Kanbach et al. (1974) in the overlap region from 70 to 100 MeV.
Neutron stars and strange matter
Cooperstein, J.
1986-01-01
The likelihood is investigated that quark matter with strangeness of order unity resides in neutron stars. In the strong coupling regime near rho/sub 0/ this is found to be unlikely. Considering higher densities where perturbative expansions are used, we find a lower bound to be at 7rho/sub 0/ for the transition density. This is higher than the inferred density of observed neutron stars, and thus the transition to quark matter is precluded. 15 refs., 3 figs.
Spectral unfolding of fast neutron energy distributions
NASA Astrophysics Data System (ADS)
Mosby, Michelle; Jackman, Kevin; Engle, Jonathan
2015-10-01
The characterization of the energy distribution of a neutron flux is difficult in experiments with constrained geometry where techniques such as time of flight cannot be used to resolve the distribution. The measurement of neutron fluxes in reactors, which often present similar challenges, has been accomplished using radioactivation foils as an indirect probe. Spectral unfolding codes use statistical methods to adjust MCNP predictions of neutron energy distributions using quantified radioactive residuals produced in these foils. We have applied a modification of this established neutron flux characterization technique to experimentally characterize the neutron flux in the critical assemblies at the Nevada National Security Site (NNSS) and the spallation neutron flux at the Isotope Production Facility (IPF) at Los Alamos National Laboratory (LANL). Results of the unfolding procedure are presented and compared with a priori MCNP predictions, and the implications for measurements using the neutron fluxes at these facilities are discussed.
Thermal neutron analysis (TNA) explosive detection based on electronic neutron generators
Lee, W.; Mahood, D.B.; Ryge, P.
1994-12-31
Thermal neutron analysis explosive detection systems have been developed and demonstrated for inspection of checked airline baggage and for detection of buried land mines. Thermal neutrons from a moderated neutron source impinge on the inspected object and the resulting capture gamma ray signatures provide detection information. Isotopic neutron sources, e.g. {sup 252}Cf, are compact, economical and reliable, but they are subject to the licensing requirements, safety concerns and public perception problems associated with radioactive material. These are mitigated by use of an electronic neutron generator - an ion accelerator with a target producing neutrons by a nuclear reaction such as D(d,n){sup 3}He or {sup 9}Be(d,n){sup 10}B. With suitable moderator designs based on neutron transport codes, operational explosive detection systems can be build and would provide effective alternatives to radioactive neutron sources. Calculations as well as laboratory and field experience with three generator types will be presented.
Neutron Imaging Developments at LANSCE
Nelson, Ronald Owen; Hunter, James F.; Schirato, Richard C.; Vogel, Sven C.; Swift, Alicia L.; Ickes, Timothy Lee; Ward, William Carl; Losko, Adrian Simon; Tremsin, Anton; Sevanto, Sanna Annika; Espy, Michelle A.; Dickman, Lee Thoresen; Malone, Michael
2015-10-29
Thermal, epithermal, and high-energy neutrons are available from two spallation sources at the 800 MeV proton accelerator. Improvements in detectors and computing have enabled new capabilities that use the pulsed beam properties at LANSCE; these include amorphous Si (aSi) detectors, intensified charge-coupled device cameras, and micro-channel plates. Applications include water flow in living specimens, inclusions and fission products in uranium oxide, and high-energy neutron imaging using an aSi flat panel with ZnS(Ag) scintillator screen. images of a metal/plastic cylinder from photons, low-energy and high-energy neutrons are compared.
Calculated analysis of experiments in fast neutron reactors
Davydov, V. K. Kalugina, K. M.; Gomin, E. A.
2012-12-15
In this paper, the results of computational simulation of experiments with the MK-I core of the JOYO fast neutron sodium-cooled reactor are presented. The MCU-KS code based on the Monte Carlo method was used for calculations. The research was aimed at additional verification of the MCU-KS code for systems with a fast neutron spectrum.
Conceptual moderator studies for the Spallation Neutron Source short-pulse second target station.
Gallmeier, F X; Lu, W; Riemer, B W; Zhao, J K; Herwig, K W; Robertson, J L
2016-06-01
Candidate moderator configurations for a short-pulse second target station (STS) at the Oak Ridge National Laboratory Spallation Neutron Source (SNS) have been identified using a global optimizer framework built around the MCNPX particle transport code. Neutron brightness metrics were selected as the figure-of-merit. We assumed that STS would use one out of six proton pulses produced by an SNS accelerator upgraded to operate at 1.3 GeV proton energy, 2.8 MW power and 60 Hz repetition rate. The simulations indicate that the peak brightness can be increased by a factor of 5 and 2.5 on a per proton pulse basis compared to the SNS first target station for both coupled and decoupled para-hydrogen moderators, respectively. Additional increases by factors of 3 and 2 were demonstrated for coupled and decoupled moderators, respectively, by reducing the area of neutron emission from 100 × 100 mm(2) to 20 × 20 mm(2). This increase in brightness has the potential to translate to an increase of beam intensity at the instruments' sample positions even though the total neutron emission of the smaller moderator is less than that of the larger. This is especially true for instruments with small samples (beam dimensions). The increased fluxes in the STS moderators come at accelerated poison and de-coupler burnout and higher radiation-induced material damage rates per unit power, which overall translate into lower moderator lifetimes. A first effort was undertaken to group decoupled moderators into a cluster collectively positioning them at the peak neutron production zone in the target and having a three-port neutron emission scheme that complements that of a cylindrical coupled moderator. PMID:27370444
Conceptual moderator studies for the Spallation Neutron Source short-pulse second target station.
Gallmeier, F X; Lu, W; Riemer, B W; Zhao, J K; Herwig, K W; Robertson, J L
2016-06-01
Candidate moderator configurations for a short-pulse second target station (STS) at the Oak Ridge National Laboratory Spallation Neutron Source (SNS) have been identified using a global optimizer framework built around the MCNPX particle transport code. Neutron brightness metrics were selected as the figure-of-merit. We assumed that STS would use one out of six proton pulses produced by an SNS accelerator upgraded to operate at 1.3 GeV proton energy, 2.8 MW power and 60 Hz repetition rate. The simulations indicate that the peak brightness can be increased by a factor of 5 and 2.5 on a per proton pulse basis compared to the SNS first target station for both coupled and decoupled para-hydrogen moderators, respectively. Additional increases by factors of 3 and 2 were demonstrated for coupled and decoupled moderators, respectively, by reducing the area of neutron emission from 100 × 100 mm(2) to 20 × 20 mm(2). This increase in brightness has the potential to translate to an increase of beam intensity at the instruments' sample positions even though the total neutron emission of the smaller moderator is less than that of the larger. This is especially true for instruments with small samples (beam dimensions). The increased fluxes in the STS moderators come at accelerated poison and de-coupler burnout and higher radiation-induced material damage rates per unit power, which overall translate into lower moderator lifetimes. A first effort was undertaken to group decoupled moderators into a cluster collectively positioning them at the peak neutron production zone in the target and having a three-port neutron emission scheme that complements that of a cylindrical coupled moderator.
Conceptual moderator studies for the Spallation Neutron Source short-pulse second target station
NASA Astrophysics Data System (ADS)
Gallmeier, F. X.; Lu, W.; Riemer, B. W.; Zhao, J. K.; Herwig, K. W.; Robertson, J. L.
2016-06-01
Candidate moderator configurations for a short-pulse second target station (STS) at the Oak Ridge National Laboratory Spallation Neutron Source (SNS) have been identified using a global optimizer framework built around the MCNPX particle transport code. Neutron brightness metrics were selected as the figure-of-merit. We assumed that STS would use one out of six proton pulses produced by an SNS accelerator upgraded to operate at 1.3 GeV proton energy, 2.8 MW power and 60 Hz repetition rate. The simulations indicate that the peak brightness can be increased by a factor of 5 and 2.5 on a per proton pulse basis compared to the SNS first target station for both coupled and decoupled para-hydrogen moderators, respectively. Additional increases by factors of 3 and 2 were demonstrated for coupled and decoupled moderators, respectively, by reducing the area of neutron emission from 100 × 100 mm2 to 20 × 20 mm2. This increase in brightness has the potential to translate to an increase of beam intensity at the instruments' sample positions even though the total neutron emission of the smaller moderator is less than that of the larger. This is especially true for instruments with small samples (beam dimensions). The increased fluxes in the STS moderators come at accelerated poison and de-coupler burnout and higher radiation-induced material damage rates per unit power, which overall translate into lower moderator lifetimes. A first effort was undertaken to group decoupled moderators into a cluster collectively positioning them at the peak neutron production zone in the target and having a three-port neutron emission scheme that complements that of a cylindrical coupled moderator.
Conceptual moderator studies for the Spallation Neutron Source short-pulse second target station
Gallmeier, F. X.; Lu, W.; Riemer, B. W.; Zhao, J. K.; Herwig, K. W.; Robertson, J. L.
