International Nuclear Information System (INIS)
Zazula, J.M.
1984-01-01
This work concerns calculation of a neutron response, caused by a neutron field perturbed by materials surrounding the source or the detector. Solution of a problem is obtained using coupling of the Monte Carlo radiation transport computation for the perturbed region and the discrete ordinates transport computation for the unperturbed system. (author). 62 refs
A time-dependent neutron transport model and its coupling to thermal-hydraulics
International Nuclear Information System (INIS)
Pautz, A.
2001-01-01
A new neutron transport code for time-dependent analyses of nuclear systems has been developed. The code system is based on the well-known Discrete Ordinates code DORT, which solves the steady-state neutron/photon transport equation in two dimensions for an arbitrary number of energy groups and the most common regular geometries. For the implementation of time-dependence a fully implicit first-order scheme was employed to minimize errors due to temporal discretization. This requires various modifications to the transport equation as well as the extensive use of elaborated acceleration mechanisms. The convergence criteria for fluxes, fission rates etc. had to be strongly tightened to ensure the reliability of results. To perform coupled analyses, an interface to the GRS system code ATHLET has been developed. The nodal power densities from the neutron transport code are passed to ATHLET to calculate thermal-hydraulic system parameters, e.g. fuel and coolant temperatures. These are in turn used to generate appropriate nuclear cross sections by interpolation of pre-calculated data sets for each time step. Finally, to demonstrate the transient capabilities of the coupled code system, the research reactor FRM-II has been analysed. Several design basis accidents were modelled, like the loss of off site power, loss of secondary heat sink and unintended control rod withdrawal. (author)
Energy Technology Data Exchange (ETDEWEB)
Baker, Randal Scott [Univ. of Arizona, Tucson, AZ (United States)
1990-01-01
The neutron transport equation is solved by a hybrid method that iteratively couples regions where deterministic (S_{N}) and stochastic (Monte Carlo) methods are applied. Unlike previous hybrid methods, the Monte Carlo and S_{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_{N} is well suited for by themselves. The fully coupled Monte Carlo/S_{N} technique consists of defining spatial and/or energy regions of a problem in which either a Monte Carlo calculation or an S_{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_{N} region. The Monte Carlo and S_{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_{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_{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_{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.
International Nuclear Information System (INIS)
Russell, G.J.; Pitcher, E.J.; Ferguson, P.D.
1995-01-01
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
A 2D/1D coupling neutron transport method based on the matrix MOC and NEM methods
Energy Technology Data Exchange (ETDEWEB)
Zhang, H.; Zheng, Y.; Wu, H.; Cao, L. [School of Nuclear Science and Technology, Xi' an Jiaotong University, No. 28, Xianning West Road, Xi' an, Shaanxi 710049 (China)
2013-07-01
A new 2D/1D coupling method based on the matrix MOC method (MMOC) and nodal expansion method (NEM) is proposed for solving the three-dimensional heterogeneous neutron transport problem. The MMOC method, used for radial two-dimensional calculation, constructs a response matrix between source and flux with only one sweep and then solves the linear system by using the restarted GMRES algorithm instead of the traditional trajectory sweeping process during within-group iteration for angular flux update. Long characteristics are generated by using the customization of commercial software AutoCAD. A one-dimensional diffusion calculation is carried out in the axial direction by employing the NEM method. The 2D and ID solutions are coupled through the transverse leakage items. The 3D CMFD method is used to ensure the global neutron balance and adjust the different convergence properties of the radial and axial solvers. A computational code is developed based on these theories. Two benchmarks are calculated to verify the coupling method and the code. It is observed that the corresponding numerical results agree well with references, which indicates that the new method is capable of solving the 3D heterogeneous neutron transport problem directly. (authors)
Energy Technology Data Exchange (ETDEWEB)
Gleicher, Frederick N.; Williamson, Richard L.; Ortensi, Javier; Wang, Yaqi; Spencer, Benjamin W.; Novascone, Stephen R.; Hales, Jason D.; Martineau, Richard C.
2014-10-01
The MOOSE neutron transport application RATTLESNAKE was coupled to the fuels performance application BISON to provide a higher fidelity tool for fuel performance simulation. This project is motivated by the desire to couple a high fidelity core analysis program (based on the self-adjoint angular flux equations) to a high fidelity fuel performance program, both of which can simulate on unstructured meshes. RATTLESNAKE solves self-adjoint angular flux transport equation and provides a sub-pin level resolution of the multigroup neutron flux with resonance treatment during burnup or a fast transient. BISON solves the coupled thermomechanical equations for the fuel on a sub-millimeter scale. Both applications are able to solve their respective systems on aligned and unaligned unstructured finite element meshes. The power density and local burnup was transferred from RATTLESNAKE to BISON with the MOOSE Multiapp transfer system. Multiple depletion cases were run with one-way data transfer from RATTLESNAKE to BISON. The eigenvalues are shown to agree well with values obtained from the lattice physics code DRAGON. The one-way data transfer of power density is shown to agree with the power density obtained from an internal Lassman-style model in BISON.
International Nuclear Information System (INIS)
Koyama, Kinji; Taji, Yukichi; Miyasaka, Shun-ichi; Minami, Kazuyoshi.
1977-07-01
The modular code system RADHEAT is for producing coupled multigroup neutron and gamma-ray cross section sets, analyzing the neutron and gamma-ray transport, and calculating the energy deposition and atomic displacements due to these radiations in a nuclear reactor or shield. The basic neutron cross sections and secondary gamma-ray production data are taken from ENDF/B and POPOP4 libraries respectively. The system (1) generates multigroup neutron cross sections, energy deposition coefficients and atomic displacement factors due to neutron reactions, (2) generates multigroup gamma-ray cross sections and energy transfer coefficients, (3) generates secondary gamma-ray production cross sections, (4) combines these cross sections into the coupled set, (5) outputs and updates the multigroup cross section libraries in convenient formats for other transport codes, (6) analyzes the neutron and gamma-ray transport and calculates the energy deposition and the number density of atomic displacements in a medium, (7) collapses the cross sections to a broad-group structure, by option, using the weighting functions obtained by one-dimensional transport calculation, and (8) plots, by option, multigroup cross sections, and neutron and gamma-ray distributions. Definitions of the input data required in various options of the code system are also given. (auth.)
Cullen, D
2000-01-01
TART2000 is a coupled neutron-photon, 3 Dimensional, combinatorial geometry, time dependent Monte Carlo radiation transport code. This code can run on any modern computer. It is a complete system to assist you with input Preparation, running Monte Carlo calculations, and analysis of output results. TART2000 is also incredibly FAST; if you have used similar codes, you will be amazed at how fast this code is compared to other similar codes. Use of the entire system can save you a great deal of time and energy. TART2000 is distributed on CD. This CD contains on-line documentation for all codes included in the system, the codes configured to run on a variety of computers, and many example problems that you can use to familiarize yourself with the system. TART2000 completely supersedes all older versions of TART, and it is strongly recommended that users only use the most recent version of TART2000 and its data files.
International Nuclear Information System (INIS)
Cullen, D.E
2000-01-01
TART2000 is a coupled neutron-photon, 3 Dimensional, combinatorial geometry, time dependent Monte Carlo radiation transport code. This code can run on any modern computer. It is a complete system to assist you with input Preparation, running Monte Carlo calculations, and analysis of output results. TART2000 is also incredibly FAST; if you have used similar codes, you will be amazed at how fast this code is compared to other similar codes. Use of the entire system can save you a great deal of time and energy. TART2000 is distributed on CD. This CD contains on-line documentation for all codes included in the system, the codes configured to run on a variety of computers, and many example problems that you can use to familiarize yourself with the system. TART2000 completely supersedes all older versions of TART, and it is strongly recommended that users only use the most recent version of TART2000 and its data files
Energy Technology Data Exchange (ETDEWEB)
Burns, Kimberly A. [Georgia Inst. of Technology, Atlanta, GA (United States)
2009-08-01
The accurate and efficient simulation of coupled neutron-photon problems is necessary for several important radiation detection applications. Examples include the detection of nuclear threats concealed in cargo containers and prompt gamma neutron activation analysis for nondestructive determination of elemental composition of unknown samples.
Energy Technology Data Exchange (ETDEWEB)
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.
Applying the response matrix method for solving coupled neutron diffusion and transport problems
International Nuclear Information System (INIS)
Sibiya, G.S.
1980-01-01
The numerical determination of the flux and power distribution in the design of large power reactors is quite a time-consuming procedure if the space under consideration is to be subdivided into very fine weshes. Many computing methods applied in reactor physics (such as the finite-difference method) require considerable computing time. In this thesis it is shown that the response matrix method can be successfully used as an alternative approach to solving the two-dimension diffusion equation. Furthermore it is shown that sufficient accuracy of the method is achieved by assuming a linear space dependence of the neutron currents on the boundaries of the geometries defined for the given space. (orig.) [de
International Nuclear Information System (INIS)
White, Morgan C.
2000-01-01
The fundamental motivation for the research presented in this dissertation was the need to development a more accurate prediction method for characterization of mixed radiation fields around medical electron accelerators (MEAs). Specifically, a model is developed for simulation of neutron and other particle production from photonuclear reactions and incorporated in the Monte Carlo N-Particle (MCNP) radiation transport code. This extension of the capability within the MCNP code provides for the more accurate assessment of the mixed radiation fields. The Nuclear Theory and Applications group of the Los Alamos National Laboratory has recently provided first-of-a-kind evaluated photonuclear data for a select group of isotopes. These data provide the reaction probabilities as functions of incident photon energy with angular and energy distribution information for all reaction products. The availability of these data is the cornerstone of the new methodology for state-of-the-art mutually coupled photon-neutron 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 and 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
Energy Technology Data Exchange (ETDEWEB)
White, Morgan C. [Univ. of Florida, Gainesville, FL (United States)
2000-07-01
The fundamental motivation for the research presented in this dissertation was the need to development a more accurate prediction method for characterization of mixed radiation fields around medical electron accelerators (MEAs). Specifically, a model is developed for simulation of neutron and other particle production from photonuclear reactions and incorporated in the Monte Carlo N-Particle (MCNP) radiation transport code. This extension of the capability within the MCNP code provides for the more accurate assessment of the mixed radiation fields. The Nuclear Theory and Applications group of the Los Alamos National Laboratory has recently provided first-of-a-kind evaluated photonuclear data for a select group of isotopes. These data provide the reaction probabilities as functions of incident photon energy with angular and energy distribution information for all reaction products. The availability of these data is the cornerstone of the new methodology for state-of-the-art mutually coupled photon-neutron transport simulations. The dissertation includes details of the model development and implementation necessary to use the new photonuclear data within MCNP simulations. A new data format has been developed to include tabular photonuclear data. Data are processed from the Evaluated Nuclear Data Format (ENDF) to the new class ''u'' A Compact ENDF (ACE) format using a standalone processing code. MCNP modifications have been completed to enable Monte Carlo sampling of photonuclear reactions. Note that both neutron and gamma production are included in the present model. The new capability has been subjected to extensive verification and validation (V&V) testing. Verification testing has established the expected basic functionality. Two validation projects were undertaken. First, comparisons were made to benchmark data from literature. These calculations demonstrate the accuracy of the new data and transport routines to better than 25 percent. Second
Neutron stochastic transport theory with delayed neutrons
International Nuclear Information System (INIS)
Munoz-Cobo, J.L.; Verdu, G.
1987-01-01
From the stochastic transport theory with delayed neutrons, the Boltzmann transport equation with delayed neutrons for the average flux emerges in a natural way without recourse to any approximation. From this theory a general expression is obtained for the Feynman Y-function when delayed neutrons are included. The single mode approximation for the particular case of a subcritical assembly is developed, and it is shown that Y-function reduces to the familiar expression quoted in many books, when delayed neutrons are not considered, and spatial and source effects are not included. (author)
Palmer, T S
2003-01-01
In this NEER project, researchers from Oregon State University have investigated the limitations of the treatment of two-phase coolants as a homogeneous mixture in neutron transport calculations. Improved methods of calculating the neutron distribution in binary stochastic mixtures have been developed over the past 10-15 years and are readily available in the transport literature. These methods are computationally more expensive than the homogeneous (or atomic mix) models, but can give much more accurate estimates of ensemble average fluxes and reaction rates provided statistical descriptions of the distributions of the two materials are know. A thorough review of the two-phase flow literature has been completed and the relevant mixture distributions have been identified. Using these distributions, we have performed Monte Carlo criticality calculations of fuel assemblies to assess the accuracy of the atomic mix approximation when compared to a resolved treatment of the two-phase coolant. To understand the ben...
Energy Technology Data Exchange (ETDEWEB)
Biondo, Elliott D [ORNL; Ibrahim, Ahmad M [ORNL; Mosher, Scott W [ORNL; Grove, Robert E [ORNL
2015-01-01
Detailed radiation transport calculations are necessary for many aspects of the design of fusion energy systems (FES) such as ensuring occupational safety, assessing the activation of system components for waste disposal, and maintaining cryogenic temperatures within superconducting magnets. Hybrid Monte Carlo (MC)/deterministic techniques are necessary for this analysis because FES are large, heavily shielded, and contain streaming paths that can only be resolved with MC. The tremendous complexity of FES necessitates the use of CAD geometry for design and analysis. Previous ITER analysis has required the translation of CAD geometry to MCNP5 form in order to use the AutomateD VAriaNce reducTion Generator (ADVANTG) for hybrid MC/deterministic transport. In this work, ADVANTG was modified to support CAD geometry, allowing hybrid (MC)/deterministic transport to be done automatically and eliminating the need for this translation step. This was done by adding a new ray tracing routine to ADVANTG for CAD geometries using the Direct Accelerated Geometry Monte Carlo (DAGMC) software library. This new capability is demonstrated with a prompt dose rate calculation for an ITER computational benchmark problem using both the Consistent Adjoint Driven Importance Sampling (CADIS) method an the Forward Weighted (FW)-CADIS method. The variance reduction parameters produced by ADVANTG are shown to be the same using CAD geometry and standard MCNP5 geometry. Significant speedups were observed for both neutrons (as high as a factor of 7.1) and photons (as high as a factor of 59.6).
International Nuclear Information System (INIS)
Fanaro, L.C.C.B.
1984-01-01
It was developed the BLINDAGE computer code for the radiation transport (neutrons and gammas) calculation. The code uses the removal - diffusion method for neutron transport and point-kernel technique with buil-up factors for gamma-rays. The results obtained through BLINDAGE code are compared with those obtained with the ANISN and SABINE computer codes. (Author) [pt
Neutron transport simulation (selected topics)
Vaz, P.
2009-10-01
Neutron transport simulation is usually performed for criticality, power distribution, activation, scattering, dosimetry and shielding problems, among others. During the last fifteen years, innovative technological applications have been proposed (Accelerator Driven Systems, Energy Amplifiers, Spallation Neutron Sources, etc.), involving the utilization of intermediate energies (hundreds of MeV) and high-intensity (tens of mA) proton accelerators impinging in targets of high Z elements. Additionally, the use of protons, neutrons and light ions for medical applications (hadrontherapy) impose requirements on neutron dosimetry-related quantities (such as kerma factors) for biologically relevant materials, in the energy range starting at several tens of MeV. Shielding and activation related problems associated to the operation of high-energy proton accelerators, emerging space-related applications and aircrew dosimetry-related topics are also fields of intense activity requiring as accurate as possible medium- and high-energy neutron (and other hadrons) transport simulation. These applications impose specific requirements on cross-section data for structural materials, targets, actinides and biologically relevant materials. Emerging nuclear energy systems and next generation nuclear reactors also impose requirements on accurate neutron transport calculations and on cross-section data needs for structural materials, coolants and nuclear fuel materials, aiming at improved safety and detailed thermal-hydraulics and radiation damage studies. In this review paper, the state-of-the-art in the computational tools and methodologies available to perform neutron transport simulation is presented. Proton- and neutron-induced cross-section data needs and requirements are discussed. Hot topics are pinpointed, prospective views are provided and future trends identified.
Neutron therapy coupling brachytherapy and boron neutron capture therapy (BNCT) techniques
International Nuclear Information System (INIS)
Chaves, Iara Ferreira.
1994-12-01
In the present dissertation, neutron radiation techniques applied into organs of the human body are investigated as oncologic radiation therapy. The proposal treatment consists on connecting two distinct techniques: Boron Neutron Capture Therapy (BNCT) and irradiation by discrete sources of neutrons, through the brachytherapy conception. Biological and radio-dosimetrical aspects of the two techniques are considered. Nuclear aspects are discussed, presenting the nuclear reactions occurred in tumoral region, and describing the forms of evaluating the dose curves. Methods for estimating radiation transmission are reviewed through the solution of the neutron transport equation, Monte Carlo methodology, and simplified analytical calculation based on diffusion equation and numerical integration. The last is computational developed and presented as a quickly way to neutron transport evaluation in homogeneous medium. The computational evaluation of the doses for distinct hypothetical situations is presented, applying the coupled techniques BNTC and brachytherapy as an possible oncologic treatment. (author). 78 refs., 61 figs., 21 tabs
Influence of the neutron transport tube on neutron resonance densitometry
Directory of Open Access Journals (Sweden)
Kitatani Fumito
2017-01-01
Full Text Available Neutron Resonance Densitometry (NRD is a non-destructive assay technique of nuclear materials in particle-like debris that contains various materials. An aim of NRD is to quantify nuclear materials in a melting fuel of Fukusima Daiichi plant, spent nuclear fuel and annihilation disposal fuel etc. NRD consists of two techniques of Neutron Resonance Transmission Analysis (NRTA and Neutron Resonance Capture Analysis (NRCA or Prompt Gamma-ray Analysis (PGA. A density of nuclear material isotopes is decided with NRTA. The materials absorbing a neutron in a wide energy range such as boron in a sample are identified by NRCA/PGA. The information of NRCA/PGA is used in NRTA analysis to quantify nuclear material isotopes. A neutron time of flight (TOF method is used in NRD measurements. A facility, consisting of a neutron source, a neutron flight path, and a detector is required. A short flight path and a strong neutron source are needed to downsize such a facility and put NRD into practical use. A neutron transport tube covers a flight path to prevent noises. In order to investigate the effect of neutron transport tube and pulse width of a neutron source, we carried out NRTA experiments with a 2-m short neutron transport tube constructed at Kyoto University Research Reactor Institute - Linear Accelerator (KURRI-LINAC, and impacts of shield of neutron transport tube and influence of pulse width of a neutron source were examined. A shield of the neutron transport tube reduced a background and had a good influence on the measurement. The resonance dips of 183W at 27 eV was successfully observed with a pulse width of a neutron source less than 2 μs.
Influence of the neutron transport tube on neutron resonance densitometry
Kitatani, Fumito; Tsuchiya, Harufumi; Koizumi, Mitsuo; Takamine, Jun; Hori, Junichi; Sano, Tadafumi
2017-09-01
Neutron Resonance Densitometry (NRD) is a non-destructive assay technique of nuclear materials in particle-like debris that contains various materials. An aim of NRD is to quantify nuclear materials in a melting fuel of Fukusima Daiichi plant, spent nuclear fuel and annihilation disposal fuel etc. NRD consists of two techniques of Neutron Resonance Transmission Analysis (NRTA) and Neutron Resonance Capture Analysis (NRCA) or Prompt Gamma-ray Analysis (PGA). A density of nuclear material isotopes is decided with NRTA. The materials absorbing a neutron in a wide energy range such as boron in a sample are identified by NRCA/PGA. The information of NRCA/PGA is used in NRTA analysis to quantify nuclear material isotopes. A neutron time of flight (TOF) method is used in NRD measurements. A facility, consisting of a neutron source, a neutron flight path, and a detector is required. A short flight path and a strong neutron source are needed to downsize such a facility and put NRD into practical use. A neutron transport tube covers a flight path to prevent noises. In order to investigate the effect of neutron transport tube and pulse width of a neutron source, we carried out NRTA experiments with a 2-m short neutron transport tube constructed at Kyoto University Research Reactor Institute - Linear Accelerator (KURRI-LINAC), and impacts of shield of neutron transport tube and influence of pulse width of a neutron source were examined. A shield of the neutron transport tube reduced a background and had a good influence on the measurement. The resonance dips of 183W at 27 eV was successfully observed with a pulse width of a neutron source less than 2 μs.
International Nuclear Information System (INIS)
Shin, Chang Hwan; Seo, Kyong Won; Chun, Tae Hyun; Kim, Kang Seog
2005-03-01
Code coupling activities have so far focused on coupling the neutronics modules with the CFD module. An interface module for the CFD-ACE/DeCART coupling was established as an alternative to the original STAR-CD/DeCART interface. The interface module for DeCART/CFD-ACE was validated by single-pin model. The optimized CFD mesh was decided through the calculation of multi-pin model. It was important to consider turbulent mixing of subchannels for calculation of fuel temperature. For the parallel calculation, the optimized decompose process was necessary to reduce the calculation costs and setting of the iteration and convergence criterion for each code was important, too
Neutron transport equation - indications on homogenization and neutron diffusion
International Nuclear Information System (INIS)
Argaud, J.P.
1992-06-01
In PWR nuclear reactor, the practical study of the neutrons in the core uses diffusion equation to describe the problem. On the other hand, the most correct method to describe these neutrons is to use the Boltzmann equation, or neutron transport equation. In this paper, we give some theoretical indications to obtain a diffusion equation from the general transport equation, with some simplifying hypothesis. The work is organised as follows: (a) the most general formulations of the transport equation are presented: integro-differential equation and integral equation; (b) the theoretical approximation of this Boltzmann equation by a diffusion equation is introduced, by the way of asymptotic developments; (c) practical homogenization methods of transport equation is then presented. In particular, the relationships with some general and useful methods in neutronic are shown, and some homogenization methods in energy and space are indicated. A lot of other points of view or complements are detailed in the text or the remarks
VHTR core modeling: coupling between neutronic and thermal-hydraulics
International Nuclear Information System (INIS)
Limaiem, I.; Damian, F.; Raepsaet, X.; Studer, E.
2005-01-01
Following the present interest in the next generation nuclear power plan (NGNP), Cea is deploying special effort to develop new models and qualify its research tools for this next generation reactors core. In this framework, the Very High Temperature Reactor concept (VHTR) has an increasing place in the actual research program. In such type of core, a strong interaction exists between neutronic and thermal-hydraulics. Consequently, the global core modelling requires accounting for the temperature feedback in the neutronic models. The purpose of this paper is to present the new neutronic and thermal-hydraulics coupling model dedicated to the High Temperature Reactors (HTR). The coupling model integrates a new version of the neutronic scheme calculation developed in collaboration between Cea and Framatome-ANP. The neutronic calculations are performed using a specific calculation processes based on the APOLLO2 transport code and CRONOS2 diffusion code which are part of the French reactor physics code system SAPHYR. The thermal-hydraulics model is characterised by an equivalent porous media and 1-D fluid/3-D thermal model implemented in the CAST3M/ARCTURUS code. The porous media approach involves the definition of both homogenous and heterogeneous models to ensure a correct temperature feedback. This study highlights the sensitivity of the coupling system's parameters (radial/axial meshing and data exchange strategy between neutronic and thermal-hydraulics code). The parameters sensitivity study leads to the definition of an optimal coupling system specification for the VHTR. Besides, this work presents the first physical analysis of the VHTR core in steady-state condition. The analysis gives information about the 3-D power peaking and the temperature coefficient. Indeed, it covers different core configurations with different helium distribution in the core bypass. (authors)
International Nuclear Information System (INIS)
Bruggeman, M.; Mandoki, R.; Van Iseghem, P.
1994-09-01
Monte Carlo simulations are used to investigate the performance and possible optimization of simple passive and active neutron assay systems for the determination of fissile material in waste packages. The active system uses external alpha-neutron or gamma-neutron sources -with mean neutron energies below 1 MeV- which continuously irradiates the waste sample. The discrimination between these source neutrons and the neutrons from induced fission in the detection process is based on the different transport properties of these neutrons. The detection limits obtained with the active system is of the order of 1 g 235 U in 1000 s measuring time
Deficiency in Monte Carlo simulations of coupled neutron-gamma-ray fields
Maleka, Peane P.; Maucec, Marko; de Meijer, Robert J.
2011-01-01
The deficiency in Monte Carlo simulations of coupled neutron-gamma-ray field was investigated by benchmarking two simulation codes with experimental data. Simulations showed better correspondence with the experimental data for gamma-ray transport only. In simulations, the neutron interactions with
Cation-Coupled Bicarbonate Transporters
Aalkjaer, Christian; Boedtkjer, Ebbe; Choi, Inyeong; Lee, Soojung
2014-01-01
Cation-coupled HCO3− transport was initially identified in the mid-1970s when pioneering studies showed that acid extrusion from cells is stimulated by CO2/HCO3− and associated with Na+ and Cl− movement. The first Na+-coupled bicarbonate transporter (NCBT) was expression-cloned in the late 1990s. There are currently five mammalian NCBTs in the SLC4-family: the electrogenic Na,HCO3-cotransporters NBCe1 and NBCe2 (SLC4A4 and SLC4A5 gene products); the electroneutral Na,HCO3-cotransporter NBCn1 ...
An introduction to the neutron transport phenomena
International Nuclear Information System (INIS)
Kulikowska, T.
2001-01-01
The main goal of the present lecture is to is to give a short description of neutron transport phenomena limited to those definitions that are necessary to understand the approach to practical solution of the problem given in the second lecture on reactor lattice transport calculations. The discussion of the neutron cross sections has been skipped as other lecturers have treated this subject in detail. (author)
Neutron transport model for standard calculation experiment
International Nuclear Information System (INIS)
Lukhminskij, B.E.; Lyutostanskij, Yu.S.; Lyashchuk, V.I.; Panov, I.V.
1989-01-01
The neutron transport calculation algorithms in complex composition media with a predetermined geometry are realized by the multigroups representations within Monte Carlo methods in the MAMONT code. The code grade was evaluated with benchmark experiments comparison. The neutron leakage spectra calculations in the spherical-symmetric geometry were carried out for iron and polyethylene. The MAMONT code utilization for metrological furnishes of the geophysics tasks is proposed. The code is orientated towards neutron transport and secondary nuclides accumulation calculations in blankets and geophysics media. 7 refs.; 2 figs
Heterogeneity effects in neutron transport computations
International Nuclear Information System (INIS)
Gelbard, E.M.
1975-01-01
A nuclear reactor is, generally, an intricate heterogeneous structure whose adjacent components may differ radically in their neutronic properties. The heterogeneities in the structure of the reactor complicate the work of the reactor analyst and tend to degrade the efficiency of the numerical methods used in reactor computations. Two types of heterogeneity effects are considered. First, certain singularities in the solution of the neutron transport equation, induced by heterogeneities, are briefly described. Second, the effect of heterogeneities on neutron leakage rates, and consequently on effective diffusion coefficients, are discussed. (5 figures) (U.S.)
Implementation of the quasi-static method for neutron transport
International Nuclear Information System (INIS)
Alcaro, Fabio; Dulla, Sandra; Ravetto, Piero; Le Tellier, Romain; Suteau, Christophe
2011-01-01
The study of the dynamic behavior of next generation nuclear reactors is a fundamental aspect for safety and reliability assessments. Despite the growing performances of modern computers, the full solution of the neutron Boltzmann equation in the time domain is still an impracticable task, thus several approximate dynamic models have been proposed for the simulation of nuclear reactor transients; the quasi-static method represents the standard tool currently adopted for the space-time solution of neutron transport problems. All the practical applications of this method that have been proposed contain a major limit, consisting in the use of isotropic quantities, such as scalar fluxes and isotropic external neutron sources, being the only data structures available in most deterministic transport codes. The loss of the angular information produces both inaccuracies in the solution of the kinetic model and the inconsistency of the quasi-static method itself. The present paper is devoted to the implementation of a consistent quasi-static method. The computational platform developed by CEA in Cadarache has been used for the creation of a kinetic package to be coupled with the existing SNATCH solver, a discrete-ordinate multi-dimensional neutron transport solver, employed for the solution of the steady-state Boltzmann equation. The work aims at highlighting the effects of the angular treatment of the neutron flux on the transient analysis, comparing the results with those produced by the previous implementations of the quasi-static method. (author)
Application of Walsh functions to neutron transport problems. I. Theory
International Nuclear Information System (INIS)
Seed, T.J.; Albrecht, R.W.
1976-01-01
An approximation to the neutron transport equation is made by representing the angular flux with an expansion of the angular dependence in the orthogonal, complete, and binary valued sets of Walsh function. The Walsh approximation is applied to the one-speed, isotropic-scattering, rectangular-geometry form of the neutron transport equation. Sets of partial differential equations for the expansion coefficients are derived along with appropriate boundary conditions for their solution. The sets of the Walsh expansion to one- and two-dimensional forms of the transport equation are also obtained. The two-dimensional expansion coefficient equations are shown to be not only hyperbolic but also transformable to a set of S/sub N/-like equations that are coupled only through the scattering term. Such transformal sets of equations are termed Walsh-derived quadrature sets
Fusion neutron damage to a charge coupled device camera
Amaden, Christopher Dean
1997-01-01
Approved for public release; distribution is unlimited A charge coupled device (CCD) camera's performance has been degraded by damage produced by 14 MeV neutrons (n) from the Rotating Target Neutron Source. High energy neutrons produce atomic dislocation in doped silicon electronics. This thesis explores changes in Dark Current (J), Charge Transfer Inefficiency (CTI), and Contrast Transfer Function (CTF) as measures of neutron damage. The camera was irradiated to a fluence, Phi, of 6.60 x ...
Parameterized Radiation Transport Model for Neutron Detection in Complex Scenes
Lavelle, C. M.; Bisson, D.; Gilligan, J.; Fisher, B. M.; Mayo, R. M.
2013-04-01
There is interest in developing the ability to rapidly compute the energy dependent neutron flux within a complex geometry for a variety of applications. Coupled with sensor response function information, this capability would allow direct estimation of sensor behavior in multitude of operational scenarios. In situations where detailed simulation is not warranted or affordable, it is desirable to possess reliable estimates of the neutron field in practical scenarios which do not require intense computation. A tool set of this kind would provide quantitative means to address the development of operational concepts, inform asset allocation decisions, and exercise planning. Monte Carlo and/or deterministic methods provide a high degree of precision and fidelity consistent with the accuracy with which the scene is rendered. However, these methods are often too computationally expensive to support the real-time evolution of a virtual operational scenario. High fidelity neutron transport simulations are also time consuming from the standpoint of user setup and post-simulation analysis. We pre-compute adjoint solutions using MCNP to generate a coarse spatial and energy grid of the neutron flux over various surfaces as an alternative to full Monte Carlo modeling. We attempt to capture the characteristics of the neutron transport solution. We report on the results of brief verification and validation measurements which test the predictive capability of this approach over soil and asphalt concrete surfaces. We highlight the sensitivity of the simulated and experimental results to the material composition of the environment.
The Effect of Anisotropic Scatter on Atmospheric Neutron Transport
2015-03-26
THE EFFECT OF ANISOTROPIC SCATTER ON ATMOSPHERIC NEUTRON TRANSPORT THESIS MARCH 2015 Nicholas J...iii AFIT-ENP-MS-15-M-085 THE EFFECT OF ANISOTROPIC SCATTER ON ATMOSPHERIC NEUTRON TRANSPORT THESIS Presented to the...EFFECT OF ANISOTROPIC SCATTER ON ATMOSPHERIC NEUTRON TRANSPORT Nicholas J. McIntee, BSE Major, USA Committee Membership: Dr. Kirk A. Mathews
International Nuclear Information System (INIS)
Sallah, M.; Margeanu, C. A.
2016-01-01
The space-fractional neutron transport equation is used to describe the neutrons transport in finite disturbed reactors. It is approximated using the Pomraning-Eddington technique to yield two space-fractional differential equations, in terms of neutron density and net neutron flux. These resultant equations are coupled into a fractional diffusion-like equation for the neutron density whose solution is obtained by using Laplace transformation method. The solution is represented in terms of the Mittag-Leffler function and its different orders. The scattering is considered as quadratic scattering to offer a more realistic, compact representation of the system, and to increase the accuracy of the estimated neutronic parameters. The results are presented graphically to illustrate the fractional parameter effect in addition to the effect of radiative-transfer properties on the physical parameters of interest (reflection coefficient, transmission coefficient, neutron energy, and net neutron flux). The neutron transport problem in finite disturbed reactor with quadratic scattering is considered in investigating the shielding effectiveness, by using MAVRIC shielding module from SCALE6 programs package. The fractional parameter can be used to adjust the analysed data on neutron energy and flux, both for the theoretical model and the neutron transport application. (authors)
Criticality of neutron transport in a slab with finite reflectors
International Nuclear Information System (INIS)
Pao, C.V.
1978-01-01
The purpose of this paper is to investigate the subcriticality and the supercriticality for the neutron transport in a slab which is surrounded by two finite reflectors. The mathematical problem is to determine when the coupled boundary-value problem has or has no positive solution. It is shown under some explicit conditions on the material properties of the transport mediums and the size of the slab length that the coupled problem has a unique solution which insures the subcriticality of the system. It is also shown under some different conditions on the same physical quantities that the system cannot have a nonnegative solution when there is an external source, and it only has the trivial solution when there is no source in the system. This conclusion leads to the supercriticality of the system. Both upper and lower bounds for the critical length of the slab are explicitly given
Thermohydraulic stability coupled to the neutronic in a BWR
International Nuclear Information System (INIS)
Calleros M, G.; Zapata Y, M.; Gomez H, R.A.; Mendez M, A.; Castlllo D, R.
2006-01-01
In a BWR type reactor the phenomenon of the nuclear fission is presented, in which are liberated in stochastic form neutrons, originating that the population of the same ones varies in statistic form around a mean value. This variation will cause that when the neutron flow impacts on the neutron detectors, its are had as a result neutron flow signals with fluctuations around an average value. In this article it is shown that it conforms it lapses the time, this variations in the neutron flow (and therefore, in the flow signal due only to the fission), they presented oscillations inside a stable range, which won't be divergent. Considering that the BWR is characterized because boiling phenomena are presented, which affect the moderation of the neutrons, additional variations will be had in the signal coming from the neutron detectors, with relationship to the fission itself, which will be influenced by the feedback of the moderator's reactivity and of the temperature of the fuel pellet. Also, as the BWR it has coupled control systems to maintain the coolant level one and of the thermal power of the reactor, for each control action it was affected the neutron population. This means that the reactor could end up straying of a stable state condition. By it previously described, the study of the thermohydraulic stability coupled to the neutronic is complex. In this work it is shown the phenomenology, the mathematical models and the theoretical behavior associated to the stability of the BWR type reactor; the variables that affect it are identified, the models that reproduce the behavior of the thermohydraulic stability coupled to the neutronic, the way to maintain stable the reactor and the instrumentation that can settle to detect and to suppress uncertainties is described. In particular, is make reference to the evolution of the methods to maintain the stability of the reactor and the detection system and suppression of uncertainties implemented in the Laguna Verde
Analysis of coupled neutron-gamma radiations, applied to shieldings in multigroup albedo method
International Nuclear Information System (INIS)
Dunley, Leonardo Souza
2002-01-01
The principal mathematical tools frequently available for calculations in Nuclear Engineering, including coupled neutron-gamma radiations shielding problems, involve the full Transport Theory or the Monte Carlo techniques. The Multigroup Albedo Method applied to shieldings is characterized by following the radiations through distinct layers of materials, allowing the determination of the neutron and gamma fractions reflected from, transmitted through and absorbed in the irradiated media when a neutronic stream hits the first layer of material, independently of flux calculations. Then, the method is a complementary tool of great didactic value due to its clarity and simplicity in solving neutron and/or gamma shielding problems. The outstanding results achieved in previous works motivated the elaboration and the development of this study that is presented in this dissertation. The radiation balance resulting from the incidence of a neutronic stream into a shielding composed by 'm' non-multiplying slab layers for neutrons was determined by the Albedo method, considering 'n' energy groups for neutrons and 'g' energy groups for gammas. It was taken into account there is no upscattering of neutrons and gammas. However, it was considered that neutrons from any energy groups are able to produce gammas of all energy groups. The ANISN code, for an angular quadrature order S 2 , was used as a standard for comparison of the results obtained by the Albedo method. So, it was necessary to choose an identical system configuration, both for ANISN and Albedo methods. This configuration was six neutron energy groups and eight gamma energy groups, using three slab layers (iron aluminum - manganese). The excellent results expressed in comparative tables show great agreement between the values determined by the deterministic code adopted as standard and, the values determined by the computational program created using the Albedo method and the algorithm developed for coupled neutron
PHISICS multi-group transport neutronic capabilities for RELAP5
International Nuclear Information System (INIS)
Epiney, A.; Rabiti, C.; Alfonsi, A.; Wang, Y.; Cogliati, J.; Strydom, G.
2012-01-01
PHISICS is a neutronic code system currently under development at INL. Its goal is to provide state of the art simulation capability to reactor designers. This paper reports on the effort of coupling this package to the thermal hydraulic system code RELAP5. This will enable full prismatic core and system modeling and the possibility to model coupled (thermal-hydraulics and neutronics) problems with more options for 3D neutron kinetics, compared to the existing diffusion theory neutron kinetics module in RELAP5 (NESTLE). The paper describes the capabilities of the coupling and illustrates them with a set of sample problems. (authors)
Asymptotic time dependent neutron transport in multidimensional systems
International Nuclear Information System (INIS)
Nagy, M.E.; Sawan, M.E.; Wassef, W.A.; El-Gueraly, L.A.
1983-01-01
A model which predicts the asymptotic time behavior of the neutron distribution in multi-dimensional systems is presented. The model is based on the kernel factorization method used for stationary neutron transport in a rectangular parallelepiped. The accuracy of diffusion theory in predicting the asymptotic time dependence is assessed. The use of neutron pulse experiments for predicting the diffusion parameters is also investigated
Limits on Tensor Coupling from Neutron $\\beta$-Decay
Pattie Jr, Robert W.; Hickerson, Kevin P.; Young, Albert R.
2013-01-01
Limits on the tensor couplings generating a Fierz interference term, b, in mixed Gamow-Teller Fermi decays can be derived by combining data from measurements of angular correlation parameters in neutron decay, the neutron lifetime, and $G_{\\text{V}}=G_{\\text{F}} V_{ud}$ as extracted from measurements of the $\\mathcal{F}t$ values from the $0^{+} \\to 0^{+}$ superallowed decays dataset. These limits are derived by comparing the neutron $\\beta$-decay rate as predicted in the standard model with t...
Cosmic-ray neutron transport at a forest field site
DEFF Research Database (Denmark)
Andreasen, Mie; Jensen, Karsten Høgh; Desilets, Darin
2017-01-01
conceptualization is found to be significant. Modeling results show that the effect of canopy interception, soil chemistry and dry bulk density of litter and mineral soil on neutron intensity is small. On the other hand, the neutron intensity decreases significantly with added litter-layer thickness, especially......-ray neutron intensity is essential (e.g., the effect of vegetation, litter layer and soil type). In this study the environmental effect is examined by performing a sensitivity analysis using neutron transport modeling. We use a neutron transport model with various representations of the forest and different...
Assessment of transport effects in LMFBR safety neutronics
International Nuclear Information System (INIS)
Cahalan, J.E.; Ott, K.O.; Ferguson, D.R.
1976-01-01
A qualitative and quantitative assessment of the significance of neutron transport effects in LMFBR core disruptive accident analysis is presented. Material relocations which might cause important neutron transport behavior are identified. A quantitative measure of the error in the neutron flux is obtained from a consistent numerical comparison of transport and diffusion theory eigenvalue solutions for models of disrupted cores. A numerical technique for the prediction of transport eigenvalues and eigenvectors is formulated and applied. The technique is based on a modified diffusion theory which is fully capable of reproducing transport theory solutions
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.
Novel Parallel Numerical Methods for Radiation and Neutron Transport
International Nuclear Information System (INIS)
Brown, P N
2001-01-01
In many of the multiphysics simulations performed at LLNL, transport calculations can take up 30 to 50% of the total run time. If Monte Carlo methods are used, the percentage can be as high as 80%. Thus, a significant core competence in the formulation, software implementation, and solution of the numerical problems arising in transport modeling is essential to Laboratory and DOE research. In this project, we worked on developing scalable solution methods for the equations that model the transport of photons and neutrons through materials. Our goal was to reduce the transport solve time in these simulations by means of more advanced numerical methods and their parallel implementations. These methods must be scalable, that is, the time to solution must remain constant as the problem size grows and additional computer resources are used. For iterative methods, scalability requires that (1) the number of iterations to reach convergence is independent of problem size, and (2) that the computational cost grows linearly with problem size. We focused on deterministic approaches to transport, building on our earlier work in which we performed a new, detailed analysis of some existing transport methods and developed new approaches. The Boltzmann equation (the underlying equation to be solved) and various solution methods have been developed over many years. Consequently, many laboratory codes are based on these methods, which are in some cases decades old. For the transport of x-rays through partially ionized plasmas in local thermodynamic equilibrium, the transport equation is coupled to nonlinear diffusion equations for the electron and ion temperatures via the highly nonlinear Planck function. We investigated the suitability of traditional-solution approaches to transport on terascale architectures and also designed new scalable algorithms; in some cases, we investigated hybrid approaches that combined both
Orthogonal polynomials in neutron transport theory
Energy Technology Data Exchange (ETDEWEB)
Dehesa, J.S. (Granada Univ. (Spain). Facultad de Ciencias)
1982-01-01
The asymptotic average properties of zeros of the polynomials gsub(k)sup(m) (x), which play a fundamental role in neutron transport and radiative transfer theories, are investigated analytically in terms of the angular expansion coefficients wsub(k) of the scattering kernel for three wide classes of scattering models. In particular it is found that the scattering models of Eccleston-McCormick (J. Nucl. Energy.; 24:23 (1970)), Shultis et al (Nucl. Sci. Eng.; 59:53 (1976)) and Henyey-Greenstein (Astrophys. J.; 93:70 (1941)) belong in one of the above-mentioned classes, and their associated polynomials gsub(k)sup(m) (x) have the same asymptotic density of zeros.
Coupled electron-photon radiation transport
International Nuclear Information System (INIS)
Lorence, L.; Kensek, R.P.; Valdez, G.D.; Drumm, C.R.; Fan, W.C.; Powell, J.L.
2000-01-01
Massively-parallel computers allow detailed 3D radiation transport simulations to be performed to analyze the response of complex systems to radiation. This has been recently been demonstrated with the coupled electron-photon Monte Carlo code, ITS. To enable such calculations, the combinatorial geometry capability of ITS was improved. For greater geometrical flexibility, a version of ITS is under development that can track particles in CAD geometries. Deterministic radiation transport codes that utilize an unstructured spatial mesh are also being devised. For electron transport, the authors are investigating second-order forms of the transport equations which, when discretized, yield symmetric positive definite matrices. A novel parallelization strategy, simultaneously solving for spatial and angular unknowns, has been applied to the even- and odd-parity forms of the transport equation on a 2D unstructured spatial mesh. Another second-order form, the self-adjoint angular flux transport equation, also shows promise for electron transport
Interfacing MCNPX and McStas for simulation of neutron transport
DEFF Research Database (Denmark)
Klinkby, Esben Bryndt; Lauritzen, Bent; Nonbøl, Erik
2013-01-01
Stas[4, 5, 6, 7]. The coupling between the two simulation suites typically consists of providing analytical fits of MCNPX neutron spectra to McStas. This method is generally successful but has limitations, as it e.g. does not allow for re-entry of neutrons into the MCNPX regime. Previous work to resolve......Simulations of target-moderator-reflector system at spallation sources are conventionally carried out using Monte Carlo codes such as MCNPX[1] or FLUKA[2, 3] whereas simulations of neutron transport from the moderator and the instrument response are performed by neutron ray tracing codes such as Mc...... geometries, backgrounds, interference between beam-lines as well as shielding requirements along the neutron guides....
Hardware accelerated high performance neutron transport computation based on AGENT methodology
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
Design of a transportable high efficiency fast neutron spectrometer
Energy Technology Data Exchange (ETDEWEB)
Roecker, C., E-mail: calebroecker@berkeley.edu [Department of Nuclear Engineering, University of California at Berkeley, CA 94720 (United States); Bernstein, A.; Bowden, N.S. [Nuclear and Chemical Sciences Division, Lawrence Livermore National Laboratory, Livermore, CA 94550 (United States); Cabrera-Palmer, B. [Radiation and Nuclear Detection Systems, Sandia National Laboratories, Livermore, CA 94550 (United States); Dazeley, S. [Nuclear and Chemical Sciences Division, Lawrence Livermore National Laboratory, Livermore, CA 94550 (United States); Gerling, M.; Marleau, P.; Sweany, M.D. [Radiation and Nuclear Detection Systems, Sandia National Laboratories, Livermore, CA 94550 (United States); Vetter, K. [Department of Nuclear Engineering, University of California at Berkeley, CA 94720 (United States); Nuclear Science Division, Lawrence Berkeley National Laboratory, Berkeley, CA 94720 (United States)
2016-08-01
A transportable fast neutron detection system has been designed and constructed for measuring neutron energy spectra and flux ranging from tens to hundreds of MeV. The transportability of the spectrometer reduces the detector-related systematic bias between different neutron spectra and flux measurements, which allows for the comparison of measurements above or below ground. The spectrometer will measure neutron fluxes that are of prohibitively low intensity compared to the site-specific background rates targeted by other transportable fast neutron detection systems. To measure low intensity high-energy neutron fluxes, a conventional capture-gating technique is used for measuring neutron energies above 20 MeV and a novel multiplicity technique is used for measuring neutron energies above 100 MeV. The spectrometer is composed of two Gd containing plastic scintillator detectors arranged around a lead spallation target. To calibrate and characterize the position dependent response of the spectrometer, a Monte Carlo model was developed and used in conjunction with experimental data from gamma ray sources. Multiplicity event identification algorithms were developed and used with a Cf-252 neutron multiplicity source to validate the Monte Carlo model Gd concentration and secondary neutron capture efficiency. The validated Monte Carlo model was used to predict an effective area for the multiplicity and capture gating analyses. For incident neutron energies between 100 MeV and 1000 MeV with an isotropic angular distribution, the multiplicity analysis predicted an effective area of 500 cm{sup 2} rising to 5000 cm{sup 2}. For neutron energies above 20 MeV, the capture-gating analysis predicted an effective area between 1800 cm{sup 2} and 2500 cm{sup 2}. The multiplicity mode was found to be sensitive to the incident neutron angular distribution.
Energy Technology Data Exchange (ETDEWEB)
Maldonado-Velázquez, M. [Posgrado en Ciencias Físicas, Universidad Nacional Autónoma de México, 04510 (Mexico); Barrón-Palos, L., E-mail: libertad@fisica.unam.mx [Instituto de Física, Universidad Nacional Autónoma de México, Apartado Postal 20-364, 01000 (Mexico); Crawford, C. [University of Kentucky, Lexington, KY 40506 (United States); Snow, W.M. [Indiana University, Bloomington, IN 47405 (United States)
2017-05-11
The neutron spin is a critical degree of freedom for many precision measurements using low-energy neutrons. Fundamental symmetries and interactions can be studied using polarized neutrons. Parity-violation (PV) in the hadronic weak interaction and the search for exotic forces that depend on the relative spin and velocity, are two questions of fundamental physics that can be studied via the neutron spin rotations that arise from the interaction of polarized cold neutrons and unpolarized matter. The Neutron Spin Rotation (NSR) collaboration developed a neutron polarimeter, capable of determining neutron spin rotations of the order of 10{sup −7} rad per meter of traversed material. This paper describes two key components of the NSR apparatus, responsible for the transport and manipulation of the spin of the neutrons before and after the target region, which is surrounded by magnetic shielding and where residual magnetic fields need to be below 100 μG. These magnetic field devices, called input and output coils, provide the magnetic field for adiabatic transport of the neutron spin in the regions outside the magnetic shielding while producing a sharp nonadiabatic transition of the neutron spin when entering/exiting the low-magnetic-field region. In addition, the coils are self contained, forcing the return magnetic flux into a compact region of space to minimize fringe fields outside. The design of the input and output coils is based on the magnetic scalar potential method.
Coupling of transport and geochemical models
International Nuclear Information System (INIS)
Noy, D.J.
1986-01-01
This report considers mass transport in the far-field of a radioactive waste repository, and detailed geochemical modelling of the ground-water in the near-field. A parallel approach to this problem of coupling transport and geochemical codes is the subject of another CEC report (ref. EUR 10226). Both studies were carried out in the framework of the CEC project MIRAGE. (Migration of radionuclides in the geosphere)
Neutron Imaging with Timepix Coupled Lithium Indium Diselenide
Directory of Open Access Journals (Sweden)
Elan Herrera
2017-12-01
Full Text Available The material lithium indium diselenide, a single crystal neutron sensitive semiconductor, has demonstrated its capabilities as a high resolution imaging device. The sensor was prepared with a 55 μ m pitch array of gold contacts, designed to couple with the Timepix imaging ASIC. The resulting device was tested at the High Flux Isotope Reactor, demonstrating a response to cold neutrons when enriched in 95% 6 Li. The imaging system performed a series of experiments resulting in a <200 μ m resolution limit with the Paul Scherrer Institute (PSI Siemens star mask and a feature resolution of 34 μ m with a knife-edge test. Furthermore, the system was able to resolve the University of Tennessee logo inscribed into a 3D printed 1 cm 3 plastic block. This technology marks the application of high resolution neutron imaging using a direct readout semiconductor.
Technical notes. Spherical harmonics approximations of neutron transport
Energy Technology Data Exchange (ETDEWEB)
Demeny, A.; Dede, K.M.; Erdei, K.
1976-12-01
A double-range spherical harmonics approximation obtained by expanding the angular flux separately in the two regions combined with the conventional single-range spherical harmonics is found to give superior description of neutron transport.
A random walk approach to stochastic neutron transport
International Nuclear Information System (INIS)
Mulatier, Clelia de
2015-01-01
One of the key goals of nuclear reactor physics is to determine the distribution of the neutron population within a reactor core. This population indeed fluctuates due to the stochastic nature of the interactions of the neutrons with the nuclei of the surrounding medium: scattering, emission of neutrons from fission events and capture by nuclear absorption. Due to these physical mechanisms, the stochastic process performed by neutrons is a branching random walk. For most applications, the neutron population considered is very large, and all physical observables related to its behaviour, such as the heat production due to fissions, are well characterised by their average values. Generally, these mean quantities are governed by the classical neutron transport equation, called linear Boltzmann equation. During my PhD, using tools from branching random walks and anomalous diffusion, I have tackled two aspects of neutron transport that cannot be approached by the linear Boltzmann equation. First, thanks to the Feynman-Kac backward formalism, I have characterised the phenomenon of 'neutron clustering' that has been highlighted for low-density configuration of neutrons and results from strong fluctuations in space and time of the neutron population. Then, I focused on several properties of anomalous (non-exponential) transport, that can model neutron transport in strongly heterogeneous and disordered media, such as pebble-bed reactors. One of the novel aspects of this work is that problems are treated in the presence of boundaries. Indeed, even though real systems are finite (confined geometries), most of previously existing results were obtained for infinite systems. (author) [fr
Strongly Coupled Chameleons and the Neutronic Quantum Bouncer
International Nuclear Information System (INIS)
Brax, Philippe; Pignol, Guillaume
2011-01-01
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 > or approx. 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 > or approx. 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 11 , is already 3 orders of magnitude better than the upper bound, β 14 , from precision tests of atomic spectra.
Numerical solution of time dependent neutron transport equation. An application
International Nuclear Information System (INIS)
Barroso, Dalton Ellery Girao
2000-01-01
In this work we show a simple method to solve numerically the time-dependent neutron transport equation which is a simple extension of the numerical methods used to solve the time-independent static transport equation. This is possible because the time-discretized transport equation has the same form as the time-independent transport equation, with only some additional terms. A general outline of the method is given and used to evaluate the neutron flux in a microexplosion calculation of a highly compressed micro fissile system composed by DT-Pu-Be microsphere. (author)
Scattered Neutron Tomography Based on A Neutron Transport Inverse Problem
International Nuclear Information System (INIS)
William Charlton
2007-01-01
Neutron radiography and computed tomography are commonly used techniques to non-destructively examine materials. Tomography refers to the cross-sectional imaging of an object from either transmission or reflection data collected by illuminating the object from many different directions
Scattered Neutron Tomography Based on A Neutron Transport Inverse Problem
Energy Technology Data Exchange (ETDEWEB)
William Charlton
2007-07-01
Neutron radiography and computed tomography are commonly used techniques to non-destructively examine materials. Tomography refers to the cross-sectional imaging of an object from either transmission or reflection data collected by illuminating the object from many different directions.
Coupled transport in field-reversed configurations
Steinhauer, L. C.; Berk, H. L.; TAE Team
2018-02-01
Coupled transport is the close interconnection between the cross-field and parallel fluxes in different regions due to topological changes in the magnetic field. This occurs because perpendicular transport is necessary for particles or energy to leave closed field-line regions, while parallel transport strongly affects evolution of open field-line regions. In most toroidal confinement systems, the periphery, namely, the portion with open magnetic surfaces, is small in thickness and volume compared to the core plasma, the portion with closed surfaces. In field-reversed configurations (FRCs), the periphery plays an outsized role in overall confinement. This effect is addressed by an FRC-relevant model of coupled particle transport that is well suited for immediate interpretation of experiments. The focus here is particle confinement rather than energy confinement since the two track together in FRCs. The interpretive tool yields both the particle transport rate χn and the end-loss time τǁ. The results indicate that particle confinement depends on both χn across magnetic surfaces throughout the plasma and τǁ along open surfaces and that they provide roughly equal transport barriers, inhibiting particle loss. The interpretation of traditional FRCs shows Bohm-like χn and inertial (free-streaming) τǁ. However, in recent advanced beam-driven FRC experiments, χn approaches the classical rate and τǁ is comparable to classic empty-loss-cone mirrors.
A finite element method for neutron transport
International Nuclear Information System (INIS)
Ackroyd, R.T.
1978-01-01
A variational treatment of the finite element method for neutron transport is given based on a version of the even-parity Boltzmann equation which does not assume that the differential scattering cross-section has a spherical harmonic expansion. The theory of minimum and maximum principles is based on the Cauchy-Schwartz equality and the properties of a leakage operator G and a removal operator C. For systems with extraneous sources, two maximum and one minimum principles are given in boundary free form, to ease finite element computations. The global error of an approximate variational solution is given, the relationship of one the maximum principles to the method of least squares is shown, and the way in which approximate solutions converge locally to the exact solution is established. A method for constructing local error bounds is given, based on the connection between the variational method and the method of the hypercircle. The source iteration technique and a maximum principle for a system with extraneous sources suggests a functional for a variational principle for a self-sustaining system. The principle gives, as a consequence of the properties of G and C, an upper bound to the lowest eigenvalue. A related functional can be used to determine both upper and lower bounds for the lowest eigenvalue from an inspection of any approximate solution for the lowest eigenfunction. The basis for the finite element is presented in a general form so that two modes of exploitation can be undertaken readily. The model can be in phase space, with positional and directional co-ordinates defining points of the model, or it can be restricted to the positional co-ordinates and an expansion in orthogonal functions used for the directional co-ordinates. Suitable sets of functions are spherical harmonics and Walsh functions. The latter set is appropriate if a discrete direction representation of the angular flux is required. (author)
Direct discrete method and its application to neutron transport problems
Directory of Open Access Journals (Sweden)
Vosoughi Naser
2003-01-01
Full Text Available The objective of this paper is to introduce a new direct method for neutronic calculations. This method, called direct discrete method, is simpler than the application of the neutron transport equation and more compatible with the physical meanings of the problem. The method, based on the physics of the problem, initially runs through meshing of the desired geometry. Next, the balance equation for each mesh interval is written. Considering the connection between the mesh intervals, the final discrete equation series are directly obtained without the need to pass through the set up of the neutron transport differential equation first. In this paper, one and multigroup neutron transport discrete equation has been produced for a cylindrical shape fuel element with and without the associated clad and the coolant regions each with two different external boundary conditions. The validity of the results from this new method is tested against the results obtained by the MCNP-4B and the ANISN codes.
Interface requirements to couple thermal-hydraulic codes to 3D neutronic codes
Energy Technology Data Exchange (ETDEWEB)
Langenbuch, S.; Austregesilo, H.; Velkov, K. [GRS, Garching (Germany)] [and others
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.
International Nuclear Information System (INIS)
Zazula, J.M.
1983-01-01
The general purpose code BALTORO was written for coupling the three-dimensional Monte-Carlo /MC/ with the one-dimensional Discrete Ordinates /DO/ radiation transport calculations. The quantity of a radiation-induced /neutrons or gamma-rays/ nuclear effect or the score from a radiation-yielding nuclear effect can be analysed in this way. (author)
International Nuclear Information System (INIS)
Tai, D.; Underhill, G.K.
1975-01-01
The Boltzmann integrodifferential neutron transport equation has been converted to an integral equation which incorporates the isotropic neutron bath boundary condition. The resulting integral equation is solved using the discrete ordinates method, resulting in solutions for the spatially-dependent and -independent fluxes in terms of transport probabilities and the neutron emission density. The transport probabilities and the lethargy-dependence solution are evaluated using a normalization condition and a neutron conservation equation, respectively, to correct for inherent error propagation. The formalism is applied to the calculation of uranium resonance integrals for a spherical, two-region lump consisting of a spherical absorber surrounded by a spherical cadmium cover. Calculated and experimental results for uranium-235 fission and uranium-238 capture resonance integrals compare favorably. 11 references. (U.S.)
Application of singular eigenfunctions method of neutron transport theory
International Nuclear Information System (INIS)
Simovicj, R.
1974-11-01
A possibility of applying analitical method of neutron transport theory was investigated in research of processes governed by linearized Boltzmann equation, especially in semiconducting media. Analitical singular eigenfunctions method was developed and improved. It was applied in solving the electron transport equation for nonpolar semiconductors in case of constant high electrical field. Energy and angular distribution of high energy electrons was obtained
Verification of fast neutron spectrum calculation in coupled system HERBE
International Nuclear Information System (INIS)
Avdic, S.; Pesic, M.; Marinkovic, P.
1995-01-01
A high-resolution semiconductor spectrometer filled with 3 He gas, in diode coincidence arrangement, is applied to measure neutron spectrum in the centre of the fast core of the coupled fast-thermal system HERBE in the 'Vinca' Institute. The neutron spectrum is evaluated from measured pulse height distribution by using the HE3 computer code developed in the Nuclear Engineering Laboratory of the Institute of Nuclear Sciences VINCA. Experimental results are compared with the relevant multigroup calculations in the energy range from 2.5 MeV to 10.5 MeV. The measured spectrum provides a sufficient overlapping with the calculated one and no serious divergence is found in the measured energy range. (author)
The isotope density inverse problem in multigroup neutron transport
International Nuclear Information System (INIS)
Zazula, J.M.
1981-01-01
The inverse problem for stationary multigroup anisotropic neutron transport is discussed in order to search for isotope densities in multielement medium. The spatial- and angular-integrated form of neutron transport equation, in terms of the flux in a group - density of an element spatial correlation, leads to a set of integral functionals for the densities weighted by the group fluxes. Some methods of approximation to make the problem uniquently solvable are proposed. Particularly P 0 angular flux information and the spherically-symetrical geometry of an infinite medium are considered. The numerical calculation using this method related to sooner evaluated direct problem data gives promising agreement with primary densities. This approach would be the basis for further application in an elemental analysis of a medium, using an isotopic neutron source and a moving, energy-dependent neutron detector. (author)
Neutron and gamma-ray transport experiments in liquid air
International Nuclear Information System (INIS)
Farley, W.E.
1976-01-01
Accurate estimates of neutron and gamma radiations from a nuclear explosion and their subsequent transport through the atmosphere are vital to nuclear-weapon employment studies: i.e., for determining safety radii for aircraft crews, casualty and collateral-damage risk radii for tactical weapons, and the kill range from a high-yield defensive burst for a maneuvering reentry vehicle. Radiation transport codes, such as the Laboratory's TARTNP, are used to calculate neutron and gamma fluences. Experiments have been performed to check and update these codes. Recently, a 1.3-m-radius liquid-air (21 percent oxygen) sphere, with a pulsed source of 14-MeV neutrons at its center, was used to measure the fluence and spectra of emerging neutrons and secondary gamma rays. Comparison of measured radiation dose with TARTNP showed agreement within 10 percent
Neutronics - thermal-hydraulics coupling: application to the helium-cooled fast reactor
International Nuclear Information System (INIS)
Vaiana, F.
