International Nuclear Information System (INIS)
Berthoud, Georges; Ducros, Gerard; Feron, Damien; Guerin, Yannick; Latge, Christian; Limoge, Yves; Santarini, Gerard; Seiler, Jean-Marie; Vernaz, Etienne; Coste-Delclaux, Mireille; M'Backe Diop, Cheikh; Nicolas, Anne; Andrieux, Catherine; Archier, Pascal; Baudron, Anne-Marie; Bernard, David; Biaise, Patrick; Blanc-Tranchant, Patrick; Bonin, Bernard; Bouland, Olivier; Bourganel, Stephane; Calvin, Christophe; Chiron, Maurice; Damian, Frederic; Dumonteil, Eric; Fausser, Clement; Fougeras, Philippe; Gabriel, Franck; Gagnier, Emmanuel; Gallo, Daniele; Hudelot, Jean-Pascal; Hugot, Francois-Xavier; Dat Huynh, Tan; Jouanne, Cedric; Lautard, Jean-Jacques; Laye, Frederic; Lee, Yi-Kang; Lenain, Richard; Leray, Sylvie; Litaize, Olivier; Magnaud, Christine; Malvagi, Fausto; Mijuin, Dominique; Mounier, Claude; Naury, Sylvie; Nicolas, Anne; Noguere, Gilles; Palau, Jean-Marc; Le Pallec, Jean-Charles; Peneliau, Yannick; Petit, Odile; Poinot-Salanon, Christine; Raepsaet, Xavier; Reuss, Paul; Richebois, Edwige; Roque, Benedicte; Royer, Eric; Saint-Jean, Cyrille de; Santamarina, Alain; Serot, Olivier; Soldevila, Michel; Tommasi, Jean; Trama, Jean-Christophe; Tsilanizara, Aime; Behar, Christophe; Provitina, Olivier; Lecomte, Michael; Forestier, Alain; Bender, Alexandra; Parisot, Jean-Francois; Finot, Pierre
2013-10-01
This bibliographical note presents a reference book which addresses the study of neutron transport in matter, the study of conditions for a chain reaction and the study of modifications of matter composition due to nuclear reactions. This book presents the main nuclear data, their measurement, assessment and processing, and the spallation. It proposes an overview of methods applied for the study of neutron transport: basic equations and their derived forms, deterministic methods and Monte Carlo method of resolution of the Boltzmann equation, methods of resolution of generalized Bateman equations, methods of time resolution of space kinetics coupled equations. It presents the main calculation codes, discusses the qualification and experimental aspects, and gives an overview of neutron transport applications: neutron transport calculation of reactors, neutron transport coupled with other disciplines, physics of fuel cycle, criticality
Radiation transport calculations for the ANS [Advanced Neutron Source] beam tubes
International Nuclear Information System (INIS)
Engle, W.W. Jr.; Lillie, R.A.; Slater, C.O.
1988-01-01
The Advanced Neutron Source facility (ANS) will incorporate a large number of both radial and no-line-of-sight (NLS) beam tubes to provide very large thermal neutron fluxes to experimental facilities. The purpose of this work was to obtain comparisons for the ANS single- and split-core designs of the thermal and damage neutron and gamma-ray scalar fluxes in these beams tubes. For experimental locations far from the reactor cores, angular flux data are required; however, for close-in experimental locations, the scalar fluxes within each beam tube provide a credible estimate of the various signal to noise ratios. In this paper, the coupled two- and three-dimensional radiation transport calculations employed to estimate the scalar neutron and gamma-ray fluxes will be described and the results from these calculations will be discussed. 6 refs., 2 figs
International Nuclear Information System (INIS)
Hayter, J.B.
1994-01-01
The Advanced Neutron Source (ANS), slated for construction start in 1994, will be a multipurpose neutron research laboratory serving academic and industrial users in chemistry, biology, condensed matter physics, nuclear and fundamental physics, materials science and engineering, and many other fields. It will be centered on the world's highest flux neutron beam reactor, operating at 330 MW, with careful design integration between the neutron source and the experiment systems. Many instruments will be situated in low backgrounds at distances up to 80 m from the reactor, using neutron guides with tailored neutron optical coatings for beam transport. Apart from the many stations for neutron scattering research, specialized stations will also be provided for isotope separation on-line, experiments with liquid hydrogen targets, neutron optical techniques such as interferometry, activation analysis, depth profiling, and positron production. Careful consideration has been given to providing a good research environment for visiting scientists, including easy access to the experimental areas, while maintaining a highly secure nuclear facility. This paper will describe the reactor and experimental facilities and give some examples of the types of research for which ANS has been designed
Advances in the solution of three-dimensional nodal neutron transport equation
International Nuclear Information System (INIS)
Pazos, Ruben Panta; Hauser, Eliete Biasotto; Vilhena, Marco Tullio de
2003-01-01
In this paper we study the three-dimensional nodal discrete-ordinates approximations of neutron transport equation in a convex domain with piecewise smooth boundaries. 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 obtaining 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. We give numerical results obtained with an algebraic computer system (for N up to 8) and with a code for higher values of N. We compare our results for the geometry of a box with a source in a vertex and a leakage zone in the opposite with others techniques used in this problem. (author)
International Nuclear Information System (INIS)
Hayter, J.B.
1989-01-01
The Advanced Neutron Source (ANS) is a new user experimental facility planned to be operational at Oak Ridge in the late 1990's. The centerpiece of the ANS will be a steady-state research reactor of unprecedented thermal neutron flux (φ th ∼ 9·10 19 m -2 ·s -1 ) accompanied by extensive and comprehensive equipment and facilities for neutron-based research. 5 refs., 5 figs
International Nuclear Information System (INIS)
Raman, S.; Hayter, J.B.
1990-01-01
The Advanced Neutron Source (ANS) is a new user experimental facility planned to be operational at Oak Ridge in the late 1990's. The centerpiece of the ANS will be a steady-state research reactor of unprecedented thermal neutron flux (φ th ∼ 8 x 10 19 m -2 ·s -1 ) accompanied by extensive and comprehensive equipment and facilities for neutron-based research
Neutron transportation simulator
International Nuclear Information System (INIS)
Uenohara, Yuzo.
1995-01-01
In the present invention, problems in an existent parallelized monte carlo method is solved, and behaviors of neutrons in a large scaled system are accurately simulated at a high speed. Namely, a neutron transportation simulator according to the monte carlo method simulates movement of each of neutrons by using a parallel computer. In this case, the system to be processed is divided based on a space region and an energy region to which neutrons belong. Simulation of neutrons in the divided regions is allotted to each of performing devices of the parallel computer. Tarry data and nuclear data of the neutrons in each of the regions are memorized dispersedly to memories of each of the performing devices. A transmission means for simulating the behaviors of the neutrons in the region by each of the performing devices, as well as transmitting the information of the neutrons, when the neutrons are moved to other region, to the performing device in a transported portion are disposed to each of the performing devices. With such procedures, simulation for the neutrons in the allotted region can be conducted with small capacity of memories. (I.S.)
International Nuclear Information System (INIS)
Peretz, F.J.
1990-01-01
The Advanced Neutron Source (ANS) is to be a multipurpose neutron research center, constructed around a high-flux reactor now being designed at the Oak Ridge National Laboratory (ORNL). Its primary purpose is to place the United States in the forefront of neutron scattering in the twenty-first century. Other research programs include nuclear and fundamental physics, isotope production, materials irradiation, and analytical chemistry. The ANS will be a unique and invaluable research tool because of the unprecedented neutron flux available from the high-intensity research reactor. But this reactor would be ineffective without world-class research facilities that allow the fullest utilization of the available neutrons. And, in turn, those research facilities will not produce new and exciting science without a broad population of users from all parts of the nation and the world, placed in a stimulating environment in which experiments can be effectively conducted and in which scientific exchange is encouraged. This paper discusses the measures being taken to ensure that the design of the ANS focuses not only on the reactor, but on providing the experiment and user support facilities needed to allow its effective use
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)
Recent advances in neutron tomography
International Nuclear Information System (INIS)
McFarland, E.; Massachusetts Inst. of Technology, Cambridge, MA; Lanza, R.
1993-01-01
Neutron imaging has been shown to be an excellent imaging tool for many nondestructive evaluation applications. Significantly improved contrast over X-ray images is possible for materials commonly found in engineering assemblies. The major limitations have been the neutron source and detection. A low cost, position sensitive neutron tomography detector system has been designed and built based on an electro-optical detector system using a LiF-ZnS scintillator screen and a cooled charge coupled device. This detector system can be used for neutron radiography as well as two and three-dimensional neutron tomography. Calculated performance of the system predicted near-quantum efficiency for position sensitive neutron detection. Experimental data was recently taken using this system at McClellan Air Force Base, Air Logistics Center, Sacramento, CA. With increased availability of low cost neutron sources and advanced image processing, neutron tomography will become an increasingly important nondestructive imaging method
Advanced Neutron Source (ANS) Project progress report
International Nuclear Information System (INIS)
McBee, M.R.; Chance, C.M.
1990-04-01
This report discusses the following topics on the advanced neutron source: quality assurance (QA) program; reactor core development; fuel element specification; corrosion loop tests and analyses; thermal-hydraulic loop tests; reactor control concepts; critical and subcritical experiments; material data, structural tests, and analysis; cold source development; beam tube, guide, and instrument development; hot source development; neutron transport and shielding; I ampersand C research and development; facility concepts; design; and safety
Advances in neutron scattering spectroscopy
International Nuclear Information System (INIS)
White, J.W.
1977-01-01
Some aspects of the application of neutron scattering to problems in polymer science, surface chemistry, and adsorption phenomena, as well as molecular biology, are reviewed. In all these areas, very significant work has been carried out using the medium flux reactors at Harwell, Juelich and Risoe, even without the use of advanced multidetector techniques or of a neutron cold source. A general tendency can also be distinguished in that, for each of these new fields, a distinct preference for colder neutrons rather than thermal neutron beams can be seen. (author)
Advanced Transportation Institute 2008.
2009-02-01
The seventh version of the Advanced Transportation Institute (ATI-08) was conducted in 2008 to encourage high school students to pursue careers in the field of transportation engineering. The University Transportation Center for Alabama partnered wit...
Advanced Transportation Institute 2009.
2009-09-01
The eighth version of the Advanced Transportation Institute (ATI-09) was conducted in 2009 to encourage high school students to pursue careers in the field of transportation engineering. The University Transportation Center for Alabama partnered with...
Advanced neutron source project
International Nuclear Information System (INIS)
Gorynina, L.V.; Proskuryakov, S.F.; Tishchenko, V.A.; Uzhanova, V.V.
1991-01-01
The project of the ANS improved neutron source intended for fundamental researches in nuclear physics and materials testing is considered. New superhigh-flux heavy-water 350 MW reactor is used for the source creation. The standard fuel is uranium silicide (U 3 Si 2 ). Reactor core volume equals 67.4 l and average power density is 4.9 MW/l. Neutron flux density is 10 16 neutron/(cm 2 xs). The facility construction begin is planned for 1996. The first experiments should be accomplished in 2000
An introduction to neutron transport
International Nuclear Information System (INIS)
Wiesenfeld, Bernard
2015-01-01
Neutron transport science is the study of neutron transport in a nuclear reactor and of associated nuclear reactions, notably fission reactions. Heat released by these reactions can be used for several purposes: electricity production, hydrogen production, sea water desalination, urban heating, naval propulsion, space propulsion, and so on. This publication contains the course proposed at Mines ParisTech and at the Arts et Metiers ParisTech. It is an introduction to neutron transport science and aims at presenting fundamental physical principles of this original branch of nuclear physics, a so called 'low energies' branch whereas 'high energy' nuclear physics focuses on elementary particles. It addresses complex computation methods which have been developed during the last decades with computation codes of always higher performance. The first part presents elements of atom physics: origin of matter, properties of nuclei and atoms, notion of quantum mechanics, interaction between radiation and matter (ray absorption, Compton Effect and scattering, photoelectric effect). The second part introduces neutron transport by addressing the following issues: nuclear structure, the various aspects of the interaction between neutrons and matter, the evolution of the reactivity of a reactor in normal operation, the chain fission reaction kinetics, and neutron slowing down. The third part addresses various aspects of neutron transport calculation: expression of neutron assessment, scattering approximation, critical condition of a nuclear reactor, introduction to transport theory, peculiarities of fast breeder reactors. The last chapter 'from theory to practice' addresses the approach of the neutron scientist, proposes an overview of the main calculation codes, and presents fields of application (within or without nuclear fission)
3-D neutron transport benchmarks
International Nuclear Information System (INIS)
Takeda, T.; Ikeda, H.
1991-03-01
A set of 3-D neutron transport benchmark problems proposed by the Osaka University to NEACRP in 1988 has been calculated by many participants and the corresponding results are summarized in this report. The results of K eff , control rod worth and region-averaged fluxes for the four proposed core models, calculated by using various 3-D transport codes are compared and discussed. The calculational methods used were: Monte Carlo, Discrete Ordinates (Sn), Spherical Harmonics (Pn), Nodal Transport and others. The solutions of the four core models are quite useful as benchmarks for checking the validity of 3-D neutron transport codes
Linear stochastic neutron transport theory
International Nuclear Information System (INIS)
Lewins, J.
1978-01-01
A new and direct derivation of the Bell-Pal fundamental equation for (low power) neutron stochastic behaviour in the Boltzmann continuum model is given. The development includes correlation of particle emission direction in induced and spontaneous fission. This leads to generalizations of the backward and forward equations for the mean and variance of neutron behaviour. The stochastic importance for neutron transport theory is introduced and related to the conventional deterministic importance. Defining equations and moment equations are derived and shown to be related to the backward fundamental equation with the detector distribution of the operational definition of stochastic importance playing the role of an adjoint source. (author)
Advanced Neutron Source enrichment study
International Nuclear Information System (INIS)
Bari, R.A.; Ludewig, H.; Weeks, J.R.
1996-01-01
A study has been performed of the impact on performance of using low-enriched uranium (20% 235 U) or medium-enriched uranium (35% 235 U) as an alternative fuel for the Advanced Neutron Source, which was initially designed to use uranium enriched to 93% 235 U. Higher fuel densities and larger volume cores were evaluated at the lower enrichments in terms of impact on neutron flux, safety, safeguards, technical feasibility, and cost. The feasibility of fabricating uranium silicide fuel at increasing material density was specifically addressed by a panel of international experts on research reactor fuels. The most viable alternative designs for the reactor at lower enrichments were identified and discussed. Several sensitivity analyses were performed to gain an understanding of the performance of the reactor at parametric values of power, fuel density, core volume, and enrichment that were interpolations between the boundary values imposed on the study or extrapolations from known technology
Neutron measurement by transportable spectrometer
International Nuclear Information System (INIS)
Anon.
1990-01-01
Two levels of neutron spectrometry are in regular use at nuclear power plants: some techniques used in the laboratory produce detailed spectra but require specialist operators, while simple instruments used by non-specialists to measure the neutron dose-rate to operators provide little spectral information. The standard portable instruments are therefore of no use when anomalous readings are obtained which require further investigation. AEA Technology at Winfrith has developed a Transportable Neutron Spectrometer (TNS) which is designed to produce reasonable spectra in routine use by staff with no specialist skill in spectroscopy, and high-quality spectra in the hands of skilled staff. The TNS provides a level of information intermediate between those currently available, and is also designed to solve the problem of imperfect dose response which is common in portable dosimeters. The TNS system consists of a power supply, a probe and a signal processing and data acquisition unit. (author)
Advanced Neutron Source (ANS) Project
International Nuclear Information System (INIS)
Campbell, J.H.; Thompson, P.B.
1994-01-01
This report covers the progress made in 1993 in the following sections: (1) project management; (2) research and development; (3) design and (4) safety. The section on research and development covers the following: (1) reactor core development; (2) fuel development; (3) corrosion loop tests and analysis; (4) thermal-hydraulic loop tests; (5) reactor control and shutdown concepts; (6) critical and subcritical experiments; (7) material data, structure tests, and analysis; (8) cold source development; (9) beam tube, guide, and instrument development; (10) neutron transport and shielding; (11) I and C research and development; and (12) facility concepts
Advanced Neutron Source operating philosophy
International Nuclear Information System (INIS)
Houser, M.M.
1993-01-01
An operating philosophy and operations cost estimate were prepared to support the Conceptual Design Report for the Advanced Neutron Source (ANS), a new research reactor planned for the Oak Ridge National Laboratory (ORNL). The operating philosophy was part of the initial effort of the ANS Human Factors Program, was integrated into the conceptual design, and addressed operational issues such as remote vs local operation; control room layout and responsibility issues; role of the operator; simulation and training; staffing levels; and plant computer systems. This paper will report on the overall plans and purpose for the operations work, the results of the work done for conceptual design, and plans for future effort
Monte Carlo simulation of neutron transport phenomena
International Nuclear Information System (INIS)
Srinivasan, P.
2009-01-01
Neutron transport is one of the central problems in nuclear reactor related studies and other applied sciences. Some of the major applications of neutron transport include nuclear reactor design and safety, criticality safety of fissile material handling, neutron detector design and development, nuclear medicine, assessment of radiation damage to materials, nuclear well logging, forensic analysis etc. Most reactor and dosimetry studies assume that neutrons diffuse from regions of high to low density just like gaseous molecules diffuse to regions of low concentration or heat flow from high to low temperature regions. However while treatment of gaseous or heat diffusion is quite accurately modeled, treatment of neutron transport as simple diffusion is quite limited. In simple diffusion, the neutron trajectories are irregular, random and zigzag - where as in neutron transport low reaction cross sections (1 barn- 10 -24 cm 2 ) lead to long mean free paths which again depend on the nature and irregularities of the medium. Hence a more accurate representation of the neutron transport evolved based on the Boltzmann equation of kinetic gas theory. In fact the neutron transport equation is a linearized version of the Boltzmann gas equation based on neutron conservation with seven independent variables. The transport equation is difficult to solve except in simple cases amenable to numerical methods. The diffusion (equation) approximation follows from removing the angular dependence of the neutron flux
Killing symmetries in neutron transport
International Nuclear Information System (INIS)
Lukacs, B.; Racz, A.
1992-10-01
Although inside the reactor zone there is no exact continuous spatial symmetry, in certain configurations neutron flux distribution is close to a symmetrical one. In such cases the symmetrical solution could provide a good starting point to determine the non-symmetrical power distribution. All possible symmetries are determined in the 3-dimensional Euclidean space, and the form of the transport equation is discussed in such a coordinate system which is adapted to the particular symmetry. Possible spontaneous symmetry breakings are pointed out. (author) 6 refs
Advanced Neutron Source (ANS) Project Progress report, FY 1991
International Nuclear Information System (INIS)
Campbell, J.H.; Selby, D.L.; Harrington, R.M.; Thompson, P.B.
1992-01-01
This report discusses the following about the Advanced Neutron Source: Project Management; Research and Development; Fuel Development; Corrosion Loop Tests and Analyses; Thermal-Hydraulic Loop Tests; Reactor Control and Shutdown Concepts; Critical and Subcritical Experiments; Material Data, Structural Tests, and Analysis; Cold-Source Development; Beam Tube, Guide, and Instrument Development; Hot-Source Development; Neutron Transport and Shielding; I ampersand C Research and Development; Design; and Safety
Some improved methods in neutron transport theory
Energy Technology Data Exchange (ETDEWEB)
Pop-Jordanov, J; Stefanovic, D; Kocic, A; Matausek, M; Bosevski, T [Boris Kidric Institute of Nuclear Sciences Vinca, Beograd (Yugoslavia)
1973-07-01
The methods described in this paper are: analytical approach to neutron spectra in case of energy dependent anisotropy of elastic scattering; Monte Carlo estimations of neutron absorption reaction rate during slowing down process; spherical harmonics treatment of space-angle-lethargy dependent slowing down transport equation; integral transport theory based on point-wise representation of variables.
(International Collaboration on Advanced Neutron Sources)
Energy Technology Data Exchange (ETDEWEB)
Hayter, J.B.
1990-11-08
The International Collaboration on Advanced Neutron Sources was started about a decade ago with the purpose of sharing information throughout the global neutron community. The collaboration has been extremely successful in optimizing the use of resources, and the discussions are open and detailed, with reasons for failure shared as well as reasons for success. Although the meetings have become increasingly oriented toward pulsed neutron sources, many of the neutron instrumentation techniques, such as the development of better monochromators, fast response detectors and various data analysis methods, are highly relevant to the Advanced Neutron Source (ANS). I presented one paper on the ANS, and another on the neutron optical polarizer design work which won a 1989 R D-100 Award. I also gained some valuable design ideas, in particular for the ANS hot source, in discussions with individual researchers from Canada, Western Europe, and Japan.
ADVANCED CUTTINGS TRANSPORT STUDY
Energy Technology Data Exchange (ETDEWEB)
Troy Reed; Stefan Miska; Nicholas Takach; Kaveh Ashenayi; Gerald Kane; Mark Pickell; Len Volk; Mike Volk; Barkim Demirdal; Affonso Lourenco; Evren Ozbayoglu; Paco Vieira; Lei Zhou
2000-01-30
This is the second quarterly progress report for Year 2 of the ACTS project. It includes a review of progress made in Flow Loop development and research during the period of time between Oct 1, 2000 and December 31, 2000. This report presents a review of progress on the following specific tasks: (a) Design and development of an Advanced Cuttings Transport Facility (Task 2: Addition of a foam generation and breaker system), (b) Research project (Task 6): ''Study of Cuttings Transport with Foam Under LPAT Conditions (Joint Project with TUDRP)'', (c) Research project (Task 7): ''Study of Cuttings Transport with Aerated Muds Under LPAT Conditions (Joint Project with TUDRP)'', (d) Research project (Task 8): ''Study of Flow of Synthetic Drilling Fluids Under Elevated Pressure and Temperature Conditions'', (e) Research project (Task 9): ''Study of Foam Flow Behavior Under EPET Conditions'', (f) Research project (Task 10): ''Study of Cuttings Transport with Aerated Mud Under Elevated Pressure and Temperature Conditions'', (g) Research on instrumentation tasks to measure: Cuttings concentration and distribution in a flowing slurry (Task 11), and Foam properties while transporting cuttings. (Task 12), (h) Development of a Safety program for the ACTS Flow Loop. Progress on a comprehensive safety review of all flow-loop components and operational procedures. (Task 1S). (i) Activities towards technology transfer and developing contacts with Petroleum and service company members, and increasing the number of JIP members. The tasks Completed During This Quarter are Task 7 and Task 8.
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
ADVANCED CUTTINGS TRANSPORT STUDY
Energy Technology Data Exchange (ETDEWEB)
Ergun Kuru; Stefan Miska; Nicholas Takach; Kaveh Ashenayi; Gerald Kane; Len Volk; Mark Pickell; Evren Ozbayoglu; Barkim Demirdal; Paco Vieira; Affonso Lourenco
1999-10-15
This report includes a review of the progress made in ACTF Flow Loop development and research during 90 days pre-award period (May 15-July 14, 1999) and the following three months after the project approval date (July15-October 15, 1999) The report presents information on the following specific subjects; (a) Progress in Advanced Cuttings Transport Facility design and development, (b) Progress report on the research project ''Study of Flow of Synthetic Drilling Fluids Under Elevated Pressure and Temperature Conditions'', (c) Progress report on the research project ''Study of Cuttings Transport with Foam Under LPAT Conditions (Joint Project with TUDRP)'', (d) Progress report on the research project ''Study of Cuttings Transport with Aerated Muds Under LPAT Conditions (Joint Project with TUDRP)'', (e) Progress report on the research project ''Study of Foam Flow Behavior Under EPET Conditions'', (f) Progress report on the instrumentation tasks (Tasks 11 and 12) (g) Activities towards technology transfer and developing contacts with oil and service company members.
An advanced fusion neutron source facility
International Nuclear Information System (INIS)
Smith, D.L.
1992-01-01
Accelerator-based 14-MeV-neutron sources based on modifications of the original Fusion Materials Irradiation Facility are currently under consideration for investigating the effects of high-fluence high-energy neutron irradiation on fusion-reactor materials. One such concept for a D-Li neutron source is based on recent advances in accelerator technology associated with the Continuous Wave Deuterium Demonstrator accelerator under construction at Argonne National Laboratory, associated superconducting technology, and advances in liquid-metal technology. In this paper a summary of conceptual design aspects based on improvements in technologies is presented
Options for the Delft advanced neutron source
International Nuclear Information System (INIS)
Gibcus, H.P.M.; Leege, P.F.A. de; Labohm, F.; Vries, J.W. de; Verkooijen, A.H.M.; Valko, J.; Feltes, W.; Heinecke, J.
2003-01-01
Results of feasibility studies are presented for options for an advanced neutron source for the Delft reactor including upgrading the HOR, a 2 MW pool-type research reactor at the Delft University of Technology. The primary utilisation of the HOR focuses on beam research applications with neutrons and positrons. The aim of being scientifically competitive in that research area requires a thermal neutron flux level of at least 1x10 14 n/cm 2 /s. The feasibility of an accelerator driven neutron source and upgrading the present core to a super compact core for reaching this goal has been investigated at large from a safety and operational point of view. For the upgraded core, a 3x3 fuel assembly arrangement and beryllium reflected at all sides was chosen. Figures on the system performance, including the merits of a cold neutron source application feeding the neutron guide system, are presented. (author)
Considerations in the design of an improved transportable neutron spectrometer
Williams, A M; Brushwood, J M; Beeley, P A
2002-01-01
The Transportable Neutron Spectrometer (TNS) has been used by the Ministry of Defence for over 15 years to characterise neutron fields in workplace environments and provide local correction factors for both area and personal dosimeters. In light of advances in neutron spectrometry, a programme to evaluate and improve TNS has been initiated. This paper describes TNS, presents its operation in known radioisotope fields and in a reactor environment. Deficiencies in the operation of the instrument are highlighted, together with proposals for updating the response functions and spectrum unfolding methodologies.
ADVANCED CUTTINGS TRANSPORT STUDY
Energy Technology Data Exchange (ETDEWEB)
Stefan Miska; Troy Reed; Ergun Kuru
2004-09-30
The Advanced Cuttings Transport Study (ACTS) was a 5-year JIP project undertaken at the University of Tulsa (TU). The project was sponsored by the U.S. Department of Energy (DOE) and JIP member companies. The objectives of the project were: (1) to develop and construct a new research facility that would allow three-phase (gas, liquid and cuttings) flow experiments under ambient and EPET (elevated pressure and temperature) conditions, and at different angle of inclinations and drill pipe rotation speeds; (2) to conduct experiments and develop a data base for the industry and academia; and (3) to develop mechanistic models for optimization of drilling hydraulics and cuttings transport. This project consisted of research studies, flow loop construction and instrumentation development. Following a one-year period for basic flow loop construction, a proposal was submitted by TU to the DOE for a five-year project that was organized in such a manner as to provide a logical progression of research experiments as well as additions to the basic flow loop. The flow loop additions and improvements included: (1) elevated temperature capability; (2) two-phase (gas and liquid, foam etc.) capability; (3) cuttings injection and removal system; (4) drill pipe rotation system; and (5) drilling section elevation system. In parallel with the flow loop construction, hydraulics and cuttings transport studies were preformed using drilling foams and aerated muds. In addition, hydraulics and rheology of synthetic drilling fluids were investigated. The studies were performed under ambient and EPET conditions. The effects of temperature and pressure on the hydraulics and cuttings transport were investigated. Mechanistic models were developed to predict frictional pressure loss and cuttings transport in horizontal and near-horizontal configurations. Model predictions were compared with the measured data. Predominantly, model predictions show satisfactory agreements with the measured data. As a
Exact solution of the neutron transport equation in spherical geometry
Energy Technology Data Exchange (ETDEWEB)
Anli, Fikret; Akkurt, Abdullah; Yildirim, Hueseyin; Ates, Kemal [Kahramanmaras Suetcue Imam Univ. (Turkey). Faculty of Sciences and Letters
2017-03-15
Solution of the neutron transport equation in one dimensional slab geometry construct a basis for the solution of neutron transport equation in a curvilinear geometry. Therefore, in this work, we attempt to derive an exact analytical benchmark solution for both neutron transport equations in slab and spherical medium by using P{sub N} approximation which is widely used in neutron transport theory.
Advances in neutron based bulk explosive detection
Gozani, Tsahi; Strellis, Dan
2007-08-01
Neutron based explosive inspection systems can detect a wide variety of national security threats. The inspection is founded on the detection of characteristic gamma rays emitted as the result of neutron interactions with materials. Generally these are gamma rays resulting from thermal neutron capture and inelastic scattering reactions in most materials and fast and thermal neutron fission in fissile (e.g.235U and 239Pu) and fertile (e.g.238U) materials. Cars or trucks laden with explosives, drugs, chemical agents and hazardous materials can be detected. Cargo material classification via its main elements and nuclear materials detection can also be accomplished with such neutron based platforms, when appropriate neutron sources, gamma ray spectroscopy, neutron detectors and suitable decision algorithms are employed. Neutron based techniques can be used in a variety of scenarios and operational modes. They can be used as stand alones for complete scan of objects such as vehicles, or for spot-checks to clear (or validate) alarms indicated by another inspection system such as X-ray radiography. The technologies developed over the last two decades are now being implemented with good results. Further advances have been made over the last few years that increase the sensitivity, applicability and robustness of these systems. The advances range from the synchronous inspection of two sides of vehicles, increasing throughput and sensitivity and reducing imparted dose to the inspected object and its occupants (if any), to taking advantage of the neutron kinetic behavior of cargo to remove systematic errors, reducing background effects and improving fast neutron signals.
Development of advanced neutron beam technology
Energy Technology Data Exchange (ETDEWEB)
Seong, B S; Lee, J S; Sim, C M [and others
2007-06-15
The purpose of this work is to timely support the national science and technology policy through development of the advanced application techniques for neutron spectrometers, built in the previous project, in order to improve the neutron spectrometer techniques up to the world-class level in both quantity and quality and to reinforce industrial competitiveness. The importance of the research and development (R and D) is as follows: 1. Technological aspects - Development of a high value-added technology through performing the advanced R and D in the broad research areas from basic to applied science and from hard to soft condensed matter using neutron scattering technique. - Achievement of an important role in development of the new technology for the following industries aerospace, defense industry, atomic energy, hydrogen fuel cell etc. by the non-destructive inspection and analysis using neutron radiography. - Development of a system supporting the academic-industry users for the HANARO facility 2. Economical and Industrial Aspects - Essential technology in the industrial application of neutron spectrometer, in the basic and applied research of the diverse materials sciences, and in NT, BT, and IT areas - Broad impact on the economics and the domestic and international collaborative research by using the neutron instruments in the mega-scale research facility, HANARO, that is a unique source of neutron in Korea. 3. Social Aspects - Creating the scientific knowledge and contributing to the advanced industrial society through the neutron beam application - Improving quality of life and building a national consensus on the application of nuclear power by developing the RT fusion technology using the HANARO facility. - Widening the national research area and strengthening the national R and D capability by performing advanced R and D using the HANARO facility.
Advanced Neutron Source: The users' perspective
International Nuclear Information System (INIS)
Peretz, F.J.
1990-01-01
User experiments will cover fields such as activation analysis of pollutants, irradiation of materials for the fusion program, and neutron scattering studies of materials as diverse as viruses, aerospace composites, and superconductors. Production capabilities must also be provided for the production of isotopes, especially of transuranic elements. The different ways in which these research areas and their required infrastructure influence the design of the Advanced Neutron Source will be the subject of this paper
Development of advanced neutron beam technology
International Nuclear Information System (INIS)
Seong, B. S.; Lee, J. S.; Sim, C. M.
2007-06-01
The purpose of this work is to timely support the national science and technology policy through development of the advanced application techniques for neutron spectrometers, built in the previous project, in order to improve the neutron spectrometer techniques up to the world-class level in both quantity and quality and to reinforce industrial competitiveness. The importance of the research and development (R and D) is as follows: 1. Technological aspects - Development of a high value-added technology through performing the advanced R and D in the broad research areas from basic to applied science and from hard to soft condensed matter using neutron scattering technique. - Achievement of an important role in development of the new technology for the following industries aerospace, defense industry, atomic energy, hydrogen fuel cell etc. by the non-destructive inspection and analysis using neutron radiography. - Development of a system supporting the academic-industry users for the HANARO facility 2. Economical and Industrial Aspects - Essential technology in the industrial application of neutron spectrometer, in the basic and applied research of the diverse materials sciences, and in NT, BT, and IT areas - Broad impact on the economics and the domestic and international collaborative research by using the neutron instruments in the mega-scale research facility, HANARO, that is a unique source of neutron in Korea. 3. Social Aspects - Creating the scientific knowledge and contributing to the advanced industrial society through the neutron beam application - Improving quality of life and building a national consensus on the application of nuclear power by developing the RT fusion technology using the HANARO facility. - Widening the national research area and strengthening the national R and D capability by performing advanced R and D using the HANARO facility
Irradiation facilities at the advanced neutron source
International Nuclear Information System (INIS)
West, C.D.
1992-01-01
The Advanced Neutron Source (ANS) is a facility, centered around a new 330MW(f) heavy-water cooled and reflected research reactor, proposed for construction at Oak Ridge. The main scientific justification for the new source is the United States' need for increased capabilities in neutron scattering and other neutron beam research, but the technical objectives of the project also cater for the need to replace the irradiation facilities at the aging High Flux Isotope Reactor and to provide other research capabilities to the scientific community. This document provides a description of the ANS facilities
Status of the advanced neutron source
International Nuclear Information System (INIS)
Hayter, J.B.
1991-01-01
Research reactors in the United States are becoming more and more outdated, at a time when neutron scattering is being recognized as an increasingly important technique in areas vital to the U.S. scientific and technological future. The last U.S. research reactor was constructed over 25 years ago, whereas new facilities have been built or are under construction in Japan, Russia and, especially, Western Europe, which now has a commanding lead in this important field. Concern over this situation in the early 1980's by a number of organizations, including the National Academy of Sciences, led to a recommendation that design work start urgently on an advanced U.S. neutron research facility. This recommendation is realized in the Advanced Neutron Source Project. The centerpiece of the Advanced Neutron Source will be a new research reactor of unprecedented flux (> 7.5x10 19 m -2 ·s -1 ), equipped with a wide variety of state-of-the-art spectrometers and diffractometers on hot, thermal, and cold neutron beams. Very cold and ultracold neutron beams will also be provided for specialized experiments. This paper will discuss the current status of the design and the plans for scattering instrumentation. (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
A Green function of neutron transport equation
International Nuclear Information System (INIS)
Simovic, R.
1993-01-01
In this paper the angularly dependent Green function of the neutron transport equation is derived analytically and approximately. By applying the analytical FDPN approximation up to eighth order, numerical values of the Green functions are obtained with the accuracy of six significant figures in the whole range of parameter c, angle cosine μ and distances x up to the ten optical lengths from the neutron source. (author)
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.)
Advanced public transportation systems benefits
1996-03-01
Benefits and cost savings for various Advanced Public Transportation Systems are outlined here. Operational efficiencies are given for Transit Management Systems in different locales, as well as compliant resolution and safety. Electronic Fare Paymen...
Study of a transportable neutron radiography system
International Nuclear Information System (INIS)
Souza, S.N.A. de.
1991-05-01
This work presents a study a transportable neutron radiography system for a 185 GBq 241 Am-Be (α, η) source with a neutron yield roughly 1,25 x 10 7 n/s. Studies about moderation, collimation and shielding are showed. In these studies, a calculation using Transport Theory was carried out by means of transport codes ANISN and DOT (3.5). Objectives were: to obtain a maximum and more homogeneous thermal neutron flux in the collimator outlet 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 of 14, for neutron fluxes up to 4,09 x 10 2 n.cm -2 .s -1 . Considering the low intensity of the source, it is a good value. Studies have also been carried out for L/D ratios of 22 and 30, giving thermal neutron fluxes at the image plain of 1,27 x 10 2 n.cm -2 .s -1 and 2,65 x 10 2 n.cm -2 .s -1 , respectively. (author). 30 refs, 39 figs, 9 tabs
A New Monte Carlo Neutron Transport Code at UNIST
International Nuclear Information System (INIS)
Lee, Hyunsuk; Kong, Chidong; Lee, Deokjung
2014-01-01
Monte Carlo neutron transport code named MCS is under development at UNIST for the advanced reactor design and research purpose. This MC code can be used for fixed source calculation and criticality calculation. Continuous energy neutron cross section data and multi-group cross section data can be used for the MC calculation. This paper presents the overview of developed MC code and its calculation results. The real time fixed source calculation ability is also tested in this paper. The calculation results show good agreement with commercial code and experiment. A new Monte Carlo neutron transport code is being developed at UNIST. The MC codes are tested with several benchmark problems: ICSBEP, VENUS-2, and Hoogenboom-Martin benchmark. These benchmarks covers pin geometry to 3-dimensional whole core, and results shows good agreement with reference results
Transportable type neutron level indicators
International Nuclear Information System (INIS)
Khatskevich, M.V.; Kalinin, O.V.; Moskovkin, V.N.; Molchanov, A.V.; Bobkov, A.D.; Rabotnov, Yu.A.
1979-01-01
Some peculiarities of designing level neutron converters (LNC) for portable indicators or level neutron relays are considered. The effect of the LNC geometry and other factors on measurement errors has been studied. Calibration results of the LNC with a neutron reflector and without it are presented. It is shown that the problem of level monitoring with the help of portable indicators can be solved practically for any volume, provided two LNC modifications with reflectors are available: the NPU-G modification with horizontal location of a counter for large volumes and the NPU-V with vertical location of a counter for lesser volumes. A possibility of perfecting LNC performances by shielding the counter with thermal neutron absorbers has been studied. The design of the NPU-V modification for the NIUP-2 level indicator is described. It is intended for tubes and cylinders 30-100 mm in diameter. Measurements carried out on different steel and aluminium vessels with a diameter ranging from 300 to 100 mm and a wall thickness of up to 16 mm with the help of the NPU-V and NPU-G modifications proved the efficiency of the LNC to control a variety of products (kerosine, gasoline, oils, acids, alkalis) [ru
Lectures on neutron transport theory
International Nuclear Information System (INIS)
Benoist, P.
1986-02-01
This note is divided in two parts. In the first one the basis of transport theory, that is, the principal forms of the transport equation and the resulting theorems, are presented. The second part is particularly devoted to the applications of integral transport theory to reactor lattice problems [fr
Advanced Neutron Source: The designer's perspective
International Nuclear Information System (INIS)
Peretz, F.J.