2016-06-14
We identified candidate moderator configurations for a short-pulse second target station (STS) at the Oak Ridge National Laboratory Spallation Neutron Source (SNS) using a global optimizer framework built around the MCNPX particle transport code. Neutron brightness metrics were selected as the figure-of-merit. We assumed that STS would use one out of six proton pulses produced by an SNS accelerator upgraded to operate at 1.3 GeV proton energy, 2.8 MW power and 60 Hz repetition rate. The simulations indicate that the peak brightness can be increased by a factor of 5 and 2.5 on a per proton pulse basis compared tomore » the SNS first target station for both coupled and decoupled para-hydrogen moderators, respectively. Additional increases by factors of 3 and 2 were demonstrated for coupled and decoupled moderators, respectively, by reducing the area of neutron emission from 100 × 100 mm2 to 20 × 20 mm2. Furthermore, this increase in brightness has the potential to translate to an increase of beam intensity at the instruments’ sample positions even though the total neutron emission of the smaller moderator is less than that of the larger. This is especially true for instruments with small samples (beam dimensions). The increased fluxes in the STS moderators come at accelerated poison and de-coupler burnout and higher radiation-induced material damage rates per unit power, which overall translate into lower moderator lifetimes. Our first effort decoupled group moderators into a cluster collectively positioning them at the peak neutron production zone in the target and having a three-port neutron emission scheme that complements that of a cylindrical coupled moderator.« less
Monte Carlo simulation of neutron scattering instruments
Seeger, P.A.; Daemen, L.L.; Hjelm, R.P. Jr.
1998-12-01
A code package consisting of the Monte Carlo Library MCLIB, the executing code MC{_}RUN, the web application MC{_}Web, and various ancillary codes is proposed as an open standard for simulation of neutron scattering instruments. The architecture of the package includes structures to define surfaces, regions, and optical elements contained in regions. A particle is defined by its vector position and velocity, its time of flight, its mass and charge, and a polarization vector. The MC{_}RUN code handles neutron transport and bookkeeping, while the action on the neutron within any region is computed using algorithms that may be deterministic, probabilistic, or a combination. Complete versatility is possible because the existing library may be supplemented by any procedures a user is able to code. Some examples are shown.
Wende, Charles W. J.
1976-08-17
A safety rod for a nuclear reactor has an inner end portion having a gamma absorption coefficient and neutron capture cross section approximately equal to those of the adjacent shield, a central portion containing materials of high neutron capture cross section and an outer end portion having a gamma absorption coefficient at least equal to that of the adjacent shield.
Fermi, E.; Zinn, W.H.; Anderson, H.L.
1958-09-16
Means are presenied for increasing the reproduction ratio of a gaphite- moderated neutronic reactor by diminishing the neutron loss due to absorption or capture by gaseous impurities within the reactor. This means comprised of a fluid-tight casing or envelope completely enclosing the reactor and provided with a valve through which the casing, and thereby the reactor, may be evacuated of atmospheric air.
Hardware accelerated high performance neutron transport computation based on AGENT methodology
NASA Astrophysics Data System (ADS)
Xiao, Shanjie
The spatial heterogeneity of the next generation Gen-IV nuclear reactor core designs brings challenges to the neutron transport analysis. The Arbitrary Geometry Neutron Transport (AGENT) AGENT code is a three-dimensional neutron transport analysis code being developed at the Laboratory for Neutronics and Geometry Computation (NEGE) at Purdue University. It can accurately describe the spatial heterogeneity in a hierarchical structure through the R-function solid modeler. The previous version of AGENT coupled the 2D transport MOC solver and the 1D diffusion NEM solver to solve the three dimensional Boltzmann transport equation. In this research, the 2D/1D coupling methodology was expanded to couple two transport solvers, the radial 2D MOC solver and the axial 1D MOC solver, for better accuracy. The expansion was benchmarked with the widely applied C5G7 benchmark models and two fast breeder reactor models, and showed good agreement with the reference Monte Carlo results. In practice, the accurate neutron transport analysis for a full reactor core is still time-consuming and thus limits its application. Therefore, another content of my research is focused on designing a specific hardware based on the reconfigurable computing technique in order to accelerate AGENT computations. It is the first time that the application of this type is used to the reactor physics and neutron transport for reactor design. The most time consuming part of the AGENT algorithm was identified. Moreover, the architecture of the AGENT acceleration system was designed based on the analysis. Through the parallel computation on the specially designed, highly efficient architecture, the acceleration design on FPGA acquires high performance at the much lower working frequency than CPUs. The whole design simulations show that the acceleration design would be able to speedup large scale AGENT computations about 20 times. The high performance AGENT acceleration system will drastically shortening the
Compact ion chamber based neutron detector
Derzon, Mark S.; Galambos, Paul C.; Renzi, Ronald F.
2015-10-27
A directional neutron detector has an ion chamber formed in a dielectric material; a signal electrode and a ground electrode formed in the ion chamber; a neutron absorbing material filling the ion chamber; readout circuitry which is electrically coupled to the signal and ground electrodes; and a signal processor electrically coupled to the readout circuitry. The ion chamber has a pair of substantially planar electrode surfaces. The chamber pressure of the neutron absorbing material is selected such that the reaction particle ion trail length for neutrons absorbed by the neutron absorbing material is equal to or less than the distance between the electrode surfaces. The signal processor is adapted to determine a path angle for each absorbed neutron based on the rise time of the corresponding pulse in a time-varying detector signal.
Monte Carlo Neutronics and Thermal Hydraulics Analysis of Reactor Cores with Multilevel Grids
NASA Astrophysics Data System (ADS)
Bernnat, W.; Mattes, M.; Guilliard, N.; Lapins, J.; Zwermann, W.; Pasichnyk, I.; Velkov, K.
2014-06-01
Power reactors are composed of assemblies with fuel pin lattices or other repeated structures with several grid levels, which can be modeled in detail by Monte Carlo neutronics codes such as MCNP6 using corresponding lattice options, even for large cores. Except for fresh cores at beginning of life, there is a varying material distribution due to burnup in the different fuel pins. Additionally, for power states the fuel and moderator temperatures and moderator densities vary according to the power distribution and cooling conditions. Therefore, a coupling of the neutronics code with a thermal hydraulics code is necessary. Depending on the level of detail of the analysis, a very large number of cells with different materials and temperatures must be regarded. The assignment of different material properties to all elements of a multilevel grid is very elaborate and may exceed program limits if the standard input procedure is used. Therefore, an internal assignment is used which overrides uniform input parameters. The temperature dependency of continuous energy cross sections, probability tables for the unresolved resonance region and thermal neutron scattering laws is taken into account by interpolation, requiring only a limited number of data sets generated for different temperatures. The method is applied with MCNP6 and proven for several full core reactor models. For the coupling of MCNP6 with thermal hydraulics appropriate interfaces were developed for the GRS system code ATHLET for liquid coolant and the IKE thermal hydraulics code ATTICA-3D for gaseous coolant. Examples will be shown for different applications for PWRs with square and hexagonal lattices, fast reactors (SFR) with hexagonal lattices and HTRs with pebble bed and prismatic lattices.
Charles A. Wemple; Joshua J. Cogliati
2005-04-01
A univel geometry, neutral particle Monte Carlo transport code, written entirely in the Java programming language, is under development for medical radiotherapy applications. The code uses ENDF-VI based continuous energy cross section data in a flexible XML format. Full neutron-photon coupling, including detailed photon production and photonuclear reactions, is included. Charged particle equilibrium is assumed within the patient model so that detailed transport of electrons produced by photon interactions may be neglected. External beam and internal distributed source descriptions for mixed neutron-photon sources are allowed. Flux and dose tallies are performed on a univel basis. A four-tap, shift-register-sequence random number generator is used. Initial verification and validation testing of the basic neutron transport routines is underway. The searchlight problem was chosen as a suitable first application because of the simplicity of the physical model. Results show excellent agreement with analytic solutions. Computation times for similar numbers of histories are comparable to other neutron MC codes written in C and FORTRAN.
Leung, Ka-Ngo; Lou, Tak Pui; Reijonen, Jani
2008-03-11
A neutron tube or generator is based on a RF driven plasma ion source having a quartz or other chamber surrounded by an external RF antenna. A deuterium or mixed deuterium/tritium (or even just a tritium) plasma is generated in the chamber and D or D/T (or T) ions are extracted from the plasma. A neutron generating target is positioned so that the ion beam is incident thereon and loads the target. Incident ions cause D-D or D-T (or T-T) reactions which generate neutrons. Various embodiments differ primarily in size of the chamber and position and shape of the neutron generating target. Some neutron generators are small enough for implantation in the body. The target may be at the end of a catheter-like drift tube. The target may have a tapered or conical surface to increase target surface area.
Cason, J.L. Jr.; Shaw, C.B.
1975-10-21
A neutron source which is particularly useful for neutron radiography consists of a vessel containing a moderating media of relatively low moderating ratio, a flux trap including a moderating media of relatively high moderating ratio at the center of the vessel, a shell of depleted uranium dioxide surrounding the moderating media of relatively high moderating ratio, a plurality of guide tubes each containing a movable source of neutrons surrounding the flux trap, a neutron shield surrounding one part of each guide tube, and at least one collimator extending from the flux trap to the exterior of the neutron source. The shell of depleted uranium dioxide has a window provided with depleted uranium dioxide shutters for each collimator. Reflectors are provided above and below the flux trap and on the guide tubes away from the flux trap.