2009-11-01
This thesis focuses on the study of interactions between neutron-kinetics and thermal-hydraulics. Neutron-kinetics allow to calculate the power in a nuclear reactor and the temperature evolution of materials where this power is deposited is known thanks to thermal-hydraulics. Moreover, when the temperatures evolve, the densities and cross sections change. These two disciplines are thus coupled. The first part of this work corresponds to the study and development of a method which allows to simulate transients in nuclear reactors and especially with a Monte-Carlo code for neutron-kinetics. An algorithm for the resolution of the neutron transport equation has been established and validated with a benchmark. In thermal-hydraulics, a porous media approach, based on another thesis, is considered. This gives the opportunity to solve the equations on the whole core without unconscionable computation time. Finally, a theoretical study has been performed on the statistical uncertainties which result from the use of a Monte-Carlo code and which spread from the reactivity to the power and from the power to the temperatures. The second part deals with the study of a misplaced control rod withdrawing in a GFR (helium-cooled fast reactor), a fourth generation reactor. Some models allowing to calculate neutron-kinetics and thermal-hydraulics in the core (which contains assemblies built up with fuel plates) were defined. In thermal-hydraulics, a model for the core based on the porous media approach and a fuel plate homogenization model have been set up. A similar homogenization model has been studied for neutron-kinetics. Finally, the control rod withdrawing transient where we can observe the power raising and the stabilisation by thermal feedback has been performed with the Monte-Carlo code Tripoli for neutron-kinetics and the code Trio-U for thermal-hydraulics. (author)
Transportable, Low-Dose Active Fast-Neutron Imaging
Energy Technology Data Exchange (ETDEWEB)
Mihalczo, John T. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Wright, Michael C. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); McConchie, Seth M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Archer, Daniel E. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Palles, Blake A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)
2017-08-01
This document contains a description of the method of transportable, low-dose active fast-neutron imaging as developed by ORNL. The discussion begins with the technique and instrumentation and continues with the image reconstruction and analysis. The analysis discussion includes an example of how a gap smaller than the neutron production spot size and detector size can be detected and characterized depending upon the measurement time.
Measurements of anomalous neutron transport in bulk graphite
International Nuclear Information System (INIS)
Bowman, C.D.; Smith, G.A.; Vogelaar, B.; Howell, C.R.; Bilpuch, E.G.; Tornow, W.
2003-01-01
The neutron absorption of bulk granular graphite has been measured in a classical exponential diffusion experiment. Our first measurements of April 2002 implementing both exponential decay and pulsed die-away experiments and using the TUNL pulsed accelerator at Duke University as a neutron source indicated a capture cross section for graphite a striking factor of three lower than the measured value for carbon of 3.4 millibarns. Therefore a new exponential experiment with an improved geometry enabling greater accuracy has been performed giving an apparent cross section for carbon in the form of bulk granular graphite of less than 0.5 millibarns. This result confirms our first result and is also consistent with less than one part per million of boron in our graphite. The bulk density of the graphite is 1.02 compared with the actual particle density of 1.60 indicating a packing fraction of 0.64 or a void fraction of 0.36. We suspect that the apparent suppression of absorption in bulk graphite may be associated with the strong coherent diffraction of neutrons that dominates neutron transport in graphite. Coherent diffraction has never been taken into account in graphite reactor design and no neutron transport code including general use codes such as MCNP incorporate diffraction effects even though diffraction dominates many practical thermal neutron transport problems. (orig.)
Calculated characteristics of subcritical assembly with anisotropic transport of neutrons
International Nuclear Information System (INIS)
Gorin, N.V.; Lipilina, E.N.; Lyutov, V.D.; Saukov, A.I.
2003-01-01
There was considered possibility of creating enough sub-critical system that multiply neutron fluence from a primary source by many orders. For assemblies with high neutron tie between parts, it is impossible. That is why there was developed a construction consisting of many units (cascades) having weak feedback with preceding cascades. The feedback attenuation was obtained placing layers of slow neutron absorber and moderators between the cascades of fission material. Anisotropy of fast neutron transport through the layers was used. The system consisted of many identical cascades aligning one by another. Each cascade consists of layers of moderator, fissile material and absorber of slow neutrons. The calculations were carried out using the code MCNP.4a with nuclear data library ENDF/B5. In this construction neutrons spread predominantly in one direction multiplying in each next fissile layer, and they attenuate considerably in the opposite direction. In a calculated construction, multiplication factor of one cascade is about 1.5 and multiplication factor of whole construction composed of n cascades is 1.5 n . Calculated keff value is 0.9 for one cascade and does not exceed 0.98 for a system containing any number of cascades. Therefore the assembly is always sub-critical and therefore it is safe in respect of criticality. There was considered using such a sub-critical assembly to create a powerful neutron fluence for neutron boron-capturing therapy. The system merits and demerits were discussed. (authors)
Coupling of transport and geochemical models
International Nuclear Information System (INIS)
Noy, D.J.
1985-01-01
This contract stipulated separate pieces of work to consider mass transport in the far-field of a repository, and more detailed geochemical modelling of the groundwater in the near-field. It was envisaged that the far-field problem would be tackled by numerical solutions to the classical advection-diffusion equation obtained by the finite element method. For the near-field problem the feasibility of coupling existing geochemical equilibrium codes to the three dimensional groundwater flow codes was to be investigated. This report is divided into two sections with one part devoted to each aspect of this contract. (author)
An iterative method for solving neutron transport equation
International Nuclear Information System (INIS)
Simovic, R.
1988-01-01
Assuming a plane geometry and isotropic form of the neutron scattering function a new iterative method for solving the one-velocity transport equation is developed. The basic point of this method is the definition of the neutron fluxes Φ n± (x, μ, μ 0 ) representing the space dependent angular distributions of neutrons scattered n-times in directions μ 0. This makes possible to construct a new system for successive calculation of Φ n± (x, μ, μ 0 ) starting with the flux of un-collided neutrons. This treatment was shown to be more efficient than the ordinary one. As examples, the infinite medium Green functions and reflection coefficients of half space were calculated and analyzed. (author)
Importance estimation in Monte Carlo modelling of neutron and photon transport
International Nuclear Information System (INIS)
Mickael, M.W.
1992-01-01
The estimation of neutron and photon importance in a three-dimensional geometry is achieved using a coupled Monte Carlo and diffusion theory calculation. The parameters required for the solution of the multigroup adjoint diffusion equation are estimated from an analog Monte Carlo simulation of the system under investigation. The solution of the adjoint diffusion equation is then used as an estimate of the particle importance in the actual simulation. This approach provides an automated and efficient variance reduction method for Monte Carlo simulations. The technique has been successfully applied to Monte Carlo simulation of neutron and coupled neutron-photon transport in the nuclear well-logging field. The results show that the importance maps obtained in a few minutes of computer time using this technique are in good agreement with Monte Carlo generated importance maps that require prohibitive computing times. The application of this method to Monte Carlo modelling of the response of neutron porosity and pulsed neutron instruments has resulted in major reductions in computation time. (Author)
International Nuclear Information System (INIS)
Stefanovic, D.
1975-09-01
The research work of this contract was oriented towards the study of different methods in neutron transport theory. Authors studied analytical solution of the neutron slowing down transport equation and extension of this solution to include the energy dependence of the anisotropy of neutron scattering. Numerical solution of the fast and resonance transport equation for the case of mixture of scatterers including inelastic effects were also reviewed. They improved the existing formalism for treating the scattering of neutrons on water molecules; Identifying modal analysis as the Galerkin method, general conditions for modal technique applications have been investigated. Inverse problems in transport theory were considered. They obtained the evaluation of an advanced level distribution function, made improvement of the standard formalism for treating the inelastic scattering and development of a cluster nuclear model for this evaluation. Authors studied the neutron transport treatment in space energy groups for criticality calculation of a reactor core, and development of the Monte Carlo sampling scheme from the neutron transport equation
Neutron transport calculations of some fast critical assemblies
Energy Technology Data Exchange (ETDEWEB)
Martinez-Val Penalosa, J. A.
1976-07-01
To analyse the influence of the input variables of the transport codes upon the neutronic results (eigenvalues, generation times, . . . ) four Benchmark calculations have been performed. Sensitivity analysis have been applied to express these dependences in a useful way, and also to get an unavoidable experience to carry out calculations achieving the required accuracy and doing them in practical computing times. (Author) 29 refs.
International Nuclear Information System (INIS)
Muzychka, A.Yu.; Pokotilovski, Yu.N.
1996-01-01
The results are presented of Monte Carlo simulation of the transport of very cold (VCN) and ultracold neutrons (UCN) in straight and curved vertical neutron guides with a rectangular cross section in the presence of neutron losses due to neutron capture and diffuse scattering on imperfectly smooth reflecting surface of the guide wall. The gravitational neutron deceleration and bending of neutron trajectories are rigorously taken into account. The nonstationary storage of UCN in experimental chambers is modelled for a low periodic or a periodic pulse neutron source. (orig.)
International Nuclear Information System (INIS)
Thiagu Supramaniam
2007-01-01
The aim of this research was to propose a new neutron collimator design for thermal neutron radiography facility using tangential beam port of PUSPATI TRIGA Mark II reactor, Malaysia Institute of Nuclear Technology Research (MINT). Best geometry and materials for neutron collimator were chosen in order to obtain a uniform beam with maximum thermal neutron flux, high L/ D ratio, high neutron to gamma ratio and low beam divergence with high resolution. Monte Carlo N-particle Transport Code version 5 (MCNP 5) was used to optimize six neutron collimator components such as beam port medium, neutron scatterer, neutron moderator, gamma filter, aperture and collimator wall. The reactor and tangential beam port setup in MCNP5 was plotted according to its actual sizes. A homogeneous reactor core was assumed and population control method of variance reduction technique was applied by using cell importance. The comparison between experimental results and simulated results of the thermal neutron flux measurement of the bare tangential beam port, shows that both graph obtained had similar pattern. This directly suggests the reliability of MCNP5 in order to obtained optimal neutron collimator parameters. The simulated results of the optimal neutron medium, shows that vacuum was the best medium to transport neutrons followed by helium gas and air. The optimized aperture component was boral with 3 cm thickness. The optimal aperture center hole diameter was 2 cm which produces 88 L/ D ratio. Simulation also shows that graphite neutron scatterer improves thermal neutron flux while reducing fast neutron flux. Neutron moderator was used to moderate fast and epithermal neutrons in the beam port. Paraffin wax with 90 cm thick was bound to be the best neutron moderator material which produces the highest thermal neutron flux at the image plane. Cylindrical shape high density polyethylene neutron collimator produces the highest thermal neutron flux at the image plane rather than divergent
International Nuclear Information System (INIS)
Miller, T.M.; Pevey, R.E.; Lillie, R.A.; Johnson, J.O.
2000-01-01
A detailed radiation transport analysis of the Spallation Neutron Source (SNS) shutters is important for the construction of the SNS because of its impact on conventional facility design, normal operation of the facility, and maintenance operations. Thus far the analysis of the SNS shutter travel gaps has been completed. This analysis was performed using coupled Monte Carlo and multi-dimensional discrete ordinates calculations
Homogenization of the critically spectral equation in neutron transport
International Nuclear Information System (INIS)
Allaire, G.; Paris-6 Univ., 75; Bal, G.
1998-01-01
We address the homogenization of an eigenvalue problem for the neutron transport equation in a periodic heterogeneous domain, modeling the criticality study of nuclear reactor cores. We prove that the neutron flux, corresponding to the first and unique positive eigenvector, can be factorized in the product of two terms, up to a remainder which goes strongly to zero with the period. On terms is the first eigenvector of the transport equation in the periodicity cell. The other term is the first eigenvector of a diffusion equation in the homogenized domain. Furthermore, the corresponding eigenvalue gives a second order corrector for the eigenvalue of the heterogeneous transport problem. This result justifies and improves the engineering procedure used in practice for nuclear reactor cores computations. (author)
Graphical User Interface for Simplified Neutron Transport Calculations
Energy Technology Data Exchange (ETDEWEB)
Schwarz, Randolph; Carter, Leland L
2011-07-18
A number of codes perform simple photon physics calculations. The nuclear industry is lacking in similar tools to perform simplified neutron physics shielding calculations. With the increased importance of performing neutron calculations for homeland security applications and defense nuclear nonproliferation tasks, having an efficient method for performing simple neutron transport calculations becomes increasingly important. Codes such as Monte Carlo N-particle (MCNP) can perform the transport calculations; however, the technical details in setting up, running, and interpreting the required simulations are quite complex and typically go beyond the abilities of most users who need a simple answer to a neutron transport calculation. The work documented in this report resulted in the development of the NucWiz program, which can create an MCNP input file for a set of simple geometries, source, and detector configurations. The user selects source, shield, and tally configurations from a set of pre-defined lists, and the software creates a complete MCNP input file that can be optionally run and the results viewed inside NucWiz.
TRINIDY: Transport of ions and neutrons in dynamic materials
Spencer, Joshua B.
The TRansport of Ions and Neutrons In DYnamic (TRINIDY) materials code is a new code designed to study the effects of high fluence ion and neutron radiation on solid surfaces. This is done in a quasi-deterministic way, in that the transport of pseudo-particles within target material is accomplished via a Monte Carlo approach while the changes within the target are calculated deterministically by use of a one-dimensional Lagrangian mesh into which each of the tracked pseudo-particles are either deposited or removed. After each cycle the mesh is allowed to relax to a solid state areal density adjusted for its new constituency. As a natural corollary to the change in material compositions in each mesh element comes the resultant change in thickness of the target. Within TRINIDY charged particles are transported by means of a Binary Collision Approximation (BCA) where the elastic nuclear and inelastic electronic stopping forces are decoupled in such a way that the projectile only interacts with one target atom at a time. TRINIDY builds on the legacy of the Transport of Ions in Matter (TRIM), TRIM-SP and TRIDYN codes, in that it uses Biersack's analytic approximation to the quantum scattering integral and a screened coulomb potential as the basic for the charged particle transport. The neutron transport within TRINIDY is based on 32-group elastic scattering and total absorption cross-section data which has been derived from the ENDF7 continuous neutron data sets for each of the naturally occurring elements Hydrogen through Uranium. This work is comprised of essentially three sections. First, there is a detailed technical description of the science behind TRINIDY. Secondly there will be a complete write-up of the validation and verification work done during the development of TRINIDY. Lastly, a series of practical demonstration of particular interest to the semi-conductor industry are presented to exemplify the use of TRINIDY within the realm of applied materials
A method for solving neutron transport equation
International Nuclear Information System (INIS)
Dimitrijevic, Z.
1993-01-01
The procedure for solving the transport equation by directly integrating for case one-dimensional uniform multigroup medium is shown. The solution is expressed in terms of linear combination of function H n (x,μ), and the coefficient is determined from given conditions. The solution is applied for homogeneous slab of critical thickness. (author)
International Nuclear Information System (INIS)
Guertin, Chantal
1995-01-01
This thesis is part of the validation process of using coupled 3D neutronics and thermal-hydraulics codes for studying accidental situations with boiling. First part is dedicated to a numerical stability analysis of neutronics and thermal-hydraulics coupled schemes. Both explicit and semi-implicit coupling schemes were applied to solve the set of equations describing the linearized neutronics and thermal-hydraulics of point reactor. Point reactor modelling was preferred to obtain analytical expressions of eigenvalues of the discretized Systems. Stability criteria, based on eigenvalues, was calculated as well as neutronic and thermalhydraulic responses of the System following insertion of a reactivity step. Results show no severe restriction of time domain, stability wise. Actual transient calculations using coupled neutronics and thermal-hydraulics codes, like COCCINELLE and THYC developed at Electricite de France, do not show stability problems. Second part introduces surface spline as a new neutronic feedback model. The cross influences of feedback parameters is now taken into account. Moderator temperature and density were modeled. This method, simple and accurate, allows an homogeneous description of cross-sections overall operating reactor situations including accidents with boiling. (author) [fr
International Nuclear Information System (INIS)
Pirouzmand, Ahmad; Hadad, Kamal
2012-01-01
Highlights: ► This paper describes the solution of time-dependent neutron transport equation. ► We use a novel method based on cellular neural networks (CNNs) coupled with the spherical harmonics method. ► We apply the CNN model to simulate step and ramp perturbation transients in a core. ► The accuracy and capabilities of the CNN model are examined for x–y geometry. - Abstract: In an earlier paper we utilized a novel method using cellular neural networks (CNNs) coupled with spherical harmonics method to solve the steady state neutron transport equation in x–y geometry. Here, the previous work is extended to the study of time-dependent neutron transport equation. To achieve this goal, an equivalent electrical circuit based on a second-order form of time-dependent neutron transport equation and one equivalent group of neutron precursor density is obtained by the CNN method. The CNN model is used to simulate step and ramp perturbation transients in a typical 2D core.
KAMCCO, a reactor physics Monte Carlo neutron transport code
International Nuclear Information System (INIS)
Arnecke, G.; Borgwaldt, H.; Brandl, V.; Lalovic, M.
1976-06-01
KAMCCO is a 3-dimensional reactor Monte Carlo code for fast neutron physics problems. Two options are available for the solution of 1) the inhomogeneous time-dependent neutron transport equation (census time scheme), and 2) the homogeneous static neutron transport equation (generation cycle scheme). The user defines the desired output, e.g. estimates of reaction rates or neutron flux integrated over specified volumes in phase space and time intervals. Such primary quantities can be arbitrarily combined, also ratios of these quantities can be estimated with their errors. The Monte Carlo techniques are mostly analogue (exceptions: Importance sampling for collision processes, ELP/MELP, Russian roulette and splitting). Estimates are obtained from the collision and track length estimators. Elastic scattering takes into account first order anisotropy in the center of mass system. Inelastic scattering is processed via the evaporation model or via the excitation of discrete levels. For the calculation of cross sections, the energy is treated as a continuous variable. They are computed by a) linear interpolation, b) from optionally Doppler broadened single level Breit-Wigner resonances or c) from probability tables (in the region of statistically distributed resonances). (orig.) [de
Neutron and gamma transport effects by heterogeneous core designs. [LMFBR
Energy Technology Data Exchange (ETDEWEB)
Lam, S.K.
1977-01-01
The use of diffusion theory for the prediction of power production near a reactor core-blanket interface and the assumption that gammas are absorbed in situ can lead to substantial errors. This is primarily due to the breakdown of Fick's law for neutron diffusion near the core-blanket boundary, and the gamma leakage from the core into the blanket. These considerations are more pronounced in a situation where a large number of internal blanket assemblies are present, such as in the large heterogeneous core designs. The power distribution is studied for both fission and gamma heating in a large heterogeneous LMFBR with 3 core zones separated by 2 internal blanket zones. Comparisons are made between diffusion and transport theory for neutronics calculations, and between in-situ absorption and rigorous transport theory calculation for gamma heating.
Mathematical models for volume rendering and neutron transport
International Nuclear Information System (INIS)
Max, N.
1994-09-01
This paper reviews several different models for light interaction with volume densities of absorbing, glowing, reflecting, or scattering material. They include absorption only, glow only, glow and absorption combined, single scattering of external illumination, and multiple scattering. The models are derived from differential equations, and illustrated on a data set representing a cloud. They are related to corresponding models in neutron transport. The multiple scattering model uses an efficient method to propagate the radiation which does not suffer from the ray effect
Neutron Star Structure in the Presence of Conformally Coupled Scalar Fields
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.
Transport of D-D fusion neutrons in thick concrete
International Nuclear Information System (INIS)
Ku, L.P.; Kolibal, J.G.
1982-07-01
By altering the collision mechanism in the numerical transport calculations, and by constructing an analytical model based on age-diffusion theory, the outstanding feature in the life history of D-D fusion neutrons penetrating deeply into ordinary concrete is shown to be the transport in the 2.3 MeV oxygen anti-resonance. This result is used to assess the impact of the cross-section uncertainties and the uncertainties due to variations in the D-D fusion spectrum and temperature
Ogawa, Y; Kosugi, N; Iwasa, H; Furusaka, M; Watanabe, N
1999-01-01
In order to obtain higher cold neutron intensity from a coupled liquid-hydrogen moderator with a premoderator for pulsed cold neutron sources, we examined a partial enhancement method, namely, narrow beam extraction for both a flat liquid-hydrogen moderator and a single-groove one. Combined with the narrow beam extraction, which is especially suitable for small-angle scattering and neutron reflectometry experiments, a single-groove moderator provides higher intensity, by about 30%, than a flat-surface moderator at the region of interest on a viewed surface. The effect of double-side beam extraction from such moderators on the intensity gain factor is also discussed. (author)
Coupled neutronics - thermal-hydraulics programs for SCWRS
Energy Technology Data Exchange (ETDEWEB)
Reiss, T. [Institute of Nuclear Techniques, Budapest University of Technology and Economics, Muegyetem rkp. 9., 1111 Budapest (Hungary)
2010-07-01
The Supercritical Water Cooled Reactor (SCWR) was chosen as one of the Generation IV reactors by GIF. At the moment, a number of concepts - thermal as well as fast ones - exist. The reference parameters for a thermal SCWR have been taken from the European High Performance Light Water Reactor (HPLWR). Since the pressure is higher than the critical pressure (22.1 MPa) there is no change in the phase of the water in the core. On the other hand, due to the significant changes in the physical properties of water at supercritical pressure, the system is susceptible to local temperature, density and power oscillations. This inclination is increased by the pseudo-critical transformation of the water used as coolant. Thus, for modelling a system of this type coupled neutronics - thermal-hydraulics programs are required. Such a program system has been developed with the following main features: great modularity which allows for easy modifications, thus several SCWR concepts can be studied; detailed assembly calculations (with MCNP) and full-core analysis (with SCALE) are supported; the differential equations of xenon poisoning are implemented to study xenon oscillations. The program system was used to examine the assembly of the HPLWR, to design the assembly and the core of the Simplified Supercritical Water Cooled Reactor (SSCWR) and to model xenon oscillations in SCWRs. (authors)
Coupling of 3D neutronics models with the system code ATHLET
International Nuclear Information System (INIS)
Langenbuch, S.; Velkov, K.
1999-01-01
The system code ATHLET for plant transient and accident analysis has been coupled with 3D neutronics models, like QUABOX/CUBBOX, for the realistic evaluation of some specific safety problems under discussion. The considerations for the coupling approach and its realization are discussed. The specific features of the coupled code system established are explained and experience from first applications is presented. (author)
Niranjan, Ram; Rout, R. K.; Srivastava, R.; Kaushik, T. C.; Gupta, Satish C.
2016-03-01
A 17 kJ transportable plasma focus (PF) device with flexible transmission lines is developed and is characterized. Six custom made capacitors are used for the capacitor bank (CB). The common high voltage plate of the CB is fixed to a centrally triggered spark gap switch. The output of the switch is coupled to the PF head through forty-eight 5 m long RG213 cables. The CB has a quarter time-period of 4 μs and an estimated current of 506 kA is delivered to the PF device at 17 kJ (60 μF, 24 kV) energy. The average neutron yield measured using silver activation detector in the radial direction is (7.1 ± 1.4) × 108 neutrons/shot over 4π sr at 5 mbar optimum D2 pressure. The average neutron yield is more in the axial direction with an anisotropy factor of 1.33 ± 0.18. The average neutron energies estimated in the axial as well as in the radial directions are (2.90 ± 0.20) MeV and (2.58 ± 0.20) MeV, respectively. The flexibility of the PF head makes it useful for many applications where the source orientation and the location are important factors. The influence of electromagnetic interferences from the CB as well as from the spark gap on applications area can be avoided by putting a suitable barrier between the bank and the PF head.
Agent code: Neutron transport benchmark example and extension to 3D lattice geometry
Directory of Open Access Journals (Sweden)
Hursin Mathieu
2005-01-01
Full Text Available The general methodology be hind 2D arbitrary geometry neutron transport AGENT code is the theory of R-functions, which al lows for simple modeling of complex geometries, and the method of characteristics, which solves the integral transport equation along characteristic neutron trajectories. This paper focuses on the extension of the methodology to ac count for 3D lattice geometries. Since the direct application of method of characteristics to 3D non-homogenized core con figuration may re quire a tremendous amount of memory and computing time, an alternative approximate solution based on coupling 2D method of characteristics and 1D diffusion solution is developed. The planar 2D method of characteristics and axial 1D diffusion solutions are coupled through the trans verse leak age. The use of a lower order 1D solution in the axial direction is justified by the fact that more heterogeneity in current PWR and BWR reactor cores occurs in the radial direction than in the axial one. In order to demonstrate the versatility and accuracy of the AGENT code, a 2D heterogeneous lattice problem, C5G7 is described in details. A theoretical description of the coupling methodology for 3D method of characteristics solution is followed by preliminary validation in comparison to the DeCART code.
Energy Technology Data Exchange (ETDEWEB)
Barcellos, Luiz Felipe F.C.; Bodmann, Bardo E.J.; Vilhena, Marco T.M.B., E-mail: luizfelipe.fcb@gmail.com, E-mail: bardo.bodmann@ufrgs.br, E-mail: mtmbvilhena@gmail.com [Universidade Federal do Rio Grande do Sul (UFRGS), Porto Alegre, RS (Brazil). Grupo de Estudos Nucleares; Leite, Sergio Q. Bogado, E-mail: sbogado@ibest.com.br [Comissao Nacional de Energia Nuclear (CNEN), Rio de Janeiro, RJ (Brazil)
2017-07-01
In this work a Monte Carlo simulator with continuous energy is used. This simulator distinguishes itself by using the sum of three probability distributions to represent the neutron spectrum. Two distributions have known shape, but have varying population of neutrons in time, and these are the fission neutron spectrum (for high energy neutrons) and the Maxwell-Boltzmann distribution (for thermal neutrons). The third distribution has an a priori unknown and possibly variable shape with time and is determined from parametrizations of Monte Carlo simulation. It is common practice in neutron transport calculations, e.g. multi-group transport, to consider that the neutrons only lose energy with each scattering reaction and then to use a thermal group with a Maxwellian distribution. Such an approximation is valid due to the fact that for fast neutrons up-scattering occurrence is irrelevant, being only appreciable at low energies, i.e. in the thermal energy region, in which it can be regarded as a Maxwell-Boltzmann distribution for thermal equilibrium. In this work the possible neutron-matter interactions are simulated with exception of the up-scattering of neutrons. In order to preserve the thermal spectrum, neutrons are selected stochastically as being part of the thermal population and have an energy attributed to them taken from a Maxwellian distribution. It is then shown how this procedure can emulate the up-scattering effect by the increase in the neutron population kinetic energy. Since the simulator uses tags to identify the reactions it is possible not only to plot the distributions by neutron energy, but also by the type of interaction with matter and with the identification of the target nuclei involved in the process. This work contains some preliminary results obtained from a Monte Carlo simulator for neutron transport that is being developed at Federal University of Rio Grande do Sul. (author)
Geant4 simulations of the neutron production and transport in the n_TOF spallation target
Lerendegui-Marco, J; Guerrero, C; Quesada,, J , M
2016-01-01
The neutron production and transport in the spallation target of the n TOF facility at CERN has been simulated with Geant4. The results obtained with the different hadronic Physics Lists provided by Geant4 have been compared with the experimental neutron flux in n TOF-EAR1. The best overall agreement in both the absolute value and the energy dependence of the flux from thermal to 1 GeV, is obtained with the INCL++ model coupled with the Fritiof Model(FTFP). This Physics List has been thus used to simulate and study the main features of the new n TOF-EAR2 beam line, currently in its commissioning phase.
Neutron transport and diffusion in inhomogeneous media. I
International Nuclear Information System (INIS)
Larsen, E.W.
1975-01-01
The asymptotic solution of the neutron transport equation is obtained for large near-critical domains D which possess a cellular, nearly periodic structure. A typical mean free path in D is taken to be of the same order of magnitude as a cell diameter, and these are taken to be small (of order epsilon) compared to a typical diameter of D. The solution is asymptotic with respect to the small parameter epsilon. It is a product of two functions, one determined by a detailed cell calculation and the other obtained as the solution of a time dependent diffusion equation. The diffusion equation contains precursor (delayed neutron) densities, equations for which are derived. The coefficients in the diffusion equation, which are determined using the results of the cell calculation, differ from those now used in engineering applications. The initial condition for the diffusion equation is derived, and the problem of determining the boundary condition is discussed
Beam-transport optimization for cold-neutron spectrometer
Directory of Open Access Journals (Sweden)
Nakajima Kenji
2015-01-01
Full Text Available We report the design of the beam-transport system (especially the vertical geometry for a cold-neutron disk-chopper spectrometer AMATERAS at J-PARC. Based on the elliptical shape, which is one of the most effective geometries for a ballistic mirror, the design was optimized to obtain, at the sample position, a neutron beam with high flux without serious degrading in divergence and spacial homogeneity within the boundary conditions required from actual spectrometer construction. The optimum focal point was examined. An ideal elliptical shape was modified to reduce its height without serious loss of transmission. The final result was adapted to the construction requirements of AMATERAS. Although the ideas studied in this paper are considered for the AMATERAS case, they can be useful also to other spectrometers in similar situations.
Generalized diffusion theory for calculating the neutron transport scalar flux
International Nuclear Information System (INIS)
Alcouffe, R.E.
1975-01-01
A generalization of the neutron diffusion equation is introduced, the solution of which is an accurate approximation to the transport scalar flux. In this generalization the auxiliary transport calculations of the system of interest are utilized to compute an accurate, pointwise diffusion coefficient. A procedure is specified to generate and improve this auxiliary information in a systematic way, leading to improvement in the calculated diffusion scalar flux. This improvement is shown to be contingent upon satisfying the condition of positive calculated-diffusion coefficients, and an algorithm that ensures this positivity is presented. The generalized diffusion theory is also shown to be compatible with conventional diffusion theory in the sense that the same methods and codes can be used to calculate a solution for both. The accuracy of the method compared to reference S/sub N/ transport calculations is demonstrated for a wide variety of examples. (U.S.)
Energy Technology Data Exchange (ETDEWEB)
Monti, Lanfranco, E-mail: lanfranco.monti@gmail.co [Karlsruhe Institute of Technology (KIT), Institute for Nuclear and Energy Technologies (IKET), Hermann-von-Helmholtzplatz 1, 76344 Eggenstein-Leopoldshafen (Germany); Starflinger, Joerg, E-mail: joerg.starflinger@ike.uni-stuttgart.d [Universitaet Stuttgart, Institut fuer Kernenergetik und Energiesysteme, Pfaffenwaldring 31, 70569 Stuttgart (Germany); Schulenberg, Thomas, E-mail: schulenberg@kit.ed [Karlsruhe Institute of Technology (KIT), Institute for Nuclear and Energy Technologies (IKET), Hermann-von-Helmholtzplatz 1, 76344 Eggenstein-Leopoldshafen (Germany)
2011-05-15
Highlights: Advanced analysis and design techniques for innovative reactors are addressed. Detailed investigation of a 3 pass core design with a multi-physics-scales tool. Coupled 40-group neutron transport/equivalent channels TH core analyses methods. Multi-scale capabilities: from equivalent channels to sub-channel pin-by-pin study. High fidelity approach: reduction of conservatism involved in core simulations. - Abstract: The High Performance Light Water Reactor (HPLWR) is a thermal spectrum nuclear reactor cooled and moderated with light water operated at supercritical pressure. It is an innovative reactor concept, which requires developing and applying advanced analysis tools as described in the paper. The relevant water density reduction associated with the heat-up, together with the multi-pass core design, results in a pronounced coupling between neutronic and thermal-hydraulic analyses, which takes into account the strong natural influence of the in-core distribution of power generation and water properties. The neutron flux gradients within the multi-pass core, together with the pronounced dependence of water properties on the temperature, require to consider a fine spatial resolution in which the individual fuel pins are resolved to provide precise evaluation of the clad temperature, currently considered as one of the crucial design criteria. These goals have been achieved considering an advanced analysis method based on the usage of existing codes which have been coupled with developed interfaces. Initially neutronic and thermal-hydraulic full core calculations have been iterated until a consistent solution is found to determine the steady state full power condition of the HPLWR core. Results of few group neutronic analyses might be less reliable in case of HPLWR 3-pass core than for conventional LWRs because of considerable changes of the neutron spectrum within the core, hence 40 groups transport theory has been preferred to the usual 2 groups
Neutron imaging of root water uptake, transport and hydraulic redistribution
Warren, J.; Bilheux, H.; Kang, M.; Voisin, S.; Cheng, C.; Horita, J.; Perfect, E.
2012-12-01
Knowledge of plant water fluxes is critical for assessing mechanistic processes linked to biogeochemical cycles, yet resolving root water transport dynamics has been a particularly daunting task. Our objectives were to demonstrate the ability to non-invasively monitor individual root functionality and water fluxes within 1-3-week old Zea mays L. (maize) and Panicum virgatum L. (switchgrass) seedlings using neutron imaging. Seedlings were propagated in a growth chamber adjacent to the HFIR CG1 Beam Line at Oak Ridge National Laboratory in cylindrical or plate-like aluminum chambers containing sand. Seedlings were maintained under fairly dry conditions, with water added only to replace daily evapotranspiration. Plants were placed into the high flux cold neutron beam line and injections of H2O or deuterium oxide (D2O) were tracked through the soil and root systems by collecting consecutive CCD radiographs through time. Water fluxes within the root systems were manipulated by cycling on a growth lamp that altered foliar demand for water and thus internal water potential driving forces. 2D and 3D neutron radiography readily illuminated root structure, root growth, and relative plant and soil water content. 2D pulse-chase irrigation experiments with H2O and D2O, which have different neutron cross sections and thus differences in resulting image contrast, successfully allowed observation of uptake and mass flow of water within the root system. After irrigation there was rapid root water uptake from the newly wetted soil, followed by progressive hydraulic redistribution of water through the root systems to roots terminating in dry soil. Water flux within individual roots responded differentially to foliar illumination based on internal water potential gradients. Using 2D radiography, absolute fluxes of H2O or D2O through the system could not be easily determined since neutron attenuation through the sample was dependent on unknown and dynamic magnitudes of both D and H
Coupling tests of parallel channels to modal neutronic kinetics
International Nuclear Information System (INIS)
Cecenas F, M.; Campos G, R.M.
2007-01-01
In this work the initial results of the joining of an arrangement of 36 channels in parallel are studied to a modal neutronic kinetic model to represent the core of a BWR type reactor. The set of channels is obtained group the assemblies of that it consists the core in an arrangement of concentric rings, for later on to subdivide each ring in four parts to assign each segment to a quadrant of the core. The obtained channels are coupled to a modal kinetics model that considers the fundamental way and the first harmonic. The obtained solution represents the radial distribution and power azimuthal, the one which is feedback to the channels to update the thermohydraulic variables. The restriction that the pressure drop is same for each channel it is only imposed as initial condition, like part of the stationary state, and it is allowed that the pressure drop in the assemblies them it is different in each channel during a reactivity interference. For the tests to the system, it is convenient to select a relatively big core and that it operates near their stability frontier, for that the channels are dimensioning according to the case 9 of the Stability Benchmark of the Ringhals Swedish plant, organized by the Nuclear Energy Agency in 1994. In general, they reproduce the results of this benchmark, when reproducing oscillations outside of phase, with the additional results that the quadrants 1 and 2 of the core present oscillations of more width that the quadrants 3 and 4. The set group nuclear-thermohydraulics it is solved numerically by means of the outline master-slave of distributed calculation implemented by means of Parallel Virtual Machine (PVM). (Author)
Neutronic and thermal-hydraulic coupling for 3D reactor core modeling combining MCB and fluent
Directory of Open Access Journals (Sweden)
Królikowski Igor P.
2015-09-01
Full Text Available Three-dimensional simulations of neutronics and thermal hydraulics of nuclear reactors are a tool used to design nuclear reactors. The coupling of MCB and FLUENT is presented, MCB allows to simulate neutronics, whereas FLUENT is computational fluid dynamics (CFD code. The main purpose of the coupling is to exchange data such as temperature and power profile between both codes. Temperature required as an input parameter for neutronics is significant since cross sections of nuclear reactions depend on temperature. Temperature may be calculated in thermal hydraulics, but this analysis needs as an input the power profile, which is a result from neutronic simulations. Exchange of data between both analyses is required to solve this problem. The coupling is a better solution compared to the assumption of estimated values of the temperatures or the power profiles; therefore the coupled analysis was created. This analysis includes single transient neutronic simulation and several steady-state thermal simulations. The power profile is generated in defined points in time during the neutronic simulation for the thermal analysis to calculate temperature. The coupled simulation gives information about thermal behavior of the reactor, nuclear reactions in the core, and the fuel evolution in time. Results show that there is strong influence of neutronics on thermal hydraulics. This impact is stronger than the impact of thermal hydraulics on neutronics. Influence of the coupling on temperature and neutron multiplication factor is presented. The analysis has been performed for the ELECTRA reactor, which is lead-cooled fast reactor concept, where the coolant fl ow is generated only by natural convection
Thermodynamically coupled mass transport processes in a saturated clay
International Nuclear Information System (INIS)
Carnahan, C.L.
1984-01-01
Gradients of temperature, pressure, and fluid composition in saturated clays give rise to coupled transport processes (thermal and chemical osmosis, thermal diffusion, ultrafiltration) in addition to the direct processes (advection and diffusion). One-dimension transport of water and a solute in a saturated clay subjected to mild gradients of temperature and pressure was simulated numerically. When full coupling was accounted for, volume flux (specific discharge) was controlled by thermal osmosis and chemical osmosis. The two coupled fluxes were oppositely directed, producing a point of stagnation within the clay column. Solute flows were dominated by diffusion, chemical osmosis, and thermal osmosis. Chemical osmosis produced a significant flux of solute directed against the gradient of solute concentration; this effect reduced solute concentrations relative to the case without coupling. Predictions of mass transport in clays at nuclear waste repositories could be significantly in error if coupled transport processes are not accounted for. 14 refs., 8 figs
International Nuclear Information System (INIS)
Laureau, Axel
2015-01-01
In this PhD thesis, we describe the development of innovative neutronic models for their coupling with thermal hydraulics such that they combine precision and reasonable computational times. One of the main cases where this method is applied is the Molten Salt Fast Reactor (MSFR) whose combines a fast neutron spectrum with a thorium cycle. In this fourth generation reactor, the motion of the delayed neutron precursors and the associated phenomena have to be taken into account due to the liquid fuel circulation. The starting point for these developments was the preliminary design of this type of system where a dedicated multi-physical representation was needed to study the reactor performance in steady and transient conditions. As a first step, a stationary coupling was developed. A neutronic model based on a stochastic approach was associated to a CFD (Computational Fluid Dynamics) code to solve the Navier Stokes equations for turbulent flows and the transport of the delayed neutron precursors. The impact of this precursor motion is taken into account by reconstructing the prompt shower that they generate. This approach, called by shower, views the critical reactor as a prompt subcritical reactor that amplifies a source of delayed neutrons. A second step consisted in developing a neutronic model based on a time dependent version of the fission matrices (Transient Fission Matrix or TFM) so as to enable reactor transient studies. With the TFM model, an initial computation of the matrices with a stochastic code (MCNP, SERPENT) allows the characterization of the global spatial and time dependent neutronic response of the reactor with a precision close to that of a Monte Carlo calculation. The information thus obtained is then used to calculate transients, while retaining the advantage of reduced computational time. The TFM model, which can be used for various system concepts, also allows the evaluation of effective kinetic parameters such as the effective fraction of
Direct measurement of lithium transport in graphite electrodes using neutrons
International Nuclear Information System (INIS)
Owejan, Jon P.; Gagliardo, Jeffrey J.; Harris, Stephen J.; Wang, Howard; Hussey, Daniel S.; Jacobson, David L.
2012-01-01
Highlights: ► Spatiotemporal measurements of lithium through the electrode thickness were quantified with high resolution neutron imaging. ► A nonuniform lithium distribution was observed early in the first intercalation cycle but relaxed as the electrode filled with lithium. ► Through-plane transport resistance in the bulk of the graphite composite electrode was measured. ► The distribution of lost capacity associated with trapped lithium was quantified and linked to regions with low intercalation rates. - Abstract: Lithium intercalation into graphite electrodes is widely studied, but few direct in situ diagnostic methods exist. Such diagnostic methods are desired to probe the influence of factors such as charge rate, electrode structure and solid electrolyte interphase layer transport resistance as they relate to lithium-ion battery performance and durability. In this work, we present a continuous measurement of through-plane lithium distributions in a composite graphite/lithium metal electrochemical cell. Capacity change in a thick graphite electrode was measured during several charge/discharge cycles with high resolution (14 μm) neutron imaging. A custom test fixture and a method for quantifying lithium are described. The measured lithium distribution within the graphite electrode is given as a function of state of charge. Bulk transport resistance is considered by comparing intercalation rates through the thickness of the electrode near the separator and current collector. The residual lithium content associated with irreversible capacity loss that results from cycling is also measured.
Error reduction techniques for Monte Carlo neutron transport calculations
International Nuclear Information System (INIS)
Ju, J.H.W.
1981-01-01
Monte Carlo methods have been widely applied to problems in nuclear physics, mathematical reliability, communication theory, and other areas. The work in this thesis is developed mainly with neutron transport applications in mind. For nuclear reactor and many other applications, random walk processes have been used to estimate multi-dimensional integrals and obtain information about the solution of integral equations. When the analysis is statistically based such calculations are often costly, and the development of efficient estimation techniques plays a critical role in these applications. All of the error reduction techniques developed in this work are applied to model problems. It is found that the nearly optimal parameters selected by the analytic method for use with GWAN estimator are nearly identical to parameters selected by the multistage method. Modified path length estimation (based on the path length importance measure) leads to excellent error reduction in all model problems examined. Finally, it should be pointed out that techniques used for neutron transport problems may be transferred easily to other application areas which are based on random walk processes. The transport problems studied in this dissertation provide exceptionally severe tests of the error reduction potential of any sampling procedure. It is therefore expected that the methods of this dissertation will prove useful in many other application areas
In situ quantification and visualization of lithium transport with neutrons.
Liu, Danny X; Wang, Jinghui; Pan, Ke; Qiu, Jie; Canova, Marcello; Cao, Lei R; Co, Anne C
2014-09-01
A real-time quantification of Li transport using a nondestructive neutron method to measure the Li distribution upon charge and discharge in a Li-ion cell is reported. By using in situ neutron depth profiling (NDP), we probed the onset of lithiation in a high-capacity Sn anode and visualized the enrichment of Li atoms on the surface followed by their propagation into the bulk. The delithiation process shows the removal of Li near the surface, which leads to a decreased coulombic efficiency, likely because of trapped Li within the intermetallic material. The developed in situ NDP provides exceptional sensitivity in the temporal and spatial measurement of Li transport within the battery material. This diagnostic tool opens up possibilities to understand rates of Li transport and their distribution to guide materials development for efficient storage mechanisms. Our observations provide important mechanistic insights for the design of advanced battery materials. © 2014 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.
Three dimensional neutronic/thermal-hydraulic coupled simulation of MSR in transient state condition
International Nuclear Information System (INIS)
Zhou, Jianjun; Zhang, Daling; Qiu, Suizheng; Su, Guanghui; Tian, Wenxi; Wu, Yingwei
2015-01-01
Highlights: • Developed a three dimensional neutronic/thermal-hydraulic coupled transient analysis code for MSR. • Investigated the neutron distribution and thermal-hydraulic characters of the core under transient condition. • Analyzed three different transient conditions of inlet temperature drop, reactivity jump and pump coastdown. - Abstract: MSR (molten salt reactor) use liquid molten salt as coolant and fuel solvent, which was the only one liquid reactor of six Generation IV reactor types. As a liquid reactor the physical property of reactor was significantly influenced by fuel salt flow and the conventional analysis methods applied in solid fuel reactors are not applicable for this type of reactors. The present work developed a three dimensional neutronic/thermal-hydraulic coupled code investigated the neutronics and thermo-hydraulics characteristics of the core in transient condition based on neutron diffusion theory and numerical heat transfer. The code consists of two group neutron diffusion equations for fast and thermal neutron fluxes and six group balance equations for delayed neutron precursors. The code was separately validated by neutron benchmark and flow and heat transfer benchmark. Three different transient conditions was analyzed with inlet temperature drop, reactivity jump and pump coastdown. The results provide some valuable information in design and research this kind of reactor
A solution of the neutron transport equation using spherical harmonics
International Nuclear Information System (INIS)
Fletcher, J.K.
1983-01-01
A solution of the neutron transport equation is obtained by expanding the flux as a series. Preliminary investigations in one dimension indicated that the first-order differential equations resulting for the unknown coefficients or moments could be solved by eliminating terms with odd L (L = order of Legendre polynomial) to give a second-order system. FORTRAN subroutines have been written to calculate the necessary coefficients and specify the relevant differentials. A finite-difference or finite-element approximation can then be used. (U.K.)
Structures of the fractional spaces generated by the difference neutron transport operator
Energy Technology Data Exchange (ETDEWEB)
Ashyralyev, Allaberen [Department of Elementary Mathematics Education, Fatih University, 34500, Istanbul (Turkey); Department of Mathematics, ITTU, Ashgabat (Turkmenistan); Taskin, Abdulgafur [Department of Mathematics, Fatih University, 34500, Istanbul (Turkey)
2015-09-18
The initial boundary value problem for the neutron transport equation is considered. The first, second and third order of accuracy difference schemes for the approximate solution of this problem are presented. Highly accurate difference schemes for neutron transport equation based on Padé approximation are constructed. In applications, stability estimates for solutions of difference schemes for the approximate solution of the neutron transport equation are obtained.The positivity of the neutron transport operator in Slobodeckij spaces is proved. Numerical techniques are developed and algorithms are tested on an example in MATLAB.
SAM-CE, Time-Dependent 3-D Neutron Transport, Gamma Transport in Complex Geometry by Monte-Carlo
International Nuclear Information System (INIS)
2003-01-01
1 - Nature of physical problem solved: The SAM-CE system comprises two Monte Carlo codes, SAM-F and SAM-A. SAM-F supersedes the forward Monte Carlo code, SAM-C. SAM-A is an adjoint Monte Carlo code designed to calculate the response due to fields of primary and secondary gamma radiation. The SAM-CE system is a FORTRAN Monte Carlo computer code designed to solve the time-dependent neutron and gamma-ray transport equations in complex three-dimensional geometries. SAM-CE is applicable for forward neutron calculations and for forward as well as adjoint primary gamma-ray calculations. In addition, SAM-CE is applicable for the gamma-ray stage of the coupled neutron-secondary gamma ray problem, which may be solved in either the forward or the adjoint mode. Time-dependent fluxes, and flux functionals such as dose, heating, count rates, etc., are calculated as functions of energy, time and position. Multiple scoring regions are permitted and these may be either finite volume regions or point detectors or both. Other scores of interest, e.g., collision and absorption densities, etc., are also made. 2 - Method of solution: A special feature of SAM-CE is its use of the 'combinatorial geometry' technique which affords the user geometric capabilities exceeding those available with other commonly used geometric packages. All nuclear interaction cross section data (derived from the ENDF for neutrons and from the UNC-format library for gamma-rays) are tabulated in point energy meshes. The energy meshes for neutrons are internally derived, based on built-in convergence criteria and user- supplied tolerances. Tabulated neutron data for each distinct nuclide are in unique and appropriate energy meshes. Both resolved and unresolved resonance parameters from ENDF data files are treated automatically, and extremely precise and detailed descriptions of cross section behaviour is permitted. Such treatment avoids the ambiguities usually associated with multi-group codes, which use flux
Two-dimensional time dependent Riemann solvers for neutron transport
International Nuclear Information System (INIS)
Brunner, Thomas A.; Holloway, James Paul
2005-01-01
A two-dimensional Riemann solver is developed for the spherical harmonics approximation to the time dependent neutron transport equation. The eigenstructure of the resulting equations is explored, giving insight into both the spherical harmonics approximation and the Riemann solver. The classic Roe-type Riemann solver used here was developed for one-dimensional problems, but can be used in multidimensional problems by treating each face of a two-dimensional computation cell in a locally one-dimensional way. Several test problems are used to explore the capabilities of both the Riemann solver and the spherical harmonics approximation. The numerical solution for a simple line source problem is compared to the analytic solution to both the P 1 equation and the full transport solution. A lattice problem is used to test the method on a more challenging problem
SUSD, Sensitivity and Uncertainty in Neutron Transport and Detector Response
International Nuclear Information System (INIS)
Furuta, Lazuo; Kondo, Shunsuke; Oka, Yoshika
1991-01-01
1 - Description of program or function: SUSD calculates sensitivity coefficients for one and two-dimensional transport problems. Variance and standard deviation of detector responses or design parameters can be obtained using cross-section covariance matrices. In neutron transport problems, this code is able to perform sensitivity-uncertainty analysis for secondary angular distribution (SAD) or secondary energy distribution (SED). 2 - Method of solution: The first-order perturbation theory is used to obtain sensitivity coefficients. The method described in the distributed report is employed to consider SAD/SED effect. 3 - Restrictions on the complexity of the problem: Variable dimension is used so that there is no limitation in each array size but the total core size
TORT/MCNP coupling method for the calculation of neutron flux around a core of BWR
International Nuclear Information System (INIS)
Kurosawa, M.
2005-01-01
For the analysis of BWR neutronics performance, accurate data are required for neutron flux distribution over the In-Reactor Pressure Vessel equipments taking into account the detailed geometrical arrangement. The TORT code can calculate neutron flux around a core of BWR in a three-dimensional geometry model, but has difficulties in fine geometrical modelling and lacks huge computer resource. On the other hand, the MCNP code enables the calculation of the neutron flux with a detailed geometry model, but requires very long sampling time to give enough number of particles. Therefore, a TORT/MCNP coupling method has been developed to eliminate the two problems mentioned above in each code. In this method, the TORT code calculates angular flux distribution on the core surface and the MCNP code calculates neutron spectrum at the points of interest using the flux distribution. The coupling method will be used as the DOT-DOMINO-MORSE code system. This TORT/MCNP coupling method was applied to calculate the neutron flux at points where induced radioactivity data were measured for 54 Mn and 60 Co and the radioactivity calculations based on the neutron flux obtained from the above method were compared with the measured data. (authors)
TORT/MCNP coupling method for the calculation of neutron flux around a core of BWR.
Kurosawa, Masahiko
2005-01-01
For the analysis of BWR neutronics performance, accurate data are required for neutron flux distribution over the In-Reactor Pressure Vessel equipments taking into account the detailed geometrical arrangement. The TORT code can calculate neutron flux around a core of BWR in a three-dimensional geometry model, but has difficulties in fine geometrical modelling and lacks huge computer resource. On the other hand, the MCNP code enables the calculation of the neutron flux with a detailed geometry model, but requires very long sampling time to give enough number of particles. Therefore, a TORT/MCNP coupling method has been developed to eliminate the two problems mentioned above in each code. In this method, the TORT code calculates angular flux distribution on the core surface and the MCNP code calculates neutron spectrum at the points of interest using the flux distribution. The coupling method will be used as the DOT-DOMINO-MORSE code system. This TORT/MCNP coupling method was applied to calculate the neutron flux at points where induced radioactivity data were measured for 54Mn and 60Co and the radioactivity calculations based on the neutron flux obtained from the above method were compared with the measured data.
International Nuclear Information System (INIS)
Massimiliano, Rosa; Azmy, Y.Y.; Morel, J.E.
2005-01-01
The general expressions for the matrix elements of the discrete Sn-equivalent integral transport operator have been derived in slab geometry. Their asymptotic behavior has been investigated both for a homogeneous slab and for a heterogeneous slab characterized by a periodic material discontinuity wherein each optically thick cell is surrounded by two optically thin cells in a repeating pattern. In the case of a homogeneous slab, the asymptotic analysis conducted in a diffusive limit obtained as the thick limit of computational cell size for a highly scattering medium, has shown that the discretized integral transport operator is approximated by a sparse matrix characterized by a tri-diagonal diffusion-like coupling stencil. Also, the tri-diagonal matrix structure, characteristic of the diffusion coupling stencil, is approached at a fast exponential rate. In the case of periodically heterogeneous slab configurations, the asymptotic behavior investigated is that in which the cells' optical thicknesses are pushed apart, i.e. the thick is made thicker while the thin is made thinner at a prescribed rate. It has been shown that in this limit the discretized integral transport operator is approximated by a penta-diagonal structure. Notwithstanding, the discrete operator is amenable to algebraic transformations leading to a matrix representation still asymptotically approaching a tri-diagonal structure at a fast exponential rate. The existence of a low order tri-diagonal approximation to the full discrete integral transport operator in the case of a periodically heterogeneous slab might provide a basic understanding of the superior convergence properties of diffusion-based acceleration schemes observed in slab geometry, even in the presence of sharp material discontinuities. The obtained results also suggest that a sparse approximation to the S n -equivalent integral transport operator might itself be used as the low-order operator in an acceleration scheme for the
MCNP: a general Monte Carlo code for neutron and photon transport
International Nuclear Information System (INIS)
1979-11-01
The general-purpose Monte Carlo code MCNP ca be used for neutron, photon, or coupled neutron-photon transport, including the capability to calculate eigenvalues for critical systems. The code treats an arbitrary three-dimensional configuration of materials in geometric cells bounded by first- and second-degree surfaces and some special fourth-degree surfaces (elliptical tori). Pointwise cross-section data are used. For neutrons, all reactions given in a particular cross-section evaluation are accounted for. Thermal neutrons are described by both the free-gas and S(α,β) models. For photons, the code takes account of incoherent and coherent scattering, the possibility of fluorescent emission following photoelectric absorption, and absorption in pair production with local emission of annihilation radiation. MCNP includes an elaborate, interactive plotting capability that allows the user to view his input geometry to help check for setup errors. Standard features which are available to improve computational efficiency include geometry splitting and Russian roulette, weight cutoff with Russian roulette, correlated sampling, analog capture or capture by weight reduction, the exponential transformation, energy splitting, forced collisions in designated cells, flux estimates at point or ring detectors, deterministically transporting pseudo-particles to designated regions, track-length estimators, source biasing, and several parameter cutoffs. Extensive summary information is provided to help the user better understand the physics and Monte Carlo simulation of his problem. The standard, user-defined output of MCNP includes two-way current as a function of direction across any set of surfaces or surface segments in the problem. Flux across any set of surfaces or surface segments is available. 58 figures, 28 tables
Coupled electric and transport phenomena in porous media
Li, Shuai
2014-01-01
The coupled electrical and transport properties of clay-containing porous media are the topics of interest in this study. Both experimental and numerical (pore network modeling) techniques are employed to gain insight into the macro-scale interaction between electrical and solute transport phenomena
International Nuclear Information System (INIS)
Perez-Manes, Jorge; Sanchez Espinoza, Victor Hugo; Chiva, Sergio
2014-01-01
The Institute for Neutron Physics and Reactor Technology (INR) at the Karlsruhe Institute of Technology (KIT) is investigating the application of the meso- and microscale analysis for the prediction of local safety parameters for light water reactors (LWR). By applying codes like CFD (computational fluid dynamics) and SP3 (simplified transport) reactor dynamics it is possible to describe the underlying phenomena in a more accurate manner than by the nodal/coarse 1D thermal hydraulic coupled codes. By coupling the transport (SP3) based neutron kinetics (NK) code DYN3D with NEPTUNE-CFD, within a parallel MPI-environment, the NHESDYN platform is created. The newly developed system will allow high fidelity simulations of LWR fuel assemblies and cores. In NHESDYN, a heat conduction solver, SYRTHES, is coupled to NEPTUNE-CFD. The driver module of NHESDYN controls the sequence of execution of the solvers as well as the communication between the solvers based on MPI. In this paper, the main features of NHESDYN are discussed and the proof of the concept is done by solving a single pin problem. The prediction capability of NHESDYN is demonstrated by a code-to-code comparison with the DYNSUB code. Finally, the future developments and validation efforts are highlighted. (authors)
Coupled water transport by rat proximal tubule.
Green, R; Giebisch, G; Unwin, R; Weinstein, A M
1991-12-01
Simultaneous microperfusion of proximal tubules and peritubular capillaries in kidneys of rats anesthetized with Inactin was used to examine water reabsorption by this epithelium. Osmolality of the luminal solution was varied with changes in NaCl concentration and by the addition of raffinose. Capillary perfusates contained either low (2 g/dl) or high (16 g/dl) concentrations of albumin. We used low-bicarbonate perfusates for both lumen and capillary so that we might apply the nonequilibrium thermodynamic model of transport for a single solute (NaCl) to interpret our observations. Linear regression with the volume flux equation Jv = -Lp delta II - Lp sigma delta C + Jav (where Jv is volume flux, Lp is hydraulic conductance, delta II is oncotic force, sigma is osmotic reflection coefficient, delta C is salt concentration difference, and Jav is the component of Jv not attributed to transepithelial hydrostatic or osmotic forces) revealed a tubule water permeability (Pf = 0.11 +/- 0.01 cm/s) and a sigma (0.74 +/- 0.08) in agreement with previous determinations. These transport parameters were unaffected by changes in peritubular protein. We also found that Jav was substantial, approximately three-fourths of the rate of isotonic transport under these perfusion conditions. Further, this component of water transport nearly doubled with the transition from low- to high-protein peritubular capillary perfusion. When expressed as a capacity for water reabsorption against an osmotic gradient, the salt concentration differences required to null volume flux were 13.2 +/- 2.4 and 29.4 +/- 4.0 mosmol/kgH2O under low and high peritubular protein. Our data suggest that this protein effect is, most likely, an increase in solute transport by the tubule epithelial cells.
Thermal neutron imaging through XRQA2 GAFCHROMIC films coupled with a cadmium radiator
Energy Technology Data Exchange (ETDEWEB)
Sacco, D. [INFN – LNF, Via E. Fermi n.40, Frascati, 00044 Roma (Italy); INAIL – DIT, Via di Fontana Candida n.1, 00040 Monteporzio Catone (Italy); Bedogni, R., E-mail: roberto.bedogni@lnf.infn.it [INFN – LNF, Via E. Fermi n.40, Frascati, 00044 Roma (Italy); Bortot, D. [Politecnico di Milano, Dipartimento di Energia, Via La Masa 34, 20156 Milano (Italy); INFN – Milano, Via Celoria16, 20133 Milano (Italy); Palomba, M. [ENEA Casaccia, Via Anguillarese, 301, S. Maria di Galeria, 00123 Roma (Italy); Pola, A. [Politecnico di Milano, Dipartimento di Energia, Via La Masa 34, 20156 Milano (Italy); INFN – Milano, Via Celoria16, 20133 Milano (Italy); Introini, M.V.; Lorenzoli, M. [Politecnico di Milano, Dipartimento di Energia, Via La Masa 34, 20156 Milano (Italy); Gentile, A. [INFN – LNF, Via E. Fermi n.40, Frascati, 00044 Roma (Italy); Strigari, L. [Laboratory of Medical Physics, Regina Elena National Cancer Institute, Via E. Chianesi 53, 00144 Roma (Italy); Pressello, C. [Department of Medical Physics, Azienda Ospedaliera San Camillo Forlanini, Circonvallazione Gianicolense 87, 00152 Roma (Italy); Soriani, A. [Laboratory of Medical Physics, Regina Elena National Cancer Institute, Via E. Chianesi 53, 00144 Roma (Italy); Gómez-Ros, J.M. [INFN – LNF, Via E. Fermi n.40, Frascati, 00044 Roma (Italy); CIEMAT, Av. Complutense 40, 28040 Madrid (Spain)
2015-10-21
A simple and inexpensive method to perform passive thermal neutron imaging on large areas was developed on the basis of XRQA2 GAFCHROMIC films, commonly employed for quality assurance in radiology. To enhance their thermal neutron response, the sensitive face of film was coupled with a 1 mm thick cadmium radiator, forming a sandwich. By exchanging the order of Cd filter and sensitive film with respect to the incident neutron beam direction, two different configurations (beam-Cd-film and beam-film-Cd) were identified. These configurations were tested at thermal neutrons fluence values in the range 10{sup 9}–10{sup 10} cm{sup −2}, using the ex-core radial thermal neutron column of the ENEA Casaccia – TRIGA reactor. The results are presented in this work.
International Nuclear Information System (INIS)
Pena, C.; Pellacani, F.; Macian Juan, R.; Chiva, S.; Barrachina, T.; Miro, R.
2011-01-01
A computational code system based on coupling the 3D neutron diffusion code PARCS v2.7 and the Ansys CFX 13.0 Computational Fluid Dynamics (CFD) code has been developed as a tool for nuclear reactor systems simulations. This paper presents the coupling methodology between the CFD and the neutronic code. The methodology to simulate a 3D-neutronic problem coupled with 1D thermal hydraulics is already a mature technology, being part of the regular calculations performed to analyze different kinds of Reactivity Insertion Accidents (RIA) and asymmetric transients in Nuclear Power Plants, with state-of-the-art coupled codes like TRAC-B/NEM, RELAP5/PARCS, TRACE/PARCS, RELAP3D, RETRAN3D, etc. This work represents one of the first attempts to couple the multiphysics of a nuclear reactor core with a 3D spatial resolution in a computer code. This will open new possibilities regarding the analysis of fuel elements, contributing to a better understanding and design of the heat transfer process and specific fluid dynamics phenomena such as cross flow among fuel elements. The transient simulation of control rod insertion, boron dilution and cold water injection will be made possible with a degree of accuracy not achievable with current methodologies based on the use of system and/or subchannel codes. The transport of neutrons depends on several parameters, like fuel temperature, moderator temperature and density, boron concentration and fuel rod insertion. These data are calculated by the CFD code with high local resolution and used as input to the neutronic code to calculate a 3D nodal power distribution that will be returned and remapped to the CFD code control volumes (cells). Since two different nodalizations are used to discretized the same system, an averaging and interpolating procedure is needed to realize an effective data exchange. These procedures have been developed by means of the Ansys CFX 'User Fortran' interface; a library with several subroutines has been
Energy Technology Data Exchange (ETDEWEB)
Pena, C.; Pellacani, F.; Macian Juan, R., E-mail: carlos.pena@ntech.mw.tum.de, E-mail: pellacani@ntech.mw.tum.de, E-mail: macian@ntech.mw.tum.de [Technische Universitaet Muenchen, Garching (Germany). Ntech Lehrstuhl fuer Nukleartechnik; Chiva, S., E-mail: schiva@emc.uji.es [Universitat Jaume I, Castellon de la Plana (Spain). Dept. de Ingenieria Mecanica y Construccion; Barrachina, T.; Miro, R., E-mail: rmiro@iqn.upv.es, E-mail: tbarrachina@iqn.upv.es [Universitat Politecnica de Valencia (ISIRYM/UPV) (Spain). Institute for Industrial, Radiophysical and Environmental Safety
2011-07-01
A computational code system based on coupling the 3D neutron diffusion code PARCS v2.7 and the Ansys CFX 13.0 Computational Fluid Dynamics (CFD) code has been developed as a tool for nuclear reactor systems simulations. This paper presents the coupling methodology between the CFD and the neutronic code. The methodology to simulate a 3D-neutronic problem coupled with 1D thermal hydraulics is already a mature technology, being part of the regular calculations performed to analyze different kinds of Reactivity Insertion Accidents (RIA) and asymmetric transients in Nuclear Power Plants, with state-of-the-art coupled codes like TRAC-B/NEM, RELAP5/PARCS, TRACE/PARCS, RELAP3D, RETRAN3D, etc. This work represents one of the first attempts to couple the multiphysics of a nuclear reactor core with a 3D spatial resolution in a computer code. This will open new possibilities regarding the analysis of fuel elements, contributing to a better understanding and design of the heat transfer process and specific fluid dynamics phenomena such as cross flow among fuel elements. The transient simulation of control rod insertion, boron dilution and cold water injection will be made possible with a degree of accuracy not achievable with current methodologies based on the use of system and/or subchannel codes. The transport of neutrons depends on several parameters, like fuel temperature, moderator temperature and density, boron concentration and fuel rod insertion. These data are calculated by the CFD code with high local resolution and used as input to the neutronic code to calculate a 3D nodal power distribution that will be returned and remapped to the CFD code control volumes (cells). Since two different nodalizations are used to discretized the same system, an averaging and interpolating procedure is needed to realize an effective data exchange. These procedures have been developed by means of the Ansys CFX 'User Fortran' interface; a library with several subroutines has
Structure of a bacterial energy-coupling factor transporter.