1990-01-01
The Advanced Neutron Source (ANS) is a research facility based on a 350 MW beam reactor, to be brought into service at the Oak Ridge National Laboratory at the end of the century. The primary objective is to provide high-flux neutron beams and guides, with cold, thermal, hot, and ultra-cold neutrons, for research in many fields of science. Secondary objectives include isotopes production, materials irradiation and activation analysis. The design of the ANS is strongly influenced by the historical development of research and power reactor concepts, and of the regulatory infrastructure of the Department of Energy (DOE) and the Nuclear Regulatory Commission (NRC). Current trends in reactor safety also impact the climate for the design of such a reactor
ALADIN - Advanced Laue Diffraction Instruments using Neutrons
International Nuclear Information System (INIS)
Lemee-Cailleau, M.H.; Ouladdiaf, B.; McIntyre, G.J.
2011-01-01
Laue diffraction techniques have proven to be very attractive to a broad user community interested in obtaining detailed structural information on very small single-crystal samples or needing data collection speeds comparable to those available with the powder diffraction technique. However our experience has clearly demonstrated the negative effect of up-stream monochromatic instruments on the quality of Laue data. In order to obtain Laue diffraction data with a statistical accuracy similar to that achieved on a monochromatic instrument (neutron or X-rays), the project ALADIN (for Advanced Laue Diffraction Instruments using Neutrons) aims to: -) construct a Laue-dedicated thermal neutron guide, with m=2 super-mirror coating, providing access to the desirable wavelength bandwidth; -) installation of one of the ILL Laue diffractometers (VIVALDI or CYCLOPS) on this new guide. (authors)
Uncertainty analysis of neutron transport calculation
International Nuclear Information System (INIS)
Oka, Y.; Furuta, K.; Kondo, S.
1987-01-01
A cross section sensitivity-uncertainty analysis code, SUSD was developed. The code calculates sensitivity coefficients for one and two-dimensional transport problems based on the first order perturbation theory. Variance and standard deviation of detector responses or design parameters can be obtained using cross section covariance matrix. The code is able to perform sensitivity-uncertainty analysis for secondary neutron angular distribution(SAD) and secondary neutron energy distribution(SED). Covariances of 6 Li and 7 Li neutron cross sections in JENDL-3PR1 were evaluated including SAD and SED. Covariances of Fe and Be were also evaluated. The uncertainty of tritium breeding ratio, fast neutron leakage flux and neutron heating was analysed on four types of blanket concepts for a commercial tokamak fusion reactor. The uncertainty of tritium breeding ratio was less than 6 percent. Contribution from SAD/SED uncertainties are significant for some parameters. Formulas to estimate the errors of numerical solution of the transport equation were derived based on the perturbation theory. This method enables us to deterministically estimate the numerical errors due to iterative solution, spacial discretization and Legendre polynomial expansion of transfer cross-sections. The calculational errors of the tritium breeding ratio and the fast neutron leakage flux of the fusion blankets were analysed. (author)
Advanced transport aircraft technology
Energy Technology Data Exchange (ETDEWEB)
Winblade, R L
1980-06-01
Various elements of the NASA aircraft energy efficiency program are described. Regarding composite structures, the development of three secondary and three medium-primary components to validate structural and fabrication technology is discussed. In laminar flow control, the design of advanced airfoils having large regions of supercritical flow with features which simplify laminarization are considered. Emphasis is placed on engine performance improvement, directed at developing advanced components to reduce fuel consumption in current production engines, and engine diagnostics aimed at identifying the sources and causes of performance deterioration in high-bypass turbofan engines. In addition, the results of propeller aerodynamic and acoustic tests have substantiated the feasibility of achieving the propeller efficiency goal of 80% and confirmed that the effect of blade sweep on reducing propeller source noise was 5-6 dB.
Neutron radiography using a transportable superconducting cyclotron
Energy Technology Data Exchange (ETDEWEB)
Allen, D.A. (School of Physics and Space Research, University of Birmingham, Birmingham, B15 2TT (United Kingdom)); Hawkesworth, M.R. (School of Physics and Space Research, University of Birmingham, Birmingham, B15 2TT (United Kingdom)); Beynon, T.D. (School of Physics and Space Research, University of Birmingham, Birmingham, B15 2TT (United Kingdom)); Green, S. (School of Physics and Space Research, University of Birmingham, Birmingham, B15 2TT (United Kingdom)); Rogers, J.D. (Rolls-Royce, Derby (United Kingdom)); Allen, M.J. (Rolls-Royce, Derby (United Kingdom)); Plummer, H.C. (Rolls-Royce, MatEval, Derby (United Kingdom)); Boulding, N.J. (Oxford Instruments (United Kingdom)); Cox, M. (Oxford Instruments (United Kingdom)); McDougall, I. (Oxford Instruments (United Kingdom))
1994-12-30
A thermal neutron radiography system based on a compact 12 MeV superconducting proton cyclotron is described. Neutrons are generated using a thick beryllium target and moderated in high density polyethylene. Monte Carlo computer simulations have been used to model the neutron and photon transport in order to optimise the performance of the system. With proton beam currents in excess of 100 [mu]A, it can provide high thermal neutron fluxes with L/D ratios of between 50 and 300 for various applications. Both film and electronic imaging are used to produce radiographs. The electronic imaging system consists of a [sup 6]Li-loaded ZnS intensifier screen, and a low light CCD or SIT camera. High resolution images can be recorded and computer-controlled data processing, analysis and display are possible. ((orig.))
ADVANCED CUTTINGS TRANSPORT STUDY
Energy Technology Data Exchange (ETDEWEB)
Stefan Miska; Nicholas Takach; Kaveh Ashenayi; Mengjiao Yu; Ramadan Ahmed; Mark Pickell; Len Volk; Lei Zhou; Zhu Chen; Aimee Washington; Crystal Redden
2003-09-30
The Quarter began with installing the new drill pipe, hooking up the new hydraulic power unit, completing the pipe rotation system (Task 4 has been completed), and making the SWACO choke operational. Detailed design and procurement work is proceeding on a system to elevate the drill-string section. The prototype Foam Generator Cell has been completed by Temco and delivered. Work is currently underway to calibrate the system. Literature review and preliminary model development for cuttings transportation with polymer foam under EPET conditions are in progress. Preparations for preliminary cuttings transport experiments with polymer foam have been completed. Two nuclear densitometers were re-calibrated. Drill pipe rotation system was tested up to 250 RPM. Water flow tests were conducted while rotating the drill pipe up to 100 RPM. The accuracy of weight measurements for cuttings in the annulus was evaluated. Additional modifications of the cuttings collection system are being considered in order to obtain the desired accurate measurement of cuttings weight in the annular test section. Cutting transport experiments with aerated fluids are being conducted at EPET, and analyses of the collected data are in progress. The printed circuit board is functioning with acceptable noise level to measure cuttings concentration at static condition using ultrasonic method. We were able to conduct several tests using a standard low pass filter to eliminate high frequency noise. We tested to verify that we can distinguish between different depths of sand in a static bed of sand. We tested with water, air and a mix of the two mediums. Major modifications to the DTF have almost been completed. A stop-flow cell is being designed for the DTF, the ACTF and Foam Generator/Viscometer which will allow us to capture bubble images without the need for ultra fast shutter speeds or microsecond flash system.
ADVANCED CUTTINGS TRANSPORT STUDY
Energy Technology Data Exchange (ETDEWEB)
Ergun Kuru; Stefan Miska; Nicholas Takach; Kaveh Ashenayi; Gerald Kane; Mark Pickell; Len Volk; Mike Volk; Barkim Demirdal; Affonso Lourenco; Evren Ozbayoglu; Paco Vieira; Neelima Godugu
2000-07-30
ACTS flow loop is now operational under elevated pressure and temperature. Currently, experiments with synthetic based drilling fluids under pressure and temperature are being conducted. Based on the analysis of Fann 70 data, empirical correlations defining the shear stress as a function of temperature, pressure and the shear rate have been developed for Petrobras synthetic drilling fluids. PVT equipment has been modified for testing Synthetic oil base drilling fluids. PVT tests with Petrobras Synthetic base mud have been conducted and results are being analyzed Foam flow experiments have been conducted and the analysis of the data has been carried out to characterize the rheology of the foam. Comparison of pressure loss prediction from the available foam hydraulic models and the test results has been made. Cuttings transport experiments in horizontal annulus section have been conducted using air, water and cuttings. Currently, cuttings transport tests in inclined test section are being conducted. Foam PVT analysis tests have been conducted. Foam stability experiments have also been conducted. Effects of salt and oil concentration on the foam stability have been investigated. Design of ACTS flow loop modification for foam and aerated mud flow has been completed. A flow loop operation procedure for conducting foam flow experiments under EPET conditions has been prepared Design of the lab-scale flow loop for dynamic foam characterization and cuttings monitoring instrumentation tests has been completed. The construction of the test loop is underway. As part of the technology transport efforts, Advisory Board Meeting with ACTS-JIP industry members has been organized on May 13, 2000.
Advanced Neutron Source overview and status report
International Nuclear Information System (INIS)
West, C.D.
1992-01-01
The new Advanced Neutron Source is a research facility centered around a new research reactor of unprecedented flux. Unique core and cooling system designs provide many inherent or passive safety features. The combination of a relatively high power level and a small core places special requirements on the response time of the reactor control system, and especially on the scram function. Similar requirements have been faced before on research reactors, and successfully met. The ANS design have evolved from those other reactors
3D neutron transport modelization
International Nuclear Information System (INIS)
Warin, X.
1996-12-01
Some nodal methods to solve the transport equation in 3D are presented. Two nodal methods presented at an OCDE congress are described: a first one is a low degree one called RTN0; a second one is a high degree one called BDM1. The two methods can be made faster with a totally consistent DSA. Some results of parallelization show that: 98% of the time is spent in sweeps; transport sweeps are easily parallelized. (K.A.)
3D neutron transport modelization
Energy Technology Data Exchange (ETDEWEB)
Warin, X.
1996-12-01
Some nodal methods to solve the transport equation in 3D are presented. Two nodal methods presented at an OCDE congress are described: a first one is a low degree one called RTN0; a second one is a high degree one called BDM1. The two methods can be made faster with a totally consistent DSA. Some results of parallelization show that: 98% of the time is spent in sweeps; transport sweeps are easily parallelized. (K.A.). 10 refs.
ADVANCED CUTTINGS TRANSPORT STUDY
Energy Technology Data Exchange (ETDEWEB)
Stefan Miska; Nicholas Takach; Kaveh Ashenayi
2004-01-31
Final design of the mast was completed (Task 5). The mast is consisting of two welded plate girders, set next to each other, and spaced 14-inches apart. Fabrication of the boom will be completed in two parts solely for ease of transportation. The end pivot connection will be made through a single 2-inch diameter x 4 feet-8 inch long 316 SS bar. During installation, hard piping make-ups using Chiksan joints will connect the annular section and 4-inch return line to allow full movement of the mast from horizontal to vertical. Additionally, flexible hoses and piping will be installed to isolate both towers from piping loads and allow recycling operations respectively. Calibration of the prototype Foam Generator Cell has been completed and experiments are now being conducted. We were able to generate up to 95% quality foam. Work is currently underway to attach the Thermo-Haake RS300 viscometer and install a view port with a microscope to measure foam bubble size and bubble size distribution. Foam rheology tests (Task 13) were carried out to evaluate the rheological properties of the proposed foam formulation. After successful completion of the first foam test, two sets of rheological tests were conducted at different foam flow rates while keeping other parameters constant (100 psig, 70F, 80% quality). The results from these tests are generally in agreement with the previous foam tests done previously during Task 9. However, an unanticipated observation during these tests was that in both cases, the frictional pressure drop in 2 inch pipe was lower than that in the 3 inch and 4 inch pipes. We also conducted the first foam cuttings transport test during this quarter. Experiments on aerated fluids without cuttings have been completed in ACTF (Task 10). Gas and liquid were injected at different flow rates. Two different sets of experiments were carried out, where the only difference was the temperature. Another set of tests was performed, which covered a wide range of
ADVANCED CUTTINGS TRANSPORT STUDY
Energy Technology Data Exchange (ETDEWEB)
Stefan Miska; Nicholas Takach; Kaveh Ashenayi
2004-07-31
We have tested the loop elevation system. We raised the mast to approximately 25 to 30 degrees from horizontal. All went well. However, while lowering the mast, it moved laterally a couple of degrees. Upon visual inspection, severe spalling of the concrete on the face of the support pillar, and deformation of the steel support structure was observed. At this time, the facility is ready for testing in the horizontal position. A new air compressor has been received and set in place for the ACTS test loop. A new laboratory has been built near the ACTS test loop Roughened cups and rotors for the viscometer (RS300) were obtained. Rheologies of aqueous foams were measured using three different cup-rotor assemblies that have different surface roughness. The relationship between surface roughness and foam rheology was investigated. Re-calibration of nuclear densitometers has been finished. The re-calibration was also performed with 1% surfactant foam. A new cuttings injection system was installed at the bottom of the injection tower. It replaced the previous injection auger. A mechanistic model for cuttings transport with aerated mud has been developed. Cuttings transport mechanisms with aerated water at various conditions were experimentally investigated. A total of 39 tests were performed. Comparisons between the model predictions and experimental measurements show a satisfactory agreement. Results from the ultrasonic monitoring system indicated that we could distinguish between different sand levels. We also have devised ways to achieve consistency of performance by securing the sensors in the caps in exactly the same manner as long as the sensors are not removed from the caps. A preliminary test was conducted on the main flow loop at 100 gpm flow rate and 20 lb/min cuttings injection rate. The measured bed thickness using the ultrasonic method showed a satisfactory agreement with nuclear densitometer readings. Thirty different data points were collected after the test
Solving the equation of neutron transport
International Nuclear Information System (INIS)
Nasfi, Rim
2009-01-01
This work is devoted to the study of some numerical methods of resolution of the problem of transport of the neutrons. We started by introducing the equation integro-differential transport of the neutrons. Then we applied the finite element method traditional for stationary and nonstationary linear problems in 2D. A great part is reserved for the presentation of the mixed numerical diagram and mixed hybrid with two types of uniform grids: triangular and rectangular. Thereafter we treated some numerical examples by implementations in Matlab in order to test the convergence of each method. To finish, we had results of simulation by the Monte Carlo method on a problem of two-dimensional transport with an aim of comparing them with the results resulting from the finite element method mixed hybrids. Some remarks and prospects conclude this work.
Reactivity studies on the advanced neutron source
International Nuclear Information System (INIS)
Ryskamp, J.M.; Redmond, E.L. II; Fletcher, C.D.
1990-01-01
An Advanced Neutron Source (ANS) with a peak thermal neutron flux of about 8.5 x 10 19 m -2 s -1 is being designed for condensed matter physics, materials science, isotope production, and fundamental physics research. The ANS is a new reactor-based research facility being planned by Oak Ridge National Laboratory (ORNL) to meet the need for an intense steady-state source of neutrons. The design effort is currently in the conceptual phase. A reference reactor design has been selected in order to examine the safety, performance, and costs associated with this one design. The ANS Project has an established, documented safety philosophy, and safety-related design criteria are currently being established. The purpose of this paper is to present analyses of safety aspects of the reference reactor design that are related to core reactivity events. These analyses include control rod worth, shutdown rod worth, heavy water voiding, neutron beam tube flooding, light water ingress, and single fuel element criticality. Understanding these safety aspects will allow us to make design modifications that improve the reactor safety and achieve the safety related design criteria. 8 refs., 3 tabs
Advanced Neutron Sources: Plant Design Requirements
International Nuclear Information System (INIS)
1990-07-01
The Advanced Neutron Source (ANS) is a new, world class facility for research using hot, thermal, cold, and ultra-cold neutrons. At the heart of the facility is a 350-MW th , heavy water cooled and moderated reactor. The reactor is housed in a central reactor building, with supporting equipment located in an adjoining reactor support building. An array of cold neutron guides fans out into a large guide hall, housing about 30 neutron research stations. Office, laboratory, and shop facilities are included to provide a complete users facility. The ANS is scheduled to begin operation at the Oak Ridge National Laboratory at the end of the decade. This Plant Design Requirements document defines the plant-level requirements for the design, construction, and operation of the ANS. This document also defines and provides input to the individual System Design Description (SDD) documents. Together, this Plant Design Requirements document and the set of SDD documents will define and control the baseline configuration of the ANS
Advanced Neutron Source: Plant Design Requirements
International Nuclear Information System (INIS)
1990-07-01
The Advanced Neutron Source will be a new world-class facility for research using hot, thermal, cold, and ultra-cold neutrons. The heart of the facility will be a 330-MW (fission), heavy-water cooled and heavy-water moderated reactor. The reactor will be housed in a central reactor building, with supporting equipment located in an adjoining reactor support building. An array of cold neutron guides will fan out into a large guide hall, housing about 30 neutron research stations. Appropriate office, laboratory, and shop facilities will be included to provide a complete facility for users. The ANS is scheduled to begin operation at the Oak Ridge National Laboratory early in the next decade. This PDR document defines the plant-level requirements for the design, construction, and operation of ANS. It also defines and provides input to the individual System Design Description (SDD) documents. Together, this PDR document and the set of SDD documents will define and control the baseline configuration of ANS
Advances in transport phenomena 2011
2014-01-01
This new volume of the annual review “Advances in Transport Phenomena” series contains three in-depth review articles on the microfluidic fabrication of vesicles, the dielectrophoresis field-flow fractionation for continuous-flow separation of particles and cells in microfluidic devices, and the thermodynamic analysis and optimization of heat exchangers, respectively.
Advanced neutron source materials surveillance program
International Nuclear Information System (INIS)
Heavilin, S.M.
1995-01-01
The Advanced Neutron Source (ANS) will be composed of several different materials, one of which is 6061-T6 aluminum. Among other components, the reflector vessel and the core pressure boundary tube (CPBT), are to be made of 6061-T6 aluminum. These components will be subjected to high thermal neutron fluences and will require a surveillance program to monitor the strength and fracture toughness of the 6061-T6 aluminum over their lifetimes. The purpose of this paper is to explain the steps that were taken in the summer of 1994 toward developing the surveillance program. The first goal was to decide upon standard specimens to use in the fracture toughness and tensile testing. Second, facilities had to be chosen for specimens representing the CPBT and the reflector vessel base, weld, and heat-affected-zone (HAZ) metals. Third, a timetable had to be defined to determine when to remove the specimens for testing
Monte Carlo method for neutron transport problems
International Nuclear Information System (INIS)
Asaoka, Takumi
1977-01-01
Some methods for decreasing variances in Monte Carlo neutron transport calculations are presented together with the results of sample calculations. A general purpose neutron transport Monte Carlo code ''MORSE'' was used for the purpose. The first method discussed in this report is the method of statistical estimation. As an example of this method, the application of the coarse-mesh rebalance acceleration method to the criticality calculation of a cylindrical fast reactor is presented. Effective multiplication factor and its standard deviation are presented as a function of the number of histories and comparisons are made between the coarse-mesh rebalance method and the standard method. Five-group neutron fluxes at core center are also compared with the result of S4 calculation. The second method is the method of correlated sampling. This method was applied to the perturbation calculation of control rod worths in a fast critical assembly (FCA-V-3) Two methods of sampling (similar flight paths and identical flight paths) are tested and compared with experimental results. For every cases the experimental value lies within the standard deviation of the Monte Carlo calculations. The third method is the importance sampling. In this report a biased selection of particle flight directions discussed. This method was applied to the flux calculation in a spherical fast neutron system surrounded by a 10.16 cm iron reflector. Result-direction biasing, path-length stretching, and no biasing are compared with S8 calculation. (Aoki, K.)
Multi-group neutron transport theory
International Nuclear Information System (INIS)
Zelazny, R.; Kuszell, A.
1962-01-01
Multi-group neutron transport theory. In the paper the general theory of the application of the K. M. Case method to N-group neutron transport theory in plane geometry is given. The eigenfunctions (distributions) for the system of Boltzmann equations have been derived and the completeness theorem has been proved. By means of general solution two examples important for reactor and shielding calculations are given: the solution of a critical and albedo problem for a slab. In both cases the system of singular integral equations for expansion coefficients into a full set of eigenfunction distributions has been reduced to the system of Fredholm-type integral equations. Some results can be applied also to some spherical problems. (author) [fr
Fast neutron therapy in advanced malignant tumour treatment
International Nuclear Information System (INIS)
Avinc, A.
1998-01-01
In this report the fast neutron therapy applications were examined by thoroughly consideration of the fast neutron sources and the interactions of the fast neutron by the medium. The efficacy of fast neutron radiotherapy with that of patients with locally advanced tumours were compared. Radiological data indicate that fast neutrons could bring benefit in the treatment of some tumour types especially salivary glands, paranasal sinuses, soft tissue sarcomas, prostatic adenocarcinomas, palliative treatment of melanoma and rectum. There is a significant improvement in local/regional control for the neutron group, but no improvement in the survival. The neutron therapy is suggested through which this benefit could be achieved
Cosmic-ray neutron transport at a forest field site
DEFF Research Database (Denmark)
Andreasen, Mie; Jensen, Karsten Høgh; Desilets, Darin
2017-01-01
-ray neutron intensity is essential (e.g., the effect of vegetation, litter layer and soil type). In this study the environmental effect is examined by performing a sensitivity analysis using neutron transport modeling. We use a neutron transport model with various representations of the forest and different...
Advanced Transport Operating Systems Program
White, John J.
1990-01-01
NASA-Langley's Advanced Transport Operating Systems Program employs a heavily instrumented, B 737-100 as its Transport Systems Research Vehicle (TRSV). The TRSV has been used during the demonstration trials of the Time Reference Scanning Beam Microwave Landing System (TRSB MLS), the '4D flight-management' concept, ATC data links, and airborne windshear sensors. The credibility obtainable from successful flight test experiments is often a critical factor in the granting of substantial commitments for commercial implementation by the FAA and industry. In the case of the TRSB MLS, flight test demonstrations were decisive to its selection as the standard landing system by the ICAO.
Progress in multidimensional neutron transport computation
International Nuclear Information System (INIS)
Lewis, E.E.
1977-01-01
The methods available for solution of the time-independent neutron transport problems arising in the analysis of nuclear systems are examined. The merits of deterministic and Monte Carlo methods are briefly compared. The capabilities of deterministic computational methods derived from the first-order form of the transport equation, from the second-order even-parity form of this equation, and from integral transport formulations are discussed in some detail. Emphasis is placed on the approaches for dealing with the related problems of computer memory requirements, computational cost, and achievable accuracy. Attention is directed to some areas where problems exist currently and where the need for further work appears to be particularly warranted
Discrete elements method of neutron transport
International Nuclear Information System (INIS)
Mathews, K.A.
1988-01-01
In this paper a new neutron transport method, called discrete elements (L N ) is derived and compared to discrete ordinates methods, theoretically and by numerical experimentation. The discrete elements method is based on discretizing the Boltzmann equation over a set of elements of angle. The discrete elements method is shown to be more cost-effective than discrete ordinates, in terms of accuracy versus execution time and storage, for the cases tested. In a two-dimensional test case, a vacuum duct in a shield, the L N method is more consistently convergent toward a Monte Carlo benchmark solution
Toward whole-core neutron transport without spatial homogenization
International Nuclear Information System (INIS)
Lewis, E. E.
2009-01-01
, iteration between lattice and homogenized calculations yields two-dimensional whole-core results without homogenization error. In the third, planar lattice transport is synthesized with lower-order axial transport to obtain approximate three-dimensional results without planar homogenization. In all three, advances in the method of characteristics play a prominent role, and each rests on some form of equivalence intervention to preserve reaction rates and currents between lattice and homogenized calculations. The talk concludes with some conjectures concerning the potential of interface current, response matrix and related domain decomposition approaches as alternative paths toward achieving whole-core neutronics without homogenization. (author)
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
Neutron transport on the connection machine
International Nuclear Information System (INIS)
Robin, F.
1991-12-01
Monte Carlo methods are heavily used at CEA and account for a a large part of the total CPU time of industrial codes. In the present work (done in the frame of the Parallel Computing Project of the CEL-V Applied Mathematics Department) we study and implement on the Connection Machine an optimised Monte Carlo algorithm for solving the neutron transport equation. This allows us to investigate the suitability of such an architecture for this kind of problem. This report describes the chosen methodology, the algorithm and its performances. We found that programming the CM-2 in CM Fortran is relatively easy and we got interesting performances as, on a 16 k, CM-2 they are the same level as those obtained on one processor of a CRAY X-MP with a well optimized vector code
Advanced spallation neutron sources for condensed matter research
International Nuclear Information System (INIS)
Lovesey, S.W.; Stirling, G.C.
1984-03-01
Advanced spallation neutron sources afford significant advantages over existing high flux reactors. The effective flux is much greater than that currently available with reactor sources. A ten-fold increase in neutron flux will be a major benefit to a wide range of condensed matter studies, and it will realise important experiments that are marginal at reactor sources. Moreover, the high intensity of epithermal neutrons open new vistas in studies of electronic states and molecular vibrations. (author)
The advanced neutron source design - A status report
International Nuclear Information System (INIS)
West, C.D.
1992-01-01
The Advanced Neutron Source (ANS) facility is being designed as a user laboratory for all types of neutron-based research, centered around a nuclear fission reactor (D 2 O cooled, moderated, and reflected), operating at approximately 300 MWth. Safety, and especially passive safety features, have been emphasized throughout the design process. The design also provides experimental facilities for neutron scattering and nuclear and fundamental physics research, transuranic and other isotope production, radiation effects research, and materials analysis. (author)
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
ANL--LASL workshop on advanced neutron detection systems
International Nuclear Information System (INIS)
Kitchens, T.A.
1979-06-01
A two-day workshop on advanced neutron detectors and associated electronics was held in Los Alamos on April 5--6, 1979, as a part of the Argonne National Laboratory--Los Alamos Scientific Laboratory Coordination on neutron scattering instrumentation. This report contains an account of the information presented and conclusions drawn at the workshop
Transport phenomena an introduction to advanced topics
Glasgow, Larry A
2010-01-01
Enables readers to apply transport phenomena principles to solve advanced problems in all areas of engineering and science This book helps readers elevate their understanding of, and their ability to apply, transport phenomena by introducing a broad range of advanced topics as well as analytical and numerical solution techniques. Readers gain the ability to solve complex problems generally not addressed in undergraduate-level courses, including nonlinear, multidimensional transport, and transient molecular and convective transport scenarios. Avoiding rote memorization, the author em
Generic programming for deterministic neutron transport codes
International Nuclear Information System (INIS)
Plagne, L.; Poncot, A.
2005-01-01
This paper discusses the implementation of neutron transport codes via generic programming techniques. Two different Boltzmann equation approximations have been implemented, namely the Sn and SPn methods. This implementation experiment shows that generic programming allows us to improve maintainability and readability of source codes with no performance penalties compared to classical approaches. In the present implementation, matrices and vectors as well as linear algebra algorithms are treated separately from the rest of source code and gathered in a tool library called 'Generic Linear Algebra Solver System' (GLASS). Such a code architecture, based on a linear algebra library, allows us to separate the three different scientific fields involved in transport codes design: numerical analysis, reactor physics and computer science. Our library handles matrices with optional storage policies and thus applies both to Sn code, where the matrix elements are computed on the fly, and to SPn code where stored matrices are used. Thus, using GLASS allows us to share a large fraction of source code between Sn and SPn implementations. Moreover, the GLASS high level of abstraction allows the writing of numerical algorithms in a form which is very close to their textbook descriptions. Hence the GLASS algorithms collection, disconnected from computer science considerations (e.g. storage policy), is very easy to read, to maintain and to extend. (authors)
New era of neutron scattering research on advanced materials
International Nuclear Information System (INIS)
Ikeda, Susumu
2001-01-01
The projects of the next generation of pulsed spallation neutron source are planed in USA, Europe and Japan. They are one order of magnitude more powerful than the most powerful existing neutron source, ISIS in UK. They offer the exciting prospects for the future, and will open the new era of neutron scattering research on advanced materials. The Japanese project is named as the 'Joint project' between JAERI and KEK on high-intensity proton accelerators. The details of the neutron science facility in the 'Joint project' and the sciences to be developed are summarized. (author)
Advanced Neutron Source radiological design criteria
International Nuclear Information System (INIS)
Westbrook, J.L.
1995-08-01
The operation of the proposed Advanced Neutron Source (ANS) facility will present a variety of radiological protection problems. Because it is desired to design and operate the ANS according to the applicable licensing standards of the Nuclear Regulatory Commission (NRC), it must be demonstrated that the ANS radiological design basis is consistent not only with state and Department of Energy (DOE) and other usual federal regulations, but also, so far as is practicable, with NRC regulations and with recommendations of such organizations as the Institute of Nuclear Power Operations (INPO) and the Electric Power Research Institute (EPRI). Also, the ANS radiological design basis is in general to be consistent with the recommendations of authoritative professional and scientific organizations, specifically the National Council on Radiation Protection and Measurements (NCRP) and the International Commission on Radiological Protection (ICRP). As regards radiological protection, the principal goals of DOE regulations and guidance are to keep occupational doses ALARA [as low as (is) reasonably achievable], given the current state of technology, costs, and operations requirements; to control and monitor contained and released radioactivity during normal operation to keep public doses and releases to the environment ALARA; and to limit doses to workers and the public during accident conditions. Meeting these general design objectives requires that principles of dose reduction and of radioactivity control by employed in the design, operation, modification, and decommissioning of the ANS. The purpose of this document is to provide basic radiological criteria for incorporating these principles into the design of the ANS. Operations, modification, and decommissioning will be covered only as they are affected by design
Nobuhara, Fumiyoshi; Kuroyanagi, Makoto; Masumoto, Kazuyoshi; Nakamura, Hajime; Toyoda, Akihiro; Takahashi, Katsuhiko
2017-09-01
In order to evaluate the state of activation in a cyclotron facility used for the radioisotope production of PET diagnostics, we measured the neutron flux by using gold foils and TLDs. Then, the spatial distribution of neutrons and induced activity inside the cyclotron vault were simulated with the Monte Calro calculation code for neutron transport and DCHAIN-SP for activation calculation. The calculated results are in good agreement with measured values within factor 3. Therefore, the adaption of the advanced evaluation procedure for activation level is proved to be important for the planning of decommissioning of these facilities.
Comparison of neutron transport calculations with NRC test results
International Nuclear Information System (INIS)
Koban, J.; Hofmann, W.
1981-02-01
For an exactly defined reactor arrangement (PCA = Pool Critical Assembly) neutron fluxes, neutron spectra and reaction rates for several neutron detectors were calculated by means of one and two dimensional transport codes. An international comparison proved the methods applied at KWU to be adequate. There were difficulties, however, in considering the three dimensions of the assembly which result mainly from its small dimension. This fact applies to all participants who didn't use three dimensional codes. (orig.) [de
Concise four-vector scheme for neutron transport calculations
International Nuclear Information System (INIS)
Kulacsy, K.; Lukacs, B.; Racz, A.
1995-01-01
An explicit Riemannian geometrical form or the vectorial Neutron Streaming Term is presented. The method applies the full Riemannian technique of general covariance. There are cases when the symmetry of the neutron flux must be smaller than that of the arrangement. However, in coordinate space there are always solutions of the Neutron Transport Equation as symmetric as the arrangement, if the latter's symmetry is at least an affine collineation of the Euclidian 3-space. (author). 7 refs
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.
Neutron transport in Eulerian coordinates with bulk material motion
Energy Technology Data Exchange (ETDEWEB)
Baker, Randal S., E-mail: rsb@lanl.gov [Los Alamos National Laboratory, Computational Physics Group, Los Alamos, NM (United States); Dahl, Jon A., E-mail: dahl@lanl.gov [Los Alamos National Laboratory, Computational Physics Group, Los Alamos, NM (United States); Fichtl, Erin J., E-mail: efichtl@lanl.gov [Los Alamos National Laboratory, Computational Physics Group, Los Alamos, NM (United States); Morel, Jim E., E-mail: morel@tamu.edu [Department of Nuclear Engineering, Texas A& M University, College Station, TX (United States)
2015-12-15
A consistent, numerically stable algorithm for the solution of the neutron transport equation in the presence of a moving material background is presented for one-dimensional spherical geometry. Manufactured solutions are used to demonstrate the correctness and stability of our numerical algorithm. The importance of including moving material corrections is shown for the r-process in proto-neutron stars.
International Nuclear Information System (INIS)
Pillon, M.; Martone, M.; Verschuur, K.A.; Jarvis, O.N.; Kaellne, J.
1989-01-01
Neutron transport calculations have been performed using fluence ray tracing (FURNACE code) and Monte Carlo particle trajectory sampling methods (MCNP code) in order to determine the neutron fluence and energy distributions at different locations in the JET tokamak. These calculations were used to calibrate the activation measurements used in the determination of the absolute fusion neutron yields from the JET plasma. We present here the neutron activation response coefficients calculated for three different materials. Comparison of the MCNP and FURNACE results helps identify the sources of error in these neutron transport calculations. The accuracy of these calculations was tested by comparing the total 2.5 MeV neutron yields derived from the activation measurements with those obtained with calibrated fission chambers; agreement at the ±15% level was demonstrate. (orig.)
Advanced geometries for ballistic neutron guides
International Nuclear Information System (INIS)
Schanzer, Christian; Boeni, Peter; Filges, Uwe; Hils, Thomas
2004-01-01
Sophisticated neutron guide systems take advantage of supermirrors being used to increase the neutron flux. However, the finite reflectivity of supermirrors becomes a major loss mechanism when many reflections occur, e.g. in long neutron guides and for long wavelengths. In order to reduce the number of reflections, ballistic neutron guides have been proposed. Usually linear tapered sections are used to enlarge the cross-section and finally, focus the beam to the sample. The disadvantages of linear tapering are (i) an inhomogeneous phase space at the sample position and (ii) a decreasing flux with increasing distance from the exit of the guide. We investigate the properties of parabolic and elliptic tapering for ballistic neutron guides, using the Monte Carlo program McStas with a new guide component dedicated for such geometries. We show that the maximum flux can indeed be shifted away from the exit of the guide. In addition we explore the possibilities of parabolic and elliptic geometries to create point like sources for dedicated experimental demands
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
Advanced neutron diagnostics for ITER fusion experiments
International Nuclear Information System (INIS)
Kaellne, J.; Giacomelli, L.; Hjalmarsson, A.; Conroy, S.; Ericsson, G.; Johnson, M.G.; Glasser, W.; Henriksson, H.; Ronchi, E.; Sjoestrand, H.; Andersson, E.S.; Thun, J.; Weiszflog, M.; Gorini, G.; Tardocchi, M.; Popovichev, S.; Sousa, J.
2005-01-01
Results are presented from the neutron emission spectroscopy (NES) diagnosis of JET plasma performed with the MPR during the DTE1 campaign of 1997 and the recent TTE of 2003. The NES diagnostic capabilities at JET are presently being drastically enhanced by an upgrade of the MPR (MPRu) and a new 2.5-MeV TOF neutron spectrometer (TOFOR). The principles of MPRu and TOFOR are described and illustrated with the diagnostic role they will play in the high performance fusion experiments in the forward program of JET largely aimed at supporting ITER. The importance for the JET NES effort for ITER is discussed. (author)
Recent advances in neutron capture therapy (NCT)
International Nuclear Information System (INIS)
Fairchild, R.G.
1985-01-01
The application of the 10 B(n,α) 7 Li reaction to cancer radiotherapy (Neutron Capture therapy, or NCT) has intrigued investigators since the discovery of the neutron. This paper briefly summarizes data describing recently developed boronated compounds with evident tumor specificity and extended biological half-lives. The implication of these compounds to NCT is evaluated in terms of Therapeutic Gain (TG). The optimization of NCT using band-pass filtered beams is described, again in terms of TG, and irradiation times with these less intense beams are estimated. 24 refs., 3 figs., 3 tabs
Advanced neutron instrumentation at FRM-II
International Nuclear Information System (INIS)
Petry, Winfried
2003-01-01
The construction of the new German high flux neutron source FRM-II is finished and FRM-II is waiting for its licence to start nuclear operation. With the beginning of the routine operation 22 instruments will be in action, including 5 irradiation facilities and 17 beam tube instruments, most of them use neutron scattering techniques. Additional instruments are under construction. Some of these instruments are unique, others are expected to be the best of their kind, all instruments are based on innovative techniques. (author)
Recent advances in laser-driven neutron sources
Alejo, A.; Ahmed, H.; Green, A.; Mirfayzi, S. R.; Borghesi, M.; Kar, S.
2016-11-01
Due to the limited number and high cost of large-scale neutron facilities, there has been a growing interest in compact accelerator-driven sources. In this context, several potential schemes of laser-driven neutron sources are being intensively studied employing laser-accelerated electron and ion beams. In addition to the potential of delivering neutron beams with high brilliance, directionality and ultra-short burst duration, a laser-driven neutron source would offer further advantages in terms of cost-effectiveness, compactness and radiation confinement by closed-coupled experiments. Some of the recent advances in this field are discussed, showing improvements in the directionality and flux of the laser-driven neutron beams.
Advancement of neutron radiography technique in JRR-3M
International Nuclear Information System (INIS)
Matsubayashi, Masahito
1999-01-01
The JRR-3M thermal neutron radiography facility (JRR-3M TNRF) was completed in the JRR-3M of the Japan Atomic Energy Research Institute in 1991 and has been utilized as research tools for various kinds of research fields such as thermal hydraulic researches, agricultural researches, medical researches, archaeological researches and so on. High performance of the JRR-3M TNRF such as high neutron flux, high collimator ratio and wide radiographing field has enabled advanced researches and stimulated developments of advanced neutron radiography (NR) systems for higher spatial resolution and for higher temporal resolution. Static NR systems using neutron imaging plates or cooled CCD camera with high spatial resolution, a real-time NR system using a silicon intensifier target tube camera and a high-frame-rate NR system using a combination of an image intensifier and a high speed digital video camera with high temporal resolution have been developed to fill the requirements from researchers. (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
An Advanced Neutron Spectrometer for Future Manned Exploration Missions
Christl, Mark; Apple, Jeffrey A.; Cox, Mark D.; Dietz, Kurtis L.; Dobson, Christopher C.; Gibson, Brian F.; Howard, David E.; Jackson, Amanda C.; Kayatin, Mathew J.; Kuznetsov, Evgeny N.;
2014-01-01
An Advanced Neutron Spectrometer (ANS) is being developed to support future manned exploration missions. This new instrument uses a refined gate and capture technique that significantly improves the identification of neutrons in mixed radiation fields found in spacecraft, habitats and on planetary surfaces. The new instrument is a composite scintillator comprised of PVT loaded with litium-6 glass scintillators. We will describe the detection concept and show preliminary results from laboratory tests and exposures at particle accelerators
Optics for Advanced Neutron Imaging and Scattering
International Nuclear Information System (INIS)
Moncton, David E.; Khaykovich, Boris
2016-01-01
During the report period, we continued the work as outlined in the original proposal. We have analyzed potential optical designs of Wolter mirrors for the neutron-imaging instrument VENUS, which is under construction at SNS. In parallel, we have conducted the initial polarized imaging experiment at Helmholtz Zentrum, Berlin, one of very few of currently available polarized-imaging facilities worldwide.