Covariance Evaluation Methodology for Neutron Cross Sections
Herman,M.; Arcilla, R.; Mattoon, C.M.; Mughabghab, S.F.; Oblozinsky, P.; Pigni, M.; Pritychenko, b.; Songzoni, A.A.
2008-09-01
We present the NNDC-BNL methodology for estimating neutron cross section covariances in thermal, resolved resonance, unresolved resonance and fast neutron regions. The three key elements of the methodology are Atlas of Neutron Resonances, nuclear reaction code EMPIRE, and the Bayesian code implementing Kalman filter concept. The covariance data processing, visualization and distribution capabilities are integral components of the NNDC methodology. We illustrate its application on examples including relatively detailed evaluation of covariances for two individual nuclei and massive production of simple covariance estimates for 307 materials. Certain peculiarities regarding evaluation of covariances for resolved resonances and the consistency between resonance parameter uncertainties and thermal cross section uncertainties are also discussed.
A fundamental study on hyper-thermal neutrons for neutron capture therapy.
Sakurai, Y; Kobayashi, T; Kanda, K
1994-12-01
The utilization of hyper-thermal neutrons, which have an energy spectrum with a Maxwellian distribution at a higher temperature than room temperature (300 K), was studied in order to improve the thermal neutron flux distribution at depth in a living body for neutron capture therapy. Simulation calculations were carried out using a Monte Carlo code 'MCNP-V3' in order to investigate the characteristics of hyper-thermal neutrons, i.e. (i) depth dependence of the neutron energy spectrum, and (ii) depth distribution of the reaction rate in a water phantom for materials with 1/v neutron absorption. It is confirmed that hyper-thermal neutron irradiation can improve the thermal neutron flux distribution in the deeper areas in a living body compared with thermal neutron irradiation. When hyper-thermal neutrons with a 3000 K Maxwellian distribution are incident on a body, the reaction rates of 1/v materials such as 14N, 10B etc are about twice that observed for incident thermal neutrons at 300 K, at a depth of 5 cm. The limit of the treatable depth for tumours having 30 ppm 10B is expected to be about 1.5 cm greater by utilizing hyper-thermal neutrons at 3000 K compared with the incidence of thermal neutrons at 300 K.
NASA Technical Reports Server (NTRS)
Pollara, Fabrizio; Hamkins, Jon; Dolinar, Sam; Andrews, Ken; Divsalar, Dariush
2006-01-01
This viewgraph presentation reviews uplink coding. The purpose and goals of the briefing are (1) Show a plan for using uplink coding and describe benefits (2) Define possible solutions and their applicability to different types of uplink, including emergency uplink (3) Concur with our conclusions so we can embark on a plan to use proposed uplink system (4) Identify the need for the development of appropriate technology and infusion in the DSN (5) Gain advocacy to implement uplink coding in flight projects Action Item EMB04-1-14 -- Show a plan for using uplink coding, including showing where it is useful or not (include discussion of emergency uplink coding).
Study of TFTR D-T neutron spectra using natural diamond detectors
Roquemore, A.L.; Krasilnikov, A.V., Gorelenkov, N.N.
1996-12-31
Three Natural Diamond Detector (NDD) based spectrometers have been used for neutron spectra measurement during Deuterium-Tritium (D-T) experiments using high power Neutral Beam Injection (NBI) and Ton Cyclotron Resonance Heating (ICRH) on the Tokamak Fusion Test Reactor (TFTR) in 1996. A 2-3 % energy resolution coupled with the high radiation resistance of NDDs (5 x 10{sup 14}n/cm{sup 2}) makes them ideal for measuring the D-T neutron spectra in the high radiation environment of TFTR tritium experiments. The compact size of the NDD made it possible to insert one of the detectors into one of the center channels of the TFTR multichannel neutron collimator to provide a vertical view perpendicular to the vessel midplane, Two other detectors were placed inside shields in TFTR test cell and provide measurements of the neutrons having angles of emission of 110- 180{degrees} and 60-12-{degrees} with respect to the direction of the plasma current. By using a 0.25 {mu}s shaping time of the Ortec 673 spectroscopy amplifier we were able to accumulate useful spectrometry data at count rates up to 1.5 x 10{sup 3} counts/sec. To model the D- T neutron spectra measured by each of three NDD`s the Neutron Source post TRANSP (NST) code and semi-analytical model were developed. A set of D-T and D-D plasmas is analyzed for the dynamics of D-T neutron spectral broadening for each of three NDD cones of view. The application of three NDD based D-T neutron -spectrometer array demonstrated the anisotropy of the ion distribution function. and provided a mature of the dynamics of the effective ion temperatures for each detector view, and determined the tangential velocity of resonant tritons during ICRH.
Neutron spectrometry with He-3 proportional counters
Manolopoulou, M.; Fragopoulou, M.; Stoulos, S.; Vagena, E.; Westmeier, W.; Zamani, M.
2011-07-01
Helium filled proportional counters are widely used in the field of neutron detection and spectrometry. In this work the response of a commercially available He-3 counter is studied experimentally and calculated with Monte Carlo for the neutron energy range from 230 keV up to about 7 MeV. The calculated response of the system is used to determine neutron yield energy distribution emitted from an extended {sup nat}U/Pb assembly irradiated with 1.6 GeV deuterons. The results are in acceptable agreement with the calculated neutron distribution with DCM-DEM code. (authors)
Rutqvist, J.; Tsang, C-F.
2003-09-01
This paper presents a fully coupled thermal-hydrological-mechanical analysis of FEBEX--a large underground heater test conducted in a bentonite and fractured rock system. System responses predicted by the numerical analysis--including temperature, moisture content, and bentonite-swelling stress--were compared to field measurements at sensors located in the bentonite. An overall good agreement between predicted and measured system responses shows that coupled thermal-hydrological-mechanical processes in a bentonite barrier are well represented by the numerical model. The most challenging aspect of this particular analysis was modeling of the bentonite's mechanical behavior, which at FEBEX turned out to be affected by gaps between prefabricated bentonite blocks. At FEBEX, the swelling pressure did not develop until a few months into the experiment when moisture swelling of bentonite blocks had closed the gaps completely. Moreover, the wetting of the bentonite took place uniformly from the rock and was not impacted by the permeability difference between the Lamprophyres dykes and surrounding rock.
Modular System for Neutronics Calculations of Fission Reactors, Fusion Blankets, and Other Systems.
1999-07-23
AUS is a neutronics code system which may be used for calculations of a wide range of fission reactors, fusion blankets and other neutron applications. The present version, AUS98, has a nuclear cross section library based on ENDF/B-VI and includes modules which provide for reactor lattice calculations, one-dimensional transport calculations, multi-dimensional diffusion calculations, cell and whole reactor burnup calculations, and flexible editing of results. Calculations of multi-region resonance shielding, coupled neutron and photon transport, energymore » deposition, fission product inventory and neutron diffusion are combined within the one code system. The major changes from the previous release, AUS87, are the inclusion of a cross-section library based on ENDF/B-VI, the addition of the POW3D multi-dimensional diffusion module, the addition of the MICBURN module for controlling whole reactor burnup calculations, and changes to the system as a consequence of moving from IBM mainframe computers to UNIX workstations.« less
Coupled simulation of the reactor core using CUPID/MASTER
Lee, J. R.; Cho, H. K.; Yoon, H. Y.; Jeong, J. J.
2012-07-01
The CUPID is a component-scale thermal hydraulics code which is aimed for the analysis of transient two-phase flows in nuclear reactor components such as the reactor vessel, steam generator, containment. This code adopts a three-dimensional, transient, two-phase and three-field model, and includes physical models and correlations of the interfacial mass, momentum, and energy transfer for the closure. In the present paper, a multi-physics simulation was performed by coupling CUPID with a three dimensional neutron kinetics code, MASTER. MASTER is merged into CUPID as a dynamic link library (DLL). The APR1400 reactor core during a control rod drop/ejection accident was simulated as an example by adopting a porous media approach to employ a fuel assembly. The following sections present the numerical modeling for the reactor core, coupling of the kinetics code, and the simulation results. And also, a preliminary study for multi-scale simulation between CUPID and system-scaled thermal hydraulics code, MARS will be introduced as well. (authors)
NASA Astrophysics Data System (ADS)
Gupta, Mayanak Kumar; Mittal, Ranjan; Zbiri, Mohamed; Singh, Ripandeep; Rols, Stephane; Schober, Helmut; Chaplot, Samrath Lal
2014-10-01
We have carried out an extensive phonon study on multiferroic GaFeO3 to elucidate its dynamical behavior. Inelastic neutron scattering measurements are performed over a wide temperature range, 150 to 1198 K. First principles lattice dynamical calculations are done for the sake of the analysis and interpretation of the observations. The comparison of the phonon spectra from magnetic and nonmagnetic calculations highlights pronounced differences. The energy range of the vibrational atomistic contributions of the Fe and O ions are found to differ significantly in the two calculation types. Therefore, magnetism induced by the active spin degrees of freedom of Fe cations plays a key role in stabilizing the structure and dynamics of GaFeO3. Moreover, the computed enthalpy in various phases of GaFeO3 is used to gain deeper insights into the high-pressure phase stability of this material. Further, the volume dependence of the phonon spectra is used to determine its thermal expansion behavior.