Wang, Tingliang; Fu, Guobin; Pan, Xiaojing; Wu, Jianping; Gong, Xinqi; Wang, Jiawei; Shi, Yigong
2013-05-09
The energy-coupling factor (ECF) transporters constitute a novel family of conserved membrane transporters in prokaryotes that have a similar domain organization to the ATP-binding cassette transporters. Each ECF transporter comprises a pair of cytosolic ATPases (the A and A' components, or EcfA and EcfA'), a membrane-embedded substrate-binding protein (the S component, or EcfS) and a transmembrane energy-coupling component (the T component, or EcfT) that links the EcfA-EcfA' subcomplex to EcfS. The structure and transport mechanism of the quaternary ECF transporter remain largely unknown. Here we report the crystal structure of a nucleotide-free ECF transporter from Lactobacillus brevis at a resolution of 3.5 Å. The T component has a horseshoe-shaped open architecture, with five α-helices as transmembrane segments and two cytoplasmic α-helices as coupling modules connecting to the A and A' components. Strikingly, the S component, thought to be specific for hydroxymethyl pyrimidine, lies horizontally along the lipid membrane and is bound exclusively by the five transmembrane segments and the two cytoplasmic helices of the T component. These structural features suggest a plausible working model for the transport cycle of the ECF transporters.
Strongly coupled chameleon fields: Possible test with a neutron Lloyd's mirror interferometer
Energy Technology Data Exchange (ETDEWEB)
Pokotilovski, Yu.N., E-mail: pokot@nf.jinr.ru [Joint Institute for Nuclear Research, 141980 Dubna, Moscow Region (Russian Federation)
2013-02-26
The consideration of possible neutron Lloyd's mirror interferometer experiment to search for strongly coupled chameleon fields is presented. The chameleon scalar fields were proposed to explain the acceleration of expansion of the Universe. The presence of a chameleon field results in a change of a particle's potential energy in vicinity of a massive body. This interaction causes a phase shift of neutron waves in the interferometer. The sensitivity of the method is estimated.
Neutron transport and Montecarlo method: analysis and revision
International Nuclear Information System (INIS)
Perlado, J.M.
1982-01-01
The resolution of the neutron transport equation by the Montecarlo method is presented. Coming from an extensive discussion on the best formulation of that equation in order to be treated through the mentioned method, the theoretical bases of the estimator and random-walk generation is extensively explained. The most general expression for the estimators in different physical situations, each with a diverse random-walk, is included in this basical theoretical part. Furthemore, a large revision on the variance reduction methods is made. Its theoretical presentation is claimed to be in connection with the need for each one of them. The use of the adjoint equation, as a part of the importance sampling, Russian Roulette, splitting, exponential transform, conditional and correlated Montecarlo, and one-collision and next-event extimators, are discussed. Finally, come comments in the presentation of the last works on the theoretical prediction of errors in the generation of estimators-random walks are made. (author)
Numerical method for solving integral equations of neutron transport. II
International Nuclear Information System (INIS)
Loyalka, S.K.; Tsai, R.W.
1975-01-01
In a recent paper it was pointed out that the weakly singular integral equations of neutron transport can be quite conveniently solved by a method based on subtraction of singularity. This previous paper was devoted entirely to the consideration of simple one-dimensional isotropic-scattering and one-group problems. The present paper constitutes interesting extensions of the previous work in that in addition to a typical two-group anisotropic-scattering albedo problem in the slab geometry, the method is also applied to an isotropic-scattering problem in the x-y geometry. These results are compared with discrete S/sub N/ (ANISN or TWOTRAN-II) results, and for the problems considered here, the proposed method is found to be quite effective. Thus, the method appears to hold considerable potential for future applications. (auth)
Parallelism in continuous energy Monte Carlo method for neutron transport
Energy Technology Data Exchange (ETDEWEB)
Uenohara, Yuji (Nuclear Engineering Lab., Toshiba Corp. (Japan))
1993-04-01
The continuous energy Monte Carlo code VIM was implemented on a prototype highly parallel computer called PRODIGY developed by TOSHIBA Corporation. The author tried to distribute nuclear data to the processing elements (PEs) for the purpose of studying domain decompositon for the velocity space. Eigenvalue problems for a 1-D plate-cell infinite lattice mockup of ZPR-6-7 wa examined. For the geometrical space, the PEs were assigned to domains corresponding to nuclear fuel bundles in a typical boiling water reactor. The author estimated the parallelization efficiencies for both highly parallel and a massively parallel computer. Negligible communication overhead derived from neutron transports resulted from the heavy computing loads of Monte Carlo simulations. In the case of highly parallel computers, the communication overheads scarcely contributed to the parallelization efficiency. In the case of massively parallel computers, the control of PEs resulted in considerable communication overheads. (orig.)
Parallelism in continuous energy Monte Carlo method for neutron transport
International Nuclear Information System (INIS)
Uenohara, Yuji
1993-01-01
The continuous energy Monte Carlo code VIM was implemented on a prototype highly parallel computer called PRODIGY developed by TOSHIBA Corporation. The author tried to distribute nuclear data to the processing elements (PEs) for the purpose of studying domain decompositon for the velocity space. Eigenvalue problems for a 1-D plate-cell infinite lattice mockup of ZPR-6-7 wa examined. For the geometrical space, the PEs were assigned to domains corresponding to nuclear fuel bundles in a typical boiling water reactor. The author estimated the parallelization efficiencies for both highly parallel and a massively parallel computer. Negligible communication overhead derived from neutron transports resulted from the heavy computing loads of Monte Carlo simulations. In the case of highly parallel computers, the communication overheads scarcely contributed to the parallelization efficiency. In the case of massively parallel computers, the control of PEs resulted in considerable communication overheads. (orig.)
Safety improvement of start-up neutron source handling work by preparing new transport containers
International Nuclear Information System (INIS)
Shimazaki, Yosuke; Sawahata, Hiroaki; Yanagida, Yoshinori; Shinohara, Masanori; Kawamoto, Taiki; Takada, Shoji
2016-01-01
The conventional transport containers that have been used in HTTR start-up neutron source replacement work are not specialized type for HTTR start-up neutron source. As the risks associated with the safe handling of neutron source holders due to the above fact, the following three risks have been confirmed: (1) exposure risk due to leakage of neutron source or gamma rays, (2) risk of erroneous fall of neutron source holders, and (3) accident due to incorrect handling of transport containers. This paper reports the risks confirmed in the handling of neutron source holders associated with transport containers and the risk reduction measures, as well as the fabrication of new transport containers. By employing the size-reduction and simple structure, new transport containers have been completed at the same level of costs compared with the continuous use of the conventional transport containers, while satisfying the criteria of Type A transport materials and serving as risk preventive measures. Thus, new transport containers aimed at the risk prevention measures for the handling work of neutron source holders have been completed, and the safety of operation has been improved. (A.O.)
Paganini, S
2005-01-01
Crews working on present-day jet aircraft are a large occupationally exposed group with a relatively high average effective dose from Galactic cosmic radiation. Crews of future high-speed commercial flying at higher altitudes would be even more exposed. To help reduce the significant uncertainties in calculations of such exposures, the male adult voxels phantom MAX, developed in the Nuclear Energy Department of Pernambuco Federal University in Brazil, has been coupled with the Monte Carlo simulation code GEANT4. This toolkit, distributed and upgraded from the international scientific community of CERN/Switzerland, simulates thermal to ultrahigh energy neutrons transport and interactions in the matter. The high energy neutrons are pointed as the component that contribute about 70% of the neutron effective dose that represent the 35% to 60% total dose at aircraft altitude. In this research calculations of conversion coefficients from fluence to effective dose are performed for neutrons of energies from 100 MeV ...
Development of a transportable neutron radiography system for non-destructive tests application
International Nuclear Information System (INIS)
Silva, Ademir X. da; Crispim, Verginia R.
1999-01-01
This paper presents a study of a transportable neutron radiography system utilizing californium-252. Studies about moderation, collimation and shielding are showed. A Monte Carlo Code, MCNP3b, has been used to obtain a maximum and more homogeneous thermal neutron flux in the collimator outlet next to the image plain, and an adequate radiation shielding to attend radiological protection rules. With the presented collimator, it was possible to obtain for the thermal neutron flux, at the collimator outlet and next to the image plain, a L/D ratio 7,5, for neutron flux up to 6 X 10 -6 cm -2 .s -1 per neutron source. (author)
Investigation on coupling characteristics of neutronics/thermal-hydraulics of PWR NPP core
International Nuclear Information System (INIS)
Zheng Yong; Peng Minjun; Xia Genglei; Liu Xinkai
2014-01-01
In this paper, an integrated neutronics/thermal-hydraulic model for the reactor of Qinshan Phase n NPP project was developed, using the RELAP5-HD as core coupled computational code. Based on the coupled model, the steady state calculation and the rod drop transient simulation were performed. The results show that the values obtained from RELAP5-HD calculation agree well with the available measured data, and the calculated accident curves can predict all major parameters trends of the transient with good accuracy. Both steady state and transient calculation results are in accordance with the theoretical analysis from the feedback aspect of coupled reactor neutronics/thermal-hydraulics, this demonstrates that a successful coupled model of Qinshan Phase n NPP core has been developed, and the established model provides a good foundation for further simulation analysis of the nuclear power plant system. (authors)
Transport Phenomena in Magnetized Plasmas across Coupling Regimes
Baalrud, Scott; Daligault, Jerome
2015-11-01
Plasmas with components that are magnetized, strongly coupled, or both arise in a variety of frontier plasma physics experiments including magnetized dusty plasmas, magnetized ICF concepts, as well as from self-generated fields in ICF. Here, a theory is described that treats classical mixtures of magnetized and unmagnetized species across coupling regimes. The approach is based on an extension of the recent effective potential transport theory to include a magnetic field. The utility of this approach is that it can be incorporated into magnetohydrodynamic descriptions by modification of the Coulomb logarithm in the transport coefficients. Like weakly coupled plasma theory, the magnetic field is found to suppress cross-field transport. However, the ratio of parallel to cross field transport rates is much closer to unity at strong coupling. Not only cross field, but also parallel, transport rates are found to be reduced by the field. Results are compared with classical molecular dynamics simulations of self-diffusion of the one component plasma, and with simulations of parallel to perpendicular temperature equilibration of an initially anisotropic distribution. The authors gratefully acknowledge support from Los Alamos National Laboratory grant LDRD 20150520ER.
Energy Technology Data Exchange (ETDEWEB)
An, P.; Yao, D. [Science and Tech. on Reactor System Design Tech. Laboratory, Chengdu (China)
2011-07-01
The MCATHAS system of coupled neutronics/Thermal-hydraulics in supercritical water reactor is described, which considers the mutual influence between the obvious axial and radial evolution of material temperature, water density and the relative power distribution. This system can obtain the main neutronics and thermal parameters along with burn-up. MCATHAS system is parallel processing coupling. The MCNP code is used for neutronics analysis with the continuous cross section library at any temperature calculated by interpolation algorithm; The sub-channel code ATHAS is for thermal-hydraulics analysis and the ORIGEN Code for burn-up calculation. We validate the code with the assembly of HPLWR and analyze the assembly SCLWR- H. (author)
Burn-up measurements coupling gamma spectrometry and neutron measurement
International Nuclear Information System (INIS)
Toubon, H.; Pin, P.; Lebrun, A.; Oriol, L.; Saurel, N.; Gain, T.
2006-01-01
The need to apply for burn-up credit arises with the increase of the initial enrichment of nuclear fuel. When burn-up credit is used in criticality safety studies, it is often necessary to confirm it by measurement. For the last 10 years, CANBERRA has manufactured the PYTHON system for such measurements. However, the method used in the PYTHON itself uses certain reactor data to arrive at burn-up estimates. Based on R and D led by CEA and COGEMA in the framework of burn-up measurement for burn-up credit and safeguards applications, CANBERRA is developing the next generation of burn-up measurement device. This new product, named SMOPY, is able to measure burn-up of any kind of irradiated fuel assembly with a combination of gamma spectrometry and passive neutron measurements. The measurement data is used as input to the CESAR depletion code, which has been developed and qualified by CEA and COGEMA for burn-up credit determinations. In this paper, we explain the complementary nature of the gamma and neutron measurements. In addition, we draw on our previous experience from PYTHON system and from COGEMA La Hague to show what types of evaluations are required to qualify the SMOPY system, to estimate its uncertainties, and to detect discrepancies in the fuel data given by the reactor plant to characterize the irradiated fuel assembly. (authors)
Modular coupling of transport and chemistry: theory and model applications
International Nuclear Information System (INIS)
Pfingsten, W.
1994-06-01
For the description of complex processes in the near-field of a radioactive waste repository, the coupling of transport and chemistry is necessary. A reason for the relatively minor use of coupled codes in this area is the high amount of computer time and storage capacity necessary for calculations by conventional codes, and lack of available data. The simple application of the sequentially coupled code MCOTAC, which couples one-dimensional advective, dispersive and diffusive transport with chemical equilibrium complexation and precipitation/dissolution reactions in a porous medium, shows some promising features with respect to applicability to relevant problems. Transport, described by a random walk of multi-species particles, and chemical equilibrium calculations are solved separately, coupled only by an exchange term to ensure mass conservation. The modular-structured code was applied to three problems: a) incongruent dissolution of hydrated silicate gels, b) dissolution of portlandite and c) calcite dissolution and hypothetical dolomite precipitation. This allows for a comparison with other codes and their applications. The incongruent dissolution of cement phases, important for degradation of cementitious materials in a repository, can be included in the model without the problems which occur with a directly coupled code. The handling of a sharp multi-mineral front system showed a much faster calculation time compared to a directly coupled code application. Altogether, the results are in good agreement with other code calculations. Hence, the chosen modular concept of MCOTAC is more open to an easy extension of the code to include additional processes like sorption, kinetically controlled processes, transport in two or three spatial dimensions, and adaptation to new developments in computing (hardware and software), an important factor for applicability. (author) figs., tabs., refs
International Nuclear Information System (INIS)
Bragt, D.D.B. van.
1995-10-01
A theoretical model for out-of-phase power oscillations in BWRs is proposed. This model describes the dynamic behavior of the neutronic and thermohydraulic subsystems during out-of-phase oscillations, and the coupling of these subsystems via the fuel temperature dynamics and void- and Doppler feedback effects. The zero-power neutron kinetics of the out-of-phase flux density mode is derived by expanding the (time- and space-dependent) neutron flux density in the static solutions of the neutron transport equation. This procedure yields the modal point-kinetic equations for the (first-harmonic) out-of-phase mode. The fuel temperature dynamics is described by a lumped parameter first-order process, characterized by a typical fuel time constant. Using the quasistatic approach, the basic equations of the channel thermohydraulics are derived from the conservation laws of mass and energy and the momentum equation. The momentum equation is coupled with the appropriate boundary condition (constant core pressure drop) for out-of phase oscillations. This procedure yields a set of nonlinear equations describing the dynamic behavior of the boiling boundary, void fraction and mass flux density in the cooling channel. A frequency-domain parametric study confirms that if the out-of-phase mode has a more negative subcriticality, reactor stability increases. On the other hand, a more negative void reactivity coefficient has a destabilizing effect. Besides these two parameters, the fuel time constant was found to be an important parameter determining stability. Where possible, the linearized equations describing the channel thermohydraulics were compare with exact solutions of the governing partial-differential channel equations. This comparison shows that in the frequency range of interest, discrepancies between the proposed quasi-static model and more complicated exact solutions are to be expected. (orig.)
Coupled 3D neutronics/thermal hydraulics modeling of the SAFARI-1 MTR
International Nuclear Information System (INIS)
Rosenkrantz, Adam; Avramova, Maria; Ivanov, Kostadin; Prinsloo, Rian; Botes, Danniëll; Elsakhawy, Khalid
2014-01-01
Highlights: • Development of 3D coupled neutronics/thermal–hydraulic model of SAFARI-1. • Verification of 3D steady-state NEM based neutronics model for SAFARI-1. • Verification of 3D COBRA-TF based thermal–hydraulic model of SAFARI-1. • Quantification of the effect of correct modeling of thermal–hydraulic feedback. - Abstract: The purpose of this study was to develop a coupled accurate multi-physics model of the SAFARI-1 Material Testing Reactor (MTR), a facility that is used for both research and the production of medical isotopes. The model was developed as part of the SAFARI-1 benchmarking project as a cooperative effort between the Pennsylvania State University (PSU) and the South African Nuclear Energy Corporation (Necsa). It was created using a multi-physics coupling of state of the art nuclear reactor simulation tools, consisting of a neutronics code and a thermal hydraulics code. The neutronics tool used was the PSU code NEM, and the results from this component were verified using the Necsa neutronics code OSCAR-4, which is utilized for SAFARI-1 core design and fuel management. On average, the multiplication factors of the neutronics models agreed to within 5 pcm and the radial assembly-averaged powers agreed to within 0.2%. The thermal hydraulics tool used was the PSU version of COBRA-TF (CTF) sub-channel code, and the results of this component were verified against another thermal hydraulics code, the RELAP5-3D system code, used at Necsa for thermal–hydraulics analysis of SAFARI-1. Although only assembly-averaged results from RELAP5-3D were available, they fell within the range of values for the corresponding assemblies in the comprehensive CTF solution. This comparison allows for the first time to perform a quantification of steady-state errors for a low-powered MTR with an advanced thermal–hydraulic code such as CTF on a per-channel basis as compared to simpler and coarser-mesh RELAP5-3D modeling. Additionally, a new cross section
Inter-dot coupling effects on transport through correlated parallel
Indian Academy of Sciences (India)
Transport through symmetric parallel coupled quantum dot system has been studied, using non-equilibrium Green function formalism. The inter-dot tunnelling with on-dot and inter-dot Coulomb repulsion is included. The transmission coefficient and Landaur–Buttiker like current formula are shown in terms of internal states ...
Coupled models in porous media: reactive transport and fractures
International Nuclear Information System (INIS)
Amir, L.
2008-12-01
This thesis deals with numerical simulation of coupled models for flow and transport in porous media. We present a new method for coupling chemical reactions and transport by using a Newton-Krylov method, and we also present a model of flow in fractured media, based on a domain decomposition method that takes into account the case of intersecting fractures. This study is composed of three parts: the first part contains an analysis, and implementation, of various numerical methods for discretizing advection-diffusion problems, in particular by using operator splitting methods. The second part is concerned with a fully coupled method for modeling transport and chemistry problems. The coupled transport-chemistry model is described, after discretization in time, by a system of nonlinear equations. The size of the system, namely the number of grid points times the number a chemical species, precludes a direct solution of the linear system. To alleviate this difficulty, we solve the system by a Newton-Krylov method, so as to avoid forming and factoring the Jacobian matrix. In the last part, we present a model of flow in 3D for intersecting fractures, by using a domain decomposition method. The fractures are treated as interfaces between sub-domains. We show existence and uniqueness of the solution, and we validate the model by numerical tests. (author)
Transport zonation limits coupled nitrification-denitrification in permeable sediments
DEFF Research Database (Denmark)
Kessler, Adam J.; Glud, Ronnie N.; Cardenas, M. Bayani
2013-01-01
into the coupling between ammonification, nitrification and denitrification in stationary sand ripples, we combined the diffusion equilibrium thin layer (DET) gel technique with a computational reactive transport biogeochemical model. The former approach provided high-resolution two-dimensional distributions of NO3......- and N-15-N-2 gas. The measured two-dimensional profiles correlate with computational model simulations, showing a deep pool of N-2 gas forming, and being advected to the surface below ripple peaks. Further isotope pairing calculations on these data indicate that coupled nitrification-denitrification......Measurement of biogeochemical processes in permeable sediments (including the hyporheic zone) is difficult because of complex multidimensional advective transport. This is especially the case for nitrogen cycling, which involves several coupled redox-sensitive reactions. To provide detailed insight...
International Nuclear Information System (INIS)
Jevremovic, Tatjana; Hursin, Mathieu; Satvat, Nader; Hopkins, John; Xiao, Shanjie; Gert, Godfree
2006-01-01
The AGENT (Arbitrary Geometry Neutron Transport) an open-architecture reactor modeling tool is deterministic neutron transport code for two or three-dimensional heterogeneous neutronic design and analysis of the whole reactor cores regardless of geometry types and material configurations. The AGENT neutron transport methodology is applicable to all generations of nuclear power and research reactors. It combines three theories: (1) the theory of R-functions used to generate real three-dimensional whole-cores of square, hexagonal or triangular cross sections, (2) the planar method of characteristics used to solve isotropic neutron transport in non-homogenized 2D) reactor slices, and (3) the one-dimensional diffusion theory used to couple the planar and axial neutron tracks through the transverse leakage and angular mesh-wise flux values. The R-function-geometrical module allows a sequential building of the layers of geometry and automatic sub-meshing based on the network of domain functions. The simplicity of geometry description and selection of parameters for accurate treatment of neutron propagation is achieved through the Boolean algebraic hierarchically organized simple primitives into complex domains (both being represented with corresponding domain functions). The accuracy is comparable to Monte Carlo codes and is obtained by following neutron propagation through real geometrical domains that does not require homogenization or simplifications. The efficiency is maintained through a set of acceleration techniques introduced at all important calculation levels. The flux solution incorporates power iteration with two different acceleration techniques: Coarse Mesh Re-balancing (CMR) and Coarse Mesh Finite Difference (CMFD). The stand-alone originally developed graphical user interface of the AGENT code design environment allows the user to view and verify input data by displaying the geometry and material distribution. The user can also view the output data such
[H(+)-coupled heavy metal transport in plants].
Migocka, Magdalena; Nowojska, Ewa; Kłobus, Grazyna
2007-01-01
It has been recently well documented that metal transport systems play a crucial role in the uptake, distribution and detoxification of heavy metals throughout the plant. A range of gene families that are likely to be involved in essential and non-essential metal transport has been now identified and their plasma membrane and/or tonoplast localization in plant cells has been recently confirmed. These include the primary metal transporters, using ATP as the source of energy and H(+)-coupling transporters, utilizing the electrochemical gradient previously generated by plasma membrane and tonoplast proton pumps. As the presence of nucleotide binding domains in the protein sequence may indicate its ATP-hydrolytic activity, it is more difficult to determine the H(+)-coupling activity of protein on the base of its structure. Thus, the H(+)-coupling activity of protein may be only proved by functional analysis of the protein. In this work, we briefly review the structure, regulation and function of the metal transporters operating as H(+)/metal cotransporters.
Transport calculation of neutron flux distribution in reflector of PW reactor
International Nuclear Information System (INIS)
Remec, I.
1982-01-01
Two-dimensional transport calculation of the neutron flux and spectrum in the equatorial plain of PW reactor, using computer program DOT 3, is presented. Results show significant differences between neutron fields in which test samples and reactor vessel are exposed. (author)
International Nuclear Information System (INIS)
Bareiss, E.H.
1977-08-01
The objectives of this research are to develop mathematically and computationally founded criteria for the design of highly efficient and reliable multidimensional neutron transport codes to solve a variety of neutron migration and radiation problems, and to analyze existing and new methods for performance
FMCEIR: a Monte Carlo program for solving the stationary neutron and gamma transport equation
International Nuclear Information System (INIS)
Taormina, A.
1978-05-01
FMCEIR is a three-dimensional Monte Carlo program for solving the stationary neutron and gamma transport equation. It is used to study the problem of neutron and gamma streaming in the GCFR and HHT reactor channels. (G.T.H.)
Prototype coupling of the CFD software ansys CFX with the 3D neutron kinetic core model DYN3D - 249
International Nuclear Information System (INIS)
Kliem, S.; Rohde, U.; Schutze, J.; Frank, Th.
2010-01-01
The CFD code ANSYS CFX has been coupled with the neutron-kinetic core model DYN3D. ANSYS CFX calculates the fluid dynamics and related transport phenomena in the reactor's coolant and provides the corresponding data to DYN3D. In the fluid flow simulation of the coolant, the core itself is modeled within the porous body approach. DYN3D calculates the neutron kinetics and the fuel behavior including the heat transfer to the coolant. The physical data interface between the codes is the volumetric heat release rate into the coolant. In the prototype that is currently available, the coupling is restricted to single-phase flow problems. In the time domain an explicit coupling of the codes has been implemented so far. Steady-state and transient verification calculations for a small-size test problem confirm the correctness of the implementation of the prototype coupling. This test problem was a mini-core consisting of nine real-size fuel assemblies. Comparison was performed with the DYN3D standalone code. In the steady state, the effective multiplication factor obtained by the ANSYS CFX/DYN3D codes shows a deviation of 9.8 pcm from the DYN3D stand-alone solution. This difference can be attributed to the use of different water property packages in the two codes. The transient test case simulated the withdrawal of the control rod from the central fuel assembly at hot zero power. Power increase during the introduction of positive reactivity and power reduction due to fuel temperature increase are calculated in the same manner by the coupled and the stand-alone codes. The maximum values reached during the power rise differ by about 1 MW at a power level of 50 MW. Beside the different water property packages, these differences are caused by the use of different flow solvers. (authors)
Cooperative learning of neutron diffusion and transport theories
International Nuclear Information System (INIS)
Robinson, Michael A.
1999-01-01
A cooperative group instructional strategy is being used to teach a unit on neutron transport and diffusion theory in a first-year-graduate level, Reactor Theory course that was formerly presented in the traditional lecture/discussion style. Students are divided into groups of two or three for the duration of the unit. Class meetings are divided into traditional lecture/discussion segments punctuated by cooperative group exercises. The group exercises were designed to require the students to elaborate, summarize, or practice the material presented in the lecture/discussion segments. Both positive interdependence and individual accountability are fostered by adjusting individual grades on the unit exam by a factor dependent upon group achievement. Group collaboration was also encouraged on homework assignments by assigning each group a single grade on each assignment. The results of the unit exam have been above average in the two classes in which the cooperative group method was employed. In particular, the problem solving ability of the students has shown particular improvement. Further,the students felt that the cooperative group format was both more educationally effective and more enjoyable than the lecture/discussion format
Neutron and photon transport calculations in fusion system. 2
Energy Technology Data Exchange (ETDEWEB)
Sato, Satoshi [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment
1998-03-01
On the application of MCNP to the neutron and {gamma}-ray transport calculations for fusion reactor system, the wide range design calculation has been carried out in the engineering design activities for the international thermonuclear fusion experimental reactor (ITER) being developed jointly by Japan, USA, EU and Russia. As the objects of shielding calculation for fusion reactors, there are the assessment of dose equivalent rate for living body shielding and the assessment of the nuclear response for the soundness of in-core structures. In the case that the detailed analysis of complicated three-dimensional shapes is required, the assessment using MCNP has been carried out. Also when the nuclear response of peripheral equipment due to the gap streaming between blanket modules is evaluated with good accuracy, the calculation with MCNP has been carried out. The analyses of the shieldings for blanket modules and NBI port are explained, and the examples of the results of analyses are shown. In the blanket modules, there are penetrating holes and continuous gap. In the case of the NBI port, shielding plug cannot be installed. These facts necessitate the MCNP analysis with high accuracy. (K.I.)
A method for transient, three-dimensional neutron transport calculations
Energy Technology Data Exchange (ETDEWEB)
Waddell, M.W. Jr. (Oak Ridge Y-12 Plant, TN (United States)); Dodds, H.L. (Tennessee Univ., Knoxville, TN (United States))
1992-12-28
This paper describes the development and evaluation of a method for solving the time-dependent, three-dimensional Boltzmann transport model with explicit representation of delayed neutrons. A hybrid stochastic/deterministic technique is utilized with a Monte Carlo code embedded inside of a quasi-static kinetics framework. The time-dependent flux amplitude, which is usually fast varying, is computed deterministically by a conventional point kinetics algorithm. The point kinetics parameters, reactivity and generation time as well as the flux shape, which is usually slowly varying in time, are computed stochastically during the random walk of the Monte Carlo calculation. To verify the accuracy of this new method, several computational benchmark problems from the Argonne National Laboratory benchmark book, ANL-7416, were calculated. The results are shown to be in reasonably good agreement with other independently obtained solutions. The results obtained in this work indicate that the method/code is working properly and that it is economically feasible for many practical applications provided a dedicated high performance workstation is available.
A method for transient, three-dimensional neutron transport calculations
Energy Technology Data Exchange (ETDEWEB)
Waddell, M.W. Jr. (Martin Marietta Energy Systems, Inc. (United States)); Dodds, H.L. (Univ. of Tennessee (United States))
1993-04-01
This paper describes the development and evaluation of a method for solving the time-dependent, three-dimensional Boltzmann transport model with explicit representation of delayed neutrons. A hybrid stochastic/deterministic technique is utilized with a Monte Carlo code embedded inside of a quasi-static kinetics framework. The time-dependent flux amplitude, which is usually fast varying, is computed deterministically by a conventional point kinetics algorithm. The point kinetics parameters, reactivity and generation time as well as the flux shape, which is usually slowly varying in time, are computed stochastically during the random walk of the Monte Carlo calculation. To verify the accuracy of this new method, several computational benchmark problems from the Argonne National Laboratory benchmark book, ANL-7416, were calculated. The results are shown to be in reasonably good agreement with other independently obtained solutions. The results obtained in this work indicate that the method/code is working properly and that it is economically feasible for many practical applications provided a dedicated high performance workstation is available. (orig.)
Application of Trotter approximation for solving time dependent neutron transport equation
International Nuclear Information System (INIS)
Stancic, V.
1987-01-01
A method is proposed to solve multigroup time dependent neutron transport equation with arbitrary scattering anisotropy. The recurrence relation thus obtained is simple, numerically stable and especially suitable for treatment of complicated geometries. (author)
Light-water-reactor coupled neutronic and thermal-hydraulic codes
International Nuclear Information System (INIS)
Diamond, D.J.
1982-01-01
An overview is presented of computer codes that model light water reactor cores with coupled neutronics and thermal-hydraulics. This includes codes for transient analysis and codes for steady state analysis which include fuel depletion and fission product buildup. Applications in nuclear design, reactor operations and safety analysis are given and the major codes in use in the USA are identified. The neutronic and thermal-hydraulic methodologies and other code features are outlined for three steady state codes (PDQ7, NODE-P/B and SIMULATE) and four dynamic codes (BNL-TWIGL, MEKIN, RAMONA-3B, RETRAN-02). Speculation as to future trends with such codes is also presented
Light-water-reactor coupled neutronic and thermal-hydraulic codes
Energy Technology Data Exchange (ETDEWEB)
Diamond, D.J.
1982-01-01
An overview is presented of computer codes that model light water reactor cores with coupled neutronics and thermal-hydraulics. This includes codes for transient analysis and codes for steady state analysis which include fuel depletion and fission product buildup. Applications in nuclear design, reactor operations and safety analysis are given and the major codes in use in the USA are identified. The neutronic and thermal-hydraulic methodologies and other code features are outlined for three steady state codes (PDQ7, NODE-P/B and SIMULATE) and four dynamic codes (BNL-TWIGL, MEKIN, RAMONA-3B, RETRAN-02). Speculation as to future trends with such codes is also presented.
Simulation of neutron transport equation using parallel Monte Carlo for deep penetration problems
International Nuclear Information System (INIS)
Bekar, K. K.; Tombakoglu, M.; Soekmen, C. N.
2001-01-01
Neutron transport equation is simulated using parallel Monte Carlo method for deep penetration neutron transport problem. Monte Carlo simulation is parallelized by using three different techniques; direct parallelization, domain decomposition and domain decomposition with load balancing, which are used with PVM (Parallel Virtual Machine) software on LAN (Local Area Network). The results of parallel simulation are given for various model problems. The performances of the parallelization techniques are compared with each other. Moreover, the effects of variance reduction techniques on parallelization are discussed
Energy Technology Data Exchange (ETDEWEB)
Brew, D.R.M. [ANSTO (Australian Nuclear Science and Technology Organisation), Menai, NSW 2234 (Australia)], E-mail: dbr@ansto.gov.au; Beer, F.C. de; Radebe, M.J.; Nshimirimana, R. [Necsa (South African Nuclear Energy Corporation), Pretoria (South Africa); McGlinn, P.J.; Aldridge, L.P.; Payne, T.E. [ANSTO (Australian Nuclear Science and Technology Organisation), Menai, NSW 2234 (Australia)
2009-06-21
In this preliminary study we use neutron radiography and tomography to examine differences in water transport through cement pastes and mortars. Bulk residual water contents and sorptivity curves determined using neutron radiography are compared with data obtained gravimetrically. In addition, macro-pore volume distributions of each sample were measured. Furthermore, it was possible to use neutron radiography to monitor the change in the mass of water when samples were dried or when water moved into the samples. The trends and absolute values of weight loss and gain obtained using both approaches are very consistent for mortars, especially when a neutron-scattering correction is applied.
Coupled Transport Phenomena in the Opalinus Clay: Implications for Radionuclide Transport
International Nuclear Information System (INIS)
Soler, J.M.
1999-09-01
Coupled phenomena (thermal and chemical osmosis, hyperfiltration, coupled diffusion, thermal diffusion, thermal filtration, Dufour effect) may play an important role in fluid, solute and heat transport in clay-rich formations, such as the Opalinus Clay (OPA), which are being considered as potential hosts for radioactive waste repositories. In this study, the potential effects of coupled phenomena on radionuclide transport in the vicinity of a repository for vitrified high-level radioactive waste (HLW) and spent nuclear fuel (SF) hosted by the Opalinus Clay, at times equal to or greater than the expected lifetime of the waste canisters (about 1000 years), have been addressed. Firstly, estimates of the solute fluxes associated with chemical osmosis, hyperfiltration, thermal diffusion and thermal osmosis have been calculated. Available experimental data concerning coupled transport phenomena in compacted clays, and the hydrogeological and geochemical conditions to which the Opalinus Clay is subject, have been used for these estimates. These estimates suggest that thermal osmosis is the only coupled transport mechanism that could have a strong impact on solute and fluid transport in the vicinity of the repository. Secondly, estimates of the heat fluxes associated with thermal filtration and the Dufour effect in the vicinity of the repository have been calculated. The calculated heat fluxes are absolutely negligible compared to the heat flux caused by thermal conduction. As a further step to obtain additional insight into the effects of coupled phenomena on solute transport, the solute fluxes associated with advection, chemical diffusion, thermal and chemical osmosis, hyperfiltration and thermal diffusion have been incorporated into a simple one-dimensional transport equation. The analytical solution of this equation, with appropriate parameters, shows again that thermal osmosis is the only coupled transport mechanism that could have a strong effect on repository
Neutron transport simulation in high speed moving media using Geant4
Li, G.; Ciungu, B.; Harrisson, G.; Rogge, R. B.; Tun, Z.; van der Ende, B. M.; Zwiers, I.
2017-12-01
A method using Geant4 to simulate neutron transport in moving media is described. The method is implanted in the source code of the software since Geant4 does not intrinsically support a moving object. The simulation utilizes the existing physical model and data library in Geant4, combined with frame transformations to account for the effect of relative velocity between neutrons and the moving media. An example is presented involving a high speed rotating cylinder to verify this method and show the effect of moving media on neutron transport.
National Research Council Canada - National Science Library
Labowski, Kristofer
2001-01-01
The Linear Characteristic (LC) method on rectangular boxoid meshes is a discrete ordinate neutron transport technique that uses both zeroth and first moments of the angular neutron flux to construct a relatively accurate...
Nuttin, A.; Capellan, N.; David, S.; Doligez, X.; El Mhari, C.; Méplan, O.
2014-06-01
Safety analysis of innovative reactor designs requires three dimensional modeling to ensure a sufficiently realistic description, starting from steady state. Actual Monte Carlo (MC) neutron transport codes are suitable candidates to simulate large complex geometries, with eventual innovative fuel. But if local values such as power densities over small regions are needed, reliable results get more difficult to obtain within an acceptable computation time. In this scope, NEA has proposed a performance test of full PWR core calculations based on Monte Carlo neutron transport, which we have used to define an optimal detail level for convergence of steady state coupled neutronics. Coupling between MCNP for neutronics and the subchannel code COBRA for thermal-hydraulics has been performed using the C++ tool MURE, developed for about ten years at LPSC and IPNO. In parallel with this study and within the same MURE framework, a simplified code of nodal kinetics based on two-group and few-point diffusion equations has been developed and validated on a typical CANDU LOCA. Methods for the computation of necessary diffusion data have been defined and applied to NU (Nat. U) and Th fuel CANDU after assembly evolutions by MURE. Simplicity of CANDU LOCA model has made possible a comparison of these two fuel behaviours during such a transient.
Optimization of spring exchange coupled ferrites, studied by in situ neutron diffraction
DEFF Research Database (Denmark)
Ahlburg, Jakob; Christensen, Mogens; Granados-Miralles, Cecilia
) is reduced to a metallic alloy CoFe (soft magnet) by heating the sample and flowing it with hydrogen gas. It is studied in situ using neutron powder diffraction with a time resolution of 12 min. The transition from spinel to pure metal goes through an intermediate step of a metal oxide before being fully...... reduced. These metal oxides are antiferromagnetically ordered an is therefore considered a parasitic phase. However by fine-tuning the reaction temperature and hydrogen flow rate the occurrence of the phase can be minimized. In order to distinguish between Co and Fe Neutrons are chosen. Since neutrons...... have a spin it will also be possible to measure a magnetic signal and investigate the exchange-coupling. After the reduction the samples was furthermore investigated using powder x-ray diffraction and VSM (vibrating sample magnetometer). To understand the reaction mechanism, a series of experiments...
Peptide Selectivity of the Proton-Coupled Oligopeptide Transporter from Neisseria meningitidis
DEFF Research Database (Denmark)
Sharma, Neha; Aduri, Nanda G; Iqbal, Anna
2016-01-01
Peptide transport in living organisms is facilitated by either primary transport, hydrolysis of ATP, or secondary transport, cotransport of protons. In this study, we focused on investigating the ligand specificity of the Neisseria meningitidis proton-coupled oligopeptide transporter (Nm...
3D neutronic codes coupled with thermal-hydraulic system codes for PWR, and BWR and VVER reactors
Energy Technology Data Exchange (ETDEWEB)
Langenbuch, S.; Velkov, K. [GRS, Garching (Germany); Lizorkin, M. [Kurchatov-Institute, Moscow (Russian Federation)] [and others
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.
Wentzel-Bardeen singularity in coupled Luttinger liquids: Transport properties
International Nuclear Information System (INIS)
Martin, T.
1994-01-01
The recent progress on 1 D interacting electrons systems and their applications to study the transport properties of quasi one dimensional wires is reviewed. We focus on strongly correlated elections coupled to low energy acoustic phonons in one dimension. The exponents of various response functions are calculated, and their striking sensitivity to the Wentzel-Bardeen singularity is discussed. For the Hubbard model coupled to phonons the equivalent of a phase diagram is established. By increasing the filling factor towards half filling the WB singularity is approached. This in turn suppresses antiferromagnetic fluctuations and drives the system towards the superconducting regime, via a new intermediate (metallic) phase. The implications of this phenomenon on the transport properties of an ideal wire as well as the properties of a wire with weak or strong scattering are analyzed in a perturbative renormalization group calculation. This allows to recover the three regimes predicted from the divergence criteria of the response functions
Vibrationally coupled electron transport through single-molecule junctions
Energy Technology Data Exchange (ETDEWEB)
Haertle, Rainer
2012-04-26
vibrational effects have a profound influence on the transport characteristics of a single-molecule contact and play therefore a fundamental role in this transport problem. Our findings demonstrate that vibrationally coupled electron transport through a molecular junction involves two types of processes: (i) transport processes, where an electron tunnels through the molecular bridge from one lead to the other, and (ii) electron-hole pair creation processes, where an electron tunnels from one of the leads onto the molecular bridge and back to the same lead again. Transport processes directly contribute to the electrical current flowing through a molecular contact and involve both excitation and deexcitation processes of the vibrational modes of the junction. Electron-hole pair creation processes do not directly contribute to the electrical current and typically involve only deexcitation processes. Nevertheless, they constitute a cooling mechanism for the vibrational modes of a single-molecule junction that is as important as cooling by transport processes. As the level of vibrational excitation determines the efficiency of electron transport processes, they have an indirect influence on the electrical current flowing through the junction. As we show, however, this influence can be substantial, in particular, if the molecule is coupled asymmetrically to the leads. Accounting for all these processes and their complex interrelationship, we analyze a number of intriguing transport phenomena, including rectification, negative differential resistance, anomalous peak broadening, mode-selective vibrational excitation and vibrationally induced decoherence. Moreover, we show that higher levels of vibrational excitation are obtained for weaker electronic-vibrational coupling. Thus, based on physical grounds, we establish a relation between the weak electronic-vibrational coupling limit and the limit of large bias voltages, where the level of vibrational excitation in a molecular junction
Parallel computing solution of Boltzmann neutron transport equation
International Nuclear Information System (INIS)
Ansah-Narh, T.
2010-01-01
The focus of the research was on developing parallel computing algorithm for solving Eigen-values of the Boltzmam Neutron Transport Equation (BNTE) in a slab geometry using multi-grid approach. In response to the problem of slow execution of serial computing when solving large problems, such as BNTE, the study was focused on the design of parallel computing systems which was an evolution of serial computing that used multiple processing elements simultaneously to solve complex physical and mathematical problems. Finite element method (FEM) was used for the spatial discretization scheme, while angular discretization was accomplished by expanding the angular dependence in terms of Legendre polynomials. The eigenvalues representing the multiplication factors in the BNTE were determined by the power method. MATLAB Compiler Version 4.1 (R2009a) was used to compile the MATLAB codes of BNTE. The implemented parallel algorithms were enabled with matlabpool, a Parallel Computing Toolbox function. The option UseParallel was set to 'always' and the default value of the option was 'never'. When those conditions held, the solvers computed estimated gradients in parallel. The parallel computing system was used to handle all the bottlenecks in the matrix generated from the finite element scheme and each domain of the power method generated. The parallel algorithm was implemented on a Symmetric Multi Processor (SMP) cluster machine, which had Intel 32 bit quad-core x 86 processors. Convergence rates and timings for the algorithm on the SMP cluster machine were obtained. Numerical experiments indicated the designed parallel algorithm could reach perfect speedup and had good stability and scalability. (au)
Coester, Annemieke M.; Smit, Watske; Struijk, Dirk G.; Krediet, Raymond T.
2009-01-01
Ultrafiltration in peritoneal dialysis occurs through endothelial water channels (free water transport) and together with solutes across small pores: solute coupled water transport. A review is given of cross-sectional studies and on the results of longitudinal follow-up
International Nuclear Information System (INIS)
Brenner, D.J.; Prael, R.E.; Little, R.C.
1987-01-01
Realistic simulations of the passage of fast neutrons through tissue require a large quantity of cross-sectional data. What are needed are differential (in particle type, energy and angle) cross sections. A computer code is described which produces such spectra for neutrons above ∼14 MeV incident on light nuclei such as carbon and oxygen. Comparisons have been made with experimental measurements of double-differential secondary charged-particle production on carbon and oxygen at energies from 27 to 60 MeV; they indicate that the model is adequate in this energy range. In order to utilize fully the results of these calculations, they should be incorporated into a neutron transport code. This requires defining a generalized format for describing charged-particle production, putting the calculated results in this format, interfacing the neutron transport code with these data, and charged-particle transport. The design and development of such a program is described. 13 refs., 3 figs
Improving Neutron Kinetics and Thermal Hydraulics coupled tools for BEPU calculations
Energy Technology Data Exchange (ETDEWEB)
Pericas, R.; Reventós, F.; Batet, Il.
2015-07-01
The BEPU methodology is capable of providing a solution in terms of increasing the nuclear power production without compromising the safety margins. This study presents different improvements performed using tools available at UPC in the field of Neutron Kinetics and Thermal Hydraulics coupled systems. The paper describes a comparison between the BEPU methodology and the Conservative Bounding methodology within the framework of the Neutron Kinetics and Thermal Hydraulics coupled systems. To perform such comparison the following tools have been selected: TRACE for thermal-hydraulic system calculations, PARCS for reactor kinetics core simulator code. A Main Steam Line Break (MSLB) in a Pressurized Water Reactor (PWR) is the selected simulated transient to show the improvements performed. (Author)
Transport behavior of coupled continuous-time random walks.
Dentz, Marco; Scher, Harvey; Holder, Devora; Berkowitz, Brian
2008-10-01
The origin of anomalous or non-Fickian transport in disordered media is the broad spectrum of transition rates intrinsic to these systems. A system that contains within it heterogeneities over multiple length scales is geological formations. The continuous time random walk (CTRW) framework, which has been demonstrated to be an effective means to model non-Fickian transport features in these systems and to have predictive capacities, has at its core this full spectrum represented as a joint probability density psi(s,t) of random space time displacements (s,t) . Transport in a random fracture network (RFN) has been calculated with a coupled psi(s,t) and has subsequently been shown to be approximated well by a decoupled form psi(s,t)=F(s)psi(t) . The latter form has been used extensively to model non-Fickian transport in conjunction with a velocity distribution Phi(xi),xi identical with 1v, where v is the velocity magnitude. The power-law behavior of psi(t) proportional to (-1-beta), which determines non-Fickian transport, derives from the large xi dependence of Phi(xi) . In this study we use numerical CTRW simulations to explore the expanded transport phenomena derived from a coupled psi(s,t) . Specifically, we introduce the features of a power-law dependence in the s distribution with different Phi(xi) distributions (including a constant v) coupled by t=s(xi) . Unlike Lévy flights in this coupled scenario the spatial moments of the plumes are well defined. The shapes of the plumes depend on the entire Phi(xi) distribution, i.e., both small and large xi dependence; there is a competition between long displacements (which depend on the small xi dependence) and large time events (which depend on a power law for large xi). These features give rise to an enhanced range of transport behavior with a broader scope of applications, e.g., to correlated migrations in a RFN and in heterogeneous permeability fields. The approximation to the decoupled case is investigated as a
Measurement of spectrum for thermal neutrons produced from H2O moderator coupled with mercury target
International Nuclear Information System (INIS)
Meigo, S.; Maekawa, F.; Kasugai, Y.; Nakashima, H.; Ikeda, Y.; Watanabe, N.
2001-01-01
In order to obtain fundamental data for the design of pulsed spallation neutron source, the slowing-down and thermalized neutrons from an H 2 O moderator coupled with the mercury target were measured using GeV proton beams at AGS (Alternative Gradient Synchrotron) in BNL (Brookhaven National Laboratory) under the ASTE (AGS-Spallation Target Experiment) collaboration. The mercury target (φ 20 cm x L 130 cm) was surrounded by a lead reflector (1 x 1 x 1 m 3 ) was irradiated by 1.94-, 12- and 24-GeV protons. The spectral intensities of thermal neutrons from the moderator are measured by the current-mode time-of-flight technique using enriched 6 Li and 7 Li glass scintillators. By this technique, only several incident pulses were needed to obtain sufficient statistics for incident energy. The results have shown that the neutron spectral intensity per proton integrated over the Maxwellian region was almost proportional to the proton energy. By moving the target along the beam direction within 15 cm, the dependence of the relative moderator position to the target on the neutron flux was also measured. With this position change, the difference with flux was found within 10%. (author)
Cosmic ray heliospheric transport study with neutron monitor data
Ahluwalia, H. S.; Ygbuhay, R. C.; Modzelewska, R.; Dorman, L. I.; Alania, M. V.
2015-10-01
Determining transport coefficients for galactic cosmic ray (GCR) propagation in the turbulent interplanetary magnetic field (IMF) poses a fundamental challenge in modeling cosmic ray modulation processes. GCR scattering in the solar wind involves wave-particle interaction, the waves being Alfven waves which propagate along the ambient field (B). Empirical values at 1 AU are determined for the components of the diffusion tensor for GCR propagation in the heliosphere using neutron monitor (NM) data. At high rigidities, particle density gradients and mean free paths at 1 AU in B can only be computed from the solar diurnal anisotropy (SDA) represented by a vector A (components Ar, Aϕ, and Aθ) in a heliospherical polar coordinate system. Long-term changes in SDA components of NMs (with long track record and the median rigidity of response Rm ~ 20 GV) are used to compute yearly values of the transport coefficients for 1963-2013. We confirm the previously reported result that the product of the parallel (to B) mean free path (λ||) and radial density gradient (Gr) computed from NM data exhibits a weak Schwabe cycle (11y) but strong Hale magnetic cycle (22y) dependence. Its value is most depressed in solar activity minima for positive (p) polarity intervals (solar magnetic field in the Northern Hemisphere points outward from the Sun) when GCRs drift from the polar regions toward the helioequatorial plane and out along the heliospheric current sheet (HCS), setting up a symmetric gradient Gθs pointing away from HCS. Gr drives all SDA components and λ|| Gr contributes to the diffusive component (Ad) of the ecliptic plane anisotropy (A). GCR transport is commonly discussed in terms of an isotropic hard sphere scattering (also known as billiard-ball scattering) in the solar wind plasma. We use it with a flat HCS model and the Ahluwalia-Dorman master equations to compute the coefficients α (=λ⊥/λ∥) and ωτ (a measure of turbulence in the solar wind) and transport
International Nuclear Information System (INIS)
Kliem, S.; Grahn, A.; Rohde, U.; Schuetze, J.; Frank, Th.
2010-01-01
The computational fluid dynamics code ANSYS CFX has been coupled with the neutron-kinetic core model DYN3D. ANSYS CFX calculates the fluid dynamics and related transport phenomena in the reactors coolant and provides the corresponding data to DYN3D. In the fluid flow simulation of the coolant, the core itself is modeled within the porous body approach. DYN3D calculates the neutron kinetics and the fuel behavior including the heat transfer to the coolant. The physical data interface between the codes is the volumetric heat release rate into the coolant. In the prototype that is currently available, the coupling is restricted to single-phase flow problems. In the time domain an explicit coupling of the codes has been implemented so far. Steady-state and transient verification calculations for two small-size test problems confirm the correctness of the implementation of the prototype coupling. The first test problem was a mini-core consisting of nine real-size fuel assemblies with quadratic cross section. Comparison was performed with the DYN3D stand-alone code. In the steady state, the effective multiplication factor obtained by the DYN3D/ANSYS CFX codes hows a deviation of 9.8 pcm from the DYN3D stand-alone solution. This difference can be attributed to the use of different water property packages in the two codes. The transient test case simulated the withdrawal of the control rod from the central fuel assembly at hot zero power in the same mini-core. Power increase during the introduction of positive reactivity and power reduction due to fuel temperature increase are calculated in the same manner by the coupled and the stand-alone codes. The maximum values reached during the power rise differ by about 1 MW at a power level of 50 MW. Beside the different water property packages, these differences are caused by the use of different flow solvers. The same calculations were carried for a mini-core with seven real-size fuel assemblies with hexagonal cross section in
Resolution of the neutron transport equation by massively parallel computer in the Cronos code
International Nuclear Information System (INIS)
Zardini, D.M.
1996-01-01
The feasibility of neutron transport problems parallel resolution by CRONOS code's SN module is here studied. In this report we give the first data about the parallel resolution by angular variable decomposition of the transport equation. Problems about parallel resolution by spatial variable decomposition and memory stage limits are also explained here. (author)
The infinite medium Green's function for neutron transport in plane geometry 40 years later
International Nuclear Information System (INIS)
Ganapol, B.D.
1993-01-01
In 1953, the first of what was supposed to be two volumes on neutron transport theory was published. The monograph, entitled open-quotes Introduction to the Theory of Neutron Diffusionclose quotes by Case et al., appeared as a Los Alamos National Laboratory report and was to be followed by a second volume, which never appeared as intended because of the death of Placzek. Instead, Case and Zweifel collaborated on the now classic work entitled Linear Transport Theory 2 in which the underlying mathematical theory of linear transport was presented. The initial monograph, however, represented the coming of age of neutron transport theory, which had its roots in radiative transfer and kinetic theory. In addition, it provided the first benchmark results along with the mathematical development for several fundamental neutron transport problems. In particular, one-dimensional infinite medium Green's functions for the monoenergetic transport equation in plane and spherical geometries were considered complete with numerical results to be used as standards to guide code development for applications. Unfortunately, because of the limited computational resources of the day, some numerical results were incorrect. Also, only conventional mathematics and numerical methods were used because the transport theorists of the day were just becoming acquainted with more modern mathematical approaches. In this paper, Green's function solution is revisited in light of modern numerical benchmarking methods with an emphasis on evaluation rather than theoretical results. The primary motivation for considering the Green's function at this time is its emerging use in solving finite and heterogeneous media transport problems
Application of neutron/gamma transport codes for the design of explosive detection systems
International Nuclear Information System (INIS)
Elias, E.; Shayer, Z.
1994-01-01
Applications of neutron and gamma transport codes to the design of nuclear techniques for detecting concealed explosives material are discussed. The methodology of integrating radiation transport computations in the development, optimization and analysis phases of these new technologies is discussed. Transport and Monte Carlo codes are used for proof of concepts, guide the system integration, reduce the extend of experimental program and provide insight into the physical problem involved. The paper concentrates on detection techniques based on thermal and fast neutron interactions in the interrogated object. (authors). 6 refs., 1 tab., 5 figs
Niranjan, Ram; Rout, R K; Srivastava, R; Kaushik, T C; Gupta, Satish C
2016-03-01
A 17 kJ transportable plasma focus (PF) device with flexible transmission lines is developed and is characterized. Six custom made capacitors are used for the capacitor bank (CB). The common high voltage plate of the CB is fixed to a centrally triggered spark gap switch. The output of the switch is coupled to the PF head through forty-eight 5 m long RG213 cables. The CB has a quarter time-period of 4 μs and an estimated current of 506 kA is delivered to the PF device at 17 kJ (60 μF, 24 kV) energy. The average neutron yield measured using silver activation detector in the radial direction is (7.1 ± 1.4) × 10(8) neutrons/shot over 4π sr at 5 mbar optimum D2 pressure. The average neutron yield is more in the axial direction with an anisotropy factor of 1.33 ± 0.18. The average neutron energies estimated in the axial as well as in the radial directions are (2.90 ± 0.20) MeV and (2.58 ± 0.20) MeV, respectively. The flexibility of the PF head makes it useful for many applications where the source orientation and the location are important factors. The influence of electromagnetic interferences from the CB as well as from the spark gap on applications area can be avoided by putting a suitable barrier between the bank and the PF head.
International Nuclear Information System (INIS)
Wuest, C.R.
1993-01-01
The coupled neutron/photon transport code TART has been used to calculate the attenuation of neutrons and the production of induced photons for neutrons incidents on 5% and 20% borated polyethylene slabs. The neutron attenuation lengths are found to be 2.4 cm and 2.9 cm for 5% and 20% borated polyethylene, respectively
Modelling of neutron absorbers in high temperature reactors by combined transport diffusion methods
Fen, V.; Lebedev, M.; Sarytchev, V.; Scherer, W.
1992-01-01
Today, the neutron-physical description of strong neutron absorbing materials for control and shut-down of nuclear power plants is performed using combined transport and diffusion methods. Two of these approaches are described and compared in this paper. The method of equivalent cross-sections has been developed at the KFA-Jülich Institute for Safety Research and Reactor Technology (ISR) and was widely used for all german HTR reactor concepts. The Obninsk Institute for Nuclear Power Engineeri...
Neutron transport for pure-triplet scattering in finite planar media with reflective boundaries
International Nuclear Information System (INIS)
Sallah, M.; Degheidy, A.R.
2008-01-01
Pure-triplet scattering in neutron transport through a finite plane-parallel medium with internal source of energy is considered. The medium is assumed to have specular- and diffusely-reflecting boundaries. The neutron partial heat fluxes for this problem are computed in terms of the albedos of the source-free problem. Pomraning-Eddington approximation is used to solve the source free problem. A weight function is introduced to force the boundary conditions to be fulfilled
Transport synthetic acceleration scheme for multi-dimensional neutron transport problems
International Nuclear Information System (INIS)
Modak, R.S.; Vinod Kumar; Menon, S.V.G.; Gupta, Anurag
2005-09-01
The numerical solution of linear multi-energy-group neutron transport equation is required in several analyses in nuclear reactor physics and allied areas. Computer codes based on the discrete ordinates (Sn) method are commonly used for this purpose. These codes solve external source problem and K-eigenvalue problem. The overall solution technique involves solution of source problem in each energy group as intermediate procedures. Such a single-group source problem is solved by the so-called Source Iteration (SI) method. As is well-known, the SI-method converges very slowly for optically thick and highly scattering regions, leading to large CPU times. Over last three decades, many schemes have been tried to accelerate the SI; the most prominent being the Diffusion Synthetic Acceleration (DSA) scheme. The DSA scheme, however, often fails and is also rather difficult to implement. In view of this, in 1997, Ramone and others have developed a new acceleration scheme called Transport Synthetic Acceleration (TSA) which is much more robust and easy to implement. This scheme has been recently incorporated in 2-D and 3-D in-house codes at BARC. This report presents studies on the utility of TSA scheme for fairly general test problems involving many energy groups and anisotropic scattering. The scheme is found to be useful for problems in Cartesian as well as Cylindrical geometry. (author)
A domian Decomposition Method for Transient Neutron Transport with Pomrning-Eddington Approximation
International Nuclear Information System (INIS)
Hendi, A.A.; Abulwafa, E.E.
2008-01-01
The time-dependent neutron transport problem is approximated using the Pomraning-Eddington approximation. This approximation is two-flux approximation that expands the angular intensity in terms of the energy density and the net flux. This approximation converts the integro-differential Boltzmann equation into two first order differential equations. The A domian decomposition method that used to solve the linear or nonlinear differential equations is used to solve the resultant two differential equations to find the neutron energy density and net flux, which can be used to calculate the neutron angular intensity through the Pomraning-Eddington approximation
Directory of Open Access Journals (Sweden)
Nobuhara Fumiyoshi
2017-01-01
Full Text Available In order to evaluate the state of activation in a cyclotron facility used for the radioisotope production of PET diagnostics, we measured the neutron flux by using gold foils and TLDs. Then, the spatial distribution of neutrons and induced activity inside the cyclotron vault were simulated with the Monte Calro calculation code for neutron transport and DCHAIN-SP for activation calculation. The calculated results are in good agreement with measured values within factor 3. Therefore, the adaption of the advanced evaluation procedure for activation level is proved to be important for the planning of decommissioning of these facilities.
International Nuclear Information System (INIS)
Panini, G.C.; Siciliano, F.; Lioi, A.
1987-01-01
The main characteristics of a P 3 coupled 219-group neutron 36-group gamma-ray library in the AMPX-II Master Interface Format obtained processing ENDF/B-IV data by means of various AMPX-II System modules are presented in this note both for the more reprocessing aspects and features of the generated component files-neutrons, photon and secondary gamma-ray production cross sections. As far as the neutron data are concerned there is the avaibility of 186 data sets regarding most significant fission products. Results of the additional validation of the neutron data pertaining to eighteen benchmark experiments are also given. Some calculational tests on both neutron and coupled data emphasize the important role of the secondary gamma-ray data in nuclear criticality safety calculations
Neutron detection with plastic scintillators coupled to solid state photomultiplier detectors
Christian, James F.; Johnson, Erik B.; Fernandez, Daniel E.; Vogel, Samuel; Frank, Rebecca; Stoddard, Graham; Stapels, Christopher; Pereira, Jorge; Zegers, Remco
2017-09-01
The recent reduction of dark current in Silicon Solid-state photomultipliers (SiSSPMs) makes them an attractive alternative to conventional photomultiplier tubes (PMTs) for scintillation detection applications. Nuclear Physics experiments often require large detector volumes made using scintillation materials, which require sensitive photodetectors, such as a PMTs. PMTs add to the size, fragility, and high-voltage requirements as well as distance requirements for experiments using magnetic fields. This work compares RMD's latest detector modules, denoted as the "year 2 prototype", of plastic scintillators that discriminate gamma and high-energy particle events from neutron events using pulse shape discrimination (PSD) coupled to a SiSSPM to the following two detector modules: a similar "year 1 prototype" and a scintillator coupled to a PMT module. It characterizes the noise floor, relative signal-to-noise ratio (SNR), the timing performance, the PSD figure-of-merit (FOM) and the neutron detection efficiency of RMD's detectors. This work also evaluates the scaling of SiSSPM detector modules to accommodate the volumes needed for many Nuclear Physics experiments. The Si SSPM detector module provides a clear advantage in Nuclear Physics experiments that require the following attributes: discrimination of neutron and gamma-ray events, operation in or near strong magnetic fields, and segmentation of the detector.