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)
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)
Neutron irradiation effects in advanced superconductors
International Nuclear Information System (INIS)
Yoshida, H.; Kodaka, H.; Miyata, K.; Hayashi, Y.; Atobe, K.
1988-01-01
This paper reports the effects of neutron irradiation on superconducting transitions studied by susceptibility and resistivity measurements for A15 type compounds, Laves-phase compounds and oxide superconductors. For A15 superconductors, the transition temperature (T c ) decreased with increasing neutron fluence and showed large drop started at about 5 x 10 18 n/cm 2 (E > 0.1 MeV). Post-irradiation annealing gave recovery of T c , but the behaviors were different for the materials with different composition and microstructure. The Laves-phase compounds showed less degradation than the A15 superconductors. For oxide superconductors very sensitive transition change was observed, including the radiation-induced superconductivity
Transport coefficients in superfluid neutron stars
Energy Technology Data Exchange (ETDEWEB)
Tolos, Laura [Instituto de Ciencias del Espacio (IEEC/CSIC) Campus Universitat Autònoma de Barcelona, Facultat de Ciències, Torre C5, E-08193 Bellaterra (Barcelona) (Spain); Frankfurt Institute for Advances Studies. Johann Wolfgang Goethe University, Ruth-Moufang-Str. 1, 60438 Frankfurt am Main (Germany); Manuel, Cristina [Instituto de Ciencias del Espacio (IEEC/CSIC) Campus Universitat Autònoma de Barcelona, Facultat de Ciències, Torre C5, E-08193 Bellaterra (Barcelona) (Spain); Sarkar, Sreemoyee [Tata Institute of Fundamental Research, Homi Bhaba Road, Mumbai-400005 (India); Tarrus, Jaume [Physik Department, Technische Universität München, D-85748 Garching (Germany)
2016-01-22
We study the shear and bulk viscosity coefficients as well as the thermal conductivity as arising from the collisions among phonons in superfluid neutron stars. We use effective field theory techniques to extract the allowed phonon collisional processes, written as a function of the equation of state and the gap of the system. The shear viscosity due to phonon scattering is compared to calculations of that coming from electron collisions. We also comment on the possible consequences for r-mode damping in superfluid neutron stars. Moreover, we find that phonon collisions give the leading contribution to the bulk viscosities in the core of the neutron stars. We finally obtain a temperature-independent thermal conductivity from phonon collisions and compare it with the electron-muon thermal conductivity in superfluid neutron stars.
Neutron transport from targets to moderators
International Nuclear Information System (INIS)
Taylor, A.D.
1981-06-01
By appropriately choosing parameters such as temperature, decoupler, thickness and effective size it is possible to tailor the moderators of a pulsed spallation neutron source in such a way that the different characteristics regarding time structure and spectral distribution as requested for the different instruments can be met very closely. This enables a unique flexibility in the design of neutron spectrometers to be used at such a source. (author)
Effect of granulation of geological samples in neutron transport measurements
International Nuclear Information System (INIS)
Woznicka, Urszula; Drozdowicz, Krzysztof; Gabanska, Barbara; Krynicka, Ewa; Igielski, Andrzej
2001-01-01
The thermal neutron absorption cross section is one of the parameters describing the transport of thermal neutrons in a medium. Theoretical descriptions and experiments which determine the absorption cross section have a wide literature for homogeneous media. The situation comes true e.g. for fluids or amorphous solids. There are many other media which should be treated as heterogeneous. Among others - geological materials. The material heterogeneity for the thermal neutron transport in a considered volume is understood here as an existence of many small regions which differ significantly in their macroscopic neutron diffusion parameters (defined by the absorption and transport cross sections). The final difference, which influences the neutron transport, comes from a combination of the absolute differences between the parameters and of sizes of regions (related to the neutron mean free paths). A rock can be naturally heterogeneous in the above meaning. Besides, it can happen that a preparation of the rock sample for a neutron measurement can increase its natural heterogeneity. (For example, when the rock material is crushed and the measured sample consists of the obtained grains). The question is which granulation is allowed to treat the sample material as still homogeneous, and from which size of the rock grains we have to consider a two-component medium. It has been experimentally proved that the effective absorption of thermal neutrons in a heterogeneous two-component material can significantly differ from the absorption in a homogeneous one which consists of the same elements. The final effect is dependent on a few factors: the macroscopic absorption cross sections of the components, their total mass contributions, and the size of the grains. The ratio of the effective absorption cross section of the heterogeneous material to the cross section of the equivalent homogeneous, is a measure of the heterogeneity effect on the thermal neutron absorption
Advances in associated-particle sealed-tube neutron probe diagnostics for substance detection
International Nuclear Information System (INIS)
Rhodes, E.; Dickerman, C.E.; Frey, M.
1995-01-01
The development and investigation of a small associated-particle sealed-tube neutron generator (APSTNG) shows potential to allow the associated-particle diagnostic method to be moved out of the laboratory into field applications. The APSTNG interrogates the inspected object with 14-MeV neutrons generated from the deuterium-tritium reaction and detects the alpha-particle associated with each neutron inside a cone encompassing the region of interest. Gamma-ray spectra of resulting neutron reactions identify many nuclides. Flight-times determined from detection times of the gamma-rays and alpha-particles can yield a separate coarse tomographic image of each identified nuclide, from a single orientation. Chemical substances are identified by comparing relative spectral line intensities with ratios of elements in reference compounds. The high-energy neutrons and gamma-rays penetrate large objects and dense materials. Generally no collimators or radiation shielding are needed. Proof-of-concept laboratory experiments have been successfully performed for simulated nuclear, chemical warfare, and conventional munitions. Most recently, inspection applications have been investigated for radioactive waste characterization, presence of cocaine in propane tanks, and uranium and plutonium smuggling. Based on lessons learned with the present APSTNG system, an advanced APSTNG tube (along with improved high voltage supply and control units) is being designed and fabricated that will be transportable and rugged, yield a substantial neutron output increase, and provide sufficiently improved lifetime to allow operation at more than an order of magnitude increase in neutron flux
Research reactor of the future: The advanced neutron source
International Nuclear Information System (INIS)
Appleton, B.; West, C.
1994-01-01
Agents for cancer detection and treatment, stronger materials, better electronic gadgets, and other consumer and industrial products - these are assured benefits of a research reactor project proposed for Oak Ridge. Just as American companies have again assumed world leadership in producing semiconductor chips as well as cars and trucks, the United States is poised to retake the lead in neutron science by building and operating the $2.9 billion Advanced Neutron Source (ANS) research reactor by the start of the next century. In 1985, the neutron community, led by ORNL researchers, proposed a pioneering project, later called the ANS. Scheduled to begin operation in 2003, the ANS is seen not only as a replacement for the aging HFIR and HFBR but also as the best laboratory in the world for conducting neutron-based research
Neutron transport from targets to moderators
International Nuclear Information System (INIS)
Taylor, A.D.
1980-01-01
The title of this meeting is 'Targets for Neutron Beam Spallation Sources', but so far all the emphasis in the talks has been on how to produce the fast neutron flux. I would like to stress that that is just the beginning of the story. What we are required to produce are beams of thermal and epithermal neutrons with time and spectral characteristics tailored to the instrumental requirements. The real source of our neutrons is not uranium arrays or thorium cylinders but a small volume of hydrogenous material, some 10 x 10 x 5 cm 3 . This is really what the whole thing is about - the target produces a copious field of fast neutrons, but if we fail to moderate them with the right energy and time characteristics, we will not match to what is happening downstream. In this talk, I am going to deal specifically with what we have done for SNS to optimise the target-moderator-reflector and decoupler system in this respect. (orig.)
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
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)
Transport of accelerator produced high energy neutrons though concrete
International Nuclear Information System (INIS)
Prabhakar Rao, G.; Sarkar, P.K.
1996-01-01
Development of a computational system for estimating the production and transport of high energy neutrons in particle accelerators is reported. The energy-angle distribution of neutrons from accelerated ions bombarding thick targets is calculated by a hybrid nuclear reaction model code, ALICE-91, modified to suit the purpose. Subsequent transmission of these neutrons through concrete slabs is treated using the anisotropic source-flux iteration technique (ASFIT) in the framework of a coupled neutron-gamma transport. Several parameters of both the codes have been optimized to obtain the transmitted dose through concrete. The calculations are found to be accurate and at the same time faster compared to the detailed Monte Carlo calculations. (author). 8 refs., 2 figs
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.
On generating neutron transport tables with the NJOY system
International Nuclear Information System (INIS)
Caldeira, Alexandre D.; Claro, Luiz H.
2013-01-01
Incorrect values for the product of the average number of neutrons released per fission and the fission microscopic cross-section were detected in several energy groups of a neutron transport table generated with the most updated version of the NJOY system. It was verified that the problem persists when older versions of this system are utilized. Although this problem exists for, at least, ten years, it is still an open question. (author)
The Advanced Neutron Source design: A status report
International Nuclear Information System (INIS)
West, C.D.
1992-01-01
The Advanced Nuetron Source (ANS) facility is being designed as a user laboratory for all types of neutron-based research, centered around a nuclear fission reactor (D 2 O cooled, moderated, and reflected), operating at approximately 300 MW th . Safety, and especially passive safety features, have been emphasized throughout the design process
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.)
Measurements of anomalous neutron transport in bulk graphite
Energy Technology Data Exchange (ETDEWEB)
Bowman, C.D.; Smith, G.A. [ADNA Corp., Los Alamos, NM (United States); Vogelaar, B. [Virginia Tech., Blacksburg, VA (United States); Howell, C.R.; Bilpuch, E.G.; Tornow, W. [Triangle Univ. Nuclear Lab., Duke Univ., Durham, NC (United States)
2003-07-01
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.)
Guideline of Monte Carlo calculation. Neutron/gamma ray transport simulation by Monte Carlo method
2002-01-01
This report condenses basic theories and advanced applications of neutron/gamma ray transport calculations in many fields of nuclear energy research. Chapters 1 through 5 treat historical progress of Monte Carlo methods, general issues of variance reduction technique, cross section libraries used in continuous energy Monte Carlo codes. In chapter 6, the following issues are discussed: fusion benchmark experiments, design of ITER, experiment analyses of fast critical assembly, core analyses of JMTR, simulation of pulsed neutron experiment, core analyses of HTTR, duct streaming calculations, bulk shielding calculations, neutron/gamma ray transport calculations of the Hiroshima atomic bomb. Chapters 8 and 9 treat function enhancements of MCNP and MVP codes, and a parallel processing of Monte Carlo calculation, respectively. An important references are attached at the end of this report.
Recent advances on uric acid transporters
Xu, Liuqing; Shi, Yingfeng; Zhuang, Shougang; Liu, Na
2017-01-01
Uric acid is the product of purine metabolism and its increased levels result in hyperuricemia. A number of epidemiological reports link hyperuricemia with multiple disorders, such as kidney diseases, cardiovascular diseases and diabetes. Recent studies also showed that expression and functional changes of urate transporters are associated with hyperuricemia. Uric acid transporters are divided into two categories: urate reabsorption transporters, including urate anion transporter 1 (URAT1), organic anion transporter 4 (OAT4) and glucose transporter 9 (GLUT9), and urate excretion transporetrs, including OAT1, OAT3, urate transporter (UAT), multidrug resistance protein 4 (MRP4/ABCC4), ABCG-2 and sodium-dependent phosphate transport protein. In the kidney, uric acid transporters decrease the reabsorption of urate and increase its secretion. These transporters’ dysfunction would lead to hyperuricemia. As the function of urate transporters is important to control the level of serum uric acid, studies on the functional role of uric acid transporter may provide a new strategy to treat hyperuricemia associated diseases, such as gout, chronic kidney disease, hyperlipidemia, hypertension, coronary heart disease, diabetes and other disorders. This review article summarizes the physiology of urate reabsorption and excretion transporters and highlights the recent advances on their roles in hyperuricemia and various diseases. PMID:29246027
The Lattice Boltzmann Method applied to neutron transport
International Nuclear Information System (INIS)
Erasmus, B.; Van Heerden, F. A.
2013-01-01
In this paper the applicability of the Lattice Boltzmann Method to neutron transport is investigated. One of the main features of the Lattice Boltzmann method is the simultaneous discretization of the phase space of the problem, whereby particles are restricted to move on a lattice. An iterative solution of the operator form of the neutron transport equation is presented here, with the first collision source as the starting point of the iteration scheme. A full description of the discretization scheme is given, along with the quadrature set used for the angular discretization. An angular refinement scheme is introduced to increase the angular coverage of the problem phase space and to mitigate lattice ray effects. The method is applied to a model problem to investigate its applicability to neutron transport and the results are compared to a reference solution calculated, using MCNP. (authors)
A modular spherical harmonics approach to the neutron transport equation
International Nuclear Information System (INIS)
Inanc, F.; Rohach, A.F.
1989-01-01
A modular nodal method was developed for solving the neutron transport equation in 2-D xy coordinates. The spherical harmonic expansion was used for approximating the second-order even-parity form of the neutron transport equation. The boundary conditions of the spherical harmonics approximation were derived in a form to have forms analogous to the partial currents in the neutron diffusion equation. Relations were developed for generating both the second-order spherical harmonic equations and the boundary conditions in an automated computational algorithm. Nodes using different orders of the spherical harmonics approximation to the transport equation were interfaced through mixed-type boundary conditions. The determination of spherical harmonic orders implemented in the nodes were determined by the scheme in an automated manner. Results of the method compared favorably to benchmark problems. (author)
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)
A Monte Carlo Green's function method for three-dimensional neutron transport
International Nuclear Information System (INIS)
Gamino, R.G.; Brown, F.B.; Mendelson, M.R.
1992-01-01
This paper describes a Monte Carlo transport kernel capability, which has recently been incorporated into the RACER continuous-energy Monte Carlo code. The kernels represent a Green's function method for neutron transport from a fixed-source volume out to a particular volume of interest. This method is very powerful transport technique. Also, since kernels are evaluated numerically by Monte Carlo, the problem geometry can be arbitrarily complex, yet exact. This method is intended for problems where an ex-core neutron response must be determined for a variety of reactor conditions. Two examples are ex-core neutron detector response and vessel critical weld fast flux. The response is expressed in terms of neutron transport kernels weighted by a core fission source distribution. In these types of calculations, the response must be computed for hundreds of source distributions, but the kernels only need to be calculated once. The advance described in this paper is that the kernels are generated with a highly accurate three-dimensional Monte Carlo transport calculation instead of an approximate method such as line-of-sight attenuation theory or a synthesized three-dimensional discrete ordinates solution
Cold source vessel development for the advanced neutron source
Energy Technology Data Exchange (ETDEWEB)
Williams, P.T.; Lucas, A.T. [Oak Ridge National Lab., TN (United States)
1995-09-01
The Advanced Neutron Source (ANS), in its conceptual design phase at Oak Ridge National Laboratory (ORNL), will be a user-oriented neutron research facility that will produce the most intense flux of neutrons in the world. Among its many scientific applications, the productions of cold neutrons is a significant research mission for the ANS. The cold neutrons come from two independent cold sources positioned near the reactor core. Contained by an aluminum alloy vessel, each cold source is a 410 mm diameter sphere of liquid deuterium that functions both as a neutron moderator and a cryogenic coolant. With nuclear heating of the containment vessel and internal baffling, steady-state operation requires close control of the liquid deuterium flow near the vessel`s inner surface. Preliminary thermal-hydraulic analyses supporting the cold source design are being performed with multi-dimensional computational fluid dynamics simulations of the liquid deuterium flow and heat transfer. This paper presents the starting phase of a challenging program and describes the cold source conceptual design, the thermal-hydraulic feasibility studies of the containment vessel, and the future computational and experimental studies that will be used to verify the final design.
International Nuclear Information System (INIS)
Iwasaki, Tomohiko; Hirakawa, Naohiro
1995-01-01
Transmutation of Minor Actinide (MA) in a well thermalized neutron field was studied. Since MA nuclides have large effective cross sections in the well thermalized neutron field, the transmutation in the well thermalized neutron field has an advantage of high transmutation rate. However, the transmutation rate largely decreases by accumulation of 246 Cm when MA is transmuted only in the well thermalized neutron field for a long period. An acceleration method of burn-up of 246 Cm was studied. High transmutation rate can be obtained by providing a neutron field with high flux in the energy region between 1 and 100 eV. Two stage transmutation using the well thermalized neutron field and this field can transmute MA rapidly. The applicability of the Advanced Neutron Source (ANS) to the transmutation of MA was examined for a typical MA with the composition in the high-level waste generated in the conventional PWR. If the ANS is applied without changing the fuel inventory, the amount of MA which corresponds to that produced by a conventional 1,175 MWe PWR in one year can be transmuted by the ANS in one year. Furthermore, the amount of the residual can be reduced to about 1g (10 -5 of the initial MA weight) by continuing the transmutation for 5 years owing to the two stage transmutation. (author)
Advanced technology for future regional transport aircraft
Williams, L. J.
1982-01-01
In connection with a request for a report coming from a U.S. Senate committee, NASA formed a Small Transport Aircraft Technology (STAT) team in 1978. STAT was to obtain information concerning the technical improvements in commuter aircraft that would likely increase their public acceptance. Another area of study was related to questions regarding the help which could be provided by NASA's aeronautical research and development program to commuter aircraft manufacturers with respect to the solution of technical problems. Attention is given to commuter airline growth, current commuter/region aircraft and new aircraft in development, prospects for advanced technology commuter/regional transports, and potential benefits of advanced technology. A list is provided of a number of particular advances appropriate to small transport aircraft, taking into account small gas turbine engine component technology, propeller technology, three-dimensional wing-design technology, airframe aerodynamics/propulsion integration, and composite structure materials.
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
Advanced transport modeling of toroidal plasmas with transport barriers
International Nuclear Information System (INIS)
Fukuyama, A.; Murakami, S.; Honda, M.; Izumi, Y.; Yagi, M.; Nakajima, N.; Nakamura, Y.; Ozeki, T.
2005-01-01
Transport modeling of toroidal plasmas is one of the most important issue to predict time evolution of burning plasmas and to develop control schemes in reactor plasmas. In order to describe the plasma rotation and rapid transition self-consistently, we have developed an advanced scheme of transport modeling based on dynamical transport equation and applied it to the analysis of transport barrier formation. First we propose a new transport model and examine its behavior by the use of conventional diffusive transport equation. This model includes the electrostatic toroidal ITG mode and the electromagnetic ballooning mode and successfully describes the formation of internal transport barriers. Then the dynamical transport equation is introduced to describe the plasma rotation and the radial electric field self-consistently. The formation of edge transport barriers is systematically studied and compared with experimental observations. The possibility of kinetic transport modeling in velocity space is also examined. Finally the modular structure of integrated modeling code for tokamaks and helical systems is discussed. (author)
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)
A status report on the Advanced Neutron Source project
International Nuclear Information System (INIS)
West, C.D.
1993-01-01
The Advanced Neutron Source (ANS) will be a new laboratory for neutron research, centered around a 330 MW(f) research reactor cooled and reflected by heavy water and including extensive experiment systems and support facilities. The major components of the baseline design, occupying about 16 heetares, are a guide hall/research support area, containing most of the neutron beam experiment systems, shops and supporting laboratories; a 60 m diameter containment building housing the reactor and its pimary coolant system, and selected scientific research facilities; an operations support building with the majority of the remaining plant systems, an office/interface complex providing a carefully designed, user friendly entry point for access control; and several other major facilities including user housing, an electrical substation, a diesel generator building, a cryorefrigerator building, and heavy water cleanup and upgrade systems
The advanced neutron source safety approach and plans
International Nuclear Information System (INIS)
Harrington, R.M.
1989-01-01
The Advanced Neutron Source (ANS) is a user facility for all areas of neutron research proposed for construction at the Oak Ridge National Laboratory. The neutron source is planned to be a 350-MW research reactor. The reactor, currently in conceptual design, will belong to the United States Department of Energy (USDOE). The safety approach and planned elements of the safety program for the ANS are described. The safety approach is to incorporate USDOE requirements [which, by reference, include appropriate requirements from the United States Nuclear Regulatory Commission (USNRC) and other national and state regulatory agencies] into the design, and to utilize probabilistic risk assessment (PRA) techniques during design to achieve extremely low probability of severe core damage. The PRA has already begun and will continue throughout the design and construction of the reactor. Computer analyses will be conducted for a complete spectrum of accidental events, from anticipated events to very infrequent occurrences. 8 refs., 2 tabs
Scientific opportunities with advanced facilities for neutron scattering
International Nuclear Information System (INIS)
Lander, G.H.; Emery, V.J.
1984-01-01
The present report documents deliberations of a large group of experts in neutron scattering and fundamental physics on the need for new neutron sources of greater intensity and more sophisticated instrumentation than those currently available. An additional aspect of the Workshop was a comparison between steady-state (reactor) and pulsed (spallation) sources. The main conclusions were: (1) the case for a new higher flux neutron source is extremely strong and such a facility will lead to qualitatively new advances in condensed matter science and fundamental physics; (2) to a large extent the future needs of the scientific community could be met with either a 5 x 10 15 n cm -2 s -1 steady state source or a 10 17 n cm -2 s -1 peak flux spallation source; and (3) the findings of this Workshop are consistent with the recommendations of the Major Materials Facilities Committee
The Advanced Neutron Source safety approach and plans
International Nuclear Information System (INIS)
Harrington, R.M.
1990-01-01
The Advanced Neutron Source (ANS) is a user facility proposed for construction at the Oak Ridge National Laboratory for all areas of neutron research. The neutron source is planned to be a 350-MW research reactor. The reactor, currently in conceptual design, will belong to the United States Department of Energy (USDOE). The safety approach and planned elements of the safety program for the ANS are described. The safety approach is to incorporate USDOE requirements (which, by reference, include appropriate requirements from the United States Nuclear Regulatory Commission (USNRC) and other national and state regulatory agencies) into the design, and to utilize probabilistic risk assessment (PRA) techniques during design to achieve extremely low probability of severe core damage. The PRA has already begun and will continue throughout the design and construction of the reactor. Computer analyses will be conducted for a complete spectrum of accidental events, from anticipated events to very infrequent occurrences
Recent advances in neutral particle transport methods and codes
International Nuclear Information System (INIS)
Azmy, Y.Y.
1996-01-01
An overview of ORNL's three-dimensional neutral particle transport code, TORT, is presented. Special features of the code that make it invaluable for large applications are summarized for the prospective user. Advanced capabilities currently under development and installation in the production release of TORT are discussed; they include: multitasking on Cray platforms running the UNICOS operating system; Adjacent cell Preconditioning acceleration scheme; and graphics codes for displaying computed quantities such as the flux. Further developments for TORT and its companion codes to enhance its present capabilities, as well as expand its range of applications are disucssed. Speculation on the next generation of neutron particle transport codes at ORNL, especially regarding unstructured grids and high order spatial approximations, are also mentioned
The advanced neutron source research and development plan
International Nuclear Information System (INIS)
Selby, D.L.
1995-08-01
The Advanced Neutron Source (ANS) is being designed as a user-oriented neutron research laboratory centered around the most intense continuous beams of thermal and subthermal neutrons in the world (an order of magnitude more intense than beams available from the most advanced existing reactors). The ANS will be built around a new research reactor of 330-MW fission power, producing an unprecedented peak thermal flux of >7 · 10 19 · m -2 · s -1 . Primarily a research facility, the ANS will accommodate more than 1000 academic, industrial, and government researchers each year. They will conduct basic research in all branches of science as well as applied research leading to better understanding of new materials, including high temperature super conductors, plastics, and thin films. Some 48 neutron beam stations will be set up in the ANS beam rooms and the neutron guide hall for neutron scattering and for fundamental and nuclear physics research. There also will be extensive facilities for materials irradiation, isotope production, and analytical chemistry. The top level work breakdown structure (WBS) for the project. As noted in this figure, one component of the project is a research and development (R ampersand D) program (WBS 1.1). This program interfaces with all of the other project level two WBS activities. Because one of the project guidelines is to meet minimum performance goals without relying on new inventions, this R ampersand D activity is not intended to produce new concepts to allow the project to meet minimum performance goals. Instead, the R ampersand D program will focus on the four objectives described
Finite element method for solving neutron transport problems
International Nuclear Information System (INIS)
Ferguson, J.M.; Greenbaum, A.
1984-01-01
A finite element method is introduced for solving the neutron transport equations. Our method falls into the category of Petrov-Galerkin solution, since the trial space differs from the test space. The close relationship between this method and the discrete ordinate method is discussed, and the methods are compared for simple test problems
Neutron transport in two dissimilar media anisotropic scattering
International Nuclear Information System (INIS)
Burkart, A.R.; Ishiguro, Y.; Siewert, C.E.
1976-01-01
The elementary solution of the one-speed neutron-transport equation with linearly anisotropic scattering are used in conjunction with Chandrasekhar's invariance principles to solve in a concise manner the Milne problem for two adjoining half-spaces and the critical reactor problem for a reflected slab
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.
Neutron transport calculations of some fast critical assemblies
International Nuclear Information System (INIS)
Martinez-Val Penalosa, J. A.
1976-01-01
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
Advanced public transportation systems : evaluation guidelines
1994-01-01
The Federal Transit Administration has developed the Advanced Public Transportation Systems (APTS) Program which is an integral part of the overall U.S. DOT Intelligent Vehicle Highway Systems (IVHS) effort. A major aim of the APTS Program is to prom...
Dynamic modeling of the advanced neutron source reactor
International Nuclear Information System (INIS)
March-Leuba, J.; Ibn-Khayat, M.
1990-01-01
The purpose of this paper is to provide a summary description and some applications of a computer model that has been developed to simulate the dynamic behavior of the advanced neutron source (ANS) reactor. The ANS dynamic model is coded in the advanced continuous simulation language (ACSL), and it represents the reactor core, vessel, primary cooling system, and secondary cooling systems. The use of a simple dynamic model in the early stages of the reactor design has proven very valuable not only in the development of the control and plant protection system but also of components such as pumps and heat exchangers that are usually sized based on steady-state calculations
Optimization of a neutron detector design using adjoint transport simulation
International Nuclear Information System (INIS)
Yi, C.; Manalo, K.; Huang, M.; Chin, M.; Edgar, C.; Applegate, S.; Sjoden, G.
2012-01-01
A synthetic aperture approach has been developed and investigated for Special Nuclear Materials (SNM) detection in vehicles passing a checkpoint at highway speeds. SNM is postulated to be stored in a moving vehicle and detector assemblies are placed on the road-side or in chambers embedded below the road surface. Neutron and gamma spectral awareness is important for the detector assembly design besides high efficiencies, so that different SNMs can be detected and identified with various possible shielding settings. The detector assembly design is composed of a CsI gamma-ray detector block and five neutron detector blocks, with peak efficiencies targeting different energy ranges determined by adjoint simulations. In this study, formulations are derived using adjoint transport simulations to estimate detector efficiencies. The formulations is applied to investigate several neutron detector designs for Block IV, which has its peak efficiency in the thermal range, and Block V, designed to maximize the total neutron counts over the entire energy spectrum. Other Blocks detect different neutron energies. All five neutron detector blocks and the gamma-ray block are assembled in both MCNP and deterministic simulation models, with detector responses calculated to validate the fully assembled design using a 30-group library. The simulation results show that the 30-group library, collapsed from an 80-group library using an adjoint-weighting approach with the YGROUP code, significantly reduced the computational cost while maintaining accuracy. (authors)
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
Homogenization of the critically spectral equation in neutron transport
Energy Technology Data Exchange (ETDEWEB)
Allaire, G. [CEA Saclay, 91 - Gif-sur-Yvette (France). Dept. de Mecanique et de Technologie]|[Paris-6 Univ., 75 (France). Lab. d' Analyse Numerique; Bal, G. [Electricite de France (EDF), 92 - Clamart (France). Direction des Etudes et Recherches
1998-07-01
We address the homogenization of an eigenvalue problem for the neutron transport equation in a periodic heterogeneous domain, modeling the criticality study of nuclear reactor cores. We prove that the neutron flux, corresponding to the first and unique positive eigenvector, can be factorized in the product of two terms, up to a remainder which goes strongly to zero with the period. On terms is the first eigenvector of the transport equation in the periodicity cell. The other term is the first eigenvector of a diffusion equation in the homogenized domain. Furthermore, the corresponding eigenvalue gives a second order corrector for the eigenvalue of the heterogeneous transport problem. This result justifies and improves the engineering procedure used in practice for nuclear reactor cores computations. (author)
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
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)
Rapid Measurement of Neutron Dose Rate for Transport Index
International Nuclear Information System (INIS)
Morris, R.L.
2000-01-01
A newly available neutron dose equivalent remmeter with improved sensitivity and energy response has been put into service at Rocky Flats Environmental Technology Site (RFETS). This instrument is being used to expedite measurement of the Transport Index and as an ALARA tool to identify locations where slightly elevated neutron dose equivalent rates exist. The meter is capable of measuring dose rates as low as 0.2 μSv per hour (20 μrem per hour). Tests of the angular response and energy response of the instrument are reported. Calculations of the theoretical instrument response made using MCNPtrademark are reported for materials typical of those being shipped
Discontinuous nodal schemes applied to the bidimensional neutron transport equation
International Nuclear Information System (INIS)
Delfin L, A.; Valle G, E. Del; Hennart B, J.P.
1996-01-01
In this paper several strong discontinuous nodal schemes are described, starting from the one that has only two interpolation parameters per cell to the one having ten. Their application to the spatial discretization of the neutron transport equation in X-Y geometry is also described, giving, for each one of the nodal schemes, the approximation for the angular neutron flux that includes the set of interpolation parameters and the corresponding polynomial space. Numerical results were obtained for several test problems presenting here the problem with the highest degree of difficulty and their comparison with published results 1,2 . (Author)
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)
The Advanced Neutron Source research and development plan
International Nuclear Information System (INIS)
Selby, D.L.
1992-01-01
The Advanced Neutron Source (ANS) is being designed as a user-oriented neutron research laboratory centered around the most intense continuous beams of thermal and subthermal neutrons in the world. The ANS will be built around a new research reactor of ∼ 330 MW fission power, producing an unprecedented peak thermal flux of > 7 x 10 19 M -2 · S -1 . Primarily a research facility, the ANS will accommodate more than 1000 academic, industrial, and government researchers each year. They will conduct basic research in all branches of science-as well as applied research-leading to better understanding of new materials, including high temperature super conductors, plastics, and thin films. Some 48 neutron beam stations will be set up in the ANS beam rooms and the neutron guide hall for neutron scattering and for fundamental and nuclear physics research. There also will be extensive facilities for materials irradiation, isotope production, and analytical chemistry. The R ampersand D program will focus on the four objectives: Address feasibility issues; provide analysis support; evaluate options for improvement in performance beyond minimum requirements; and provide prototype demonstrations for unique facilities. The remainder of this report presents (1) the process by which the R ampersand D activities are controlled and (2) a discussion of the individual tasks that have been identified for the R ampersand D program, including their justification, schedule and costs. The activities discussed in this report will be performed by Martin Marietta Energy Systems, Inc. (MMES) through the Oak Ridge National Laboratory (ORNL) and through subcontracts with industry, universities, and other national laboratories. It should be noted that in general a success path has been assumed for all tasks
Cadmium depletion impacts on hardening neutron spectrum for advanced fuel testing in ATR
International Nuclear Information System (INIS)
Chang, Gray S.
2011-01-01
For transmuting long-lived isotopes contained in spent nuclear fuel into shorter-lived fission products effectively is in a fast neutron spectrum reactor. In the absence of a fast spectrum test reactor in the United States of America (USA), initial irradiation testing of candidate fuels can be performed in a thermal test reactor that has been modified to produce a test region with a hardened neutron spectrum. A test region is achieved with a Cadmium (Cd) filter which can harden the neutron spectrum to a spectrum similar (although still somewhat softer) to that of the liquid metal fast breeder reactor (LMFBR). A fuel test loop with a Cd-filter has been installed within the East Flux Trap (EFT) of the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL). A detailed comparison analyses between the cadmium (Cd) filter hardened neutron spectrum in the ATR and the LMFBR fast neutron spectrum have been performed using MCWO. MCWO is a set of scripting tools that are used to couple the Monte Carlo transport code MCNP with the isotope depletion and buildup code ORIGEN-2.2. The MCWO-calculated results indicate that the Cd-filter can effectively flatten the Rim-Effect and reduce the linear heat rate (LHGR) to meet the advanced fuel testing project requirements at the beginning of irradiation (BOI). However, the filtering characteristics of Cd as a strong absorber quickly depletes over time, and the Cd-filter must be replaced for every two typical operating cycles within the EFT of the ATR. The designed Cd-filter can effectively depress the LHGR in experimental fuels and harden the neutron spectrum enough to adequately flatten the Rim-Effect in the test region. (author)
Stability and accuracy of 3D neutron transport simulations using the 2D/1D method in MPACT
International Nuclear Information System (INIS)
Collins, Benjamin; Stimpson, Shane; Kelley, Blake W.; Young, Mitchell T.H.; Kochunas, Brendan; Graham, Aaron; Larsen, Edward W.; Downar, Thomas; Godfrey, Andrew
2016-01-01
A consistent “2D/1D” neutron transport method is derived from the 3D Boltzmann transport equation, to calculate fuel-pin-resolved neutron fluxes for realistic full-core Pressurized Water Reactor (PWR) problems. The 2D/1D method employs the Method of Characteristics to discretize the radial variables and a lower order transport solution to discretize the axial variable. This paper describes the theory of the 2D/1D method and its implementation in the MPACT code, which has become the whole-core deterministic neutron transport solver for the Consortium for Advanced Simulations of Light Water Reactors (CASL) core simulator VERA-CS. Several applications have been performed on both leadership-class and industry-class computing clusters. Results are presented for whole-core solutions of the Watts Bar Nuclear Power Station Unit 1 and compared to both continuous-energy Monte Carlo results and plant data.
Stability and accuracy of 3D neutron transport simulations using the 2D/1D method in MPACT
Energy Technology Data Exchange (ETDEWEB)
Collins, Benjamin, E-mail: collinsbs@ornl.gov [Oak Ridge National Laboratory, One Bethel Valley Rd., Oak Ridge, TN 37831 (United States); Stimpson, Shane, E-mail: stimpsonsg@ornl.gov [Oak Ridge National Laboratory, One Bethel Valley Rd., Oak Ridge, TN 37831 (United States); Kelley, Blake W., E-mail: kelleybl@umich.edu [Department of Nuclear Engineering and Radiological Sciences, University of Michigan, Ann Arbor, MI 48109 (United States); Young, Mitchell T.H., E-mail: youngmit@umich.edu [Department of Nuclear Engineering and Radiological Sciences, University of Michigan, Ann Arbor, MI 48109 (United States); Kochunas, Brendan, E-mail: bkochuna@umich.edu [Department of Nuclear Engineering and Radiological Sciences, University of Michigan, Ann Arbor, MI 48109 (United States); Graham, Aaron, E-mail: aarograh@umich.edu [Department of Nuclear Engineering and Radiological Sciences, University of Michigan, Ann Arbor, MI 48109 (United States); Larsen, Edward W., E-mail: edlarsen@umich.edu [Department of Nuclear Engineering and Radiological Sciences, University of Michigan, Ann Arbor, MI 48109 (United States); Downar, Thomas, E-mail: downar@umich.edu [Department of Nuclear Engineering and Radiological Sciences, University of Michigan, Ann Arbor, MI 48109 (United States); Godfrey, Andrew, E-mail: godfreyat@ornl.gov [Oak Ridge National Laboratory, One Bethel Valley Rd., Oak Ridge, TN 37831 (United States)
2016-12-01
A consistent “2D/1D” neutron transport method is derived from the 3D Boltzmann transport equation, to calculate fuel-pin-resolved neutron fluxes for realistic full-core Pressurized Water Reactor (PWR) problems. The 2D/1D method employs the Method of Characteristics to discretize the radial variables and a lower order transport solution to discretize the axial variable. This paper describes the theory of the 2D/1D method and its implementation in the MPACT code, which has become the whole-core deterministic neutron transport solver for the Consortium for Advanced Simulations of Light Water Reactors (CASL) core simulator VERA-CS. Several applications have been performed on both leadership-class and industry-class computing clusters. Results are presented for whole-core solutions of the Watts Bar Nuclear Power Station Unit 1 and compared to both continuous-energy Monte Carlo results and plant data.
International Nuclear Information System (INIS)
Penttila, S.I.; Fitzsimmons, M.R.; Delheij, P.J.
1998-01-01
The authors describe work on the development of polarized gaseous 3 He cells, which are intended for use as neutron polarizers. Laser diode arrays polarize Rb vapor in a sample cell and the 3 He is polarized via collisions. They describe development and tests of such a system at LANSCE
Energy Technology Data Exchange (ETDEWEB)
Cook, J.C.; Barker, J.G.; Rowe, J.M.; Williams, R.E. [NIST Center for Neutron Research, National Institute of Standards and Technology, 100 Bureau Drive, Mail Stop 6100, Gaithersburg, MD 20899-6100 (United States); Gagnon, C. [Department of Materials Science and Engineering, University of Maryland, College Park, MD 20742 (United States); Lindstrom, R.M. [Scientist Emeritus, Chemical Sciences Division, National Institute of Standards and Technology, 100 Bureau Drive, Mail Stop 8395, Gaithersburg, MD 20899-8395 (United States); Ibberson, R.M.; Neumann, D.A. [NIST Center for Neutron Research, National Institute of Standards and Technology, 100 Bureau Drive, Mail Stop 6100, Gaithersburg, MD 20899-6100 (United States)
2015-08-21
The recent expansion of the National Institute of Standards and Technology (NIST) Center for Neutron Research facility has offered a rare opportunity to perform an accurate measurement of the cold neutron spectrum at the exit of a newly-installed neutron guide. Using a combination of a neutron time-of-flight measurement, a gold foil activation measurement, and Monte Carlo simulation of the neutron guide transmission, we obtain the most reliable experimental characterization of the Advanced Liquid Hydrogen Cold Neutron Source brightness to date. Time-of-flight measurements were performed at three distinct fuel burnup intervals, including one immediately following reactor startup. Prior to the latter measurement, the hydrogen was maintained in a liquefied state for an extended period in an attempt to observe an initial radiation-induced increase of the ortho (o)-hydrogen fraction. Since para (p)-hydrogen has a small scattering cross-section for neutron energies below 15 meV (neutron wavelengths greater than about 2.3 Å), changes in the o- p hydrogen ratio and in the void distribution in the boiling hydrogen influence the spectral distribution. The nature of such changes is simulated with a continuous-energy, Monte Carlo radiation-transport code using 20 K o and p hydrogen scattering kernels and an estimated hydrogen density distribution derived from an analysis of localized heat loads. A comparison of the transport calculations with the mean brightness function resulting from the three measurements suggests an overall o- p ratio of about 17.5(±1) % o- 82.5% p for neutron energies<15 meV, a significantly lower ortho concentration than previously assumed.