Extension of the CENTAR system simulation code to thermionic space nuclear reactors
NASA Astrophysics Data System (ADS)
Nassersharif, Bahram; Gaeta, Michael J.; Berge, Francoise; Guffee, Laura; Williams, Ken
The Code for Extended Nonlinear Transient Analysis of Extraterrestial Reactors (CENTAR) is a general-purpose reactor system simulation code capable of modeling coupled heat transfer, fluid flow, neutronic, and control in an arbitrary reactor system (or subsystem) configuration. CENTAR 4.0 has been enhanced to support thermionic solid-core systems as well as liquid-metal systems. Several new models have been added. The fuel model has been enhanced to support two-dimensional heat transfer with multiple material gap interfaces. Gaps may contain thermionic converters. A quasi-steady-state heat-pipe component model has been added to predict operating limits. The thermoelectric/electromagnetic (TEM) pump model in CENTAR has been extended to also model electromagnetic pumps. A thermionic energy conversion model has been added. Numerous other enhancements to code architecture, user interface, and I/O have also been added in version 4.0.
Bernander, N.K. et al.
1960-10-18
An apparatus is described for producing neutrons through target bombardment with deuterons. Deuterium gas is ionized by electron bombardment and the deuteron ions are accelerated through a magnetic field to collimate them into a continuous high intensity beam. The ion beam is directed against a deuteron pervious metal target of substantially the same nnaterial throughout to embed the deuterous therein and react them to produce neutrons. A large quantity of neutrons is produced in this manner due to the increased energy and quantity of ions bombarding the target.
High-resolution coupled physics solvers for analysing fine-scale nuclear reactor design problems
Mahadevan, Vijay S.; Merzari, Elia; Tautges, Timothy; Jain, Rajeev; Obabko, Aleksandr; Smith, Michael; Fischer, Paul
2014-01-01
An integrated multi-physics simulation capability for the design and analysis of current and future nuclear reactor models is being investigated, to tightly couple neutron transport and thermal-hydraulics physics under the SHARP framework. Over several years, high-fidelity, validated mono-physics solvers with proven scalability on petascale architectures have been developed independently. Based on a unified component-based architecture, these existing codes can be coupled with a mesh-data backplane and a flexible coupling-strategy-based driver suite to produce a viable tool for analysts. The goal of the SHARP framework is to perform fully resolved coupled physics analysis of a reactor on heterogeneous geometry, in order to reduce the overall numerical uncertainty while leveraging available computational resources. The coupling methodology and software interfaces of the framework are presented, along with verification studies on two representative fast sodium-cooled reactor demonstration problems to prove the usability of the SHARP framework. PMID:24982250
High-resolution coupled physics solvers for analysing fine-scale nuclear reactor design problems.
Mahadevan, Vijay S; Merzari, Elia; Tautges, Timothy; Jain, Rajeev; Obabko, Aleksandr; Smith, Michael; Fischer, Paul
2014-08-01
An integrated multi-physics simulation capability for the design and analysis of current and future nuclear reactor models is being investigated, to tightly couple neutron transport and thermal-hydraulics physics under the SHARP framework. Over several years, high-fidelity, validated mono-physics solvers with proven scalability on petascale architectures have been developed independently. Based on a unified component-based architecture, these existing codes can be coupled with a mesh-data backplane and a flexible coupling-strategy-based driver suite to produce a viable tool for analysts. The goal of the SHARP framework is to perform fully resolved coupled physics analysis of a reactor on heterogeneous geometry, in order to reduce the overall numerical uncertainty while leveraging available computational resources. The coupling methodology and software interfaces of the framework are presented, along with verification studies on two representative fast sodium-cooled reactor demonstration problems to prove the usability of the SHARP framework. PMID:24982250
High-resolution coupled physics solvers for analysing fine-scale nuclear reactor design problems.
Mahadevan, Vijay S; Merzari, Elia; Tautges, Timothy; Jain, Rajeev; Obabko, Aleksandr; Smith, Michael; Fischer, Paul
2014-08-01
An integrated multi-physics simulation capability for the design and analysis of current and future nuclear reactor models is being investigated, to tightly couple neutron transport and thermal-hydraulics physics under the SHARP framework. Over several years, high-fidelity, validated mono-physics solvers with proven scalability on petascale architectures have been developed independently. Based on a unified component-based architecture, these existing codes can be coupled with a mesh-data backplane and a flexible coupling-strategy-based driver suite to produce a viable tool for analysts. The goal of the SHARP framework is to perform fully resolved coupled physics analysis of a reactor on heterogeneous geometry, in order to reduce the overall numerical uncertainty while leveraging available computational resources. The coupling methodology and software interfaces of the framework are presented, along with verification studies on two representative fast sodium-cooled reactor demonstration problems to prove the usability of the SHARP framework.
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.
Thermal neutron detection system
Peurrung, Anthony J.; Stromswold, David C.
2000-01-01
According to the present invention, a system for measuring a thermal neutron emission from a neutron source, has a reflector/moderator proximate the neutron source that reflects and moderates neutrons from the neutron source. The reflector/moderator further directs thermal neutrons toward an unmoderated thermal neutron detector.
Analysis of active neutron multiplicity data for Y-12 skull oxide samples
Krick, M.S.; Ensslin, N.; Ceo, R.N.; May, P.K.
1996-09-01
Previous work on active neutron multiplicity measurements and analyses is summarized. New active multiplicity measurements are described for samples of Y-12 skull oxide using an Active Well Coincidence Counter and MSR4 multiplicity electronics. Neutron multiplication values for the samples were determined from triples/doubles ratios. Neutron multiplication values were also obtained from Monte Carlo calculations using the MCNP code and the results compared with the experimental values. A calibration curve of AmLi source-sample coupling vs neutron multiplication was determined and used for active multiplicity assay of the skull oxides. The results are compared with those obtained from assay with the conventional calibration-curve technique, where the doubles rate is calibrated vs the {sup 235}U mass. The coupling-multiplication relationship determined for the skull oxides is compared with that determined earlier for pure high-enrichment uranium metal and pure uranium oxide. Conclusions are drawn about the application of active multiplicity techniques to uranium assay. Additional active multiplicity measurements and calculations are recommended.
Cadmium Depletion Impacts on Hardening Neutron6 Spectrum for Advanced Fuel Testing in ATR
Gray S. Chang
2011-05-01
For transmuting long-lived isotopes contained in spent nuclear fuel into shorter-lived fission products effectively is in a fast neutron spectrum reactor. In the absence of a fast spectrum test reactor in the United States of America (USA), initial irradiation testing of candidate fuels can be performed in a thermal test reactor that has been modified to produce a test region with a hardened neutron spectrum. A test region is achieved with a Cadmium (Cd) filter which can harden the neutron spectrum to a spectrum similar (although still somewhat softer) to that of the liquid metal fast breeder reactor (LMFBR). A fuel test loop with a Cd-filter has been installed within the East Flux Trap (EFT) of the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL). A detailed comparison analyses between the cadmium (Cd) filter hardened neutron spectrum in the ATR and the LMFBR fast neutron spectrum have been performed using MCWO. MCWO is a set of scripting tools that are used to couple the Monte Carlo transport code MCNP with the isotope depletion and buildup code ORIGEN-2.2. The MCWO-calculated results indicate that the Cd-filter can effectively flatten the Rim-Effect and reduce the linear heat rate (LHGR) to meet the advanced fuel testing project requirements at the beginning of irradiation (BOI). However, the filtering characteristics of Cd as a strong absorber quickly depletes over time, and the Cd-filter must be replaced for every two typical operating cycles within the EFT of the ATR. The designed Cd-filter can effectively depress the LHGR in experimental fuels and harden the neutron spectrum enough to adequately flatten the Rim Effect in the test region.
Kubilius, Jonas
2014-01-01
Sharing code is becoming increasingly important in the wake of Open Science. In this review I describe and compare two popular code-sharing utilities, GitHub and Open Science Framework (OSF). GitHub is a mature, industry-standard tool but lacks focus towards researchers. In comparison, OSF offers a one-stop solution for researchers but a lot of functionality is still under development. I conclude by listing alternative lesser-known tools for code and materials sharing.
Wade, E.J.
1958-09-16
This patent relates to a reflector means for a neutronic reactor. A reflector comprised of a plurality of vertically movable beryllium control members is provided surrounding the sides of the reactor core. An absorber of fast neutrons comprised of natural uramum surrounds the reflector. An absorber of slow neutrons surrounds the absorber of fast neutrons and is formed of a plurality of beryllium blocks having natural uranium members distributcd therethrough. in addition, a movable body is positioned directly below the core and is comprised of a beryllium reflector and an absorbing member attached to the botiom thereof, the absorbing member containing a substance selected from the goup consisting of natural urantum and Th/sup 232/.
Fraas, A.P.; Mills, C.B.
1961-11-21
A neutronic reactor in which neutron moderation is achieved primarily in its reflector is described. The reactor structure consists of a cylindrical central "island" of moderator and a spherical moderating reflector spaced therefrom, thereby providing an annular space. An essentially unmoderated liquid fuel is continuously passed through the annular space and undergoes fission while contained therein. The reactor, because of its small size, is particularly adapted for propulsion uses, including the propulsion of aircraft. (AEC)
Richmond, J.L.; Wells, C.E.