The stability of tidally deformed neutron stars to three- and four-mode coupling
International Nuclear Information System (INIS)
Venumadhav, Tejaswi; Zimmerman, Aaron; Hirata, Christopher M.
2014-01-01
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.
Coupling of Groundwater Transport and Plant Uptake Models
DEFF Research Database (Denmark)
Rein, Arno; Bauer-Gottwein, Peter; Trapp, Stefan
2010-01-01
Plants significantly influence contaminant transport and fate. Important processes are uptake of soil and groundwater contaminants, as well as biodegradation in plants and their root zones. Models for the prediction of chemical uptake into plants are required for the setup of mass balances...... in environmental systems at different scale. Feedback mechanisms between plants and hydrological systems can play an important role, however having received little attention to date. Here, a new model concept for dynamic plant uptake models applying analytical matrix solutions is presented, which can be coupled...... to groundwater transport simulation tools. Exemplary simulations of plant uptake were carried out, in order to estimate concentrations in the soilplant- air system and the influence of plants on contaminant mass fluxes from soil to groundwater....
Mechanisms of acetylcholine synthesis: Coupling with choline transport
International Nuclear Information System (INIS)
Rylett, R.J.
1986-01-01
Comparative studies were performed to assess the utilization of choline transported by synaptosomal sodium-dependent, high-affinity choline carriers for the synthesis of ACh; it was determined that a significantly higher percentage of tritium-choline transported into rat forebrain synaptosomes was acetylated immediately compared to that of guinea-pig. Studies were performed to determine whether inhibition of synaptosomal ChAT was produced by incubating guinea-pig brain synaptosomes with ChMAz, comparable to that observed with rat brain synaptosomes. Very little ChAT activity was measured in guinea-pig brain; that this difference could reflect differing subcellular localizations of ChAT and different relativities with respect to coupling with choline carriers is speculative and currtly being investigated
Cellular automaton model of coupled mass transport and chemical reactions
International Nuclear Information System (INIS)
Karapiperis, T.
1994-01-01
Mass transport, coupled with chemical reactions, is modelled as a cellular automaton in which solute molecules perform a random walk on a lattice and react according to a local probabilistic rule. Assuming molecular chaos and a smooth density function, we obtain the standard reaction-transport equations in the continuum limit. The model is applied to the reactions a + b ↔c and a + b →c, where we observe interesting macroscopic effects resulting from microscopic fluctuations and spatial correlations between molecules. We also simulate autocatalytic reaction schemes displaying spontaneous formation of spatial concentration patterns. Finally, we propose and discuss the limitations of a simple model for mineral-solute interaction. (author) 5 figs., 20 refs
Spallation neutron production and the current intra-nuclear cascade and transport codes
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.
Spallation neutron production and the current intra-nuclear cascade and transport codes
International Nuclear Information System (INIS)
Filges, D.; Goldenbaum, F.
2001-01-01
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. (orig.)
Energy Technology Data Exchange (ETDEWEB)
Azmy, Yousry
2014-06-10
We employ the Integral Transport Matrix Method (ITMM) as the kernel of new parallel solution methods for the discrete ordinates approximation of the within-group neutron transport equation. The ITMM abandons the repetitive mesh sweeps of the traditional source iterations (SI) scheme in favor of constructing stored operators that account for the direct coupling factors among all the cells' fluxes and between the cells' and boundary surfaces' fluxes. The main goals of this work are to develop the algorithms that construct these operators and employ them in the solution process, determine the most suitable way to parallelize the entire procedure, and evaluate the behavior and parallel performance of the developed methods with increasing number of processes, P. The fastest observed parallel solution method, Parallel Gauss-Seidel (PGS), was used in a weak scaling comparison with the PARTISN transport code, which uses the source iteration (SI) scheme parallelized with the Koch-baker-Alcouffe (KBA) method. Compared to the state-of-the-art SI-KBA with diffusion synthetic acceleration (DSA), this new method- even without acceleration/preconditioning-is completitive for optically thick problems as P is increased to the tens of thousands range. For the most optically thick cells tested, PGS reduced execution time by an approximate factor of three for problems with more than 130 million computational cells on P = 32,768. Moreover, the SI-DSA execution times's trend rises generally more steeply with increasing P than the PGS trend. Furthermore, the PGS method outperforms SI for the periodic heterogeneous layers (PHL) configuration problems. The PGS method outperforms SI and SI-DSA on as few as P = 16 for PHL problems and reduces execution time by a factor of ten or more for all problems considered with more than 2 million computational cells on P = 4.096.
Coupled transport/hyperelastic model for nastic materials
Homison, Chris; Weiland, Lisa M.
2006-03-01
Nastic materials are high energy density active materials that mimic processes used in the plant kingdom to produce large deformations through the conversion of chemical energy. These materials utilize the controlled transport of charge and fluid across a selectively-permeable membrane to achieve bulk deformation in a process referred to in the plant kingdom as nastic movements. The nastic material being developed consists of synthetic membranes containing biological ion pumps, ion channels, and ion exchangers surrounding fluid-filled cavities embedded within a polymer matrix. In this paper the formulation of a biological transport model and its coupling with a hyperelastic finite element model of the polymer matrix is discussed. The transport model includes contributions from ion pumps, ion exchangers, and solvent flux. This work will form the basis for a feedback loop in material synthesis efforts. The goal of these studies is to determine the relative importance of the various parameters associated with both the polymer matrix and the biological transport components.
Least-squares finite element discretizations of neutron transport equations in 3 dimensions
Energy Technology Data Exchange (ETDEWEB)
Manteuffel, T.A [Univ. of Colorado, Boulder, CO (United States); Ressel, K.J. [Interdisciplinary Project Center for Supercomputing, Zurich (Switzerland); Starkes, G. [Universtaet Karlsruhe (Germany)
1996-12-31
The least-squares finite element framework to the neutron transport equation introduced in is based on the minimization of a least-squares functional applied to the properly scaled neutron transport equation. Here we report on some practical aspects of this approach for neutron transport calculations in three space dimensions. The systems of partial differential equations resulting from a P{sub 1} and P{sub 2} approximation of the angular dependence are derived. In the diffusive limit, the system is essentially a Poisson equation for zeroth moment and has a divergence structure for the set of moments of order 1. One of the key features of the least-squares approach is that it produces a posteriori error bounds. We report on the numerical results obtained for the minimum of the least-squares functional augmented by an additional boundary term using trilinear finite elements on a uniform tesselation into cubes.
Solution and study of nodal neutron transport equation applying the LTSN-DiagExp method
International Nuclear Information System (INIS)
Hauser, Eliete Biasotto; Pazos, Ruben Panta; Vilhena, Marco Tullio de; Barros, Ricardo Carvalho de
2003-01-01
In this paper we report advances about the three-dimensional nodal discrete-ordinates approximations of neutron transport equation for Cartesian geometry. We use the combined collocation method of the angular variables and nodal approach for the spatial variables. By nodal approach we mean the iterated transverse integration of the S N equations. This procedure leads to the set of one-dimensional averages angular fluxes in each spatial variable. The resulting system of equations is solved with the LTS N method, first applying the Laplace transform to the set of the nodal S N equations and then obtained the solution by symbolic computation. We include the LTS N method by diagonalization to solve the nodal neutron transport equation and then we outline the convergence of these nodal-LTS N approximations with the help of a norm associated to the quadrature formula used to approximate the integral term of the neutron transport equation. (author)
Development of coupled neutronics/thermal-hydraulics test case for HPLWR
Pham, P.; Gamtsemlidze, I. D.; Bahdanovich, R. B.; Nikonov, S. P.; Smirnov, A. D.
2017-01-01
The High-Performance Light Water Reactor (HPLWR) is the European concept of a supercritical water reactor (SCWR) which is one of the most promising and innovative designs of the Generation IV nuclear reactor concepts. The thermal-hydraulics behavior of supercritical water is significantly different from water at sub-critical pressure because of the difference in the specific heat value. Coupled analysis of HPLWR assembly neutronics and thermal-hydraulics has become important because of the strong influence of the water density on the neutron spectrum and power distribution. Programs MCU (Monte-Carlo Universal) and ATHLET (Analysis of Thermal-hydraulics of Leaks and Transients) were used for better estimation of power and temperature distribution in HPLWR assembly.
Analytical benchmarks for nuclear engineering applications. Case studies in neutron transport theory
International Nuclear Information System (INIS)
2008-01-01
The developers of computer codes involving neutron transport theory for nuclear engineering applications seldom apply analytical benchmarking strategies to ensure the quality of their programs. A major reason for this is the lack of analytical benchmarks and their documentation in the literature. The few such benchmarks that do exist are difficult to locate, as they are scattered throughout the neutron transport and radiative transfer literature. The motivation for this benchmark compendium, therefore, is to gather several analytical benchmarks appropriate for nuclear engineering applications under one cover. We consider the following three subject areas: neutron slowing down and thermalization without spatial dependence, one-dimensional neutron transport in infinite and finite media, and multidimensional neutron transport in a half-space and an infinite medium. Each benchmark is briefly described, followed by a detailed derivation of the analytical solution representation. Finally, a demonstration of the evaluation of the solution representation includes qualified numerical benchmark results. All accompanying computer codes are suitable for the PC computational environment and can serve as educational tools for courses in nuclear engineering. While this benchmark compilation does not contain all possible benchmarks, by any means, it does include some of the most prominent ones and should serve as a valuable reference. (author)
Analysis and development of spatial hp-refinement methods for solving the neutron transport equation
International Nuclear Information System (INIS)
Fournier, D.
2011-01-01
The different neutronic parameters have to be calculated with a higher accuracy in order to design the 4. generation reactor cores. As memory storage and computation time are limited, adaptive methods are a solution to solve the neutron transport equation. The neutronic flux, solution of this equation, depends on the energy, angle and space. The different variables are successively discretized. The energy with a multigroup approach, considering the different quantities to be constant on each group, the angle by a collocation method called SN approximation. Once the energy and angle variable are discretized, a system of spatially-dependent hyperbolic equations has to be solved. Discontinuous finite elements are used to make possible the development of hp-refinement methods. Thus, the accuracy of the solution can be improved by spatial refinement (h-refinement), consisting into subdividing a cell into sub-cells, or by order refinement (p-refinement), by increasing the order of the polynomial basis. In this thesis, the properties of this methods are analyzed showing the importance of the regularity of the solution to choose the type of refinement. Thus, two error estimators are used to lead the refinement process. Whereas the first one requires high regularity hypothesis (analytical solution), the second one supposes only the minimal hypothesis required for the solution to exist. The comparison of both estimators is done on benchmarks where the analytic solution is known by the method of manufactured solutions. Thus, the behaviour of the solution as a regard of the regularity can be studied. It leads to a hp-refinement method using the two estimators. Then, a comparison is done with other existing methods on simplified but also realistic benchmarks coming from nuclear cores. These adaptive methods considerably reduces the computational cost and memory footprint. To further improve these two points, an approach with energy-dependent meshes is proposed. Actually, as the
Coupling of unidimensional neutron kinetics to thermal hydraulics in parallel channels
International Nuclear Information System (INIS)
Cecenas F, M.; Campos G, R.M.
2003-01-01
In this work the dynamic behavior of a consistent system in fifteen channels in parallel that represent the reactor core of a BWR type, coupled of a kinetic neutronic model in one dimension is studied by means of time series. The arrangement of channels is obtained collapsing the assemblies that it consists the core to an arrangement of channels prepared in straight lines, and it is coupled to the unidimensional solution of the neutron diffusion equation. This solution represents the radial power distribution, and initially the static solution is obtained to verify that the one modeling core is critic. The coupled set nuclear-thermal hydraulics it is solved numerically by means of a net of CPUs working in the outline teacher-slave by means of Parallel Virtual Machine (PVM), subject to the restriction that the pressure drop is equal for each channel, which is executed iterating on the refrigerant distribution. The channels are dimensioned according to the one Stability Benchmark of the Ringhals swedish plant, organized by the Nuclear Energy Agency in 1994. From the information of this benchmark it is obtained the axial power profile for each channel, which is assumed as invariant in the time. To obtain the time series, the system gets excited with white noise (sequence that statistically obeys to a normal distribution with zero media), so that the power generated in each channel it possesses the same ones characteristics of a typical signal obtained by means of the acquisition of those signals of neutron flux in a BWR reactor. (Author)
Tsai, C. H.; Yeh, G. T.
2015-12-01
In this investigation, a coupled model of multiphase flow, reactive biogeochemical transport, thermal transport and geo-mechanics in subsurface media is presented. It iteratively solves the mass conservation equation for fluid flow, thermal transport equation for temperature, reactive biogeochemical transport equations for concentration distributions, and solid momentum equation for displacement with successive linearization algorithm. With species-based equations of state, density of a phase in the system is obtained by summing up concentrations of all species. This circumvents the problem of having to use empirical functions. Moreover, reaction rates of all species are incorporated in mass conservation equation for fluid flow. Formation enthalpy of all species is included in the law of energy conservation as a source-sink term. Finite element methods are used to discretize the governing equations. Numerical experiments are presented to examine the accuracy and robustness of the proposed model. The results demonstrate the feasibility and capability of present model in subsurface media.
International Nuclear Information System (INIS)
Mika, J.
1975-09-01
Originally the work was oriented towards two main topics: a) difference and integral methods in neutron transport theory. Two computers were used for numerical calculations GIER and CYBER-72. During the first year the main effort was shifted towards basic theoretical investigations. At the first step the ANIS code was adopted and later modified to check various finite difference approaches against each other. Then the general finite element method and the singular perturbation method were developed. The analysis of singularities of the one-dimensional neutron transport equation in spherical geometry has been done and presented. Later the same analysis for the case of cylindrical symmetry has been carried out. The second and the third year programme included the following topics: 1) finite difference methods in stationary neutron transport theory; 2)mathematical fundamentals of approximate methods for solving the transport equation; 3) singular perturbation method for the time-dependent transport equation; 4) investigation of various iterative procedures in reactor calculations. This investigation will help to better understanding of the mathematical basis for existing and developed numerical methods resulting in more effective algorithms for reactor computer codes
Energy Technology Data Exchange (ETDEWEB)
Calleros M, G.; Zapata Y, M.; Gomez H, R.A.; Mendez M, A. [Comision Federal de Electricidad, Central Nucleoelectrica de Laguna Verde, Carretera Cardel-Nautla Km. 42.5, Mpio. Alto Lucero, Veracruz (Mexico); Castlllo D, R. [ININ, Carretera Mexico-Toluca Km 36.5, La Marquesa, Estado de Mexico (Mexico)]. e-mail: gcm9acpp@cfe.gob.mx
2006-07-01
In a BWR type reactor the phenomenon of the nuclear fission is presented, in which are liberated in stochastic form neutrons, originating that the population of the same ones varies in statistic form around a mean value. This variation will cause that when the neutron flow impacts on the neutron detectors, its are had as a result neutron flow signals with fluctuations around an average value. In this article it is shown that it conforms it lapses the time, this variations in the neutron flow (and therefore, in the flow signal due only to the fission), they presented oscillations inside a stable range, which won't be divergent. Considering that the BWR is characterized because boiling phenomena are presented, which affect the moderation of the neutrons, additional variations will be had in the signal coming from the neutron detectors, with relationship to the fission itself, which will be influenced by the feedback of the moderator's reactivity and of the temperature of the fuel pellet. Also, as the BWR it has coupled control systems to maintain the coolant level one and of the thermal power of the reactor, for each control action it was affected the neutron population. This means that the reactor could end up straying of a stable state condition. By it previously described, the study of the thermohydraulic stability coupled to the neutronic is complex. In this work it is shown the phenomenology, the mathematical models and the theoretical behavior associated to the stability of the BWR type reactor; the variables that affect it are identified, the models that reproduce the behavior of the thermohydraulic stability coupled to the neutronic, the way to maintain stable the reactor and the instrumentation that can settle to detect and to suppress uncertainties is described. In particular, is make reference to the evolution of the methods to maintain the stability of the reactor and the detection system and suppression of uncertainties implemented in the
International Nuclear Information System (INIS)
Koch, K.R.
1985-01-01
A new analysis method specially suited for the inherent difficulties of fusion neutronics was developed to provide detailed studies of the fusion neutron transport physics. These studies should provide a better understanding of the limitations and accuracies of typical fusion neutronics calculations. The new analysis method is based on the direct integration of the integral form of the neutron transport equation and employs a continuous energy formulation with the exact treatment of the energy angle kinematics of the scattering process. In addition, the overall solution is analyzed in terms of uncollided, once-collided, and multi-collided solution components based on a multiple collision treatment. Furthermore, the numerical evaluations of integrals use quadrature schemes that are based on the actual dependencies exhibited in the integrands. The new DITRAN computer code was developed on the Cyber 205 vector supercomputer to implement this direct integration multiple-collision fusion neutronics analysis. Three representative fusion reactor models were devised and the solutions to these problems were studied to provide suitable choices for the numerical quadrature orders as well as the discretized solution grid and to understand the limitations of the new analysis method. As further verification and as a first step in assessing the accuracy of existing fusion-neutronics calculations, solutions obtained using the new analysis method were compared to typical multigroup discrete ordinates calculations
International Nuclear Information System (INIS)
Sanchez, J.
2010-10-01
A standard numerical procedure for the solution of singular integral equations is applied to the one-dimensional transport equation for monoenergetic neutrons. As is usual in quadrature methods, the procedure yields an Eigen system whose solution provide, for the critical slab, both the eigenvalue which is proportional to the number of secondary neutrons per collision, and the density as a function of position. The results obtained with two versions of the procedure, differing only in the extent of the basic region to which they are applied, are compared with analytically derived results available for benchmarking. The procedures considered yield consistent results for the calculated neutron densities and eigenvalues. Since the one-dimensional transport kernel and its spatial moments are integrable and their integrals can be put in terms of exponential integral functions, the resulting approximations to the neutron density yield somewhat lengthy but closed, forms. These approximate expressions of the neutron density can be used to render, after they are operated on, closed-form formulas for build-up factors, extrapolation distances or angular densities or employed for other purposes that require an analytical expression of the neutron density. As an example of this latter capability, the results of the calculation of the angular density at the surface of the slab are provided. (Author)
SAID analysis of meson photoproduction: Determination of neutron and proton EM couplings
Directory of Open Access Journals (Sweden)
Strakovsky Igor
2014-06-01
Full Text Available We present an overview of the GW SAID group effort to analyze on new pion photoproduction on both proton- and neutron-targets. The main database contribution came from the recent CLAS and MAMI unpolarized and polarized measurements. The differential cross section for the processes γn → π−p was extracted from new measurements accounting for Fermi motion effects in the impulse approximation (IA as well as NN- and πN effects beyond the IA. The electromagnetic coupling results are compared to other recent studies.
The study of neutron transport by oscillation method
International Nuclear Information System (INIS)
Raievski, V.
1959-01-01
The oscillation method is of very general use for studying the behavior of thermal neutrons in media. The main experiments are described and a general theory of them is given. This theory, which is presented in the first part, is established using the two-group approximation which has proved its efficiency in the case of thermal neutron piles. The validity of the two-group approximation is recalled. This allows definition of the meaning of the parameters used in the theory and which are measured in these experiments. The experiments carried out by this method are described, especially those performed at the Centre d'Etudes Nucleaires de Saclay where the method has been extensively used. These experiments are interpreted by means of the general theory given previously. In this way, the identity of parameters measured by this method and those given by the theory is proved. This is particularly conclusive is the case of the mean life of neutrons in a pile. (author) [fr
TMCC: a transient three-dimensional neutron transport code by the direct simulation method - 222
International Nuclear Information System (INIS)
Shen, H.; Li, Z.; Wang, K.; Yu, G.
2010-01-01
A direct simulation method (DSM) is applied to solve the transient three-dimensional neutron transport problems. DSM is based on the Monte Carlo method, and can be considered as an application of the Monte Carlo method in the specific type of problems. In this work, the transient neutronics problem is solved by simulating the dynamic behaviors of neutrons and precursors of delayed neutrons during the transient process. DSM gets rid of various approximations which are always necessary to other methods, so it is precise and flexible in the requirement of geometric configurations, material compositions and energy spectrum. In this paper, the theory of DSM is introduced first, and the numerical results obtained with the new transient analysis code, named TMCC (Transient Monte Carlo Code), are presented. (authors)
International Nuclear Information System (INIS)
Sanchez G, J.
2007-01-01
A standard procedure for the solution of singular integral equations is applied to the one-dimensional transport equation for monoenergetic neutrons. The results obtained with two versions of the procedure, differing only in the extent of the basic region to which they are applied, are compared with analytically derived results available for benchmarking. The procedures considered yield consistent results for the calculated neutron densities and eigenvalues. Several approximate expressions of the neutron density are used to render closed-form formulas for the densities which can then be analytically operated on to obtain expressions for extrapolation distances or angular densities or serve other purposes that require an analytical expression of the neutron density. (Author)
Interplay of vacuolar transporters for coupling primary and secondary active transport
Directory of Open Access Journals (Sweden)
Michèle Siek
2016-10-01
Full Text Available Secondary active transporters are driven by the proton motif force which is generated by primary active transporters such as the vacuolar proton pumps V-ATPase and V-PPase. The vacuole occupies up to 90 % of the mature cell and acidification of the vacuolar lumen is a challenging and energy-consuming task for the plant cell. Therefore, a direct coupling of primary and secondary active transporters is expected to enhance transport efficiency and to reduce energy consumption by transport processes across the tonoplast. This has been addressed by analyzing physical and functional interactions between the V-ATPase and a selection of vacuolar transporters including the primary active proton pump AVP1, the calcium ion/proton exchanger CAX1, the potassium ion/proton symporter KUP5, the sodium ion/proton exchanger NHX1, and the anion/proton exchanger CLC-c. Physical interaction was demonstrated in vivo for the V-ATPase and the secondary active transporters CAX1 and CLC-c, which are responsible for calcium- and anion-accumulation in the vacuole, respectively. Measurements of V-ATPase activity and vacuolar pH revealed a functional interaction of V-ATPase and CAX1, CLC-c that is likely caused by the observed physical interaction. The complex of the V-ATPase further interacts with the nitrate reductase 2, and as a result, nitrate assimilation is directly linked to the energization of vacuolar nitrate accumulation by secondary active anion/proton exchangers.
Coupled Model of channels in parallel and neutron kinetics in two dimensions
International Nuclear Information System (INIS)
Cecenas F, M.; Campos G, R.M.; Valle G, E. del
2004-01-01
In this work an arrangement of thermohydraulic channels is presented that represent those four quadrants of a nucleus of reactor type BWR. The channels are coupled to a model of neutronic in two dimensions that allow to generate the radial profile of power of the reactor. Nevertheless that the neutronic pattern is of two dimensions, it is supplemented with axial additional information when considering the axial profiles of power for each thermo hydraulic channel. The stationary state is obtained the one it imposes as frontier condition the same pressure drop for all the channels. This condition is satisfied to iterating on the flow of coolant in each channel to equal the pressure drop in all the channels. This stationary state is perturbed later on when modifying the values for the effective sections corresponding to an it assembles. The calculation in parallel of the neutronic and the thermo hydraulic is carried out with Vpm (Virtual parallel machine) by means of an outline teacher-slave in a local net of computers. (Author)
Coupled neutronic core and subchannel analysis of nanofluids in VVER-1000 type reactor
Energy Technology Data Exchange (ETDEWEB)
Zarifi, Ehsan; Sepanloo, Kamran [Nuclear Science and Technology Research Institute (NSTRI), Tehran (Iran, Islamic Republic of). Reactor and Nuclear Safety School; Jahanfarnia, Golamreza [Islamic Azad Univ., Tehran (Iran, Islamic Republic of). Dept. of Nuclear Engineering, Science and Research Branch
2017-05-15
This study is aimed to perform the coupled thermal-hydraulic/neutronic analysis of nanofluids as the coolant in the hot fuel assembly of VVER-1000 reactor core. Water-based nanofluid containing various volume fractions of Al{sub 2}O{sub 3} nanoparticle is analyzed. WIMS and CITATION codes are used for neutronic simulation of the reactor core, calculating neutron flux and thermal power distribution. In the thermal-hydraulic modeling, the porous media approach is used to analyze the thermal behavior of the reactor core and the subchannel analysis is used to calculate the hottest fuel assembly thermal-hydraulic parameters. The derived conservation equations for coolant and conduction heat transfer equation for fuel and clad are discretized by Finite volume method and solved numerically using visual FORTRAN program. Finally the analysis results for nanofluids and pure water are compared together. The achieved results show that at low concentration (0.1 percent volume fraction) alumina is the optimum nanoparticles for normal reactor operation.
Doliwa, B.; Arévalo, E.; Weiland, T.
2007-10-01
The study of beam dynamics and the localization of potential sources of instabilities are important tasks in the design of modern, high-intensity particle accelerators. In the case of synchrotrons and storage rings, coupling impedance data are needed to characterize the parasitic interaction of critical components with the beam. In this article we demonstrate the application of numerical field simulations to the computation of transverse kicker coupling impedances. Based on the 3D simulation results, a parametrized model is developed to incorporate the impedance of an arbitrary pulse-forming network attached to the kicker. Detailed comparisons of numerical results with twin-wire and direct measurements are discussed at the example of the Spallation Neutron Source extraction kicker.
Directory of Open Access Journals (Sweden)
B. Doliwa
2007-10-01
Full Text Available The study of beam dynamics and the localization of potential sources of instabilities are important tasks in the design of modern, high-intensity particle accelerators. In the case of synchrotrons and storage rings, coupling impedance data are needed to characterize the parasitic interaction of critical components with the beam. In this article we demonstrate the application of numerical field simulations to the computation of transverse kicker coupling impedances. Based on the 3D simulation results, a parametrized model is developed to incorporate the impedance of an arbitrary pulse-forming network attached to the kicker. Detailed comparisons of numerical results with twin-wire and direct measurements are discussed at the example of the Spallation Neutron Source extraction kicker.
Optimization of spring exchange coupled ferrites, studied by in situ neutron diffraction
DEFF Research Database (Denmark)
Ahlburg, Jakob; Christensen, Mogens; Granados-Miralles, Cecilia
the magnetic energy product. For the exchange coupling to happen it is crucial to have the right ratio between the hard and the soft phase but also to control the size of the particles since exchange coupling is a very small range effect. In this study, nanoparticles of the spinel CoFe2O4 (hard magnet......) is reduced to a metallic alloy CoFe (soft magnet) by heating the sample and flowing it with hydrogen gas. It is studied in situ using neutron powder diffraction with a time resolution of 12 min. The transition from spinel to pure metal goes through an intermediate step of a metal oxide before being fully......Strong permanent magnets with a high energy-product are vital for a great number of electronic devices, these can be found in transformers, loudspeakers, windmills etc. Normally the preferred type of magnets are Rare Earth Metals (REM) containing magnets. REM excels since the magnetic contribution...
Pramanick, Abhijit; Wang, Xun-Li
2013-01-01
Ferromagnetic shape memory alloys (FSMA) are a new class of smart materials with unique properties and applications. The shape memory properties of FSMAs are due to a strong coupling between their elastic and magnetic properties. Understanding the physical origins of magnetoelastic coupling is therefore critical for future materials development and device design. To this end, a thorough study is required of the atomistic and microscopic structural arrangements and how these structural arrangements are related to the orientation of the magnetic moments. The different characterization techniques developed recently to probe these features are reviewed. A special emphasis is placed on in situ techniques such as neutron diffraction through which the microscopic origins of the collective material behavior could be ascertained.
International Nuclear Information System (INIS)
Moraes, Pedro Gabriel B.; Leite, Michel C.A.; Barros, Ricardo C.
2013-01-01
In this work we developed a software to model and generate results in tables and graphs of one-dimensional neutron transport problems in multi-group formulation of energy. The numerical method we use to solve the problem of neutron diffusion is analytic, thus eliminating the truncation errors that appear in classical numerical methods, e.g., the method of finite differences. This numerical analytical method increases the computational efficiency, since they are not refined spatial discretization necessary because for any spatial discretization grids used, the numerical result generated for the same point of the domain remains unchanged unless the rounding errors of computational finite arithmetic. We chose to develop a computational application in MatLab platform for numerical computation and program interface is simple and easy with knobs. We consider important to model this neutron transport problem with a fixed source in the context of shielding calculations of radiation that protects the biosphere, and could be sensitive to ionizing radiation
The importance of anisotropic scattering in high energy neutron transport problems
International Nuclear Information System (INIS)
Prillinger, G.; Mattes, M.
1984-01-01
To describe the highly anisotropic scattering of very fast neutrons adequately the transport code ANISN has been improved. Fokker-Planck terms have been introduced into the transport equation which accurately describe the small changes in energy and angle. The new code has been tested for a d(50)-Be neutron source in a deep penetration iron problem. The influence of the forward peaked elastic scattering on the fast neutron spectrum is shown to be significant and can be handled efficiently in the new ANISN version. Since common cross-section libraries are limited by Legendre expansion, or by their upper energy boundary, or exclude elastic scattering above 20 MeV a special library has been created. (Auth.)
Energy Technology Data Exchange (ETDEWEB)
Raievski, V
1959-07-01
The oscillation method is of very general use for studying the behavior of thermal neutrons in media. The main experiments are described and a general theory of them is given. This theory, which is presented in the first part, is established using the two-group approximation which has proved its efficiency in the case of thermal neutron piles. The validity of the two-group approximation is recalled. This allows definition of the meaning of the parameters used in the theory and which are measured in these experiments. The experiments carried out by this method are described, especially those performed at the Centre d'Etudes Nucleaires de Saclay where the method has been extensively used. These experiments are interpreted by means of the general theory given previously. In this way, the identity of parameters measured by this method and those given by the theory is proved. This is particularly conclusive is the case of the mean life of neutrons in a pile. (author) [French] La methode de modulation est un procede tres general d'etude des proprietes neutroniques des milieux contenant des neutrons thermiques. Le present rapport a pour but de decrire les principales de ces experiences et d'en donner une theorie generale. Cette theorie, exposee dans la premiere partie, est etablie dons le cadre de l'approximation a deux groupes de vitesse qui a prouve son efficacite dons le cas des piles a neutrons thermiques. Le domaine de validite de l'approximation a deux groupes est rappele au debut, ce qui permet de definir avec precision la signification des parametres qui entrent dons la theorie et qui font l'objet de ces mesures. La deuxieme partie decrit les experiences realisees, en particulier celles effectuees au Centre d'Etudes Nucleaires de Saclay ou la methode a ete considerablement developpee. Ces experiences sont interpretees dans le cadre de la theorie generale exposee precedemment. On prouve ainsi l'identite des parametres mesures
International Nuclear Information System (INIS)
Iranzo, Alfredo; Boillat, Pierre; Oberholzer, Pierre; Guerra, José
2014-01-01
A novel modelling framework for the simulation of the diffusive mass transport limitations occurring at GDL local scale of PEFCs is presented, in particular in relation with the distribution of liquid water in the porous media. The distinctive characteristic of this framework is the fact that the distribution of liquid water is not predicted by the model but it is instead mapped into the simulation model from available experimental measurements, obtained with neutron imaging. The presence of liquid water is thus included in the model as a modifier for the gas diffusion transport, and not directly calculated by the model. This allows for a coupling of experimental measurements and model development that is expected to allow a further progress of highly reliable models for the understanding of local fuel cell phenomena. A 1D cell is analyzed, and the effective diffusion coefficient and the n exponent in the diffusion correction factor is calculated from the results of the combination of modelling and experimental data. An extension of the method for a 2D cell is also introduced. - Highlights: • Novel modelling framework for the research of GDL diffusive mass transport. • Liquid water distribution is mapped into the model from neutron imaging data. • The approach provides a stronger coupling of experimental measurements and modelling. • Analysis of 1D and 2D cells presented as introduction to the development approach
Kyutoku, Koutarou; Kiuchi, Kenta; Sekiguchi, Yuichiro; Shibata, Masaru; Taniguchi, Keisuke
2018-01-01
We study the merger of black hole-neutron star binaries by fully general-relativistic neutrino-radiation-hydrodynamics simulations throughout the coalescence, particularly focusing on the role of neutrino irradiation in dynamical mass ejection. Neutrino transport is incorporated by an approximate transfer scheme based on the truncated moment formalism. While we fix the mass ratio of the black hole to the neutron star to be 4 and the dimensionless spin parameter of the black hole to be 0.75, the equations of state for finite-temperature neutron-star matter are varied. The hot accretion disk formed after tidal disruption of the neutron star emits a copious amount of neutrinos with the peak total luminosity ˜1 - 3 ×1053 erg s-1 via thermal pair production and subsequent electron/positron captures on free nucleons. Nevertheless, the neutrino irradiation does not modify significantly the electron fraction of the dynamical ejecta from the neutrinoless β -equilibrium value at zero temperature of initial neutron stars. The mass of the wind component driven by neutrinos from the remnant disk is negligible compared to the very neutron-rich dynamical component, throughout our simulations performed until a few tens milliseconds after the onset of merger, for the models considered in this study. These facts suggest that the ejecta from black hole-neutron star binaries are very neutron rich and are expected to accommodate strong r -process nucleosynthesis, unless magnetic or viscous processes contribute substantially to the mass ejection from the disk. We also find that the peak neutrino luminosity does not necessarily increase as the disk mass increases, because tidal disruption of a compact neutron star can result in a remnant disk with a small mass but high temperature.
Energy Technology Data Exchange (ETDEWEB)
Dumazert, Jonathan; Coulon, Romain; Carrel, Frédérick; Corre, Gwenolé; Normand, Stéphane [CEA, LIST, Laboratoire Capteurs Architectures Electroniques, 91191 Gif-sur-Yvette (France); Méchin, Laurence [CNRS, UCBN, Groupe de Recherche en Informatique, Image, Automatique et Instrumentation de Caen, 14050 Caen (France); Hamel, Matthieu [CEA, LIST, Laboratoire Capteurs Architectures Electroniques, 91191 Gif-sur-Yvette (France)
2016-08-21
Neutron detection forms a critical branch of nuclear-related issues, currently driven by the search for competitive alternative technologies to neutron counters based on the helium-3 isotope. The deployment of plastic scintillators shows a high potential for efficient detectors, safer and more reliable than liquids, more easily scalable and cost-effective than inorganic. In the meantime, natural gadolinium, through its 155 and mostly 157 isotopes, presents an exceptionally high interaction probability with thermal neutrons. This paper introduces a dual system including a metal gadolinium core inserted at the center of a high-scale plastic scintillator sphere. Incident fast neutrons are thermalized by the scintillator shell and then may be captured with a significant probability by gadolinium 155 and 157 nuclei in the core. The deposition of a sufficient fraction of the capture high-energy prompt gamma signature inside the scintillator shell will then allow discrimination from background radiations by energy threshold, and therefore neutron detection. The scaling of the system with the Monte Carlo MCNPX2.7 code was carried out according to a tradeoff between the moderation of incident fast neutrons and the probability of slow neutron capture by a moderate-cost metal gadolinium core. Based on the parameters extracted from simulation, a first laboratory prototype for the assessment of the detection method principle has been synthetized. The robustness and sensitivity of the neutron detection principle are then assessed by counting measurement experiments. Experimental results confirm the potential for a stable, highly sensitive, transportable and cost-efficient neutron detector and orientate future investigation toward promising axes.
Energy Technology Data Exchange (ETDEWEB)
T' Joen, C., E-mail: c.g.a.tjoen@tudelft.nl [Delft University of Technology, Department Radiation, Radionuclides and Reactors, Mekelweg 15, 2629 JB Delft (Netherlands); Ghent University, Department of Flow, Heat and Combustion Mechanics, Sint-Pietersnieuwstraat 41, 9000 Gent (Belgium); Rohde, M. [Delft University of Technology, Department Radiation, Radionuclides and Reactors, Mekelweg 15, 2629 JB Delft (Netherlands)
2012-01-15
Highlights: Black-Right-Pointing-Pointer No pure thermo-hydraulic instabilities were recorded. Black-Right-Pointing-Pointer A large unstable zone was found for the coupled thermo-hydraulic-neutronic mode. Black-Right-Pointing-Pointer The instabilities are similar to the type I instabilities of boiling systems. Black-Right-Pointing-Pointer The low power stability threshold crosses the equivalent reference line h{sub out} = h{sub pc}. - Abstract: The HPLWR (high performance light water reactor) is the European concept design for a SCWR (supercritical water reactor). This unique reactor design consists of a three pass core with intermediate mixing plena. As the supercritical water passes through the core, it experiences a significant density reduction. This large change in density could be used as the driving force for natural circulation of the coolant, adding an inherent safety feature to this concept design. The idea of natural circulation has been explored in the past for boiling water reactors (BWR). From those studies, it is known that the different feedback mechanisms can trigger flow instabilities. These can be purely thermo-hydraulic (driven by the friction - mass flow rate or gravity - mass flow rate feedback of the system), or they can be coupled thermo-hydraulic-neutronic (driven by the coupling between friction, mass flow rate and power production). The goal of this study is to explore the stability of a natural circulation HPLWR considering the thermo-hydraulic-neutronic feedback. This was done through a unique experimental facility, DeLight, which is a scaled model of the HPLWR using Freon R23 as a scaling fluid. An artificial neutronic feedback was incorporated into the system based on the average measured density. To model the heat transfer dynamics in the rods, a simple first order model was used with a fixed time constant of 6 s. The results include the measurements of the varying decay ratio (DR) and frequency over a wide range of operating
DIAMANT2 - A multigroup neutron transport program for triangular and hexagonal geometry
International Nuclear Information System (INIS)
Kuefner, K.; Heger, R.
1980-09-01
DIAMANT2 evolved out of the DIAMANT-code. DIAMANT2 solves the multigroup neutron transport equation in planar geometry using the Ssub(N) method. Spatial discretization is accomplished by taking finite differences on a meshgrid composed of equilateral triangles. This report contains a detailed documentation of the program and the input description. (orig./HJ) [de
The neutron transport code DTF-Traca users manual and input data
International Nuclear Information System (INIS)
Ahnert, C.
1979-01-01
This is a users manual of the neutron transport code DTF-TRACA, which is a version of the original DTF-IV with some modifications made at JEN. A detailed input data descriptions is given. The new options developed at JEN are included too. (Author) 18 refs
The neutron transport code DTF-Traca users manual and input data
Energy Technology Data Exchange (ETDEWEB)
Ahnert, C.
1979-07-01
This is a users manual of the neutron transport code DTF-TRACA, which is a version of the original DTF-IV with some modifications made at JEN. A detailed input data descriptions is given. The new options developed at JEN are included too. (Author) 18 refs.
International Nuclear Information System (INIS)
Trukhanov, G.Ya.
2005-01-01
Time-dependent neutron transport theory of G.Ya. Trukhanov and S.A. Podosenov is developed. Errors of calculating of power series expansion coefficients, γ k , in this theory were estimated. It has been found that power series convergence radius R=|χ 1,2 |= 0.9595. Power series convergence speed were estimated [ru
Two-group neutron transport theory in adjacent space with lineary anisotropic scattering
International Nuclear Information System (INIS)
Maiorino, J.R.
1978-01-01
A solution method for two-group neutron transport theory with anisotropic scattering is introduced by the combination of case method (expansion method of self singular function) and the invariant imbedding (invariance principle). The numerical results for the Milne problem in light water and borated water is presented to demonstrate the avalibility of the method [pt
In situ neutron depth profiling: A powerful method to probe lithium transport in micro-batteries
Oudenhoven, J.F.M.; Labohm, F.; Mulder, M.; Niessen, R.A.H.; Mulder, F.M.; Notten, P.H.L.
2011-01-01
In situ neutron depth profiling (NDP) offers the possibility to observe lithium transport inside micro-batteries during battery operation. It is demonstrated that NDP results are consistent with the results of electrochemical measurements, and that the use of an enriched6LiCoO2 cathode offers more
Ligand-induced dynamical change of G-protein-coupled receptor revealed by neutron scattering
Shrestha, Utsab R.; Bhowmik, Debsindhu; Mamontov, Eugene; Chu, Xiang-Qiang
Light activation of the visual G-protein-coupled receptor rhodopsin leads to the significant change in protein conformation and structural fluctuations, which further activates the cognate G-protein (transducin) and initiates the biological signaling. In this work, we studied the rhodopsin activation dynamics using state-of-the-art neutron scattering technique. Our quasi-elastic neutron scattering (QENS) results revealed a broadly distributed relaxation rate of the hydrogen atom in rhodopsin on the picosecond to nanosecond timescale (beta-relaxation region), which is crucial for the protein function. Furthermore, the application of mode-coupling theory to the QENS analysis uncovers the subtle changes in rhodopsin dynamics due to the retinal cofactor. Comparing the dynamics of the ligand-free apoprotein, opsin versus the dark-state rhodopsin, removal of the retinal cofactor increases the relaxation time in the beta-relaxation region, which is due to the possible open conformation. Moreover, we utilized the concept of free-energy landscape to explain our results for the dark-state rhodopsin and opsin dynamics, which can be further applied to other GPCR systems to interpret various dynamic behaviors in ligand-bound and ligand-free protein.
MC++: A parallel, portable, Monte Carlo neutron transport code in C++
International Nuclear Information System (INIS)
Lee, S.R.; Cummings, J.C.; Nolen, S.D.
1997-01-01
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
Quantum Thermodynamics in Strong Coupling: Heat Transport and Refrigeration
Directory of Open Access Journals (Sweden)
Gil Katz
2016-05-01
Full Text Available The performance characteristics of a heat rectifier and a heat pump are studied in a non-Markovian framework. The device is constructed from a molecule connected to a hot and cold reservoir. The heat baths are modelled using the stochastic surrogate Hamiltonian method. The molecule is modelled by an asymmetric double-well potential. Each well is semi-locally connected to a heat bath composed of spins. The dynamics are driven by a combined system–bath Hamiltonian. The temperature of the baths is regulated by a secondary spin bath composed of identical spins in thermal equilibrium. A random swap operation exchange spins between the primary and secondary baths. The combined system is studied in various system–bath coupling strengths. In all cases, the average heat current always flows from the hot towards the cold bath in accordance with the second law of thermodynamics. The asymmetry of the double well generates a rectifying effect, meaning that when the left and right baths are exchanged the heat current follows the hot-to-cold direction. The heat current is larger when the high frequency is coupled to the hot bath. Adding an external driving field can reverse the transport direction. Such a refrigeration effect is modelled by a periodic driving field in resonance with the frequency difference of the two potential wells. A minimal driving amplitude is required to overcome the heat leak effect. In the strong driving regime the cooling power is non-monotonic with the system–bath coupling.
Coupled Modeling of Rhizosphere and Reactive Transport Processes
Roque-Malo, S.; Kumar, P.
2017-12-01
The rhizosphere, as a bio-diverse plant root-soil interface, hosts many hydrologic and biochemical processes, including nutrient cycling, hydraulic redistribution, and soil carbon dynamics among others. The biogeochemical function of root networks, including the facilitation of nutrient cycling through absorption and rhizodeposition, interaction with micro-organisms and fungi, contribution to biomass, etc., plays an important role in myriad Critical Zone processes. Despite this knowledge, the role of the rhizosphere on watershed-scale ecohydrologic functions in the Critical Zone has not been fully characterized, and specifically, the extensive capabilities of reactive transport models (RTMs) have not been applied to these hydrobiogeochemical dynamics. This study uniquely links rhizospheric processes with reactive transport modeling to couple soil biogeochemistry, biological processes, hydrologic flow, hydraulic redistribution, and vegetation dynamics. Key factors in the novel modeling approach are: (i) bi-directional effects of root-soil interaction, such as simultaneous root exudation and nutrient absorption; (ii) multi-state biomass fractions in soil (i.e. living, dormant, and dead biological and root materials); (iii) expression of three-dimensional fluxes to represent both vertical and lateral interconnected flows and processes; and (iv) the potential to include the influence of non-stationary external forcing and climatic factors. We anticipate that the resulting model will demonstrate the extensive effects of plant root dynamics on ecohydrologic functions at the watershed scale and will ultimately contribute to a better characterization of efflux from both agricultural and natural systems.
International Nuclear Information System (INIS)
Ganapol, B.D.; Kornreich, D.E.
1997-01-01
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
Monte Carlo study in the mechanisms of transport of fast neutrons in various media
International Nuclear Information System (INIS)
Ku, L.
1976-01-01
The life histories of fast neutrons created by the straight Monte Carlo method in various attenuation media were examined. The media studied range from the one with simple, featureless properties (Na) to iron with very complicated cross section structure. The life histories of exceptional neutrons, i.e. those staying very close to the source, or those going very far from the source, were compared with those of the general population. When the exceptional neutrons exploited a particular collision property in a narrow energy band in order to reach a given detector, the method of analyzing Monte Carlo histories was able to provide a clear physical picture and single out the influence of that property on the macroscopic behavior of the neutrons. Two such phenomena were demonstrated by using this technique. In one, transport in a cross section minimum dominates the deep penetration of the neutrons. In such a circumstance most of the spatial transport is accomplished by the traveling at energies in and near the minimum, while little transport occurs at any other energies. The second example involves the effect of inelastic scattering on the low-energy leakage spectra for small bare assemblies. It is shown that, for a small bare iron sphere and for a fission source, the exit current spectrum below 100 keV is extremely sensitive to the details of the inelastic scattering near threshold. It often happened that in some exceptional situations the number of histories available for the analysis was too few to give statistically significant results. The most important conclusion to be drawn here is that the analysis of Monte Carlo histories can provide information on the details of transport mechanisms that is not available through forward or even adjoint deterministic transport calculations. 47 figures, 21 tables
Energy Technology Data Exchange (ETDEWEB)
Ganapol, B.D.; Kornreich, D.E. [Univ. of Arizona, Tucson, AZ (United States). Dept. of Nuclear Engineering
1997-07-01
Because of the requirement of accountability and quality control in the scientific world, a demand for high-quality analytical benchmark calculations has arisen in the neutron transport community. The intent of these benchmarks is to provide a numerical standard to which production neutron transport codes may be compared in order to verify proper operation. The overall investigation as modified in the second year renewal application includes the following three primary tasks. Task 1 on two dimensional neutron transport is divided into (a) single medium searchlight problem (SLP) and (b) two-adjacent half-space SLP. Task 2 on three-dimensional neutron transport covers (a) point source in arbitrary geometry, (b) single medium SLP, and (c) two-adjacent half-space SLP. Task 3 on code verification, includes deterministic and probabilistic codes. The primary aim of the proposed investigation was to provide a suite of comprehensive two- and three-dimensional analytical benchmarks for neutron transport theory applications. This objective has been achieved. The suite of benchmarks in infinite media and the three-dimensional SLP are a relatively comprehensive set of one-group benchmarks for isotropically scattering media. Because of time and resource limitations, the extensions of the benchmarks to include multi-group and anisotropic scattering are not included here. Presently, however, enormous advances in the solution for the planar Green`s function in an anisotropically scattering medium have been made and will eventually be implemented in the two- and three-dimensional solutions considered under this grant. Of particular note in this work are the numerical results for the three-dimensional SLP, which have never before been presented. The results presented were made possible only because of the tremendous advances in computing power that have occurred during the past decade.
International Nuclear Information System (INIS)
Pirouzmand, Ahmad; Hadad, Kamal
2011-01-01
Highlights: → This paper describes the solution of time-independent neutron transport equation. → Using a novel method based on cellular neural networks (CNNs) coupled with P N method. → Utilize the CNN model to simulate spatial scalar flux distribution in steady state. → The accuracy, stability, and capabilities of CNN model are examined in x-y geometry. - Abstract: This paper describes a novel method based on using cellular neural networks (CNN) coupled with spherical harmonics method (P N ) to solve the time-independent neutron transport equation in x-y geometry. To achieve this, an equivalent electrical circuit based on second-order form of neutron transport equation and relevant boundary conditions is obtained using CNN method. We use the CNN model to simulate spatial response of scalar flux distribution in the steady state condition for different order of spherical harmonics approximations. The accuracy, stability, and capabilities of CNN model are examined in 2D Cartesian geometry for fixed source and criticality problems.
Numerical solution of neutron transport equations in discrete ordinates and slab geometry
International Nuclear Information System (INIS)
Serrano Pedraza, F.
1985-01-01
An unified formalism to solve numerically, between other equation, the neutron transport in discrete ordinates, slab geometry, several energy groups and independents of time, has been developed recently. Such a formalism cover some of the conventional schemes as diamond difference, (WDD) characteristic step (SC) lineal characteristic (LC), quadratic characteristic (QC) and lineal discontinuous. Unified formation gives before hand the convergence order of the previously selected scheme. In fact it allows besides to generate a big amount of numerical schemes, with which is also possible to solve numerical equations as soon as neutron transport. The essential purpose of this work was to solve the neutron transport equations in slab geometry and discrete ordinates considering several energy groups without to take under advisement time dependence based in the above mentioned unified formalism. To reach this purpose it was necesary to design a computer code with the name TNOD1 (Neutron transport in discrete ordinates and 1 dimension) which includes each one of the schemes already pointed out. there exist two numerical schemes, also recently developed, quadratic continuous (QC) and cubic continuous (CN), although covered by unified formalism, it has been possible to include them inside this computer code without make substantial changes in its structure. In chapter I, derivative of neutron transport equation independent of time is taken, for angular flux, including boundary conditions and discontinuity. In chapter II the neutron transport equations are obtained in multigroups, independents of time, for approximation of discrete ordinates. Description of theory related with unified formalism and its relationship with mentioned discretization schemes is presented in chapter III. Chapter IV describes the computer code developed and finally, in chapter V different numerical results obtained with TNOD1 program are shown. In Appendix A theorems and mathematical arguments used
Casinini, F.; Petrillo, C.; Sacchetti, F.
2012-05-01
In the next years the slow neutron scattering community is waiting for a continuous improvement of the neutron detectors because of the development of the new and more intense neutron sources and to obtain a better performance of the neutron instrumentation to face the higher demands and new capabilities necessary for the novel experiments. In particular detectors having a faster response and a better shape of the time response must be produced, while new and more flexible acquisition systems must be introduced in order to collect in the proper way the information carried by the scattered neutrons. At present inside the neutron detector community the lack for detectors having a spatial resolution below 1 mm is evident. In the past it has been already demonstrated that a silicon microstrip detector coupled to a Gadolinium foil, used as neutron converter, provides a good performance neutron detector. In the present paper we present a 128 channel detector which has been designed for operation in the thermal neutron region with 0.55 mm spatial resolution, 100 ns time resolution and 25 ns time stamp accuracy. We present a new approach for the acquisition of the neutron arrival time, based on a single event storage by manipulating the detector digital output using a programmable acquisition system which takes advantage from high performance industrial standard hardware employing a FPGA and a real-time on board processor. We suggest the use of the single neutron event storing to make the time to energy transformation more efficient in the case of time of flight inelastic scattering, where the conversion from angle and time to momentum and energy is necessary.
Kirstein, O; Prager, M; Schneider, G J
2009-06-07
Methyl group rotations in methyl fluoride were studied using the high flux backscattering spectrometer SPHERES at FRM-II. The asymmetry and width of the low temperature tunneling peak was used to determine if coupled rotations between neighboring methyl fluoride molecules exist. The temperature dependent broadening of the tunneling peak was used to determine the first librational transition and compared to the temperature dependent shift of the position of the tunneling peak. The results obtained by using inelastic neutron scattering confirm previous models that assume rotational coupling. This is the first neutron backscattering experiment with sub-microeV resolution at energy transfers up to 31 microeV.
Hybrid variational principles and synthesis method for finite element neutron transport calculations
International Nuclear Information System (INIS)
Ackroyd, R.T.; Nanneh, M.M.
1990-01-01
A family of hybrid variational principles is derived using a generalised least squares method. Neutron conservation is automatically satisfied for the hybrid principles employing two trial functions. No interfaces or reflection conditions need to be imposed on the independent even-parity trial function. For some hybrid principles a single trial function can be employed by relating one parity trial function to the other, using one of the parity transport equation in relaxed form. For other hybrid principles the trial functions can be employed sequentially. Synthesis of transport solutions, starting with the diffusion theory approximation, has been used as a way of reducing the scale of the computation that arises with established finite element methods for neutron transport. (author)
OECD/NEA benchmark for time-dependent neutron transport calculations without spatial homogenization
Energy Technology Data Exchange (ETDEWEB)
Hou, Jason, E-mail: jason.hou@ncsu.edu [Department of Nuclear Engineering, North Carolina State University, Raleigh, NC 27695 (United States); Ivanov, Kostadin N. [Department of Nuclear Engineering, North Carolina State University, Raleigh, NC 27695 (United States); Boyarinov, Victor F.; Fomichenko, Peter A. [National Research Centre “Kurchatov Institute”, Kurchatov Sq. 1, Moscow (Russian Federation)
2017-06-15
Highlights: • A time-dependent homogenization-free neutron transport benchmark was created. • The first phase, known as the kinetics phase, was described in this work. • Preliminary results for selected 2-D transient exercises were presented. - Abstract: A Nuclear Energy Agency (NEA), Organization for Economic Co-operation and Development (OECD) benchmark for the time-dependent neutron transport calculations without spatial homogenization has been established in order to facilitate the development and assessment of numerical methods for solving the space-time neutron kinetics equations. The benchmark has been named the OECD/NEA C5G7-TD benchmark, and later extended with three consecutive phases each corresponding to one modelling stage of the multi-physics transient analysis of the nuclear reactor core. This paper provides a detailed introduction of the benchmark specification of Phase I, known as the “kinetics phase”, including the geometry description, supporting neutron transport data, transient scenarios in both two-dimensional (2-D) and three-dimensional (3-D) configurations, as well as the expected output parameters from the participants. Also presented are the preliminary results for the initial state 2-D core and selected transient exercises that have been obtained using the Monte Carlo method and the Surface Harmonic Method (SHM), respectively.
International Nuclear Information System (INIS)
Hartmann, C.; Sanchez, V.; Tietsch, W.; Stieglitz, R.
2012-01-01
The KIT is involved in the development and qualification of best estimate methodologies for BWR transient analysis in cooperation with industrial partners. The goal is to establish the most advanced thermal hydraulic system codes coupled with 3D reactor dynamic codes to be able to perform a more realistic evaluation of the BWR behavior under accidental conditions. For this purpose a computational chain based on the lattice code (SCALE6/GenPMAXS), the coupled neutronic/thermal hydraulic code (TRACE/PARCS) as well as a Monte Carlo based uncertainty and sensitivity package (SUSA) has been established and applied to different kind of transients of a Boiling Water Reactor (BWR). This paper will describe the multidimensional models of the plant elaborated for TRACE and PARCS to perform the investigations mentioned before. For the uncertainty quantification of the coupled code TRACE/PARCS and specifically to take into account the influence of the kinetics parameters in such studies, the PARCS code has been extended to facilitate the change of model parameters in such a way that the SUSA package can be used in connection with TRACE/PARCS for the U and S studies. This approach will be presented in detail. The results obtained for a rod drop transient with TRACE/PARCS using the SUSA-methodology showed clearly the importance of some kinetic parameters on the transient progression demonstrating that the coupling of a best-estimate coupled codes with uncertainty and sensitivity tools is very promising and of great importance for the safety assessment of nuclear reactors. (authors)
Coupled fully 3D neutron kinetics thermal-hydraulic computations for DNB analysis on PWRs
International Nuclear Information System (INIS)
Pitot, Samuel; Alborghetti, Nicolas
2007-01-01
Departure from Nucleate Boiling (DNB) is one of the major limiting factors of Pressurized Water Reactors (PWRs). Safety requires that occurrence of DNB should be precluded under normal or incidental operating conditions. To perform Main Steam Line Break (MSLB) accident calculations EDF have developed its own numerical tool OSCARD based on: the thermal-hydraulic THYC code for DNB analysis, the neutron kinetics COCCINELLE code for power distribution computations, the thermal-hydraulic CATHARE code to provide boundary conditions analysis with system scale computation. With OSCARD a fully three-dimensional (3D) representation of the core is proposed in conjunction with a two-phase flow porous-body approach (THYC) and two-group diffusion equations in the axial and lateral directions with Doppler and void reactivity feedback effects (COCCINELLE). OSCARD provides EDF with an alternative and independent way of evaluating fuel performance and safety margins. In the licensed approach, the coupled 3D neutron kinetics and thermal-hydraulic part of OSCARD steady computations is used to produce 3D power distribution in the reactor core at the most penalizing moment of the transient. Then this distribution is used as an input for THYC to perform thermal-hydraulic subchannel analysis. This 3 steps approach is used with simple conservative and bounding analysis assumptions, that can not occur in reality. In a prospective approach, OSCARD enables to combine thermal-hydraulic subchannel analysis with the neutron kinetics radial average channel model using a nodalization of one quarter of fuel assembly in order to perform one step DNB analysis. (author)
Dosimetric analysis of BNCT - Boron Neutron Capture Therapy - coupled to 252Cf brachytherapy
International Nuclear Information System (INIS)
Brandao, Samia F.; Campos, Tarcisio P.R.
2009-01-01
The incidence of brain tumors is increasing in world population; however, the treatments employed in this type of tumor have a high rate of failure and in some cases have been considered palliative, depending on histology and staging of tumor. Its necessary to achieve the control tumor dose without the spread irradiation cause damage in the brain, affecting patient neurological function. Stereotactic radiosurgery is a technique that achieves this; nevertheless, other techniques that can be used on the brain tumor control must be developed, in order to guarantee lower dose on health surroundings tissues other techniques must be developing. The 252 Cf brachytherapy applied to brain tumors has already been suggested, showing promising results in comparison to photon source, since the active source is placed into the tumor, providing greater dose deposition, while more distant regions are spared. BNCT - Boron Neutron Capture Therapy - is another technique that is in developing to brain tumors control, showing theoretical superiority on the rules of conventional treatments, due to a selective irradiation of neoplasics cells, after the patient receives a borate compound infusion and be subjected to a epithermal neutrons beam. This work presents dosimetric studies of the coupling techniques: BNCT with 252 Cf brachytherapy, conducted through computer simulation in MCNP5 code, using a precise and well discretized voxel model of human head, which was incorporated a representative Glioblastoma Multiform tumor. The dosimetric results from MCNP5 code were exported to SISCODES program, which generated isodose curves representing absorbed dose rate in the brain. Isodose curves, neutron fluency, and dose components from BNCT and 252 Cf brachytherapy are presented in this paper. (author)
International Nuclear Information System (INIS)
Miller, W.F. Jr.
1975-10-01
The coarse-mesh rebalance method, based on neutron conservation, is used in discrete ordinates neutron transport codes to accelerate convergence of the within-group scattering source. Though very powerful for this application, the method is ineffective in accelerating the iteration on the discrete-ordinates-to-spherical-harmonics fictitious sources used for ray-effect elimination. This is largely because this source makes a minimum contribution to the neutron balance equation. The traditional rebalance approach is derived in a variational framework and compared with new rebalance approaches tailored to be compatible with the fictitious source. The new approaches are compared numerically to determine their relative advantages. It is concluded that there is little incentive to use the new methods. (3 tables, 5 figures)
Coupled neutronics and thermal hydraulics modelling in reactor dynamics codes TRAB-3D and HEXTRAN
International Nuclear Information System (INIS)
Kyrki-Rajamaeki, R.; Raety, H.
1999-01-01
The reactor dynamics codes for transient and accident analyses inherently include the coupling of neutronics and thermal hydraulics modelling. In Finland a number of codes with 1D and 3D neutronic models have been developed, which include models also for the cooling circuits. They have been used mainly for the needs of Finnish power plants, but some of the codes have also been utilized elsewhere. The continuous validation, simultaneous development, and experiences obtained in commercial applications have considerably improved the performance and range of application of the codes. The fast operation of the codes has enabled realistic analysis of 3D core combined to a full model of the cooling circuit even in such long reactivity scenarios as ATWS. The reactor dynamics methods are developed further and new more detailed models are created for tasks related to increased safety requirements. For thermal hydraulics calculations, an accurate general flow model based on a new solution method has been developed. Although mainly intended for analysis purposes, the reactor dynamics codes also provide reference solutions for simulator applications. As computer capability increases, these more sophisticated methods can be taken into use also in simulator environments. (author)
ARCADIAR - A New Generation of Coupled Neutronics / Core Thermal- Hydraulics Code System at AREVA NP
International Nuclear Information System (INIS)
Curca-Tivig, Florin; Merk, Stephan; Pautz, Andreas; Thareau, Sebastien
2007-01-01
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 R and concludes on customer benefits. ARCADIA 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 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)
Numerical solution of the neutron transport equation using cellular neural networks
International Nuclear Information System (INIS)
Boroushaki, Mehrdad
2009-01-01
Various methods have been used for solving the neutron transport equation in the past, and a number of computer codes have been developed based on these solution methods. This paper describes a novel method for the solution of the steady-state and time-dependent neutron transport equation using the duality between neutronic parameters in the method of characteristic (MOC) and the electrical parameters in the cellular neural networks (CNN). The relevant electrical circuit can be simulated by professional electrical circuit simulator software, HSPICE. This software is used for numerical solution of the transport equation only by preparation of appropriate inputs. This method does not need inner and outer iterations, which is a necessary step in the other deterministic methods. One of the main applications of the proposed method may be the development of a new hardware by VLSI technology for online spatio-temporal calculations of the transport equation for nuclear reactor core. The accuracy and capability of this method are examined in a 2D steady-state problem for a BWR fuel assembly, and a 2D time-dependent TWIGL seed/blanket problem
Comparison of neutronic transport equation resolution nodal methods
International Nuclear Information System (INIS)
Zamonsky, O.M.; Gho, C.J.