Cook, J. C.; Barker, J. G.; Rowe, J. M.; Williams, R. E.; Gagnon, C.; Lindstrom, R. M.; Ibberson, R. M.; Neumann, D. A.
2015-08-01
The recent expansion of the National Institute of Standards and Technology (NIST) Center for Neutron Research facility has offered a rare opportunity to perform an accurate measurement of the cold neutron spectrum at the exit of a newly-installed neutron guide. Using a combination of a neutron time-of-flight measurement, a gold foil activation measurement, and Monte Carlo simulation of the neutron guide transmission, we obtain the most reliable experimental characterization of the Advanced Liquid Hydrogen Cold Neutron Source brightness to date. Time-of-flight measurements were performed at three distinct fuel burnup intervals, including one immediately following reactor startup. Prior to the latter measurement, the hydrogen was maintained in a liquefied state for an extended period in an attempt to observe an initial radiation-induced increase of the ortho (o)-hydrogen fraction. Since para (p)-hydrogen has a small scattering cross-section for neutron energies below 15 meV (neutron wavelengths greater than about 2.3 Å), changes in the o- p hydrogen ratio and in the void distribution in the boiling hydrogen influence the spectral distribution. The nature of such changes is simulated with a continuous-energy, Monte Carlo radiation-transport code using 20 K o and p hydrogen scattering kernels and an estimated hydrogen density distribution derived from an analysis of localized heat loads. A comparison of the transport calculations with the mean brightness function resulting from the three measurements suggests an overall o- p ratio of about 17.5(±1) % o- 82.5% p for neutron energies<15 meV, a significantly lower ortho concentration than previously assumed.
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
Green Propulsion Technologies for Advanced Air Transports
Del Rosario, Ruben
2015-01-01
Air transportation is critical to U.S. and Global economic vitality. However, energy and climate issues challenge aviations ability to be sustainable in the long term. Aviation must dramatically reduce fuel use and related emissions. Energy costs to U.S. airlines nearly tripled between 1995 and 2011, and continue to be the highest percentage of operating costs. The NASA Advanced Air Transports Technology Project addresses the comprehensive challenge of enabling revolutionary energy efficiency improvements in subsonic transport aircraft combined with dramatic reductions in harmful emissions and perceived noise to facilitate sustained growth of the air transportation system. Advanced technologies and the development of unconventional aircraft systems offer the potential to achieve these improvements. The presentation will highlight the NASA vision of revolutionary systems and propulsion technologies needed to achieve these challenging goals. Specifically, the primary focus is on the N+3 generation; that is, vehicles that are three generations beyond the current state of the art, requiring mature technology solutions in the 2025-30 timeframe, which are envisioned as being powered by Hybrid Electric Propulsion Systems.
Fast neutron irradiation for locally advanced pancreatic cancer
International Nuclear Information System (INIS)
Smith, F.P.; Schein, P.S.; MacDonald, J.S.; Woolley, P.V.; Ornitz, R.; Rogers, C.
1981-01-01
Nineteen patients with locally advanced pancreatic cancer and one patient with islet cell cancer were treated with 1700-1500 neutron rad alone or in combination with 5-fluorouracil to exploit the theoretic advantages of higher linear energy of transfer, and lower oxygen enhancement ratio of neutrons. Only 5 of 14 (36%) obtained partial tumor regression. The median survival for all patients with pancreatic cancer was 6 months, which is less than that reported with 5-fluorouracil and conventional photon irradiation. Gastrointestinal toxicity was considerable; hemorhagic gastritis in five patients, colitis in two and esophagitis in one. One patient developed radiation myelitis. We therefore, caution any enthusiasm for this modality of therapy until clear evidence of a therapeutic advantage over photon therapy is demonstrated in controlled clinical trials
Preliminary design concepts for the advanced neutron source reactor systems
International Nuclear Information System (INIS)
Peretz, F.J.
1988-01-01
This paper describes the initial design work to develop the reactor systems hardware concepts for the advanced neutron source (ANS) reactor. This project has not yet entered the conceptual design phase; thus, design efforts are quite preliminary. This paper presents the collective work of members of the Oak Ridge National Laboratory, Martin Marietta Energy Systems, Inc., Engineering Division, and other participating organizations. The primary purpose of this effort is to show that the ANS reactor concept is realistic from a hardware standpoint and to show that project objectives can be met. It also serves to generate physical models for use in neutronic and thermal-hydraulic core design efforts and defines the constraints and objectives for the design. Finally, this effort will develop the criteria for use in the conceptual design of the reactor
Fast neutron irradiation for locally advanced pancreatic cancer
Energy Technology Data Exchange (ETDEWEB)
Smith, F.P. (Georgetown Univ. Medical Center, Washington, DC); Schein, P.S.; MacDonald, J.S.; Woolley, P.V.; Ornitz, R.; Rogers, C.
1981-11-01
Nineteen patients with locally advanced pancreatic cancer and one patient with islet cell cancer were treated with 1700-1500 neutron rad alone or in combination with 5-fluorouracil to exploit the theoretic advantages of higher linear energy of transfer, and lower oxygen enhancement ratio of neutrons. Only 5 of 14 (36%) obtained partial tumor regression. The median survival for all patients with pancreatic cancer was 6 months, which is less than that reported with 5-fluorouracil and conventional photon irradiation. Gastrointestinal toxicity was considerable; hemorhagic gastritis in five patients, colitis in two and esophagitis in one. One patient developed radiation myelitis. We therefore, caution any enthusiasm for this modality of therapy until clear evidence of a therapeutic advantage over photon therapy is demonstrated in controlled clinical trials.
Spatial domain decomposition for neutron transport problems
International Nuclear Information System (INIS)
Yavuz, M.; Larsen, E.W.
1989-01-01
A spatial Domain Decomposition method is proposed for modifying the Source Iteration (SI) and Diffusion Synthetic Acceleration (DSA) algorithms for solving discrete ordinates problems. The method, which consists of subdividing the spatial domain of the problem and performing the transport sweeps independently on each subdomain, has the advantage of being parallelizable because the calculations in each subdomain can be performed on separate processors. In this paper we describe the details of this spatial decomposition and study, by numerical experimentation, the effect of this decomposition on the SI and DSA algorithms. Our results show that the spatial decomposition has little effect on the convergence rates until the subdomains become optically thin (less than about a mean free path in thickness)
Symmetrized neutron transport equation and the fast Fourier transform method
International Nuclear Information System (INIS)
Sinh, N.Q.; Kisynski, J.; Mika, J.
1978-01-01
The differential equation obtained from the neutron transport equation by the application of the source iteration method in two-dimensional rectangular geometry is transformed into a symmetrized form with respect to one of the angular variables. The discretization of the symmetrized equation leads to finite difference equations based on the five-point scheme and solved by use of the fast Fourier transform method. Possible advantages of the approach are shown on test calculations
Deterministic methods to solve the integral transport equation in neutronic
International Nuclear Information System (INIS)
Warin, X.
1993-11-01
We present a synthesis of the methods used to solve the integral transport equation in neutronic. This formulation is above all used to compute solutions in 2D in heterogeneous assemblies. Three kinds of methods are described: - the collision probability method; - the interface current method; - the current coupling collision probability method. These methods don't seem to be the most effective in 3D. (author). 9 figs
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
Complex eigenvalues for neutron transport equation with quadratically anisotropic scattering
International Nuclear Information System (INIS)
Sjoestrand, N.G.
1981-01-01
Complex eigenvalues for the monoenergetic neutron transport equation in the buckling approximation have been calculated for various combinations of linearly and quadratically anisotropic scattering. The results are discussed in terms of the time-dependent case. Tables are given of complex bucklings for real decay constants and of complex decay constants for real bucklings. The results fit nicely into the pattern of real and purely imaginary eigenvalues obtained earlier. (author)
Sensitivity of neutron air transport to nitrogen cross section uncertainties
International Nuclear Information System (INIS)
Niiler, A.; Beverly, W.B.; Banks, N.E.
1975-01-01
The sensitivity of the transport of 14-MeV neutrons in sea level air to uncertainties in the ENDF/B-III values of the various Nitrogen cross sections has been calculated using the correlated sampling Monte Carlo neutron transport code SAMCEP. The source consisted of a 14.0- to 14.9-MeV band of isotropic neutrons and the fluences (0.5 to 15.0 MeV) were calculated at radii from 50 to 1500 metres. The maximum perturbations, assigned to the ENDF/B-III or base cross section set in the 6.0- to 14.5-MeV energy range were; (1) 2 percent to the total, (2) 10 percent to the total elastic, (3) 40 percent to the inelastic and absorption and (4) 20 percent to the first Legendre coefficient and 10 percent to the second Legendre coefficient of the elastic angular distribtuions. Transport calculations were carried out using various physically realistic sets of perturbed cross sections, bounded by evaluator-assigned uncertainties, as well as the base set. Results show that in some energy intervals at 1500 metres, the differential fluence level with a perturbed set differed by almost a factor of two from the differential fluence level with the base set. 5 figures
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
Neutron physics of a high converting advanced pressurized water reactor
International Nuclear Information System (INIS)
Berger, H.D.
1985-01-01
The neutron physics of an APWR are analysed by single pin-cell calculations as well as two-dimensional whole-reactor computations. The calculational methods of the two codes employed for this study, viz. the cell code SPEKTRA and the diffusion-burnup code DIBU, are presented in detail. The APWR-investigations carried out concentrate on the void coefficient characteristics of tight UO 2 /PuO 2 -lattices, control rod worths, burnup behaviour and spatial power distributions in APWR cores. The principal physics design differences between advanced pressurized water reactors and present-day PWRs are identified and discussed. (orig./HP) [de
Creep analysis of fuel plates for the Advanced Neutron Source
International Nuclear Information System (INIS)
Swinson, W.F.; Yahr, G.T.
1994-11-01
The reactor for the planned Advanced Neutron Source will use closely spaced arrays of fuel plates. The plates are thin and will have a core containing enriched uranium silicide fuel clad in aluminum. The heat load caused by the nuclear reactions within the fuel plates will be removed by flowing high-velocity heavy water through narrow channels between the plates. However, the plates will still be at elevated temperatures while in service, and the potential for excessive plate deformation because of creep must be considered. An analysis to include creep for deformation and stresses because of temperature over a given time span has been performed and is reported herein
Finite element analysis of advanced neutron source fuel plates
International Nuclear Information System (INIS)
Luttrell, C.R.
1995-08-01
The proposed design for the Advanced Neutron Source reactor core consists of closely spaced involute fuel plates. Coolant flows between the plates at high velocities. It is vital that adjacent plates do not come in contact and that the coolant channels between the plates remain open. Several scenarios that could result in problems with the fuel plates are studied. Finite element analyses are performed on fuel plates under pressure from the coolant flowing between the plates at a high velocity, under pressure because of a partial flow blockage in one of the channels, and with different temperature profiles
The advanced neutron source facility: Safety philosophy and studies
International Nuclear Information System (INIS)
Greene, S.R.; Harrington, R.M.
1988-01-01
The Advanced Neutron Source (ANS) is currently the only new civilian nuclear reactor facility proposed for construction in the United States. Even though the thermal power of this research-oriented reactor is a relatively low 300 MW, the design will undoubtedly receive intense scrutiny before construction is allowed to proceed. Safety studies are already under way to ensure that the maximum degree of safety in incorporated into the design and that the design is acceptable to the Department of Energy (DOE) and can meet the Nuclear Regulatory Commission regulations. This document discusses these safety studies
The Advanced Neutron Source liquid deuterium cold source
International Nuclear Information System (INIS)
Lucas, A.T.
1995-08-01
The Advanced Neutron Source will employ two cold sources to moderate neutrons to low energy (<10 meV). The cold neutrons produced are then passed through beam guides to various experiment stations. Each cold source moderator is a sphere of 410-mm internal diameter. The moderator material is liquid deuterium flowing at a rate of 1 kg/s and maintained at subcooled temperatures at all points of the circuit, to prevent boiling. Nuclear beat deposited within the liquid deuterium and its containment structure totals more than 30 kW. All of this heat is removed by the liquid deuterium, which raises its temperature by 5 K. The liquid prime mover is a cryogenic circulator that is situated in the return leg of the flow loop. This arrangement minimizes the heat added to the liquid between the heat exchanger and the moderator vessel, allowing the moderator to be operated at the minimum practical temperature. This report describes the latest thinking at the time of project termination. It also includes the status of various systems at that time and outlines anticipated directions in which the design would have progressed. In this regard, some detail differences between this report and official design documents reflect ideas that were not approved at the time of closure but are considered noteworthy
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)
Smith, L.A.; Gehin, J.C.; Worley, B.A.; Renier, J.P.
1994-01-01
The FOEHN critical experiments were analyzed to validate the use of multigroup cross sections in the design of the Advanced Neutron Source. Eleven critical configurations were evaluated using the KENO, DORT, and VENTURE neutronics codes. Eigenvalue and power density profiles were computed and show very good agreement with measured values
Neutron tolerance of advanced SiC-fiber/CVI-SiC composites
International Nuclear Information System (INIS)
Katoh, Y.; Kohyama, A.; Snead, L.L.; Hinoki, T.; Hasegawa, A.
2003-01-01
Fusion blankets employing a silicon carbide (SiC) fiber-reinforced SiC matrix composite (SiC/SiC composite) as the structural material provide attractive features represented by high cycle efficiency and extremely low induced radioactivity. Recent advancement in processing and utilization techniques and application studies in ceramic gas turbine and advanced transportation systems, SiC/SiC composites are steadily getting matured as industrial materials. Reference SiC/SiC composites for fusion structural applications have been produced by a forced-flow chemical vapor infiltration (FCVI) method using conventional and advanced near-stoichiometric SiC fibers and extensively evaluated primarily in Japan-US collaborative JUPITER program. In this work, effect of neutron irradiation at elevated temperatures on mechanical property of these composites is characterized. Unlike in conventional SiC/SiC composites, practically no property degradation was identified in advanced composites with a thin carbon interphase by a neutron fluence level of approximately 8dpa at 800C. (author)
Advanced Neutron Source (ANS) Project progress report, FY 1994
Energy Technology Data Exchange (ETDEWEB)
Campbell, J.H.; King-Jones, K.H. [eds.; Selby, D.L.; Harrington, R.M. [Oak Ridge National Lab., TN (United States); Thompson, P.B. [Martin Marietta Energy Systems, Inc., Oak Ridge, TN (United States). Central Engineering Services
1995-01-01
The President`s budget request for FY 1994 included a construction project for the Advanced Neutron Source (ANS). However, the budget that emerged from the Congress did not, and so activities during this reporting period were limited to continued research and development and to advanced conceptual design. A significant effort was devoted to a study, requested by the US Department of Energy (DOE) and led by Brookhaven National Laboratory, of the performance and cost impacts of reducing the uranium fuel enrichment below the baseline design value of 93%. The study also considered alternative core designs that might mitigate those impacts. The ANS Project proposed a modified core design, with three fuel elements instead of two, that would allow operation with only 50% enriched uranium and use existing fuel technology. The performance penalty would be 15--20% loss of thermal neutron flux; the flux would still just meet the minimum design requirement set by the user community. At the time of this writing, DOE has not established an enrichment level for ANS, but two advisory committees have recommended adopting the new core design, provided the minimum flux requirements are still met.
Advanced Neutron Source (ANS) Project progress report, FY 1994
International Nuclear Information System (INIS)
Campbell, J.H.; King-Jones, K.H.; Thompson, P.B.
1995-01-01
The President's budget request for FY 1994 included a construction project for the Advanced Neutron Source (ANS). However, the budget that emerged from the Congress did not, and so activities during this reporting period were limited to continued research and development and to advanced conceptual design. A significant effort was devoted to a study, requested by the US Department of Energy (DOE) and led by Brookhaven National Laboratory, of the performance and cost impacts of reducing the uranium fuel enrichment below the baseline design value of 93%. The study also considered alternative core designs that might mitigate those impacts. The ANS Project proposed a modified core design, with three fuel elements instead of two, that would allow operation with only 50% enriched uranium and use existing fuel technology. The performance penalty would be 15--20% loss of thermal neutron flux; the flux would still just meet the minimum design requirement set by the user community. At the time of this writing, DOE has not established an enrichment level for ANS, but two advisory committees have recommended adopting the new core design, provided the minimum flux requirements are still met
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)
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.)
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.
International Nuclear Information System (INIS)
Bussac, J.; Reuss, P.
1985-01-01
This book presents the main physical bases of neutron theory and nuclear reactor calculation. 1) Interactions of neutrons with matter and basic principles of neutron transport; 2) Neutron transport in homogeneous medium and the neutron field: kinetic behaviour, slowing-down, resonance absorption, diffusion equation, processing methods; 3) Theory of a reactor constituted with homogeneous zones: critical condition, kinetics, separation of variables, calculation and neutron balance of the fundamental mode, one-group and multigroup theories; 4) Study of heterogeneous cell lattices: fast fission factor, resonance absorption, thermal output factor, diffusion coefficient, computer codes; 5) Operation and control of reactors: perturbation theory, reactivity, fuel properties evolution, poisoning by fission products, calculation of a reactor and fuel management; 6) Study of some types of reactors: PWR and fast breeder reactors, the main reactor types of the present French program [fr
A three-dimensional neutron transport benchmark solution
International Nuclear Information System (INIS)
Ganapol, B.D.; Kornreich, D.E.
1993-01-01
For one-group neutron transport theory in one dimension, several powerful analytical techniques have been developed to solve the neutron transport equation, including Caseology, Wiener-Hopf factorization, and Fourier and Laplace transform methods. In addition, after a Fourier transform in the transverse plane and formulation of a pseudo problem, two-dimensional (2-D) and three-dimensional (3-D) problems can be solved using the techniques specifically developed for the one-dimensional (1-D) case. Numerical evaluation of the resulting expressions requiring an inversion in the transverse plane have been successful for 2-D problems but becomes exceedingly difficult in the 3-D case. In this paper, we show that by using the symmetry along the beam direction, a 2-D problem can be transformed into a 3-D problem in an infinite medium. The numerical solution to the 3-D problem is then demonstrated. Thus, a true 3-D transport benchmark solution can be obtained from a well-established numerical solution to a 2-D problem
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
An application of reactor noise techniques to neutron transport problems in a random medium
International Nuclear Information System (INIS)
Sahni, D.C.
1989-01-01
Neutron transport problems in a random medium are considered by defining a joint Markov process describing the fluctuations of one neutron population and the random changes in the medium. Backward Chapman-Kolmogorov equations are derived which yield an adjoint transport equation for the average neutron density. It is shown that this average density also satisfied the direct transport equation as given by the phenomenological model. (author)
MINARET: Towards a time-dependent neutron transport parallel solver
International Nuclear Information System (INIS)
Baudron, A.M.; Lautard, J.J.; Maday, Y.; Mula, O.
2013-01-01
We present the newly developed time-dependent 3D multigroup discrete ordinates neutron transport solver that has recently been implemented in the MINARET code. The solver is the support for a study about computing acceleration techniques that involve parallel architectures. In this work, we will focus on the parallelization of two of the variables involved in our equation: the angular directions and the time. This last variable has been parallelized by a (time) domain decomposition method called the para-real in time algorithm. (authors)
Criticality problems in energy dependent neutron transport theory
International Nuclear Information System (INIS)
Victory, H.D. Jr.
1979-01-01
The criticality problem is considered for energy dependent neutron transport in an isotropically scattering, homogeneous slab. Under a positivity assumption on the scattering kernel, an expression can be found relating the thickness of the slab to a parameter characterizing production by fission. This is accomplished by exploiting the Perron-Frobenius-Jentsch characterization of positive operators (i.e. those leaving invariant a normal, reproducing cone in a Banach space). It is pointed out that those techniques work for classes of multigroup problems were the Case singular eigenfunction approach is not as feasible as in the one-group theory, which is also analyzed
Reliability analysis of neutron transport simulation using Monte Carlo method
International Nuclear Information System (INIS)
Souza, Bismarck A. de; Borges, Jose C.
1995-01-01
This work presents a statistical and reliability analysis covering data obtained by computer simulation of neutron transport process, using the Monte Carlo method. A general description of the method and its applications is presented. Several simulations, corresponding to slowing down and shielding problems have been accomplished. The influence of the physical dimensions of the materials and of the sample size on the reliability level of results was investigated. The objective was to optimize the sample size, in order to obtain reliable results, optimizing computation time. (author). 5 refs, 8 figs
Structures of the fractional spaces generated by the difference neutron transport operator
International Nuclear Information System (INIS)
Ashyralyev, Allaberen; Taskin, Abdulgafur
2015-01-01
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
Property change of advanced tungsten alloys due to neutron irradiation
International Nuclear Information System (INIS)
Fukuda, Makoto; Hasegawa, Akira; Tanno, Takashi; Nogami, Shuhei; Kurishita, Hiroaki
2013-01-01
This study investigates the effect of neutron irradiation on the functional properties of pure tungsten (W) and advanced tungsten alloys (e.g., lanthanum (La)-doped W, potassium (K)-doped W, and ultra-fine-grained (UFG) W–TiC alloys) tested in the Japan Materials Testing Reactor (JMTR) or experimental fast reactor Joyo. The irradiation temperature and damage were in the range 804–1073 K and 0.15–0.47 dpa, respectively. TEM images of all samples after 0.42 dpa irradiation at 1023 K showed voids, black dots, and dislocation loops, indicating that similar damage structures were formed in pure W, La-doped W, K-doped W, and UFG W–0.5 wt% TiC. The electrical resistivity of all specimens increased following neutron irradiation. Nearly identical electrical resistivity and irradiation hardening were observed in pure W, La-doped W, and K-doped W. The electrical resistivity of UFG W–TiC was higher than that of other specimens before and after irradiation, which may be attributed to its ultra-fine-grain structure, as well as the presence of impurities introduced during the alloying process. Compared to the other specimens, the UFG W–TiC was more resistant to irradiation hardening
Hydrogen retention behavior of beryllides as advanced neutron multipliers
Directory of Open Access Journals (Sweden)
Y. Fujii
2016-12-01
Full Text Available Beryllium intermetallic compounds (beryllides are the most promising candidate materials for use as advanced neutron multipliers in future fusion reactors because of their low swelling and high stability at high temperatures. Recently, beryllium–titanium beryllide pebbles such as Be12Ti have been successfully fabricated using a novel granulation process. In this study, the fundamental aspects of the behavior of hydrogen isotopes in Be12Ti pebbles were investigated via thermal desorption spectroscopy and transmission electron microscopy. In addition, atomistic calculations using first principles electronic-structure methods were applied to determine the solution energy of hydrogen in Be12Ti. The results showed simpler and weaker hydrogen-trapping efficiency for Be12Ti than for pure Be.
Optimization of the Neutronics of the Advanced High Temperature Reactor
International Nuclear Information System (INIS)
Zakova, Jitka; Talamo, Alberto
2006-01-01
In these studies, we have investigated the neutronic and safety performance of the Advanced High Temperature Reactor (AHTR) for plutonium and uranium fuels and we extended the analysis to five different coolants. The AHTR is a graphite-moderated and molten salt-cooled high temperature reactor, which takes advantage of the TRISO particles technology for the fuel utilization. The conceptual design of the core, proposed at the Oak Ridge National Laboratory, aims to provide an alternative to helium as coolant of high-temperature reactors for industrial applications like hydrogen production. We evaluated the influence of the radial reflector on the criticality of the core for the uranium and plutonium fuels and we focused on the void coefficient of 5 different molten salts; since the safety of the reactor is enhanced also by the large and negative coefficient of temperature, we completed our investigation by observing the keff changes when the graphite temperature varies from 300 to 1800 K. (authors)
Transport modeling and advanced computer techniques
International Nuclear Information System (INIS)
Wiley, J.C.; Ross, D.W.; Miner, W.H. Jr.
1988-11-01
A workshop was held at the University of Texas in June 1988 to consider the current state of transport codes and whether improved user interfaces would make the codes more usable and accessible to the fusion community. Also considered was the possibility that a software standard could be devised to ease the exchange of routines between groups. It was noted that two of the major obstacles to exchanging routines now are the variety of geometrical representation and choices of units. While the workshop formulated no standards, it was generally agreed that good software engineering would aid in the exchange of routines, and that a continued exchange of ideas between groups would be worthwhile. It seems that before we begin to discuss software standards we should review the current state of computer technology --- both hardware and software to see what influence recent advances might have on our software goals. This is done in this paper
International Nuclear Information System (INIS)
Smith, L.A.; Gallmeier, F.X.; Gehin, J.C.
1995-05-01
The FOEHN critical experiment was analyzed to validate the use of multigroup cross sections and Oak Ridge National Laboratory neutronics computer codes in the design of the Advanced Neutron Source. The ANSL-V 99-group master cross section library was used for all the calculations. Three different critical configurations were evaluated using the multigroup KENO Monte Carlo transport code, the multigroup DORT discrete ordinates transport code, and the multigroup diffusion theory code VENTURE. The simple configuration consists of only the fuel and control elements with the heavy water reflector. The intermediate configuration includes boron endplates at the upper and lower edges of the fuel element. The complex configuration includes both the boron endplates and components in the reflector. Cross sections were processed using modules from the AMPX system. Both 99-group and 20-group cross sections were created and used in two-dimensional models of the FOEHN experiment. KENO calculations were performed using both 99-group and 20-group cross sections. The DORT and VENTURE calculations were performed using 20-group cross sections. Because the simple and intermediate configurations are azimuthally symmetric, these configurations can be explicitly modeled in R-Z geometry. Since the reflector components cannot be modeled explicitly using the current versions of these codes, three reflector component homogenization schemes were developed and evaluated for the complex configuration. Power density distributions were calculated with KENO using 99-group cross sections and with DORT and VENTURE using 20-group cross sections. The average differences between the measured values and the values calculated with the different computer codes range from 2.45 to 5.74%. The maximum differences between the measured and calculated thermal flux values for the simple and intermediate configurations are ∼ 13%, while the average differences are < 8%
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
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
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
Study of influence of transport performance of the neutron guide
International Nuclear Information System (INIS)
Li Xinxi; Wang Yan; Huang Chaoqiang; Chen Bo; Chen Liang
2009-01-01
For the sake of improving the performance of the neutron scattering instrument, usually we need use the neutron guide, it's very important to select the right type and optimizing of neutron guide. The papers calculate the focus neutron guide and the single channel neutron guide by numeric method. The results shows that the choice of neutron guide should consult the resolution requirement of neutron scattering instrument, and the length of the neutron guide should be optimized. The calculation results can be the theoretical reference for the design of neutron scattering instrument. (authors)
Advanced transport systems analysis, modeling, and evaluation of performances
Janić, Milan
2014-01-01
This book provides a systematic analysis, modeling and evaluation of the performance of advanced transport systems. It offers an innovative approach by presenting a multidimensional examination of the performance of advanced transport systems and transport modes, useful for both theoretical and practical purposes. Advanced transport systems for the twenty-first century are characterized by the superiority of one or several of their infrastructural, technical/technological, operational, economic, environmental, social, and policy performances as compared to their conventional counterparts. The advanced transport systems considered include: Bus Rapid Transit (BRT) and Personal Rapid Transit (PRT) systems in urban area(s), electric and fuel cell passenger cars, high speed tilting trains, High Speed Rail (HSR), Trans Rapid Maglev (TRM), Evacuated Tube Transport system (ETT), advanced commercial subsonic and Supersonic Transport Aircraft (STA), conventionally- and Liquid Hydrogen (LH2)-fuelled commercial air trans...
Advanced public transportation system deployment in the United States
1999-01-01
This report documents work performed under FTA's Advanced Public Transportation Systems (APTS) Program, a program structured to undertake research and development of innovative applications of advanced navigation, information, and communication techn...
Advanced public transportation systems : the state of the art
1991-03-01
This report documents one of the early initiatives of UMTA's Advanced Public Transportation Systems (APTS) Program, a program structured to undertake research and development of innovative applications of advanced navigation, information, and communi...
Benefits assessment of advanced public transportation system technologies, update 2000
This report was performed under the Federal Transit Administration's (FTA) Advanced Public Transportation Systems (APTS) Program. This program focuses on the development and demonstration of innovative advanced navigation, information and communicati...
Energy Technology Data Exchange (ETDEWEB)
Archambault, Brian C. [Sagamore Adams Laboratories, Lafayette, IN (United States); Webster, Jeffrey A.; Grimes, Thomas F.; Fischer, Kevin F.; Hagen, Alex R. [School of Nuclear Engineering, Purdue University, 400 Central Avenue, West Lafayette, IN 47907 (United States); Taleyakhan, Rusi P., E-mail: rusi@purdue.edu [Sagamore Adams Laboratories, Lafayette, IN (United States); School of Nuclear Engineering, Purdue University, 400 Central Avenue, West Lafayette, IN 47907 (United States)
2015-06-01
Advancements in the development of a direction and position sensing fast neutron detector which utilizes the directional acoustic tensioned metastable fluid detector (D-ATMFD) are described. The resulting D-ATMFD sensor is capable of determining the direction of neutron radiation with a single compact detector versus use of arrays of detectors in conventional directional systems. Directional neutron detection and source positioning offer enhanced detection speeds in comparison to traditional proximity searching; including enabling determination of the neutron source shape, size, and strength in near real time. This paper discusses advancements that provide the accuracy and precision of ascertaining directionality and source localization information utilizing enhanced signal processing-cum-signal analysis, refined computational algorithms, and on-demand enlargement capability of the detector sensitive volume. These advancements were accomplished utilizing experimentation and theoretical modeling. Benchmarking and qualifications studies were successfully conducted with random and fission based special nuclear material (SNM) neutron sources ({sup 239}Pu–Be and {sup 252}Cf). These results of assessments have indicated that the D-ATMFD compares well in technical performance with banks of competing directional fast neutron detector technologies under development worldwide, but it does so with a single detector unit, an unlimited field of view, and at a significant reduction in both cost and size while remaining completely blind to common background (e.g., beta-gamma) radiation. Rapid and direct SNM neutron source imaging with two D-ATMFD sensors was experimentally demonstrated, and furthermore, validated via multidimensional nuclear particle transport simulations utilizing MCNP-PoliMi. Characterization of a scaled D-ATMFD based radiation portal monitor (RPM) as a cost-effective and efficient {sup 3}He sensor replacement was performed utilizing MCNP-PoliMi simulations
International Nuclear Information System (INIS)
2001-05-01
An OECD Workshop on Advanced Thermal-Hydraulic and Neutronic Codes Applications was held from 10 to 13 April 2000, in Barcelona, Spain, sponsored by the Committee on the Safety of Nuclear Installations (CSNI) of the OECD Nuclear Energy Agency (NEA). It was organised in collaboration with the Spanish Nuclear Safety Council (CSN) and hosted by CSN and the Polytechnic University of Catalonia (UPC) in collaboration with the Spanish Electricity Association (UNESA). The objectives of the Workshop were to review the developments since the previous CSNI Workshop held in Annapolis [NEA/CSNI/ R(97)4; NUREG/CP-0159], to analyse the present status of maturity and remnant needs of thermal-hydraulic (TH) and neutronic system codes and methods, and finally to evaluate the role of these tools in the evolving regulatory environment. The Technical Sessions and Discussion Sessions covered the following topics: - Regulatory requirements for Best-Estimate (BE) code assessment; - Application of TH and neutronic codes for current safety issues; - Uncertainty analysis; - Needs for integral plant transient and accident analysis; - Simulators and fast running codes; - Advances in next generation TH and neutronic codes; - Future trends in physical modeling; - Long term plans for development of advanced codes. The focus of the Workshop was on system codes. An incursion was made, however, in the new field of applying Computational Fluid Dynamic (CFD) codes to nuclear safety analysis. As a general conclusion, the Barcelona Workshop can be considered representative of the progress towards the targets marked at Annapolis almost four years ago. The Annapolis Workshop had identified areas where further development and specific improvements were needed, among them: multi-field models, transport of interfacial area, 2D and 3D thermal-hydraulics, 3-D neutronics consistent with level of details of thermal-hydraulics. Recommendations issued at Annapolis included: developing small pilot/test codes for
International Nuclear Information System (INIS)
2001-01-01
An OECD Workshop on Advanced Thermal-Hydraulic and Neutronic Codes Applications was held from 10 to 13 April 2000, in Barcelona, Spain, sponsored by the Committee on the Safety of Nuclear Installations (CSNI) of the OECD Nuclear Energy Agency (NEA). It was organised in collaboration with the Spanish Nuclear Safety Council (CSN) and hosted by CSN and the Polytechnic University of Catalonia (UPC) in collaboration with the Spanish Electricity Association (UNESA). The objectives of the Workshop were to review the developments since the previous CSNI Workshop held in Annapolis [NEA/CSNI/ R(97)4; NUREG/CP-0159], to analyse the present status of maturity and remnant needs of thermal-hydraulic (TH) and neutronic system codes and methods, and finally to evaluate the role of these tools in the evolving regulatory environment. The Technical Sessions and Discussion Sessions covered the following topics: - Regulatory requirements for Best-Estimate (BE) code assessment; - Application of TH and neutronic codes for current safety issues; - Uncertainty analysis; - Needs for integral plant transient and accident analysis; - Simulators and fast running codes; - Advances in next generation TH and neutronic codes; - Future trends in physical modeling; - Long term plans for development of advanced codes. The focus of the Workshop was on system codes. An incursion was made, however, in the new field of applying Computational Fluid Dynamic (CFD) codes to nuclear safety analysis. As a general conclusion, the Barcelona Workshop can be considered representative of the progress towards the targets marked at Annapolis almost four years ago. The Annapolis Workshop had identified areas where further development and specific improvements were needed, among them: multi-field models, transport of interfacial area, 2D and 3D thermal-hydraulics, 3-D neutronics consistent with level of details of thermal-hydraulics. Recommendations issued at Annapolis included: developing small pilot/test codes for
Transport calculations for a 14.8 MeV neutron beam in a water phantom
International Nuclear Information System (INIS)
Goetsch, S.J.
1981-01-01
A coupled neutron/photon Monte Carlo radiation transport code (MORSE-CG) has been used to calculate neutron and photon doses in a water phantom irradiated by 14.8 MeV neutrons from the Gas Target Neutron Source. The source-collimator-phantom geometry was carefully simulated. Results of calculations utilizing two different statistical estimators (next-collision and track-length) are presented
Simulations of neutron transport at low energy: a comparison between GEANT and MCNP.
Colonna, N; Altieri, S
2002-06-01
The use of the simulation tool GEANT for neutron transport at energies below 20 MeV is discussed, in particular with regard to shielding and dose calculations. The reliability of the GEANT/MICAP package for neutron transport in a wide energy range has been verified by comparing the results of simulations performed with this package in a wide energy range with the prediction of MCNP-4B, a code commonly used for neutron transport at low energy. A reasonable agreement between the results of the two codes is found for the neutron flux through a slab of material (iron and ordinary concrete), as well as for the dose released in soft tissue by neutrons. These results justify the use of the GEANT/MICAP code for neutron transport in a wide range of applications, including health physics problems.
International Nuclear Information System (INIS)
Hoogenboom, J.E.
1980-01-01
1 - Description of problem or function: FOCUS enables the calculation of any quantity related to neutron transport in reactor or shielding problems, but was especially designed to calculate differential quantities, such as point values at one or more of the space, energy, direction and time variables of quantities like neutron flux, detector response, reaction rate, etc. or averages of such quantities over a small volume of the phase space. Different types of problems can be treated: systems with a fixed neutron source which may be a mono-directional source located out- side the system, and Eigen function problems in which the neutron source distribution is given by the (unknown) fundamental mode Eigen function distribution. Using Monte Carlo methods complex 3- dimensional geometries and detailed cross section information can be treated. Cross section data are derived from ENDF/B, with anisotropic scattering and discrete or continuous inelastic scattering taken into account. Energy is treated as a continuous variable and time dependence may also be included. 2 - Method of solution: A transformed form of the adjoint Boltzmann equation in integral representation is solved for the space, energy, direction and time variables by Monte Carlo methods. Adjoint particles are defined with properties in some respects contrary to those of neutrons. Adjoint particle histories are constructed from which estimates are obtained of the desired quantity. Adjoint cross sections are defined with which the nuclide and reaction type are selected in a collision. The energy after a collision is selected from adjoint energy distributions calculated together with the adjoint cross sections in advance of the actual Monte Carlo calculation. For multiplying systems successive generations of adjoint particles are obtained which will die out for subcritical systems with a fixed neutron source and will be kept approximately stationary for Eigen function problems. Completely arbitrary problems can
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)
MCNP speed advances for boron neutron capture therapy
International Nuclear Information System (INIS)
Goorley, J.T.; McKinney, G.; Adams, K.; Estes, G.
1998-04-01
The Boron Neutron Capture Therapy (BNCT) treatment planning process of the Beth Israel Deaconess Medical Center-M.I.T team relies on MCNP to determine dose rates in the subject's head for various beam orientations. In this time consuming computational process, four or five potential beams are investigated. Of these, one or two final beams are selected and thoroughly evaluated. Recent advances greatly decreased the time needed to do these MCNP calculations. Two modifications to the new MCNP4B source code, lattice tally and tracking enhancements, reduced the wall-clock run times of a typical one million source neutrons run to one hour twenty five minutes on a 200 MHz Pentium Pro computer running Linux and using the GNU FORTRAN compiler. Previously these jobs used a special version of MCNP4AB created by Everett Redmond, which completed in two hours two minutes. In addition to this 30% speedup, the MCNP4B version was adapted for use with Parallel Virtual Machine (PVM) on personal computers running the Linux operating system. MCNP, using PVM, can be run on multiple computers simultaneously, offering a factor of speedup roughly the same as the number of computers used. With two 200 MHz Pentium Pro machines, the run time was reduced to forty five minutes, a 1.9 factor of improvement over the single Linux computer. While the time of a single run was greatly reduced, the advantages associated with PVM derive from using computational power not already used. Four possible beams, currently requiring four separate runs, could be run faster when each is individually run on a single machine under Windows NT, rather than using Linux and PVM to run one after another with each multiprocessed across four computers. It would be advantageous, however, to use PVM to distribute the final two beam orientations over four computers
Advanced Neutron Radiographic Equipment within the US Army
2014-01-29
The first of the two new neutron generators developed for ARDEC was built and delivered by Phoenix Nuclear Labs ( PNL ) located in Madison, WI...PRELIMINARY NEUTRON RADIOGRAPHS AND RESULTS During the hardware and software testing of the PNL neutron generator several test shots were
Recent advances in fast neutron radiography for cargo inspection
International Nuclear Information System (INIS)
Sowerby, B.D.; Tickner, J.R.