1963-01-15
A neutron source is obtained without employing any separate beryllia receptacle, as was formerly required. The new method is safer and faster, and affords a source with both improved yield and symmetry of neutron emission. A Be container is used to hold and react with Pu. This container has a thin isolating layer that does not obstruct the desired Pu--Be reaction and obviates procedures previously employed to disassemble and remove a beryllia receptacle. (AEC)
Microstructured silicon neutron detectors for security applications
NASA Astrophysics Data System (ADS)
Esteban, S.; Fleta, C.; Guardiola, C.; Jumilla, C.; Pellegrini, G.; Quirion, D.; Rodriguez, J.; Lozano, M.
2014-12-01
In this paper we present the design and performance of a perforated thermal neutron silicon detector with a 6LiF neutron converter. This device was manufactured within the REWARD project workplace whose aim is to develop and enhance technologies for the detection of nuclear and radiological materials. The sensor perforated structure results in a higher efficiency than that obtained with an equivalent planar sensor. The detectors were tested in a thermal neutron beam at the nuclear reactor at the Instituto Superior Técnico in Lisbon and the intrinsic detection efficiency for thermal neutrons and the gamma sensitivity were obtained. The Geant4 Monte Carlo code was used to simulate the experimental conditions, i.e. thermal neutron beam and the whole detector geometry. An intrinsic thermal neutron detection efficiency of 8.6%±0.4% with a discrimination setting of 450 keV was measured.
Coupled-Channel Models of Direct-Semidirect Capture via Giant-Dipole Resonances
NASA Astrophysics Data System (ADS)
Thompson, I. J.; Escher, J. E.; Arbanas, G.
2014-04-01
Semidirect capture, a two-step process that excites a giant-dipole resonance followed by its radiative de-excitation, is a dominant process near giant-dipole resonances, that is, for incoming neutron energies within 5-20 MeV. At lower energies such processes may affect neutron capture rates that are relevant to astrophysical nucleosynthesis models. We implement a semidirect capture model in the coupled-channel reaction code Fresco and validate it by comparing the cross section for direct-semidirect capture 208Pb(n,γ)209Pb to experimental data. We also investigate the effect of low-energy electric dipole strength in the pygmy resonance. We use a conventional single-particle direct-semidirect capture code Cupido for comparison. Furthermore, we present and discuss our results for direct-semidirect capture reaction 130Sn(n,γ)131Sn, the cross section of which is known to have a significant effect on nucleosynthesis models.
Coupled-Channel Models of Direct-Semidirect Capture via Giant-Dipole Resonances
Thompson, I J; Escher, Jutta E; Arbanas, Goran
2013-01-01
Semidirect capture, a two-step process that excites a giant-dipole resonance followed by its radiative de-excitation, is a dominant process near giant-dipole resonances, that is, for incoming neutron energies within 5 20 MeV. At lower energies such processes may affect neutron capture rates that are relevant to astrophysical nucleosynthesis models. We implement a semidirect capture model in the coupled-channel reaction code Fresco and validate it by comparing the cross section for direct-semidirect capture 208Pb(n,g)209Pb to experimental data. We also investigate the effect of low-energy electric dipole strength in the pygmy resonance. We use a conventional single-particle direct-semidirect capture code Cupido for comparison. Furthermore, we present and discuss our results for direct-semidirect capture reaction 130Sn(n,g)131Sn, the cross section of which is known to have a significant effect on nucleosynthesis models.
Multiple component codes based generalized LDPC codes for high-speed optical transport.
Djordjevic, Ivan B; Wang, Ting
2014-07-14
A class of generalized low-density parity-check (GLDPC) codes suitable for optical communications is proposed, which consists of multiple local codes. It is shown that Hamming, BCH, and Reed-Muller codes can be used as local codes, and that the maximum a posteriori probability (MAP) decoding of these local codes by Ashikhmin-Lytsin algorithm is feasible in terms of complexity and performance. We demonstrate that record coding gains can be obtained from properly designed GLDPC codes, derived from multiple component codes. We then show that several recently proposed classes of LDPC codes such as convolutional and spatially-coupled codes can be described using the concept of GLDPC coding, which indicates that the GLDPC coding can be used as a unified platform for advanced FEC enabling ultra-high speed optical transport. The proposed class of GLDPC codes is also suitable for code-rate adaption, to adjust the error correction strength depending on the optical channel conditions.
Development of a parallelization strategy for the VARIANT code
Hanebutte, U.R.; Khalil, H.S.; Palmiotti, G.; Tatsumi, M.
1996-12-31
The VARIANT code solves the multigroup steady-state neutron diffusion and transport equation in three-dimensional Cartesian and hexagonal geometries using the variational nodal method. VARIANT consists of four major parts that must be executed sequentially: input handling, calculation of response matrices, solution algorithm (i.e. inner-outer iteration), and output of results. The objective of the parallelization effort was to reduce the overall computing time by distributing the work of the two computationally intensive (sequential) tasks, the coupling coefficient calculation and the iterative solver, equally among a group of processors. This report describes the code`s calculations and gives performance results on one of the benchmark problems used to test the code. The performance analysis in the IBM SPx system shows good efficiency for well-load-balanced programs. Even for relatively small problem sizes, respectable efficiencies are seen for the SPx. An extension to achieve a higher degree of parallelism will be addressed in future work. 7 refs., 1 tab.
Wigner, E.P.
1958-04-22
A nuclear reactor for isotope production is described. This reactor is designed to provide a maximum thermal neutron flux in a region adjacent to the periphery of the reactor rather than in the center of the reactor. The core of the reactor is generally centrally located with respect tn a surrounding first reflector, constructed of beryllium. The beryllium reflector is surrounded by a second reflector, constructed of graphite, which, in tune, is surrounded by a conventional thermal shield. Water is circulated through the core and the reflector and functions both as a moderator and a coolant. In order to produce a greatsr maximum thermal neutron flux adjacent to the periphery of the reactor rather than in the core, the reactor is designed so tbat the ratio of neutron scattering cross section to neutron absorption cross section averaged over all of the materials in the reflector is approximately twice the ratio of neutron scattering cross section to neutron absorption cross section averaged over all of the material of the core of the reactor.
IllinoisGRMHD: an open-source, user-friendly GRMHD code for dynamical spacetimes
NASA Astrophysics Data System (ADS)
Etienne, Zachariah B.; Paschalidis, Vasileios; Haas, Roland; Mösta, Philipp; Shapiro, Stuart L.
2015-09-01
In the extreme violence of merger and mass accretion, compact objects like black holes and neutron stars are thought to launch some of the most luminous outbursts of electromagnetic and gravitational wave energy in the Universe. Modeling these systems realistically is a central problem in theoretical astrophysics, but has proven extremely challenging, requiring the development of numerical relativity codes that solve Einstein's equations for the spacetime, coupled to the equations of general relativistic (ideal) magnetohydrodynamics (GRMHD) for the magnetized fluids. Over the past decade, the Illinois numerical relativity (ILNR) group's dynamical spacetime GRMHD code has proven itself as a robust and reliable tool for theoretical modeling of such GRMHD phenomena. However, the code was written ‘by experts and for experts’ of the code, with a steep learning curve that would severely hinder community adoption if it were open-sourced. Here we present IllinoisGRMHD, which is an open-source, highly extensible rewrite of the original closed-source GRMHD code of the ILNR group. Reducing the learning curve was the primary focus of this rewrite, with the goal of facilitating community involvement in the code's use and development, as well as the minimization of human effort in generating new science. IllinoisGRMHD also saves computer time, generating roundoff-precision identical output to the original code on adaptive-mesh grids, but nearly twice as fast at scales of hundreds to thousands of cores.
Ion chamber based neutron detectors
Derzon, Mark S; Galambos, Paul C; Renzi, Ronald F
2014-12-16
A neutron detector with monolithically integrated readout circuitry, including: a bonded semiconductor die; an ion chamber formed in the bonded semiconductor die; a first electrode and a second electrode formed in the ion chamber; a neutron absorbing material filling the ion chamber; and the readout circuitry which is electrically coupled to the first and second electrodes. The bonded semiconductor die includes an etched semiconductor substrate bonded to an active semiconductor substrate. The readout circuitry is formed in a portion of the active semiconductor substrate. The ion chamber has a substantially planar first surface on which the first electrode is formed and a substantially planar second surface, parallel to the first surface, on which the second electrode is formed. The distance between the first electrode and the second electrode may be equal to or less than the 50% attenuation length for neutrons in the neutron absorbing material filling the ion chamber.
Design and simulations of the neutron dump for the back-streaming white neutron beam at CSNS
NASA Astrophysics Data System (ADS)
Zhang, L. Y.; Jing, H. T.; Tang, J. Y.; Wang, X. Q.
2016-10-01
For nuclear data measurements with a white neutron source, to control the background at the detector is a key issue. The neutron dump which locates at the end of the white neutron beam line at CSNS has a very important impact to the neutron and gamma backgrounds in the endstation. A sophisticated neutron dump was designed to reduce the backgrounds to the level of about 10-8 relative to the neutron flux. In this paper, the method to suppress both neutron and gamma backgrounds near a white-spectrum neutron dump is introduced. The optimized geometry structure and materials of the dump are described, and the neutron and gamma space distributions have been calculated by using the FLUKA code for different operation settings which are defined by beam spots of Φ30 mm, Φ60 mm and 90 mm×90 mm, respectively.