1990-01-01
In this work, some transport equation resolution nodal methods are comparatively studied: the constant-constant (CC), linear-nodal (LN) and the constant-quadratic (CQ). A nodal scheme equivalent to finite differences has been used for its programming, permitting its inclusion in existing codes. Some bidimensional problems have been solved, showing that linear-nodal (LN) are, in general, obtained with accuracy in CPU shorter times. (Author) [es
Coupled neutronic-thermal-hydraulics analysis in a coolant subchannel of a PWR using CFD techniques
Energy Technology Data Exchange (ETDEWEB)
Ribeiro, Felipe P.; Su, Jian, E-mail: sujian@nuclear.ufrj.br [Coordenacao de Pos-Graduacao e Pesquisa de Engenharia (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Programa de Engenharia Nuclear
2017-07-01
The high capacity of Computational Fluid Dynamics code to predict multi-dimensional thermal-hydraulics behaviour and the increased availability of capable computer systems are making that method a good tool to simulate phenomena of thermal-hydraulics nature in nuclear reactors. However, since there are no neutron kinetics models available in commercial CFD codes to the present day, the application of CFD in the nuclear reactor safety analyses is still limited. The present work proposes the implementation of the point kinetics model (PKM) in ANSYS - Fluent to predict the neutronic behaviour in a Westinghouse Sequoyah nuclear reactor, coupling with the phenomena of heat conduction in the rod and thermal-hydraulics in the cooling fluid, via the reactivity feedback. Firstly, a mesh convergence and turbulence model study was performed, using the Reynolds-Average Navier-Stokes method, with square arrayed rod bundle featuring pitch to diameter ratio of 1:32. Secondly, simulations using the k-! SST turbulence model were performed with an axial distribution of the power generation in the fuel to analyse the heat transfer through the gap and cladding, and its in fluence on the thermal-hydraulics behaviour of the cooling fluid. The wall shear stress distribution for the centre-line rods and the dimensionless velocity were evaluated to validate the model, as well as the in fluence of the mass flow rate variation on the friction factor. The coupled model enabled to perform a dynamic analysis of the nuclear reactor during events of insertion of reactivity and shutdown of primary coolant pumps. (author)
International Nuclear Information System (INIS)
Jimenez, J.; Herrero, J. J.; Cuervo, D.; Aragones, J. M.
2010-10-01
Nowadays coupled 3-dimensional neutron kinetics and thermal-hydraulic core calculations are performed by applying a radial average channel approach using a meshing of one quarter of assembly in the best case. This approach does not take into account the subchannels effects due to the averaging of the physical fields and the loose of heterogeneity in the thermal-hydraulic model. Therefore the models do not have enough resolution to predict those subchannels effects which are important for the fuel design safety margins, because it is in the local scale, where we can search the hottest pellet or the maximum heat flux. The Polytechnic University of Madrid advanced multi-scale neutron-kinetics and thermal-hydraulics methodologies being implemented in COBAYA3 include domain decomposition by alternate core dissections for the local 3-dimensional fine-mesh scale problems (pin cells/subchannels) and an analytical nodal diffusion solver for the coarse mesh scale coupled with the thermal-hydraulic using a model of one channel per assembly or per quarter of assembly. In this work, we address the domain decomposition by the alternate core dissections methodology applied to solve coupled 3-dimensional neutronic-thermal-hydraulic problems at the fine-mesh scale. The neutronic-thermal-hydraulic coupling at the cell-subchannel scale allows the treatment of the effects of the detailed thermal-hydraulic feedbacks on cross-sections, thus resulting in better estimates of the local safety margins at the pin level. (Author)
Coupled neutronics and thermal hydraulics of high density cores for FRM II
Energy Technology Data Exchange (ETDEWEB)
Breitkreutz, Harald
2011-03-04
According to the 'Verwaltungsvereinbarung zwischen Bund und Land vom 30.5.2003' and its updating on 13.11.2010, the Forschungs-Neutronenquelle Heinz Maier-Leibnitz, Frm II, has to convert its fuel element to an uranium enrichment which is significantly lower than the current 93%, in case this is economically reasonable and doesn't impact the reactor performance immoderate. In the framework of this conversion, new calculations regarding neutronics and thermal hydraulics for the anticipated core configurations have to be made. The computational power available nowadays allows for detailed 3D calculations, on the neutronic as well as on the thermal hydraulic side. In this context, a new program system, 'X{sup 2}', was developed. It couples the Monte Carlo code McnpX, the computational fluid dynamics code Cfx and the burn-up code sequence MonteBurns. The codes were modified and extended to meet the requirements of the coupled calculation concept. To verify the new program system, highly detailed calculations for the current fuel element were made and compared to simulations and measurements that were performed in the past. The results strengthen the works performed so far and show that the original, conservative approach overestimates all critical thermal hydraulic values. Using the CFD software, effects like the impact of the combs that fix the fuel plates and the pressure drop at the edges of the fuel plates were studied in great detail for the first time. Afterwards, a number of possible new fuel elements with lower enrichment, based on disperse and monolithic UMo (uranium with 8 wt.-% Mo) were analysed. A number of straight-forward conversion scenarios was discussed, showing that a further compaction of the fuel element, an extended cycle length or an increased reactor power is needed to compensate the flux loss, which is caused by the lower enrichment. This flux loss is in excess of 7%. The discussed new fuel elements include a 50
Coupled neutronics and thermal hydraulics of high density cores for FRM II
International Nuclear Information System (INIS)
Breitkreutz, Harald
2011-01-01
According to the 'Verwaltungsvereinbarung zwischen Bund und Land vom 30.5.2003' and its updating on 13.11.2010, the Forschungs-Neutronenquelle Heinz Maier-Leibnitz, Frm II, has to convert its fuel element to an uranium enrichment which is significantly lower than the current 93%, in case this is economically reasonable and doesn't impact the reactor performance immoderate. In the framework of this conversion, new calculations regarding neutronics and thermal hydraulics for the anticipated core configurations have to be made. The computational power available nowadays allows for detailed 3D calculations, on the neutronic as well as on the thermal hydraulic side. In this context, a new program system, 'X 2 ', was developed. It couples the Monte Carlo code McnpX, the computational fluid dynamics code Cfx and the burn-up code sequence MonteBurns. The codes were modified and extended to meet the requirements of the coupled calculation concept. To verify the new program system, highly detailed calculations for the current fuel element were made and compared to simulations and measurements that were performed in the past. The results strengthen the works performed so far and show that the original, conservative approach overestimates all critical thermal hydraulic values. Using the CFD software, effects like the impact of the combs that fix the fuel plates and the pressure drop at the edges of the fuel plates were studied in great detail for the first time. Afterwards, a number of possible new fuel elements with lower enrichment, based on disperse and monolithic UMo (uranium with 8 wt.-% Mo) were analysed. A number of straight-forward conversion scenarios was discussed, showing that a further compaction of the fuel element, an extended cycle length or an increased reactor power is needed to compensate the flux loss, which is caused by the lower enrichment. This flux loss is in excess of 7%. The discussed new fuel elements include a 50% enriched disperse UMo core with
Energy Technology Data Exchange (ETDEWEB)
Zwermann, W.; Aures, A.; Bernnat, W.; and others
2013-06-15
This report documents the status of the research and development goals reached within the reactor safety research project RS1503 ''Development and Application of Neutron Transport Methods and Uncertainty Analyses for Reactor Core Calculations'' as of the 1{sup st} quarter of 2013. The superordinate goal of the project is the development, validation, and application of neutron transport methods and uncertainty analyses for reactor core calculations. These calculation methods will mainly be applied to problems related to the core behaviour of light water reactors and innovative reactor concepts. The contributions of this project towards achieving this goal are the further development, validation, and application of deterministic and stochastic calculation programmes and of methods for uncertainty and sensitivity analyses, as well as the assessment of artificial neutral networks, for providing a complete nuclear calculation chain. This comprises processing nuclear basis data, creating multi-group data for diffusion and transport codes, obtaining reference solutions for stationary states with Monte Carlo codes, performing coupled 3D full core analyses in diffusion approximation and with other deterministic and also Monte Carlo transport codes, and implementing uncertainty and sensitivity analyses with the aim of propagating uncertainties through the whole calculation chain from fuel assembly, spectral and depletion calculations to coupled transient analyses. This calculation chain shall be applicable to light water reactors and also to innovative reactor concepts, and therefore has to be extensively validated with the help of benchmarks and critical experiments.
Effect of Fast Neutron Irradiation on Current Transport Properties of HTS Materials
Ballarino, A; Kruglov, V S; Latushkin, S T; Lubimov, A N; Ryazanov, A I; Shavkin, S V; Taylor, T M; Volkov, P V
2004-01-01
The effect of fast neutron irradiation with energy up to 35 MeV and integrated fluence of up to 5 x 10**15 cm-2 on the current transport properties of HTS materials Bi-2212 and Bi-2223 has been studied, both at liquid nitrogen and at room temperatures. The samples irradiated were selected after verification of the stability of their superconducting properties after temperature cycling in the range of 77 K - 293 K. It has been found that the irradiation by fast neutrons up to the above dose does not produce a significant degradation of critical current. The effect of room temperature annealing on the recovery of transport properties of the irradiated samples is also reported, as is a preliminary microstructure investigation of the effect of irradiation on the soldered contacts.
On equiconvergence of ultraspherical polynomials solution of one-speed neutron transport equation
International Nuclear Information System (INIS)
Yilmazer, Ayhan; Tombakoglu, Mehmet
2006-01-01
Ultraspherical polynomial approximation is used in slab criticality calculations for strongly anisotropic scattering. A unique and general formulation is developed for slab criticality condition whose sub-cases are spherical harmonics approximation, Chebyshev polynomial approximation of first and second kind. Since Legendre polynomials, Chebyshev Polynomials of first and second kinds are special cases of ultraspherical polynomials; our formulation inherently covers all these approximations and lets one to employ any other ultraspherical polynomial approximation in the solution of one-speed neutron transport equation. Our calculations showed that solution of one-speed neutron transport equation for various degrees of anisotropy and cross-section parameters is almost insensitive to the choice of ultraspherical polynomial with the present days' computing capabilities. In other words, as much as high order ultraspherical polynomial approximation is used the solution converges to the same value for a specified problem regardless the type of the ultraspherical polynomial assigned in the solution as equiconvergence theorem of Jacobi polynomials states
Presentation of some methods for the solution of the monoenergetic neutrons transport equation
International Nuclear Information System (INIS)
Valle G, E. del.
1978-01-01
The neutrons transport theory problems whose solution has been reached were collected in order to show that the transport equation is so complicated that different techniques were developed so as to give approximative numerical solutions to problems concerning the practical application. Such a technique, which had not been investigated in the literature dealing with these problems, is described here. The results which were obtained through this technique in undimensional problems of criticity are satisfactory and speaking in a conceptual way this method is extremely simple because it times. There is no limitation to deal with problems related neutrons sources with an arbitrary distribution and in principle the application of this technique can be extended to unhomogeneous environments. (author)
Monte Carlo method for neutron transport calculations in graphics processing units (GPUs)
International Nuclear Information System (INIS)
Pellegrino, Esteban
2011-01-01
Monte Carlo simulation is well suited for solving the Boltzmann neutron transport equation in an inhomogeneous media for complicated geometries. However, routine applications require the computation time to be reduced to hours and even minutes in a desktop PC. The interest in adopting Graphics Processing Units (GPUs) for Monte Carlo acceleration is rapidly growing. This is due to the massive parallelism provided by the latest GPU technologies which is the most promising solution to the challenge of performing full-size reactor core analysis on a routine basis. In this study, Monte Carlo codes for a fixed-source neutron transport problem were developed for GPU environments in order to evaluate issues associated with computational speedup using GPUs. Results obtained in this work suggest that a speedup of several orders of magnitude is possible using the state-of-the-art GPU technologies. (author) [es
Modular, object-oriented redesign of a large-scale Monte Carlo neutron transport program
International Nuclear Information System (INIS)
Moskowitz, B.S.
2000-01-01
This paper describes the modular, object-oriented redesign of a large-scale Monte Carlo neutron transport program. This effort represents a complete 'white sheet of paper' rewrite of the code. In this paper, the motivation driving this project, the design objectives for the new version of the program, and the design choices and their consequences will be discussed. The design itself will also be described, including the important subsystems as well as the key classes within those subsystems
Normal and adjoint integral and integrodifferential neutron transport equations. Pt. 2
International Nuclear Information System (INIS)
Velarde, G.
1976-01-01
Using the simplifying hypotheses of the integrodifferential Boltzmann equations of neutron transport, given in JEN 334 report, several integral equations, and theirs adjoint ones, are obtained. Relations between the different normal and adjoint eigenfunctions are established and, in particular, proceeding from the integrodifferential Boltzmann equation it's found out the relation between the solutions of the adjoint equation of its integral one, and the solutions of the integral equation of its adjoint one (author)
International Nuclear Information System (INIS)
Pessine, E.J.
1978-01-01
Typical half-space problems in two-group neutron transport theory are solved numerically using the singular-eigenfunction-expansion technique, considering isotropic-and linearly anisotropic scattering. Numerical results are reported for the Albedo, Milne and Constant-Source problems in a half-space pure light-water medium using isotropic scattering data set of Metacalf and Zweifel and considering various degrees of anisotropy [pt
Neutronic / thermal-hydraulic coupling with the code system Trace / Parcs
International Nuclear Information System (INIS)
Mejia S, D. M.; Del Valle G, E.
2015-09-01
The developed models for Parcs and Trace codes corresponding for the cycle 15 of the Unit 1 of the Laguna Verde nuclear power plant are described. The first focused to the neutronic simulation and the second to thermal hydraulics. The model developed for Parcs consists of a core of 444 fuel assemblies wrapped in a radial reflective layer and two layers, a superior and another inferior, of axial reflector. The core consists of 27 total axial planes. The model for Trace includes the vessel and its internal components as well as various safety systems. The coupling between the two codes is through two maps that allow its intercommunication. Both codes are used in coupled form performing a dynamic simulation that allows obtaining acceptably a stable state from which is carried out the closure of all the main steam isolation valves (MSIVs) followed by the performance of safety relief valves (SRVs) and ECCS. The results for the power and reactivities introduced by the moderator density, the fuel temperature and total temperature are shown. Data are also provided like: the behavior of the pressure in the steam dome, the water level in the downcomer, the flow through the MSIVs and SRVs. The results are explained for the power, the pressure in the steam dome and the water level in the downcomer which show agreement with the actions of the MSIVs, SRVs and ECCS. (Author)
Testing universal relations of neutron stars with a nonlinear matter-gravity coupling theory
International Nuclear Information System (INIS)
Sham, Y.-H.; Lin, L.-M.; Leung, P. T.
2014-01-01
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.
Development of a finite element method for neutron transport equation approximations
Vidal Ferràndiz, Antoni
2018-01-01
La ecuación del transporte neutrónico describe la población de neutrones y las reacciones nucleares dentro de un reactor nuclear. Primero, introducimos esta ecuación y las aproximaciones de la misma. Entonces, estudiamos la ecuación de la difusión neutrónica, la aproximación al transporte más utilizada. Para el caso estacionario, esta aproximación da lugar a un problema diferencial de valores propios. Para resolver la ecuación de la difusión se ha desarrollado un método de elementos finitos h...
A portable, parallel, object-oriented Monte Carlo neutron transport code in C++
International Nuclear Information System (INIS)
Lee, S.R.; Cummings, J.C.; Nolen, S.D.
1997-01-01
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 α-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 α-eigenvalues in seconds to minutes rather than hours to days. Future plans and the implementation of a general transport physics framework are also discussed
Neutron and gamma ray transport calculations in shielding system
Energy Technology Data Exchange (ETDEWEB)
Masukawa, Fumihiro; Sakamoto, Hiroki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment
1998-03-01
In the shields for radiation in nuclear facilities, the penetrating holes of various kinds and irregular shapes are made for the reasons of operation, control and others. These penetrating holes and gaps are filled with air or the substances with relatively small shielding performance, and radiation flows out through them, which is called streaming. As the calculation techniques for the shielding design or analysis related to the streaming problem, there are the calculations by simplified evaluation, transport calculation and Monte Carlo method. In this report, the example of calculation by Monte Carlo method which is represented by MCNP code is discussed. A number of variance reduction techniques which seem effective for the analysis of streaming problem were tried. As to the investigation of the applicability of MCNP code to streaming analysis, the object of analysis which are the concrete walls without hole and with horizontal hole, oblique hole and bent oblique hole, the analysis procedure, the composition of concrete, and the conversion coefficient of dose equivalent, and the results of analysis are reported. As for variance reduction technique, cell importance was adopted. (K.I.)
International Nuclear Information System (INIS)
Aniel-Buchheit, Sylvie; Podowski, Michael Z.
2006-01-01
The purpose of this paper is to discuss the development in progress of a complete space- and time-dependent model of the coupled neutron kinetic and reactor thermal-hydraulics. The neutron kinetics model is based on two-group diffusion equations with Doppler and void reactivity feedback effects. This model is coupled with the model of two-phase flow and heat transfer in parallel coolant channels. The modeling concepts considered for this purpose include one-dimensional drift flux and two-fluid models, as well a CFD model implemented in the NPHASE advanced computational multiphase fluid dynamics (CMFD) computer code. Two methods of solution for the overall model are proposed. One is based on direct numerical integration of the spatially-discretized governing equations. The other approach is based on a quasi-analytical modal approach to the neutronics model, in which a complete set of eigenvectors is found for step-wise temporal changes of the cross-sections of core materials (fuel and coolant/moderator). The issues investigated in the paper include details of model formulation, as well as the results of calculations for neutronically-coupled density-wave oscillations. (authors)
Žukauskaite, A; Plukiene, R; Plukis, A
2007-01-01
Particle accelerators and other high energy facilities produce penetrating ionizing radiation (neutrons and γ-rays) that must be shielded. The objective of this work was to model photon and neutron transport in various materials, usually used as shielding, such as concrete, iron or graphite. Monte Carlo method allows obtaining answers by simulating individual particles and recording some aspects of their average behavior. In this work several nuclear experiments were modeled: AVF 65 – γ-ray beams (1-10 MeV), HIMAC and ISIS-800 – high energy neutrons (20-800 MeV) transport in iron and concrete. The results were then compared with experimental data.
Patel, R.A.; Perko, J.; Jaques, D.; De Schutter, G.; Ye, G.; Van Breugel, K.
2013-01-01
A Lattice Boltzmann (LB) based reactive transport model intended to capture reactions and solid phase changes occurring at the pore scale is presented. The proposed approach uses LB method to compute multi component mass transport. The LB multi-component transport model is then coupled with the
Energy Coupling Factor-Type ABC Transporters for Vitamin Uptake in Prokaryotes
Erkens, Guus B.; Dosz-Majsnerowska, Maria; ter Beek, Josy; Slotboom, Dirk Jan
2012-01-01
Energy coupling factor (ECF) transporters are a subgroup of ATP-binding cassette (ABC) transporters involved in the uptake of vitamins and micronutrients in prokaryotes. In contrast to classical ABC importers, ECF transporters do not make use of water-soluble substrate binding proteins or domains
International Nuclear Information System (INIS)
Takahashi, A.; Rusch, D.
1979-07-01
Some recent neutronics experiments for fusion reactor blankets show that the precise treatment of anisotropic secondary emissions for all types of neutron scattering is needed for neutron transport calculations. In the present work new rigorous methods, i.e. based on non-approximative microscopic neutron balance equations, are applied to treat the anisotropic collision source term in transport equations. The collision source calculation is free from approximations except for the discretization of energy, angle and space variables and includes the rigorous treatment of nonelastic collisions, as far as nuclear data are given. Two methods are presented: first the Ii-method, which relies on existing nuclear data files and then, as an ultimate goal, the I*-method, which aims at the use of future double-differential cross section data, but which is also applicable to the present single-differential data basis to allow a smooth transition to the new data type. An application of the Ii-method is given in the code system NITRAN which employs the Ssub(N)-method to solve the transport equations. Both rigorous methods, the Ii- and the I*-method, are applicable to all radiation transport problems and they can be used also in the Monte-Carlo-method to solve the transport problem. (orig./RW) [de
How LeuT shapes our understanding of the mechanisms of sodium-coupled neurotransmitter transporters.
Penmatsa, Aravind; Gouaux, Eric
2014-03-01
Neurotransmitter transporters are ion-coupled symporters that drive the uptake of neurotransmitters from neural synapses. In the past decade, the structure of a bacterial amino acid transporter, leucine transporter (LeuT), has given valuable insights into the understanding of architecture and mechanism of mammalian neurotransmitter transporters. Different conformations of LeuT, including a substrate-free state, inward-open state, and competitive and non-competitive inhibitor-bound states, have revealed a mechanistic framework for the transport and transport inhibition of neurotransmitters. The current review integrates our understanding of the mechanistic and pharmacological properties of eukaryotic neurotransmitter transporters obtained through structural snapshots of LeuT.
International Nuclear Information System (INIS)
Abreu, Marcos Pimenta de
2006-01-01
In this article, we extend the one-speed multi-layer models to neutron reflection and transmission developed in our earlier work (de Abreu, M.P., 2005. Multi-layer models to neutron reflection and transmission for whole-core transport calculations, Annals of Nuclear Energy 32, 215) to multigroup transport theory. We begin by considering a two-layer boundary region, and we develop for such a region discrete ordinates models to the diffuse reflection and transmission of neutrons for multigroup nuclear reactor core problems with anisotropic scattering. We perform numerical experiments to show that our models to neutron reflection and transmission can be used to replace efficiently and accurately two nonactive boundary layers in whole-core transport calculations. We conclude this article with an inductive extension of our two-layer results to a boundary region with an arbitrary number of layers
International Nuclear Information System (INIS)
Johnson, Jeffrey O.; Gallmeier, Franz X.; Popova, Irina
2002-01-01
Determining the bulk shielding requirements for accelerator environments is generally an easy task compared to analyzing the radiation transport through the complex shield configurations and penetrations typically associated with the detailed Title II design efforts of a facility. Shielding calculations for penetrations in the SNS accelerator environment are presented based on hybrid Monte Carlo and discrete ordinates particle transport methods. This methodology relies on coupling tools that map boundary surface leakage information from the Monte Carlo calculations to boundary sources for one-, two-, and three-dimensional discrete ordinates calculations. The paper will briefly introduce the coupling tools for coupling MCNPX to the one-, two-, and three-dimensional discrete ordinates codes in the DOORS code suite. The paper will briefly present typical applications of these tools in the design of complex shield configurations and penetrations in the SNS proton beam transport system
Positive solution of a time and energy dependent neutron transport problem
International Nuclear Information System (INIS)
Pao, C.V.
1975-01-01
A constructive method is given for the determination of a solution and an existence--uniqueness theorem for some nonlinear time and energy dependent neutron transport problems, including the linear transport system. The geometry of the medium under consideration is allowed to be either bounded or unbounded which includes the geometry of a finite or infinite cylinder, a half-space and the whole space R/subm/ (m=1,2,center-dotcenter-dotcenter-dot). Our approach to the problem is by successive approximation which leads to various recursion formulas for the approximations in terms of explicit integrations. It is shown under some Lipschitz conditions on the nonlinear functions, which describe the process of neutrons absorption, fission, and scattering, that the sequence of approximations converges to a unique positive solution. Since these conditions are satisfied by the linear transport equation, all the results for the nonlinear system are valid for the linear transport problem. In the general nonlinear problem, the existence of both local and global solutions are discussed, and an iterative process for the construction of the solution is given
International Nuclear Information System (INIS)
Arreola V, G.; Vazquez R, R.; Guzman A, J. R.
2012-10-01
In this work a comparative analysis of the results for the neutrons dispersion in a not multiplicative semi-infinite medium is presented. One of the frontiers of this medium is located in the origin of coordinates, where a neutrons source in beam form, i.e., μο=1 is also. The neutrons dispersion is studied on the statistical method of Monte Carlo and through the unidimensional transport theory and for an energy group. The application of transport theory gives a semi-analytic solution for this problem while the statistical solution for the flow was obtained applying the MCNPX code. The dispersion in light water and heavy water was studied. A first remarkable result is that both methods locate the maximum of the neutrons distribution to less than two mean free trajectories of transport for heavy water, while for the light water is less than ten mean free trajectories of transport; the differences between both methods is major for the light water case. A second remarkable result is that the tendency of both distributions is similar in small mean free trajectories, while in big mean free trajectories the transport theory spreads to an asymptote value and the solution in base statistical method spreads to zero. The existence of a neutron current of low energy and toward the source is demonstrated, in contrary sense to the neutron current of high energy coming from the own source. (Author)
International Nuclear Information System (INIS)
Shanjie, Xiao; Tatjana, Jevremovic
2010-01-01
The accurate, detailed and 3D neutron transport analysis for Gen-IV reactors is still time-consuming regardless of advanced computational hardware available in developed countries. This paper introduces a new concept in addressing the computational time while persevering the detailed and accurate modeling; a specifically designed FPGA co-processor accelerates robust AGENT methodology for complex reactor geometries. For the first time this approach is applied to accelerate the neutronics analysis. The AGENT methodology solves neutron transport equation using the method of characteristics. The AGENT methodology performance was carefully analyzed before the hardware design based on the FPGA co-processor was adopted. The most time-consuming kernel part is then transplanted into the FPGA co-processor. The FPGA co-processor is designed with data flow-driven non von-Neumann architecture and has much higher efficiency than the conventional computer architecture. Details of the FPGA co-processor design are introduced and the design is benchmarked using two different examples. The advanced chip architecture helps the FPGA co-processor obtaining more than 20 times speed up with its working frequency much lower than the CPU frequency. (authors)
International Nuclear Information System (INIS)
Teixeira, Paulo Cleber Mendonca
2002-12-01
In this study, an analytical solution of the neutron transport equation in an annular reactor is presented with a short and rotating neutron source of the type S(x) δ (x- Vt), where V is the speed of annular pulsed reactor. The study is an extension of a previous study by Williams [12] carried out with a pulsed source of the type S(x) δ (t). In the new concept of annular pulsed reactor designed to produce continuous high flux, the core consists of a subcritical annular geometry pulsed by a rotating modulator, producing local super prompt critical condition, thereby giving origin to a rotating neutron pulse. An analytical solution is obtained by opening up of the annular geometry and applying one energy group transport theory in one dimension using applied mathematical techniques of Laplace transform and Complex Variables. The general solution for the flux consists of a fundamental mode, a finite number of harmonics and a transient integral. A condition which limits the number of harmonics depending upon the circumference of the annular geometry has been obtained. Inverse Laplace transform technique is used to analyse instability condition in annular reactor core. A regenerator parameter in conjunction with perimeter of the ring and nuclear properties is used to obtain stable and unstable harmonics and to verify if these exist. It is found that the solution does not present instability in the conditions stated in the new concept of annular pulsed reactor. (author)
International Nuclear Information System (INIS)
Rowe, M.; Manalo, K.; Plower, T.; Sjoden, G.
2009-01-01
Evaluation of silicon carbide (SiC) semiconductor detectors for use in power monitoring is of significant interest because of their distinct advantages, including small size, small mass, and their inactivity both chemically and neutronically. The main focus of this paper includes evaluating the predicted response of a SiC detector when placed in a 17 x 17 Westinghouse PWR assembly, using the PENTRAN code system for the 3-D deterministic adjoint transport computations. Adjoint transport results indicated maximum adjoint values of 1, 0.507 and 0.308 were obtained for the thermal, epithermal and fast neutron energy groups, respectively. Within a radial distance of 6.08 cm from the SiC detector, local fuel pins contribute 75.33% at this radius within the thermal group response. A total of 35.85% of the response in the epithermal group is accounted for in the same 6.08 cm radius; similarly, 21.58% of the fast group response is accounted for in the same radius. This means that for neutrons, the effective monitoring range of the SiC detectors is on the order of five fuel pins away from the detector; pins outside this range in the fuel lattice are minimally 'seen' by the SiC detector. (authors)
GNES-R: Global nuclear energy simulator for reactors task 1: High-fidelity neutron transport
International Nuclear Information System (INIS)
Clarno, K.; De Almeida, V.; D'Azevedo, E.; De Oliveira, C.; Hamilton, S.
2006-01-01
A multi-laboratory, multi-university collaboration has formed to advance the state-of-the-art in high-fidelity, coupled-physics simulation of nuclear energy systems. We are embarking on the first-phase in the development of a new suite of simulation tools dedicated to the advancement of nuclear science and engineering technologies. We seek to develop and demonstrate a new generation of multi-physics simulation tools that will explore the scientific phenomena of tightly coupled physics parameters within nuclear systems, support the design and licensing of advanced nuclear reactors, and provide benchmark quality solutions for code validation. In this paper, we have presented the general scope of the collaborative project and discuss the specific challenges of high-fidelity neutronics for nuclear reactor simulation and the inroads we have made along this path. The high-performance computing neutronics code system utilizes the latest version of SCALE to generate accurate, problem-dependent cross sections, which are used in NEWTRNX - a new 3-D, general-geometry, discrete-ordinates solver based on the Slice-Balance Approach. The Global Nuclear Energy Simulator for Reactors (GNES-R) team is embarking on a long-term simulation development project that encompasses multiple laboratories and universities for the expansion of high-fidelity coupled-physics simulation of nuclear energy systems. (authors)
International Nuclear Information System (INIS)
Peng Muzhang; Zhang Quan; Wang Guoli; Zhang Yuman
1988-01-01
TISKTH-3 is a coupled neutronics/thermal-hydraulics code for the transient analysis. A 3-dimensional neutron kinetics equation solved by the Nodal Green's Function Method is used for the neutronics model of the code. A homogeneous equilibrium model with a complete boiling curve and two numerical solutions of the implicit and explicit scheme is used for the thermal-hydraulics model of the code. A 2-dimensional heat conduction equation with variable conductivity solved by the method of weighted residuals is used for the fuel rod heat transfer model of the code. TISKTH-3 is able to analyze the fast transient process and complicate accident situations in the core. The initative applications have shown that the stability and convergency in the calculations with the code are satisfactory
CAD-Based Monte Carlo Neutron Transport KSTAR Analysis for KSTAR
Seo, Geon Ho; Choi, Sung Hoon; Shim, Hyung Jin
2017-09-01
The Monte Carlo (MC) neutron transport analysis for a complex nuclear system such as fusion facility may require accurate modeling of its complicated geometry. In order to take advantage of modeling capability of the computer aided design (CAD) system for the MC neutronics analysis, the Seoul National University MC code, McCARD, has been augmented with a CAD-based geometry processing module by imbedding the OpenCASCADE CAD kernel. In the developed module, the CAD geometry data are internally converted to the constructive solid geometry model with help of the CAD kernel. An efficient cell-searching algorithm is devised for the void space treatment. The performance of the CAD-based McCARD calculations are tested for the Korea Superconducting Tokamak Advanced Research device by comparing with results of the conventional MC calculations using a text-based geometry input.
Response matrix method for neutron transport in reactor lattices using group symmetry properties
International Nuclear Information System (INIS)
Mund, E.H.
1991-01-01
This paper describes a response matrix method for the approximate solution of one-velocity, multi-dimensional transport problems in reactor lattices, with isotropic neutron scattering. The transport equation is solved on a homogeneous cell by using a Petrov-Galerkin technique based on a set of trial and test functions (including polynomials and exponential functions) closely related to transport problems in infinite media. The number of non-zero elements of the response matrices reduces to a minimum when the symmetry properties of the cell are included ab initio in the span of the basis functions. To include these properties, use is made of projection operations which are performed very efficiently on symbolic manipulation programs. Numerical results of model problems in square geometry show a good agreement with reference solutions
The use of symbolic computation in radiative, energy, and neutron transport calculations
Frankel, J. I.
This investigation uses symbolic computation in developing analytical methods and general computational strategies for solving both linear and nonlinear, regular and singular, integral and integro-differential equations which appear in radiative and combined mode energy transport. This technical report summarizes the research conducted during the first nine months of the present investigation. The use of Chebyshev polynomials augmented with symbolic computation has clearly been demonstrated in problems involving radiative (or neutron) transport, and mixed-mode energy transport. Theoretical issues related to convergence, errors, and accuracy have also been pursued. Three manuscripts have resulted from the funded research. These manuscripts have been submitted to archival journals. At the present time, an investigation involving a conductive and radiative medium is underway. The mathematical formulation leads to a system of nonlinear, weakly-singular integral equations involving the unknown temperature and various Legendre moments of the radiative intensity in a participating medium. Some preliminary results are presented illustrating the direction of the proposed research.
Radiation Transport Analysis in Chalcogenide-Based Devices and a Neutron Howitzer Using MCNP
Bowler, Herbert
As photons, electrons, and neutrons traverse a medium, they impart their energy in ways that are analytically difficult to describe. Monte Carlo methods provide valuable insight into understanding this behavior, especially when the radiation source or environment is too complex to simplify. This research investigates simulating various radiation sources using the Monte Carlo N-Particle (MCNP) transport code, characterizing their impact on various materials, and comparing the simulation results to general theory and measurements. A total of five sources were of interest: two photon sources of different incident particle energies (3.83 eV and 1.25 MeV), two electron sources also of different energies (30 keV and 100 keV), and a californium-252 (Cf-252) spontaneous fission neutron source. Lateral and vertical programmable metallization cells (PMCs) were developed by other researchers for exposure to these photon and electron sources, so simplified PMC models were implemented in MCNP to estimate the doses and fluences. Dose rates measured around the neutron source and the predicted maximum activity of activation foils exposed to the neutrons were determined using MCNP and compared to experimental results obtained from gamma-ray spectroscopy. The analytical fluence calculations for the photon and electron cases agreed with MCNP results, and differences are due to MCNP considering particle movements that hand calculations do not. Doses for the photon cases agreed between the analytical and simulated results, while the electron cases differed by a factor of up to 4.8. Physical dose rate measurements taken from the neutron source agreed with MCNP within the 10% tolerance of the measurement device. The activity results had a percent error of up to 50%, which suggests a need to further evaluate the spectroscopy setup.
Energy Technology Data Exchange (ETDEWEB)
Pazianotto, Mauricio Tizziani; Carlson, Brett Vern [Instituto Tecnologico de Aeronautica (ITA), Sao Jose dos Campos, SP (Brazil); Federico, Claudio Antonio; Goncalez, Odair Lelis [Centro Tecnico Aeroespacial (CTA), Sao Jose dos Campos, SP (Brazil). Instituto de Estudos Avancados
2011-07-01
Full text: Great effort is required to understand better the cosmic radiation (CR) dose received by sensitive equipment, on-board computers and aircraft crew members at Brazil airspace, because there is a large area of South America and Brazil subject to the South Atlantic Anomaly (SAA). High energy neutrons are produced by interactions between primary cosmic ray and atmospheric atoms, and also undergo moderation resulting in a wider spectrum of energy ranging from thermal energies (0:025eV ) to energies of several hundreds of MeV. Measurements of the cosmic radiation dose on-board aircrafts need to be followed with an integral flow monitor on the ground level in order to register CR intensity variations during the measurements. The Long Counter (LC) neutron detector was designed as a directional neutron flux meter standard because it presents fairly constant response for energy under 10MeV. However we would like to use it as a ground based neutron monitor for cosmic ray induced neutron spectrum (CRINS) that presents an isotropic fluency and a wider spectrum of energy. The LC was modeled and tested using a Monte Carlo transport simulation for irradiations with known neutron sources ({sup 241}Am-Be and {sup 251}Cf) as a benchmark. Using this geometric model its efficiency was calculated to CRINS isotropic flux, introducing high energy neutron interactions models. The objective of this work is to present the model for simulation of the isotropic neutron source employing the MCNPX code (Monte Carlo N-Particle eXtended) and then access the LC efficiency to compare it with experimental results for cosmic ray neutrons measures on ground level. (author)
Energy Technology Data Exchange (ETDEWEB)
Kramer, K J; Latkowski, J F; Abbott, R P; Boyd, J K; Powers, J J; Seifried, J E
2008-10-24
Lawrence Livermore National Laboratory is currently developing a hybrid fusion-fission nuclear energy system, called LIFE, to generate power and burn nuclear waste. We utilize inertial confinement fusion to drive a subcritical fission blanket surrounding the fusion chamber. It is composed of TRISO-based fuel cooled by the molten salt flibe. Low-yield (37.5 MJ) targets and a repetition rate of 13.3 Hz produce a 500 MW fusion source that is coupled to the subcritical blanket, which provides an additional gain of 4-8, depending on the fuel. In the present work, we describe the neutron transport and nuclear burnup analysis. We utilize standard analysis tools including, the Monte Carlo N-Particle (MCNP) transport code, ORIGEN2 and Monteburns to perform the nuclear design. These analyses focus primarily on a fuel composed of depleted uranium not requiring chemical reprocessing or enrichment. However, other fuels such as weapons grade plutonium and highly-enriched uranium are also under consideration. In addition, we have developed a methodology using {sup 6}Li as a burnable poison to replace the tritium burned in the fusion targets and to maintain constant power over the lifetime of the engine. The results from depleted uranium analyses suggest up to 99% burnup of actinides is attainable while maintaining full power at 2GW for more than five decades.
Energy Technology Data Exchange (ETDEWEB)
Fournier, D.; Le Tellier, R.; Suteau, C., E-mail: damien.fournier@cea.fr, E-mail: romain.le-tellier@cea.fr, E-mail: christophe.suteau@cea.fr [CEA, DEN, DER/SPRC/LEPh, Cadarache, Saint Paul-lez-Durance (France); Herbin, R., E-mail: raphaele.herbin@cmi.univ-mrs.fr [Laboratoire d' Analyse et de Topologie de Marseille, Centre de Math´ematiques et Informatique (CMI), Universit´e de Provence, Marseille Cedex (France)
2011-07-01
The solution of the time-independent neutron transport equation in a deterministic way invariably consists in the successive discretization of the three variables: energy, angle and space. In the SNATCH solver used in this study, the energy and the angle are respectively discretized with a multigroup approach and the discrete ordinate method. A set of spatial coupled transport equations is obtained and solved using the Discontinuous Galerkin Finite Element Method (DGFEM). Within this method, the spatial domain is decomposed into elements and the solution is approximated by a hierarchical polynomial basis in each one. This approach is time and memory consuming when the mesh becomes fine or the basis order high. To improve the computational time and the memory footprint, adaptive algorithms are proposed. These algorithms are based on an error estimation in each cell. If the error is important in a given region, the mesh has to be refined (h−refinement) or the polynomial basis order increased (p−refinement). This paper is related to the choice between the two types of refinement. Two ways to estimate the error are compared on different benchmarks. Analyzing the differences, a hp−refinement method is proposed and tested. (author)
International Nuclear Information System (INIS)
Fournier, D.; Le Tellier, R.; Suteau, C.; Herbin, R.
2011-01-01
The solution of the time-independent neutron transport equation in a deterministic way invariably consists in the successive discretization of the three variables: energy, angle and space. In the SNATCH solver used in this study, the energy and the angle are respectively discretized with a multigroup approach and the discrete ordinate method. A set of spatial coupled transport equations is obtained and solved using the Discontinuous Galerkin Finite Element Method (DGFEM). Within this method, the spatial domain is decomposed into elements and the solution is approximated by a hierarchical polynomial basis in each one. This approach is time and memory consuming when the mesh becomes fine or the basis order high. To improve the computational time and the memory footprint, adaptive algorithms are proposed. These algorithms are based on an error estimation in each cell. If the error is important in a given region, the mesh has to be refined (h−refinement) or the polynomial basis order increased (p−refinement). This paper is related to the choice between the two types of refinement. Two ways to estimate the error are compared on different benchmarks. Analyzing the differences, a hp−refinement method is proposed and tested. (author)
A Coupled Chemical and Mass Transport Model for Concrete Durability
DEFF Research Database (Denmark)
Jensen, Mads Mønster; Johannesson, Björn; Geiker, Mette Rica
2012-01-01
In this paper a general continuum theory is used to evaluate the service life of cement based materials, in terms of mass transport processes and chemical degradation of the solid matrix. The model established is a reactive mass transport model, based on an extended version of the Poisson-Nernst-...
Temporal moment analysis of solute transport in a coupled fracture ...
Indian Academy of Sciences (India)
Study on fluid flow and transport of solute through fractures has been an area of great interest among hydro-geologists during past few ... affect the solute transport through fracture are advection, dispersion, matrix diffusion, sorption and degradation. Among the ...... α0 = Local fracture dispersivity, [L]; ρs = Bulk density of the ...
Evolution of Neutron Star Magnetic Fields
Indian Academy of Sciences (India)
R. Narasimhan (Krishtel eMaging) 1461 1996 Oct 15 13:05:22
The magnetic field of a neutron star determines the evolution of its spin, its radia- tive properties and its interaction with the ... resulting in metal-like transport properties (electrical and heat conductivities) in this region (Yakovlev & Urpin ... from the spinning neutron star via magnetic coupling. The shorter the decay time scale.
Effect of spin rotation coupling on spin transport
International Nuclear Information System (INIS)
Chowdhury, Debashree; Basu, B.
2013-01-01
We have studied the spin rotation coupling (SRC) as an ingredient to explain different spin-related issues. This special kind of coupling can play the role of a Dresselhaus like coupling in certain conditions. Consequently, one can control the spin splitting, induced by the Dresselhaus like term, which is unusual in a semiconductor heterostructure. Within this framework, we also study the renormalization of the spin-dependent electric field and spin current due to the k → ⋅p → perturbation, by taking into account the interband mixing in the rotating system. In this paper we predict the enhancement of the spin-dependent electric field resulting from the renormalized spin rotation coupling. The renormalization factor of the spin electric field is different from that of the SRC or Zeeman coupling. The effect of renormalized SRC on spin current and Berry curvature is also studied. Interestingly, in the presence of this SRC-induced SOC it is possible to describe spin splitting as well as spin galvanic effect in semiconductors. -- Highlights: •Studied effect of spin rotation coupling on the spin electric field, spin current and Berry curvature. •In the k → ⋅p → framework we study the renormalization of spin electric field and spin current. •For an inertial system we have discussed the spin splitting. •Expression for the Berry phase in the inertial system is discussed. •The inertial spin galvanic effect is studied
Israelashvili, I.; Coimbra, A. E. C.; Vartsky, D.; Arazi, L.; Shchemelinin, S.; Caspi, E. N.; Breskin, A.
2017-09-01
Gamma-ray and fast-neutron imaging was performed with a novel liquid xenon (LXe) scintillation detector read out by a Gaseous Photomultiplier (GPM). The 100 mm diameter detector prototype comprised a capillary-filled LXe converter/scintillator, coupled to a triple-THGEM imaging-GPM, with its first electrode coated by a CsI UV-photocathode, operated in Ne/5%CH4 at cryogenic temperatures. Radiation localization in 2D was derived from scintillation-induced photoelectron avalanches, measured on the GPM's segmented anode. The localization properties of 60Co gamma-rays and a mixed fast-neutron/gamma-ray field from an AmBe neutron source were derived from irradiation of a Pb edge absorber. Spatial resolutions of 12± 2 mm and 10± 2 mm (FWHM) were reached with 60Co and AmBe sources, respectively. The experimental results are in good agreement with GEANT4 simulations. The calculated ultimate expected resolutions for our application-relevant 4.4 and 15.1 MeV gamma-rays and 1-15 MeV neutrons are 2-4 mm and ~ 2 mm (FWHM), respectively. These results indicate the potential applicability of the new detector concept to Fast-Neutron Resonance Radiography (FNRR) and Dual-Discrete-Energy Gamma Radiography (DDEGR) of large objects.
Energy Technology Data Exchange (ETDEWEB)
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.
International Nuclear Information System (INIS)
Valdes Parra, J.J.
1986-01-01
One of the main problems in reactor physics is to determine the neutron distribution in reactor core, since knowing that, it is possible to calculate the rapidity of occurrence of different nuclear reaction inside the reactor core. Within different theories existing in nuclear reactor physics, is neutron transport the one in which equation who govern the exact behavior of neutronic distribution are developed even inside the proper neutron transport theory, there exist different methods of solution which are approximations to exact solution; still more, with the purpose to reach a more precise solution, the majority of methods have been approached to the obtention of solutions in numerical form with the aim of take the advantages of modern computers, and for this reason a great deal of effort is dedicated to numerical solution of the equations of neutron transport. In agreement with the above mentioned, in this work has been developed a computer program which uses a relatively new techniques known as 'acceleration of synthetic diffusion' which has been applied to solve the neutron transport equation with 'classical schemes of spatial integration' obtaining results with a smaller quantity of interactions, if they compare to done without using such equation (Author)
Stability and accuracy of 3D neutron transport simulations using the 2D/1D method in MPACT
Collins, Benjamin; Stimpson, Shane; Kelley, Blake W.; Young, Mitchell T. H.; Kochunas, Brendan; Graham, Aaron; Larsen, Edward W.; Downar, Thomas; Godfrey, Andrew
2016-12-01
A consistent "2D/1D" neutron transport method is derived from the 3D Boltzmann transport equation, to calculate fuel-pin-resolved neutron fluxes for realistic full-core Pressurized Water Reactor (PWR) problems. The 2D/1D method employs the Method of Characteristics to discretize the radial variables and a lower order transport solution to discretize the axial variable. This paper describes the theory of the 2D/1D method and its implementation in the MPACT code, which has become the whole-core deterministic neutron transport solver for the Consortium for Advanced Simulations of Light Water Reactors (CASL) core simulator VERA-CS. Several applications have been performed on both leadership-class and industry-class computing clusters. Results are presented for whole-core solutions of the Watts Bar Nuclear Power Station Unit 1 and compared to both continuous-energy Monte Carlo results and plant data.
Application of the three-dimensional transport code to analysis of the neutron streaming experiment
International Nuclear Information System (INIS)
Chatani, K.; Slater, C.O.
1990-01-01
The neutron streaming through an experimental mock-up of a Clinch River Breeder Reactor (CRBR) prototypic coolant pipe chaseway was recalculated with a three-dimensional discrete ordinates code. The experiment was conducted at the Tower Shielding Facility at Oak Ridge National Laboratory in 1976 and 1977. The measurement of the neutron flux, using Bonner ball detectors, indicated nine orders of attenuation in the empty pipeway, which contained two 90-deg bends and was surrounded by concrete walls. The measurement data were originally analyzed using the DOT3.5 two-dimensional discrete ordinates radiation transport code. However, the results did not agree with measurement data at the bend because of the difficulties in modeling the three-dimensional configurations using two-dimensional methods. The two-dimensional calculations used a three-step procedure in which each of the three legs making the two 90-deg bends was a separate calculation. The experiment was recently analyzed with the TORT three-dimensional discrete ordinates radiation transport code, not only to compare the calculational results with the experimental results, but also to compare with results obtained from analyses in Japan using DOT3.5, MORSE, and ENSEMBLE, which is a three-dimensional discrete ordinates radiation transport code developed in Japan
International Nuclear Information System (INIS)
Chung, Yong Sam; Choi, Kwang Soon; Moon, Jong Hwa; Kim, Sun Ha; Lim, Jong Myoung; Kim, Young Jin; Quraishi, Shamshad Begum
2003-05-01
Elemental analyses for certified reference materials were carried out using instrumental neutron activation analysis and inductively coupled plasma-atomic emission spectrometry. Five Certified Reference Materials (CRM) were selected for the study on comparative analysis of environmental samples. The CRM are Soil (NIST SRM 2709), Coal fly ash (NIST SRM 1633a), urban dust (NIST SRM 1649a) and air particulate on filter media (NIST SRM 2783 and human hair (GBW 09101)
Energy Technology Data Exchange (ETDEWEB)
Reis, Patricia A.L.; Costa, Antonella L.; Hamers, Adolfo R.; Pereira, Claubia; Rodrigues, Thiago D.A.; Mantecon, Javier G.; Veloso, Maria A.F., E-mail: patricialire@yahoo.com.br, E-mail: antonella@nuclear.ufmg.br, E-mail: adolforomerohamers@hotmail.com, E-mail: claubia@nuclear.ufmg.br, E-mail: thiagodanielbh@gmail.com, E-mail: mantecon1987@gmail.com, E-mail: dora@nuclear.ufmg.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear; Instituto Nacional de Ciencias e Tecnologia de Reatores Nucleares Inovadores (INCT/CNPq), Belo Horizonte (Brazil); Miro, Rafael; Verdu, Gumersindo, E-mail: rmiro@iqn.upv.es, E-mail: gverdu@iqn.upv.es [Universidad Politecnica de Valencia (Spain). Departamento de Ingenieria Quimica y Nuclear
2015-07-01
The computational advances observed in the last two decades have been provided direct impact on the researches related to nuclear simulations, which use several types of computer codes, including coupled between them, allowing representing with very accuracy the behavior of nuclear plants. Studies of complex scenarios in nuclear reactors have been improved by the use of thermal-hydraulic (TH) and neutron kinetics (NK) coupled codes. This technique consists in incorporating three-dimensional (3D) neutron modeling of the reactor core into codes, mainly to simulate transients that involve asymmetric core spatial power distributions and strong feedback effects between neutronics and reactor thermal-hydraulics. Therefore, this work presents preliminary results of TH RELAP5 and the NK PARCS calculations applied to model of the Angra 2 reactor. The WIMSD-5B code has been used to generate the macroscopic cross sections used in the NK code. The results obtained are satisfactory and represent important part of the development of this methodology. The next step is to couple the codes. (author)
Whole core neutronics modeling of a TRIGA reactor using integral transport theory
International Nuclear Information System (INIS)
Schwinkendorf, K.N.; Toffer, H.
1990-01-01
An innovative analysis approach for performing whole core reactor physics calculations for TRIGA reactors has been employed recently at the Westinghouse Hanford Company. A deterministic transport theory model with sufficient geometric complexity to evaluate asymmetric loading patterns was used. Calculations of this complexity have been performed in the past using Monte Carlo simulation, such as the MCNP code. However, the Monte Carlo calculations are more difficult to prepare and require more computer time. On the Hanford Site CRAY XMP-18 computer, the new methods required less than one-third of the central processing unit time per calculation as compared to an MCNP calculation using 100,000 neutron histories
Simulation of neutron transport process, photons and charged particles within the Monte Carlo method
International Nuclear Information System (INIS)
Androsenko, A.A.; Androsenko, P.A.; Artamonov, S.N.; Bolonkina, G.V.; Lomtev, V.L.; Pupko, S.V.
1991-01-01
Description is given to the program system BRAND designed for the accurate solution of non-stationary transport equation of neutrons, photons and charged particles in the conditions of real three-dimensional geometry. An extensive set of local and non-local estimates provides an opportunity of calculating a great set of linear functionals normally being of interest in the calculation of reactors, radiation protection and experiment simulation. The process of particle interaction with substance is simulated on the basis of individual non-group data on each isotope of the composition. 24 refs
International Nuclear Information System (INIS)
Goncalves, G.A.; Bogado Leite, S.Q.; Vilhena, M.T. de
2009-01-01
An analytical solution has been obtained for the one-speed stationary neutron transport problem, in an infinitely long cylinder with anisotropic scattering by the decomposition method. Series expansions of the angular flux distribution are proposed in terms of suitably constructed functions, recursively obtainable from the isotropic solution, to take into account anisotropy. As for the isotropic problem, an accurate closed-form solution was chosen for the problem with internal source and constant incident radiation, obtained from an integral transformation technique and the F N method
Benchmark results for the critical slab and sphere problem in one-speed neutron transport theory
International Nuclear Information System (INIS)
Rawat, Ajay; Mohankumar, N.
2011-01-01
Research highlights: → The critical slab and sphere problem in neutron transport under Case eigenfunction formalism is considered. → These equations reduce to integral expressions involving X functions. → Gauss quadrature is not ideal but DE quadrature is well-suited. → Several fold decrease in computational effort with improved accuracy is realisable. - Abstract: In this paper benchmark numerical results for the one-speed criticality problem with isotropic scattering for the slab and sphere are reported. The Fredholm integral equations of the second kind based on the Case eigenfunction formalism are numerically solved by Neumann iterations with the Double Exponential quadrature.
Adjacent-cell Preconditioners for solving optically thick neutron transport problems
International Nuclear Information System (INIS)
Azmy, Y.Y.
1994-01-01
We develop, analyze, and test a new acceleration scheme for neutron transport methods, the Adjacent-cell Preconditioner (AP) that is particularly suited for solving optically thick problems. Our method goes beyond Diffusion Synthetic Acceleration (DSA) methods in that it's spectral radius vanishes with increasing cell thickness. In particular, for the ID case the AP method converges immediately, i.e. in one iteration, to 10 -4 pointwise relative criterion in problems with dominant cell size of 10 mfp or thicker. Also the AP has a simple formalism and is cell-centered hence, multidimensional and high order extensions are easier to develop, and more efficient to implement
Impact of pore size variability and network coupling on electrokinetic transport in porous media
Alizadeh, Shima; Bazant, Martin Z.; Mani, Ali
2016-11-01
We have developed and validated an efficient and robust computational model to study the coupled fluid and ion transport through electrokinetic porous media, which are exposed to external gradients of pressure, electric potential, and concentration. In our approach a porous media is modeled as a network of many pores through which the transport is described by the coupled Poisson-Nernst-Planck-Stokes equations. When the pore sizes are random, the interactions between various modes of transport may provoke complexities such as concentration polarization shocks and internal flow circulations. These phenomena impact mixing and transport in various systems including deionization and filtration systems, supercapacitors, and lab-on-a-chip devices. In this work, we present simulations of massive networks of pores and we demonstrate the impact of pore size variation, and pore-pore coupling on the overall electrokinetic transport in porous media.
Latkowski, J F; Sanz, J
2000-01-01
Recent modifications to the TART Monte Carlo neutron and photon transport code allow enable calculation of 566-group neutron spectra. This expanded group structure represents a significant improvement over the 50- and 175-group structures that have been previously available. To support use of this new capability, neutron activation cross-section libraries have been created in the 175- and 566-group structures starting from the FENDL/A-2.0 pointwise data. Neutron spectra have been calculated for the first walls of the HYLIFE-II and Sombrero inertial fusion energy power plant designs and have been used in subsequent neutron activation calculations. The results obtained using the two different group structures are compared with each other as well as to those obtained using a 175-group version of the EAF3.1 activation cross-section library.
Žukauskaitėa, A; Plukienė, R; Ridikas, D
2007-01-01
Particle accelerators and other high energy facilities produce penetrating ionizing radiation (neutrons and γ-rays) that must be shielded. The objective of this work was to model photon and neutron transport in various materials, usually used as shielding, such as concrete, iron or graphite. Monte Carlo method allows obtaining answers by simulating individual particles and recording some aspects of their average behavior. In this work several nuclear experiments were modeled: AVF 65 (AVF cyclotron of Research Center of Nuclear Physics, Osaka University, Japan) – γ-ray beams (1-10 MeV), HIMAC (heavy-ion synchrotron of the National Institute of Radiological Sciences in Chiba, Japan) and ISIS-800 (ISIS intensive spallation neutron source facility of the Rutherford Appleton laboratory, UK) – high energy neutron (20-800 MeV) transport in iron and concrete. The calculation results were then compared with experimental data.compared with experimental data.
Guskov, Albert; Jensen, Sonja; Faustino, Ignacio; Marrink, Siewert J.; Slotboom, Dirk Jan
2016-01-01
Glutamate transporters catalyse the thermodynamically unfavourable transport of anionic amino acids across the cell membrane by coupling it to the downhill transport of cations. This coupling mechanism is still poorly understood, in part because the available crystal structures of these transporters
Preliminary radiation transport analysis for the proposed National Spallation Neutron Source (NSNS)
International Nuclear Information System (INIS)
Johnson, J.O.; Lillie, R.A.
1997-01-01
The use of neutrons in science and industry has increased continuously during the past 50 years with applications now widely used in physics, chemistry, biology, engineering, and medicine. Within this history, the relative merits of using pulsed accelerator spallation sources versus reactors for neutron sources as the preferred option for the future. To address this future need, the Department of Energy (DOE) has initiated a pre-conceptual design study for the National Spallation Neutron Source (NSNS) and given preliminary approval for the proposed facility to be built at Oak Ridge National Laboratory (ORNL). The DOE directive is to design and build a short pulse spallation source in the 1 MS power range with sufficient design flexibility that it can be upgraded and operated at a significantly higher power at a later stage. The pre-conceptualized design of the NSNS initially consists of an accelerator system capable of delivering a 1 to 2 GeV proton beam with 1 MW of beam power in an approximate 0.5 microsecond pulse at a 60 Hz frequency onto a single target station. The NSNS will be upgraded in stages to a 5 MW facility with two target stations (a high power station operating at 60 Hz and a low power station operating at 10 Hz). Each target station will contain four moderators (combinations of cryogenic and ambient temperature) and 18 beam liens for a total of 36 experiment stations. This paper summarizes the radiation transport analysis strategies for the proposed NSNS facility
Effect of high fluence neutron irradiation on transport properties of thermoelectrics
Wang, H.; Leonard, K. J.
2017-07-01
Thermoelectric materials were subjected to high fluence neutron irradiation in order to understand the effect of radiation damage on transport properties. This study is relevant to the NASA Radioisotope Thermoelectric Generator (RTG) program in which thermoelectric elements are exposed to radiation over a long period of time in space missions. Selected n-type and p-type bismuth telluride materials were irradiated at the High Flux Isotope Reactor with a neutron fluence of 1.3 × 1018 n/cm2 (E > 0.1 MeV). The increase in the Seebeck coefficient in the n-type material was partially off-set by an increase in electrical resistivity, making the power factor higher at lower temperatures. For the p-type materials, although the Seebeck coefficient was not affected by irradiation, electrical resistivity decreased slightly. The figure of merit, zT, showed a clear drop in the 300-400 K range for the p-type material and an increase for the n-type material. Considering that the p-type and n-type materials are connected in series in a module, the overall irradiation damages at the device level were limited. These results, at neutron fluences exceeding a typical space mission, are significant to ensure that the radiation damage to thermoelectrics does not affect the performance of RTGs.
Controllable single-photon transport between remote coupled-cavity arrays
Qin, Wei; Nori, Franco
2015-01-01
We develop a new approach for controllable single-photon transport between two remote one-dimensional coupled-cavity arrays, used as quantum registers, mediated by an additional one-dimensional coupled-cavity array, acting as a quantum channel. A single two-level atom located inside one cavity of the intermediate channel is used to control the long-range coherent quantum coupling between two remote registers, thereby functioning as a quantum switch. With a time-independent perturbative treatm...
International Nuclear Information System (INIS)
Vondy, D.R.; Fowler, T.B.; Cunningham, G.W.
1977-11-01
The report documents the computer code block VENTURE designed to solve multigroup neutronics problems with application of the finite-difference diffusion-theory approximation to neutron transport (or alternatively simple P 1 ) in up to three-dimensional geometry. It uses and generates interface data files adopted in the cooperative effort sponsored by the Reactor Physics Branch of the Division of Reactor Research and Development of the Energy Research and Development Administration. Several different data handling procedures have been incorporated to provide considerable flexibility; it is possible to solve a wide variety of problems on a variety of computer configurations relatively efficiently
Energy Technology Data Exchange (ETDEWEB)
Vondy, D.R.; Fowler, T.B.; Cunningham, G.W.
1977-11-01
The report documents the computer code block VENTURE designed to solve multigroup neutronics problems with application of the finite-difference diffusion-theory approximation to neutron transport (or alternatively simple P/sub 1/) in up to three-dimensional geometry. It uses and generates interface data files adopted in the cooperative effort sponsored by the Reactor Physics Branch of the Division of Reactor Research and Development of the Energy Research and Development Administration. Several different data handling procedures have been incorporated to provide considerable flexibility; it is possible to solve a wide variety of problems on a variety of computer configurations relatively efficiently.
International Nuclear Information System (INIS)
Vondy, D.R.; Fowler, T.B.; Cunningham, G.W.
1975-10-01
The computer code block VENTURE, designed to solve multigroup neutronics problems with application of the finite-difference diffusion-theory approximation to neutron transport (or alternatively simple P 1 ) in up to three-dimensional geometry is described. A variety of types of problems may be solved: the usual eigenvalue problem, a direct criticality search on the buckling, on a reciprocal velocity absorber (prompt mode), or on nuclide concentrations, or an indirect criticality search on nuclide concentrations, or on dimensions. First-order perturbation analysis capability is available at the macroscopic cross section level
Temporal moment analysis of solute transport in a coupled fracture ...
Indian Academy of Sciences (India)
precipitation of quartz in a single fracture. Later, Sekhar et al (2006) and Suresh Kumar (2008) have investigated the effect of linear and non-linear sorption on dispersivity and macro-dispersion coefficient in a coupled fracture-matrix system.
Keeley, N.; Mackintosh, R. S.