2007-01-01
Fast neutron radiography techniques are attractive for screening cargo for contraband such as narcotics and explosives. Neutrons have the required penetration, they interact with matter in a manner complementary to X-rays and they can be used to determine elemental composition. Compared to neutron interrogation techniques that measure secondary radiation (neutron or gamma-rays), neutron radiography systems are much more efficient and rapid and they are much more amenable to imaging. However, for neutron techniques to be successfully applied to cargo screening, they must demonstrate significant advantages over well-established X-ray techniques. This paper reviews recent developments and applications of fast neutron radiography for cargo inspection. These developments include a fast neutron and gamma-ray radiography system that utilizes a 14 MeV neutron generator as well as fast neutron resonance radiography systems that use variable energy quasi-monoenergetic neutrons and pulsed broad energy neutron beams. These systems will be discussed and compared with particular emphasis on user requirements, sources, detector systems, imaging ability and performance
The transport of neutrons and gamma-rays in the air
International Nuclear Information System (INIS)
Adamski, J.
1980-01-01
The transport of neutrons and gamma rays in the infinite homogeneous air has been investigated. For the calculations has been used the Multigroup One Dimensional Discrete Ordinates Transport Code ANISN-W. The calculations have been performed for three types of neutron sources. The neutrons and gamma ray doses in the air have been analyzed, and comparison to the other authors' results has been given. (author)
Advanced Hydrogen Transport Membrane for Coal Gasification
Energy Technology Data Exchange (ETDEWEB)
Schwartz, Joseph [Praxair, Inc., Tonawanda, NY (United States); Porter, Jason [Colorado School of Mines, Golden, CO (United States); Patki, Neil [Colorado School of Mines, Golden, CO (United States); Kelley, Madison [Colorado School of Mines, Golden, CO (United States); Stanislowski, Josh [Univ. of North Dakota, Grand Forks, ND (United States); Tolbert, Scott [Univ. of North Dakota, Grand Forks, ND (United States); Way, J. Douglas [Colorado School of Mines, Golden, CO (United States); Makuch, David [Praxair, Inc., Tonawanda, NY (United States)
2015-12-23
A pilot-scale hydrogen transport membrane (HTM) separator was built that incorporated 98 membranes that were each 24 inches long. This separator used an advanced design to minimize the impact of concentration polarization and separated over 1000 scfh of hydrogen from a hydrogen-nitrogen feed of 5000 scfh that contained 30% hydrogen. This mixture was chosen because it was representative of the hydrogen concentration expected in coal gasification. When tested with an operating gasifier, the hydrogen concentration was lower and contaminants in the syngas adversely impacted membrane performance. All 98 membranes survived the test, but flux was lower than expected. Improved ceramic substrates were produced that have small surface pores to enable membrane production and large pores in the bulk of the substrate to allow high flux. Pd-Au was chosen as the membrane alloy because of its resistance to sulfur contamination and good flux. Processes were developed to produce a large quantity of long membranes for use in the demonstration test.
Parallel computing for homogeneous diffusion and transport equations in neutronics
International Nuclear Information System (INIS)
Pinchedez, K.
1999-06-01
Parallel computing meets the ever-increasing requirements for neutronic computer code speed and accuracy. In this work, two different approaches have been considered. We first parallelized the sequential algorithm used by the neutronics code CRONOS developed at the French Atomic Energy Commission. The algorithm computes the dominant eigenvalue associated with PN simplified transport equations by a mixed finite element method. Several parallel algorithms have been developed on distributed memory machines. The performances of the parallel algorithms have been studied experimentally by implementation on a T3D Cray and theoretically by complexity models. A comparison of various parallel algorithms has confirmed the chosen implementations. We next applied a domain sub-division technique to the two-group diffusion Eigen problem. In the modal synthesis-based method, the global spectrum is determined from the partial spectra associated with sub-domains. Then the Eigen problem is expanded on a family composed, on the one hand, from eigenfunctions associated with the sub-domains and, on the other hand, from functions corresponding to the contribution from the interface between the sub-domains. For a 2-D homogeneous core, this modal method has been validated and its accuracy has been measured. (author)
Finite element based composite solution for neutron transport problems
International Nuclear Information System (INIS)
Mirza, A.N.; Mirza, N.M.
1995-01-01
A finite element treatment for solving neutron transport problems is presented. The employs region-wise discontinuous finite elements for the spatial representation of the neutron angular flux, while spherical harmonics are used for directional dependence. Composite solutions has been obtained by using different orders of angular approximations in different parts of a system. The method has been successfully implemented for one dimensional slab and two dimensional rectangular geometry problems. An overall reduction in the number of nodal coefficients (more than 60% in some cases as compared to conventional schemes) has been achieved without loss of accuracy with better utilization of computational resources. The method also provides an efficient way of handling physically difficult situations such as treatment of voids in duct problems and sharply changing angular flux. It is observed that a great wealth of information about the spatial and directional dependence of the angular flux is obtained much more quickly as compared to Monte Carlo method, where most of the information in restricted to the locality of immediate interest. (author)
Nodal methods for problems in fluid mechanics and neutron transport
International Nuclear Information System (INIS)
Azmy, Y.Y.
1985-01-01
A new high-accuracy, coarse-mesh, nodal integral approach is developed for the efficient numerical solution of linear partial differential equations. It is shown that various special cases of this general nodal integral approach correspond to several high efficiency nodal methods developed recently for the numerical solution of neutron diffusion and neutron transport problems. The new approach is extended to the nonlinear Navier-Stokes equations of fluid mechanics; its extension to these equations leads to a new computational method, the nodal integral method which is implemented for the numerical solution of these equations. Application to several test problems demonstrates the superior computational efficiency of this new method over previously developed methods. The solutions obtained for several driven cavity problems are compared with the available experimental data and are shown to be in very good agreement with experiment. Additional comparisons also show that the coarse-mesh, nodal integral method results agree very well with the results of definitive ultra-fine-mesh, finite-difference calculations for the driven cavity problem up to fairly high Reynolds numbers
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
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)
Approximate solution to neutron transport equation with linear anisotropic scattering
International Nuclear Information System (INIS)
Coppa, G.; Ravetto, P.; Sumini, M.
1983-01-01
A method to obtain an approximate solution to the transport equation, when both sources and collisions show a linearly anisotropic behavior, is outlined and the possible implications for numerical calculations in applied neutronics as well as shielding evaluations are investigated. The form of the differential system of equations taken by the method is quite handy and looks simpler and more manageable than any other today available technique. To go deeper into the efficiency of the method, some typical calculations concerning critical dimension of multiplying systems are then performed and the results are compared with the ones coming from the classical Ssub(N) approximations. The outcome of such calculations leads us to think of interesting developments of the method which could be quite useful in alternative to other today widespread approximate procedures, for any geometry, but especially for curved ones. (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)
Massively parallel performance of neutron transport response matrix algorithms
International Nuclear Information System (INIS)
Hanebutte, U.R.; Lewis, E.E.
1993-01-01
Massively parallel red/black response matrix algorithms for the solution of within-group neutron transport problems are implemented on the Connection Machines-2, 200 and 5. The response matrices are dericed from the diamond-differences and linear-linear nodal discrete ordinate and variational nodal P 3 approximations. The unaccelerated performance of the iterative procedure is examined relative to the maximum rated performances of the machines. The effects of processor partitions size, of virtual processor ratio and of problems size are examined in detail. For the red/black algorithm, the ratio of inter-node communication to computing times is found to be quite small, normally of the order of ten percent or less. Performance increases with problems size and with virtual processor ratio, within the memeory per physical processor limitation. Algorithm adaptation to courser grain machines is straight-forward, with total computing time being virtually inversely proportional to the number of physical processors. (orig.)
On the Solution of the Neutron Transport Equation
Energy Technology Data Exchange (ETDEWEB)
Depken, S
1962-12-15
The neutron transport equation has occupied the attention of many authors since Placzek, Wick and others made their first attempts to solve it, Even in the simple case of energy independent cross-sections, and disregarding the motion of the scattering nucleons, it is difficult to find a solution in an analytical form which is easily surveyable and fitted for numerical calculations. In Part I of this paper some new viewpoints will be introduced which enable the solution to be presented in its simplest possible form. Part II is devoted to an investigation of some functions introduced in Part I. In Part III the results are applied to the case of large energy lethargy, and the validity of derived formulas is discussed.
Parallel processing of neutron transport in fuel assembly calculation
International Nuclear Information System (INIS)
Song, Jae Seung
1992-02-01
Group constants, which are used for reactor analyses by nodal method, are generated by fuel assembly calculations based on the neutron transport theory, since one or a quarter of the fuel assembly corresponds to a unit mesh in the current nodal calculation. The group constant calculation for a fuel assembly is performed through spectrum calculations, a two-dimensional fuel assembly calculation, and depletion calculations. The purpose of this study is to develop a parallel algorithm to be used in a parallel processor for the fuel assembly calculation and the depletion calculations of the group constant generation. A serial program, which solves the neutron integral transport equation using the transmission probability method and the linear depletion equation, was prepared and verified by a benchmark calculation. Small changes from the serial program was enough to parallelize the depletion calculation which has inherent parallel characteristics. In the fuel assembly calculation, however, efficient parallelization is not simple and easy because of the many coupling parameters in the calculation and data communications among CPU's. In this study, the group distribution method is introduced for the parallel processing of the fuel assembly calculation to minimize the data communications. The parallel processing was performed on Quadputer with 4 CPU's operating in NURAD Lab. at KAIST. Efficiencies of 54.3 % and 78.0 % were obtained in the fuel assembly calculation and depletion calculation, respectively, which lead to the overall speedup of about 2.5. As a result, it is concluded that the computing time consumed for the group constant generation can be easily reduced by parallel processing on the parallel computer with small size CPU's
Conceptual design of a high-intensity positron source for the Advanced Neutron Source
International Nuclear Information System (INIS)
Hulett, L.D.; Eberle, C.C.
1994-12-01
The Advanced Neutron Source (ANS) is a planned new basic and applied research facility based on a powerful steady-state research reactor that provides neutrons for measurements and experiments in the fields of materials science and engineering, biology, chemistry, materials analysis, and nuclear science. The useful neutron flux will be at least five times more than is available in the world's best existing reactor facility. Construction of the ANS provides a unique opportunity to build a positron spectroscopy facility (PSF) with very-high-intensity beams based on the radioactive decay of a positron-generating isotope. The estimated maximum beam current is 1000 to 5000 times higher than that available at the world's best existing positron research facility. Such an improvement in beam capability, coupled with complementary detectors, will reduce experiment durations from months to less than one hour while simultaneously improving output resolution. This facility will remove the existing barriers to the routine use of positron-based analytical techniques and will be a giant step toward realization of the full potential of the application of positron spectroscopy to materials science. The ANS PSF is based on a batch cycle process using 64 Cu isotope as the positron emitter and represents the status of the design at the end of last year. Recent work not included in this report, has led to a proposal for placing the laboratory space for the positron experiments outside the ANS containment; however, the design of the positron source is not changed by that relocation. Hydraulic and pneumatic flight tubes transport the source material between the reactor and the positron source where the beam is generated and conditioned. The beam is then transported through a beam pipe to one of several available detectors. The design presented here includes all systems necessary to support the positron source, but the beam pipe and detectors have not been addressed yet
A spin-transport system for a longitudinally polarized epithermal neutron beam
International Nuclear Information System (INIS)
Crawford, B.E.; Bowman, J.D.; Penttilae, S.I.; Roberson, N.R.
2001-01-01
The TRIPLE (Time Reversal and Parity at Low Energies) collaboration uses a polarized epithermal neutron beam and a capture γ-ray detector to study parity violation in neutron-nucleus reactions. In order to preserve the spin polarization of the neutrons as they travel the 60-m path to the target, the beam pipes are wrapped with wire to produce a solenoidal magnetic field of about 10 G along the beam direction. The flanges and bellows between sections of the beam pipe cause gaps in the windings which in turn produce radial fields that can depolarize the neutron spins. A computer code has been developed that numerically evaluates the effect of these gaps on the polarization. A measurement of the neutron depolarization for neutrons in the actual spin-transport system agrees with a calculation of the neutron depolarization for the TRIPLE system. Features that will aid in designing similar spin-transport systems are discussed
Energy Technology Data Exchange (ETDEWEB)
Suzuki, Jun-ichi [Center for Neutron Science, Tokai Research Establishment, Japan Atomic Energy Research Institute, Tokai, Ibaraki (Japan); Itoh, Shinichi [Neutron Science Laboratory, High Energy Accelerator Research Organization, Tsukuba, Ibaraki (JP)
2001-03-01
The fifteenth meeting of the International Collaboration on Advanced Neutron Sources (ICANS-XV) was held at Epocal Tsukuba, International Congress Center on 6-9 November 2000. It was hosted by Japan Atomic Energy Research Institute (JAERI) and High Energy Accelerator Research Organization (KEK). This meeting focused on 'Neutron Sources toward the 21st Century' and research activities related to targets and moderators, neutron scattering instruments and accelerators were presented. The 151 of the presented papers are indexed individually. (J.P.N.)
Energy Technology Data Exchange (ETDEWEB)
Suzuki, Jun-ichi [Center for Neutron Science, Tokai Research Establishment, Japan Atomic Energy Research Institute, Tokai, Ibaraki (Japan); Itoh, Shinichi [Neutron Science Laboratory, High Energy Accelerator Research Organization, Tsukuba, Ibaraki (JP)] (eds.)
2001-03-01
The fifteenth meeting of the International Collaboration on Advanced Neutron Sources (ICANS-XV) was held at Epocal Tsukuba, International Congress Center on 6-9 November 2000. It was hosted by Japan Atomic Energy Research Institute (JAERI) and High Energy Accelerator Research Organization (KEK). This meeting focused on 'Neutron Sources toward the 21st Century' and research activities related to targets and moderators, neutron scattering instruments and accelerators were presented. The 151 of the presented papers are indexed individually. (J.P.N.)
Transport calculation of medium-energy protons and neutrons by Monte Carlo method
International Nuclear Information System (INIS)
Ban, Syuuichi; Hirayama, Hideo; Katoh, Kazuaki.
1978-09-01
A Monte Carlo transport code, ARIES, has been developed for protons and neutrons at medium energy (25 -- 500 MeV). Nuclear data provided by R.G. Alsmiller, Jr. were used for the calculation. To simulate the cascade development in the medium, each generation was represented by a single weighted particle and an average number of emitted particles was used as the weight. Neutron fluxes were stored by the collisions density method. The cutoff energy was set to 25 MeV. Neutrons below the cutoff were stored to be used as the source for the low energy neutron transport calculation upon the discrete ordinates method. Then transport calculations were performed for both low energy neutrons (thermal -- 25 MeV) and secondary gamma-rays. Energy spectra of emitted neutrons were calculated and compared with those of published experimental and calculated results. The agreement was good for the incident particles of energy between 100 and 500 MeV. (author)
International Nuclear Information System (INIS)
Ford, W.E. III; Arwood, J.W.; Greene, N.M.; Moses, D.L.; Petrie, L.M.; Primm, R.T. III; Slater, C.O.; Westfall, R.M.; Wright, R.Q.
1990-09-01
Pseudo-problem-independent, multigroup cross-section libraries were generated to support Advanced Neutron Source (ANS) Reactor design studies. The ANS is a proposed reactor which would be fueled with highly enriched uranium and cooled with heavy water. The libraries, designated ANSL-V (Advanced Neutron Source Cross Section Libraries based on ENDF/B-V), are data bases in AMPX master format for subsequent generation of problem-dependent cross-sections for use with codes such as KENO, ANISN, XSDRNPM, VENTURE, DOT, DORT, TORT, and MORSE. Included in ANSL-V are 99-group and 39-group neutron, 39-neutron-group 44-gamma-ray-group secondary gamma-ray production (SGRP), 44-group gamma-ray interaction (GRI), and coupled, 39-neutron group 44-gamma-ray group (CNG) cross-section libraries. The neutron and SGRP libraries were generated primarily from ENDF/B-V data; the GRI library was generated from DLC-99/HUGO data, which is recognized as the ENDF/B-V photon interaction data. Modules from the AMPX and NJOY systems were used to process the multigroup data. Validity of selected data from the fine- and broad-group neutron libraries was satisfactorily tested in performance parameter calculations
Advanced neutron source reactor probabilistic flow blockage assessment
International Nuclear Information System (INIS)
Ramsey, C.T.
1995-08-01
The Phase I Level I Probabilistic Risk Assessment (PRA) of the conceptual design of the Advanced Neutron Source (ANS) Reactor identified core flow blockage as the most likely internal event leading to fuel damage. The flow blockage event frequency used in the original ANS PRA was based primarily on the flow blockage work done for the High Flux Isotope Reactor (HFIR) PRA. This report examines potential flow blockage scenarios and calculates an estimate of the likelihood of debris-induced fuel damage. The bulk of the report is based specifically on the conceptual design of ANS with a 93%-enriched, two-element core; insights to the impact of the proposed three-element core are examined in Sect. 5. In addition to providing a probability (uncertainty) distribution for the likelihood of core flow blockage, this ongoing effort will serve to indicate potential areas of concern to be focused on in the preliminary design for elimination or mitigation. It will also serve as a loose-parts management tool
Assessment of the roles of the Advanced Neutron Source Operators
International Nuclear Information System (INIS)
Hill, W.E.; Houser, M.M.; Knee, H.E.; Spelt, P.F.
1995-03-01
The Advanced Neutron Source (ANS) is unique in the extent to which human factors engineering (HFE) principles are being applied at the conceptual design stage. initial HFE accomplishments include the development of an ANS HFE program plan, operating philosophy, and functional analysis. In FY 1994, HFE activities focused on the role of the ANS control room reactor operator (RO). An operator-centered control room model was used in conjunction with information gathered from existing ANS system design descriptions and other literature to define a list of RO responsibilities. From this list, a survey instrument was developed and administered to ANS design engineers, operations management personnel at Oak Ridge National Laboratory's High Flux Isotope Reactor (HFIR), and HFIR ROs to detail the nature of the RO position. Initial results indicated that the RO will function as a high-level system supervisor with considerable monitoring, verification, and communication responsibilities. The relatively high level of control automation has resulted in a reshaping of the RO's traditional safety and investment protection roles
Fabrication development for the Advanced Neutron Source Reactor
International Nuclear Information System (INIS)
Pace, B.W.; Copeland, G.L.
1995-08-01
This report presents the fuel fabrication development for the Advanced Neutron Source (ANS) reactor. The fuel element is similar to that successfully fabricated and used in the High Flux Isotope Reactor (HFIR) for many years, but there are two significant differences that require some development. The fuel compound is U 3 Si 2 rather than U 3 O 8 , and the fuel is graded in the axial as well as the radial direction. Both of these changes can be accomplished with a straightforward extension of the HFIR technology. The ANS also requires some improvements in inspection technology and somewhat more stringent acceptance criteria. Early indications were that the fuel fabrication and inspection technology would produce a reactor core meeting the requirements of the ANS for the low volume fraction loadings needed for the highly enriched uranium design (up to 1.7 Mg U/m 3 ). Near the end of the development work, higher volume fractions were fabricated that would be required for a lower- enrichment uranium core. Again, results look encouraging for loadings up to ∼3.5 Mg U/m 3 ; however, much less evaluation was done for the higher loadings
Advanced Neutron Source Reactor zoning, shielding, and radiological optimization guide
International Nuclear Information System (INIS)
Westbrook, J.L.; DeVore, J.R.
1995-08-01
In the design of major nuclear facilities, it is important to protect both humans and equipment excessive radiation dose. Past experience has shown that it is very effective to apply dose reduction principles early in the design of a nuclear facility both to specific design features and to the manner of operation of the facility, where they can aid in making the facility more efficient and cost-effective. Since the appropriate choice of radiological controls and practices varies according to the case, each area of the facility must be analyzed for its radiological impact, both by itself and in interactions with other areas. For the Advanced Neutron Source (ANS) project, a large relational database will be used to collect facility information by system and relate it to areas. The database will also hold the facility dose and shielding information as it is produced during the design process. This report details how the ANS zoning scheme was established and how the calculation of doses and shielding are to be done
Reactor installation and maintenance for the Advanced Neutron Source
International Nuclear Information System (INIS)
Smith, B.R.
1993-01-01
Advanced Neutron Source (ANS) reactor assembly components have been modeled in great detail in IGRIP in order to realistically simulate preliminary installation and maintenance processes. Animation of these processes has been captured in a 15-minute video with narration. Approximately 90% of the parts were initially translated from CADAM (a two-dimensional drawing package) to IGRIP and then revolved or extruded. IGRIP's IGES translator greatly reduced the time required to perform this operation. The interfacing of devices in the work cell has identified numerous design inconsistencies. Most of the modeled reactor components are devices with a single degree of freedom (DOF) however, some of the slanted experiments required 6 DOF so that they could be removed at an angle in order to clear the reflector vessel flanges. IGRIP's collision detection feature proved to be extremely helpful in determining interferences when removing the experiments. The combination of three-dimensional visualization and collision detection allows engineers to clearly and easily visualize potential design problems before the construction phase of the project
Advanced neutron source design: burnout heat flux correlation development
International Nuclear Information System (INIS)
Gambill, W.R.; Mochizuki, T.
1988-01-01
In the advanced neutron source reactor (ANSR) fuel element region, heat fluxes will be elevated. Early designs corresponded to average and estimated hot-spot fluxes of 11 to 12 and 21 to 22 MW/m 2 , respectively. Design changes under consideration may lower these values to ∼ 9 and 17 MW/m 1 . In either event, the development of a satisfactory burnout heat flux correlation is an important element among the many thermal-hydraulic design issues, since the critical power ratio will depend in part on its validity. Relatively little work in the area of subcooled-flow burnout has been published over the past 12 yr. The authors have compared seven burnout correlations and modifications therefore with several sets of experimental data, of which the most relevant to the ANS core are those referenced. The best overall agreement between the correlations tested and these data is currently provided by a modification of Thorgerson et al. correlation. The variable ranges of the experimental data are outlined and the results of the correlation comparisons are summarized
The advanced neutron source three-element-core fuel grading
International Nuclear Information System (INIS)
Gehin, J.C.
1995-01-01
The proposed Advanced Neutron Source (ANS) pre-conceptual design consists of a two-element 330 MW f nuclear reactor fueled with highly-enriched uranium and is cooled, moderated, and reflected with heavy water. Recently, the ANS design has been changed to a three-element configuration in order to permit a reduction of the enrichment, if required, while maintaining or improving the thermal-hydraulic margins. The core consists of three annular fuel elements composed of involute-shaped fuel plates. Each fuel plate has a thickness of 1.27 mm and consists of a fuel meat region Of U 3 Si 2 -Al (50% enriched in one case that was proposed) and an aluminum filler region between aluminum cladding. The individual plates are separated by a 1.27 mm coolant channel. The three element core has a fuel loading of 31 kg of 235 U which is sufficient for a 17-day fuel cycle. The goal in obtaining a new fuel grading is to maximize important temperature margins. The limits imposed axe: (1) Limit the temperature drop over the cladding oxide layer to less than 119 degrees C to avoid oxide spallation. (2) Limit the fuel centerline temperature to less than 400 degrees C to avoid fuel damage. (3) Limit the cladding wall temperature to less than the coolant. incipient-boiling temperature to avoid coolant boiling. Other thermal hydraulic conditions, such as critical heat flux, are also considered
International Nuclear Information System (INIS)
Taleyarkhan, R.P.; Georgevich, V.; Kim, S.H.; Valenti, S.; Simpson, D.B.; Sawruk, W.
1994-01-01
This paper describes salient aspects of the modeling, analyses, and evaluations for hydrogen detonation in selected regions of the Advanced Neutron Source (ANS) containment during hypothetical severe accident conditions. Shock wave generation and transport modeling and analyses were conducted for two stratified configurations in the dome region of the high bay. Principal tools utilized for these purposes were the CTH and CET89 computer codes. Dynamic pressure loading functions were generated for key locations and used for evaluating structural response behavior for which a finite-element model was developed using the ANSYS code. For the range of conditions analyzed in the two critical dome regions, it was revealed that the ANS containment would be able to withstand detonation loads without failure. (author)
Analysis of core-concrete interaction event with flooding for the Advanced Neutron Source reactor
International Nuclear Information System (INIS)
Kim, S.H.; Taleyarkhan, R.P.; Georgevich, V.; Navarro-Valenti, S.
1993-01-01
This paper discusses salient aspects of the methodology, assumptions, and modeling of various features related to estimation of source terms from an accident involving a molten core-concrete interaction event (with and without flooding) in the Advanced Neutron Source (ANS) reactor at the Oak Ridge National Laboratory. Various containment configurations are considered for this postulated severe accident. Several design features (such as rupture disks) are examined to study containment response during this severe accident. Also, thermal-hydraulic response of the containment and radionuclide transport and retention in the containment are studied. The results are described as transient variations of source terms, which are then used for studying off-site radiological consequences and health effects for the support of the Conceptual Safety Analysis Report for ANS. The results are also to be used to examine the effectiveness of subpile room flooding during this type of severe accident
International Nuclear Information System (INIS)
Kim, S.H.; Taleyarkhan, R.P.; Georgevich, V.
1992-10-01
This paper discusses salient aspects of methodology, assumptions, and modeling of various features related to estimation of source terms from two conservatively scoped severe accident scenarios in the Advanced Neutron Source (ANS) reactor at the Oak Ridge National Laboratory. Various containment configurations are considered for steaming-pool-type accidents and an accident involving molten core-concrete interaction. Several design features (such as rupture disks) are examined to study containment response during postulated severe accidents. Also, thermal-hydraulic response of the containment and radionuclide transport and retention in the containment are studied. The results are described as transient variations of source terms for each scenario, which are to be used for studying off-site radiological consequences and health effects for these postulated severe accidents. Also highlighted will be a comparison of source terms estimated by two different versions of the MELCOR code
International Nuclear Information System (INIS)
Taleyarkhan, R.P.; Georgevich, V.; Kim, S.H.; Valenti, S.N.; Simpson, D.B.; Sawruk, W.
1994-07-01
This paper describes salient aspects of the modeling, analyses, and evaluations for hydrogen detonation in selected regions of the Advanced Neutron Source (ANS) containment during hypothetical severe accident conditions. Shock wave generation and transport modeling and analyses were conducted for two stratified configurations in the dome region of the high bay. Principal tools utilized for these purposes were the CTH and CET89 computer codes. Dynamic pressure loading functions were generated for key locations and used for evaluating structural response behavior for which a finite-element model was developed using the ANSYS code. For the range of conditions analyzed in the two critical dome regions, it was revealed that the ANS containment would be able to withstand detonation loads without failure
International Nuclear Information System (INIS)
Hayter, J.B.
1994-01-01
New neutron sources being planned, such as the Advanced Neutron Source (ANS) or the European Spallation Source (ESS), will provide an order of magnitude flux increase over what is available today, but neutron scattering will still remain a signal-limited technique. At the same time, the development of new materials, such as polymer and ceramic composites or a variety of complex fluids, will increasingly require neutron-based research. This paper will discuss some of the new techniques which will allow us to make. better use of the available neutrons, either through improved instrumentation or through sample manipulation. Discussion will center primarily on unpolarized neutron techniques since polarized neutrons will be the subject of the next paper
The advanced neutron source - A world-class research reactor facility
International Nuclear Information System (INIS)
Thompson, P.B.; Meek, W.E.
1993-01-01
The advanced neutron source (ANS) is a new facility being designed at the Oak Ridge National Laboratory that is based on a heavy-water-moderated reactor and extensive experiment and user-support facilities. The primary purpose of the ANS is to provide world-class facilities for neutron scattering research, isotope production, and materials irradiation in the United States. The neutrons provided by the reactor will be thermalized to produce sources of hot, thermal, cold, very cold, and ultracold neutrons usable at the experiment stations. Beams of cold neutrons will be directed into a large guide hall using neutron guide technology, greatly enhancing the number of research stations possible in the project. Fundamental and nuclear physics, materials analysis, and other research pro- grams will share the neutron beam facilities. Sufficient laboratory and office space will be provided to create an effective user-oriented environment
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.)
ADVANCES IN ZERO ENERGY TRANSPORTATION SYSTEMS
Ahmad, Othman
2017-01-01
Hyperloop mass transportation systems are activelydeveloped at the moment. They represent the forefront development of the ZeroEnergy Transportation systems where air drag is minimized by travelling in avacuum and friction is reduced by non-contact bearings. Hyperloop supportersare confident that the cost of their transportation systems would be lowcompared to existing transportation systems because of the low loss andtherefore low energy consumption as well as other cost-saving techniquesdoc...
Development of advanced radiation monitors for pulsed neutron fields
AUTHOR|(CDS)2081895
The need of radiation detectors capable of efficiently measuring in pulsed neutron fields is attracting widespread interest since the 60s. The efforts of the scientific community substantially increased in the last decade due to the increasing number of applications in which this radiation field is encountered. This is a major issue especially at particle accelerator facilities, where pulsed neutron fields are present because of beam losses at targets, collimators and beam dumps, and where the correct assessment of the intensity of the neutron fields is fundamental for radiation protection monitoring. LUPIN is a neutron detector that combines an innovative acquisition electronics based on logarithmic amplification of the collected current signal and a special technique used to derive the total number of detected neutron interactions, which has been specifically conceived to work in pulsed neutron fields. Due to its special working principle, it is capable of overcoming the typical saturation issues encountere...
New opportunities in neutron capture research using advanced pulsed neutron sources
International Nuclear Information System (INIS)
Bowman, C.D.
1987-08-01
The extraordinary neutron intensities available from the new spallation pulsed neutron sources open up exciting opportunities for basic and applied research in neutron nuclear physics. Prospective experiments are reviewed with particular attention to those with a strong connection to capture gamma-ray spectroscopy
Experimental validation of GADRAS's coupled neutron-photon inverse radiation transport solver
International Nuclear Information System (INIS)
Mattingly, John K.; Mitchell, Dean James; Harding, Lee T.
2010-01-01
Sandia National Laboratories has developed an inverse radiation transport solver that applies nonlinear regression to coupled neutron-photon deterministic transport models. The inverse solver uses nonlinear regression to fit a radiation transport model to gamma spectrometry and neutron multiplicity counting measurements. The subject of this paper is the experimental validation of that solver. This paper describes a series of experiments conducted with a 4.5 kg sphere of α-phase, weapons-grade plutonium. The source was measured bare and reflected by high-density polyethylene (HDPE) spherical shells with total thicknesses between 1.27 and 15.24 cm. Neutron and photon emissions from the source were measured using three instruments: a gross neutron counter, a portable neutron multiplicity counter, and a high-resolution gamma spectrometer. These measurements were used as input to the inverse radiation transport solver to evaluate the solver's ability to correctly infer the configuration of the source from its measured radiation signatures.
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)
International Nuclear Information System (INIS)
Kotegawa, Hiroshi; Sasamoto, Nobuo; Tanaka, Shun-ichi
1987-02-01
Both ''measured radioactive inventory due to neutron activation in the shield concrete of JPDR'' and ''measured intermediate and low energy neutron spectra penetrating through a graphite sphere'' are analyzed using a continuous energy model Monte Carlo code MCNP so as to estimate calculational accuracy of the code for neutron transport in thermal and epithermal energy regions. Analyses reveal that MCNP calculates thermal neutron spectra fairly accurately, while it apparently over-estimates epithermal neutron spectra (of approximate 1/E distribution) as compared with the measurements. (author)
Advances in the development of Cr-39 based neutron dosimeters
International Nuclear Information System (INIS)
Hadlock, D.E.; Parkhurst, M.A.
1987-12-01
A combination thermoluminescent dosimeter (TLD) and track etch dosimeter (TED), which can be used for detecting neutrons over a wide energy range, has been developed through recent research in passive neutron dosimetery. This dosimeter uses Li-600 TLDs to detect thermal and low energy neutrons reflected from the body, and the TED polymer of CR-39, to detect fast neutrons from proton recoil interactions with the polyethylene radiator or with CR-39 itself. Some form of the combination dosimeter is currently in use at several US Department of Energy (DOE) facilities, and its use is expected to expand over the next year to include all DOE facilities where significant neutron exposures may occur. The extensive research conducted on the TED component over the past six years has continually focused on material improvements, reduction in processing time and dosimeter handling, and ease of sample readout with the goal of automating the process as much as possible. 1 fig
A Review of Significant Advances in Neutron Imaging from Conception to the Present
Brenizer, J. S.
This review summarizes the history of neutron imaging with a focus on the significant events and technical advancements in neutron imaging methods, from the first radiograph to more recent imaging methods. A timeline is presented to illustrate the key accomplishments that advanced the neutron imaging technique. Only three years after the discovery of the neutron by English physicist James Chadwick in 1932, neutron imaging began with the work of Hartmut Kallmann and Ernst Kuhn in Berlin, Germany, from 1935-1944. Kallmann and Kuhn were awarded a joint US Patent issued in January 1940. Little progress was made until the mid-1950's when Thewlis utilized a neutron beam from the BEPO reactor at Harwell, marking the beginning of the application of neutron imaging to practical applications. As the film method was improved, imaging moved from a qualitative to a quantitative technique, with applications in industry and in nuclear fuels. Standards were developed to aid in the quantification of the neutron images and the facility's capabilities. The introduction of dynamic neutron imaging (initially called real-time neutron radiography and neutron television) in the late 1970's opened the door to new opportunities and new challenges. As the electronic imaging matured, the introduction of the CCD imaging devices and solid-state light intensifiers helped address some of these challenges. Development of improved imaging devices for the medical community has had a major impact on neutron imaging. Additionally, amorphous silicon sensors provided improvements in temporal resolution, while providing a reasonably large imaging area. The development of new neutron imaging sensors and the development of new neutron imaging techniques in the past decade has advanced the technique's ability to provide insight and understanding of problems that other non-destructive techniques could not provide. This rapid increase in capability and application would not have been possible without the
IMPROVED COMPUTATIONAL NEUTRONICS METHODS AND VALIDATION PROTOCOLS FOR THE ADVANCED TEST REACTOR
Energy Technology Data Exchange (ETDEWEB)
David W. Nigg; Joseph W. Nielsen; Benjamin M. Chase; Ronnie K. Murray; Kevin A. Steuhm
2012-04-01
The Idaho National Laboratory (INL) is in the process of modernizing the various reactor physics modeling and simulation tools used to support operation and safety assurance of the Advanced Test Reactor (ATR). Key accomplishments so far have encompassed both computational as well as experimental work. A new suite of stochastic and deterministic transport theory based reactor physics codes and their supporting nuclear data libraries (HELIOS, KENO6/SCALE, NEWT/SCALE, ATTILA, and an extended implementation of MCNP5) has been installed at the INL. Corresponding models of the ATR and ATRC are now operational with all five codes, demonstrating the basic feasibility of the new code packages for their intended purpose. Of particular importance, a set of as-run core depletion HELIOS calculations for all ATR cycles since August 2009 was successfully completed during 2011. This demonstration supported a decision late in the year to proceed with the phased incorporation of the HELIOS methodology into the ATR fuel cycle management process beginning in 2012. On the experimental side of the project, new hardware was fabricated, measurement protocols were finalized, and the first four of six planned physics code validation experiments based on neutron activation spectrometry were conducted at the ATRC facility. Data analysis for the first three experiments, focused on characterization of the neutron spectrum in one of the ATR flux traps, has been completed. The six experiments will ultimately form the basis for a flexible, easily-repeatable ATR physics code validation protocol that is consistent with applicable ASTM standards.
International Nuclear Information System (INIS)
Simakov, S.P.; Fischer, U.; Moellendorff, U. von; Schmuck, I.; Konobeev, A.Yu.; Korovin, Yu.A.; Pereslavtsev, P.
2002-01-01
A newly developed computational procedure is presented for the generation of d-Li source neutrons in Monte Carlo transport calculations based on the use of evaluated double-differential d+ 6,7 Li cross section data. A new code M c DeLicious was developed as an extension to MCNP4C to enable neutronics design calculations for the d-Li based IFMIF neutron source making use of the evaluated deuteron data files. The M c DeLicious code was checked against available experimental data and calculation results of M c DeLi and MCNPX, both of which use built-in analytical models for the Li(d, xn) reaction. It is shown that M c DeLicious along with newly evaluated d+ 6,7 Li data is superior in predicting the characteristics of the d-Li neutron source. As this approach makes use of tabulated Li(d, xn) cross sections, the accuracy of the IFMIF d-Li neutron source term can be steadily improved with more advanced and validated data
Simakov, S P; Moellendorff, U V; Schmuck, I; Konobeev, A Y; Korovin, Y A; Pereslavtsev, P
2002-01-01
A newly developed computational procedure is presented for the generation of d-Li source neutrons in Monte Carlo transport calculations based on the use of evaluated double-differential d+ sup 6 sup , sup 7 Li cross section data. A new code M sup c DeLicious was developed as an extension to MCNP4C to enable neutronics design calculations for the d-Li based IFMIF neutron source making use of the evaluated deuteron data files. The M sup c DeLicious code was checked against available experimental data and calculation results of M sup c DeLi and MCNPX, both of which use built-in analytical models for the Li(d, xn) reaction. It is shown that M sup c DeLicious along with newly evaluated d+ sup 6 sup , sup 7 Li data is superior in predicting the characteristics of the d-Li neutron source. As this approach makes use of tabulated Li(d, xn) cross sections, the accuracy of the IFMIF d-Li neutron source term can be steadily improved with more advanced and validated data.