Risner, J.M.; Wiarda, D.; Miller, T.M.; Peplow, D.E.; Patton, B.W.; Dunn, M.E.; Parks, B.T.
2011-07-01
The U.S. Nuclear Regulatory Commission's Regulatory Guide 1.190 states that calculational methods used to estimate reactor pressure vessel (RPV) fluence should use the latest version of the evaluated nuclear data file (ENDF). The VITAMIN-B6 fine-group library and BUGLE-96 broad-group library, which are widely used for RPV fluence calculations, were generated using ENDF/B-VI.3 data, which was the most current data when Regulatory Guide 1.190 was issued. We have developed new fine-group (VITAMIN-B7) and broad-group (BUGLE-B7) libraries based on ENDF/B-VII.0. These new libraries, which were processed using the AMPX code system, maintain the same group structures as the VITAMIN-B6 and BUGLE-96 libraries. Verification and validation of the new libraries were accomplished using diagnostic checks in AMPX, 'unit tests' for each element in VITAMIN-B7, and a diverse set of benchmark experiments including critical evaluations for fast and thermal systems, a set of experimental benchmarks that are used for SCALE regression tests, and three RPV fluence benchmarks. The benchmark evaluation results demonstrate that VITAMIN-B7 and BUGLE-B7 are appropriate for use in RPV fluence calculations and meet the calculational uncertainty criterion in Regulatory Guide 1.190. (authors)
Risner, Joel M; Wiarda, Dorothea; Miller, Thomas Martin; Peplow, Douglas E.; Patton, Bruce W; Dunn, Michael E; Parks, Benjamin T
2011-01-01
The U.S. Nuclear Regulatory Commission s Regulatory Guide 1.190 states that calculational methods used to estimate reactor pressure vessel (RPV) fluence should use the latest version of the Evaluated Nuclear Data File (ENDF). The VITAMIN-B6 fine-group library and BUGLE-96 broad-group library, which are widely used for RPV fluence calculations, were generated using ENDF/B-VI data, which was the most current data when Regulatory Guide 1.190 was issued. We have developed new fine-group (VITAMIN-B7) and broad-group (BUGLE-B7) libraries based on ENDF/B-VII. These new libraries, which were processed using the AMPX code system, maintain the same group structures as the VITAMIN-B6 and BUGLE-96 libraries. Verification and validation of the new libraries was accomplished using diagnostic checks in AMPX, unit tests for each element in VITAMIN-B7, and a diverse set of benchmark experiments including critical evaluations for fast and thermal systems, a set of experimental benchmarks that are used for SCALE regression tests, and three RPV fluence benchmarks. The benchmark evaluation results demonstrate that VITAMIN-B7 and BUGLE-B7 are appropriate for use in LWR shielding applications, and meet the calculational uncertainty criterion in Regulatory Guide 1.190.
Elizondo-Decanini, Juan M.
2016-08-02
Short pulse neutron generators are described herein. In a general embodiment, the short pulse neutron generator includes a Blumlein structure. The Blumlein structure includes a first conductive plate, a second conductive plate, a third conductive plate, at least one of an inductor or a resistor, a switch, and a dielectric material. The first conductive plate is positioned relative to the second conductive plate such that a gap separates these plates. A vacuum chamber is positioned in the gap, and an ion source is positioned to emit ions in the vacuum chamber. The third conductive plate is electrically grounded, and the switch is operable to electrically connect and disconnect the second conductive plate and the third conductive plate. The at least one of the resistor or the inductor is coupled to the first conductive plate and the second conductive plate.
NASA Astrophysics Data System (ADS)
Sutton, Michele Rhea
2001-12-01
Fluence-to-dose conversion coefficients for the radiation protection quantity effective dose were calculated for neutrons, photons and protons with energies up to 2 GeV using the MCNPX code. The calculations were performed using the Pacific Northwest National Laboratory versions of the MIRD-V male and female anthropomorphic phantoms modified to include the skin and esophagus. The latest high-energy neutron evaluated cross-section libraries and the recommendations given in ICRP Publication 60 and ICRP Publication 74 were utilized to perform the calculations. Sets of fluence-to- effective dose conversion coefficients are given for anterior-posterior, posterior-anterior, left-lateral, right-lateral and rotational irradiation geometries. This is the first set of dose conversion coefficients over this energy range calculated for the L-LAT irradiation geometry. A unique set of high-energy neutron depth-dose benchmark experiments were performed at the Los Alamos Neutron Science Center/Weapons Neutron Research (LANSCE/WNR) complex. The experiments consisted of filtered neutron beams with energies up to 800 MeV impinging on a 30 x 30 x 30 cm3 tissue-equivalent phantom. The absorbed dose was measured in the phantom at various depths with tissue-equivalent ion chambers. The phantom and the experimental set-up were modeled using MCNPX. Comparisons of the experimental and computational depth- dose distributions indicate that the absorbed dose calculated by MCNPX is within 13% for neutrons with energies up to 750 MeV. This experiment will serve as a benchmark experiment for the testing of high-energy radiation transport codes for the international radiation protection community.
Moridis, George; Freeman, Craig
2013-09-30
We developed two new EOS additions to the TOUGH+ family of codes, the RealGasH2O and RealGas . The RealGasH2O EOS option describes the non-isothermal two-phase flow of water and a real gas mixture in gas reservoirs, with a particular focus in ultra-tight (such as tight-sand and shale gas) reservoirs. The gas mixture is treated as either a single-pseudo-component having a fixed composition, or as a multicomponent system composed of up to 9 individual real gases. The RealGas option has the same general capabilities, but does not include water, thus describing a single-phase, dry-gas system. In addition to the standard capabilities of all members of the TOUGH+ family of codes (fully-implicit, compositional simulators using both structured and unstructured grids), the capabilities of the two codes include: coupled flow and thermal effects in porous and/or fractured media, real gas behavior, inertial (Klinkenberg) effects, full micro-flow treatment, Darcy and non-Darcy flow through the matrix and fractures of fractured media, single- and multi-component gas sorption onto the grains of the porous media following several isotherm options, discrete and fracture representation, complex matrix-fracture relationships, and porosity-permeability dependence on pressure changes. The two options allow the study of flow and transport of fluids and heat over a wide range of time frames and spatial scales not only in gas reservoirs, but also in problems of geologic storage of greenhouse gas mixtures, and of geothermal reservoirs with multi-component condensable (H2O and CH4) and non-condensable gas mixtures. The codes are verified against available analytical and semi-analytical solutions. Their capabilities are demonstrated in a series of problems of increasing complexity, ranging from isothermal flow in simpler 1D and 2D conventional gas reservoirs, to non-isothermal gas flow in 3D fractured shale gas reservoirs involving 4 types of fractures, micro-flow, non-Darcy flow and gas
NASA Astrophysics Data System (ADS)
Moridis, George J.; Freeman, Craig M.
2014-04-01
We developed two new EOS additions to the TOUGH+ family of codes, the RealGasH2O and RealGas. The RealGasH2O EOS option describes the non-isothermal two-phase flow of water and a real gas mixture in gas reservoirs, with a particular focus in ultra-tight (such as tight-sand and shale gas) reservoirs. The gas mixture is treated as either a single-pseudo-component having a fixed composition, or as a multicomponent system composed of up to 9 individual real gases. The RealGas option has the same general capabilities, but does not include water, thus describing a single-phase, dry-gas system. In addition to the standard capabilities of all members of the TOUGH+ family of codes (fully-implicit, compositional simulators using both structured and unstructured grids), the capabilities of the two codes include coupled flow and thermal effects in porous and/or fractured media, real gas behavior, inertial (Klinkenberg) effects, full micro-flow treatment, Darcy and non-Darcy flow through the matrix and fractures of fractured media, single- and multi-component gas sorption onto the grains of the porous media following several isotherm options, discrete and fracture representation, complex matrix-fracture relationships, and porosity-permeability dependence on pressure changes. The two options allow the study of flow and transport of fluids and heat over a wide range of time frames and spatial scales not only in gas reservoirs, but also in problems of geologic storage of greenhouse gas mixtures, and of geothermal reservoirs with multi-component condensable (H2O and CH4) and non-condensable gas mixtures. The codes are verified against available analytical and semi-analytical solutions. Their capabilities are demonstrated in a series of problems of increasing complexity, ranging from isothermal flow in simpler 1D and 2D conventional gas reservoirs, to non-isothermal gas flow in 3D fractured shale gas reservoirs involving 4 types of fractures, micro-flow, non-Darcy flow and gas
NASA Astrophysics Data System (ADS)
Singh, Anar; Senyshyn, Anatoliy; Fuess, Hartmut; Chatterji, Tapan; Pandey, Dhananjai
2011-02-01
We report here the results of a high-resolution neutron powder diffraction study on the multiferroic solid solution system (Bi0.8Ba0.2)(Fe0.8Ti0.2)O3 in the temperature range 4 to 700 K. Using irreducible representation theory to analyze the magnetic structure by Rietveld refinement, we show that the magnetic structure is collinear G-type antiferromagnetic. Further, we confirm the occurrence of an isostructural phase transition (IPT) accompanying the magnetic ordering around ˜625 K in (Bi0.8Ba0.2)(Fe0.8Ti0.2)O3. It is shown that as a result of the IPT, the positions of all the atoms change significantly in the magnetically ordered phase, leading to an excess polarization which scales linearly with the sublattice magnetization obtained by Rietveld refinement of the magnetic structure. Structural evidence for magnetoelastic coupling for the magnetic transitions below room temperature is also presented.