2018-01-01
Background: Precise fitting of scattering observables suggests that the nucleon-nucleus interaction is l dependent. Such l dependence has been shown to be S -matrix equivalent to an undulatory l -independent potential. The undulations include radial regions where the imaginary term is emissive. Purpose: To study the dynamical polarization potential (DPP) generated in proton-16O and neutron-16O interaction potentials by coupling to pickup channels. Undulatory features occurring in these DPPs can be compared with corresponding features of empirical optical model potentials (OMPs). Furthermore, the additional inclusion of coupling to vibrational states of the target will provide evidence for dynamically generated nonlocality. Methods: The fresco code provides the elastic channel S -matrix Sl j for chosen channel couplings. Inversion, Sl j→V (r ) +l .s VSO(r ) , followed by subtraction of the bare potential, yields an l -independent and local representation of the DPP due to the chosen couplings. Results: The DPPs have strongly undulatory features, including radial regions of emissivity. Certain features of empirical DPPs appear, e.g., the full inverted potential has emissive regions. The DPPs for different collective states are additive except near the nuclear center, whereas the collective and reaction channel DPPs are distinctly nonadditive over a considerable radial range, indicating dynamical nonlocality. Substantial differences between the DPPs due to pickup coupling for protons and neutrons occur; these imply a greater difference between proton and neutron OMPs than the standard phenomenological prescription. Conclusions: The onus is on those who object to undularity in the local and l -independent representation of nucleon elastic scattering to show why such undulations do not occur. This work suggests that it is not legitimate to halt model-independent fits to high-quality data at the appearance of undularity.
Crystal Structure of a Phosphorylation-coupled Saccharide Transporter
Energy Technology Data Exchange (ETDEWEB)
Y Cao; X Jin; E Levin; H Huang; Y Zong; W Hendrickson; J Javitch; K Rajashankar; M Zhou; et al.
2011-12-31
Saccharides have a central role in the nutrition of all living organisms. Whereas several saccharide uptake systems are shared between the different phylogenetic kingdoms, the phosphoenolpyruvate-dependent phosphotransferase system exists almost exclusively in bacteria. This multi-component system includes an integral membrane protein EIIC that transports saccharides and assists in their phosphorylation. Here we present the crystal structure of an EIIC from Bacillus cereus that transports diacetylchitobiose. The EIIC is a homodimer, with an expansive interface formed between the amino-terminal halves of the two protomers. The carboxy-terminal half of each protomer has a large binding pocket that contains a diacetylchitobiose, which is occluded from both sides of the membrane with its site of phosphorylation near the conserved His250 and Glu334 residues. The structure shows the architecture of this important class of transporters, identifies the determinants of substrate binding and phosphorylation, and provides a framework for understanding the mechanism of sugar translocation.
Geometry and transport in a model of two coupled quadratic nonlinear waveguides
DEFF Research Database (Denmark)
Stirling, James R.; Bang, Ole; Christiansen, Peter Leth
2008-01-01
This paper applies geometric methods developed to understand chaos and transport in Hamiltonian systems to the study of power distribution in nonlinear waveguide arrays. The specific case of two linearly coupled X(2) waveguides is modeled and analyzed in terms of transport and geometry in the pha...
International Nuclear Information System (INIS)
Pop-Jordanov, J.; Bosevski, T.; Kocic, A.; Altiparmakov, D.
1980-01-01
A Space-Point Energy-Group integral transport theory method (SPEG) is developed and applied to the local and global calculations of the Yugoslav RA reactor. Compared to other integral transport theory methods, the SPEG distinguishes by (1) the arbitrary order of the polynomial, (2) the effective determination of integral parameters through point flux values, (3) the use of neutron balance condition. as a posterior measure of the accuracy of the calculation and (4) the elimination of the subdivisions- into zones, in realistic cases. In addition, different direct (collision probability) and indirect (Monte Carlo) approaches to integral transport theory have been investigated and Some effective acceleration procedures introduced. The study was performed on three test problems in plane and cylindrical geometry, as well as on the nine-region cell of the RA reactor. In particular, the limitations of the integral transport theory including its non-applicability to optically large material regions and to global reactor calculations were examined. The proposed strictly multipoint approach, avoiding the subdivision into zones and groups, seems to provide a good starting point to overcome these limitations of the integral transport theory. (author)
An analytic solution to the critical problem of neutron transport in plane geometry
International Nuclear Information System (INIS)
Coppa, G.; Ravetto, P.; Sumini, M.
1987-01-01
The linear transport equation in slab geometry is given an analytic solution by means of a series expansion of the unknown, using Helmholtz eigenfunctions, and suitably introducing a space independent term which can account for vacuum boundary conditions. A second-order form of the transport equation is taken into consideration. Only materially homogeneous systems showing isotropic scattering properties are investigated and the absence of external sources is supposed. An exact infinite system of equations for the coefficients of the expansion is derived. The truncation of the series leads to approximations, some numerical results of which are presented. The total neutron current and its relationship to the total flux, together with the presence of a zero-gradient current, will also be discussed. (orig.) [de
Sparks, Samuel; Temel, Deniz B; Rout, Michael P; Cowburn, David
2018-03-06
The largely intrinsically disordered phenylalanine-glycine-rich nucleoporins (FG Nups) underline a selectivity mechanism that enables the rapid translocation of transport factors (TFs) through the nuclear pore complexes (NPCs). Conflicting models of NPC transport have assumed that FG Nups undergo different conformational transitions upon interacting with TFs. To selectively characterize conformational changes in FG Nups induced by TFs we performed small-angle neutron scattering (SANS) with contrast matching. Conformational-ensembles derived from SANS data indicated an increase in the overall size of FG Nups is associated with TF interaction. Moreover, the organization of the FG motif in the interacting state is consistent with prior experimental analyses defining that FG motifs undergo conformational restriction upon interacting with TFs. These results provide structural insights into a highly dynamic interaction and illustrate how functional disorder imparts rapid and selective FG Nup-TF interactions. Copyright © 2018 Elsevier Ltd. All rights reserved.
Low-energy beam transport studies supporting the spallation neutron source 1-MW beam operation.
Han, B X; Kalvas, T; Tarvainen, O; Welton, R F; Murray, S N; Pennisi, T R; Santana, M; Stockli, M P
2012-02-01
The H(-) injector consisting of a cesium enhanced RF-driven ion source and a 2-lens electrostatic low-energy beam transport (LEBT) system supports the spallation neutron source 1 MW beam operation with ∼38 mA beam current in the linac at 60 Hz with a pulse length of up to ∼1.0 ms. In this work, two important issues associated with the low-energy beam transport are discussed: (1) inconsistent dependence of the post-radio frequency quadrupole accelerator beam current on the ion source tilt angle and (2) high power beam losses on the LEBT electrodes under some off-nominal conditions compromising their reliability.
Application of the finite element method to the neutron transport equation
International Nuclear Information System (INIS)
Martin, W.R.
1976-01-01
This paper examines the theoretical and practical application of the finite element method to the neutron transport equation. It is shown that in principle the system of equations obtained by application of the finite element method can be solved with certain physical restrictions concerning the criticality of the medium. The convergence of this approximate solution to the exact solution with mesh refinement is examined, and a non-optical estimate of the convergence rate is obtained analytically. It is noted that the numerical results indicate a faster convergence rate and several approaches to obtain this result analytically are outlined. The practical application of the finite element method involved the development of a computer code capable of solving the neutron transport equation in 1-D plane geometry. Vacuum, reflecting, or specified incoming boundary conditions may be analyzed, and all are treated as natural boundary conditions. The time-dependent transport equation is also examined and it is shown that the application of the finite element method in conjunction with the Crank-Nicholson time discretization method results in a system of algebraic equations which is readily solved. Numerical results are given for several critical slab eigenvalue problems, including anisotropic scattering, and the results compare extremely well with benchmark results. It is seen that the finite element code is more efficient than a standard discrete ordinates code for certain problems. A problem with severe heterogeneities is considered and it is shown that the use of discontinuous spatial and angular elements results in a marked improvement in the results. Finally, time-dependent problems are examined and it is seen that the phenomenon of angular mode separation makes the numerical treatment of the transport equation in slab geometry a considerable challenge, with the result that the angular mesh has a dominant effect on obtaining acceptable solutions
A Coupled Chemical and Mass Transport Model for Concrete Durability
DEFF Research Database (Denmark)
Jensen, Mads Mønster; Johannesson, Björn; Geiker, Mette Rica
2012-01-01
by the rate of mass transport only. A consequence of the source or sink term, is the assumption that equilibrium is reached instantaneously in each time step considered. Some numerical problems was found, where the residual requirements for the chemical equilibrium was not reached. Small imbalances, in e...
Inter-dot coupling effects on transport through correlated parallel ...
Indian Academy of Sciences (India)
a substantial attention for theoretical [13–17] and experimental [18,19] mesoscopic transport phenomena. Being more complicated than series CQD, most of the pre- vious theoretical work on parallel CQD is either in the presence of magnetic flux. [13,14] or in equilibrium condition [17] or is devoted to the interference effects.
Progress on RMC: a Monte Carlo neutron transport code for reactor analysis
International Nuclear Information System (INIS)
Wang, Kan; Li, Zeguang; She, Ding; Liu, Yuxuan; Xu, Qi; Shen, Huayun; Yu, Ganglin
2011-01-01
This paper presents a new 3-D Monte Carlo neutron transport code named RMC (Reactor Monte Carlo code), specifically intended for reactor physics analysis. This code is being developed by Department of Engineering Physics in Tsinghua University and written in C++ and Fortran 90 language with the latest version of RMC 2.5.0. The RMC code uses the method known as the delta-tracking method to simulate neutron transport, the advantages of which include fast simulation in complex geometries and relatively simple handling of complicated geometrical objects. Some other techniques such as computational-expense oriented method and hash-table method have been developed and implemented in RMC to speedup the calculation. To meet the requirements of reactor analysis, the RMC code has the calculational functions including criticality calculation, burnup calculation and also kinetics simulation. In this paper, comparison calculations of criticality problems, burnup problems and transient problems are carried out using RMC code and other Monte Carlo codes, and the results show that RMC performs quite well in these kinds of problems. Based on MPI, RMC succeeds in parallel computation and represents a high speed-up. This code is still under intensive development and the further work directions are mentioned at the end of this paper. (author)
International Nuclear Information System (INIS)
Karlsson, J.K.H.; Linden, P.
1997-01-01
The neutron transport in a bare, cylindrical and homogeneous reactor, with and without the presence of a central partially inserted control rod, has been simulated by using a Monte Carlo transport code. The behaviour of both the flux and current in this system have been investigated. We have found that the flux and especially the current are strongly affected by the presence of the control rod in its close vicinity. The results indicate the feasibility to identify the position and especially the tip of the rod from the flux and current. Further, the direction to the rod can be found from the current vector. The information content regarding the position of the rod, in both the neutron flux and the current, decays strongly as a function of distance and it is dependent on the size of the rod. In our model, the practical range over which the flux or current can be a useful indicator of the position of the tip of the rod is about 10-15 cm for a rod with a diameter of 2 cm. The practical range for identification of the position of the rod is greater for a rod of larger diameter
An analytical approach for a nodal scheme of two-dimensional neutron transport problems
International Nuclear Information System (INIS)
Barichello, L.B.; Cabrera, L.C.; Prolo Filho, J.F.
2011-01-01
Research highlights: → Nodal equations for a two-dimensional neutron transport problem. → Analytical Discrete Ordinates Method. → Numerical results compared with the literature. - Abstract: In this work, a solution for a two-dimensional neutron transport problem, in cartesian geometry, is proposed, on the basis of nodal schemes. In this context, one-dimensional equations are generated by an integration process of the multidimensional problem. Here, the integration is performed for the whole domain such that no iterative procedure between nodes is needed. The ADO method is used to develop analytical discrete ordinates solution for the one-dimensional integrated equations, such that final solutions are analytical in terms of the spatial variables. The ADO approach along with a level symmetric quadrature scheme, lead to a significant order reduction of the associated eigenvalues problems. Relations between the averaged fluxes and the unknown fluxes at the boundary are introduced as the usually needed, in nodal schemes, auxiliary equations. Numerical results are presented and compared with test problems.
International Nuclear Information System (INIS)
Delfin L, A.
1996-01-01
The purpose of this work is to solve the neutron transport equation in discrete-ordinates and X-Y geometry by developing and using the strong discontinuous and strong modified discontinuous nodal finite element schemes. The strong discontinuous and modified strong discontinuous nodal finite element schemes go from two to ten interpolation parameters per cell. They are describing giving a set D c and polynomial space S c corresponding for each scheme BDMO, RTO, BL, BDM1, HdV, BDFM1, RT1, BQ and BDM2. The solution is obtained solving the neutron transport equation moments for each nodal scheme by developing the basis functions defined by Pascal triangle and the Legendre moments giving in the polynomial space S c and, finally, looking for the non singularity of the resulting linear system. The linear system is numerically solved using a computer program for each scheme mentioned . It uses the LU method and forward and backward substitution and makes a partition of the domain in cells. The source terms and angular flux are calculated, using the directions and weights associated to the S N approximation and solving the angular flux moments to find the effective multiplication constant. The programs are written in Fortran language, using the dynamic allocation of memory to increase efficiently the available memory of the computing equipment. (Author)
N. Natarajan, G. Suresh Kumar
2011-01-01
A numerical model is developed for studying the transport of colloid facilitated radionuclide transport in a coupled fracture-matrix system. The radionuclides and the colloids are assumed to decay, sorb on the fracture surface, as well as diffuse into the rock matrix. The sorption of the radionuclides onto the mobile and immobile colloids within the fracture is assumed to be linear. The governing equations describing the radionuclide and colloidal transport along the fracture axis and diffusi...
Axial SPN and radial MOC coupled whole core transport calculation
International Nuclear Information System (INIS)
Cho, Jin-Young; Kim, Kang-Seog; Lee, Chung-Chan; Zee, Sung-Quun; Joo, Han-Gyu
2007-01-01
The Simplified P N (SP N ) method is applied to the axial solution of the two-dimensional (2-D) method of characteristics (MOC) solution based whole core transport calculation. A sub-plane scheme and the nodal expansion method (NEM) are employed for the solution of the one-dimensional (1-D) SP N equations involving a radial transverse leakage. The SP N solver replaces the axial diffusion solver of the DeCART direct whole core transport code to provide more accurate, transport theory based axial solutions. In the sub-plane scheme, the radial equivalent homogenization parameters generated by the local MOC for a thick plane are assigned to the multiple finer planes in the subsequent global three-dimensional (3-D) coarse mesh finite difference (CMFD) calculation in which the NEM is employed for the axial solution. The sub-plane scheme induces a much less nodal error while having little impact on the axial leakage representation of the radial MOC calculation. The performance of the sub-plane scheme and SP N nodal transport solver is examined by solving a set of demonstrative problems and the C5G7MOX 3-D extension benchmark problems. It is shown in the demonstrative problems that the nodal error reaching upto 1,400 pcm in a rodded case is reduced to 10 pcm by introducing 10 sub-planes per MOC plane and the transport error is reduced from about 150 pcm to 10 pcm by using SP 3 . Also it is observed, in the C5G7MOX rodded configuration B problem, that the eigenvalues and pin power errors of 180 pcm and 2.2% of the 10 sub-planes diffusion case are reduced to 40 pcm and 1.4%, respectively, for SP 3 with only about a 15% increase in the computing time. It is shown that the SP 5 case gives very similar results to the SP 3 case. (author)
A methodology for the coupling of RAMONA-3B neutron kinetics and TRAC-BF1 thermal-hydraulics
International Nuclear Information System (INIS)
Lopez, Arsenio Procopio; Morales Sandoval, Jaime B.
2005-01-01
The initial objective of this project was to directly couple the RAMONA and TRAC codes running on different PCs. The idea was to use the best part of each one and eliminate some of their limitations and widen the applicability of these codes to simulate different BWR and system components. It was required to try to minimize the amount of changes to present code subroutines and calculation procedures. If possible, just substitute values obtained in the parallel code. Preliminary results indicated that using a CHAN component of TRAC to model thermal-hydraulic phenomena for each neutronic channel modeled in RAMONA is rather difficult. Large amounts of CPU time consumption are obtained and lots of PCs would make this solution difficult, besides considerable large transients are introduced by the differences in thermal-hydraulic results of these codes. The substitution of the thermal-hydraulics of RAMONA, by the TRAC channel calculations, is possible but simulation of a null transient on both codes must be planed and a gradual change must be controlled by an additional supervisory subroutine. An indirect coupling of these codes, it is therefore proposed, in order to eliminate most of these limitations. In this indirect coupling, a thermal-hydraulic model of the average tube in a bundle and the global channel cooling fluid dynamics is programmed for each neutronic channel while the global reactor vessel and core is modeled by TRAC with just four channels and four rings. Results are more reliable, coupling is simpler and faster simulations are possible
DEFF Research Database (Denmark)
Larsen, Erik Hviid; Sørensen, Jakob Balslev; Sørensen, Jens Nørkær
2000-01-01
A mathematical model of an absorbing leaky epithelium is developed for analysis of solute coupled water transport. The non-charged driving solute diffuses into cells and is pumped from cells into the lateral intercellular space (lis). All membranes contain water channels with the solute passing...... increases with hydraulic conductance of the pathway carrying water from mucosal solution into lis. Uphill water transport is accomplished, but with high hydraulic conductance of cell membranes strength of transport is obscured by water flow through cells. Anomalous solvent drag occurs when back flux...
International Nuclear Information System (INIS)
1996-01-01
A - Nature of problem or function: DOT solves the Boltzmann transport equation in two-dimensional geometries. Principal applications are to neutron and/or photon transport, although the code can be applied to transport problems for any particles not subject to external force fields. Both homogeneous and external-source problems can be solved. Searches on multiplication factor, time absorption, nuclide concentration, and zone thickness are available for reactor problems. Numerous edits and output data sets for subsequent use are available. DOT-3.5 improves the space-scaling algorithm. DOT-3.5/CAB contains group by group UPSCATTER scaling method. DUCT calculates perturbations to the scalar flux caused by the presence of ducts filled with coolant. VIP is a program for cross section sensitivity analysis using two- dimensional discrete ordinates transport calculations. DGRAD calculates the directional flux gradients from DOT-3 diffusion theory flux tapes. In conjunction with VIP and TPERT, it allows the use of diffusion theory fluxes to obtain exact and first-order perturbation reactivity changes. In order to calculate the reactivity associated with changes in reactor compositions using diffusion theory, it is necessary to fold not only the scalar fluxes with the appropriate cross sections, but also the average flux gradients with the diffusion coefficients. Since DOT diffusion theory does not directly calculate these gradients, it was necessary to calculate the needed quantities external to the DOT code. TPERT is a perturbation code to obtain exact and first-order reactivity changes. TPERT is coupled to VIP which generates adjoint forward flux tables using DOT-3 scalar flux tape information. GRTUNCL calculates an analytical first-collision source for subsequent use in DOT. B - Method of solution: The method of discrete ordinates is used. Balance equations are solved for the density of particles moving along discrete directions in each cell of a two-dimensional spatial
Application of preconditioned GMRES to the numerical solution of the neutron transport equation
International Nuclear Information System (INIS)
Patton, B.W.; Holloway, J.P.
2002-01-01
The generalized minimal residual (GMRES) method with right preconditioning is examined as an alternative to both standard and accelerated transport sweeps for the iterative solution of the diamond differenced discrete ordinates neutron transport equation. Incomplete factorization (ILU) type preconditioners are used to determine their effectiveness in accelerating GMRES for this application. ILU(τ), which requires the specification of a dropping criteria τ, proves to be a good choice for the types of problems examined in this paper. The combination of ILU(τ) and GMRES is compared with both DSA and unaccelerated transport sweeps for several model problems. It is found that the computational workload of the ILU(τ)-GMRES combination scales nonlinearly with the number of energy groups and quadrature order, making this technique most effective for problems with a small number of groups and discrete ordinates. However, the cost of preconditioner construction can be amortized over several calculations with different source and/or boundary values. Preconditioners built upon standard transport sweep algorithms are also evaluated as to their effectiveness in accelerating the convergence of GMRES. These preconditioners show better scaling with such problem parameters as the scattering ratio, the number of discrete ordinates, and the number of spatial meshes. These sweeps based preconditioners can also be cast in a matrix free form that greatly reduces storage requirements
Cynod: A Neutronics Code for Pebble Bed Modular Reactor Coupled Transient Analysis
Energy Technology Data Exchange (ETDEWEB)
Hikaru Hiruta; Abderrafi M. Ougouag; Hans D. Gougar; Javier Ortensi
2008-09-01
The Pebble Bed Reactor (PBR) is one of the two concepts currently considered for development into the Next Generation Nuclear Plant (NGNP). This interest is due, in particular, to the concept’s inherent safety characteristics. In order to verify and confirm the design safety characteristics of the PBR computational tools must be developed that treat the range of phenomena that are expected to be important for this type of reactors. This paper presents a recently developed 2D R-Z cylindrical nodal kinetics code and shows some of its capabilities by applying it to a set of known and relevant benchmarks. The new code has been coupled to the thermal hydraulics code THERMIX/KONVEK[1] for application to the simulation of very fast transients in PBRs. The new code, CYNOD, has been written starting with a fixed source solver extracted from the nodal cylindrical geometry solver contained within the PEBBED code. The fixed source solver was then incorporated into a kinetic solver.. The new code inherits the spatial solver characteristics of the nodal solver within PEBBED. Thus, the time-dependent neutron diffusion equation expressed analytically in each node of the R-Z cylindrical geometry sub-domain (or node) is transformed into one-dimensional equations by means of the usual transverse integration procedure. The one-dimensional diffusion equations in each of the directions are then solved using the analytic Green’s function method. The resulting equations for the entire domain are then re-cast in the form of the Direct Coarse Mesh Finite Difference (D-CMFD) for convenience of solution. The implicit Euler method is used for the time variable discretization. In order to correctly treat the cusping effect for nodes that contain a partially inserted control rod a method is used that takes advantage of the Green’s function solution available in the intrinsic method. In this corrected treatment, the nodes are re-homogenized using axial flux shapes reconstructed based on the
International Nuclear Information System (INIS)
Kodeli, I.; Diop, C.M.; Nimal, J.C.
1994-01-01
In the framework of future Boron Neutron Capture Therapy (BNCT) experiments, where cells and animals irradiations are planned at the research reactor of Strasbourg University, the feasibility to obtain a suitable epithermal neutron beam is investigated. The neutron fluence and spectra calculations in the reactor are performed using the 3D Monte Carlo code TRIPOLI-3 and the 2D SN code TWODANT. The preliminary analysis of Al 2 O 3 and Al-Al 2 O 3 filters configurations are carried out in an attempt to optimize the flux characteristics in the beam tube facility. 7 figs., 7 refs
Chan, Cheong Xin; Zäuner, Simone; Wheeler, Glen; Grossman, Arthur R; Prochnik, Simon E; Blouin, Nicolas A; Zhuang, Yunyun; Benning, Christoph; Berg, Gry Mine; Yarish, Charles; Eriksen, Renée L; Klein, Anita S; Lin, Senjie; Levine, Ira; Brawley, Susan H; Bhattacharya, Debashish
2012-04-01
Membrane transporters play a central role in many cellular processes that rely on the movement of ions and organic molecules between the environment and the cell, and between cellular compartments. Transporters have been well characterized in plants and green algae, but little is known about transporters or their evolutionary histories in the red algae. Here we examined 482 expressed sequence tag contigs that encode putative membrane transporters in the economically important red seaweed Porphyra (Bangiophyceae, Rhodophyta). These contigs are part of a comprehensive transcriptome dataset from Porphyra umbilicalis and Porphyra purpurea. Using phylogenomics, we identified 30 trees that support the expected monophyly of red and green algae/plants (i.e. the Plantae hypothesis) and 19 expressed sequence tag contigs that show evidence of endosymbiotic/horizontal gene transfer involving stramenopiles. The majority (77%) of analyzed contigs encode transporters with unresolved phylogenies, demonstrating the difficulty in resolving the evolutionary history of genes. We observed molecular features of many sodium-coupled transport systems in marine algae, and the potential for coregulation of Porphyra transporter genes that are associated with fatty acid biosynthesis and intracellular lipid trafficking. Although both the tissue-specific and subcellular locations of the encoded proteins require further investigation, our study provides red algal gene candidates associated with transport functions and novel insights into the biology and evolution of these transporters.
Energy Technology Data Exchange (ETDEWEB)
Moraes, Pedro Gabriel B.; Leite, Michel C.A.; Barros, Ricardo C., E-mail: pgbmoraes@gmail.com, E-mail: chell_leite@hotmail.com, E-mail: rcbarros@pq.cnpq.br [Universidade do Estado do Rio de Janeiro (UERJ), Nova Friburgo, RJ (Brazil). Instituto Politecnico. Departamento de Modelagem Computacional
2013-07-01
In this work we developed a software to model and generate results in tables and graphs of one-dimensional neutron transport problems in multi-group formulation of energy. The numerical method we use to solve the problem of neutron diffusion is analytic, thus eliminating the truncation errors that appear in classical numerical methods, e.g., the method of finite differences. This numerical analytical method increases the computational efficiency, since they are not refined spatial discretization necessary because for any spatial discretization grids used, the numerical result generated for the same point of the domain remains unchanged unless the rounding errors of computational finite arithmetic. We chose to develop a computational application in MatLab platform for numerical computation and program interface is simple and easy with knobs. We consider important to model this neutron transport problem with a fixed source in the context of shielding calculations of radiation that protects the biosphere, and could be sensitive to ionizing radiation.
Computing and the electrical transport properties of coupled quantum networks
Cain, Casey Andrew
In this dissertation a number of investigations were conducted on ballistic quantum networks in the mesoscopic range. In this regime, the wave nature of electron transport under the influence of transverse magnetic fields leads to interesting applications for digital logic and computing circuits. The work specifically looks at characterizing a few main areas that would be of interest to experimentalists who are working in nanostructure devices, and is organized as a series of papers. The first paper analyzes scaling relations and normal mode charge distributions for such circuits in both isolated and open (terminals attached) form. The second paper compares the flux-qubit nature of quantum networks to the well-established spintronics theory. The results found exactly contradict the conventional school of thought for what is required for quantum computation. The third paper investigates the requirements and limitations of extending the Thevenin theorem in classic electric circuits to ballistic quantum transport. The fourth paper outlines the optimal functionally complete set of quantum circuits that can completely satisfy all sixteen Boolean logic operations for two variables.
Energy Technology Data Exchange (ETDEWEB)
Jaeger, W.; Sanchez, V.; Cheng, X. [Karlsruhe Institute of Technology (KIT), Eggenstein-Leopoldshafen (Germany). Inst. for Neutron Physics and Reactor Technology; Monti, L. [Karlsruhe Institute of Technology (KIT), Eggenstein-Leopoldshafen (Germany). Inst. for Nuclear and Energy Technologies; Hurtado, A. [Technical Univ. of Dresden (Germany). Inst. of Power Engineering
2011-07-01
At the Institute for Neutron Physics and Reactor Technology (INR) at the Karlsruhe Institute of Technology (KIT), the development and validation of coupled codes systems is one major activity. In this paper, a 2-step method is proposed to perform uncertainty and sensitivity analysis of a nuclear fuel bundle. At first, the SUSA package (Software system for Uncertainty and Sensitivity Analysis), 2 is applied to the thermal hydraulic results of the TRACE (TRACE/RELAP Advanced Computational Engine) code to identify crucial thermal hydraulic parameter combinations which are successively used in the TH/NP coupled system TRACEERANOS to account for the neutronic feedbacks. This 2-step method was applied since the TRACE-ERANOS system runs 1 input in approximately 1 day (depending on the computer configurations). Since the uncertainty and sensitivity analysis requires about 100 runs of the thermal hydraulic input (with altered parameters, running within minutes) an integral TRACE-SUSA-ERANOS analysis would need around 100 days. For this analysis a fuel assembly model of the HPLWR (High Performance Light Water Reactor) was selected. Due to the general structure of the coupling and code communication scripts, the system can be used for any kind of reactor/system which can be described with TRACE and ERANOS (e.g., fast systems) while SUSA can be applied to anything. (orig.)
International Nuclear Information System (INIS)
Matausek, M.
1972-01-01
A new proposed method for solving the space-energy dependent spherical harmonics equations represents a methodological contribution to neutron transport theory. The proposed method was applied for solving the problem of spec-energy transport of fast and resonance neutrons in multi-zone, cylindrical y symmetric infinite reactor cell and is related to previously developed procedure for treating the thermal energy region. The advantages of this method are as follows: a unique algorithm was obtained for detailed determination of spatial and energy distribution of neutrons (from thermal to fast) in the reactor cell; these detailed distributions enable more precise calculations of criticality conditions, obtaining adequate multigroup data and better interpretation of experimental data; computing time is rather short
Beam transient analyses of Accelerator Driven Subcritical Reactors based on neutron transport method
Energy Technology Data Exchange (ETDEWEB)
He, Mingtao; Wu, Hongchun [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an 710049, Shaanxi (China); Zheng, Youqi, E-mail: yqzheng@mail.xjtu.edu.cn [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an 710049, Shaanxi (China); Wang, Kunpeng [Nuclear and Radiation Safety Center, PO Box 8088, Beijing 100082 (China); Li, Xunzhao; Zhou, Shengcheng [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an 710049, Shaanxi (China)
2015-12-15
Highlights: • A transport-based kinetics code for Accelerator Driven Subcritical Reactors is developed. • The performance of different kinetics methods adapted to the ADSR is investigated. • The impacts of neutronic parameters deteriorating with fuel depletion are investigated. - Abstract: The Accelerator Driven Subcritical Reactor (ADSR) is almost external source dominated since there is no additional reactivity control mechanism in most designs. This paper focuses on beam-induced transients with an in-house developed dynamic analysis code. The performance of different kinetics methods adapted to the ADSR is investigated, including the point kinetics approximation and space–time kinetics methods. Then, the transient responds of beam trip and beam overpower are calculated and analyzed for an ADSR design dedicated for minor actinides transmutation. The impacts of some safety-related neutronics parameters deteriorating with fuel depletion are also investigated. The results show that the power distribution varying with burnup leads to large differences in temperature responds during transients, while the impacts of kinetic parameters and feedback coefficients are not very obvious. Classification: Core physic.
Neutron slowing down and transport in a medium of constant cross section. I. Spatial moments
International Nuclear Information System (INIS)
Cacuci, D.G.; Goldstein, H.
1977-01-01
Some aspects of the problem of neutron slowing down and transport have been investigated in an infinite medium consisting of a single nuclide scattering elastically and isotropically without absorption and with energy-independent cross sections. The method of singular eigenfunctions has been applied to the Boltzmann equation governing the Laplace transform (with respect to the lethargy variable) of the neutron flux. Formulas have been obtained for the lethargy dependent spatial moments of the scalar flux applicable in the limit of large lethargy. In deriving these formulas, use has been made of the well-known connection between the spatial moments of the Laplace-transformed scalar flux and the moments of the flux in the ''eigenvalue space.'' The calculations have been greatly aided by the construction of a closed general expression for these ''eigenvalue space'' moments. Extensive use has also been made of the methods of combinatorial analysis and of computer evaluation, via FORMAC, of complicated sequences of manipulations. It has been possible to obtain for materials of any atomic weight explicit corrections to the age theory formulas for the spatial moments M/sub 2n/(u), of the scalar flux, valid through terms of order of u -5 . Higher order correction terms could be obtained at the expense of additional computer time. The evaluation of the coefficients of the powers of n, as explicit functions of the nuclear mass, represent the end product of this investigation
Using the transportable, computer-operated, liquid-scintillator fast-neutron spectrometer system
International Nuclear Information System (INIS)
Thorngate, J.H.
1988-01-01
When a detailed energy spectrum is needed for radiation-protection measurements from approximately 1 MeV up to several tens of MeV, organic-liquid scintillators make good neutron spectrometers. However, such a spectrometer requires a sophisticated electronics system and a computer to reduce the spectrum from the recorded data. Recently, we added a Nuclear Instrument Module (NIM) multichannel analyzer and a lap-top computer to the NIM electronics we have used for several years. The result is a transportable fast-neutron spectrometer system. The computer was programmed to guide the user through setting up the system, calibrating the spectrometer, measuring the spectrum, and reducing the data. Measurements can be made over three energy ranges, 0.6--2 MeV, 1.1--8 MeV, or 1.6--16 MeV, with the spectrum presented in 0.1-MeV increments. Results can be stored on a disk, presented in a table, and shown in graphical form. 5 refs., 51 figs
An optimized ultra-fine energy group structure for neutron transport calculations
International Nuclear Information System (INIS)
Huria, Harish; Ouisloumen, Mohamed
2008-01-01
This paper describes an optimized energy group structure that was developed for neutron transport calculations in lattices using the Westinghouse lattice physics code PARAGON. The currently used 70-energy group structure results in significant discrepancies when the predictions are compared with those from the continuous energy Monte Carlo methods. The main source of the differences is the approximations employed in the resonance self-shielding methodology. This, in turn, leads to ambiguous adjustments in the resonance range cross-sections. The main goal of developing this group structure was to bypass the self-shielding methodology altogether thereby reducing the neutronic calculation errors. The proposed optimized energy mesh has 6064 points with 5877 points spanning the resonance range. The group boundaries in the resonance range were selected so that the micro group cross-sections matched reasonably well with those derived from reaction tallies of MCNP for a number of resonance absorbers of interest in reactor lattices. At the same time, however, the fast and thermal energy range boundaries were also adjusted to match the MCNP reaction rates in the relevant ranges. The resulting multi-group library was used to obtain eigenvalues for a wide variety of reactor lattice numerical benchmarks and also the Doppler reactivity defect benchmarks to establish its adequacy. (authors)
International Nuclear Information System (INIS)
Hammer, C.; Paffrath, M.; Boeer, R.; Finnemann, H.; Jackson, C.J.
1996-01-01
The light water reactor core simulation code PANBOX has been coupled with the transient analysis code RELAP5 for the purpose of performing plant safety analyses with a three-dimensional (3-D) neutron kinetics model. The system has been parallelized to improve the computational efficiency. The paper describes the features of this system with emphasis on performance aspects. Performance results are given for different types of parallelization, i. e. for using an automatic parallelizing compiler, using the portable PVM platform on a workstation cluster, using PVM on a shared memory multiprocessor, and for using machine dependent interfaces. (author)
International Nuclear Information System (INIS)
Fontana, A; Rossi, F; Viliani, G; Caponi, S; Fabiani, E; Baldi, G; Ruocco, G; Dal Maschio, R
2007-01-01
We report new inelastic Raman and neutron scattering spectra for glasses with different degrees of fragility, v-SiO 2 , v-GeO 2 (AgI) 0.5 (Ag 2 O-B 2 O 3 ) 0.5 (AgI) x (AgPO 3 ) 1-x . The data are compared for each sample to obtain the Raman coupling function C(ω). The study indicates a general linear behaviour of C(ω) near the boson peak maximum, and evidences a correlation between vibrational and relaxational properties, confirming the results of recent publications
Spectrum of the multigroup neutron transport operator for bounded spatial domains
International Nuclear Information System (INIS)
Larsen, E.W.
1979-01-01
The spectrum of the multigroup neutron transport operator A is studied for bounded spatial regions D which consist of a finite number of material subregions. Our main results provide simple conditions on the material cross sections which guarantee that (1) A possesses eigenvalues in the finite plane; (2) A possesses a ''leading'' eigenvalue lambda 0 which is real, not less than the real part of any other eigenvalue, and to which there corresponds at least one nonnegative eigenfunction psi/sub lambda/0; and (3) A possesses a ''dominant'' eigenvalue lambda 0 which is real, simple, greater than the real part of any other eigenvalue, and whose eigenfunction psi/sub lambda/0 satisfies psi/sub lambda/0> or =0 and ∫psi/sub lambda/0d 2 Ω>0. We give examples to illustrate the results and to show that a leading eigenvalue need not be simple, nor its eigenfunction(s) positive
Finite element analysis of the neutron transport equation in spherical geometry
International Nuclear Information System (INIS)
Kim, Yong Ill; Kim, Jong Kyung; Suk, Soo Dong
1992-01-01
The Galerkin formulation of the finite element method is applied to the integral law of the first-order form of the one-group neutron transport equation in one-dimensional spherical geometry. Piecewise linear or quadratic Lagrange polynomials are utilized in the integral law for the angular flux to establish a set of linear algebraic equations. Numerical analyses are performed for the scalar flux distribution in a heterogeneous sphere as well as for the criticality problem in a uniform sphere. For the criticality problems in the uniform sphere, the results of the finite element method, with the use of continuous finite elements in space and angle, are compared with the exact solutions. In the heterogeneous problem, the scalar flux distribution obtained by using discontinuous angular and spatical finite elements is in good agreement with that from the ANISN code calculation. (Author)
A massively parallel discrete ordinates response matrix method for neutron transport
International Nuclear Information System (INIS)
Hanebutte, U.R.; Lewis, E.E.
1992-01-01
In this paper a discrete ordinates response matrix method is formulated with anisotropic scattering for the solution of neutron transport problems on massively parallel computers. The response matrix formulation eliminates iteration on the scattering source. The nodal matrices that result from the diamond-differenced equations are utilized in a factored form that minimizes memory requirements and significantly reduces the number of arithmetic operations required per node. The red-black solution algorithm utilizes massive parallelism by assigning each spatial node to one or more processors. The algorithm is accelerated by a synthetic method in which the low-order diffusion equations are also solved by massively parallel red-black iterations. The method is implemented on a 16K Connection Machine-2, and S 8 and S 16 solutions are obtained for fixed-source benchmark problems in x-y geometry
The Development of 3D Graphics for Simple Implementation of Photon and Neutron Transport Code
International Nuclear Information System (INIS)
Siangsanan, P.
2014-01-01
The Simple Implementation of Photon and Neutron Transport code (SIPHON) was developed and tested at Office of Atoms for Peace around 1998 using nuclear data from MCNP code. The input of SIPHON is in the form of text file so that user could set the dimension of simulation model with accuracy. Whereas the code can check the correctness of geometry of the model during running time, the point of error will be found only if a simulated particle has crossed the erratic geometry and might take a lot of time to be found in a very complex system. The three-dimensional graphical view was implemented into SIPHON to solve this problem and was found later that it is also useful in educational purpose.
GPU-based high performance Monte Carlo simulation in neutron transport
International Nuclear Information System (INIS)
Heimlich, Adino; Mol, Antonio C.A.; Pereira, Claudio M.N.A.
2009-01-01
Graphics Processing Units (GPU) are high performance co-processors intended, originally, to improve the use and quality of computer graphics applications. Since researchers and practitioners realized the potential of using GPU for general purpose, their application has been extended to other fields out of computer graphics scope. The main objective of this work is to evaluate the impact of using GPU in neutron transport simulation by Monte Carlo method. To accomplish that, GPU- and CPU-based (single and multicore) approaches were developed and applied to a simple, but time-consuming problem. Comparisons demonstrated that the GPU-based approach is about 15 times faster than a parallel 8-core CPU-based approach also developed in this work. (author)
GPU-based high performance Monte Carlo simulation in neutron transport
Energy Technology Data Exchange (ETDEWEB)
Heimlich, Adino; Mol, Antonio C.A.; Pereira, Claudio M.N.A. [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil). Lab. de Inteligencia Artificial Aplicada], e-mail: cmnap@ien.gov.br
2009-07-01
Graphics Processing Units (GPU) are high performance co-processors intended, originally, to improve the use and quality of computer graphics applications. Since researchers and practitioners realized the potential of using GPU for general purpose, their application has been extended to other fields out of computer graphics scope. The main objective of this work is to evaluate the impact of using GPU in neutron transport simulation by Monte Carlo method. To accomplish that, GPU- and CPU-based (single and multicore) approaches were developed and applied to a simple, but time-consuming problem. Comparisons demonstrated that the GPU-based approach is about 15 times faster than a parallel 8-core CPU-based approach also developed in this work. (author)
The solution of the multigroup neutron transport equation using spherical harmonics
International Nuclear Information System (INIS)
Fletcher, K.
1981-01-01
A solution of the multi-group neutron transport equation in up to three space dimensions is presented. The flux is expanded in a series of unnormalised spherical harmonics. Using the various recurrence formulae a linked set of first order differential equations is obtained for the moments psisup(g)sub(lm)(r), γsup(g)sub(lm)(r). Terms with odd l are eliminated resulting in a second order system which is solved by two methods. The first is a finite difference formulation using an iterative procedure, secondly, in XYZ and XY geometry a finite element solution is given. Results for a test problem using both methods are exhibited and compared. (orig./RW) [de
Analysis of EBR-II neutron and photon physics by multidimensional transport-theory techniques
International Nuclear Information System (INIS)
Jacqmin, R.P.; Finck, P.J.; Palmiotti, G.
1994-01-01
This paper contains a review of the challenges specific to the EBR-II core physics, a description of the methods and techniques which have been developed for addressing these challenges, and the results of some validation studies relative to power-distribution calculations. Numerical tests have shown that the VARIANT nodal code yields eigenvalue and power predictions as accurate as finite difference and discrete ordinates transport codes, at a small fraction of the cost. Comparisons with continuous-energy Monte Carlo results have proven that the errors introduced by the use of the diffusion-theory approximation in the collapsing procedure to obtain broad-group cross sections, kerma factors, and photon-production matrices, have a small impact on the EBR-II neutron/photon power distribution
International Nuclear Information System (INIS)
Ganapol, B.D.
1986-01-01
In a course on neutron transport theory and also in the analytical neutron transport theory literature, the pioneering work of Case et al. (CdHP) is often referenced. This work was truly a monumental effort in that it treated the fundamental mathematical properties of the one-group neutron Boltzmann equation in detail as well as the numerical evaluation of most of the resulting solutions. Many mathematically and numerically oriented dissertations were based on this classic monograph. In light of the considerable advances made both in numerical methods and computer technology since 1953, when the historic CdHP monograph first appeared, it seems appropriate to reevaluate the numerical benchmark solutions found therein with present-day computational technology. In most transport theory courses, the subject of proper benchmarking of numerical algorithms and transport codes is seldom addressed at any great length. This may be the reason that the benchmarking procedure is so rarely practiced in the nuclear community and when practiced is improperly applied. In this presentation, the development of a new benchmark for the one-group neutron flux in an infinite medium will be detailed with emphasis placed on the educational aspects of the benchmarking activity
International Nuclear Information System (INIS)
Williams, M.M.R.
1985-01-01
A multigroup formalism is developed for the backward-forward-isotropic scattering model of neutron transport. Some exact solutions are obtained in two-group theory for slab and spherical geometry. The results are useful for benchmark problems involving multigroup anisotropic scattering. (author)
International Nuclear Information System (INIS)
Ahnert, C.; Aragones, J. M.
1981-01-01
This Is a users manual of the neutron transport code TWOTRAN-TRACA, which is a version of the original TWOTRAN-GG from the Los Alamos Laboratory, with some modifications made at JEN. A detailed input data description is given as well as the new modifications developed at JEN. (Author) 8 refs
Chung, Eric
2015-12-11
In this paper, we develop a mass conservative multiscale method for coupled flow and transport in heterogeneous porous media. We consider a coupled system consisting of a convection-dominated transport equation and a flow equation. We construct a coarse grid solver based on the Generalized Multiscale Finite Element Method (GMsFEM) for a coupled system. In particular, multiscale basis functions are constructed based on some snapshot spaces for the pressure and the concentration equations and some local spectral decompositions in the snapshot spaces. The resulting approach uses a few multiscale basis functions in each coarse block (for both the pressure and the concentration) to solve the coupled system. We use the mixed framework, which allows mass conservation. Our main contributions are: (1) the development of a mass conservative GMsFEM for the coupled flow and transport; (2) the development of a robust multiscale method for convection-dominated transport problems by choosing appropriate test and trial spaces within Petrov-Galerkin mixed formulation. We present numerical results and consider several heterogeneous permeability fields. Our numerical results show that with only a few basis functions per coarse block, we can achieve a good approximation.
Effects of coupled thermal, hydrological and chemical processes on nuclide transport
International Nuclear Information System (INIS)
Carnahan, C.L.
1987-03-01
Coupled thermal, hydrological and chemical processes can be classified in two categories. One category consists of the ''Onsager'' type of processes driven by gradients of thermodynamic state variables. These processes occur simultaneously with the direct transport processes. In particular, thermal osmosis, chemical osmosis and ultrafiltration may be prominent in semipermeable materials such as clays. The other category consists of processes affected indirectly by magnitudes of thermodynamic state variables. An important example of this category is the effect of temperature on rates of chemical reactions and chemical equilibria. Coupled processes in both categories may affect transport of radionuclides. Although computational models of limited extent have been constructed, there exists no model that accounts for the full set of THC-coupled processes. In the category of Onsager coupled processes, further model development and testing is severely constrained by a deficient data base of phenomenological coefficients. In the second category, the lack of a general description of effects of heterogeneous chemical reactions on permeability of porous media inhibits progress in quantitative modeling of hydrochemically coupled transport processes. Until fundamental data necessary for further model development have been acquired, validation efforts will be limited necessarily to testing of incomplete models of nuclide transport under closely controlled experimental conditions. 34 refs., 2 tabs
Edge and coupled core-edge transport modelling in tokamaks
International Nuclear Information System (INIS)
Lodestro, L.L.; Casper, T.A.; Cohen, R.H.
2001-01-01
Recent advances in the theory and modelling of tokamak edge, scrape-off-layer (SOL) and divertor plasmas are described. The effects of the poloidal ExB drift on inner/outer divertor-plate asymmetries within a 1D analysis are shown to be in good agreement with experimental trends; above a critical v ExB, the model predicts transitions to supersonic SOL flow at the inboard midplane. 2D simulations show the importance of ExB flow in the private-flux region and of ∇ B-drifts. A theory of rough plasma-facing surfaces is given, predicting modifications to the SOL plasma. The parametric dependence of detached-plasma states in slab geometry has been explored; with sufficient pumping, the location of the ionization front can be controlled; otherwise only fronts near the plate or the X-point are stable. Studies with a more accurate Monte-Carlo neutrals model and a detailed non-LTE radiation-transport code indicate various effects are important for quantitative modelling. Detailed simulations of the DIII-D core and edge are presented; impurity and plasma flow are discussed and shown to be well modelled with UEDGE. (author)
Edge and coupled core/edge transport modelling in tokamaks
International Nuclear Information System (INIS)
Lodestro, L.L.; Casper, T.A.; Cohen, R.H.
1999-01-01
Recent advances in the theory and modelling of tokamak edge, scrape-off-layer (SOL) and divertor plasmas are described. The effects of the poloidal E x B drift on inner/outer divertor-plate asymmetries within a 1D analysis are shown to be in good agreement with experimental trends; above a critical v ExB , the model predicts transitions to supersonic flow at the inboard midplane. 2D simulations show the importance of E x B flow in the private-flux region and of ∇ B-drifts. A theory of rough plasma-facing surfaces is given, predicting modifications to the SOL plasma. The parametric dependence of detached-plasma states in slab geometry has been explored; with sufficient pumping, the location of the ionization front can be controlled; otherwise only fronts near the plate or the X-point are stable. Studies with a more accurate Monte-Carlo neutrals model and a detailed non-LTE radiation-transport code indicate various effects are important for quantitative modelling. Detailed simulations of the DIII-D core and edge are presented; impurity and plasma flow are discussed and shown to be well modelled with UEDGE. (author)
International Nuclear Information System (INIS)
Apperson, C.E. Jr.
1981-01-01
A method is presented for studying the influence of fission product transpot on delayed neutron precursors and decay heat sources during Liquid Metal Fast Breeder Reactor (LMFBR) unprotected accidents. The model represents the LMFBR core as a closed homogeneous cell. Thermodynamic phase equilibrium theory is used to predict fission product mobility. Reactor kinetics behavior is analyzed by an extension of point kinetics theory. Group dependent delayed neutron precursor and decay heat source retention factors, which represent the fraction of each group retained in the fuel, are developed to link the kinetics and thermodynamics analysis. Application of the method to a highly simplified model of an unprotected loss-of-flow accident shows a time delay on the order of 10 ms is introduced in the predisassembly power history if fission product motion is considered when compared to the traditional transient solution. The post-transient influence of fission product transport calculated by the present model is a 24 percent reduction in the decay heat level in the fuel material which is similar to traditional approximations. Isotopes of the noble gases, Kr and Xe, and the elements I and Br are shown to be very mobile and are responsible for a major part of the observed effects. Isotopes of the elements Cs, Se, Rb, and Te were found to be moderately mobile and contribute to a lesser extent to the observed phenomena. These results obtained from the application of the described model confirm the initial hypothesis that sufficient fission product transport can occur to influence a transient. For these reasons, it is concluded that extension of this model into a multi-cell transient analysis code is warranted
CO2-ECBM related coupled physical and mechanical transport processes
Gensterblum, Yves; Satorius, Michael; Busch, Andreas; Krooß, Bernhard
2013-04-01
The interrelation of cleat transport processes and mechanical properties was investigated by permeability tests at different stress levels (60% to 130% of in-situ stress) with sorbing (CH4, CO2) and inert gases (N2, Ar, He) on a sub bituminous A coal from the Surat Basin, Queensland Australia. From the flow tests under controlled triaxial stress conditions the Klinkenberg-corrected "true" permeability coefficients and the Klinkenberg slip factors were derived. The "true"-, absolute or Klinkenberg corrected permeability shows a gas type dependence. Following the approach of Seidle et al. (1992) the cleat volume compressibility (cf) was calculated from observed changes in apparent permeability upon variation of external stress (at equal mean gas pressures). The observed effects also show a clear dependence on gas type. Due to pore or cleat compressibility the cleat aperture decreases with increasing effective stress. Vice versa we observe with increasing mean pressure at lower confining pressure an increase in permeability which we attribute to a cleat aperture widening. The cleat volume compressibility (cf) also shows a dependence on the mean pore pressure. Non-sorbing gases like helium and argon show higher apparent permeabilities than sorbing gases like methane. Permeability coefficients measured with successively increasing mean gas pressures were consistently lower than those determined at decreasing mean gas pressures. This permeability hysteresis is in accordance with results reported by Harpalani and McPherson (1985). The kinetics of matrix transport processes were studied by sorption tests on different particle sizes at various moisture contents and temperatures (cf. Busch et al., 2006). Methane uptake rates were determined from the pressure decline curves recorded for each particle-size fraction, and "diffusion coefficients" were calculated using several unipore and bidisperse diffusion models. While the CH4 sorption capacity of moisture-equilibrated coals
Neutron electric dipole moment from supersymmetric anomalous W-boson coupling
International Nuclear Information System (INIS)
Kadoyoshi, T.; Oshimo, N.
1997-01-01
In the supersymmetric standard model (SSM) the W boson could have a nonvanishing electric dipole moment (EDM) through a one-loop diagram mediated by the charginos and neutralinos. Then the W-boson EDM induces the EDMs of the neutron and the electron. We discuss these EDMs, taking into consideration the constraints from the neutron and electron EDMs at the one-loop level induced by the charginos and squarks or sleptons. It is shown that the neutron and the electron could, respectively, have EDMs of the order of 10 -26 ecm and 10 -27 ecm, solely owing to the W-boson EDM. Since these EDMs do not depend on the values of the SSM parameters for the squark or slepton sector, they provide less ambiguous predictions for CP violation in the SSM. copyright 1997 The American Physical Society
The role of Rashba spin-orbit coupling in valley-dependent transport of Dirac fermions
Energy Technology Data Exchange (ETDEWEB)
Hasanirok, Kobra; Mohammadpour, Hakimeh
2017-01-01
At this work, spin- and valley-dependent electron transport through graphene and silicene layers are studied in the presence of Rashba spin- orbit coupling. We find that the transport properties of the related ferromagnetic/normal/ferromagnetic structure depend on the relevant parameters. A fully valley- and spin- polarized current is obtained. As another result, Rashba spin-orbit interaction plays important role in controlling the transmission characteristics.
Directory of Open Access Journals (Sweden)
Surian Pinem
2014-01-01
Full Text Available A coupled neutronics thermal-hydraulics code NODAL3 has been developed based on the few-group neutron diffusion equation in 3-dimensional geometry for typical PWR static and transient analyses. The spatial variables are treated by using a polynomial nodal method while for the neutron dynamic solver the adiabatic and improved quasistatic methods are adopted. In this paper we report the benchmark calculation results of the code against the OECD/NEA CRP PWR rod ejection cases. The objective of this work is to determine the accuracy of NODAL3 code in analysing the reactivity initiated accident due to the control rod ejection. The NEACRP PWR rod ejection cases are chosen since many organizations participated in the NEA project using various methods as well as approximations, so that, in addition to the reference solutions, the calculation results of NODAL3 code can also be compared to other codes’ results. The transient parameters to be verified are time of power peak, power peak, final power, final average Doppler temperature, maximum fuel temperature, and final coolant temperature. The results of NODAL3 code agree well with the PHANTHER reference solutions in 1993 and 1997 (revised. Comparison with other validated codes, DYN3D/R and ANCK, shows also a satisfactory agreement.
Zhang, Jinglong; Moore, Michael E; Wang, Zhonghai; Rong, Zhou; Yang, Chaowen; Hayward, Jason P
2017-10-01
Choosing a digitizer with an appropriate sampling rate is often a trade-off between performance and economy. The influence of sampling rates on the neutron-gamma Pulse Shape Discrimination (PSD) with a solid stilbene scintillator coupled to a Silicon Photomultiplier was investigated in this work. Sampling rates from 125MSPS to 2GSPS from a 10-bit digitizer were used to collect detector pulses produced by the interactions of a Cf-252 source. Due to the decreased signal-to-noise ratio (SNR), the PSD performance degraded with reduced sampling rates. The reason of PSD performance degradation was discussed. Then, an efficient combination of filtering and digital signal processing (DSP) was then applied to suppress the timing noise and electronic background noise. The results demonstrate an improved PSD performance especially at low sampling rates, down to 125MSPS. Using filtering and DSP, the ascribed Figure of Merit (FOM) at 125keV ee (± 10keV ee ) increased from 0.95 to 1.02 at 125MSPS. At 300keV ee and above, all the FOMs are better than 2.00. Our study suggests that 250MSPS is a good enough sampling rate for neutron-gamma discrimination in this system in order to be sensitive to neutrons at and above ~ 125keV ee . Copyright © 2017 Elsevier Ltd. All rights reserved.
International Nuclear Information System (INIS)
Staalek, Mathias
2008-03-01
Coupled calculations are important for the simulation of nuclear power plants when there is a strong feedback between the neutron kinetics and the thermal-hydraulics. A general coupled model of the Ringhals-3 Pressurized Water Reactor has been developed for this purpose. The development is outlined in the thesis with details given in the appended papers. A PARCS model was developed for the core calculations and a RELAP5 model for the thermal-hydraulic calculations. The RELAP5 model has 157 channels for modelling the flow in the fuel assemblies. This means that there is a one-one correspondence radially between the neutronic and thermal-hydraulic nodalization. This detailed mapping between the neutron kinetics and the thermal-hydraulics makes it possible to use the model for all kinds of transient. To provide realistic material data to the PARCS model, a cross-section interface was developed. With this interface one can import material data from a binary CASMO-4 library file into PARCS. Due to the one-to-one mapping, any any core loading can easily be considered. The PARCS model was benchmarked against measurements of the steady-state power distribution of Ringhals-3. The power shape was well reproduced by the model. Validational work for steady-state conditions of the thermal-hydraulic was also successfully performed. The most challenging part of the validation of a coupled model is for transients. This is much more difficult since the dynamics of the system becomes very important. Two transients that occurred at Ringhals-3 were chosen for the validational work. The first transient was a Load Rejection Transient. In general the model gave good results but some problems were experienced, e.g. the pressurizer pressure turned out to be more difficult to be correctly simulated. The second transient was a Loss of Feed Water transient. A malfunctioning feed water control valve closed, and therefore shut down the feed water supply to the steam generator in one of the
Energy Technology Data Exchange (ETDEWEB)
Schunert, Sebastian; Schwen, Daniel; Ghassemi, Pedram; Baker, Benjamin; Zabriskie, Adam; Ortensi, Javier; Wang, Yaqi; Gleicher, Frederick; DeHart, Mark; Martineau, Richard
2017-04-01
This work presents a multi-physics, multi-scale approach to modeling the Transient Test Reactor (TREAT) currently prepared for restart at the Idaho National Laboratory. TREAT fuel is made up of microscopic fuel grains (r ˜ 20µm) dispersed in a graphite matrix. The novelty of this work is in coupling a binary collision Monte-Carlo (BCMC) model to the Finite Element based code Moose for solving a microsopic heat-conduction problem whose driving source is provided by the BCMC model tracking fission fragment energy deposition. This microscopic model is driven by a transient, engineering scale neutronics model coupled to an adiabatic heating model. The macroscopic model provides local power densities and neutron energy spectra to the microscpic model. Currently, no feedback from the microscopic to the macroscopic model is considered. TREAT transient 15 is used to exemplify the capabilities of the multi-physics, multi-scale model, and it is found that the average fuel grain temperature differs from the average graphite temperature by 80 K despite the low-power transient. The large temperature difference has strong implications on the Doppler feedback a potential LEU TREAT core would see, and it underpins the need for multi-physics, multi-scale modeling of a TREAT LEU core.
Flow resistance of orifices and spacers of BWR thermal-hydraulic and neutronic coupling loop
International Nuclear Information System (INIS)
Iguchi, Tadashi; Asaka, Hideaki; Nakamura, Hideo
2002-03-01
Authors are performing THYNC experiments to study thermal-hydraulic instability under neutronic and thermal-hydraulic coupling. In THYNC experiments, the orifices are installed at the exit of the test section and the spacers are installed in the test section, in order to properly simulate in-core thermal-hydraulics in the reactor core. It is necessary to know the flow resistance of the orifices and spacers for the analysis of THYNC experimental results. Consequently, authors measured the flow resistance of orifice and spacer under single-phase and two-phase flows. Using the experimental results, authors investigated the dependency of the flow resistances on the parameters, such as pressure, mass flux, an geometries. Furthermore, authors investigated the applicability of the basic two-phase flow models, for example the separate flow model, to the two-phase flow multiplier. As the result of the investigation on the single-phase flow experiment, it was found (1) that the effects of pressure and mass flux flow resistance are described by a function of Reynolds number, and (2) that flow resistances of the orifice and the spacer are calculated with the previous prediction methods. However, it was necessary to introduce an empirical coefficient, since it was difficult to predict accurately the flow resistance only with the previous prediction method due to the complicated geometry dependency, for example a flow area blockage ratio. On the other hand, according to the investigation on two-phase flow experiment, the followings were found. (1) Relation between the two-phase flow multiplier and the quality is regarded to be linear under pressure of 2MPa - 7MPa. The relation is dependent on pressure and geometry, and is little dependent on mass flux. (2) Relation between the two-phase flow multiplier and void fraction is little dependent on pressure, mass flux, and geometry under pressure of 0.2MPa - 7MPa and void fraction less than 0.6. The relation is less dependent on
PAD: a one-dimensional, coupled neutronic-thermodynamic-hydrodynamic computer code
International Nuclear Information System (INIS)
Peterson, D.M.; Stratton, W.R.; McLaughlin, T.P.
1976-12-01
Theoretical and numerical foundations, utilization guide, sample problems, and program listing and glossary are given for the PAD computer code which describes dynamic systems with interactive neutronics, thermodynamics, and hydrodynamics in one-dimensional spherical, cylindrical, and planar geometries. The code has been applied to prompt critical excursions in various fissioning systems (solution, metal, LMFBR, etc.) as well as to nonfissioning systems
International Nuclear Information System (INIS)
Ackroyd, R.T.
1982-01-01
Some minimum and maximum variational principles for even-parity neutron transport are reviewed and the corresponding principles for odd-parity transport are derived by a simple method to show why the essential boundary conditions associated with these maximum principles have to be imposed. The method also shows why both the essential and some of the natural boundary conditions associated with these minimum principles have to be imposed. These imposed boundary conditions for trial functions in the variational principles limit the choice of the finite element used to represent trial functions. The reasons for the boundary conditions imposed on the principles for even- and odd-parity transport point the way to a treatment of composite neutron transport, for which completely boundary-free maximum and minimum principles are derived from a functional identity. In general a trial function is used for each parity in the composite neutron transport, but this can be reduced to one without any boundary conditions having to be imposed. (author)
Gravitational effects on planetary neutron flux spectra
Feldman, W. C.; Drake, D. M.; O'Dell, R. D.; Brinkley, F. W., Jr.; 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.
Schmeeckle, M. W.; Leary, K. P.