Advances in neutronics calculation of fast neutron reactors - Demonstration on Super-Phenix reactor
International Nuclear Information System (INIS)
Czernecki, Sebastien
1998-01-01
The fast reactor european neutronics calculations system, ERANOS, has integrated recent improvements both in nuclear data, with the use of the adjusted nuclear library ERALIB 1 from the JEF2.2 library, and calculation methods, with the use of the new european cell code, ECCO, and the deterministic code, TGV/VARIANT. This code performs full 3-D reactor calculation in the transport theory with variational method. The aim of this work is to create and validate a new calculational scheme for fast spectrum systems offering good compromise between accuracy and running time. The new scheme is based on these improvements plus a special procedure accounting for control rod heterogeneity, which uses a reactivity equivalence homogenization. The new scheme has been validated by means of experiment/calculation comparisons, using the extensive start-up program measurements performed in Super-Phenix reactor. The validation uses also recent measurements performed in the Phenix reactor. The results are very satisfactory and show a significant improvement for almost all core parameters, especially for critical mass, control rod worth and radial subassembly power distribution. A detailed analysis of the discrepancies between the old scheme and the new one for this parameter allows to understand the separate effects of methods and nuclear data on the radial power distribution shape. (author) [fr
AUS, Neutron Transport and Gamma Transport System for Fission Reactors and Fusion Reactors
International Nuclear Information System (INIS)
1990-01-01
1 - Description of program or function: AUS is a neutronics code system which may be used for calculations of a wide range of fission reactors, fusion blankets and other neutron applications. The present version, AUS98, has a nuclear cross section library based on ENDF/B-VI and includes modules which provide for reactor lattice calculations, one-dimensional transport calculations, multi-dimensional diffusion calculations, cell and whole reactor burnup calculations, and flexible editing of results. Calculations of multi-region resonance shielding, coupled neutron and photon transport, energy deposition, fission product inventory and neutron diffusion are combined within the one code system. The major changes from the previous release, AUS87, are the inclusion of a cross-section library based on ENDF/B-VI, the addition of the POW3D multi-dimensional diffusion module, the addition of the MICBURN module for controlling whole reactor burnup calculations, and changes to the system as a consequence of moving from IBM mainframe computers to UNIX workstations. 2 - Method of solution: AUS98 is a modular system in which the modules are complete programs linked by a path given in the input stream. A simple path is simply a sequence of modules, but the path is actually pre-processed and compiled using the Fortran 77 compiler. This provides for complex module linking if required. Some of the modules included in AUS98 are: MIRANDA Cross-section generation in a multi-region resonance subgroup calculation and preliminary group condensation. ANAUSN One-dimensional discrete ordinates calculation. ICPP Isotropic collision probability calculation in one dimension and for rod clusters. POW3D Multi-dimensional neutron diffusion calculation including feedback-free kinetics. AUSIDD One-dimensional diffusion calculation. EDITAR Reaction-rate editing and group collapsing following a transport calculation. CHAR Lattice and global burnup calculation. MICBURN Control of global burnup
Numerical study of the particle transport in fast neutron detectors with conversion layer
International Nuclear Information System (INIS)
Sedlackova, K.; Zatko, B.; Necas, V.
2012-01-01
This paper deals with fast neutron and recoil proton transport simulation using statistical analysis of Monte Carlo radiation transport code (MCNPX). Its possibilities in the detector design and optimization are presented. MCNPX proved as a very advantageous self-contained simulation program for fast neutron and secondary proton tracking. Simulations of respective particle transport through conversion layer of HDPE and further in the active volume of detector let us to follow important characteristics as neutron/proton flux density, reaction rate of elastic scattering on hydrogen nuclei and deposited energy as well as their dependencies on incident neutron energy and conversion layer/active region thickness. The efficiency of neutrons to protons conversion has been calculated and its maximum was reached for 500 μm thick conversion layer. The minimum active region thickness has been estimated to be about 300 μm.(authors)
Monte Carlo study of the mechanisms of transport of fast neutrons in various media
International Nuclear Information System (INIS)
Ku, L.
1976-01-01
The technique of analyzing Monte Carlo histories was used to study the details of neutron transport and slowing down mechanisms. The statistical properties of life histories of ''exceptional'' neutrons, i.e., those staying closer to the source or penetrating to larger distances from the source, were compared to those of the general population. The macroscopic behavior of ''exceptional'' neutrons was also related to the interaction mechanics and to the microscopic properties of the medium
The Advanced Neutron Source (ANS) project: A world-class research reactor facility
International Nuclear Information System (INIS)
Thompson, P.B.; Meek, W.E.
1993-01-01
This paper provides an overview of the Advanced Neutron Source (ANS), a new research facility being designed at Oak Ridge National Laboratory. The facility is based on a 330 MW, heavy-water cooled and reflected reactor as the neutron source, with a thermal neutron flux of about 7.5x10 19 m -2 ·sec -1 . Within the reflector region will be one hot source which will serve 2 hot neutron beam tubes, two cryogenic cold sources serving fourteen cold neutron beam tubes, two very cold beam tubes, and seven thermal neutron beam tubes. In addition there will be ten positions for materials irradiation experiments, five of them instrumented. The paper touches on the project status, safety concerns, cost estimates and scheduling, a description of the site, the reactor, and the arrangements of the facilities
Terrestrial neutron-induced soft errors in advanced memory devices
Nakamura, Takashi; Ibe, Eishi; Yahagi, Yasuo; Kameyama, Hideaki
2008-01-01
Terrestrial neutron-induced soft errors in semiconductor memory devices are currently a major concern in reliability issues. Understanding the mechanism and quantifying soft-error rates are primarily crucial for the design and quality assurance of semiconductor memory devices. This book covers the relevant up-to-date topics in terrestrial neutron-induced soft errors, and aims to provide succinct knowledge on neutron-induced soft errors to the readers by presenting several valuable and unique features. Sample Chapter(s). Chapter 1: Introduction (238 KB). Table A.30 mentioned in Appendix A.6 on
Adjuvant neutron therapy in complex treatment of patients with locally advanced breast cancer
Lisin, V. A.; Velikaya, V. V.; Startseva, Zh. A.; Popova, N. O.; Goldberg, V. E.
2017-09-01
The study included 128 patients with stage T2-4N0-3M0 locally advanced breast cancer. All patients were divided into two groups. Group I (study group) consisted of 68 patients, who received neutron therapy, and group II (control group) comprised 60 patients, who received electron beam therapy. Neutron therapy was well tolerated by the patients and 1-2 grade radiation skin reactions were the most common. Neutron therapy was shown to be effective in multimodality treatment of the patients with locally advanced breast cancer. The 8-year recurrence-free survival rate in the patients with locally advanced breast cancer was 94.5 ± 4.1% after neutron therapy and 81.4 ± 5.9% after electron beam therapy (p = 0.05).
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.
1975-01-01
The objectives of the research remain the same as outlined in the original proposal. They are in short as follows: Develop mathematically and computationally founded criteria for the design of highly efficient and reliable multi-dimensional neutron transport codes to solve a variety of neutron migration and radiation problems and analyze existing and new methods for performance. (U.S.)
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.)
On the reciprocity-like relations in linear neutron transport theory
International Nuclear Information System (INIS)
Modak, R.S.; Sahni, D.C.
1997-01-01
The existence of certain reciprocity-like relations in neutron transport theory was shown earlier under some quite restrictive conditions. Here, these relations are shown to be valid in more general situations by using a different approach based on individual neutron trajectories. (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
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.)
Advanced Public Transportation Systems. Technical Assistance Brief 2
1993-01-01
The Advanced Public Transportation Systems (APTS) Program requires evaluation : of pilot projects. This technical assistance brief discusses the guidelines set : for developing evaluation framework and methodology for local projects. 4p.
International Nuclear Information System (INIS)
Hoogenboom, J. Eduard
2003-01-01
Adjoint Monte Carlo may be a useful alternative to regular Monte Carlo calculations in cases where a small detector inhibits an efficient Monte Carlo calculation as only very few particle histories will cross the detector. However, in general purpose Monte Carlo codes, normally only the multigroup form of adjoint Monte Carlo is implemented. In this article the general methodology for continuous-energy adjoint Monte Carlo neutron transport is reviewed and extended for photon and coupled neutron-photon transport. In the latter cases the discrete photons generated by annihilation or by neutron capture or inelastic scattering prevent a direct application of the general methodology. Two successive reaction events must be combined in the selection process to accommodate the adjoint analog of a reaction resulting in a photon with a discrete energy. Numerical examples illustrate the application of the theory for some simplified problems
Identification of materials by an advanced neutronic method
International Nuclear Information System (INIS)
El Kanawati, W.
2011-01-01
The EURITRACK inspection system, based on the associated particle technique, aims at detecting explosives and narcotics in cargo containers with 14 MeV neutrons produced by the D(T,γ)n reaction. Alpha particle and neutron are emitted almost back to back. Reactions induced by fast neutrons produce gamma rays which are detected in coincidence with the alpha particle to determine the neutron direction. Neutron time-of-flight allows to determine gamma-ray origin inside the container. Information concerning material composition is obtained by unfolding the gamma spectrum into elemental signatures using a database of elemental spectra (C, O, N, Fe...). Carbon, oxygen, and nitrogen count ratios are converted into chemical proportions to distinguish illicit and benign organic materials. Conversion factors based on Monte Carlo simulations have been calculated and validated experimentally, taking into account neutron slowing down and photon attenuation in cargo materials. Application to the elemental characterisation of radioactive wastes is also studied by numerical simulation, with shields and collimators to limit the background due to waste radiations. (author)
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)
TEMPS, 1-Group Time-Dependent Pulsed Source Neutron Transport
International Nuclear Information System (INIS)
Ganapol, B.D.
1988-01-01
1 - Description of program or function: TEMPS numerically determines the scalar flux as given by the one-group neutron transport equation with a pulsed source in an infinite medium. Standard plane, point, and line sources are considered as well as a volume source in the negative half-space in plane geometry. The angular distribution of emitted neutrons can either be isotropic or mono-directional (beam) in plane geometry and isotropic in spherical and cylindrical geometry. A general anisotropic scattering Kernel represented in terms of Legendre polynomials can be accommodated with a time- dependent number of secondaries given by c(t)=c 0 (t/t 0 ) β , where β is greater than -1 and less than infinity. TEMPS is designed to provide the flux to a high degree of accuracy (4-5 digits) for use as a benchmark to which results from other numerical solutions or approximations can be compared. 2 - Method of solution: A semi-analytic Method of solution is followed. The main feature of this approach is that no discretization of the transport or scattering operators is employed. The numerical solution involves the evaluation of an analytical representation of the solution by standard numerical techniques. The transport equation is first reformulated in terms of multiple collisions with the flux represented by an infinite series of collisional components. Each component is then represented by an orthogonal Legendre series expansion in the variable x/t where the distance x and time t are measured in terms of mean free path and mean free time, respectively. The moments in the Legendre reconstruction are found from an algebraic recursion relation obtained from Legendre expansion in the direction variable mu. The multiple collision series is evaluated first to a prescribed relative error determined by the number of digits desired in the scalar flux. If the Legendre series fails to converge in the plane or point source case, an accelerative transformation, based on removing the
PHISICS multi-group transport neutronic capabilities for RELAP5
Energy Technology Data Exchange (ETDEWEB)
Epiney, A.; Rabiti, C.; Alfonsi, A.; Wang, Y.; Cogliati, J.; Strydom, G. [Idaho National Laboratory (INL), 2525 N. Fremont Ave., Idaho Falls, ID 83402 (United States)
2012-07-01
PHISICS is a neutronic code system currently under development at INL. Its goal is to provide state of the art simulation capability to reactor designers. This paper reports on the effort of coupling this package to the thermal hydraulic system code RELAP5. This will enable full prismatic core and system modeling and the possibility to model coupled (thermal-hydraulics and neutronics) problems with more options for 3D neutron kinetics, compared to the existing diffusion theory neutron kinetics module in RELAP5 (NESTLE). The paper describes the capabilities of the coupling and illustrates them with a set of sample problems. (authors)
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
Calculation of neutron and gamma transport at the FOA:type of problems and calculation methods
International Nuclear Information System (INIS)
Lefvert, T.
1975-11-01
Protection against the effects of nuclear warfare involves the analysis of the forms of results of a nuclear charge explosion producing neutron and gamma radiation. It brings out problems leading to the calculation of criticality, leakage, and deep transmission. Methods have been developed for various kinds of particle transport problems. Applications to radiation therapy, storage of fissile materials, and fast reactors are discussed. A list (with brief description) of all neutron and gamma transport programmes of the FOA is given. (J.S.)
Application of direct discrete method (DDM) to multigroup neutron transport problems
International Nuclear Information System (INIS)
Vosoughi, Naser; Salehi, Ali Akbar; Shahriari, Majid
2003-01-01
The Direct Discrete Method (DDM), which produced excellent results for one-group neutron transport problems, has been developed for multigroup energy. A multigroup neutron transport discrete equation has been produced for a cylindrical shape fuel element with and without associated coolant regions with two boundary conditions. The calculations are illustrated for two-group energy by graphs showing the fast and thermal fluxes. The validity of the results are tested against the results obtained by the ANISN code. (author)
Transport calculation of thermal and cold neutrons using NMTC/JAERI-MCNP4A code system
International Nuclear Information System (INIS)
Iga, Kiminori; Takada, Hiroshi; Nagao, Tadashi.
1998-01-01
In order to investigate the applicability of the NMTC/JAERI-MCNP4A code system to the neutronics design study in the neutron science research project of JAERI, transport calculations of thermal and cold neutrons are performed with the code system on a spallation neutron source composed of light water cooled tantalum target with a moderator and a reflector system. The following neutronic characteristics are studied in the calculation : the variation of the intensity of neutrons emitted from a light water moderator or a liquid hydrogen with/without the B 4 C decoupler, which are installed to produce sharp pulse, and that dependent on the position of external source neutrons in the tantalum target. The calculated neutron energy spectra are reproduced well by the semi-empirical formula with the parameter values reliable in physical meanings. It is found to be necessary to employ proper importance sampling technique in the statistics. It is confirmed from this work that the NMTC/JAERI-MCNP4A code system is applicable to the neutronics design study of spallation neutron sources proposed for the neutron science research project. (author)
Transport calculation of thermal and cold neutrons using NMTC/JAERI-MCNP4A code system
Energy Technology Data Exchange (ETDEWEB)
Iga, Kiminori [Kyushu Univ., Fukuoka (Japan); Takada, Hiroshi; Nagao, Tadashi
1998-01-01
In order to investigate the applicability of the NMTC/JAERI-MCNP4A code system to the neutronics design study in the neutron science research project of JAERI, transport calculations of thermal and cold neutrons are performed with the code system on a spallation neutron source composed of light water cooled tantalum target with a moderator and a reflector system. The following neutronic characteristics are studied in the calculation : the variation of the intensity of neutrons emitted from a light water moderator or a liquid hydrogen with/without the B{sub 4}C decoupler, which are installed to produce sharp pulse, and that dependent on the position of external source neutrons in the tantalum target. The calculated neutron energy spectra are reproduced well by the semi-empirical formula with the parameter values reliable in physical meanings. It is found to be necessary to employ proper importance sampling technique in the statistics. It is confirmed from this work that the NMTC/JAERI-MCNP4A code system is applicable to the neutronics design study of spallation neutron sources proposed for the neutron science research project. (author)
Neutronics investigation of advanced self-cooled liquid blanket systems in helical reactor
International Nuclear Information System (INIS)
Tanaka, T.; Sagara, A.; Muroga, T.; Youssef, M.Z.
2006-10-01
Neutronics performances of advanced self-cooled liquid blanket systems have been investigated in design activity of the helical-type reactor FFHR2. In the present study, a new three-dimensional (3-D) neutronics calculation system has been developed for the helical-type reactor to enhance quick feedback between neutronics evaluation and design modification. Using this new calculation system, advanced Flibe-cooled and Li-cooled liquid blanket systems proposed for FFHR2 have been evaluated to make clear design issues to enhance neutronics performance. Based on calculated results, modification of the blanket dimensions and configuration have been attempted to achieve the adequate tritium breeding ability and neutron shielding performance in the helical reactor. The total tritium breeding ratios (TBRs) obtained after modifying the blanket dimensions indicated that all the advanced blanket systems proposed for FFHR2 would achieve adequate tritium self-sufficiency by dimension adjustment and optimization of structures in the breeder layers. Issues in neutron shielding performance have been investigated quantitatively using 3-D geometry of the helical blanket system, support structures, poloidal coils etc. Shielding performance of the helical coils against direct neutrons from core plasma would achieve design target by further optimization of shielding materials. However, suppression of the neutron streaming and reflection through the divertor pumping areas in the original design is important issue to protect the poloidal coils and helical coils, respectively. Investigation of the neutron wall loading indicated that the peaking factor of the neutron wall load distribution would be moderated by the toroidal and helical effect of the plasma distribution in the helical reactor. (author)
Continuous energy adjoint Monte Carlo for coupled neutron-photon transport
Energy Technology Data Exchange (ETDEWEB)
Hoogenboom, J.E. [Delft Univ. of Technology (Netherlands). Interfaculty Reactor Inst.
2001-07-01
Although the theory for adjoint Monte Carlo calculations with continuous energy treatment for neutrons as well as for photons is known, coupled neutron-photon transport problems present fundamental difficulties because of the discrete energies of the photons produced by neutron reactions. This problem was solved by forcing the energy of the adjoint photon to the required discrete value by an adjoint Compton scattering reaction or an adjoint pair production reaction. A mathematical derivation shows the exact procedures to follow for the generation of an adjoint neutron and its statistical weight. A numerical example demonstrates that correct detector responses are obtained compared to a standard forward Monte Carlo calculation. (orig.)
International Nuclear Information System (INIS)
Singleterry, R.C. Jr.; Wilson, J.W.
1997-01-01
Extension of the high charge and energy (HZE) transport computer program HZETRN for angular transport of neutrons is considered. For this paper, only light ion transport, He 4 and lighter, will be analyzed using a pure solar proton source. The angular transport calculator is the ANISN/PC program which is being controlled by the HZETRN program. The neutron flux values are compared for straight-ahead transport and angular transport in one dimension. The shield material is aluminum and the target material is water. The thickness of these materials is varied; however, only the largest model calculated is reported which is 50 gm/cm 2 of aluminum and 100 gm/cm 2 of water. The flux from the ANISN/PC calculation is about two orders of magnitude lower than the flux from HZETRN for very low energy neutrons. It is only a magnitude lower for the neutrons in the 10 to 20 MeV range in the aluminum and two orders lower in the water. The major reason for this difference is in the transport modes: straight-ahead versus angular. The angular treatment allows a longer path length than the straight-ahead approximation. Another reason is the different cross section sets used by the ANISN/PC-BUGLE-80 mode and the HZETRN mode. The next step is to investigate further the differences between the two codes and isolate the differences to just the angular versus straight-ahead transport mode. Then, create a better coupling between the angular neutron transport and the charged particle transport
The Application of Neutron Transport Green's Functions to Threat Scenario Simulation
Thoreson, Gregory G.; Schneider, Erich A.; Armstrong, Hirotatsu; van der Hoeven, Christopher A.
2015-02-01
Radiation detectors provide deterrence and defense against nuclear smuggling attempts by scanning vehicles, ships, and pedestrians for radioactive material. Understanding detector performance is crucial to developing novel technologies, architectures, and alarm algorithms. Detection can be modeled through radiation transport simulations; however, modeling a spanning set of threat scenarios over the full transport phase-space is computationally challenging. Previous research has demonstrated Green's functions can simulate photon detector signals by decomposing the scenario space into independently simulated submodels. This paper presents decomposition methods for neutron and time-dependent transport. As a result, neutron detector signals produced from full forward transport simulations can be efficiently reconstructed by sequential application of submodel response functions.
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.
Final LDRD report : advanced plastic scintillators for neutron detection.
Energy Technology Data Exchange (ETDEWEB)
Vance, Andrew L.; Mascarenhas, Nicholas; O' Bryan, Greg; Mrowka, Stanley
2010-09-01
This report summarizes the results of a one-year, feasibility-scale LDRD project that was conducted with the goal of developing new plastic scintillators capable of pulse shape discrimination (PSD) for neutron detection. Copolymers composed of matrix materials such as poly(methyl methacrylate) (PMMA) and blocks containing trans-stilbene (tSB) as the scintillator component were prepared and tested for gamma/neutron response. Block copolymer synthesis utilizing tSBMA proved unsuccessful so random copolymers containing up to 30% tSB were prepared. These copolymers were found to function as scintillators upon exposure to gamma radiation; however, they did not exhibit PSD when exposed to a neutron source. This project, while falling short of its ultimate goal, demonstrated the possible utility of single-component, undoped plastics as scintillators for applications that do not require PSD.
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)
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...
Advanced technologies for intelligent transportation systems
Picone, Marco; Amoretti, Michele; Zanichelli, Francesco; Ferrari, Gianluigi
2015-01-01
This book focuses on emerging technologies in the field of Intelligent Transportation Systems (ITSs) namely efficient information dissemination between vehicles, infrastructures, pedestrians and public transportation systems. It covers the state-of-the-art of Vehicular Ad-hoc Networks (VANETs), with centralized and decentralized (Peer-to-Peer) communication architectures, considering several application scenarios. With a detailed treatment of emerging communication paradigms, including cross networking and distributed algorithms. Unlike most of the existing books, this book presents a multi-layer overview of information dissemination systems, from lower layers (MAC) to high layers (applications). All those aspects are investigated considering the use of mobile devices, such as smartphones/tablets and embedded systems, i.e. technologies that during last years completely changed the current market, the user expectations, and communication networks. The presented networking paradigms are supported and validate...
Simulation study on the cold neutron guides in China advanced research reactor
International Nuclear Information System (INIS)
Guo Liping; Yang Tonghua; Wang Hongli; Sun Kai; Zhao Zhixiang
2003-01-01
The designs of the two cold neutron guides, CNG1 and CNG2, to be built in China advanced research reactor (CARR) are studied with Monte-Carlo simulation technique. The neutron flux density at the exit of the both guides can reach above 1 x10 9 cm -2 ·s -1 under the assumed flux spectrum of the cold neutron source. The transmission efficiency is 50% and 42%, and the maximum divergence is about 2.2 degree and 1.9 degree, respectively for CNG1 and CNG2. Neutron distribution along horizontal direction is quite uniform for both guides, with maximum fluctuation of less than 3%. Gravity can affect neutron distribution along vertical direction considerably
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
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)
New generation of cryogen free advanced superconducting magnets for neutron scattering experiments
International Nuclear Information System (INIS)
Kirichek, O; Adroja, D T; Manuel, P; Bewley, R I; Brown, J; Kouzmenko, G; Wotherspoon, R
2012-01-01
Recent advances in superconducting technology and cryocooler refrigeration have resulted in a new generation of advanced superconducting magnets for neutron beam applications. These magnets have outstanding parameters such as high homogeneity and stability at highest magnetic fields possible, a reasonably small stray field, low neutron scattering background and larger exposure to neutron detectors. At the same time the pulse tube refrigeration technology provides a complete re-condensing regime which allows to minimise the requirements for cryogens without introducing additional noise and mechanical vibrations. The magnets can be used with dilution refrigerator insert which expands the temperature range from 20mK to 300K. Here we are going to present design, test results and the operational data of the 14T magnet for neutron diffraction and the 9T wide angle chopper magnet for neutron spectroscopy developed by Oxford Instruments in collaboration with ISIS neutron source. First scientific results obtained from the neutron scattering experiments with these magnets are also going to be discussed.
Energy Technology Data Exchange (ETDEWEB)
Maor, Moshe H; Errington, R Douglas; Caplan, Richard J; Griffin, Thomas W; Laramore, George E; Parker, Robert G; Burnison, Michele; Stetz, Joann; Zink, Sandra; Davis, Lawrence W; Peters, Lester J
1995-06-15
Purpose: To compare the efficacy of fast-neutron radiotherapy with that of conventionally fractionated photon therapy in the management of patients with locally advanced squamous cell carcinoma of the head and neck. Methods and Materials: Patients with Stage III or IV disease were randomized to receive either 20.4 Gy/12 fractions/4 weeks of neutrons or 70 Gy/35 fractions/7 weeks of photons (control). Between April 1986 and March 1991, 178 patients were entered, 169 of whom were eligible for analysis. The treatment arms were balanced for age, stage, and performance status, but not for primary site of origin. Results: Complete response occurred in 70 and 52% with neutrons and photons, respectively (p = 0.006). Local regional failure at 3 years for all patients was 63% for neutrons and 68% for photons. Actuarial overall survival curves were virtually identical in both study arms, falling to 27% at 3 years. Acute toxicity was similar in the two arms, but late grade 3-5 toxicity was 40% with neutrons compared to 18% with photons (p = 0.008). Conclusion: Although the initial response rate was higher with neutrons, permanent local control and survival were not improved, and the incidence of late normal tissue toxicity was increased. As a result, fast-neutron therapy for advanced squamous cell carcinoma of the head and neck can only be recommended for patients in whom the logistic benefit of treatment in 12 sessions over 4 weeks outweighs the increased risk of late toxicity.
International Nuclear Information System (INIS)
Maor, Moshe H.; Errington, R. Douglas; Caplan, Richard J.; Griffin, Thomas W.; Laramore, George E.; Parker, Robert G.; Burnison, Michele; Stetz, Joann; Zink, Sandra; Davis, Lawrence W.; Peters, Lester J.
1995-01-01
Purpose: To compare the efficacy of fast-neutron radiotherapy with that of conventionally fractionated photon therapy in the management of patients with locally advanced squamous cell carcinoma of the head and neck. Methods and Materials: Patients with Stage III or IV disease were randomized to receive either 20.4 Gy/12 fractions/4 weeks of neutrons or 70 Gy/35 fractions/7 weeks of photons (control). Between April 1986 and March 1991, 178 patients were entered, 169 of whom were eligible for analysis. The treatment arms were balanced for age, stage, and performance status, but not for primary site of origin. Results: Complete response occurred in 70 and 52% with neutrons and photons, respectively (p = 0.006). Local regional failure at 3 years for all patients was 63% for neutrons and 68% for photons. Actuarial overall survival curves were virtually identical in both study arms, falling to 27% at 3 years. Acute toxicity was similar in the two arms, but late grade 3-5 toxicity was 40% with neutrons compared to 18% with photons (p = 0.008). Conclusion: Although the initial response rate was higher with neutrons, permanent local control and survival were not improved, and the incidence of late normal tissue toxicity was increased. As a result, fast-neutron therapy for advanced squamous cell carcinoma of the head and neck can only be recommended for patients in whom the logistic benefit of treatment in 12 sessions over 4 weeks outweighs the increased risk of late toxicity
Energy Technology Data Exchange (ETDEWEB)
Santi, Peter Angelo [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Cutler, Theresa Elizabeth [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Favalli, Andrea [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Koehler, Katrina Elizabeth [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Henzl, Vladimir [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Henzlova, Daniela [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Parker, Robert Francis [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Croft, Stephen [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)
2015-12-01
In order to improve the accuracy and capabilities of neutron multiplicity counting, additional quantifiable information is needed in order to address the assumptions that are present in the point model. Extracting and utilizing higher order moments (Quads and Pents) from the neutron pulse train represents the most direct way of extracting additional information from the measurement data to allow for an improved determination of the physical properties of the item of interest. The extraction of higher order moments from a neutron pulse train required the development of advanced dead time correction algorithms which could correct for dead time effects in all of the measurement moments in a self-consistent manner. In addition, advanced analysis algorithms have been developed to address specific assumptions that are made within the current analysis model, namely that all neutrons are created at a single point within the item of interest, and that all neutrons that are produced within an item are created with the same energy distribution. This report will discuss the current status of implementation and initial testing of the advanced dead time correction and analysis algorithms that have been developed in an attempt to utilize higher order moments to improve the capabilities of correlated neutron measurement techniques.
Recent advances in mass transport in materials
Ochsner, Andreas
2012-01-01
The present topical volume presents a representative cross-section of some recent advances made in the area of diffusion. The range of topics covered is very large, and, this reflects the enormous breadth of the topic of diffusion. The areas covered include diffusion in intermetallics, phenomenological diffusion theory, diffusional creep, kinetics of steel-making, diffusion in thin films, precipitation, diffusional phase transformations, atomistic diffusion simulations, epitaxial growth and diffusion in porous media. Review from Book News Inc.: In 13 invited and peer-reviewed papers, scientist
Comparison of 2D and 3D Neutron Transport Analyses on Yonggwang Unit 3 Reactor
International Nuclear Information System (INIS)
Maeng, Aoung Jae; Kim, Byoung Chul; Lim, Mi Joung; Kim, Kyung Sik; Jeon, Young Kyou; Yoo, Choon Sung
2012-01-01
10 CFR Part 50 Appendix H requires periodical surveillance program in the reactor vessel (RV) belt line region of light water nuclear power plant to check vessel integrity resulting from the exposure to neutron irradiation and thermal environment. Exact exposure analysis of the neutron fluence based on right modeling and simulations is the most important in the evaluation. Traditional 2 dimensional (D) and 1D synthesis methodologies have been widely applied to evaluate the fast neutron (E > 1.0 MeV) fluence exposure to RV. However, 2D and 1D methodologies have not provided accurate fast neutron fluence evaluation at elevations far above or below the active core region. RAPTOR-M3G (RApid Parallel Transport Of Radiation - Multiple 3D Geometries) program for 3D geometries calculation was therefore developed both by Westinghouse Electronic Company, USA and Korea Reactor Integrity Surveillance Technology (KRIST) for the analysis of In-Vessel Surveillance Test and Ex-Vessel Neutron Dosimetry (EVND). Especially EVND which is installed at active core height between biological shielding material and concrete also evaluates axial neutron fluence by placing three dosimetries each at Top, Middle and Bottom part of the angle representing maximum neutron fluence. The EVND programs have been applied to the Korea Nuclear Plants. The objective of this study is therefore to compare the 3D and the 2D Neutron Transport Calculations and Analyses on the Yonggwang unit 3 Reactor as an example
Special Features of the Air to Space Neutron Transport Problem
2017-09-14
an atmosphere model. Radioactive Decay Free neutrons are not stable elementary particles. They decay radioactively with a half- life of around ten...milliseconds to seconds, so that radioactive decay of neutrons is negligible. (The probability of decay in 100 milliseconds with a 10 minute half- life is...the bottom and top of a layer are 1bZ - and bZ respectively. The methods developed here apply to any planet with an atmosphere and an orbiting
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
TDTORT: Time-Dependent, 3-D, Discrete Ordinates, Neutron Transport Code System with Delayed Neutrons
International Nuclear Information System (INIS)
2002-01-01
1 - Description of program or function: TDTORT solves the time-dependent, three-dimensional neutron transport equation with explicit representation of delayed neutrons to estimate the fission yield from fissionable material transients. This release includes a modified version of TORT from the C00650MFMWS01 DOORS3.1 code package plus the time-dependent TDTORT code. GIP is also included for cross-section preparation. TORT calculates the flux or fluence of particles due to particles incident upon the system from extraneous sources or generated internally as a result of interaction with the system in two- or three-dimensional geometric systems. The principle application is to the deep-penetration transport of neutrons and photons. Reactor eigenvalue problems can also be solved. Numerous printed edits of the results are available, and results can be transferred to output files for subsequent analysis. TDTORT reads ANISN-format cross-section libraries, which are not included in the package. Users may choose from several available in RSICC's data library collection which can be identified by the keyword 'ANISN FORMAT'. 2 - Methods:The time-dependent spatial flux is expressed as a product of a space-, energy-, and angle-dependent shape function, which is usually slowly varying in time and a purely time-dependent amplitude function. The shape equation is solved for the shape using TORT; and the result is used to calculate the point kinetics parameters (e.g., reactivity) by using their inner product definitions, which are then used to solve the time-dependent amplitude and precursor equations. The amplitude function is calculated by solving the kinetics equations using the LSODE solver. When a new shape calculation is needed, the flux is calculated using the newly computed amplitude function. The Boltzmann transport equation is solved using the method of discrete ordinates to treat the directional variable and weighted finite-difference methods, in addition to Linear Nodal
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
International Nuclear Information System (INIS)
Jahshan, S.N.; Wemple, C.A.; Ganapol, B.D.
1993-01-01
A comparison of the numerical solutions of the transport equation describing the steady neutron slowing down in an infinite medium with constant cross sections is made with stochastic solutions obtained from tracking successive neutron histories in the same medium. The transport equation solution is obtained using a numerical Laplace transform inversion algorithm. The basis for the algorithm is an evaluation of the Bromwich integral without analytical continuation. Neither the transport nor the stochastic solution is limited in the number of scattering species allowed. The medium may contain an absorption component as well. (orig.)
International Nuclear Information System (INIS)
Bareiss, E.H.
1976-05-01
The objectives of the work are to develop mathematically and computationally founded 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. As new analytical insights are gained, new numerical methods are developed and tested. Significant results obtained include implementation of the integer-preserving Gaussian elimination method (two-step method) in a CDC 6400 computer code, modes analysis for one-dimensional transport solutions, and a new method for solving the 1-T transport equation. Some of the work dealt with the interface and corner problem in diffusion theory
Advancing Transportation through Vehicle Electrification - PHEV
Energy Technology Data Exchange (ETDEWEB)
Bazzi, Abdullah [Chrysler Group LLC, Auburn Hills, MI (United States); Barnhart, Steven [Chrysler Group LLC, Auburn Hills, MI (United States)
2014-12-31
FCA US LLC viewed the American Recovery and Reinvestment Act (ARRA) as an historic opportunity to learn about and develop PHEV technologies and create the FCA US LLC engineering center for Electrified Powertrains. The ARRA funding supported FCA US LLC’s light-duty electric drive vehicle and charging infrastructure-testing activities and enabled FCA US LLC to utilize the funding on advancing Plug-in Hybrid Electric Vehicle (PHEV) technologies for production on future programs. FCA US LLC intended to develop the next-generations of electric drive and energy batteries through a properly paced convergence of standards, technology, components and common modules. To support the development of a strong, commercially viable supplier base, FCA US LLC also utilized this opportunity to evaluate various designated component and sub-system suppliers. The original proposal of this project was submitted in May 2009 and selected in August 2009. The project ended in December 2014.
Development of Cold Neutron Scattering Kernels for Advanced Moderators
International Nuclear Information System (INIS)
Granada, J. R.; Cantargi, F.
2010-01-01
The development of scattering kernels for a number of molecular systems was performed, including a set of hydrogeneous methylated aromatics such as toluene, mesitylene, and mixtures of those. In order to partially validate those new libraries, we compared predicted total cross sections with experimental data obtained in our laboratory. In addition, we have introduced a new model to describe the interaction of slow neutrons with solid methane in phase II (stable phase below T = 20.4 K, atmospheric pressure). Very recently, a new scattering kernel to describe the interaction of slow neutrons with solid Deuterium was also developed. The main dynamical characteristics of that system are contained in the formalism, the elastic processes involving coherent and incoherent contributions are fully described, as well as the spin-correlation effects.
Study on MPI/OpenMP hybrid parallelism for Monte Carlo neutron transport code
International Nuclear Information System (INIS)
Liang Jingang; Xu Qi; Wang Kan; Liu Shiwen
2013-01-01
Parallel programming with mixed mode of messages-passing and shared-memory has several advantages when used in Monte Carlo neutron transport code, such as fitting hardware of distributed-shared clusters, economizing memory demand of Monte Carlo transport, improving parallel performance, and so on. MPI/OpenMP hybrid parallelism was implemented based on a one dimension Monte Carlo neutron transport code. Some critical factors affecting the parallel performance were analyzed and solutions were proposed for several problems such as contention access, lock contention and false sharing. After optimization the code was tested finally. It is shown that the hybrid parallel code can reach good performance just as pure MPI parallel program, while it saves a lot of memory usage at the same time. Therefore hybrid parallel is efficient for achieving large-scale parallel of Monte Carlo neutron transport. (authors)
Design studies for a high-resolution, transportable neutron radiography/radioscopy system
International Nuclear Information System (INIS)
Gillespie, G.H.; Micklich, B.J.; McMichael, G.E.
1996-01-01
A preliminary design has been developed for a high-resolution, transportable neutron radiology system (TNRS) concept. The primary system requirement is taken to be a thermal neutron flux of 10[sup 6] n/(cm[sup 2]-sec) with a L/D ratio of 100. The approach is to use an accelerator-driven neutron source, with a radiofrequency quadrupole (RFQ) as the primary accelerator component. Initial concepts for all of the major components of the system have been developed,and selected key parts have been examined further. An overview of the system design is presented, together with brief summaries of the concepts for the ion source, low energy beam transport (LEBT), RFQ, high energy beam transport (HEBT), target, moderator, collimator, image collection, power, cooling, vacuum, structure, robotics, control system, data analysis, transport vehicle, and site support. The use of trade studies for optimizing the TNRS concept are also described
Advances in passive neutron instruments for safeguards use
International Nuclear Information System (INIS)
Menlove, H.O.; Krick, M.S.; Langner, D.G.; Miller, M.C.; Stewart, J.E.
1994-01-01
Passive neutron and other nondestructive assay techniques have been used extensively by the International Atomic Energy Agency to verify plutonium metal, powder, mixed oxide, pellets, rods, assemblies, scrap, and liquids. Normally, the coincidence counting rate is used to measure the 240 Pu-effective mass and gamma-ray spectrometry or mass spectrometry is used to verify the plutonium isotopic ratios. During the past few years, the passive neutron detectors have been installed in plants and operated in the unattended/continuous mode. These radiation data with time continuity have made it possible to use the totals counting rate to monitor the movement of nuclear material. Monte Carlo computer codes have been used to optimize the detector designs for specific applications. The inventory sample counter (INVS-III) has been designed to have a higher efficiency (43%) and a larger uniform counting volume than the original INVS. Data analyses techniques have been developed, including the ''known alpha'' and ''known multiplication'' methods that depend on the sample. For scrap and other impure or poorly characterized samples, we have developed multiplicity counting, initially implemented in the plutonium scrap multiplicity counter. For large waste containers such as 200-L drums, we have developed the add-a-source technique to give accurate corrections for the waste-matrix materials. This paper summarizes recent developments in the design and application of passive neutron assay systems
Thermal Evaluation of Storage Rack with an Advanced Neutron Absorber during Normal Operation
Energy Technology Data Exchange (ETDEWEB)
Lee, Hee-Jae; Kim, Mi-Jin; Sohn, Dong-Seong [UNIST, Ulsan (Korea, Republic of)
2016-10-15
The storage capacity of the domestic wet storage site is expected to reach saturation from Hanbit in 2024 to Sin-wolseong in 2038 and accordingly management alternatives are urgently taken. Since installation of the dense rack is considered in the short term, it is necessary to urgently develop an advanced neutron absorber which can be applied to a spent nuclear fuel storage facility. Neutron absorber is the material for controlling the reactivity. A material which has excellent thermal neutron absorption ability, high strength and corrosion resistance must be selected as the neutron absorber. Existing neutron absorbers are made of boron which has a good thermal absorption ability such as BORAL and METAMIC. However, possible problems have been reported in using the boron-based neutron absorber for wet storage facility. Gadolinium is known to have higher neutron absorption cross-section than that of boron. And the strength of duplex stainless steel is about 1.5 times higher than stainless steel 304 which has been frequently used as a structural material. Therefore, duplex stainless steel which contains gadolinium is in consideration as an advanced neutron absorber. Temperature distribution is shown in figure 4. In pool bottom region near the inlet shows a relatively low tendency and heat generated from the fuel assemblies is transmitted to the pool upper region by the vertical flow. Also, temperature gradient appear in rack structures for the axial direction and temperature is uniformly distributed in the pool upper region. Table 1 presents the calculated results. The maximum temperature is 306.63K and does not exceed the 333.15K (60℃). The maximum temperature of the neutron absorber is 306.48K.
Advances in the safe transport of irradiated Magnox fuel
International Nuclear Information System (INIS)
Jackson, C.N.