Code System to Calculate Integral Parameters with Reaction Rates from WIMS Output.
1994-10-25
Version 00 REACTION calculates different integral parameters related to neutron reactions on reactor lattices, from reaction rates calculated with WIMSD4 code, and comparisons with experimental values.
High flux compact neutron generators
Reijonen, J.; Lou, T.-P.; Tolmachoff, B.; Leung, K.-N.; Verbeke, J.; Vujic, J.
2001-06-15
Compact high flux neutron generators are developed at the Lawrence Berkeley National Laboratory. The neutron production is based on D-D or D-T reaction. The deuterium or tritium ions are produced from plasma using either a 2 MHz or 13.56 MHz radio frequency (RF) discharge. RF-discharge yields high fraction of atomic species in the beam which enables higher neutron output. In the first tube design, the ion beam is formed using a multiple hole accelerator column. The beam is accelerated to energy of 80 keV by means of a three-electrode extraction system. The ion beam then impinges on a titanium target where either the 2.4 MeV D-D or 14 MeV D-T neutrons are generated. The MCNP computation code has predicted a neutron flux of {approximately}10{sup 11} n/s for the D-D reaction at beam intensity of 1.5 A at 150 kV. The neutron flux measurements of this tube design will be presented. Recently new compact high flux tubes are being developed which can be used for various applications. These tubes also utilize RF-discharge for plasma generation. The design of these tubes and the first measurements will be discussed in this presentation.
Neutron detector resolution for scattering
Kolda, S.A.
1997-03-01
A resolution function has been determined for scattered neutron experiments at Rensselaer Polytechnic Institute (RPI). This function accounts for the shifting and broadening of the resonance peak due to the additional path length, traveled by the neutron after scattering and prior to detection, along with the broadening of the resonance peak due to the bounce target. This resolution function has been parameterized both in neutron energy and size of the sample disk. Monte Carlo Neutron and Photon (MCNP) modeling has been used to determine the shape of the detector resolution function while assuming that the sample nucleus has an infinite mass. The shape of the function for a monoenergetic neutron point source has been compared to the analytical solution. Additionally, the parameterized detector resolution function has been used to broaden the scatter yield calculated from Evaluated Neutron Data File ENDF/B-VI cross section data for {sup 238}U. The target resolution function has been empirically determined by comparison of the broadened scatter yield and the experimental yield for {sup 238}U. The combined resolution function can be inserted into the SAMMY code to allow resonance analysis for scattering measurements.
a Portable Pulsed Neutron Generator
NASA Astrophysics Data System (ADS)
Skoulakis, A.; Androulakis, G. C.; Clark, E. L.; Hassan, S. M.; Lee, P.; Chatzakis, J.; Bakarezos, M.; Dimitriou, V.; Petridis, C.; Papadogiannis, N. A.; Tatarakis, M.
2014-02-01
The design and construction of a pulsed plasma focus device to be used as a portable neutron source for material analysis such as explosive detection using gamma spectroscopy is presented. The device is capable of operating at a repetitive rate of a few Hz. When deuterium gas is used, up to 105 neutrons per shot are expected to be produced with a temporal pulse width of a few tens of nanoseconds. The pulsed operation of the device and its portable size are its main advantage in comparison with the existing continuous neutron sources. Parts of the device include the electrical charging unit, the capacitor bank, the spark switch (spark gap), the trigger unit and the vacuum-fuel chamber / anode-cathode. Numerical simulations are used for the simulation of the electrical characteristics of the device including the scaling of the capacitor bank energies with total current, the pinch current, and the scaling of neutron yields with energies and currents. The MCNPX code is used to simulate the moderation of the produced neutrons in a simplified geometry and subsequently, the interaction of thermal neutrons with a test target and the corresponding prompt γ-ray generation.
Toloczko, M.B. ); Garner, F.A. ); Eiholzer, C.R. )
1991-11-01
Irradiation creep data from FFTF-MOTA at {approximately}400{degrees}C were analyzed for nine 20% cold-worked titanium-modified type 316 stainless steels, each of which exhibits a different duration for the transient regime of swelling. One of these steels was the fusion prime candidate alloy designated PCA. The others were various developmental breeder reactor heats. The analysis was based on the assumption that the B{sub 0} + DS creep model applies to these steels at this temperature. This assumption was found to be valid. A creep-swelling coupling coefficient of D {approx} 0.6 {times} 10{sup {minus}2} MPa{sup {minus}1} was found for all steels that had developed a significant level of swelling. This result is in excellent agreement with the results of earlier studies conducted in EBR-II using annealed AISI 304L and also 10% and 20% cold-worked AISI 316 stainless steels. There appears to be some enhancement of swelling by stress, contradicting an important assumption in the analysis and leading to an apparent but misleading nonlinearity of creep with respect to stress.
Toloczko, M.B.; Garner, F.A.; Eiholzer, C.R.
1991-11-01
Irradiation creep data from FFTF-MOTA at {approximately}400{degrees}C were analyzed for nine 20% cold-worked titanium-modified type 316 stainless steels, each of which exhibits a different duration for the transient regime of swelling. One of these steels was the fusion prime candidate alloy designated PCA. The others were various developmental breeder reactor heats. The analysis was based on the assumption that the B{sub 0} + DS creep model applies to these steels at this temperature. This assumption was found to be valid. A creep-swelling coupling coefficient of D {approx} 0.6 {times} 10{sup {minus}2} MPa{sup {minus}1} was found for all steels that had developed a significant level of swelling. This result is in excellent agreement with the results of earlier studies conducted in EBR-II using annealed AISI 304L and also 10% and 20% cold-worked AISI 316 stainless steels. There appears to be some enhancement of swelling by stress, contradicting an important assumption in the analysis and leading to an apparent but misleading nonlinearity of creep with respect to stress.
Mirzajani, N; Ciolini, R; Di Fulvio, A; Esposito, J; d'Errico, F
2014-06-01
Experimental activities are underway at INFN Legnaro National Laboratories (LNL) (Padua, Italy) and Pisa University aimed at angular-dependent neutron energy spectra measurements produced by the (9)Be(p,xn) reaction, under a 5MeV proton beam. This work has been performed in the framework of INFN TRASCO-BNCT project. Bonner Sphere Spectrometer (BSS), based on (6)LiI (Eu) scintillator, was used with the shadow-cone technique. Proper unfolding codes, coupled to BSS response function calculated by Monte Carlo code, were finally used. The main results are reported here.
Developments for neutron-induced fission at IGISOL-4
NASA Astrophysics Data System (ADS)
Gorelov, D.; Penttilä, H.; Al-Adili, A.; Eronen, T.; Hakala, J.; Jokinen, A.; Kankainen, A.; Kolhinen, V. S.; Koponen, J.; Lantz, M.; Mattera, A.; Moore, I. D.; Pohjalainen, I.; Pomp, S.; Rakopoulos, V.; Reinikainen, J.; Rinta-Antila, S.; Simutkin, V.; Solders, A.; Voss, A.; Äystö, J.
2016-06-01
At the IGISOL-4 facility, neutron-rich, medium mass nuclei have usually been produced via charged particle-induced fission of natural uranium and thorium. Neutron-induced fission is expected to have a higher production cross section of the most neutron-rich species. Development of a neutron source along with a new ion guide continues to be one of the major goals since the commissioning of IGISOL-4. Neutron intensities at different angles from a beryllium neutron source have been measured in an on-line experiment with a 30 MeV proton beam. Recently, the new ion guide coupled to the neutron source has been tested as well. Details of the neutron source and ion guide design together with preliminary results from the first neutron-induced fission experiment at IGISOL-4 are presented in this report.
NASA Technical Reports Server (NTRS)
Frigerio, N. A.; Nellans, H. N.; Shaw, M. J.
1969-01-01
Reports relate applications of neutrons to the problem of cancer therapy. The biochemical and biophysical aspects of fast-neutron therapy, neutron-capture and neutron-conversion therapy with intermediate-range neutrons are presented. Also included is a computer program for neutron-gamma radiobiology.
Hurwitz, H. Jr.; Brooks, H.; Mannal, C.; Payne, J.H.; Luebke, E.A.
1959-03-24
A reactor of the heterogeneous, liquid cooled type is described. This reactor is comprised of a central region of a plurality of vertically disposed elongated tubes surrounded by a region of moderator material. The central region is comprised of a central core surrounded by a reflector region which is surrounded by a fast neutron absorber region, which in turn is surrounded by a slow neutron absorber region. Liquid sodium is used as the primary coolant and circulates through the core which contains the fuel elements. Control of the reactor is accomplished by varying the ability of the reflector region to reflect neutrons back into the core of the reactor. For this purpose the reflector is comprised of moderator and control elements having varying effects on reactivity, the control elements being arranged and actuated by groups to give regulation, shim, and safety control.
MCNP (Monte Carlo Neutron Photon) capabilities for nuclear well logging calculations
Forster, R.A.; Little, R.C.; Briesmeister, J.F.