2016-12-01
We investigate the spatiotemporal coupling of sediment transport over dunes using a turbulence- and particle-resolving numerical model and high-speed video in a laboratory flume. The model utilizes the Large Eddy Simulation (LES) for the fluid turbulence and a Discrete Element Method (DEM) simulation for the sediment. Previous experiments assessing the effects of flow separation on downstream fluid turbulent structures and bedload transport suggest that localized, intermittent, high-magnitude transport events, called permeable splat events, play an important role in both downstream and cross-stream transport near flow reattachment. The flume was lined with 17 concrete ripples that had a 2 cm high crest and were 30 cm long. A high-speed camera observed sediment transport along the entirety of the bedform at 250 Hz. Downstream and vertical fluid velocity was observed at 1mm and 3 mm above the bed using Laser Doppler Velocitmetry (LDV) at 15 distances along bedform profile. As observed in our previous backward-facing step experiments and simulations, mean downstream fluid velocity increases nonlinearly with increasing distance along the ripple. Observed sediment transport, however, increases linearly with increasing distance along the ripple with an exception at the crest of the bedform, where both mean downstream fluid velocity and sediment transport decrease significantly. Previous experiments assessing only the effect of flow separation showed that calculating sediment transport as a function of boundary shear stress using a Meyer-Peter Müller type equation, produced a zone of underestimated transport near flow reattachment. Results reported here show that calculating sediment transport in this way underestimates observed sediment transport along the entire profile of the bedform, not just near flow reattachment. Preliminary sediment transport time-series data show a zone of high-magnitude cross-stream transport near flow reattachment, suggesting that permeable
On the adequacy of message-passing parallel supercomputers for solving neutron transport problems
International Nuclear Information System (INIS)
Azmy, Y.Y.
1990-01-01
A coarse-grained, static-scheduling parallelization of the standard iterative scheme used for solving the discrete-ordinates approximation of the neutron transport equation is described. The parallel algorithm is based on a decomposition of the angular domain along the discrete ordinates, thus naturally producing a set of completely uncoupled systems of equations in each iteration. Implementation of the parallel code on Intcl's iPSC/2 hypercube, and solutions to test problems are presented as evidence of the high speedup and efficiency of the parallel code. The performance of the parallel code on the iPSC/2 is analyzed, and a model for the CPU time as a function of the problem size (order of angular quadrature) and the number of participating processors is developed and validated against measured CPU times. The performance model is used to speculate on the potential of massively parallel computers for significantly speeding up real-life transport calculations at acceptable efficiencies. We conclude that parallel computers with a few hundred processors are capable of producing large speedups at very high efficiencies in very large three-dimensional problems. 10 refs., 8 figs
Generalized Coarse-Mesh Rebalance Method for Acceleration of Neutron Transport Calculations
International Nuclear Information System (INIS)
Yamamoto, Akio
2005-01-01
This paper proposes a new acceleration method for neutron transport calculations: the generalized coarse-mesh rebalance (GCMR) method. The GCMR method is a unified scheme of the traditional coarse-mesh rebalance (CMR) and the coarse-mesh finite difference (CMFD) acceleration methods. Namely, by using an appropriate acceleration factor, formulation of the GCMR method becomes identical to that of the CMR or CMFD method. This also indicates that the convergence property of the GCMR method can be controlled by the acceleration factor since the convergence properties of the CMR and CMFD methods are generally different. In order to evaluate the convergence property of the GCMR method, a linearized Fourier analysis was carried out for a one-group homogeneous medium, and the results clarified the relationship between the acceleration factor and the spectral radius. It was also shown that the spectral radius of the GCMR method is smaller than those of the CMR and CMFD methods. Furthermore, the Fourier analysis showed that when an appropriate acceleration factor was used, the spectral radius of the GCMR method did not exceed unity in this study, which was in contrast to the results of the CMR or the CMFD method. Application of the GCMR method to practical calculations will be easy when the CMFD acceleration is already adopted in a transport code. By multiplying a suitable acceleration factor to a coefficient (D FD ) of a finite difference formulation, one can improve the numerical instability of the CMFD acceleration method
The TORT three-dimensional discrete ordinates neutron/photon transport code (TORT version 3)
Energy Technology Data Exchange (ETDEWEB)
Rhoades, W.A.; Simpson, D.B.
1997-10-01
TORT calculates the flux or fluence of neutrons and/or photons throughout three-dimensional systems due to particles incident upon the system`s external boundaries, due to fixed internal sources, or due to sources generated by interaction with the system materials. The transport process is represented by the Boltzman transport equation. The method of discrete ordinates is used to treat the directional variable, and a multigroup formulation treats the energy dependence. Anisotropic scattering is treated using a Legendre expansion. Various methods are used to treat spatial dependence, including nodal and characteristic procedures that have been especially adapted to resist numerical distortion. A method of body overlay assists in material zone specification, or the specification can be generated by an external code supplied by the user. Several special features are designed to concentrate machine resources where they are most needed. The directional quadrature and Legendre expansion can vary with energy group. A discontinuous mesh capability has been shown to reduce the size of large problems by a factor of roughly three in some cases. The emphasis in this code is a robust, adaptable application of time-tested methods, together with a few well-tested extensions.
International Nuclear Information System (INIS)
Mugica R, C.A.; Valle G, E. del
2005-01-01
In 2002, E. del Valle and Ernest H. Mund developed a technique to solve numerically the Neutron transport equations in discrete ordinates and hexagonal geometry using two nodal schemes type finite element weakly discontinuous denominated WD 5,3 and WD 12,8 (of their initials in english Weakly Discontinuous). The technique consists on representing each hexagon in the union of three rhombuses each one of which it is transformed in a square in the one that the methods WD 5,3 and WD 12,8 were applied. In this work they are solved the mentioned equations of transport using the same discretization technique by hexagon but using two nodal schemes type finite element strongly discontinuous denominated SD 3 and SD 8 (of their initials in english Strongly Discontinuous). The application in each case as well as a reference problem for those that results are provided for the effective multiplication factor is described. It is carried out a comparison with the obtained results by del Valle and Mund for different discretization meshes so much angular as spatial. (Author)
International Nuclear Information System (INIS)
Riyait, N.S.; Ackroyd, R.T.
1987-01-01
Proof-tests on 1-D multigroup neutron transport problems are reported for strong anisotropic scattering. These tests have been undertaken as part of the validation of the 3-D multigroup finite-element transport code FELTRAN for anisotropic scattering media. To illustrate the treatment of within-group and intergroup anisotropic scattering in the finite-element method the relevant theory is outlined. Ingroup scattering is checked using the backward-forward-isotropic (BFI) scattering law for source and eigenvalue problems. With this law anisotropic scattering problems can be transformed into equivalent isotropic scattering problems. In this way the well-validated isotropic scattering version of FELTRAN is used to validate the anisotropic version. Intergroup scattering effects are checked by solving few-group source problems for P 1 and P 3 scattering and the BFI scattering law. For P 1 and P 3 scattering checks are made with the discrete-ordinate finite-difference code ANISN and the spherical harmonics finite-difference code MARC/PN. For the BFI scattering law comparison is made with two-group exact solutions of Williams (1985) for 1-D systems. (author)
MCNP: a general Monte Carlo code for neutron and photon transport
Energy Technology Data Exchange (ETDEWEB)
Forster, R.A.; Godfrey, T.N.K.
1985-01-01
MCNP is a very general Monte Carlo neutron photon transport code system with approximately 250 person years of Group X-6 code development invested. It is extremely portable, user-oriented, and a true production code as it is used about 60 Cray hours per month by about 150 Los Alamos users. It has as its data base the best cross-section evaluations available. MCNP contains state-of-the-art traditional and adaptive Monte Carlo techniques to be applied to the solution of an ever-increasing number of problems. Excellent user-oriented documentation is available for all facets of the MCNP code system. Many useful and important variants of MCNP exist for special applications. The Radiation Shielding Information Center (RSIC) in Oak Ridge, Tennessee is the contact point for worldwide MCNP code and documentation distribution. A much improved MCNP Version 3A will be available in the fall of 1985, along with new and improved documentation. Future directions in MCNP development will change the meaning of MCNP to Monte Carlo N Particle where N particle varieties will be transported.
Numerical modeling of coupled water flow and heat transport in soil and snow
Thijs J. Kelleners; Jeremy Koonce; Rose Shillito; Jelle Dijkema; Markus Berli; Michael H. Young; John M. Frank; William Massman
2016-01-01
A one-dimensional vertical numerical model for coupled water flow and heat transport in soil and snow was modified to include all three phases of water: vapor, liquid, and ice. The top boundary condition in the model is driven by incoming precipitation and the surface energy balance. The model was applied to three different terrestrial systems: A warm desert bare...
Electron-vibron coupling effects on electron transport via a single-molecule magnet
McCaskey, A.; Yamamoto, Y.; Warnock, M.; Burzuri, E.; Van der Zant, H.S.J.; Park, K.
2015-01-01
We investigate how the electron-vibron coupling influences electron transport via an anisotropic magnetic molecule, such as a single-molecule magnet (SMM) Fe4, by using a model Hamiltonian with parameter values obtained from density-functional theory (DFT). The magnetic anisotropy parameters,
A Coupled Continuous Time Random Walk Approach For Transport in Highly Heterogeneous Porous Media
Dentz, M.; Scher, H.; Holder, D.; Berkowitz, B.
2008-12-01
We present a coupled continuous time random walk (CTRW) approach as an effective model for transport in highly heterogeneous media. This approach models solute transport by a coupled system of Langevin equations for random movements in the spatial and temporal domains. Motivated by transport in random fracture networks, here we consider a model that is characterized by given distributions of transition lengths (fracture length) and velocities. Thus, transition lengths and times are intrinsically related. Fracture length and velocity define the transition time. A maximum transition time is given by the diffusion time over the fracture length. Diffusion into the matrix can be modeled explicitly by a distribution of retention times. We study spatial distributions, and effective apparent transport coefficients as well as first arrival time distributions for a series of scenarios. The scaling behavior of such a fully coupled walk is different from the one observed in uncoupled walks. We investigate the competition between long jumps and long waiting times in this fully coupled continuous time random walk and determine scaling laws for the spatial moments of concentration.
The coupled influence of input suspension concentration (Ci), ionic strength (IS) and hydrodynamics on the transport and retention of 1.1 'm carboxyl modified latex colloids in saturated quartz sand (150 'm) was investigated. Results from batch experiments and interaction energy calculations indica...
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.
Finite moments approach to the time-dependent neutron transport equation
International Nuclear Information System (INIS)
Kim, Sang Hyun
1994-02-01
Currently, nodal techniques are widely used in solving the multidimensional diffusion equation because of savings in computing time and storage. Thanks to the development of computer technology, one can now solve the transport equation instead of the diffusion equation to obtain more accurate solution. The finite moments method, one of the nodal methods, attempts to represent the fluxes in the cell and on cell surfaces more rigorously by retaining additional spatial moments. Generally, there are two finite moments schemes to solve the time-dependent transport equation. In one, the time variable is treated implicitly with finite moments method in space variable (implicit finite moments method), the other method uses finite moments method in both space and time (space-time finite moments method). In this study, these two schemes are applied to two types of time-dependent neutron transport problems. One is a fixed source problem, the other a heterogeneous fast reactor problem with delayed neutrons. From the results, it is observed that the two finite moments methods give almost the same solutions in both benchmark problems. However, the space-time finite moments method requires a little longer computing time than that of the implicit finite moments method. In order to reduce the longer computing time in the space-time finite moments method, a new iteration strategy is exploited, where a few time-stepwise calculation, in which original time steps are grouped into several coarse time divisions, is performed sequentially instead of performing iterations over the entire time steps. This strategy results in significant reduction of the computing time and we observe that 2-or 3-stepwise calculation is preferable. In addition, we propose a new finite moments method which is called mixed finite moments method in this thesis. Asymptotic analysis for the finite moments method shows that accuracy of the solution in a heterogeneous problem mainly depends on the accuracy of the
Energy Coupling Efficiency in the Type I ABC Transporter GlnPQ.
Lycklama A Nijeholt, Jelger A; Vietrov, Ruslan; Schuurman-Wolters, Gea K; Poolman, Bert
2018-03-16
Solute transport via ATP binding cassette (ABC) importers involves receptor-mediated substrate binding, which is followed by ATP-driven translocation of the substrate across the membrane. How these steps are exactly initiated and coupled, and how much ATP it takes to complete a full transport cycle, are subject of debate. Here, we reconstitute the ABC importer GlnPQ in nanodiscs and in proteoliposomes and determine substrate-(in)dependent ATP hydrolysis and transmembrane transport. We determined the conformational states of the substrate-binding domains (SBDs) by single-molecule Förster resonance energy transfer measurements. We find that the basal ATPase activity (ATP hydrolysis in the absence of substrate) is mainly caused by the docking of the closed-unliganded state of the SBDs onto the transporter domain of GlnPQ and that, unlike glutamine, arginine binds both SBDs but does not trigger their closing. Furthermore, comparison of the ATPase activity in nanodiscs with glutamine transport in proteoliposomes shows that the stoichiometry of ATP per substrate is close to two. These findings help understand the mechanism of transport and the energy coupling efficiency in ABC transporters with covalently linked SBDs, which may aid our understanding of Type I ABC importers in general. Copyright © 2018 Elsevier Ltd. All rights reserved.
Energy Technology Data Exchange (ETDEWEB)
Gerhard Strydom; Cristian Rabiti; Andrea Alfonsi
2012-10-01
PHISICS is a neutronics code system currently under development at the Idaho National Laboratory (INL). Its goal is to provide state of the art simulation capability to reactor designers. The different modules for PHISICS currently under development are a nodal and semi-structured transport core solver (INSTANT), a depletion module (MRTAU) and a cross section interpolation (MIXER) module. The INSTANT module is the most developed of the mentioned above. Basic functionalities are ready to use, but the code is still in continuous development to extend its capabilities. This paper reports on the effort of coupling the nodal kinetics code package PHISICS (INSTANT/MRTAU/MIXER) to the thermal hydraulics system code RELAP5-3D, to enable full core and system modeling. This will enable the possibility to model coupled (thermal-hydraulics and neutronics) problems with more options for 3D neutron kinetics, compared to the existing diffusion theory neutron kinetics module in RELAP5-3D (NESTLE). In the second part of the paper, an overview of the OECD/NEA MHTGR-350 MW benchmark is given. This benchmark has been approved by the OECD, and is based on the General Atomics 350 MW Modular High Temperature Gas Reactor (MHTGR) design. The benchmark includes coupled neutronics thermal hydraulics exercises that require more capabilities than RELAP5-3D with NESTLE offers. Therefore, the MHTGR benchmark makes extensive use of the new PHISICS/RELAP5-3D coupling capabilities. The paper presents the preliminary results of the three steady state exercises specified in Phase I of the benchmark using PHISICS/RELAP5-3D.
International Nuclear Information System (INIS)
Loewenhaupt, M; Witte, U
2003-01-01
In general, elementary excitations in solids such as crystal field (CF) transitions and phonons are considered decoupled and the determination and interpretation of the measured spectra of the two phenomena, i.e. the CF level schemes and the phonon dispersion relations, are performed independently of each other. In addition, the spectra of these two excitations are generally quite complex and hence any unusual features are difficult to detect. A signature of a strong coupling between the two phenomena is the observation of an unusual behaviour in both subsystems. To prove the coupling unambiguously it is therefore necessary to investigate e.g. the phonon dispersion relations of an isostructural compound where the magnetic rare-earth ion (Ce, Yb) is replaced by a non-magnetic, but chemically equivalent ion (Y, La, Lu) and to determine the CF schemes of the same compound with the rare-earth ion replaced by a 'normal' magnetic rare earth. This requires, of course, time-consuming, detailed investigations. With these considerations in mind, it is not a surprise that there are only a few examples known where a coupling between electronic (CF transitions) and lattice (phonons) degrees of freedom have been reported. Here we will discuss results on three rare-earth compounds where the coupling between CF transitions and phonons has been unambiguously shown by inelastic neutron scattering experiments: CeAl 2 , YbPO 4 and CeCu 2
International Nuclear Information System (INIS)
Masiello, Emiliano; Martin, Brunella; Do, Jean-Michel
2011-01-01
A new development for the IDT solver is presented for large reactor core applications in XYZ geometries. The multigroup discrete-ordinate neutron transport equation is solved using a Domain-Decomposition (DD) method coupled with the Coarse-Mesh Finite Differences (CMFD). The later is used for accelerating the DD convergence rate. In particular, the external power iterations are preconditioned for stabilizing the oscillatory behavior of the DD iterative process. A set of critical 2-D and 3-D numerical tests on a single processor will be presented for the analysis of the performances of the method. The results show that the application of the CMFD to the DD can be a good candidate for large 3D full-core parallel applications. (author)
Energy Technology Data Exchange (ETDEWEB)
Paszkuta, M
2005-06-15
Low permeability materials containing clay play an important role in practical life and natural environment. Indeed, the ability of clay soils to act as semi permeable membranes, that inhibit the passage of electrolytes, is of great interest. The major objective of this thesis is to evaluate the transport properties of natural clays and in particular coupled transports when a pressure gradient, an electrical field, a concentration gradient and a temperature gradient interact. The material is a compact argillite extracted in East France from a Callovo-Oxfordian formation which was supplied to us by ANDRA. NaCl was used as the main solute. Two series of experiments were performed to measure permeability, diffusion, conductivity, the electro-osmotic coefficient and the Soret coefficient. (author)
Analytical calculations of neutron slowing down and transport in the constant-cross-section problem
International Nuclear Information System (INIS)
Cacuci, D.G.
1978-01-01
Some aspects of the problem of neutron slowing down and transport in an infinite medium consisting of a single nuclide that scatters elastically and isotropically and has energy-independent cross sections were investigated. The method of singular eigenfunctions was applied to the Boltzmann equation governing the Laplace transform (with respect to the lethargy variable) of the neutron flux. A new sufficient condition for the convergence of the coefficients of the expansion of the scattering kernel in Legendre polynomials was rigorously derived for this energy-dependent problem. Formulas were obtained for the lethargy-dependent spatial moments of the scalar flux that are valid for medium to large lethargies. In deriving these formulas, use was made of the well-known connection between the spatial moments of the Laplace-transformed scalar flux and the moments of the flux in the ''eigenvalue space.'' The calculations were greatly aided by the construction of a closed general expression for these ''eigenvalue space'' moments. Extensive use was also made of the methods of combinatorial analysis and of computer evaluation, via FORMAC, of complicated sequences of manipulations. For the case of no absorption it was possible to obtain for materials of any atomic weight explicit corrections to the age-theory formulas for the spatial moments M/sub 2n/(u) of the scalar flux that are valid through terms of the order of u -5 . The evaluation of the coefficients of the powers of n, as explicit functions of the nuclear mass, is one of the end products of this investigation. In addition, an exact expression for the second spatial moment, M 2 (u), valid for arbitrary (constant) absorption, was derived. It is now possible to calculate analytically and rigorously the ''age'' for the constant-cross-section problem for arbitrary (constant) absorption and nuclear mass. 5 figures, 1 table
Analytical calculations of neutron slowing down and transport in the constant-cross-section problem
International Nuclear Information System (INIS)
Cacuci, D.G.
1978-04-01
Aspects of the problem of neutron slowing down and transport in an infinite medium consisting of a single nuclide that scatters elastically and isotropically and has energy-independent cross sections were investigated. The method of singular eigenfunctions was applied to the Boltzmann Equation governing the Laplace transform (with respect to the lethargy variable) of the neutron flux. A new sufficient condition for the convergence of the coefficients of the expansion of the scattering kernel in Legendre polynomials was rigorously derived for this energy-dependent problem. Formulas were obtained for the lethargy-dependent spatial moments of the scalar flux that are valid for medium to large lethargies. Use was made of the well-known connection between the spatial moments of the Laplace-transformed scalar flux and the moments of the flux in the ''eigenvalue space.'' The calculations were aided by the construction of a closed general expression for these ''eigenvalue space'' moments. Extensive use was also made of the methods of combinatorial analysis and of computer evaluation of complicated sequences of manipulations. For the case of no absorption it was possible to obtain for materials of any atomic weight explicit corrections to the age-theory formulas for the spatial moments M/sub 2n/(u) of the scalar flux that are valid through terms of the order of u -5 . The evaluation of the coefficients of the powers of n, as explicit functions of the nuclear mass, represent one of the end products of this investigation. In addition, an exact expression for the second spatial moment, M 2 (u), valid for arbitrary (constant) absorption, was derived. It is now possible to calculate analytically and rigorously the ''age'' for the constant-cross-section problem for arbitrary (constant) absorption and nuclear mass. 5 figures, 1 table
Analytical calculations of neutron slowing down and transport in the constant-cross-section problem
Energy Technology Data Exchange (ETDEWEB)
Cacuci, D.G.
1978-04-01
Aspects of the problem of neutron slowing down and transport in an infinite medium consisting of a single nuclide that scatters elastically and isotropically and has energy-independent cross sections were investigated. The method of singular eigenfunctions was applied to the Boltzmann Equation governing the Laplace transform (with respect to the lethargy variable) of the neutron flux. A new sufficient condition for the convergence of the coefficients of the expansion of the scattering kernel in Legendre polynomials was rigorously derived for this energy-dependent problem. Formulas were obtained for the lethargy-dependent spatial moments of the scalar flux that are valid for medium to large lethargies. Use was made of the well-known connection between the spatial moments of the Laplace-transformed scalar flux and the moments of the flux in the ''eigenvalue space.'' The calculations were aided by the construction of a closed general expression for these ''eigenvalue space'' moments. Extensive use was also made of the methods of combinatorial analysis and of computer evaluation of complicated sequences of manipulations. For the case of no absorption it was possible to obtain for materials of any atomic weight explicit corrections to the age-theory formulas for the spatial moments M/sub 2n/(u) of the scalar flux that are valid through terms of the order of u/sup -5/. The evaluation of the coefficients of the powers of n, as explicit functions of the nuclear mass, represent one of the end products of this investigation. In addition, an exact expression for the second spatial moment, M/sub 2/(u), valid for arbitrary (constant) absorption, was derived. It is now possible to calculate analytically and rigorously the ''age'' for the constant-cross-section problem for arbitrary (constant) absorption and nuclear mass. 5 figures, 1 table.
Liu, Yingzi; Koltick, David; Byrne, Patrick; Wang, Haoyu; Zheng, Wei; Nie, Linda H
2013-12-01
This study was conducted to investigate the methodology and feasibility of developing a transportable neutron activation analysis (NAA) system to quantify manganese (Mn) in bone using a portable deuterium-deuterium (DD) neutron generator as the neutron source. Since a DD neutron generator was not available in our laboratory, a deuterium-tritium (DT) neutron generator was used to obtain experimental data and validate the results from Monte Carlo (MC) simulations. After validation, MC simulations using a DD generator as the neutron source were then conducted. Different types of moderators and reflectors were simulated, and the optimal thicknesses for the moderator and reflector were determined. To estimate the detection limit (DL) of the system, and to observe the interference of the magnesium (Mg) γ line at 844 keV to the Mn γ line at 847 keV, three hand phantoms with Mn concentrations of 30 parts per million (ppm), 150 ppm, and 500 ppm were made and irradiated by the DT generator system. The Mn signals in these phantoms were then measured using a 50% high-efficiency high-purity germanium (HPGe) detector. The DL was calculated to be about 4.4 ppm for the chosen irradiation, decay, and measurement time. This was calculated to be equivalent to a DL of about 3.3 ppm for the DD generator system. To achieve this DL with one 50% high-efficiency HPGe detector, the dose to the hand was simulated to be about 37 mSv, with the total body equivalent dose being about 23µSv. In conclusion, it is feasible to develop a transportable NAA system to quantify Mn in bone in vivo with an acceptable radiation exposure to the subject.
Mechanism of coupling drug transport reactions located in two different membranes
Directory of Open Access Journals (Sweden)
Helen I. Zgurskaya
2015-02-01
Full Text Available Gram- negative bacteria utilize a diverse array of multidrug transporters to pump toxic compounds out of cells. Some transporters together with periplasmic membrane fusion proteins (MFPs and outer membrane channels assemble trans-envelope complexes that expel multiple antibiotics across outer membranes of Gram-negative bacteria and into the external medium. Others further potentiate this efflux by pumping drugs across the inner membrane into the periplasm. Together these transporters create a powerful network of efflux that protect bacteria against a broad range of antimicrobial agents. This review is focused on the mechanism of coupling transport reactions located in two different membranes of Gram-negative bacteria. Using a combination of biochemical, genetic and biophysical approaches we have reconstructed the sequence of events leading to the assembly of trans-envelope drug efflux complexes and characterized the roles of periplasmic and outer membrane proteins in this process. Our recent data suggest a critical step in the activation of intermembrane efflux pumps, which is controlled by MFPs. We propose that the reaction cycles of transporters are tightly coupled to the assembly of the trans-envelope complexes. Transporters and MFPs exist in the inner membrane as dormant complexes. The activation of complexes is triggered by MFP binding to the outer membrane channel, which leads to a conformational change in the membrane proximal domain of MFP needed for stimulation of transporters. The activated MFP-transporter complex engages the outer membrane channel to expel substrates across the outer membrane. The recruitment of the channel is likely triggered by binding of effectors (substrates to MFP or MFP-transporter complexes. This model together with recent structural and functional advances in the field of drug efflux provides a fairly detailed understanding of the mechanism of drug efflux across the two membranes.
Pollock, Rachel A.
Mesoporous materials are interesting as catalyst supports, because molecules can move efficiently in and out of the pore network, but they must be stable in water if they are to be used for the production of biofuels. Before investigating hydrothermal stability and transport properties, the pore structure of SBA-15 was characterized using small angle neutron scattering (SANS) and non-local density functional theory (NLDFT) analysis of nitrogen sorption isotherms. A new Contrast Matching SANS method, using a range of probe molecules to directly probe the micropore size, gave a pore size distribution onset of 6 ± 0.2 Å, consistent with cylindrical pores formed from polymer template strands that unravel into the silica matrix. Diffraction intensity analysis of SANS measurements, combined with pore size distributions calculated from NLDFT, showed that the secondary pores are distributed relatively uniformly throughout the silica framework. The hydrothermal stability of SBA-15 was evaluated using a post-calcination hydrothermal treatment in both liquid and vapor phase water. The results were consistent with a degradation mechanism in which silica dissolves from regions of small positive curvature, e.g. near the entrance to the secondary pores, and is re-deposited deeper into the framework. Under water treatment at 115 °C, the mesopore diameter increases and the intra-wall void fraction decreases significantly. The behavior is similar for steam treatment, but occurs more slowly, suggesting that transport is faster when condensation occurs in the pores. Quasielastic neutron scattering (QENS) measurements of methane in SBA-15 probed the rotational and translational motion as a function of temperature and loading. A qualitative analysis of the QENS data suggested that for the initial dose of methane at 100 K, the self diffusion constant is similar in magnitude to literature values for methane in ZSM-5 and Y-zeolite, showing that the secondary pores trap methane and limit
Influences of a Side-Coupled Triple Quantum Dot on Kondo Transport Through a Quantum Dot
International Nuclear Information System (INIS)
Jiang Zhaotan; Yang Yannan; Qin Zhijie
2010-01-01
Kondo transport properties through a Kondo-type quantum dot (QD) with a side-coupled triple-QD structure are systematically investigated by using the non-equilibrium Green's function method. We firstly derive the formulae of the current, the linear conductance, the transmission coefficient, and the local density of states. Then we carry out the analytical and numerical studies and some universal conductance properties are obtained. It is shown that the number of the conductance valleys is intrinsically determined by the side-coupled QDs and at most equal to the number of the QDs included in the side-coupled structure in the asymmetric limit. In the process of forming the conductance valleys, the side-coupled QD system plays the dominant role while the couplings between the Kondo-type QD and the side-coupled structure play the subsidiary and indispensable roles. To testify the validity of the universal conductance properties, another different kinds of side-coupled triple-QD structures are considered. It should be emphasized that these universal properties are applicable in understanding this kind of systems with arbitrary many-QD side structures.
Electronic transport through a quantum dot chain with strong dot-lead coupling
International Nuclear Information System (INIS)
Liu, Yu; Zheng, Yisong; Gong, Weijiang; Gao, Wenzhu; Lue, Tianquan
2007-01-01
By means of the non-equilibrium Green function technique, the electronic transport through an N-quantum-dot chain is theoretically studied. By calculating the linear conductance spectrum and the local density of states in quantum dots, we find the resonant peaks in the spectra coincides with the eigen-energies of the N-quantum-dot chain when the dot-lead coupling is relatively weak. With the increase of the dot-lead coupling, such a correspondence becomes inaccurate. When the dot-lead coupling exceeds twice the interdot coupling, such a mapping collapses completely. The linear conductance turn to reflect the eigen-energies of the (N-2)- or (N-1)-quantum dot chain instead. The two peripheral quantum dots do not manifest themselves in the linear conductance spectrum. More interestingly, with the further increase of the dot-lead coupling, the system behaves just like an (N-2)- or (N-1)-quantum dot chain in weak dot-lead coupling limit, since the resonant peaks becomes narrower with the increase of dot-lead coupling
Neutron Transport in Spatially Random Media: An Assessment of the Accuracy of First Order Smoothing
International Nuclear Information System (INIS)
Williams, M.M.R.
2000-01-01
A formalism has been developed for studying the transmission of neutrons through a spatially stochastic medium. The stochastic components are represented by absorbing plates of randomly varying strength and random position. This type of geometry enables the Feinberg-Galanin-Horning method to be employed and leads to the solution of a coupled set of linear equations for the flux at the plate positions. The matrix of the coefficients contains members that are random and these are solved by simulation. That is, the strength and plate positions are sampled from uniform distributions and the equations solved many times (in this case 10 5 simulations are carried out). Probability distributions for the plate transmission and reflection factors are constructed from which the mean and variance can be computed.These essentially exact solutions enable closure approximations to be assessed for accuracy. To this end, we have compared the mean and variance obtained from the first order smoothing approximation of Keller with the exact results and have found excellent agreement for the mean values but note deviations of up to 40% for the variance. Nevertheless, for the problems considered here, first order smoothing appears to be of practical value and is very efficient numerically in comparison with simulation
Coupled processes of fluid flow, solute transport, and geochemical reactions in reactive barriers
Energy Technology Data Exchange (ETDEWEB)
Kim, Jeongkon; Schwartz, Franklin W.; Xu, Tianfu; Choi, Heechul, and Kim, In S.
2004-01-02
A complex pattern of coupling between fluid flow and mass transport develops when heterogeneous reactions occur. For instance, dissolution and precipitation reactions can change a porous medium's physical properties, such as pore geometry and thus permeability. These changes influence fluid flow, which in turn impacts the composition of dissolved constituents and the solid phases, and the rate and direction of advective transport. Two-dimensional modeling studies using TOUGHREACT were conducted to investigate the coupling between flow and transport developed as a consequence of differences in density, dissolution precipitation, and medium heterogeneity. The model includes equilibrium reactions for aqueous species, kinetic reactions between the solid phases and aqueous constituents, and full coupling of porosity and permeability changes resulting from precipitation and dissolution reactions in porous media. In addition, a new permeability relationship is implemented in TOUGHREACT to examine the effects of geochemical reactions and density difference on plume migration in porous media. Generally, the evolutions in the concentrations of the aqueous phase are intimately related to the reaction-front dynamics. Plugging of the medium contributed to significant transients in patterns of flow and mass transport.
International Nuclear Information System (INIS)
Hadad, Kamal; Pirouzmand, Ahmad; Ayoobian, Navid
2008-01-01
This paper describes the application of a multilayer cellular neural network (CNN) to model and solve the time dependent one-speed neutron transport equation in slab geometry. We use a neutron angular flux in terms of the Chebyshev polynomials (T N ) of the first kind and then we attempt to implement the equations in an equivalent electrical circuit. We apply this equivalent circuit to analyze the T N moments equation in a uniform finite slab using Marshak type vacuum boundary condition. The validity of the CNN results is evaluated with numerical solution of the steady state T N moments equations by MATLAB. Steady state, as well as transient simulations, shows a very good comparison between the two methods. We used our CNN model to simulate space-time response of total flux and its moments for various c (where c is the mean number of secondary neutrons per collision). The complete algorithm could be implemented using very large-scale integrated circuit (VLSI) circuitry. The efficiency of the calculation method makes it useful for neutron transport calculations
International Nuclear Information System (INIS)
Reitsma, Frederik
2007-01-01
Description of benchmark: This international benchmark, concerns Pebble-Bed Modular Reactor (PBMR) coupled neutronics/thermal hydraulics transients based on the PBMR-400 MW design. The deterministic neutronics, thermal-hydraulics and transient analysis tools and methods available to design and analyse PBMRs lag, in many cases, behind the state of the art compared to other reactor technologies. This has motivated the testing of existing methods for HTGRs but also the development of more accurate and efficient tools to analyse the neutronics and thermal-hydraulic behaviour for the design and safety evaluations of the PBMR. In addition to the development of new methods, this includes defining appropriate benchmarks to verify and validate the new methods in computer codes. The scope of the benchmark is to establish well-defined problems, based on a common given set of cross sections, to compare methods and tools in core simulation and thermal hydraulics analysis with a specific focus on transient events through a set of multi-dimensional computational test problems. The benchmark exercise has the following objectives: - Establish a standard benchmark for coupled codes (neutronics/thermal-hydraulics) for PBMR design; - Code-to-code comparison using a common cross section library ; - Obtain a detailed understanding of the events and the processes; - Benefit from different approaches, understanding limitations and approximations. Major Design and Operating Characteristics of the PBMR (PBMR Characteristic and Value): Installed thermal capacity: 400 MW(t); Installed electric capacity: 165 MW(e); Load following capability: 100-40-100%; Availability: ≥ 95%; Core configuration: Vertical with fixed centre graphite reflector; Fuel: TRISO ceramic coated U-235 in graphite spheres; Primary coolant: Helium; Primary coolant pressure: 9 MPa; Moderator: Graphite; Core outlet temperature: 900 C.; Core inlet temperature: 500 C.; Cycle type: Direct; Number of circuits: 1; Cycle
Coupling of the 3D neutron kinetic core model DYN3D with the CFD software ANSYS-CFX
International Nuclear Information System (INIS)
Grahn, Alexander; Kliem, Sören; Rohde, Ulrich
2015-01-01
Highlights: • Improved thermal hydraulic description of nuclear reactor cores. • Possibility of three-dimensional flow phenomena in the core, such as cross flow, flow reversal, flow around obstacles. • Simulation at higher spatial resolution as compared to system codes. - Abstract: This article presents the implementation of a coupling between the 3D neutron kinetic core model DYN3D and the commercial, general purpose computational fluid dynamics (CFD) software ANSYS-CFX. In the coupling approach, parts of the thermal hydraulic calculation are transferred to CFX for its better ability to simulate the three-dimensional coolant redistribution in the reactor core region. The calculation of the heat transfer from the fuel into the coolant remains with DYN3D, which incorporates well tested and validated heat transfer models for rod-type fuel elements. On the CFX side, the core region is modeled based on the porous body approach. The implementation of the code coupling is verified by comparing test case results with reference solutions of the DYN3D standalone version. Test cases cover mini and full core geometries, control rod movement and partial overcooling transients
International Nuclear Information System (INIS)
Silva, Vitor Vasconcelos Araújo
2016-01-01
The development of a fine mesh coupled neutronics/thermal-hydraulics framework mainly using open source software is presented. The contributions proposed go in two different directions: one, is the focus on the open software development, a concept widely spread in many fields of knowledge but rarely explored in the nuclear engineering field; the second, is the use of operating system shared memory as a fast and reliable storage area to couple the computational fluid dynamics (CFD) software OpenFOAM to the free and flexible reactor core analysis code Milonga. This concept was applied to simulate the behavior of the TRIGA Mark 1 IPR-R1 reactor fuel pin in steady-state mode. The macroscopic cross-sections for the model, a set of two-group cross-sections data, were generated using WIMSD-5B code. The results show that this innovative coupled system gives consistent results, encouraging system further development and its use for complex nuclear systems. (author)
Transport properties of a Kondo dot with a larger side-coupled noninteracting quantum dot
International Nuclear Information System (INIS)
Liu, Y S; Fan, X H; Xia, Y J; Yang, X F
2008-01-01
We investigate theoretically linear and nonlinear quantum transport through a smaller quantum dot in a Kondo regime connected to two leads in the presence of a larger side-coupled noninteracting quantum dot, without tunneling coupling to the leads. To do this we employ the slave boson mean field theory with the help of the Keldysh Green's function at zero temperature. The numerical results show that the Kondo conductance peak may develop multiple resonance peaks and multiple zero points in the conductance spectrum owing to constructive and destructive quantum interference effects when the energy levels of the large side-coupled noninteracting dot are located in the vicinity of the Fermi level in the leads. As the coupling strength between two quantum dots increases, the tunneling current through the quantum device as a function of gate voltage applied across the two leads is suppressed. The spin-dependent transport properties of two parallel coupled quantum dots connected to two ferromagnetic leads are also investigated. The numerical results show that, for the parallel configuration, the spin current or linear spin differential conductance are enhanced when the polarization strength in the two leads is increased
Heat transport through quantum Hall edge states: Tunneling versus capacitive coupling to reservoirs
Aita, Hugo; Arrachea, Liliana; Naón, Carlos; Fradkin, Eduardo
2013-08-01
We study the heat transport along an edge state of a two-dimensional electron gas in the quantum Hall regime, in contact to two reservoirs at different temperatures. We consider two exactly solvable models for the edge state coupled to the reservoirs. The first one corresponds to filling ν=1 and tunneling coupling to the reservoirs. The second one corresponds to integer or fractional filling of the sequence ν=1/m (with m odd), and capacitive coupling to the reservoirs. In both cases, we solve the problem by means of nonequilibrium Green function formalism. We show that heat propagates chirally along the edge in the two setups. We identify two temperature regimes, defined by Δ, the mean level spacing of the edge. At low temperatures, TΔ, finite-size effects become irrelevant, but the heat transport strongly depends on the strength of the edge-reservoir interactions, in both cases. The thermal conductance for tunneling coupling grows linearly with T, whereas for the capacitive case, it saturates to a value that depends on the coupling strengths and the filling factors of the edge and the contacts.
Electron-vibron coupling effects on electron transport via a single-molecule magnet
McCaskey, Alexander; Yamamoto, Yoh; Warnock, Michael; Burzurí, Enrique; van der Zant, Herre S. J.; Park, Kyungwha
2015-03-01
We investigate how the electron-vibron coupling influences electron transport via an anisotropic magnetic molecule, such as a single-molecule magnet (SMM) Fe4, by using a model Hamiltonian with parameter values obtained from density-functional theory (DFT). The magnetic anisotropy parameters, vibrational energies, and electron-vibron coupling strengths of the Fe4 are computed using DFT. A giant spin model is applied to the Fe4 with only two charge states, specifically a neutral state with a total spin S =5 and a singly charged state with S =9 /2 , which is consistent with our DFT result and experiments on Fe4 single-molecule transistors. In sequential electron tunneling, we find that the magnetic anisotropy gives rise to new features in the conductance peaks arising from vibrational excitations. In particular, the peak height shows a strong, unusual dependence on the direction as well as magnitude of applied B field. The magnetic anisotropy also introduces vibrational satellite peaks whose position and height are modified with the direction and magnitude of applied B field. Furthermore, when multiple vibrational modes with considerable electron-vibron coupling have energies close to one another, a low-bias current is suppressed, independently of gate voltage and applied B field, although that is not the case for a single mode with a similar electron-vibron coupling. In the former case, the conductance peaks reveal a stronger B -field dependence than in the latter case. The new features appear because the magnetic anisotropy barrier is of the same order of magnitude as the energies of vibrational modes with significant electron-vibron coupling. Our findings clearly show the interesting interplay between magnetic anisotropy and electron-vibron coupling in electron transport via the Fe4. Similar behavior can be observed in transport via other anisotropic magnetic molecules.
Bodek, Kazimierz
2012-09-01
The Standard Model (SM) predictions of T-violation for weak decays of systems built up of u and d quarks are by 7 to 10 orders of magnitude lower than the experimental accuracies attainable at present. It is a general presumption that time reversal phenomena are caused by a tiny admixture of exotic interaction terms. Therefore, weak decays provide a favorable testing ground in a search for such feeble forces. Physics with very slow, polarized neutrons has a great potential in this respect. An experiment seeking for small deviations from the SM in two observables, N and R, that are for the first time addressed experimentally in free neutron decay and that are exclusively sensitive to real and imaginary parts of the same linear combination of the scalar and tensor interaction coupling constants has been completed at the Paul Scherrer Institute, Villigen, Switzerland. The analysis of the experimental data has been completed recently leading to, among others, the best direct constraint for the imaginary part of the R-parity violating MSSM contribution. The success of the applied technique results in a new project devoted to the simultaneous measurement of seven correlation coefficients: H, L, N, R, S, U and V. Five of them (H, L, S, U and V) have never before been measured in weak decays. Such a systematic exploration of the transverse electron polarization will generate from the neutron decay alone a complete set of constraints for the real and imaginary parts of the weak scalar and tensor interactions on the level of 5 × 10-4 or better.
International Nuclear Information System (INIS)
Pazianotto, Mauricio T.; Carlson, Brett V.; Federico, Claudio A.; Gonzalez, Odair L.
2011-01-01
Neutrons generated by the interaction of cosmic rays with the atmosphere make an important contribution to the dose accumulated in electronic circuits and aircraft crew members at flight altitude. High-energy neutrons are produced in spallation reactions and intranuclear cascade processes by primary cosmic-ray particle interactions with atoms in the atmosphere. These neutrons can produce secondary neutrons and also undergo a moderation process due to atmosphere interactions, resulting in a wider energy spectrum, ranging from thermal energies (0.025 eV) to energies of several hundreds of MeV. The Long-Counter (LC) detector is a widely used neutron detector designed to measure the directional flux of neutrons with about constant response over a wide energy range (thermal to 20 MeV). ). Its calibration process and the determination of its energy response for the wide-energy of cosmic ray induced neutron spectrum is a very difficult process due to the lack of installations with these capabilities. The goal of this study is to assess the behavior of the response of a Long Counter using the Monte Carlo (MC) computational code MCNPX (Monte Carlo N-Particle eXtended). The dependence of the Long Counter response on the angle of incidence, as well as on the neutron energy, will be carefully investigated, compared with the experimental data previously obtained with 241 Am-Be and 252 Cf neutron sources and extended to the neutron spectrum produced by cosmic rays. (Author)
Coupled ion binding and structural transitions along the transport cycle of glutamate transporters
Verdon, Grégory; Oh, SeCheol; Serio, Ryan N; Boudker, Olga
2014-01-01
eLife digest Molecules of glutamate can carry messages between cells in the brain, and these signals are essential for thought and memory. Glutamate molecules can also act as signals to build new connections between brain cells and to prune away unnecessary ones. However, too much glutamate outside of the cells kills the brain tissue and can lead to devastating brain diseases. In a healthy brain, special pumps called glutamate transporters move these molecules back into the brain cells, where...
ITS - The integrated TIGER series of coupled electron/photon Monte Carlo transport codes
International Nuclear Information System (INIS)
Halbleib, J.A.; Mehlhorn, T.A.
1985-01-01
The TIGER series of time-independent coupled electron/photon Monte Carlo transport codes is a group of multimaterial, multidimensional codes designed to provide a state-of-the-art description of the production and transport of the electron/photon cascade. The codes follow both electrons and photons from 1.0 GeV down to 1.0 keV, and the user has the option of combining the collisional transport with transport in macroscopic electric and magnetic fields of arbitrary spatial dependence. Source particles can be either electrons or photons. The most important output data are (a) charge and energy deposition profiles, (b) integral and differential escape coefficients for both electrons and photons, (c) differential electron and photon flux, and (d) pulse-height distributions for selected regions of the problem geometry. The base codes of the series differ from one another primarily in their dimensionality and geometric modeling. They include (a) a one-dimensional multilayer code, (b) a code that describes the transport in two-dimensional axisymmetric cylindrical material geometries with a fully three-dimensional description of particle trajectories, and (c) a general three-dimensional transport code which employs a combinatorial geometry scheme. These base codes were designed primarily for describing radiation transport for those situations in which the detailed atomic structure of the transport medium is not important. For some applications, it is desirable to have a more detailed model of the low energy transport. The system includes three additional codes that contain a more elaborate ionization/relaxation model than the base codes. Finally, the system includes two codes that combine the collisional transport of the multidimensional base codes with transport in macroscopic electric and magnetic fields of arbitrary spatial dependence
Experimental and modeling analysis of coupled non-Fickian transport and sorption in natural soils.
Rubin, Shira; Dror, Ishai; Berkowitz, Brian
2012-05-01
We present experimental breakthrough curve (BTC) data and a modeling investigation of conservative and sorbing tracer transport in natural soils. By analyzing the data using the continuous time random walk (CTRW) model, we probe the traditional approach of using conservative tracer model parameters as a basis for quantifying the transport of sorbing solutes in the same domain when non-Fickian transport is present. Many known contaminants in groundwater are sorbed to the host solid porous medium, to varying extents, while being transported; this enhances the long tailing of BTCs which often already occurs because of the inherent non-Fickian nature of the transport. The CTRW framework has been shown to account very well for non-Fickian conservative (nonsorbing) transport. Here, we examine two BTC data sets in laboratory columns packed with natural soils; the first (previously analyzed by Mao and Ren (2004)) comprises transport of (conservative) bromide and (sorbing) atrazine tracers, while the second presents new data with bromide and tribromoneopentyl alcohol (TBNPA), a key flame retardant, as a sorbing solute. TBNPA has received little attention in the past, and is shown to be sorbed onto Bet Dagan soil in a nonlinear manner. We find that the transport behavior of bromide is non-Fickian in all cases, which is caused by the heterogeneity of the soil. Comparative model analysis of the non-Fickian BTCs of the conservative, and sorbing tracers and examination of the fitting parameters, exemplify the coupling between transport and adsorption/desorption processes. The difference in transport parameters used to match the conservative and sorbing data sets shows that conservative tracer parameters (average velocity and dispersion coefficient) are not valid for the transport of reactive tracers. Copyright © 2012 Elsevier B.V. All rights reserved.
Li, Yuanchang
2016-01-01
It is generally believed that the inter-edge coupling destroys the quantum spin Hall (QSH) effect along with the gap opening at the Dirac points. Using first-principles calculations, we find that the quantized edge transport persists in the presence of inter-edge coupling in Ta intercalated epitaxial graphene on SiC(0001), being a QSH insulator with the non-trivial gap of 81 meV. In this case, the band is characterized by two perfect Dirac cones with different Fermi velocities, yet only one m...
International Nuclear Information System (INIS)
Abreu, Marcos Pimenta de
1998-01-01
We describe a numerical method applied to the first-order form of one-speed slab-geometry discrete ordinates equations modelling time-independent neutron transport problems with anisotropic scattering, with no interior source and defined in a nonmultiplying homogeneous host medium. Our numerical method is concerned with the generation of the spectrum and of a vector basis for the null space of the one-speed slab-geometry discrete ordinates operator. Moreover, it allows us to overcome the difficulties introduced in previous methods by anisotropic scattering and by angular quadrature sets of high order. To illustrate the positive features of our numerical method, we present numerical results for one-speed slab-geometry neutron transport model problems with anisotropic scattering
International Nuclear Information System (INIS)
Yildiz, C.
1998-01-01
The critical slab problem is studied in one-speed neutron transport theory using a linearly anisotropic kernel which combines forward and backward scattering. It is shown that, the recently observed non-monotonic variation of the thickness also exists in this strongly anisotropic case. In addition, the influence of the linear anisotropy on the critical thickness is analysed in detail. Numerical analysis for the critical thickness are performed using the spherical harmonics method and results are tabulated for selected illustrative cases as a function of different degrees of anisotropic scattering. Finally, some results are discussed and compared with those already obtained by other methods, the agreement is satisfactory. The spherical harmonic method gives generally accurate results in one dimensional geometry, and it is very suitable for the numerical solution of the neutron transport equation with linearly anisotropic scattering
Zegers Risopatron, G., Sr.; Navarro, L.; Montserrat, S., Sr.; McPhee, J. P.; Niño, Y.
2017-12-01
The geochemistry of water and sediments, coupled with hydrodynamic transport in mountainous channels, is of particular interest in central Chilean Andes due to natural occurrence of acid waters. In this paper, we present a coupled transport and geochemical model to estimate and understand transport processes and fate of minerals at the Yerba Loca Basin, located near Santiago, Chile. In the upper zone, water presentes low pH ( 3) and high concentrations of iron, aluminum, copper, manganese and zinc. Acidity and minerals are the consequence of water-rock interactions in hydrothermal alteration zones, rich in sulphides and sulphates, covered by seasonal snow and glaciers. Downstream, as a consequence of neutral to alkaline lateral water contributions (pH >7) along the river, pH increases and concentration of solutes decreases. The mineral transport model has three components: (i) a hydrodynamic model, where we use HEC-RAS to solve 1D Saint-Venant equations, (ii) a sediment transport model to estimate erosion and sedimentation rates, which quantify minerals transference between water and riverbed and (iii) a solute transport model, based on the 1D OTIS model which takes into account the temporal delay in solutes transport that typically is observed in natural channels (transient storage). Hydrochemistry is solved using PHREEQC, a software for speciation and batch reaction. Our results show that correlation between mineral precipitation and dissolution according to pH values changes along the river. Based on pH measurements (and according to literature) we inferred that main minerals in the water system are brochantite, ferrihydrite, hydrobasaluminite and schwertmannite. Results show that our model can predict the transport and fate of minerals and metals in the Yerba Loca Basin. Mineral dissolution and precipitation process occur for limited ranges of pH values. When pH values are increased, iron minerals (schwertmannite) are the first to precipitate ( 2.5
Directory of Open Access Journals (Sweden)
Frederic Salaun
2018-01-01
Full Text Available The Canadian Supercritical Water-cooled Reactor (SCWR, a GEN IV reactor design, is a hybrid design of the well-established CANDU™ and Boiling Water Reactor with water above its thermodynamic critical point. Given the batch fueled design, control rods are used to manage the reactivity throughout the fuel cycle. This paper examines the consequences of a control rod drop accident (CRDA for the Canadian SCWR. The asymmetry generated by the dropped rod requires an accurate 3-dimensional neutron kinetics calculation coupled to a detailed thermal-hydraulic model. Before simulating the CRDAs, the proper implementation of the 3D reactivity feedback was verified and various sensitivity studies were performed. This work demonstrates that the proposed safety systems for the SCWR core are capable of terminating the CRDA sequence prior to exceeding maximum sheath and centerline temperatures. In one instance involving a rod on the periphery of the core, the proposed trip setpoint (115% FP was not exceeded and a new steady state was reached. Therefore it is recommended that the design also include provisions for a high-log rate and/or local Neutron Overpower Protection (NOP trips, similar to existing CANDU designs such that reactor shutdown can be assured for such spatial anomalies.
Energy Technology Data Exchange (ETDEWEB)
DeHart, Mark David [Texas A & M Univ., College Station, TX (United States)
1992-12-01
A method for applying the discrete ordinates method for solution of the neutron transport equation in arbitary two-dimensional meshes has been developed. The finite difference approach normally used to approximate spatial derivatives in extrapolating angular fluxes across a cell is replaced by direct solution of the characteristic form of the transport equation for each discrete direction. Thus, computational cells are not restricted to the traditional shape of a mesh element within a given coordinate system. However, in terms of the treatment of energy and angular dependencies, this method resembles traditional discrete ordinates techniques. Using the method developed here, a general two-dimensional space can be approximated by an irregular mesh comprised of arbitrary polygons. The present work makes no assumptions about the orientations or the number of sides in a given cell, and computes all geometric relationships between each set of sides in each cell for each discrete direction. A set of non-reentrant polygons can therefore be used to represent any given two dimensional space. Results for a number of test problems have been compared to solutions obtained from traditional methods, with good agreement. Comparisons include benchmarks against analytical results for problems with simple geometry, as well numerical results obtained from traditional discrete ordinates methods by applying the ANISN and TWOTRAN computer programs. Numerical results were obtained for problems ranging from simple one-dimensional geometry to complicated multidimensional configurations. These results have demonstrated the ability of the developed method to closely approximate complex geometrical configurations and to obtain accurate results for problems that are extremely difficult to model using traditional methods.
Energy Technology Data Exchange (ETDEWEB)
Waata, C.L.
2006-07-15
The use of water at supercritical pressure as coolant and moderator introduces a challenge in the design of a High-Performance Light-Water Reactor (HPLWR) fuel assembly. At supercritical pressure condition (P=25 MPa), the thermal-hydraulics behaviour of water differs strongly from that at sub-critical pressure due to a rapid variation of the thermal-physical properties across the pseudo-critical line. Due of the strong link between the water (moderation) and the neutron spectrum and subsequently the power distribution, a coupling of neutronics and thermal-hydraulics has become a necessity for reactor concepts operating at supercritical pressure condition. The effect of neutron moderation on the local parameters of thermal-hydraulics and vice-verse in a fuel assembly has to be considered for an accurate design analysis. In this study, the Monte Carlo N-Particle code (MCNP) and the sub-channel code STAFAS (Sub-channel Thermal-hydraulics Analysis of a Fuel Assembly under Supercritical conditions) have been coupled for the design analysis of a fuel assembly with supercritical water as coolant and moderator. Both codes are well known for complex geometry modelling. The MCNP code is used for neutronics analyses and for the prediction of power profiles of individual fuel rods. The sub-channel code STAFAS for the thermal-hydraulics analyses takes into account the coolant properties beyond the critical point as well as separate moderator channels. The coupling procedure is realized automatically. MCNP calculates the power distribution in each fuel rod, which is then transferred into STAFAS to obtain the corresponding thermal-hydraulic conditions in each sub-channel. The new thermal-hydraulic conditions are used to generate a new input deck for the next MCNP calculation. This procedure is repeated until a converged state is achieved. The coupled code system was tested on a proposed fuel assembly design of a HPLWR. An under-relaxation was introduced to achieve convergence
International Nuclear Information System (INIS)
Waata, C.L.
2006-07-01
The use of water at supercritical pressure as coolant and moderator introduces a challenge in the design of a High-Performance Light-Water Reactor (HPLWR) fuel assembly. At supercritical pressure condition (P=25 MPa), the thermal-hydraulics behaviour of water differs strongly from that at sub-critical pressure due to a rapid variation of the thermal-physical properties across the pseudo-critical line. Due of the strong link between the water (moderation) and the neutron spectrum and subsequently the power distribution, a coupling of neutronics and thermal-hydraulics has become a necessity for reactor concepts operating at supercritical pressure condition. The effect of neutron moderation on the local parameters of thermal-hydraulics and vice-verse in a fuel assembly has to be considered for an accurate design analysis. In this study, the Monte Carlo N-Particle code (MCNP) and the sub-channel code STAFAS (Sub-channel Thermal-hydraulics Analysis of a Fuel Assembly under Supercritical conditions) have been coupled for the design analysis of a fuel assembly with supercritical water as coolant and moderator. Both codes are well known for complex geometry modelling. The MCNP code is used for neutronics analyses and for the prediction of power profiles of individual fuel rods. The sub-channel code STAFAS for the thermal-hydraulics analyses takes into account the coolant properties beyond the critical point as well as separate moderator channels. The coupling procedure is realized automatically. MCNP calculates the power distribution in each fuel rod, which is then transferred into STAFAS to obtain the corresponding thermal-hydraulic conditions in each sub-channel. The new thermal-hydraulic conditions are used to generate a new input deck for the next MCNP calculation. This procedure is repeated until a converged state is achieved. The coupled code system was tested on a proposed fuel assembly design of a HPLWR. An under-relaxation was introduced to achieve convergence
International Nuclear Information System (INIS)
Seed, T.J.; Miller, W.F. Jr.; Brinkley, F.W. Jr.
1977-03-01
TRIDENT solves the two-dimensional-multigroup-transport equations in rectangular (x-y) and cylindrical (r-z) geometries using a regular triangular mesh. Regular and adjoint, inhomogeneous and homogeneous (k/sub eff/ and eigenvalue searches) problems subject to vacuum, reflective, white, or source boundary conditions are solved. General anisotropic scattering is allowed and anisotropic-distributed sources are permitted. The discrete-ordinates approximation is used for the neutron directional variables. An option is included to append a fictitious source to the discrete-ordinates equations that is defined such that spherical-harmonics solutions (in x-y geometry) or spherical-harmonics-like solutions (in r-z geometry) are obtained. A spatial-finite-element method is used in which the angular flux is expressed as a linear polynomial in each triangle that is discontinous at triangle boundaries. Unusual Features of the program: Provision is made for creation of standard interface output files for S/sub N/ constants, angle-integrated (scalar) fluxes, and angular fluxes. Standard interface input files for S/sub N/ constants, inhomogeneous sources, cross sections, and the scalar flux may be read. Flexible edit options as well as a dump and restart capability are provided
International Nuclear Information System (INIS)
Hoffman, Adam J.; Lee, John C.
2016-01-01
A new time-dependent Method of Characteristics (MOC) formulation for nuclear reactor kinetics was developed utilizing angular flux time-derivative propagation. This method avoids the requirement of storing the angular flux at previous points in time to represent a discretized time derivative; instead, an equation for the angular flux time derivative along 1D spatial characteristics is derived and solved concurrently with the 1D transport characteristic equation. This approach allows the angular flux time derivative to be recast principally in terms of the neutron source time derivatives, which are approximated to high-order accuracy using the backward differentiation formula (BDF). This approach, called Source Derivative Propagation (SDP), drastically reduces the memory requirements of time-dependent MOC relative to methods that require storing the angular flux. An SDP method was developed for 2D and 3D applications and implemented in the computer code DeCART in 2D. DeCART was used to model two reactor transient benchmarks: a modified TWIGL problem and a C5G7 transient. The SDP method accurately and efficiently replicated the solution of the conventional time-dependent MOC method using two orders of magnitude less memory.
Energy Technology Data Exchange (ETDEWEB)
Lizorkin, M.; Nikonov, S. [Kurchatov Institute for Atomic Energy, Moscow (Russian Federation); Langenbuch, S.; Velkov, K. [Gesellschaft fur Anlagen- und Reaktorsicherheit (GRS) mbH, Garching (Germany)
2006-07-01
The coupled thermal-hydraulics and neutron-kinetics code ATHLET/BIPR-VVER was developed within a co-operation between the RRC Kurchatov Institute (KI) and GRS. The modeling capability of this coupled code as well as the status of validation by benchmark activities and comparison with plant measurements are described. The paper is focused on the modeling of flow mixing in the reactor pressure vessel including its validation and the application for the safety justification of VVER plants. (authors)
International Nuclear Information System (INIS)
Bildstein, O.
2010-06-01
The author gives an overview of his research and teaching activities. His researches first dealt with the development of a simulation of the chemistry/transport coupling and of the retroactive effects on transport parameters, then with the chemistry/transport modelling and its coupling with mechanics, and finally with the multi-scale investigation of porous materials. Perspectives are discussed and publications are indicated
Charge transport across a single-Cooper-pair transistor coupled to a resonant transmission line
Energy Technology Data Exchange (ETDEWEB)
Leppaekangas, Juha [Institut fuer Theoretische Festkoerperphysik, Karlsruhe Institute of Technology, D-76128 Karlsruhe (Germany); Department of Physical Sciences, University of Oulu, FI-90014 Oulu (Finland); Pashkin, Yuri [NEC Nano Electronics Research Laboratories, RIKEN Advanced Science Institute, Tsukuba, Ibaraki 305-8501 (Japan); Thuneberg, Erkki [Department of Physical Sciences, University of Oulu, FI-90014 Oulu (Finland)
2010-07-01
We have investigated charge transport in ultrasmall superconducting single and double Josephson junctions coupled to a transmission-line resonator. The microstrip resonator is naturally formed by the on-chip leads and the sample holder. We observe equidistant peaks in the transport characteristics of both types of devices and attribute them to the process involving simultaneous tunneling of Cooper pairs and photon emission into the resonator. The experimental data is well reproduced with the orthodox model of Cooper pair tunneling that accounts for the microwave photon emission into the resonator.
Energy Conservation Tests of a Coupled Kinetic-kinetic Plasma-neutral Transport Code
Energy Technology Data Exchange (ETDEWEB)
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.
Yu, Minghao; Kihara, Hisashi; Abe, Ken-ichi; Takahashi, Yusuke
2015-06-01
A relatively simple method for calculating accurately the third-order electron transport properties of nitrogen and air thermal plasmas is presented. The electron transport properties, such as the electrical conductivity and the electron thermal conductivity, were computed with the best and latest available collision cross-section data in the temperature and pressure ranges of T = 300 - 15000 K and p = 0.01 - 1.0 atm, respectively. The results obtained under the atmospheric pressure condition showed good agreements with the experimental and the high-accuracy theoretical results. The presently-introduced method has good application potential in numerical simulations of nitrogen and air inductively-coupled plasmas.
Energy Technology Data Exchange (ETDEWEB)
Duerigen, Susan
2013-05-15
The superior advantage of a nodal method for reactor cores with hexagonal fuel assemblies discretized as cells consisting of equilateral triangles is its mesh refinement capability. In this thesis, a diffusion and a simplified P{sub 3} (or SP{sub 3}) neutron transport nodal method are developed based on trigonal geometry. Both models are implemented in the reactor dynamics code DYN3D. As yet, no other well-established nodal core analysis code comprises an SP{sub 3} transport theory model based on trigonal meshes. The development of two methods based on different neutron transport approximations but using identical underlying spatial trigonal discretization allows a profound comparative analysis of both methods with regard to their mathematical derivations, nodal expansion approaches, solution procedures, and their physical performance. The developed nodal approaches can be regarded as a hybrid NEM/AFEN form. They are based on the transverse-integration procedure, which renders them computationally efficient, and they use a combination of polynomial and exponential functions to represent the neutron flux moments of the SP{sub 3} and diffusion equations, which guarantees high accuracy. The SP{sub 3} equations are derived in within-group form thus being of diffusion type. On this basis, the conventional diffusion solver structure can be retained also for the solution of the SP{sub 3} transport problem. The verification analysis provides proof of the methodological reliability of both trigonal DYN3D models. By means of diverse hexagonal academic benchmark and realistic detailed-geometry full-transport-theory problems, the superiority of the SP{sub 3} transport over the diffusion model is demonstrated in cases with pronounced anisotropy effects, which is, e.g., highly relevant to the modeling of fuel assemblies comprising absorber material.