1997-01-01
This paper reviews the significant advances that have been made by Magnox Electric plc in ensuring that the Mk M2 Magnox flasks maintain the highest level of safety during transport and are used in the most efficient manner in meeting Company objectives. These advances have been achieved by improvement to the seal design, introduction of modern, state-of-the-art leak test equipment and optimisation of the generic Safety Case underpinning the UK Competent Authority (the Department of Transport, DoT) Approval Certificates. A step-by-step approach has been adopted in implementing these advances, consulting the DoT at each stage, to ensure that the safe transport of spent Magnox fuel, achieved over the past 35 years, with its enviable track record, continues into the next century. (Author)
Transport synthetic acceleration scheme for multi-dimensional neutron transport problems
Energy Technology Data Exchange (ETDEWEB)
Modak, R S; Kumar, Vinod; Menon, S V.G. [Theoretical Physics Div., Bhabha Atomic Research Centre, Mumbai (India); Gupta, Anurag [Reactor Physics Design Div., Bhabha Atomic Research Centre, Mumbai (India)
2005-09-15
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)
Discrete-ordinates electron transport calculations using standard neutron transport codes
International Nuclear Information System (INIS)
Morel, J.E.
1979-01-01
The primary purpose of this work was to develop a method for using standard neutron transport codes to perform electron transport calculations. The method is to develop approximate electron cross sections which are sufficiently well-behaved to be treated with standard S/sub n/ methods, but which nonetheless yield flux solutions which are very similar to the exact solutions. The main advantage of this approach is that, once the approximate cross sections are constructed, their multigroup Legendre expansion coefficients can be calculated and input to any standard S/sub n/ code. Discrete-ordinates calculations were performed to determine the accuracy of the flux solutions for problems corresponding to 1.0-MeV electrons incident upon slabs of aluminum and gold. All S/sub n/ calculations were compared with similar calculations performed with an electron Monte Carlo code, considered to be exact. In all cases, the discrete-ordinates solutions for integral flux quantities (i.e., scalar flux, energy deposition profiles, etc.) are generally in agreement with the Monte Carlo solutions to within approximately 5% or less. The central conclusion is that integral electron flux quantities can be efficiently and accurately calculated using standard S/sub n/ codes in conjunction with approximate cross sections. Furthermore, if group structures and approximate cross section construction are optimized, accurate differential flux energy spectra may also be obtainable without having to use an inordinately large number of energy groups. 1 figure
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)
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....
The advanced neutron source--designing to meet the needs of the user community
International Nuclear Information System (INIS)
Peretz, F.J.
1989-01-01
The Advanced Neutron Source (ANS) is to be a multi-purpose neutron research center, constructed around a high-flux reactor now being designed at the Oak Ridge National Laboratory (ORNL). Its primary purpose is to place the United States in the forefront of neutron scattering in the twenty-first century. Other research programs include nuclear and fundamental physics, isotopes production, materials irradiation, and analytical chemistry. The Advanced Neutron Source will be a unique and invaluable research tool because of the unprecedented neutron flux available from the high intensity research reactor. But that reactor would be ineffective without world-class research facilities that allow the fullest utilization of the available neutrons. And, in turn, those research facilities will not produce new and exciting science without a broad population of users coming from all parts of the nation, and the world, placed in a simulating environment in which experiments can be effectively conducted, and in which scientific exchange is encouraged. This paper discusses the measures being taken to ensure that the design of the ANS focuses not only on the reactor, but on providing the experiment and user support facilities needed to allow its effective use. 5 refs., 4 figs
How to polarise all neutrons in one beam: a high performance polariser and neutron transport system
Rodriguez, D. Martin; Bentley, P. M.; Pappas, C.
2016-09-01
Polarised neutron beams are used in disciplines as diverse as magnetism,soft matter or biology. However, most of these applications often suffer from low flux also because the existing neutron polarising methods imply the filtering of one of the spin states, with a transmission of 50% at maximum. With the purpose of using all neutrons that are usually discarded, we propose a system that splits them according to their polarisation, flips them to match the spin direction, and then focuses them at the sample. Monte Carlo (MC) simulations show that this is achievable over a wide wavelength range and with an outstanding performance at the price of a more divergent neutron beam at the sample position.
Performance of the advanced cold neutron source and optics upgrades at the NIST Research Reactor
International Nuclear Information System (INIS)
Williams, R.E.; Kopetka, P.; Cook, J.C.; Rowe, J.M.
2003-01-01
On March 6, 2002, the NIST Research Reactor resumed routine operation following a six-month shutdown for facility upgrades and maintenance. During the shutdown, the original liquid hydrogen cold neutron source was removed, and the advanced cold source was installed. An optical filter was installed on one of the neutron guides, NG-3, replacing a crystal filter for the 30-m SANS instrument and the guide used between the chopper disks of the Disk Chopper time-of-flight Spectrometer (DCS) installed on NG-4 has been recently reconfigured. Additional improvements in the neutron optics of various instruments are being made. The advanced liquid hydrogen cold neutron source performs as expected, nearly doubling the flux available to most instruments. The measured gains range from about 1.4 at 2 A, to over a factor of two at 15 A. Also as expected, the heat load in the new source increased to 1200 watts, but the previously existing refrigerator has easily accommodated the increase. With intensity gains of a factor of two in the important long wavelength region of the spectrum, the advanced cold source significantly enhances the measurement capability of the cold neutron scattering instrumentation at NIST. The optical filter on NG-3 is also very successful; the 30-m SANS has an additional gain of two at 17 A. A system of refracting lenses and prisms near the SANS sample position has made possible measurements at low Q (0.0005 A -1 ) that were previously not feasible. The DCS has also seen additional intensity gain factors in excess of two for the majority of experiments and at short neutron wavelengths the gains exceed three. In addition, two new triple axis spectrometers will feature double-focusing monochromators in order to exploit the full size of the available thermal and cold neutron beam tubes. The success of the advanced cold source and enhanced neutron optics contributed to the recognition of the NIST Center for Neutron Research as 'the premiere neutron scattering
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
Abbott, B. P.; Abbott, R.; Abbott, T. D.; Abernathy, M. R.; Acernese, E.; Ackley, K.; Adams, C.; Phythian-Adams, A.T.; Addesso, P.; Adhikari, R. X.; Adya, V. B.; Affeldt, C.; Agathos, M.; Agatsuma, K.; Aggarwal, N.T.; Aguiar, O. D.; Aiello, L.; Ain, A.; Ajith, P.; Allen, B.; Allocca, A.; Altin, P. A.; Anderson, S. B.; Anderson, W. G.; Arai, K.; Araya, M. C.; Arceneaux, C. C.; Areeda, J. S.; Arnaud, N.; Arun, K. G.; Ascenzi, S.; Ashton, G.; Ast, M.; Aston, S. M.; Astone, P.; Aufmuth, P.; Aulbert, C.; Babak, S.; Bacon, P.; Bader, M. K. M.; Baker, P. T.; Baldaccini, F.; Ballardin, G.; Ballmer, S. W.; Barayoga, J. C.; Barclay, S. E.; Barish, B. C.; Barker, R.D.; Barone, E.; Barr, B.; Barsotti, L.; Barsuglia, M.; Barta, D.; Bartlett, J.; Bartos, I.; Bassiri, R.; Basti, A.; Batch, J. C.; Baune, C.; Bavigadda, V.; Bazzan, M.; Bejger, M.; Bell, A. S.; Berger, B. K.; Bergmann, G.; Berry, C. P. L.; Bersanetti, D.; Bertolini, A.; Betzwieser, J.; Bhagwat, S.; Bhandare, R.; Bilenko, I. A.; Billingsley, G.; Birch, M.J.; Birney, R.; Biscans, S.; Bisht, A.; Bitossi, M.; Biwer, C.; Bizouard, M. A.; Blackburn, J. K.; Blair, C. D.; Blair, D. G.; Blair, R. M.; Bloemen, A.L.S.; Bock, O.; Boer, M.; Bogaert, J.G.; Bogan, C.; Bohe, A.; Bond, T.C; Bondu, F.; Bonnand, R.; Boom, B. A.; Bork, R.; Boschi, V.; Bose, S.; Bouffanais, Y.; Bozzi, A.; Bradaschia, C.; Brady, P. R.; Braginsky, V. B.; Branchesi, M.; Brau, J. E.; Briant, T.; Brillet, A.; Brinkmann, M.; Brisson, V.; Brockill, P.; Broida, J. E.; Brooks, A. F.; Brown, A.D.; Brown, D.; Brown, N. M.; Brunett, S.; Buchanan, C. C.; Buikema, A.; Bulik, T.; Bulten, H. J.; Buonanno, A.; Buskulic, D.; Buy, C.; Byer, R. L.; Cabero, M.; Cadonati, L.; Cagnoli, G.; Cahillane, C.; Calderon Bustillo, J.; Callister, T. A.; Calloni, E.; Camp, J. B.; Cannon, K. C.; Cao, J.; Capano, C. D.; Capocasa, E.; Carbognani, F.; Caride, S.; Casanueva Diaz, J.; Casentini, C.; Caudill, S.; Cavaglia, M.; Cavalier, F.; Cavalieri, R.; Cella, G.; Cepeda, C. B.; Cerboni Baiardi, L.; Cerretani, G.; Cesarini, E.; Chamberlin, S. J.; Chan, M.; Chao, D. S.; Chen, Y; Chassande-Mottin, E.; Cheeseboro, B. D.; Chen, H. Y.; Chen, Y; Cheng, C.; Chincarini, A.; Chiummo, A.; Cho, H. S.; Cho, M.; Chow, J. H.; Christensen, N.; Chu, Qian; Chua, S. E.; Chung, E.S.; Ciani, G.; Clara, E.; Clark, J. A.; Cleva, E.; Coccia, E.; Cohadon, P. -E; Colla, A.; Collette, C. G.; Cominsky, L.; Constancio, M., Jr.; Conte, A.; Conti, L.; Cook, D.; Corbitt, T. R.; Cornish, N.; Corsi, A.; Cortese, S.; Costa, A.C.; Coughlin, M. W.; Coughlin, S. B.; Coulon, J. -P.; Countryman, S. T.; Couvares, P.; Cowan, E. E.; Coward, D. M.; Cowart, M. J.; Coyne, D. C.; Coyne, R.; Craig, K.; Creighton, J. D. E.; Cripe, J.; Crowder, S. G.; Cumming, A.; Cunningham, A.L.; Cuoco, E.; Dal Canton, T.; Danilishin, S. L.; D'Antonio, S.; Danzmann, K.; Darman, N. S.; Dasgupta, A.; Da Silva Costa, C. F.; Dattilo, V.; Dave, I.; Davier, M.; Davies, G. S.; Daw, E. J.; Day, R.; De, S.; Debra, D.; Debreczeni, G.; Degallaix, J.; De laurentis, M.; Deleglise, S.; Del Pozzo, W.; Denker, T.; Dent, T.; Dergachev, V.A.; Rosa, R.; DeRosa, R. T.; DeSalvo, R.; Devine, R. C.; Dhurandhar, S.; Diaz, M. C.; Di Fiore, L.; Giovanni, M.G.; Di Girolamo, T.; Di Lieto, A.; Di Pace, S.; Di Palma, I.; Di Virgilio, A.; Dolique, V.; Donovan, F.; Dooley, K. L.; Doravari, S.; Douglas, R.; Downes, T. P.; Drago, M.; Drever, R. W. P.; Driggers, J. C.; Ducrot, M.; Dwyer, S. E.; Edo, T. B.; Edwards, M. C.; Effler, A.; Eggenstein, H. -B.; Ehrens, P.; Eichholz, J.; Eikenberry, S. S.; Engels, W.; Essick, R. C.; Etzel, T.; Evans, T. M.; Evans, T. M.; Everett, R.; Factourovich, M.; Fafone, V.; Fair, H.; Fairhurst, S.; Fan, X.M.; Fang, Q.; Farinon, S.; Farr, B.; Farr, W. M.; Favata, M.; Fays, M.; Fehrmann, H.; Fejer, M. M.; Fenyvesi, E.; Ferrante, I.; Ferreira, E. C.; Ferrini, F.; Fidecaro, F.; Fiori, I.; Fiorucci, D.; Fisher, R. P.; Flaminio, R.; Fletcher, M; Fournier, J. -D.; Frasca, S.; Frasconi, F.; Frei, Z.; Freise, A.; Frey, R.; Frey, V.; Fritschel, P.; Frolov, V. V.; Fulda, P.; Fyffe, M.; Gabbard, H. A. G.; Gair, J. R.; Gammaitoni, L.; Gaonkar, S. G.; Garunfi, E.; Gaur, G.; Gehrels, N.; Gemme, G.; Geng, P.; Genin, E.; Gennai, A.; George, J.; Gergely, L.; Germain, V.; Ghosh, Abhirup; Ghosh, Archisman; Ghosh, S.; Giaime, J. A.; Giardina, K. D.; Giazotto, A.; Gill, K.P.; Glaefke, A.; Goetz, E.; Goetz, R.; Gondan, L.; Gonzalez, R.G.; Gonzalez Castro, J. M.; Gopakumar, A.; Gordon, N. A.; Gorodetsky, M. L.; Gossan, S. E.; Lee-Gosselin, M.; Gouaty, R.; Grado, A.; Graef, C.; Graff, P. B.; Granata, M.; Grant, A.; Gras, S.; Gray, C.M.; Greco, G.; Green, A. C.; Groot, P.; Grote, H.; Grunewald, S.; Guidi, G. M.; Guo, X.; Gupta, A.; Gupta, M. K.; Gushwa, K. E.; Gustafson, E. K.; Gustafson, R.; Hacker, J. J.; Buffoni-Hall, R.; Hall, E. D.; Hammond, G.L.; Haney, M.; Hanke, M. M.; Hanks, J.; Hanna, C.; Hannam, M. D.; Hanson, P.J.; Hardwick, T.; Harms, J.; Harry, G. M.; Harry, I. W.; Hart, M. J.; Hartman, M. T.; Haster, C. -J.; Haughian, K.; Heidmann, A.; Heintze, M. C.; Heitmann, H.; Hello, P.; Hemming, G.; Hendry, M.; Heng, I. S.; Hennig, J.; Henry, J.A.; Heptonstall, A. W.; Heurs, M.; Hild, S.; Hoak, D.; Hofman, D.; Holt, K.; Holz, D. E.; Hopkins, P.; Hough, J.; Houston, E. A.; Howell, E. J.; Hu, Y. M.; Huang, S.; Huerta, E. A.; Huet, D.; Hughey, B.; Husa, S.; Huttner, S. H.; Huynh-Dinh, T.; Indik, N.; Ingram, D. R.; Inta, R.; Isa, H. N.; Isac, J. -M.; Isi, M.; Isogai, T.; Iyer, B. R.; Izumi, K.; Jacqmin, T.; Jang, D.H.; Jani, K.; Jaranowski, P.; Jawahar, S.; Jian, L.; Jimenez-Forteza, E.; Johnson, W.; Jones, I.D.; Jones, R.; Jonker, R. J. G.; Ju, L.; Haris, K.; Kalaghatgi, C. V.; Kalogera, V.; Kandhasamy, S.; Kang, G.H.; Kanner, J. B.; Kapadia, S. J.; Karki, S.; Karvinen, K. S.; Kasprzack, M.; Katsavounidis, E.; Katzman, W.; Kaufer, S.; Kaur, T.; Kawabe, K.; Kefelian, F.; Kehl, M. S.; Keitel, D.; Kelley, D. B.; Kells, W.; Kennedy, R.E.; Key, J. S.; Khalili, F. Y.; Khan, I.; Khan., S.; Khan, Z.; Khazanov, E. A.; Kusunchoo, N.; Kim, Chi-Woong; Kim, Chunglee; Kim, J.; Kim, K.; Kim, Namjun; Kim, W.; Kim, Y.M.; Kimbrell, S. J.; King, E. J.; King, P. J.; Kissel, J. S.; Klein, B.; Kleybolte, L.; Klimenko, S.; Koehlenbeck, S. M.; Koley, S.; Kondrashov, V.; Kontos, A.; Korobko, M.; Korth, W. Z.; Kowalska, I.; Kozak, D. B.; Kringel, V.; Krishnan, B.; Krolak, A.; Krueger, C.; Kuehn, G.; Kumar, P.; Kumar, R.; Kuo, L.; Kutynia, A.; Lackey, B. D.; Landry, M.; Lange, J.; Lantz, B.; Lasky, P. D.; Laxen, M.; Lazzarini, A.; Lazzar, C.; Leaci, P.; Leavey, S.; Lebigot, E. O.; Lee, C.H.; Lee, K.H.; Lee, M.H.; Lee, K.; Lenon, A.; Leonardi, M.; Leong, J. R.; Leroy, N.; Letendre, N.; Levin, Y.; Lewis, J. B.; Li, T. G. F.; Libson, A.; Littenberg, T. B.; Lockerbie, N. A.; Lombardi, A. L.; London, L. T.; Lord, J. E.; Lorenzini, M.; Loriette, V.; Lormand, M.; Losurdo, G.; Lough, J. D.; Luck, H.; Lundgren, A. P.; Lynch, R.; Ma, Y.; Machenschalk, B.; MacInnis, M.; Macleod, D. M.; Magana-Sandoval, F.; Zertuche, L. Magana; Magee, R. M.; Majorana, E.; Maksimovic, I.; Malvezzi, V.; Man, N.; Mandic, V.; Mangano, V.; Mansell, G. L.; Manske, M.; Mantovani, M.; Marchesoni, E.; Marion, F.; Marka, S.; Marka, Z.; Markosyan, A. S.; Maros, E.; Martelli, E.; Martellini, L.; Martin, I. W.; Martynov, D. V.; Marx, J. N.; Mason, K.; Masserot, A.; Massinger, T. J.; Masso-Reid, M.; Mastrogiovanni, S.; Matichard, F.; Matone, L.; Mavalvala, N.; Mazumder, N.; McCarthy, R.; McClelland, D. E.; McCormick, S.; McGuire, S. C.; McIntyre, G.; McIver, J.; McManus, D. J.; McRae, T.; McWilliams, S. T.; Meacher, D.; Meadors, G. D.; Meidam, J.; Melatos, A.; Mendell, G.; Mercer, R. A.; Merilh, E. L.; Merzougui, M.; Meshkov, S.; Messenger, C.; Messick, C.; Metzdorff, R.; Meyers, P. M.; Mezzani, F.; Miao, H.; Michel, C.; Middleton, H.; Mikhailov, E. E.; Milano, L.; Miller, A. L.; Miller, A. L.; Miller, B.; Miller, J.; Millhouse, M.; Minenkov, Y.; Ming, J.; Mirshekari, S.; Mishra, C.; Mitra, S.; Mitrofanov, V. P.; Mitselmakher, G.; Mittleman, R.; Moggi, A.; Mohan, M.; Mohapatra, S. R. P.; Montani, M.; Moore, B.C.; Moore, J.C.; Moraru, D.; Gutierrez Moreno, M.; Morriss, S. R.; Mossavi, K.; Mours, B.; Mow-Lowry, C. M.; Mueller, G.; Muir, A. W.; Mukherjee, Arunava; Mukherjee, S.D.; Mukherjee, S.; Mukund, N.; Mullavey, A.; Munch, J.; Murphy, D. J.; Murray, P.G.; Mytidis, A.; Nardecchia, I.; Naticchioni, L.; Nayak, R. K.; Nedkova, K.; Nelemans, G.; Nelson, T. J. N.; Gutierrez-Neri, M.; Neunzert, A.; Newton-Howes, G.; Nguyen, T. T.; Nielsen, A. B.; Nissanke, S.; Nitz, A.; Nocera, F.; Nolting, D.; Normandin, M. E. N.; Nuttall, L. K.; Oberling, J.; Ochsner, E.; O'Deill, J.; Oelker, E.; Ogin, G. H.; Oh, J.; Oh, S. H.; Ohme, F.; Oliver, M. B.; Oppermann, P.; Oram, Richard J.; O'Reilly, B.; O'Shaughnessy, R.; Ottaway, D. J.; Overmier, H.; Owen, B. J.; Pai, A.; Pai, S. A.; Palamos, J. R.; Palashov, O.; Palomba, C.; Pal-Singh, A.; Pan, H.; Pankow, C.; Pannarale, F.; Pant, B. C.; Paoletti, E.; Paoli, A.; Papa, M. A.; Paris, H. R.; Parker, W.S; Pascucci, D.; Pasqualetti, A.; Passahieti, R.; Passuello, D.; Patricelli, B.; Patrick, Z.; Pearlstone, B. L.; Pedraza, M.; Pedurand, R.; Pekowsky, L.; Pele, A.; Penn, S.; Perreca, A.; Perri, L. M.; Phelps, M.; Piccinni, O. J.; Pichot, M.; Piergiovanni, F.; Pierro, V.; Pillant, G.; Pinard, L.; Pinto, I. M.; Pitkin, M.; Poe, M.; Poggiani, R.; Popolizio, P.; Post, A.; Powell, J.; Prasad, J.; Predoi, V.; Prestegard, T.; Price, L. R.; Prijatelj, M.; Principe, M.; Privitera, S.; Prix, R.; Prodi, G. A.; Proxhorov, L.; Puncken, O.; Punturo, M.; Puppo, P.; Purrer, M.; Qi, H.; Qin, J.; Qiu, S.; Quetschke, V.; Quintero, E. A.; Quitzow-James, R.; Raab, E. J.; Rabeling, D. S.; Radkins, H.; Raffai, P.; Raja, S.; Rajan, C.; Rakhmanov, M.; Rapagnani, P.; Raymond, V.; Razzano, M.; Re, V.; Read, J.; Reed, C. M.; Regimbau, T.; Rei, L.; Reid, S.; Reitze, D. H.; Rew, H.; Reyes, S. D.; Ricci, F.; Riles, K.; Rizzo, D.M.; Robertson, N. A.; Robie, R.; Robinet, F.; Rocchi, A.; Rolland, L.; Rollins, J. G.; Roma, V. J.; Romano, R.; Romanov, G.; Romie, J. H.; Rosinska, D.; Rowan, S.; Rudiger, A.; Ruggi, P.; Ryan, K.A.; Sachdev, P.S.; Sadecki, T.; Sadeghian, L.; Sakellariadou, M.; Salconi, L.; Saleem, M.; Salemi, F.; Samajdar, A.; Sammut, L.; Sanchez, E. J.; Sandberg, V.; Sandeen, B.; Sanders, J. R.; Sassolas, B.; Sathyaprakash, B. S.; Saulson, P. R.; Sauter, O. E. S.; Savage, R. L.; Sawadsky, A.; Schale, P.; Schilling, R.; Schmidt, J; Schmidt, P.; Schnabel, R.B.; Schofield, R. M. S.; Schonbecx, A.; Schreiber, K.E.C.; Schuette, D.; Schutz, B. F.; Scott, J.; Scott, M.S.; Sellers, D.; Sengupta, A. S.; Sentenac, D.; Sequino, V.; Sergeev, A.; Setyawati, Y.; Shaddock, D. A.; Shaffer, T. J.; Shahriar, M. S.; Shaltevi, M.; Shapiro, B.; Shawhan, P.; Sheperd, A.; Shoemaker, D. H.; Shoemaker, D. M.; Siellez, K.; Siemens, X.; Sieniawska, M.; Sigg, D.; Silva, António Dias da; Singer, A; Singer, L. P.; Singh, A.; Singh, R.; Singhal, A.; Sintes, A. M.; Slagmolen, B. J. J.; Smith, R. J. E.; Smith, N.D.; Smith, R. J. E.; Son, E. J.; Sorazu, B.; Sorrentino, F.; Souradeep, T.; Srivastava, A. K.; Staley, A.; Steinke, M.; Steinlechner, J.; Steinlechner, S.; Steinmeyer, D.; Stephens, B. C.; Stone, J.R.; Strain, K. A.; Straniero, N.; Stratta, G.; Strauss, N. A.; Strigin, S. E.; Sturani, R.; Stuver, A. L.; Summerscales, T. Z.; Sun, L.; Sunil, S.; Sutton, P. J.; Swinkels, B. L.; Szczepanczyk, M. J.; Tacca, M.D.; Talukder, D.; Tanner, D. B.; Tapai, M.; Tarabrin, S. P.; Taracchini, A.; Taylor, W.R.; Theeg, T.; Thirugnanasambandam, M. P.; Thomas, E. G.; Thomas, M.; Thomas, P.; Thorne, K. A.; Thrane, E.; Tiwari, S.; Tiwari, V.; Toxmakov, K. V.; Toland, K.; Tomlinson, C.; Tonelli, M.; Tornasi, Z.; Torres, C. V.; Torrie, C. I.; Toyra, D.; Travasso, F.; Traylor, G.; Trifiro, D.; Tringali, M. C.; Trozzo, L.; Tse, M.; Turconi, M.; Tuyenbayev, D.; Ugolini, D.; Unnikrishnan, C. S.; Urban, A. L.; Usman, S. A.; Vahlbruch, H.; Vajente, G.; Valdes, G.; van Bakel, N.; van Beuzekom, M.G.; van den Brand, J. E. J.; Van Den Broeck, C.F.F.; Vander-Hyde, D. C.; van der Schaaf, L.; van Heuningen, J. V.; van Veggel, A. A.; Vardaro, M.; Vass, S.; Vasuth, M.; Vaulin, R.; Vecchio, A.; Vedovato, G.; Veitch, J.; Veitch, P.J.; Venkateswara, K.; Verkindt, D.; Vetrano, E.; Vicere, A.; Vinciguerra, S.; Vine, D. J.; Vinet, J. -Y.; Vitale, S.; Vo, T.; Vocca, H.; Vorvick, C.; Voss, D. V.; Vousden, W. D.; Vyatchanin, S. P.; Wade, A. R.; Wade, L. E.; Wade, MT; Walker, M.; Wallace, L.; Walsh, S.; Wang, G.; Wang, H.; Wang, M.; Wang, X.; Wang, Y.; Ward, R. L.; Warner, J.; Was, M.; Weaver, B.; Wei, L. -W.; Weinert, M.; Weinstein, A. J.; Weiss, R.; Wen, L.M.; Wessels, P.; Westphal, T.; Wette, K.; Whelan, J. T.; Whiting, B. F.; Williams, D.R.; Williamson, A. R.; Willis, J. L.; Willke, B.; Wimmer, M. H.; Winkler, W.; Wipf, C. C.; Wittel, H.; Woan, G.; Woehler, J.; Worden, J.; Wright, J.L.; Wu, D.S.; Wu, G.; Yablong, J.; Yam, W.; Yamamoto, H.; Yancey, C. C.; Yu, H.; Yvert, M.; Zadrozny, A.; Zangrando, L.; Zanolin, M.; Zendri, J. -P.; Zevin, M.; Zhang, L.; Zhang, M.; Zhang, Y.; Zhao, C.; Zhou, M.; Zhou, Z.; Zhu, X. J.; Zucker, M. E.; Zuraw, S. E.; Zweizig, J.
2016-01-01
We report here the non-detection of gravitational waves from the merger of binary-neutron star systems and neutron star-black hole systems during the first observing run of the Advanced Laser Interferometer Gravitational-wave Observatory (LIGO). In particular, we searched for gravitational-wave
Energy Technology Data Exchange (ETDEWEB)
Souza, S N.A. de
1991-05-01
This work presents a study a transportable neutron radiography system for a 185 GBq {sup 241} Am-Be ({alpha}, {eta}) source with a neutron yield roughly 1,25 x 10{sup 7} n/s. Studies about moderation, collimation and shielding are showed. In these studies, a calculation using Transport Theory was carried out by means of transport codes ANISN and DOT (3.5). Objectives were: to obtain a maximum and more homogeneous thermal neutron flux in the collimator outlet 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 of 14, for neutron fluxes up to 4,09 x 10{sup 2} n.cm{sup -2}.s{sup -1}. Considering the low intensity of the source, it is a good value. Studies have also been carried out for L/D ratios of 22 and 30, giving thermal neutron fluxes at the image plain of 1,27 x 10{sup 2} n.cm{sup -2}.s{sup -1} and 2,65 x 10{sup 2} n.cm{sup -2}.s{sup -1}, respectively. (author). 30 refs, 39 figs, 9 tabs.
International Nuclear Information System (INIS)
Jewer, S.; Buchan, A.G.; Pain, C.C.; Cacuci, D.G.
2014-01-01
Highlights: • A new method of coupled radiation transport, heat and momentum exchanges on fluids, and heat transfer simulations. • Simulation of the thermal hydraulics and radiative properties within whole PWR assemblies. • An immersed body method for modelling complex solid domains on practical computational meshes. - Abstract: A recently developed immersed body method is adapted and used to model a typical pressurised water reactor (PWR) fuel assembly. The approach is implemented with the numerical framework of the finite element, transient criticality code, FETCH which is composed of the neutron transport code, EVENT, and the CFD code, FLUIDITY. Within this framework the neutron transport equation, Navier–Stokes equations and a fluid energy conservation equation are solved in a coupled manner on a coincident structured or unstructured mesh. The immersed body method has been used to model the solid fuel pins. The key feature of this method is that the fluid/neutronic domain and the solid domain are represented by overlapping and non-conforming meshes. The main difficulty of this approach, for which a solution is proposed in this work, is the conservative mapping of the energy and momentum exchange between the fluid/neutronic mesh and the solid fuel pin mesh. Three numerical examples are presented which include a validation of the fuel pin submodel against an analytical solution; an uncoupled (no neutron transport solution) PWR fuel assembly model with a specified power distribution which was validated against the COBRA-EN subchannel analysis code; and finally a coupled model of a PWR fuel assembly with reflective neutron boundary conditions. Coupling between the fluid and neutron transport solutions is through the nuclear cross sections dependence on Doppler fuel temperature, coolant density and temperature, which was taken into account by using pre-calculated cross-section lookup tables generated using WIMS9a. The method was found to show good agreement
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.
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.)
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.
The Use of Advanced Transportation Monitoring Data for Official Statistics
Y. Ma (Yinyi)
2016-01-01
markdownabstractTraﬃc and transportation statistics are mainly published as aggregated information, and are traditionally based on surveys or secondary data sources, like public registers and companies’ administrations. Nowadays, advanced monitoring systems are installed in the road network, oﬀering
Maucec, M.; Rigollet, C.
The performance of a detection system based on the pulsed fast/thermal neutron analysis technique was assessed using Monte Carlo simulations. The aim was to develop and implement simulation methods, to support and advance the data analysis techniques of the characteristic gamma-ray spectra,
Advances in thermal hydraulic and neutronic simulation for reactor analysis and safety
International Nuclear Information System (INIS)
Tentner, A.M.; Blomquist, R.N.; Canfield, T.R.; Ewing, T.F.; Garner, P.L.; Gelbard, E.M.; Gross, K.C.; Minkoff, M.; Valentin, R.A.
1993-01-01
This paper describes several large-scale computational models developed at Argonne National Laboratory for the simulation and analysis of thermal-hydraulic and neutronic events in nuclear reactors and nuclear power plants. The impact of advanced parallel computing technologies on these computational models is emphasized
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)
The spectral element approach for the solution of neutron transport problems
International Nuclear Information System (INIS)
Barbarino, A.; Dulla, S.; Ravetto, P.; Mund, E.H.
2011-01-01
In this paper a possible application of the Spectral Element Method to neutron transport problems is presented. The basic features of the numerical scheme on the one-dimensional diffusion equation are illustrated. Then, the AN model for neutron transport is introduced, and the basic steps for the construction of a bi-dimensional solver are described. The AN equations are chosen for their structure, involving a system of coupled elliptic-type equations. Some calculations are carried out on typical benchmark problems and results are compared with the Finite Element Method, in order to evaluate their performances. (author)
Multigroup neutron transport equation in the diffusion and P{sub 1} approximation
Energy Technology Data Exchange (ETDEWEB)
Obradovic, D [Boris Kidric Institute of nuclear sciences Vinca, Belgrade (Yugoslavia)
1970-07-01
Investigations of the properties of the multigroup transport operator, width and without delayed neutrons in the diffusion and P{sub 1} approximation, is performed using Keldis's theory of operator families as well as a technique . recently used for investigations into the properties of the general linearized Boltzmann operator. It is shown that in the case without delayed neutrons, multigroup transport operator in the diffusion and P{sub 1} approximation possesses a complete set of generalized eigenvectors. A formal solution to the initial value problem is also given. (author)
International Nuclear Information System (INIS)
Tewari, S.P.
1975-01-01
A method of studying neutron transport properties in H 2 O-D 2 O mixtures, both liquid and solid, which extrapolates the neutron thermalization parameters of H 2 O is described. The decay of pulsed neutrons in the media has been investigated as an example of the application of the method. The results of the calcutions agree with the experiment for concentrations up to 50 percent D 2 O. (1 figure) (U.S.)
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
International Nuclear Information System (INIS)
Devillers, C.
1973-01-01
1 - Nature of physical problem solved: The ANISN system treats neutron and gamma transport in one-dimensional plane, spherical and cylinder geometry. The multigroup cross sections prepared by the programs LIANE and SUPERTOG are processed by the program RETTOG, which produces a binary library with Legendre expansions. The binary library can be updated and edited with the program LGR/B. The photon multigroup cross sections are created with the program GAMLEG/A. If the bulk of the data is too large, the program TAPEMA produces a special group-by-group library. The volume sources are calculated from a reduced set of input data and punched in a format suitable for input to ANISN, using the program PRESOU. The program ANISN calculates fluxes by groups, space intervals, angle and any number of reaction rates. The energy and space dependent fluxes are stored on tape and can be reprocessed, edited and plotted with the program ANISEX, which also permits to calculate supplementary reaction rates. The program ANISN can condense cross sections into a reduced number of groups. The ANISN system is used as a reference system for the evaluation of approximation methods (space-diffusion or point- kernel) or for the preparation of multigroup libraries for 2- dimensional transport codes (DOT). In particular it is used for shielding problems with high attenuation in water reactors and fast reactors. 2 - Method of solution: Method of discrete ordinates. The program has been designed to treat deep penetration with detailed calculation of spectrum as function of angle. Tests for pointwise convergence have also been introduced. 3 - Restrictions on the complexity of the problem: The complexity of the problem is limited by the storage size
Effects of fuel particle size distributions on neutron transport in stochastic media
International Nuclear Information System (INIS)
Liang, Chao; Pavlou, Andrew T.; Ji, Wei
2014-01-01
Highlights: • Effects of fuel particle size distributions on neutron transport are evaluated. • Neutron channeling is identified as the fundamental reason for the effects. • The effects are noticeable at low packing and low optical thickness systems. • Unit cells of realistic reactor designs are studied for different size particles. • Fuel particle size distribution effects are not negligible in realistic designs. - Abstract: This paper presents a study of the fuel particle size distribution effects on neutron transport in three-dimensional stochastic media. Particle fuel is used in gas-cooled nuclear reactor designs and innovative light water reactor designs loaded with accident tolerant fuel. Due to the design requirements and fuel fabrication limits, the size of fuel particles may not be perfectly constant but instead follows a certain distribution. This brings a fundamental question to the radiation transport computation community: how does the fuel particle size distribution affect the neutron transport in particle fuel systems? To answer this question, size distribution effects and their physical interpretations are investigated by performing a series of neutron transport simulations at different fuel particle size distributions. An eigenvalue problem is simulated in a cylindrical container consisting of fissile fuel particles with five different size distributions: constant, uniform, power, exponential and Gaussian. A total of 15 parametric cases are constructed by altering the fissile particle volume packing fraction and its optical thickness, but keeping the mean chord length of the spherical fuel particle the same at different size distributions. The tallied effective multiplication factor (k eff ) and the spatial distribution of fission power density along axial and radial directions are compared between different size distributions. At low packing fraction and low optical thickness, the size distribution shows a noticeable effect on neutron
Neutron Transport Methods for Accelerator-Driven Systems
International Nuclear Information System (INIS)
Nicholas Tsoulfanidis; Elmer Lewis
2005-01-01
The objective of this project has been to develop computational methods that will enable more effective analysis of Accelerator Driven Systems (ADS). The work is centered at the University of Missouri at Rolla, with a subcontract at Northwestern University, and close cooperation with the Nuclear Engineering Division at Argonne National Laboratory. The work has fallen into three categories. First, the treatment of the source for neutrons originating from the spallation target which drives the neutronics calculations of the ADS. Second, the generalization of the nodal variational method to treat the R-Z geometry configurations frequently needed for scoping calculations in Accelerator Driven Systems. Third, the treatment of void regions within variational nodal methods as needed to treat the accelerator beam tube
Thermal-hydraulic studies of the Advanced Neutron Source cold source
International Nuclear Information System (INIS)
Williams, P.T.; Lucas, A.T.
1995-08-01
The Advanced Neutron Source (ANS), in its conceptual design phase at Oak Ridge National Laboratory, was to be a user-oriented neutron research facility producing the most intense steady-state flux of thermal and cold neutrons in the world. Among its many scientific applications, the production of cold neutrons was a significant research mission for the ANS. The cold neutrons come from two independent cold sources positioned near the reactor core. Contained by an aluminum alloy vessel, each cold source is a 410-mm-diam sphere of liquid deuterium that functions both as a neutron moderator and a cryogenic coolant. With nuclear heating of the containment vessel and internal baffling, steady-state operation requires close control of the liquid deuterium flow near the vessel's inner surface. Preliminary thermal-hydraulic analyses supporting the cold source design were performed with heat conduction simulations of the vessel walls and multidimensional computational fluid dynamics simulations of the liquid deuterium flow and heat transfer. This report presents the starting phase of a challenging program and describes the cold source conceptual design, the thermal-hydraulic feasibility studies of the containment vessel, and the future computational and experimental studies that were planned to verify the final design
Advanced neutron imaging methods with a potential to benefit from pulsed sources
International Nuclear Information System (INIS)
Strobl, M.; Kardjilov, N.; Hilger, A.; Penumadu, D.; Manke, I.
2011-01-01
During the last decade neutron imaging has seen significant improvements in instrumentation, detection and spatial resolution. Additionally, a variety of new applications and methods have been explored. As a consequence of an outstanding development nowadays various techniques of neutron imaging go far beyond a two- and three-dimensional mapping of the attenuation coefficients for a broad range of samples. Neutron imaging has become sensitive to neutron scattering in the small angle scattering range as well as with respect to Bragg scattering. Corresponding methods potentially provide spatially resolved and volumetric data revealing microstructural inhomogeneities, texture variations, crystalline phase distributions and even strains in bulk samples. Other techniques allow for the detection of refractive index distribution through phase sensitive measurements and the utilization of polarized neutrons enables radiographic and tomographic investigations of magnetic fields and properties as well as electrical currents within massive samples. All these advanced methods utilize or depend on wavelength dependent signals, and are hence suited to profit significantly from pulsed neutron sources as will be discussed.
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
PELAN - a transportable, neutron-based UXO identification technique
International Nuclear Information System (INIS)
Vourvopoulos, G.