1989-01-01
The Los Alamos Radiation Transport Code System (LARTCS) consists of state-of-the-art Monte Carlo and discrete ordinates transport codes and data libraries. The general-purpose continuous-energy Monte Carlo code MCNP (Monte Carlo Neutron Photon), part of the LARTCS, provides a computational predictive capability for many applications of interest to the nuclear well logging community. The generalized three-dimensional geometry of MCNP is well suited for borehole-tool models. SABRINA, another component of the LARTCS, is a graphics code that can be used to interactively create a complex MCNP geometry. Users can define many source and tally characteristics with standard MCNP features. The time-dependent capability of the code is essential when modeling pulsed sources. Problems with neutrons, photons, and electrons as either single particle or coupled particles can be calculated with MCNP. The physics of neutron and photon transport and interactions is modeled in detail using the latest available cross-section data. A rich collections of variance reduction features can greatly increase the efficiency of a calculation. MCNP is written in FORTRAN 77 and has been run on variety of computer systems from scientific workstations to supercomputers. The next production version of MCNP will include features such as continuous-energy electron transport and a multitasking option. Areas of ongoing research of interest to the well logging community include angle biasing, adaptive Monte Carlo, improved discrete ordinates capabilities, and discrete ordinates/Monte Carlo hybrid development. Los Alamos has requested approval by the Department of Energy to create a Radiation Transport Computational Facility under their User Facility Program to increase external interactions with industry, universities, and other government organizations. 21 refs.
HADES, A Radiographic Simulation Code
Aufderheide, M.B.; Slone, D.M.; Schach von Wittenau, A.E.
2000-08-18
We describe features of the HADES radiographic simulation code. We begin with a discussion of why it is useful to simulate transmission radiography. The capabilities of HADES are described, followed by an application of HADES to a dynamic experiment recently performed at the Los Alamos Neutron Science Center. We describe quantitative comparisons between experimental data and HADES simulations using a copper step wedge. We conclude with a short discussion of future work planned for HADES.
EMPIRE: Nuclear Reaction Model Code System for Data Evaluation
NASA Astrophysics Data System (ADS)
Herman, M.; Capote, R.; Carlson, B. V.; Obložinský, P.; Sin, M.; Trkov, A.; Wienke, H.; Zerkin, V.
2007-12-01
EMPIRE is a modular system of nuclear reaction codes, comprising various nuclear models, and designed for calculations over a broad range of energies and incident particles. A projectile can be a neutron, proton, any ion (including heavy-ions) or a photon. The energy range extends from the beginning of the unresolved resonance region for neutron-induced reactions (∽ keV) and goes up to several hundred MeV for heavy-ion induced reactions. The code accounts for the major nuclear reaction mechanisms, including direct, pre-equilibrium and compound nucleus ones. Direct reactions are described by a generalized optical model (ECIS03) or by the simplified coupled-channels approach (CCFUS). The pre-equilibrium mechanism can be treated by a deformation dependent multi-step direct (ORION + TRISTAN) model, by a NVWY multi-step compound one or by either a pre-equilibrium exciton model with cluster emission (PCROSS) or by another with full angular momentum coupling (DEGAS). Finally, the compound nucleus decay is described by the full featured Hauser-Feshbach model with γ-cascade and width-fluctuations. Advanced treatment of the fission channel takes into account transmission through a multiple-humped fission barrier with absorption in the wells. The fission probability is derived in the WKB approximation within the optical model of fission. Several options for nuclear level densities include the EMPIRE-specific approach, which accounts for the effects of the dynamic deformation of a fast rotating nucleus, the classical Gilbert-Cameron approach and pre-calculated tables obtained with a microscopic model based on HFB single-particle level schemes with collective enhancement. A comprehensive library of input parameters covers nuclear masses, optical model parameters, ground state deformations, discrete levels and decay schemes, level densities, fission barriers, moments of inertia and γ-ray strength functions. The results can be converted into ENDF-6 formatted files using the
2006-10-27
FAA Smoke Transport Code, a physics-based Computational Fluid Dynamics tool, which couples heat, mass, and momentum transfer, has been developed to provide information on smoke transport in cargo compartments with various geometries and flight conditions. The software package contains a graphical user interface for specification of geometry and boundary conditions, analysis module for solving the governing equations, and a post-processing tool. The current code was produced by making substantial improvements and additions to a codemore » obtained from a university. The original code was able to compute steady, uniform, isothermal turbulent pressurization. In addition, a preprocessor and postprocessor were added to arrive at the current software package.« less
Seals Code Development Workshop
NASA Technical Reports Server (NTRS)
Hendricks, Robert C. (Compiler); Liang, Anita D. (Compiler)
1996-01-01
Seals Workshop of 1995 industrial code (INDSEAL) release include ICYL, GCYLT, IFACE, GFACE, SPIRALG, SPIRALI, DYSEAL, and KTK. The scientific code (SCISEAL) release includes conjugate heat transfer and multidomain with rotordynamic capability. Several seals and bearings codes (e.g., HYDROFLEX, HYDROTRAN, HYDROB3D, FLOWCON1, FLOWCON2) are presented and results compared. Current computational and experimental emphasis includes multiple connected cavity flows with goals of reducing parasitic losses and gas ingestion. Labyrinth seals continue to play a significant role in sealing with face, honeycomb, and new sealing concepts under investigation for advanced engine concepts in view of strict environmental constraints. The clean sheet approach to engine design is advocated with program directions and anticipated percentage SFC reductions cited. Future activities center on engine applications with coupled seal/power/secondary flow streams.
Broekman, J. D.; Nigg, D. W.; Hawthorne, M. F.
2013-07-01
Parameter studies, design calculations and neutronic performance measurements have been completed for a new thermal neutron beamline constructed for neutron capture therapy cell and small-animal radiobiology studies at the University of Missouri Research Reactor. The beamline features the use of single-crystal silicon and bismuth sections for neutron filtering and for reduction of incident gamma radiation. The computational models used for the final beam design and performance evaluation are based on coupled discrete-ordinates and Monte Carlo techniques that permit detailed modeling of the neutron transmission properties of the filtering crystals with very few approximations. Validation protocols based on neutron activation spectrometry measurements and rigorous least-square adjustment techniques show that the beam produces a neutron spectrum that has the anticipated level of thermal neutron flux and a somewhat higher than expected, but radio-biologically insignificant, epithermal neutron flux component. (authors)
Study of neutron focusing at the Texas Cold Neutron Source. Final report
Wehring, B.W.; Uenlue, K.
1996-12-19
The goals of this three-year study were: (1) design a neutron focusing system for use with the Texas Cold Neutron Source (TCNS) to produce an intense beam of cold neutrons appropriate for prompt gamma activation analysis (PGAA); (2) orchestrate the construction of the focusing system, integrate it into the TCNS neutron guide complex, and measure its performance; and (3) design, setup, and test a cold-neutron PGAA system which utilizes the guided focused cold neutron beam. During the first year of the DOE grant, a new procedure was developed and used to design a focusing converging guide consisting of truncated rectangular cone sections. Detailed calculations were performed using a 3-D Monte Carlo code which the authors wrote to trace neutrons through the curved guide of the TCNS into the proposed converging guide. Using realistic reflectivities for Ni-Ti supermirrors, the authors obtained gains of 3 to 5 for 4 different converging guide geometries. During the second year of the DOE grant, the subject of this final report, Ovonic Synthetic Materials Company was contracted to build a converging neutron guide focusing system to the specifications. Considerable time and effort were spent working with Ovonics on selecting the materials for the converging neutron guide system. The major portion of the research on the design of a cold-neutron PGAA system was also completed during the second year. At the beginning of the third year of the grant, a converging neutron guide focusing system had been ordered, and a cold-neutron PGAA system had been designed. Since DOE did not fund the third year, there was no money to purchase the required equipment for the cold-neutron PGAA system and no money to perform tests of either the converging neutron guide or the cold-neutron PGAA system. The research already accomplished would have little value without testing the systems which had been designed. Thus the project was continued at a pace that could be sustained with internal funding.
Fermi, E.; Szilard, L.
1957-09-24
Reactors of the type employing plates of natural uranium in a moderator are discussed wherein the plates are um-formly disposed in parallel relationship to each other thereby separating the moderator material into distinct and individual layers. Each plate has an uninterrupted sunface area substantially equal to the cross-sectional area of the active portion of the reactor, the particular size of the plates and the volume ratio of moderator to uranium required to sustain a chain reaction being determinable from the known purity of these materials and other characteristics such as the predictable neutron losses due to the formation of radioactive elements of extremely high neutron capture cross section.
Wigner, E.P.; Weinberg, A.W.; Young, G.J.
1958-04-15
A nuclear reactor which uses uranium in the form of elongated tubes as fuel elements and liquid as a coolant is described. Elongated tubular uranium bodies are vertically disposed in an efficient neutron slowing agent, such as graphite, for example, to form a lattice structure which is disposed between upper and lower coolant tanks. Fluid coolant tubes extend through the uranium bodies and communicate with the upper and lower tanks and serve to convey the coolant through the uranium body. The reactor is also provided with means for circulating the cooling fluid through the coolant tanks and coolant tubes, suitable neutron and gnmma ray shields, and control means.