Solution and study of nodal neutron transport equation applying the LTS{sub N}-DiagExp method
Energy Technology Data Exchange (ETDEWEB)
Hauser, Eliete Biasotto; Pazos, Ruben Panta [Pontificia Univ. Catolica do Rio Grande do Sul, Porto Alegre, RS (Brazil). Faculdade de Matematica]. E-mail: eliete@pucrs.br; rpp@mat.pucrs.br; Vilhena, Marco Tullio de [Pontificia Univ. Catolica do Rio Grande do Sul, Porto Alegre, RS (Brazil). Instituto de Matematica]. E-mail: vilhena@mat.ufrgs.br; Barros, Ricardo Carvalho de [Universidade do Estado, Nova Friburgo, RJ (Brazil). Instituto Politecnico]. E-mail: ricardo@iprj.uerj.br
2003-07-01
In this paper we report advances about the three-dimensional nodal discrete-ordinates approximations of neutron transport equation for Cartesian geometry. We use the combined collocation method of the angular variables and nodal approach for the spatial variables. By nodal approach we mean the iterated transverse integration of the S{sub N} equations. This procedure leads to the set of one-dimensional averages angular fluxes in each spatial variable. The resulting system of equations is solved with the LTS{sub N} method, first applying the Laplace transform to the set of the nodal S{sub N} equations and then obtained the solution by symbolic computation. We include the LTS{sub N} method by diagonalization to solve the nodal neutron transport equation and then we outline the convergence of these nodal-LTS{sub N} approximations with the help of a norm associated to the quadrature formula used to approximate the integral term of the neutron transport equation. (author)
A turbulent transport network model in MULTIFLUX coupled with TOUGH2
International Nuclear Information System (INIS)
Danko, G.; Bahrami, D.; Birkholzer, J.T.
2011-01-01
A new numerical method is described for the fully iterated, conjugate solution of two discrete submodels, involving (a) a transport network model for heat, moisture, and airflows in a high-permeability, air-filled cavity; and (b) a variably saturated fractured porous medium. The transport network submodel is an integrated-parameter, computational fluid dynamics solver, describing the thermal-hydrologic transport processes in the flow channel system of the cavity with laminar or turbulent flow and convective heat and mass transport, using MULTIFLUX. The porous medium submodel, using TOUGH2, is a solver for the heat and mass transport in the fractured rock mass. The new model solution extends the application fields of TOUGH2 by integrating it with turbulent flow and transport in a discrete flow network system. We present demonstrational results for a nuclear waste repository application at Yucca Mountain with the most realistic model assumptions and input parameters including the geometrical layout of the nuclear spent fuel and waste with variable heat load for the individual containers. The MULTIFLUX and TOUGH2 model elements are fully iterated, applying a programmed reprocessing of the Numerical Transport Code Functionalization model-element in an automated Outside Balance Iteration loop. The natural, convective airflow field and the heat and mass transport in a representative emplacement drift during postclosure are explicitly solved in the new model. The results demonstrate that the direction and magnitude of the air circulation patterns and all transport modes are strongly affected by the heat and moisture transport processes in the surrounding rock, justifying the need for a coupled, fully iterated model solution such as the one presented in the paper.
ZZ ENDLIB, Coupled Electron and Photon Transport Library in ENDL Format
International Nuclear Information System (INIS)
2002-01-01
Description of program or function: The LLNL Evaluated Nuclear Data Library has existed since 1958 in a succession of forms and formats. The present form is as a machine-independent character file format and contains data for the evaluated atomic relaxation data library (EADL), the evaluated photon interaction data library (EPDL), and the evaluated electron interaction data library (EEDL). The purpose of these libraries is to furnish data for coupled electron-photon transport calculations. In order to perform coupled photon-electron transport calculations, all three libraries are required. The UCRL-ID-117796 report included in the documentation for this package provides information on the contents and formats for all three libraries, which are included in this package. All of these libraries span atomic numbers, Z, from 1 to 100. Additionally the electron and photon interaction libraries cover the incident particle energy range from 10 eV to 100 GeV
Fluid transportation mechanisms by a coupled system of elastic membranes and magnetic fluids
International Nuclear Information System (INIS)
Ido, Y.; Tanaka, K.; Sugiura, Y.
2002-01-01
The basic properties of the fluid transportation mechanism that is produced by the coupled waves propagating along a thin elastic membrane covering a magnetic fluid layer in a shallow and long rectangular vessel are investigated. It is shown that the progressive magnetic field induced by the rectangular pulses generates sinusoidal vibration of the displacement of elastic membrane and makes the system work more efficiently than the magnetic field induced by the pulse-width-modulation method
Lolkema, Juke S.; Robillard, George T.
The bacterial phosphotransferase systems are believed to catalyze the concomitant transport and phosphorylation of hexoses and hexitols. The transport is from the outside to the inside of the cell. An absolute coupling between transport and phosphorylation has however been questioned in the
Energy Technology Data Exchange (ETDEWEB)
Brogi, Bharat Bhushan, E-mail: brogi-221179@yahoo.in; Ahluwalia, P. K. [Department of Physics, Himachal Pradesh University, Shimla-171005 (India); Chand, Shyam [University Institute of Information Technology, H.P. University Shimla-171005 (India)
2015-06-24
Theoretical study of the Coulomb blockade effect on transport properties (Transmission Probability and I-V characteristics) for varied configuration of coupled quantum dot system has been studied by using Non Equilibrium Green Function(NEGF) formalism and Equation of Motion(EOM) method in the presence of magnetic flux. The self consistent approach and intra-dot Coulomb interaction is being taken into account. As the key parameters of the coupled quantum dot system such as dot-lead coupling, inter-dot tunneling and magnetic flux threading through the system can be tuned, the effect of asymmetry parameter and magnetic flux on this tuning is being explored in Coulomb blockade regime. The presence of the Coulomb blockade due to on-dot Coulomb interaction decreases the width of transmission peak at energy level ε + U and by adjusting the magnetic flux the swapping effect in the Fano peaks in asymmetric and symmetric parallel configuration sustains despite strong Coulomb blockade effect.
Hydro-geochemistry: coupling chemical reactors with a particle-transport model
International Nuclear Information System (INIS)
Sauty, J.P.; Fabriol, R.
1993-01-01
Pollutant migration in groundwater and ore deposition have in common that they depend upon mechanisms coupling fluid flow and the chemical reactions between dissolved species and the solid matrix. The use of coupled models is indispensable for understanding and predicting such mechanisms. To this end, an original coupling method is presented between chemical reactors, custom-built with the help of a simulation generator, and a particle-transport model. After discussion of the principles and techniques that were called upon, two applications are presented. The first consists in a verification of the program compared to an exact theoretical solution defined as part of the international CHEMVAL rest; the second validates the program by simulating the experimental results obtained by percolation through a sandstone column. (authors). 21 refs., 6 figs
International Nuclear Information System (INIS)
Tsang, C.F.
1984-10-01
The present paper gives a general overview of mass transport in low permeability rocks under the coupled thermomechanical and hydrochemical effects associated with a nuclear waste repository. A classification of coupled processes is given. Then an ess is presented. example of a coupled process is presented. Discussions of coupled processes based on a recent LBL Panel meeting are summarized. 5 references, 3 figures, 4 tables
Energy Technology Data Exchange (ETDEWEB)
Fournier, D.
2011-10-10
flux behaviour is very different depending on the energy, there is no reason to use the same spatial discretization. Such an approach implies to modify the initial estimators in order to take into account the coupling between groups. This study is done from a theoretical as well as from a numerical point of view. (author) [French] Pour la conception des coeurs de reacteurs de 4eme generation, une precision accrue est requise pour les calculs des differents parametres neutroniques. Les ressources memoire et le temps de calcul etant limites, une solution consiste a utiliser des methodes de raffinement de maillage afin de resoudre l'equation de transport des neutrons. Le flux neutronique, solution de cette equation, depend de l'energie, l'angle et l'espace. Les differentes variables sont discretisees de maniere successive. L'energie avec une approche multigroupe, considerant les differentes grandeurs constantes sur chaque groupe, l'angle par une methode de collocation, dite approximation SN. Apres discretisation energetique et angulaire, un systeme d'equations hyperboliques couplees ne dependant plus que de la variable d'espace doit etre resolu. Des elements finis discontinus sont alors utilises afin de permettre la mise en place de methodes de raffinement dite hp. La precision de la solution peut alors etre amelioree via un raffinement en espace (h-raffinement), consistant a subdiviser une cellule en sous-cellules, ou en ordre (p-raffinement) en augmentant l'ordre de la base de polynomes utilisee. Dans cette these, les proprietes de ces methodes sont analysees et montrent l'importance de la regularite de la solution dans le choix du type de raffinement. Ainsi deux estimateurs d'erreurs permettant de mener le raffinement ont ete utilises. Le premier, suppose des hypotheses de regularite tres fortes (solution analytique) alors que le second utilise seulement le fait que la solution est a variations bornees. La
Vande Geest, Jonathan P; Simon, B R; Rigby, Paul H; Newberg, Tyler P
2011-04-01
Finite element models (FEMs) including characteristic large deformations in highly nonlinear materials (hyperelasticity and coupled diffusive/convective transport of neutral mobile species) will allow quantitative study of in vivo tissues. Such FEMs will provide basic understanding of normal and pathological tissue responses and lead to optimization of local drug delivery strategies. We present a coupled porohyperelastic mass transport (PHEXPT) finite element approach developed using a commercially available ABAQUS finite element software. The PHEXPT transient simulations are based on sequential solution of the porohyperelastic (PHE) and mass transport (XPT) problems where an Eulerian PHE FEM is coupled to a Lagrangian XPT FEM using a custom-written FORTRAN program. The PHEXPT theoretical background is derived in the context of porous media transport theory and extended to ABAQUS finite element formulations. The essential assumptions needed in order to use ABAQUS are clearly identified in the derivation. Representative benchmark finite element simulations are provided along with analytical solutions (when appropriate). These simulations demonstrate the differences in transient and steady state responses including finite deformations, total stress, fluid pressure, relative fluid, and mobile species flux. A detailed description of important model considerations (e.g., material property functions and jump discontinuities at material interfaces) is also presented in the context of finite deformations. The ABAQUS-based PHEXPT approach enables the use of the available ABAQUS capabilities (interactive FEM mesh generation, finite element libraries, nonlinear material laws, pre- and postprocessing, etc.). PHEXPT FEMs can be used to simulate the transport of a relatively large neutral species (negligible osmotic fluid flux) in highly deformable hydrated soft tissues and tissue-engineered materials.
International Nuclear Information System (INIS)
D'Auria, Francesco; Moreno, Jose Luis Gago; Galassi, Giorgio Maria; Grgic, Davor; Spadoni, Antonino
2003-01-01
A comprehensive analysis of the double ended main steam line break (MSLB) accident assumed to occur in the Babcock and Wilcox Three Mile Island Unit 1 (TMI-1) has been carried out at the Dipartimento di Ingegneria Meccanica, Nucleare e della Produzione of the University of Pisa, Italy, in cooperation with the University of Zagreb, Croatia. The overall activity has been completed within the framework of the participation in the Organization for Economic Cooperation and Development-Committee on the Safety of Nuclear Installations-Nuclear Science Committee pressurized water reactor MSLB benchmark.Thermal-hydraulic system codes (various versions of Relap5), three-dimensional (3-D) neutronics codes (Parcs, Quabbox, and Nestle), and one subchannel code (Cobra) have been adopted for the analysis. Results from the following codes (or code versions) are assumed as reference:1. Relap5/mod3.2.2, beta version, coupled with the 3-D neutron kinetics Parcs code parallel virtual machine (PVM) coupling2. Relap5/mod3.2.2, gamma version, coupled with the 3-D neutron kinetics Quabbox code (direct coupling)3. Relap5/3D code coupled with the 3-D neutron kinetics Nestle code.The influence of PVM and of direct coupling is also discussed.Boundary and initial conditions of the system, including those relevant to the fuel status, have been supplied by Pennsylvania State University in cooperation with GPU Nuclear Corporation (the utility, owner of TMI) and the U.S. Nuclear Regulatory Commission. The comparison among the results obtained by adopting the same thermal-hydraulic nodalization and the coupled code version is discussed in this paper.The capability of the control rods to recover the accident has been demonstrated in all the cases as well as the capability of all the codes to predict the time evolution of the assigned transient. However, one stuck control rod caused some 'recriticality' or 'return to power' whose magnitude is largely affected by boundary and initial conditions
International Nuclear Information System (INIS)
Sawan, M.; Wilson, P.; El-Guebaly, L.; Henderson, D.; Sviatoslavsky, G.; Bohm, T.; Kiedrowski, B.; Ibrahim, A.; Smith, B.; Slaybaugh, R.; Tautges, T.
2007-01-01
Fusion systems are, in general, geometrically complex requiring detailed three-dimensional (3-D) nuclear analysis. This analysis is required to address tritium self-sufficiency, nuclear heating, radiation damage, shielding, and radiation streaming issues. To facilitate such calculations, we developed an innovative computational tool that is based on the continuous energy Monte Carlo code MCNP and permits the direct use of CAD-based solid models in the ray-tracing. This allows performing the neutronics calculations in a model that preserves the geometrical details without any simplification, eliminates possible human error in modeling the geometry for MCNP, and allows faster design iterations. In addition to improving the work flow for simulating complex 3- D geometries, it allows a richer representation of the geometry compared to the standard 2nd order polynomial representation. This newly developed tool has been successfully tested for a detailed 40 degree sector benchmark of the International Thermonuclear Experimental Reactor (ITER). The calculations included determining the poloidal variation of the neutron wall loading, flux and nuclear heating in the divertor components, nuclear heating in toroidal field coils, and radiation streaming in the mid-plane port. The tool has been applied to perform 3-D nuclear analysis for several fusion designs including the ARIES Compact Stellarator (ARIES-CS), the High Average Power Laser (HAPL) inertial fusion power plant, and ITER first wall/shield (FWS) modules. The ARIES-CS stellarator has a first wall shape and a plasma profile that varies toroidally within each field period compared to the uniform toroidal shape in tokamaks. Such variation cannot be modeled analytically in the standard MCNP code. The impact of the complex helical geometry and the non-uniform blanket and divertor on the overall tritium breeding ratio and total nuclear heating was determined. In addition, we calculated the neutron wall loading variation in
Energy Technology Data Exchange (ETDEWEB)
Teixeira, Paulo Cleber Mendonca
2002-12-01
In this study, an analytical solution of the neutron transport equation in an annular reactor is presented with a short and rotating neutron source of the type S(x) {delta} (x- Vt), where V is the speed of annular pulsed reactor. The study is an extension of a previous study by Williams [12] carried out with a pulsed source of the type S(x) {delta} (t). In the new concept of annular pulsed reactor designed to produce continuous high flux, the core consists of a subcritical annular geometry pulsed by a rotating modulator, producing local super prompt critical condition, thereby giving origin to a rotating neutron pulse. An analytical solution is obtained by opening up of the annular geometry and applying one energy group transport theory in one dimension using applied mathematical techniques of Laplace transform and Complex Variables. The general solution for the flux consists of a fundamental mode, a finite number of harmonics and a transient integral. A condition which limits the number of harmonics depending upon the circumference of the annular geometry has been obtained. Inverse Laplace transform technique is used to analyse instability condition in annular reactor core. A regenerator parameter in conjunction with perimeter of the ring and nuclear properties is used to obtain stable and unstable harmonics and to verify if these exist. It is found that the solution does not present instability in the conditions stated in the new concept of annular pulsed reactor. (author)
Optimization of spring exchange coupled ferrites, studied by in situ neutron diffraction
DEFF Research Database (Denmark)
Ahlburg, Jakob; Christensen, Mogens; Granados-Miralles, Cecilia
have a spin it will also be possible to measure a magnetic signal and investigate the exchange-coupling. After the reduction the samples was furthermore investigated using powder x-ray diffraction and VSM (vibrating sample magnetometer). To understand the reaction mechanism, a series of experiments...
Fowler, S. J.; Driesner, T.; Kulik, D.; Wagner, T.
2010-12-01
We present a novel computational tool for modelling temporally and spatially varying chemical interactions between hydrothermal fluids and rocks that may affect the long-term performance of geothermal reservoirs. The code is written in C++. It incorporates fluid-rock interaction and scale formation self-consistently, via a modular coupling approach that combines the Complex System Modelling Platform (CSMP++) code for fluid flow in porous and fractured media (Matthai et al., 2007) with the numerical kernel (GEMIPM2K) of the GEM-Selektor Gibbs free energy minimization package (Kulik, Wagner et al., 2007). CSMP++ uses finite element-finite volume spatial discretization, implicit or explicit time discretization, and an operator splitting approach to solve equations. The GEM-Selektor package supports a wide range of equation of state and activity models, facilitating calculation of complex fluid-mineral equilibria. Coupled code input includes temperature, pressure, a charge balance, and total amounts of system chemical elements, as well as domain and boundary condition specifications. Speciation, thermodynamic, and physical properties of the system are output. Critical advantages of the coupled code compared to existing hydrothermal reactive transport models are: (1) simultaneous consideration of complex solid solutions (e.g., clay minerals) and non-ideal aqueous solutions (GEMIPM2K), and (2) a discretization scheme that can be applied to mass and heat transport in irregular, geologically realistic geometries (CSMP++). Each coupled simulation results in a thermodynamically-based description of the geochemical and physical state of a hydrothermal system evolving along a complex P-T-X path. The code design allows for efficient and flexible incorporation of numerical and thermodynamic database improvements. We apply the coupled code to a number of geologic applications to test its accuracy and performance. Kulik, D., Wagner, T. et al. (2007). GEM-Selektor (GEMS-PSI) home
Analysis of the sodium recirculation theory of solute-coupled water transport in small intestine
DEFF Research Database (Denmark)
Larsen, Erik Hviid; Sørensen, Jakob Balslev; Sørensen, Jens Nørkaer
2002-01-01
Our previous mathematical model of solute-coupled water transport through the intestinal epithelium is extended for dealing with electrolytes rather than electroneutral solutes. A 3Na+-2K+ pump in the lateral membranes provides the energy-requiring step for driving transjunctional and translateral...... permeabilities and pump constants of fluxes of water and electrolytes, volumes and ion concentrations of cell and lateral intercellular space (lis), and membrane potentials and conductances. Simulating physiological bioelectrical features together with cellular and paracellular fluxes of the sodium ion......, computations predict that the concentration differences between lis and bathing solutions are small for all three ions. Nevertheless, the diffusion fluxes of the ions out of lis significantly exceed their mass transports. It is concluded that isotonic transport requires recirculation of all three ions...
Separation of some metal ions using coupled transport supported liquid membranes
International Nuclear Information System (INIS)
Chaudhary, M.A.
1993-01-01
Liquid membrane extraction processes has become very popular due to their superiority in many ways over other separation techniques. In coupled transport membranes the metal ions can be transported across the membrane against their concentration gradient under the influence of chemical potential difference. Liquid membranes consisting of a carrier-cum-diluent, supported in microporous polymeric hydrophobic films have been studied for transport of metal ions like U(VI), Cr(VI), Be(II), V(V), Ti(IV), Zn(II), Cd(II), Hf(IV), W(VI), and Co(II). The present paper presents basic data with respect to flux and permeabilities of these metal ions across membranes based on experimental results and theoretical equations, using different carriers and diluents and provides a brief reference to possibility of such membranes for large scale applications. (author)
An analytical model for predicting transport in a coupled vadose/phreatic system
International Nuclear Information System (INIS)
Tomasko, D.
1997-05-01
A simple analytical model is presented for predicting the transport of a contaminant in both the unsaturated (vadose) and saturated (phreatic) zones following a surficial spill. The model incorporates advection, dispersion, adsorption, and first-order decay in both zones and couples the transport processes at the water table. The governing equation is solved by using the method of Laplace transforms, with numerical inversion of the Laplace space equation for concentration. Because of the complexity of the functional form for the Laplace space solution, a numerical methodology using the real and imaginary parts of a Fourier series was implemented. To reduce conservatism in the model, dilution at the water table was also included. Verification of the model is demonstrated by its ability to reproduce the source history at the surface and to replicate appropriate one-dimensional transport through either the vadose or phreatic zone. Because of its simplicity and lack of detailed input data requirements, the model is recommended for scoping calculations
Wissmeier, L. C.; Barry, D. A.
2009-12-01
Computer simulations of water availability and quality play an important role in state-of-the-art water resources management. However, many of the most utilized software programs focus either on physical flow and transport phenomena (e.g., MODFLOW, MT3DMS, FEFLOW, HYDRUS) or on geochemical reactions (e.g., MINTEQ, PHREEQC, CHESS, ORCHESTRA). In recent years, several couplings between both genres of programs evolved in order to consider interactions between flow and biogeochemical reactivity (e.g., HP1, PHWAT). Software coupling procedures can be categorized as ‘close couplings’, where programs pass information via the memory stack at runtime, and ‘remote couplings’, where the information is exchanged at each time step via input/output files. The former generally involves modifications of software codes and therefore expert programming skills are required. We present a generic recipe for remotely coupling the PHREEQC geochemical modeling framework and flow and solute transport (FST) simulators. The iterative scheme relies on operator splitting with continuous re-initialization of PHREEQC and the FST of choice at each time step. Since PHREEQC calculates the geochemistry of aqueous solutions in contact with soil minerals, the procedure is primarily designed for couplings to FST’s for liquid phase flow in natural environments. It requires the accessibility of initial conditions and numerical parameters such as time and space discretization in the input text file for the FST and control of the FST via commands to the operating system (batch on Windows; bash/shell on Unix/Linux). The coupling procedure is based on PHREEQC’s capability to save the state of a simulation with all solid, liquid and gaseous species as a PHREEQC input file by making use of the dump file option in the TRANSPORT keyword. The output from one reaction calculation step is therefore reused as input for the following reaction step where changes in element amounts due to advection
International Nuclear Information System (INIS)
Abadie, P.
2004-01-01
TN trademark Resin Vyal, a patent pending composite, is a new neutron shielding material developed by COGEMA LOGISTICS for transport/storage casks of radioactive materials (spent fuel, reprocessed fuel,..). This material is composed of a thermosetting resin (vinylester resin in solution of styrene) and two mineral fillers (alumine hydrate and zinc borate). Its shielding ability for neutron radiation is related to a high hydrogen content (for slowing down neutrons) and a high boron content (for absorbing neutrons). Source of hydrogen is organic matrix (resin) and alumine hydrate; source of boron is zinc borate. Atomic concentrations are equal to 5.10 22 at/cm 3 for hydrogen and 9.10 20 at/cm 3 for boron. Due to the flame retardant character of components, the final material has a good fire resistance (it is auto-extinguishable). Its density is equal to 1,8. The manufacturing process of these material is easy: it consists in mixing the fillers and pouring in-situ (in cask); so, the curing of this composite, which leads to a three-dimensional structure, takes place at ambient temperature. Temperature resistance of this material was evaluated by performing exposition tests of samples at different temperatures (150 C to 170 C). According to tests results, its maximal temperature of use was confirmed equal to 160 C
Green's-function approach to the atmospheric albedo neutron-transport problem. Master's thesis
Energy Technology Data Exchange (ETDEWEB)
Culp, D.R.
1991-03-01
This study investigated the reflection of neutron radiation off of the earth's upper atmosphere, with the goal of generating a quick-running computer algorithm to estimate the albedo free field flux at any point above the atmosphere. This thesis involved analytic development in the construction of the algorithm and employed Monte Carlo simulation for generating the energy and angle distributions of the reflected radiation. The Green's function approach to modeling the neutron transport process involved approximating each energy bin of the source spectrum as a Dirac pulse in energy and summing the contribution from each source bin. The computer program integrates over the surface of the atmosphere and uses the Monte Carlo data to calculate the albedo flux at any specified time and location. Run time was maximum of six minutes for a flux calculation, but a gain on the order of one thousand should be achieved on mainframe computer systems. The albedo flux from an instantaneous point source raises quickly to a maximum and then falls off over time. Albedo fluxes as much as 10(-16) (neutrons/square cm sec/source neutron) were calculated. The accuracy of the algorithm is greatly affected by the fineness of the energy bins involved.
Energy Technology Data Exchange (ETDEWEB)
Ford, W.E. III; Roussin, R.W.; Petrie, L.M.; Diggs, B.R.; Comolander, H.E.
1979-01-01
Contents of the IBM version of the APMX system distributed by the Radiation Shielding Information Center (APMX-II) are described. Sample problems which demonstrate the procedure for implementing AMPX-II modules to generate point cross sections; generate multigroup neutron, photon production, and photon interaction cross sections for various transport codes; collapse multigroup cross sections; check, edit, and punch multigroup cross sections; and execute a one-dimensional discrete ordinates transport calculation are detailed. 25 figures, 9 tables.
Numerical solutions of the monoenergetic neutron transport equation with anisotropic scattering
International Nuclear Information System (INIS)
Dahl, B.
1985-01-01
The Boltzmann equation for monoenergetic neutrons has been solved numerically with high accuracy for homogeneous slabs and spheres with various degree of linear anisotropy. Vacuum boundary conditions are used. The numerical method is based on previous work by Carlvik. Benchmark values of the criticality factor and higher order eigenvalues are given for multiplying systems of thickness or diameter from 10 -5 to 20 mean free paths and with anisotropy coefficients from 0.0 to 0.3. For slab geometry, both even and odd mode eigenvalues are treated. With increasing anisotropy, an increasing number of complex eigenvalues is observer. The total flux is calculated from the eigenvector and tables of the fundamental mode flux are given. Accurate extrapolation distances are derived for various dimensions and anisotropy coefficients from our eigenvalue results on slabs and spheres and from the work by Sanchez on infinite cylinders.The time eigenvalue spectrum in subcritical systems has also been studied. First, the connection between the eigenvalues arising from the time dependent and stationary transport equation is established. Based on this, the spectrum of real time eigenvalues in slabs and spheres is calculated. For spheres, the existence of complex time eigenvalues in the region beyond the value corresponding to the Corngold limit is numerically established. The presence of such eigenvalues has earlier not been proved. It is further shown that the Boltzmann equation for a sphere is significantly simplified when the decay constant is at the Corngold limit. The spectrum of sphere diameters corresponding to this decay constant is calculated for various linear anisotropies, and detailed numerical results are given. (Author)
International Nuclear Information System (INIS)
Goncalves, Glenio A.; Bodmann, Bardo; Bogado, Sergio; Vilhena, Marco T.
2008-01-01
Analytical solutions for neutron transport in cylindrical geometry is available for isotropic problems, but to the best of our knowledge for anisotropic problems are not available, yet. In this work, an analytical solution for the neutron transport equation in an infinite cylinder assuming anisotropic scattering is reported. Here we specialize the solution, without loss of generality, for the linearly anisotropic problem using the combined decomposition and HTS N methods. The key feature of this method consists in the application of the decomposition method to the anisotropic problem by virtue of the fact that the inverse of the operator associated to isotropic problem is well know and determined by the HTS N approach. So far, following the idea of the decomposition method, we apply this operator to the integral term, assuming that the angular flux appearing in the integrand is considered to be equal to the HTS N solution interpolated by polynomial considering only even powers. This leads to the first approximation for an anisotropic solution. Proceeding further, we replace this solution for the angular flux in the integral and apply again the inverse operator for the isotropic problem in the integral term and obtain a new approximation for the angular flux. This iterative procedure yields a closed form solution for the angular flux. This methodology can be generalized, in a straightforward manner, for transport problems with any degree of anisotropy. For the sake of illustration, we report numerical simulations for linearly anisotropic transport problems. (author)
Andreasen, Mie; Jensen, Karsten H.; Desilets, Darin; Zreda, Marek; Bogena, Heye R.; Looms, Majken C.
2017-04-01
Cosmic-ray neutron intensity is inversely correlated to all hydrogen present in the upper decimeters of the subsurface and the first few hectometers of the atmosphere above the ground surface. This correlation forms the base of the cosmic-ray neutron soil moisture estimation method. The method is, however, complicated by the fact that several hydrogen pools other than soil moisture affect the neutron intensity. In order to improve the cosmic-ray neutron soil moisture estimation method and explore the potential for additional applications, knowledge about the environmental effect on cosmic-ray neutron intensity is essential (e.g., the effect of vegetation, litter layer and soil type). In this study the environmental effect is examined by performing a sensitivity analysis using neutron transport modeling. We use a neutron transport model with various representations of the forest and different parameters describing the subsurface to match measured height profiles and time series of thermal and epithermal neutron intensities at a field site in Denmark. Overall, modeled thermal and epithermal neutron intensities are in satisfactory agreement with measurements; however, the choice of forest canopy conceptualization is found to be significant. Modeling results show that the effect of canopy interception, soil chemistry and dry bulk density of litter and mineral soil on neutron intensity is small. On the other hand, the neutron intensity decreases significantly with added litter-layer thickness, especially for epithermal neutron energies. Forest biomass also has a significant influence on the neutron intensity height profiles at the examined field site, altering both the shape of the profiles and the ground-level thermal-to-epithermal neutron ratio. This ratio increases with increasing amounts of biomass, and was confirmed by measurements from three sites representing agricultural, heathland and forest land cover. A much smaller effect of canopy interception on the ground
Energy Technology Data Exchange (ETDEWEB)
Cui, Shijie; Zhang, Dalin, E-mail: dlzhang@mail.xjtu.edu.cn; Cheng, Jie; Tian, Wenxi; Su, G.H.
2017-01-15
As one of the candidate tritium breeding blankets for Chinese Fusion Engineering Test Reactor (CFETR), a conceptual structure of the helium cooled solid breeder blanket has recently been proposed. The neutronic, thermal-hydraulic and mechanical characteristics of the blanket directly affect its tritium breeding and safety performance. Therefore, neutronic/thermal-hydraulic/mechanical coupling analyses are of vital importance for a reliable blanket design. In this work, first, three-dimensional neutronics analysis and optimization of the typical outboard equatorial blanket module (No. 12) were performed for the comprehensive optimal scheme. Then, thermal and fluid dynamic analyses of the scheme under both normal and critical conditions were performed and coupled with the previous neutronic calculation results. With thermal-hydraulic boundaries, thermo-mechanical analyses of the structure materials under normal, critical and blanket over-pressurization conditions were carried out. In addition, several parametric sensitivity studies were also conducted to investigate the influences of the main parameters on the blanket temperature distributions. In this paper, the coupled analyses verify the reasonability of the optimized conceptual design preliminarily and can provide an important reference for the further analysis and optimization design of the CFETR helium cooled solid breeder blanket.
Energy Technology Data Exchange (ETDEWEB)
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.
Analysis of the sodium recirculation theory of solute-coupled water transport in small intestine.
Larsen, Erik Hviid; Sørensen, Jakob Balslev; Sørensen, Jens Nørkaer
2002-07-01
Our previous mathematical model of solute-coupled water transport through the intestinal epithelium is extended for dealing with electrolytes rather than electroneutral solutes. A 3Na+-2K+ pump in the lateral membranes provides the energy-requiring step for driving transjunctional and translateral flows of water across the epithelium with recirculation of the diffusible ions maintained by a 1Na+-1K+-2Cl- cotransporter in the plasma membrane facing the serosal compartment. With intracellular non-diffusible anions and compliant plasma membranes, the model describes the dependence on membrane permeabilities and pump constants of fluxes of water and electrolytes, volumes and ion concentrations of cell and lateral intercellular space (lis), and membrane potentials and conductances. Simulating physiological bioelectrical features together with cellular and paracellular fluxes of the sodium ion, computations predict that the concentration differences between lis and bathing solutions are small for all three ions. Nevertheless, the diffusion fluxes of the ions out of lis significantly exceed their mass transports. It is concluded that isotonic transport requires recirculation of all three ions. The computed sodium recirculation flux that is required for isotonic transport corresponds to that estimated in experiments on toad small intestine. This result is shown to be robust and independent of whether the apical entrance mechanism for the sodium ion is a channel, a SGLT1 transporter driving inward uphill water flux, or an electroneutral Na+-K+-2Cl- cotransporter.
Modeling the coupled mechanics, transport, and growth processes in collagen tissues.
Energy Technology Data Exchange (ETDEWEB)
Holdych, David J.; Nguyen, Thao D.; Klein, Patrick A.; in' t Veld, Pieter J.; Stevens, Mark Jackson
2006-11-01
The purpose of this project is to develop tools to model and simulate the processes of self-assembly and growth in biological systems from the molecular to the continuum length scales. The model biological system chosen for the study is the tendon fiber which is composed mainly of Type I collagen fibrils. The macroscopic processes of self-assembly and growth at the fiber scale arise from microscopic processes at the fibrillar and molecular length scales. At these nano-scopic length scales, we employed molecular modeling and simulation method to characterize the mechanical behavior and stability of the collagen triple helix and the collagen fibril. To obtain the physical parameters governing mass transport in the tendon fiber we performed direct numerical simulations of fluid flow and solute transport through an idealized fibrillar microstructure. At the continuum scale, we developed a mixture theory approach for modeling the coupled processes of mechanical deformation, transport, and species inter-conversion involved in growth. In the mixture theory approach, the microstructure of the tissue is represented by the species concentration and transport and material parameters, obtained from fibril and molecular scale calculations, while the mechanical deformation, transport, and growth processes are governed by balance laws and constitutive relations developed within a thermodynamically consistent framework.
Comolli, A.; Dentz, M.
2015-12-01
Solute transport in geological media is in general non-Fickian as it cannot be explained in terms of equivalent homogeneous media. This anomalous character can be traced back to the existence of multiscale heterogeneity and strong correlations within the medium. Here we investigate the impact of fast heterogeneous mass transfer properties as represented by a spatially varying retardation coefficient (mass exchange between mobile and immobile regions, linear sorption-desorption reactions, variable porosity). In order to estimate the effects of spatial correlation, and disorder distribution on the average transport, we consider 2D media characterized by complex multiscale geometries and point distributions of retardation of increasing heterogeneity. Within a Lagrangian framework, we coarse-grain the Langevin equation for the transport of solute particles due to advection and diffusion in the heterogeneous medium. The large-scale transport properties are derived within a stochastic modeling approach by ensemble averaging of the coarse-grained Langevin equation . This approach shows that the effective particle motion can be described by a coupled CTRW that is fully parametrized by the distribution of the retardation coefficient and the spatial medium organization. This allows for the explicit relation of the heterogeneous medium properties to observed anomalous transport in terms of solute dispersion, breakthrough curves and spatial concentration profiles.
Directory of Open Access Journals (Sweden)
Armen N. Kocharian
2016-05-01
Full Text Available Rashba spin-orbit effects and electron correlations in the two-dimensional cylindrical lattices of square geometries are assessed using mesoscopic two-, three- and four-leg ladder structures. Here the electron transport properties are systematically calculated by including the spin-orbit coupling in tight binding and Hubbard models threaded by a magnetic flux. These results highlight important aspects of possible symmetry breaking mechanisms in square ladder geometries driven by the combined effect of a magnetic gauge field spin-orbit interaction and temperature. The observed persistent current, spin and charge polarizations in the presence of spin-orbit coupling are driven by separation of electron and hole charges and opposite spins in real-space. The modeled spin-flip processes on the pairing mechanism induced by the spin-orbit coupling in assembled nanostructures (as arrays of clusters engineered in various two-dimensional multi-leg structures provide an ideal playground for understanding spatial charge and spin density inhomogeneities leading to electron pairing and spontaneous phase separation instabilities in unconventional superconductors. Such studies also fall under the scope of current challenging problems in superconductivity and magnetism, topological insulators and spin dependent transport associated with numerous interfaces and heterostructures.
International Nuclear Information System (INIS)
Thomas, Sarah A; Uhoya, Walter O; Tsoi, Georgiy M; Wenger, Lowell E; Vohra, Yogesh K; Chesnut, Gary N; Weir, Samuel T; Tulk, Christopher A; Dos Santos, Antonio M
2012-01-01
Neutron diffraction and electrical transport measurements have been made on the heavy rare earth metal holmium at high pressures and low temperatures in order to elucidate its transition from a paramagnetic (PM) to a helical antiferromagnetic (AFM) ordered phase as a function of pressure. The electrical resistance measurements show a change in the resistance slope as the temperature is lowered through the antiferromagnetic Néel temperature. The temperature of this antiferromagnetic transition decreases from approximately 122 K at ambient pressure at a rate of -4.9 K GPa -1 up to a pressure of 9 GPa, whereupon the PM-to-AFM transition vanishes for higher pressures. Neutron diffraction measurements as a function of pressure at 89 and 110 K confirm the incommensurate nature of the phase transition associated with the antiferromagnetic ordering of the magnetic moments in a helical arrangement and that the ordering occurs at similar pressures as determined from the resistance results for these temperatures. (paper)
Thomas, Sarah A; Uhoya, Walter O; Tsoi, Georgiy M; Wenger, Lowell E; Vohra, Yogesh K; Chesnut, Gary N; Weir, Samuel T; Tulk, Christopher A; dos Santos, Antonio M
2012-05-30
Neutron diffraction and electrical transport measurements have been made on the heavy rare earth metal holmium at high pressures and low temperatures in order to elucidate its transition from a paramagnetic (PM) to a helical antiferromagnetic (AFM) ordered phase as a function of pressure. The electrical resistance measurements show a change in the resistance slope as the temperature is lowered through the antiferromagnetic Néel temperature. The temperature of this antiferromagnetic transition decreases from approximately 122 K at ambient pressure at a rate of -4.9 K GPa(-1) up to a pressure of 9 GPa, whereupon the PM-to-AFM transition vanishes for higher pressures. Neutron diffraction measurements as a function of pressure at 89 and 110 K confirm the incommensurate nature of the phase transition associated with the antiferromagnetic ordering of the magnetic moments in a helical arrangement and that the ordering occurs at similar pressures as determined from the resistance results for these temperatures.
Frampton, Andrew; Destouni, Georgia
2016-04-01
In cold regions, flow in the unsaturated zone is highly dynamic with seasonal variability and changes in temperature, moisture, and heat and water fluxes, all of which affect ground freeze-thaw processes and influence transport of inert and reactive waterborne substances. In arctic permafrost environments, near-surface groundwater flow is further restricted to a relatively shallow and seasonally variable active layer, confined by perennially frozen ground below. The active layer is typically partially saturated with ice, liquid water and air, and is strongly dependent on seasonal temperature fluctuations, thermal forcing and infiltration patterns. Here there is a need for improved understanding of the mechanisms controlling subsurface solute transport in the partially saturated active layer zone. Studying solute transport in cold regions is relevant to improve the understanding of how natural and anthropogenic pollution may change as activities in arctic and sub-arctic regions increase. It is also particularly relevant for understanding how dissolved carbon is transported in coupled surface and subsurface hydrological systems under climate change, in order to better understand the permafrost-hydrological-carbon climate feedback. In this contribution subsurface solute transport under surface warming and degrading permafrost conditions is studied using a physically based model of coupled cryotic and hydrogeological flow processes combined with a particle tracking method. Changes in subsurface water flows and solute transport travel times are analysed for different modelled geological configurations during a 100-year warming period. Results show that for all simulated cases, the minimum and mean travel times increase non-linearly with warming irrespective of geological configuration and heterogeneity structure. The travel time changes are shown to depend on combined warming effects of increase in pathway length due to deepening of the active layer, reduced transport
Neutronic/Thermalhydraulic Coupling Technigues for Sodium Cooled Fast Reactor Simulations
Energy Technology Data Exchange (ETDEWEB)
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.
Neutronic/Thermal-hydraulic Coupling Technigues for Sodium Cooled Fast Reactor Simulations
International Nuclear Information System (INIS)
Ragusa, Jean; Siegel, Andrew; Ruggieri, Jean-Michel
2010-01-01
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.
TART96: a coupled neutron-photon 3-D, combinatorial geometry Monte Carlo transport code
International Nuclear Information System (INIS)
Cullen, D.E.
1996-11-01
The original TARTND has been used and distributed from LLNL for many years. TART95, released in July 1995, was the first version of the code designed to be used on virtually any computer. TART96 is designed to extend the general utility of the code to more areas of application, by concentrating on improving the physics used by the code. TART96 completely supersedes all older versions of TART, and it is strongly recommended that users only use the most recent version of TART96 and its data files
Neutronics of pulsed spallation neutron sources
Watanabe, N
2003-01-01
Various topics and issues on the neutronics of pulsed spallation neutron sources, mainly for neutron scattering experiments, are reviewed to give a wide circle of readers a better understanding of these sources in order to achieve a high neutronic performance. Starting from what neutrons are needed, what the spallation reaction is and how to produce slow-neutrons more efficiently, the outline of the target and moderator neutronics are explained. Various efforts with some new concepts or ideas have already been devoted to obtaining the highest possible slow-neutron intensity with desired pulse characteristics. This paper also reviews the recent progress of such efforts, mainly focused on moderator neutronics, since moderators are the final devices of a neutron source, which determine the source performance. Various governing parameters for neutron-pulse characteristics such as material issues, geometrical parameters (shape and dimensions), the target-moderator coupling scheme, the ortho-para-hydrogen ratio, po...
Mészáros, Cs.; Kirschner, I.; Bálint, Á.
2014-07-01
A general description of the basic system of ordinary differential equations of coupled transport processes is given within framework of a linear approximation and treated by tools of matrix analysis and group representation theory. It is shown that the technique of hyperdyads directly generalizes the method of simple dyadic decomposition of operators used earlier in the classical linear irreversible thermodynamics and leads to possible new applications of the concept of quasi-polynomials at descriptions of coupled transport processes.
Electrical and thermal transport in the quasiatomic limit of coupled Luttinger liquids
Szasz, Aaron; Ilan, Roni; Moore, Joel E.
2017-02-01
We introduce a new model for quasi-one-dimensional materials, motivated by intriguing but not yet well-understood experiments that have shown two-dimensional polymer films to be promising materials for thermoelectric devices. We consider a two-dimensional material consisting of many one-dimensional systems, each treated as a Luttinger liquid, with weak (incoherent) coupling between them. This approximation of strong interactions within each one-dimensional chain and weak coupling between them is the "quasiatomic limit." We find integral expressions for the (interchain) transport coefficients, including the electrical and thermal conductivities and the thermopower, and we extract their power law dependencies on temperature. Luttinger liquid physics is manifested in a violation of the Wiedemann-Franz law; the Lorenz number is larger than the Fermi liquid value by a factor between γ2 and γ4, where γ ≥1 is a measure of the electron-electron interaction strength in the system.
Capponi, S; Arbe, A; Alvarez, F; Colmenero, J; Frick, B; Embs, J P
2009-11-28
Quasielastic neutron scattering experiments (time-of-flight, neutron spin echo, and backscattering) on protonated poly(vinyl methyl ether) (PVME) have revealed the hydrogen dynamics above the glass-transition temperature. Fully atomistic molecular dynamics simulations properly validated with the neutron scattering results have allowed further characterization of the atomic motions accessing the correlation functions directly in real space. Deviations from Gaussian behavior are found in the high-momentum transfer range, which are compatible with the predictions of mode coupling theory (MCT). We have applied the MCT phenomenological version to the self-correlation functions of PVME atoms calculated from our simulation data, obtaining consistent results. The unusually large value found for the lambda-exponent parameter is close to that recently reported for polybutadiene and simple polymer models with intramolecular barriers.
Line photon transport in a non-homogeneous plasma using radiative coupling coefficients
International Nuclear Information System (INIS)
Florido, R.; Gil, J.M.; Rodriguez, R.; Rubiano, J.G.; Martel, P.; Florido, R.; Gil, J.M.; Rodriguez, R.; Rubiano, J.G.; Martel, P.; Minguez, E.
2006-01-01
We present a steady-state collisional-radiative model for the calculation of level populations in non-homogeneous plasmas with planar geometry. The line photon transport is taken into account following an angle- and frequency-averaged escape probability model. Several models where the same approach has been used can be found in the literature, but the main difference between our model and those ones is that the details of geometry are exactly treated in the definition of coupling coefficients and a local profile is taken into account in each plasma cell. (authors)
Reactive transport modeling of coupled inorganic and organic processes in groundwater
Energy Technology Data Exchange (ETDEWEB)
Brun, Adam
1997-12-31
The main goals of this project are to develop and apply a reactive transport code for simulation of coupled organic and inorganic processes in the pollution plume in the ground water down-gradient from the Vejen landfill, Denmark. The detailed field investigations in this aquifer have previously revealed a complex pattern of strongly interdependent organic and inorganic processes. These processes occur simultaneously in a flow and transport system where the mixing of reactive species is influenced by the rather complex geology in the vicinity of the landfill. The removal of organic matter is influenced by the presence of various electron acceptors that also are involved in various inorganic geochemical reactions. It was concluded from the investigations that degradation of organic matter, complexation, mineral precipitation and dissolution, ion-exchange and inorganic redox reactions, as a minimum, should be included in the formulation of the model. The coupling of the organic and inorganic processes is developed based on a literature study. All inorganic processes are as an approximation described as equilibriumm processes. The organic processes are described by a maximum degradation rate that is decreased according to the availability of the participants in the processes, the actual pH, and the presence of inhibiting species. The reactive transport code consists of three separate codes, a flow and transport code, a geochemical code, and a biodegradation code. An iterative solution scheme couples the three codes. The coupled code was successfully verified for simple problems for which analytical solutions exist. For more complex problems the code was tested on synthetic cases and expected plume behavior was successfully simulated. Application of the code to the Vejen landfill aquifer was successful to the degree that the redox zonation down-gradient from the landfill was simulated correctly and that several of the simulated plumes showed a reasonable agreement with
Application of the Arbitrarily High Order Method to Coupled Electron Photon Transport
International Nuclear Information System (INIS)
Duo, Jose Ignacio
2004-01-01
This work is about the application of the Arbitrary High Order Nodal Method to coupled electron photon transport.A Discrete Ordinates code was enhanced and validated which permited to evaluate the advantages of using variable spatial development order per particle.The results obtained using variable spatial development and adaptive mesh refinement following an a posteriori error estimator are encouraging.Photon spectra for clinical accelerator target and, dose and charge depositio profiles are simulated in one-dimensional problems using cross section generated with CEPXS code.Our results are in good agreement with ONELD and MCNP codes
Driving-induced metamorphosis of transport in arrays of coupled resonators
Li, Huanan; Kottos, Tsampikos; Shapiro, Boris
2018-02-01
We propose a driving scheme, wherein different parts of a system are driven with different, generally incommensurate, frequencies. Such driving provides a flexible handle to control various properties of the system and to obtain new types of effective (static) Hamiltonians with arbitrary static on-site potential, be it deterministic or random. This allows us to obtain reconfigurable changes in transport, from ballistic to localized (including sub- and superdiffusion), depending on the driving protocol. The versatile reconfigurability extends also to scattering from (locally) driven extended targets. We demonstrate our scheme using an analytically solvable example of a one-dimensional tight-binding chain with appropriately driven couplings between nearby sites.
Use of coupled geochemical and transport calculations for nuclear waste problems
International Nuclear Information System (INIS)
Neretnieks, I.; Nyman, C.
1993-01-01
The dissolution and migration of radionuclides from a final repository for radioactive waste is complex chemically. The dissolution and release rates depend on the chemistry and on advective and diffusive transport. The presence of the waste may influence the chemistry by e.g. radiolysis which may change the solubility of several important nuclides as well as of the uranium of the waste matrix itself, if it is spent fuel. Redox and other fronts may evolve in the backfill and rock around the waste. Similar processes have been observed in natural systems. The evolution of a redox front in a uranium mine over long times and distances is used as an example to test the capability and illustrate the use of some coupled transport and chemical codes
International Nuclear Information System (INIS)
Michenaud, J.-P.; Lenoir, B.
2007-01-01
Magneto-transport properties of bismuth and graphite are reconsidered taking into account the magneto-elastic coupling. This coupling effect is responsible for a change of carrier density as well as a change of crystallographic symmetry. The change of carrier density can lead to a negative component in the magnetoresistance, when the change is positive
To investigate the coupled effects of solution chemistry and vadose zone processes on the mobility of quantum dot (QD) nanoparticles, laboratory scale transport experiments were performed. The complex coupled effects of ionic strength, size of QD aggregates, surface tension, contact angle, infiltrat...
International Nuclear Information System (INIS)
Hayasaka, Hideo
1983-01-01
The thermodynamics and the energy distribution function of the neutron gas in a constant power reactor are considered, taking into account the burn-up of fuel. To separate the secular motion of neutrons owing to fuel burn-up and the microscopic fluctuations of neutrons around this motion, a long time of the order of several months is divided into m equal intervals, and the respective states corresponding to m small time intervals are treated as quasi-stationary states. The local energy distribution function of the neutron gas in the quasi-stationary state is given by a generalized Boltzmann distribution specified by the respective generalized activity coefficient for each subsystem. The effects of fuel burn-up on the respective distribution functions for successive small time intervals are taken into account through various quantities relating to reactor physics, depending upon the fuel burn-up, by successive approximation. (author)
Wu, Yican
2017-01-01
This book provides a systematic and comprehensive introduction to fusion neutronics, covering all key topics from the fundamental theories and methodologies, as well as a wide range of fusion system designs and experiments. It is the first-ever book focusing on the subject of fusion neutronics research. Compared with other nuclear devices such as fission reactors and accelerators, fusion systems are normally characterized by their complex geometry and nuclear physics, which entail new challenges for neutronics such as complicated modeling, deep penetration, low simulation efficiency, multi-physics coupling, etc. The book focuses on the neutronics characteristics of fusion systems and introduces a series of theories and methodologies that were developed to address the challenges of fusion neutronics, and which have since been widely applied all over the world. Further, it introduces readers to neutronics design’s unique principles and procedures, experimental methodologies and technologies for fusion systems...
Energy Technology Data Exchange (ETDEWEB)
Xi, Xi [CNNC Key Laboratory on Nuclear Reactor Thermal Hydraulics Technology, Nuclear Power Institute of China, Chengdu 610041 (China); Xiao, Zejun, E-mail: fabulous_2012@sina.com [CNNC Key Laboratory on Nuclear Reactor Thermal Hydraulics Technology, Nuclear Power Institute of China, Chengdu 610041 (China); Yan, Xiao; Li, Yongliang; Huang, Yanping [CNNC Key Laboratory on Nuclear Reactor Thermal Hydraulics Technology, Nuclear Power Institute of China, Chengdu 610041 (China)
2013-05-15
Highlights: ► CFX and MCNP codes are suitable to calculate the axial power profile of the FA. ► The partition method in the calculation will affect the final result. ► The density feedback has little effect on the axial power profile of CSR1000 FA. -- Abstract: SCWR (super critical water reactor) is one of the IV generation nuclear reactors in the world. In a typical SCWR the water enters the reactor from the cold leg with a temperature of 280 °C and then leaves the core with a temperature of 500 °C. Due to the sharp change in temperature, there is a huge density change of the water along the axial direction of the fuel assembly (FA), which will affect the moderating power of the water. So the axial power distribution of the SCWR FA could be different from the traditional PWR FA.In this paper, it is the first time that the thermal hydraulics code CFX and neutronics code MCNP are used to analyze the axial power distribution of the SCWR FA. First, the factors in the coupled method which could affect the result are analyzed such as the initialization value or the partition method especially in the MCNP code. Then the axial power distribution of the Europe HPLWR FA is obtained by the coupled method with the two codes and the result is compared with that obtained by Waata and Reiss. There is a good agreement among the three kinds of results. At last, this method is used to calculate the axial power distribution of the Chinese SCWR (CSR1000) FA. It is found the axial power profile of the CSR1000 FA is not so sensitive to the change of the moderator density.
Coupled hydrogeological and reactive transport modelling of the Simpevarp area (Sweden)
International Nuclear Information System (INIS)
Molinero, Jorge; Raposo, Juan R.; Galindez, Juan M.; Arcos, David; Guimera, Jordi
2008-01-01
The Simpevarp area is one of the alternative sites being considered for the deep geological disposal of high level radioactive waste in Sweden. In this paper, a coupled regional groundwater flow and reactive solute transport model of the Simpevarp area is presented that integrates current hydrogeological and hydrochemical data of the area. The model simulates the current hydrochemical pattern of the groundwater system in the area. To that aim, a conceptual hydrochemical model was developed in order to represent the dominant chemical processes. Groundwater flow conditions were reproduced by taking into account fluid-density-dependent groundwater flow and regional hydrogeologic boundary conditions. Reactive solute transport calculations were performed on the basis of the velocity field so obtained. The model was calibrated and sensitivity analyses were carried out in order to investigate the effects of heterogeneities of hydraulic conductivity in the subsurface medium. Results provided by the reactive transport model are in good agreement with much of the measured hydrochemical data. This paper emphasizes the appropriateness of the use of reactive solute transport models when water-rock interaction reactions are involved, and demonstrates what powerful tools they are for the interpretation of hydrogeological and hydrochemical data from site geological repository characterization programs, by providing a qualitative framework for data analysis and testing of conceptual assumptions in a process-oriented approach
International Nuclear Information System (INIS)
Liu, X.; Ortoleva, P.
1996-01-01
The use of reaction-transport modeling for reservoir assessment and management in the context of deep well waste injection is evaluated. The study is based on CIRF.A (Chemical Interaction of Rock and Fluid), a fully coupled multiphase flow, contaminant transport, and fluid and mineral reaction model. Although SWIFT (Sandia Waste-Isolation Flow and Transport Model) is often the numerical model of choice, it can not account for chemical reactions involving rock, wastes, and formation fluids and their effects on contaminant transport, rock permeability and porosity, and the integrity of the reservoir and confining units. CIRF.A can simulate all these processes. Two field cases of waste injection were simulated by CIRF.A. Both observation data and simulation results show mineral precipitation in one case and rock dissolution in another case. Precipitation and dissolution change rock porosity and permeability, and hence the pattern of fluid migration. The model is shown to be invaluable in analyzing near borehole and reservoir-scale effects during waste injection and predicting the 10,000 year fate of the waste plume. The benefits of using underpressured compartments as waste repositories were also demonstrated by CIRF.A simulations
Energy Technology Data Exchange (ETDEWEB)
Filho, J. F. P. [Institute de Matematica, Estatistica e Fisica, Universidade Federal do Rio Grande, Av. Italia, s/n, 96203-900 Rio Grande, RS (Brazil); Barichello, L. B. [Institute de Matematica, Universidade Federal do Rio Grande do Sul, Av. Bento Goncalves, 9500, 91509-900 Porto Alegre, RS (Brazil)
2013-07-01
In this work, an analytical discrete ordinates method is used to solve a nodal formulation of a neutron transport problem in x, y-geometry. The proposed approach leads to an important reduction in the order of the associated eigenvalue systems, when combined with the classical level symmetric quadrature scheme. Auxiliary equations are proposed, as usually required for nodal methods, to express the unknown fluxes at the boundary introduced as additional unknowns in the integrated equations. Numerical results, for the problem defined by a two-dimensional region with a spatially constant and isotropically emitting source, are presented and compared with those available in the literature. (authors)
International Nuclear Information System (INIS)
Sahni, D.C.; Sharma, A.
2000-01-01
The integral form of one-speed, spherically symmetric neutron transport equation with isotropic scattering is considered. Two standard problems are solved using normal mode expansion technique. The expansion coefficients are obtained by solving their singular integral equations. It is shown that these expansion coefficients provide a representation of all spherical harmonics moments of the angular flux as a superposition of Bessel functions. It is seen that large errors occur in the computation of higher moments unless we take certain precautions. The reasons for this phenomenon are explained. They throw some light on the failure of spherical harmonics method in treating spherical geometry problems as observed by Aronsson
Ferromagnetic Cu-O-Cu coupling in CaCu3Sn4O12 probed by neutron diffraction.
Kayser, P; Retuerto, M; Sánchez-Benítez, J; Martínez-Lope, M J; Fernández-Díaz, M T; Alonso, J A
2012-12-12
The A-site ordered perovskite oxide with the formula CaCu(3)Sn(4)O(12) has been synthesized in polycrystalline form under moderate pressure conditions (3.5 GPa) in combination with high temperature (1000 °C). This oxide crystallizes in the cubic space group [Formula: see text] (no. 204) with the unit-cell parameter a = 7.64535(6) Å at 300 K. The SnO(6) network is extremely tilted, giving rise to a square planar coordination for Cu(2+) cations. The non-magnetic character of Sn(4+) offers an excellent opportunity to probe the magnetism of Cu(2+) at the A sublattice in CaCu(3)Sn(4)O(12). Magnetic susceptibility shows that this compound is ferromagnetic below T(C) = 10 K, which is an unusual magnetic behaviour in cuprates. This peculiar aspect has been examined by neutron powder diffraction. The refinement of the magnetic structure at 4 K indeed indicates a parallel coupling between Cu(2+) spins with a magnetic moment of 0.5 μ(B)/Cu atom.
Directory of Open Access Journals (Sweden)
C. Mesado
2012-01-01
Full Text Available In nuclear safety analysis, it is very important to be able to simulate the different transients that can occur in a nuclear power plant with a very high accuracy. Although the best estimate codes can simulate the transients and provide realistic system responses, the use of nonexact models, together with assumptions and estimations, is a source of uncertainties which must be properly evaluated. This paper describes a Rod Ejection Accident (REA simulated using the coupled code RELAP5/PARCSv2.7 with a perturbation on the cross-sectional sets in order to determine the uncertainties in the macroscopic neutronic information. The procedure to perform the uncertainty and sensitivity (U&S analysis is a sampling-based method which is easy to implement and allows different procedures for the sensitivity analyses despite its high computational time. DAKOTA-Jaguar software package is the selected toolkit for the U&S analysis presented in this paper. The size of the sampling is determined by applying the Wilks’ formula for double tolerance limits with a 95% of uncertainty and with 95% of statistical confidence for the output variables. Each sample has a corresponding set of perturbations that will modify the cross-sectional sets used by PARCS. Finally, the intervals of tolerance of the output variables will be obtained by the use of nonparametric statistical methods.
Plasma Membrane Na+-Coupled Citrate Transporter (SLC13A5 and Neonatal Epileptic Encephalopathy
Directory of Open Access Journals (Sweden)
Yangzom D. Bhutia
2017-02-01
Full Text Available SLC13A5 is a Na+-coupled transporter for citrate that is expressed in the plasma membrane of specific cell types in the liver, testis, and brain. It is an electrogenic transporter with a Na+:citrate3− stoichiometry of 4:1. In humans, the Michaelis constant for SLC13A5 to transport citrate is ~600 μM, which is physiologically relevant given that the normal concentration of citrate in plasma is in the range of 150–200 μM. Li+ stimulates the transport function of human SLC13A5 at concentrations that are in the therapeutic range in patients on lithium therapy. Human SLC13A5 differs from rodent Slc13a5 in two important aspects: the affinity of the human transporter for citrate is ~30-fold less than that of the rodent transporter, thus making human SLC13A5 a low-affinity/high-capacity transporter and the rodent Slc13a5 a high-affinity/low-capacity transporter. In the liver, SLC13A5 is expressed exclusively in the sinusoidal membrane of the hepatocytes, where it plays a role in the uptake of circulating citrate from the sinusoidal blood for metabolic use. In the testis, the transporter is expressed only in spermatozoa, which is also only in the mid piece where mitochondria are located; the likely function of the transporter in spermatozoa is to mediate the uptake of citrate present at high levels in the seminal fluid for subsequent metabolism in the sperm mitochondria to generate biological energy, thereby supporting sperm motility. In the brain, the transporter is expressed mostly in neurons. As astrocytes secrete citrate into extracellular medium, the potential function of SLC13A5 in neurons is to mediate the uptake of circulating citrate and astrocyte-released citrate for subsequent metabolism. Slc13a5-knockout mice have been generated; these mice do not have any overt phenotype but are resistant to experimentally induced metabolic syndrome. Recently however, loss-of-function mutations in human SLC13A5 have been found to cause severe epilepsy
VITAMIN E: a multipurpose ENDF/B-V coupled neutron-gamma cross section library
International Nuclear Information System (INIS)
Barhen, J.; Cacuci, D.G.; Ford, W.E. III; Roussin, R.W.; Wagschal, J.J.; Weisbin, C.R.; White, J.E.; Wright, R.Q.
1979-01-01
The US Department of Energy Office of Fusion Energy and the Division of Reactor Research and Technology jointly sponsored the development of a coupled fine-group cross section library (VITAMIN-C). The experience gained in the generation, validation, and utilization of the VITAMIN-C library along with its broad range of applicability has led to the request for updating this data set using ENDF/B-V. Additional support in this regard has been provided by the Defense Nuclear Agency (DNA) and by EPRI in support of weapons analyses and light water reactor shielding and dosimetry problems, respectively. The rationale for developing the multipurpose ENDF/B-V-based VITAMIN-E library is presented, with special emphasis on new models used in the data generation algorithms. The library specifications and testing procedures are also discussed in detail. The distribution of the VITAMIN-E library is currently subject to the same restrictions as the distribution of the ENDF/B-V data. 2 tables
Energy Technology Data Exchange (ETDEWEB)
O' Brien, M. J.; Brantley, P. S.
2015-01-20
In order to run Monte Carlo particle transport calculations on new supercomputers with hundreds of thousands or millions of processors, care must be taken to implement scalable algorithms. This means that the algorithms must continue to perform well as the processor count increases. In this paper, we examine the scalability of:(1) globally resolving the particle locations on the correct processor, (2) deciding that particle streaming communication