1998-01-01
An elemental characterization method is used to differentiate between inert projectiles and UXO's. This method identifies in a non-intrusive, nondestructive manner, the elemental composition of the projectile contents. Most major and minor chemical elements within the interrogated object (hydrogen, carbon, nitrogen, oxygen, fluorine, phosphorus, chlorine, arsenic, etc.) are identified and quantified. The method is based on PELAN - Pulsed Elemental Analysis with Neutrons. PELAN uses pulsed neutrons produced from a compact, sealed tube neutron generator. Using an automatic analysis computer program, the quantities of each major and minor chemical element are determined. A decision-making tree identifies the object by comparing its elemental composition with stored elemental composition libraries of substances that could be contained within the projectile. In a series of blind tests, PELAN was able to identify without failure, the contents of each shell placed in front of it. The PELAN probe does not need to be in contact with the interrogated projectile. If the object is buried, the interrogation can take place in situ provided the probe can be inserted a few centimeters from the object's surface. (author)
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)
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
Advanced Engineering Environments for Space Transportation System Development
Thomas, L. Dale; Smith, Charles A.; Beveridge, James
2000-01-01
There are significant challenges facing today's launch vehicle industry. Global competition, more complex products, geographically-distributed design teams, demands for lower cost, higher reliability and safer vehicles, and the need to incorporate the latest technologies quicker, all face the developer of a space transportation system. Within NASA, multiple technology development and demonstration projects are underway toward the objectives of safe, reliable, and affordable access to space. New information technologies offer promising opportunities to develop advanced engineering environments to meet these challenges. Significant advances in the state-of-the-art of aerospace engineering practice are envisioned in the areas of engineering design and analytical tools, cost and risk tools, collaborative engineering, and high-fidelity simulations early in the development cycle. At the Marshall Space Flight Center, work has begun on development of an advanced engineering environment specifically to support the design, modeling, and analysis of space transportation systems. This paper will give an overview of the challenges of developing space transportation systems in today's environment and subsequently discuss the advanced engineering environment and its anticipated benefits.
Energy Technology Data Exchange (ETDEWEB)
Corcoran, E.C. [Department of Chemistry and Chemical Engineering, Royal Military College of Canada, P.O. Box 17000, St. Forces, Kingston, Ont., K7K 7B4 (Canada)], E-mail: emily.corcoran@rmc.ca; Lewis, B.J.; Thompson, W.T. [Department of Chemistry and Chemical Engineering, Royal Military College of Canada, P.O. Box 17000, St. Forces, Kingston, Ont., K7K 7B4 (Canada); Hood, J. [Atomic Energy of Canada Ltd., Sheridan Park, 2251 Speakman Drive, Mississauga, Ont., L5K 1B2 (Canada); Akbari, F.; He, Z. [Atomic Energy of Canada Ltd., Chalk River Laboratories, Chalk River, Ont., K0J 1J0 (Canada); Reid, P. [Atomic Energy of Canada Ltd., Sheridan Park, 2251 Speakman Drive, Mississauga, Ont., L5K 1B2 (Canada)
2009-03-31
Burnable neutron absorbing materials are expected to be an integral part of the new fuel design for the Advanced CANDU [CANDU is as a registered trademark of Atomic Energy of Canada Limited.] Reactor. The neutron absorbing material is composed of gadolinia and dysprosia dissolved in an inert cubic-fluorite yttria-stabilized zirconia matrix. A thermodynamic model based on Gibbs energy minimization has been created to provide estimated phase equilibria as a function of composition and temperature. This work includes some supporting experimental studies involving X-ray diffraction.
Some preliminary design considerations for the ANS [Advanced Neutron Source] reactor cold source
International Nuclear Information System (INIS)
Henderson, D.L.
1988-01-01
Two areas concerned with the design of the Advanced Neutron Source (ANS) cold source have been investigated by simple one-dimensional calculations. The gain factors computed for a possible liquid nitrogen-15 cold source moderator are considerably below those computed for the much colder liquid deuterium moderator, as is reasonable considering the difference in moderator temperature. Nevertheless, nitrogen-15 does represent a viable option should safety related issues prohibit the use of deuterium as a moderating material. The slab geometry calculations have indicated that reflection of neutrons may be the dominant moderating mechanism and should be a consideration in the design of the cold source. 9 refs., 2 figs
Directory of Open Access Journals (Sweden)
Marziye Ebrahimkhani
2016-02-01
Full Text Available Calculation of the core neutronic parameters is one of the key components in all nuclear reactors. In this research, the energy spectrum and spatial distribution of the neutron flux in a uranium target have been calculated. In addition, sensitivity of the core neutronic parameters in accelerator-driven subcritical advanced liquid metal reactors, such as electron beam energy (Ee and source multiplication coefficient (ks, has been investigated. A Monte Carlo code (MCNPX_2.6 has been used to calculate neutronic parameters such as effective multiplication coefficient (keff, net neutron multiplication (M, neutron yield (Yn/e, energy constant gain (G0, energy gain (G, importance of neutron source (φ∗, axial and radial distributions of neutron flux, and power peaking factor (Pmax/Pave in two axial and radial directions of the reactor core for four fuel loading patterns. According to the results, safety margin and accelerator current (Ie have been decreased in the highest case of ks, but G and φ∗ have increased by 88.9% and 21.6%, respectively. In addition, for LP1 loading pattern, with increasing Ee from 100 MeV up to 1 GeV, Yn/e and G improved by 91.09% and 10.21%, and Ie and Pacc decreased by 91.05% and 10.57%, respectively. The results indicate that placement of the Np–Pu assemblies on the periphery allows for a consistent keff because the Np–Pu assemblies experience less burn-up.
Quantifying moisture transport in cementitious materials using neutron radiography
Lucero, Catherine L.
A portion of the concrete pavements in the US have recently been observed to have premature joint deterioration. This damage is caused in part by the ingress of fluids, like water, salt water, or deicing salts. The ingress of these fluids can damage concrete when they freeze and expand or can react with the cementitious matrix causing damage. To determine the quality of concrete for assessing potential service life it is often necessary to measure the rate of fluid ingress, or sorptivity. Neutron imaging is a powerful method for quantifying fluid penetration since it can describe where water has penetrated, how quickly it has penetrated and the volume of water in the concrete or mortar. Neutrons are sensitive to light atoms such as hydrogen and thus clearly detect water at high spatial and temporal resolution. It can be used to detect small changes in moisture content and is ideal for monitoring wetting and drying in mortar exposed to various fluids. This study aimed at developing a method to accurately estimate moisture content in mortar. The common practice is to image the material dry as a reference before exposing to fluid and normalizing subsequent images to the reference. The volume of water can then be computed using the Beer-Lambert law. This method can be limiting because it requires exact image alignment between the reference image and all subsequent images. A model of neutron attenuation in a multi-phase cementitious composite was developed to be used in cases where a reference image is not available. The attenuation coefficients for water, un-hydrated cement, and sand were directly calculated from the neutron images. The attenuation coefficient for the hydration products was then back-calculated. The model can estimate the degree of saturation in a mortar with known mixture proportions without using a reference image for calculation. Absorption in mortars exposed to various fluids (i.e., deionized water and calcium chloride solutions) were investigated
International Nuclear Information System (INIS)
Nakagawa, Yoshinobu; Kobayashi, Tooru; Fukuda, Hiroshi
2006-01-01
The Twelfth International Congress on Neutron Capture Therapy (ICNCT-12) is being held from October 9th to 13th, 2006 at the Kagawa International Congress Hall in Takamatsu, Kagawa, Japan. The main theme of the congress is From the past to the Future'. Five symposiums were organized to accommodate all the contributions from the international scientific committees of the International Society for Neutron Capture Therapy (ISNCT), and two symposiums were added to balance the number of fields of specialties. The seven symposiums for ICNCT-12 are as follows: 1) Clinical Results of BNCT for Brain Tumors, 2) Dosimetry, 3) Treatment Planning system, 4) Drug Delivery System, 5) Biomedical and General Matters, 6) BNCT Systems using Accelerators, 7) New Applications and Protocols for BNCT. There are a total of 195 presentations in this congress: 3 special lectures, 34 symposium presentations, 10 presentations in two special sessions from the recipients of the Ralph G. Fairchild Award, 70 presentations in the oral parallel sessions and 78 presentations in the poster sessions. A compilation of 169 papers are published in this proceedings. The 165 of the presented papers are indexed individually. (J.P.N.)
International Nuclear Information System (INIS)
Velarde, G.; Minguez, E.; Bravo, E.
2003-01-01
We present advances in advanced fusion cycles, atomic physics and radiation hydrodynamics. With ARWEN code we analyze a target design for ICF based on jet production. ARWEN is 2D Adaptive Mesh Refinement fluid dynamic and multigroup radiation transport. We are designing, by using also ARWEN, a target for laboratory simulation of astrophysical phenomena. We feature an experimental device to reproduce collisions of two shock waves, scaled to roughly represent cosmic supernova remnants. Opacity calculations are obtained with ANALOP code, which uses parametric potentials fitting to self-consistent potentials. It includes temperature and density effects by linearized Debye-Hueckel and it treats excited configurations and H+He-like lines. Advanced fusion cycles, as the a neutronic proton-boron 11 reaction, require very high ignition temperatures. Plasma conditions for a fusion-burning wave to propagate at such temperatures are rather extreme and complex, because of the overlapping effects of the main energy transport mechanisms. Calculations on the most appropriate ICF regimes for this purpose are presented. (author)
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)
ZZ AIRFEWG, Gamma, Neutron Transport Calculation in Air Using FEWG1 Cross-Section
International Nuclear Information System (INIS)
1985-01-01
1 - Description of program or function: Format: ANISN; Number of groups: 37 neutron / 21 gamma-ray; Nuclides: air (79% N and 21% O); Origin: DLC-0031/FEWG1 cross sections (ENDF/B-IV). Weighting spectrum: 1/E. The AIRFEWG library has been generated by an ANISN multigroup calculation of gamma-ray, neutron, and secondary gamma-ray transport in infinite homogeneous air using DLC-0031/FEWG1 cross sections. 2 - Method of solution: The results were generated with a P3, ANISN run with a source in a single energy group. Thus, 58 such runs were required. For sources in the 37 neutron groups, both neutron and secondary gamma-ray fluence results were calculated. For gamma-ray sources only gamma-ray fluences were calculated
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)
International Nuclear Information System (INIS)
Ganapol, B.D.
1987-01-01
For almost 20 yr, the main thrust of the author's research has been the generation of as many benchmark solutions to the time-dependent monoenergetic neutron transport equation as possible. The major motivation behind this effort has been to provide code developers with highly accurate numerical solutions to serve as standards in the assessment of numerical transport algorithms. In addition, these solutions provide excellent educational tools since the important physical features of neutron transport are still present even though the problems solved are idealized. A secondary motivation, though of equal importance, is the intellectual stimulation and understanding provided by the combination of the analytical, numerical, and computational techniques required to obtain these solutions. Therefore, to further the benchmark development, the added complication of time-dependent cross sections in the one-group transport equation is considered here
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 par cette methode et de ceux figurant
Monte Carlo simulations of the particle transport in semiconductor detectors of fast neutrons
International Nuclear Information System (INIS)
Sedlačková, Katarína; Zaťko, Bohumír; Šagátová, Andrea; Nečas, Vladimír
2013-01-01
Several Monte Carlo all-particle transport codes are under active development around the world. In this paper we focused on the capabilities of the MCNPX code (Monte Carlo N-Particle eXtended) to follow the particle transport in semiconductor detector of fast neutrons. Semiconductor detector based on semi-insulating GaAs was the object of our investigation. As converter material capable to produce charged particles from the (n, p) interaction, a high-density polyethylene (HDPE) was employed. As the source of fast neutrons, the 239 Pu–Be neutron source was used in the model. The simulations were performed using the MCNPX code which makes possible to track not only neutrons but also recoiled protons at all interesting energies. Hence, the MCNPX code enables seamless particle transport and no other computer program is needed to process the particle transport. The determination of the optimal thickness of the conversion layer and the minimum thickness of the active region of semiconductor detector as well as the energy spectra simulation were the principal goals of the computer modeling. Theoretical detector responses showed that the best detection efficiency can be achieved for 500 μm thick HDPE converter layer. The minimum detector active region thickness has been estimated to be about 400 μm. -- Highlights: ► Application of the MCNPX code for fast neutron detector design is demonstrated. ► Simulations of the particle transport through conversion film of HDPE are presented. ► Simulations of the particle transport through detector active region are presented. ► The optimal thickness of the HDPE conversion film has been calculated. ► Detection efficiency of 0.135% was reached for 500 μm thick HDPE conversion film
Applications of advanced transport aircraft in developing countries
Gobetz, F. W.; Assarabowski, R. J.; Leshane, A. A.
1978-01-01
Four representative market scenarios were studied to evaluate the relative performance of air-and surface-based transportation systems in meeting the needs of two developing contries, Brazil and Indonesia, which were selected for detailed case studies. The market scenarios were: remote mining, low-density transport, tropical forestry, and large cargo aircraft serving processing centers in resource-rich, remote areas. The long-term potential of various aircraft types, together with fleet requirements and necessary technology advances, is determined for each application.
Advanced Transport Operating System (ATOPS) control display unit software description
Slominski, Christopher J.; Parks, Mark A.; Debure, Kelly R.; Heaphy, William J.
1992-01-01
The software created for the Control Display Units (CDUs), used for the Advanced Transport Operating Systems (ATOPS) project, on the Transport Systems Research Vehicle (TSRV) is described. Module descriptions are presented in a standardized format which contains module purpose, calling sequence, a detailed description, and global references. The global reference section includes subroutines, functions, and common variables referenced by a particular module. The CDUs, one for the pilot and one for the copilot, are used for flight management purposes. Operations performed with the CDU affects the aircraft's guidance, navigation, and display software.
International Nuclear Information System (INIS)
Shafii, Mohammad Ali; Meidianti, Rahma; Wildian,; Fitriyani, Dian; Tongkukut, Seni H. J.; Arkundato, Artoto
2014-01-01
Theoretical analysis of integral neutron transport equation using collision probability (CP) method with quadratic flux approach has been carried out. In general, the solution of the neutron transport using the CP method is performed with the flat flux approach. In this research, the CP method is implemented in the cylindrical nuclear fuel cell with the spatial of mesh being conducted into non flat flux approach. It means that the neutron flux at any point in the nuclear fuel cell are considered different each other followed the distribution pattern of quadratic flux. The result is presented here in the form of quadratic flux that is better understanding of the real condition in the cell calculation and as a starting point to be applied in computational calculation
International Nuclear Information System (INIS)
Talamo, Alberto
2013-01-01
This study presents three numerical algorithms to solve the time dependent neutron transport equation by the method of the characteristics. The algorithms have been developed taking into account delayed neutrons and they have been implemented into the novel MCART code, which solves the neutron transport equation for two-dimensional geometry and an arbitrary number of energy groups. The MCART code uses regular mesh for the representation of the spatial domain, it models up-scattering, and takes advantage of OPENMP and OPENGL algorithms for parallel computing and plotting, respectively. The code has been benchmarked with the multiplication factor results of a Boiling Water Reactor, with the analytical results for a prompt jump transient in an infinite medium, and with PARTISN and TDTORT results for cross section and source transients. The numerical simulations have shown that only two numerical algorithms are stable for small time steps
Energy Technology Data Exchange (ETDEWEB)
Talamo, Alberto, E-mail: alby@anl.gov [Nuclear Engineering Division, Argonne National Laboratory, 9700 South Cass Avenue, Lemont, IL 60439 (United States)
2013-05-01
This study presents three numerical algorithms to solve the time dependent neutron transport equation by the method of the characteristics. The algorithms have been developed taking into account delayed neutrons and they have been implemented into the novel MCART code, which solves the neutron transport equation for two-dimensional geometry and an arbitrary number of energy groups. The MCART code uses regular mesh for the representation of the spatial domain, it models up-scattering, and takes advantage of OPENMP and OPENGL algorithms for parallel computing and plotting, respectively. The code has been benchmarked with the multiplication factor results of a Boiling Water Reactor, with the analytical results for a prompt jump transient in an infinite medium, and with PARTISN and TDTORT results for cross section and source transients. The numerical simulations have shown that only two numerical algorithms are stable for small time steps.
Energy Technology Data Exchange (ETDEWEB)
Shafii, Mohammad Ali, E-mail: mashafii@fmipa.unand.ac.id; Meidianti, Rahma, E-mail: mashafii@fmipa.unand.ac.id; Wildian,, E-mail: mashafii@fmipa.unand.ac.id; Fitriyani, Dian, E-mail: mashafii@fmipa.unand.ac.id [Department of Physics, Andalas University Padang West Sumatera Indonesia (Indonesia); Tongkukut, Seni H. J. [Department of Physics, Sam Ratulangi University Manado North Sulawesi Indonesia (Indonesia); Arkundato, Artoto [Department of Physics, Jember University Jember East Java Indonesia (Indonesia)
2014-09-30
Theoretical analysis of integral neutron transport equation using collision probability (CP) method with quadratic flux approach has been carried out. In general, the solution of the neutron transport using the CP method is performed with the flat flux approach. In this research, the CP method is implemented in the cylindrical nuclear fuel cell with the spatial of mesh being conducted into non flat flux approach. It means that the neutron flux at any point in the nuclear fuel cell are considered different each other followed the distribution pattern of quadratic flux. The result is presented here in the form of quadratic flux that is better understanding of the real condition in the cell calculation and as a starting point to be applied in computational calculation.
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.)
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
International Nuclear Information System (INIS)
Chen Qichang; Wu Hongchun; Cao Liangzhi
2008-01-01
A new 2D neutron transport code AutoMOC for arbitrary geometry has been developed. This code is based on the method of characteristics (MOCs) and the customization of AutoCAD. The MOC solves the neutron transport equation along characteristic lines. It is independent of the geometric shape of boundaries and regions. So theoretically, this method can be used to solve the neutron transport equation in highly complex geometries. However, it is important to describe the geometry and calculate intersection points of each characteristic line with every boundary and region in advance. In complex geometries, due to the complications of treating the arbitrary domain, the selection of geometric shapes and efficiency of ray tracing are generally limited. The geometry treatment through the customization of AutoCAD, a widely used computer-aided design software package, is given in this paper. Thanks to the powerful capability of AutoCAD, the description of arbitrary geometry becomes quite convenient. Moreover, with the language Visual Basic for Applications (VBAs), AutoCAD can be customized to carry out the ray tracing procedure with a high flexibility in geometry. The numerical results show that AutoMOC can solve 2D neutron transport problems in a complex geometry accurately and effectively
Energy Technology Data Exchange (ETDEWEB)
Chen Qichang; Wu Hongchun [School of Nuclear Science and Technology, Xi' an Jiaotong University, Xi' an Shaanxi 710049 (China); Cao Liangzhi [School of Nuclear Science and Technology, Xi' an Jiaotong University, Xi' an Shaanxi 710049 (China)], E-mail: caolz@mail.xjtu.edu.cn
2008-10-15
A new 2D neutron transport code AutoMOC for arbitrary geometry has been developed. This code is based on the method of characteristics (MOCs) and the customization of AutoCAD. The MOC solves the neutron transport equation along characteristic lines. It is independent of the geometric shape of boundaries and regions. So theoretically, this method can be used to solve the neutron transport equation in highly complex geometries. However, it is important to describe the geometry and calculate intersection points of each characteristic line with every boundary and region in advance. In complex geometries, due to the complications of treating the arbitrary domain, the selection of geometric shapes and efficiency of ray tracing are generally limited. The geometry treatment through the customization of AutoCAD, a widely used computer-aided design software package, is given in this paper. Thanks to the powerful capability of AutoCAD, the description of arbitrary geometry becomes quite convenient. Moreover, with the language Visual Basic for Applications (VBAs), AutoCAD can be customized to carry out the ray tracing procedure with a high flexibility in geometry. The numerical results show that AutoMOC can solve 2D neutron transport problems in a complex geometry accurately and effectively.
Neutronics assessment for the ARIES advanced reactor studies
International Nuclear Information System (INIS)
El-Guebaly, L.A.
1995-01-01
The ARIES tokamak designs have incorporated environmental and safety constraints in the design from the beginning. Low activation materials such as SiC or SiC composites, vanadium alloy, and modified HT-9 ferritic steel were utilized as the main structures in ARIES-IV, II, and III, respectively. All designs employ D-T fuel cycles except ARIES-III which is D- 3 He fuelled. An overall tritium breeding ratio of 1.12 seems adequate for ARIES-II and IV. The Li 2 O breeder requires a beryllium multiplier to achieve T self-sufficiency in the ARIES-IV design while the lithium has the ability to breed sufficient T in ARIES-II without a multiplier. Radiation damage concerns for the structures are the burn-up of the SiC and SiC composites and the atomic displacement in the vanadium. The first wall and blanket require frequent replacement (every 3-4 years) during reactor operation. The end-of-life fluences are 16.5MW yearsm -2 and 13MW yearsm -2 based on the 200dpa and 3% burn-up limits for the V and SiC structures respectively. Because of the lower neutron production, the ARIES-III first wall and shield are permanent components and require no replacement over the plant lifetime. A variety of shield options was examined and the ability of various materials to protect the magnets was assessed. At least 1.2m and 1.4m of inboard blanket-shield are required for magnet protection in ARIES-II and ARIES-IV respectively. The lack of T breeding and the lower wall loading result in a much thinner shield (0.65m) for ARIES-III. (orig.)
Radiation Transport Simulation for Boron Neutron Capture Therapy (BNCT)
Energy Technology Data Exchange (ETDEWEB)
Ziegner, M.; Blaickner, M. [AIT Austrian Institute of Technology GmbH, Health and Environment Department, Molecular Medicine, Muthgasse 11, 1190 Wien (Austria); Ziegner, M.; Khan, R.; Boeck, H. [Vienna University of Technology, Institute of Atomic and Subatomic Physics, Stadionallee 2, 1020 Wien (Austria); Bortolussi, S.; Altieri, S. [Department of Nuclear and Theoretical Physics, University of Pavia, National Institute of Nuclear Physics (INFN) Pavia Section, Pavia (Italy); Schmitz, T.; Hampel, G. [Nuclear Chemistry, University of Mainz, Fritz Strassmann Weg 2, 55099 Mainz (Germany)
2011-07-01
This work is part of a larger project initiated by the University of Mainz and aiming to use the university's TRIGA reactor to develop a treatment for liver metastases based on Boron Neutron Capture Therapy (BNCT). Diffuse distribution of cancerous cells within the organ makes complete resection difficult and the vicinity to radiosensitive organs impedes external irradiation. Therefore the method of 'autotransplantation', first established at the University of Pavia, is used. The liver is taken out of the body, irradiated in the thermal column of the reactor, therewith purged of metastases and then reimplanted. A highly precise dosimetry system is to be developed by means of measurements at the University of Mainz and computational calculations at the AIT. The stochastic MCNP-5 Monte Carlo-Code, developed by Los Alamos Laboratories, is applied. To verify the calculations of the flux and the absorbed dose in matter a number of measurements are performed irradiating different phantoms and liver sections in a 20cm x 20cm beam tube, which was created by removing graphite blocks from the thermal column of the reactor. The detector material consists of L- {alpha} -alanine pellets which are thought to be the most suitable because of their good tissue equivalence, small size and their wide response range. Another experiment focuses on the determination of the relative biological effectiveness (RBE-factor) of the neutron and photon dose for liver cells. Therefore cell culture plates with the cell medium enriched with {sup 157}Gd and {sup 10}B at different concentrations are irradiated. With regard to the alanine pellets MCNP-5 calculations give stable results. Nevertheless the absorbed dose is underestimated compared to the measurements, a phenomenon already observed in previous works. The cell culture calculations showed the enormous impact of the added isotopes with high thermal neutron cross sections, especially {sup 157}Gd, on the absorbed dose
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
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
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.
General-purpose Monte Carlo codes for neutron and photon transport calculations. MVP version 3
International Nuclear Information System (INIS)
Nagaya, Yasunobu
2017-01-01
JAEA has developed a general-purpose neutron/photon transport Monte Carlo code MVP. This paper describes the recent development of the MVP code and reviews the basic features and capabilities. In addition, capabilities implemented in Version 3 are also described. (author)
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
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. User's manual and input data
International Nuclear Information System (INIS)
Anhert, C.
1979-01-01
A user's 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 description is given. The new options developped at JEN are included too. (author)
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
Optimizing a three-element core design for the Advanced Neutron Source Reactor
International Nuclear Information System (INIS)
West, C.D.
1995-01-01
Source of neutrons in the proposed Advanced Neutron Source facility is a multipurpose research reactor providing 5-10 times the flux, for neutron beams, of the best existing facilities. Baseline design for the reactor core, based on the ''no new inventions'' rule, was an assembly of two annular fuel elements similar to those used in the Oak Ridge and Grenoble high flux reactors, containing highly enriched U silicide particles. DOE commissioned a study of the use of medium- or low-enriched U; a three-element core design was studied as a means to provide extra volume to accommodate the additional U compound required when the fissionable 235 U has to be diluted with 238 U to reduce the enrichment. This paper describes the design and optimization of that three-element core
Preparation and benchmarking of ANSL-V cross sections for advanced neutron source reactor studies
International Nuclear Information System (INIS)
Arwood, J.W.; Ford, W.E. III; Greene, N.M.; Petrie, L.M.; Primm, R.T. III; Waddell, M.W.; Webster, C.C.; Westfall, R.M.; Wright, R.Q.
1987-01-01
Research and development for the advanced neutron source (ANS) reactor is being funded by the US Dept. of Energy. This reactor is to provide the world's most intense steady-state source of low-energy neutrons for a national experimental user facility. Pseudo-problem-independent, multigroup cross-section libraries were generated to support ANS design work. The libraries, designated ANSL-V, are data bases in AMPX master format for subsequent generation of problem-dependent cross sections for use with codes such as KENO, ANISN, XSDRNPM, VENTURE, DOT, and MORSE. Included in ANSL-V are 123-material P 3 neutron, 46-material P 0 or P 6 secondary gamma-ray production (SGRP), and 34-material P 6 gamma-ray interaction (GRI) libraries
International Nuclear Information System (INIS)
Gast, R.C.
1981-08-01
A procedure for defining diffusion coefficients from Monte Carlo calculations that results in suitable ones for use in neutron diffusion theory calculations is not readily obtained. This study provides a survey of the methods used to define diffusion coefficients from deterministic calculations and provides a discussion as to why such traditional methods cannot be used in Monte Carlo. This study further provides the empirical procedure used for defining diffusion coefficients from the RCP01 Monte Carlo program
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.
International Nuclear Information System (INIS)
Dumazert, Jonathan; Coulon, Romain; Carrel, Frédérick; Corre, Gwenolé; Normand, Stéphane; Méchin, Laurence; Hamel, Matthieu
2016-01-01
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)
Roth, Markus [Technische Univ. Darmstadt (Germany); Vogel, Sven C. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Bourke, Mark Andrew M. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Fernandez, Juan Carlos [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Mocko, Michael Jeffrey [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Glenzer, Siegfried [Stanford Univ., CA (United States); Leemans, Wim [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Siders, Craig [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Haefner, Constantin [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States)
2017-04-19
A variety of opportunities for characterization of fresh nuclear fuels using thermal (~25meV) and epithermal (~10eV) neutrons have been documented at Los Alamos National Laboratory. They include spatially resolved non-destructive characterization of features, isotopic enrichment, chemical heterogeneity and stoichiometry. The LANSCE spallation neutron source is well suited in neutron fluence and temporal characteristics for studies of fuels. However, recent advances in high power short pulse lasers suggest that compact neutron sources might, over the next decade, become viable at a price point that would permit their consideration for poolside characterization on site at irradiation facilities. In a laser-driven neutron source the laser is used to accelerate deuterium ions into a beryllium target where neutrons are produced. At this time, the technology is new and their total neutron production is approximately four orders of magnitude less than a facility like LANSCE. However, recent measurements on a sub-optimized system demonstrated >10^{10} neutrons in sub-nanosecond pulses in predominantly forward direction. The compactness of the target system compared to a spallation target may allow exchanging the target during a measurement to e.g. characterize a highly radioactive sample with thermal, epithermal, and fast neutrons as well as hard X-rays, thus avoiding sample handling. At this time several groups are working on laser-driven neutron production and are advancing concepts for lasers, laser targets, and optimized neutron target/moderator systems. Advances in performance sufficient to enable poolside fuels characterization with LANSCE-like fluence on sample within a decade may be possible. This report describes the underlying physics and state-of-the-art of the laser-driven neutron production process from the perspective of the DOE/NE mission. It also discusses the development and understanding that will be necessary to provide customized capability for
Minaret, a deterministic neutron transport solver for nuclear core calculations
International Nuclear Information System (INIS)
Moller, J-Y.; Lautard, J-J.
2011-01-01
We present here MINARET a deterministic transport solver for nuclear core calculations to solve the steady state Boltzmann equation. The code follows the multi-group formalism to discretize the energy variable. It uses discrete ordinate method to deal with the angular variable and a DGFEM to solve spatially the Boltzmann equation. The mesh is unstructured in 2D and semi-unstructured in 3D (cylindrical). Curved triangles can be used to fit the exact geometry. For the curved elements, two different sets of basis functions can be used. Transport solver is accelerated with a DSA method. Diffusion and SPN calculations are made possible by skipping the transport sweep in the source iteration. The transport calculations are parallelized with respect to the angular directions. Numerical results are presented for simple geometries and for the C5G7 Benchmark, JHR reactor and the ESFR (in 2D and 3D). Straight and curved finite element results are compared. (author)
Minaret, a deterministic neutron transport solver for nuclear core calculations
Energy Technology Data Exchange (ETDEWEB)
Moller, J-Y.; Lautard, J-J., E-mail: jean-yves.moller@cea.fr, E-mail: jean-jacques.lautard@cea.fr [CEA - Centre de Saclay , Gif sur Yvette (France)
2011-07-01
We present here MINARET a deterministic transport solver for nuclear core calculations to solve the steady state Boltzmann equation. The code follows the multi-group formalism to discretize the energy variable. It uses discrete ordinate method to deal with the angular variable and a DGFEM to solve spatially the Boltzmann equation. The mesh is unstructured in 2D and semi-unstructured in 3D (cylindrical). Curved triangles can be used to fit the exact geometry. For the curved elements, two different sets of basis functions can be used. Transport solver is accelerated with a DSA method. Diffusion and SPN calculations are made possible by skipping the transport sweep in the source iteration. The transport calculations are parallelized with respect to the angular directions. Numerical results are presented for simple geometries and for the C5G7 Benchmark, JHR reactor and the ESFR (in 2D and 3D). Straight and curved finite element results are compared. (author)
Advanced Transport Operating System (ATOPS) utility library software description
Clinedinst, Winston C.; Slominski, Christopher J.; Dickson, Richard W.; Wolverton, David A.
1993-01-01
The individual software processes used in the flight computers on-board the Advanced Transport Operating System (ATOPS) aircraft have many common functional elements. A library of commonly used software modules was created for general uses among the processes. The library includes modules for mathematical computations, data formatting, system database interfacing, and condition handling. The modules available in the library and their associated calling requirements are described.
International Nuclear Information System (INIS)
Mikata, Y.
2014-01-01
Highlights: • An exact solution for the one-speed neutron transport equation is obtained. • This solution as well as its derivation are believed to be new. • Neutron flux for a purely absorbing material with a point neutron source off the origin is obtained. • Spherically as well as cylindrically piecewise constant cross sections are studied. • Neutron flux expressions for a point neutron source off the origin are believed to be new. - Abstract: An exact analytical solution of the time-independent monoenergetic neutron transport equation is obtained in this paper. The solution is applied to systems with a point source. Systematic analysis of the solution of the time-independent neutron transport equation, and its applications represent the primary goal of this paper. To the best of the author’s knowledge, certain key results on the scalar neutron flux as well as their derivations are new. As an application of these results, a scalar neutron flux for a purely absorbing medium with a spherically piecewise constant cross section and an isotropic point neutron source off the origin as well as that for a cylindrically piecewise constant cross section with a point neutron source off the origin are obtained. Both of these results are believed to be new
International Nuclear Information System (INIS)
Santi, P.; Favalli, A.; Hauck, D.; Henzl, V.; Henzlova, D.; Ianakiev, K.; Iliev, M.; Swinhoe, M.; Croft, S.; Worrall, L.
2015-01-01
One of the most distinctive and informative signatures of special nuclear materials is the emission of correlated neutrons from either spontaneous or induced fission. Because the emission of correlated neutrons is a unique and unmistakable signature of nuclear materials, the ability to effectively detect, process, and analyze these emissions will continue to play a vital role in the non-proliferation, safeguards, and security missions. While currently deployed neutron measurement techniques based on 3He proportional counter technology, such as neutron coincidence and multiplicity counters currently used by the International Atomic Energy Agency, have proven to be effective over the past several decades for a wide range of measurement needs, a number of technical and practical limitations exist in continuing to apply this technique to future measurement needs. In many cases, those limitations exist within the algorithms that are used to process and analyze the detected signals from these counters that were initially developed approximately 20 years ago based on the technology and computing power that was available at that time. Over the past three years, an effort has been undertaken to address the general shortcomings in these algorithms by developing new algorithms that are based on fundamental physics principles that should lead to the development of more sensitive neutron non-destructive assay instrumentation. Through this effort, a number of advancements have been made in correcting incoming data for electronic dead time, connecting the two main types of analysis techniques used to quantify the data (Shift register analysis and Feynman variance to mean analysis), and in the underlying physical model, known as the point model, that is used to interpret the data in terms of the characteristic properties of the item being measured. The current status of the testing and evaluation of these advancements in correlated neutron analysis techniques will be discussed
Reducing global NOx emissions: developing advanced energy and transportation technologies.
Bradley, Michael J; Jones, Brian M
2002-03-01
Globally, energy demand is projected to continue to increase well into the future. As a result, global NOx emissions are projected to continue on an upward trend for the foreseeable future as developing countries increase their standards of living. While the US has experienced improvements in reducing NOx emissions from stationary and mobile sources to reduce ozone, further progress is needed to reduce the health and ecosystem impacts associated with NOx emissions. In other parts of the world, (in developing countries in particular) NOx emissions have been increasing steadily with the growth in demand for electricity and transportation. Advancements in energy and transportation technologies may help avoid this increase in emissions if appropriate policies are implemented. This paper evaluates commercially available power generation and transportation technologies that produce fewer NOx emissions than conventional technologies, and advanced technologies that are on the 10-year commercialization horizon. Various policy approaches will be evaluated which can be implemented on the regional, national and international levels to promote these advanced technologies and ultimately reduce NOx emissions. The concept of the technology leap is offered as a possibility for the developing world to avoid the projected increases in NOx emissions.
Transport modelling and gyrokinetic analysis of advanced high performance discharges
International Nuclear Information System (INIS)
Kinsey, J.E.; Imbeaux, F.; Staebler, G.M.; Budny, R.; Bourdelle, C.; Fukuyama, A.; Garbet, X.; Tala, T.; Parail, V.
2005-01-01
Predictive transport modelling and gyrokinetic stability analyses of demonstration hybrid (HYBRID) and advanced tokamak (AT) discharges from the International Tokamak Physics Activity (ITPA) profile database are presented. Both regimes have exhibited enhanced core confinement (above the conventional ITER reference H-mode scenario) but differ in their current density profiles. Recent contributions to the ITPA database have facilitated an effort to study the underlying physics governing confinement in these advanced scenarios. In this paper, we assess the level of commonality of the turbulent transport physics and the relative roles of the transport suppression mechanisms (i.e. E x B shear and Shafranov shift (α) stabilization) using data for select HYBRID and AT discharges from the DIII-D, JET and AUG tokamaks. GLF23 transport modelling and gyrokinetic stability analysis indicate that E x B shear and Shafranov shift stabilization play essential roles in producing the improved core confinement in both HYBRID and AT discharges. Shafranov shift stabilization is found to be more important in AT discharges than in HYBRID discharges. We have also examined the competition between the stabilizing effects of E x B shear and Shafranov shift stabilization and the destabilizing effects of higher safety factors and parallel velocity shear. Linear and nonlinear gyrokinetic simulations of idealized low and high safety factor cases reveal some interesting consequences. A low safety factor (i.e. HYBRID relevant) is directly beneficial in reducing the transport, and E x B shear stabilization can dominate parallel velocity shear destabilization allowing the turbulence to be quenched. However, at low-q/high current, Shafranov shift stabilization plays less of a role. Higher safety factors (as found in AT discharges), on the other hand, have larger amounts of Shafranov shift stabilization, but parallel velocity shear destabilization can prevent E x B shear quenching of the turbulent
Transport modeling and gyrokinetic analysis of advanced high performance discharges
International Nuclear Information System (INIS)
Kinsey, J.; Imbeaux, F.; Bourdelle, C.; Garbet, X.; Staebler, G.; Budny, R.; Fukuyama, A.; Tala, T.; Parail, V.
2005-01-01
Predictive transport modeling and gyrokinetic stability analyses of demonstration hybrid (HYBRID) and Advanced Tokamak (AT) discharges from the International Tokamak Physics Activity (ITPA) profile database are presented. Both regimes have exhibited enhanced core confinement (above the conventional ITER reference H-mode scenario) but differ in their current density profiles. Recent contributions to the ITPA database have facilitated an effort to study the underlying physics governing confinement in these advanced scenarios. In this paper, we assess the level of commonality of the turbulent transport physics and the relative roles of the transport suppression mechanisms (i.e. ExB shear and Shafranov shift (α) stabilization) using data for select HYBRID and AT discharges from the DIII-D, JET, and AUG tokamaks. GLF23 transport modeling and gyrokinetic stability analysis indicates that ExB shear and Shafranov shift stabilization play essential roles in producing the improved core confinement in both HYBRID and AT discharges. Shafranov shift stabilization is found to be more important in AT discharges than in HYBRID discharges. We have also examined the competition between the stabilizing effects of ExB shear and Shafranov shift stabilization and the destabilizing effects of higher safety factors and parallel velocity shear. Linear and nonlinear gyrokinetic simulations of idealized low and high safety factor cases reveals some interesting consequences. A low safety factor (i.e. HYBRID relevant) is directly beneficial in reducing the transport, and ExB shear stabilization can win out over parallel velocity shear destabilization allowing the turbulence to be quenched. However, at low-q/high current, Shafranov shift stabilization plays less of a role. Higher safety factors (as found in AT discharges), on the other hand, have larger amounts of Shafranov shift stabilization, but parallel velocity shear destabilization can prevent ExB shear quenching of the turbulent
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
Cf-252 neutron brachytherapy: an advance for bulky localized cancer therapy
International Nuclear Information System (INIS)
Maruyama, Y.
1984-01-01
The physical and radiobiogical basis as well as the rationale for neutron brachytherapy, using Cf-252, in human cancer therapy is reviewed. Cf-252 brachytherapy represents an economical and effective form of neutron radiotherapy that is readily and safely applied clinically. It can be used anywhere in the world without unusual personnel, equipment or facilities, or prohibitive expenses or maintenance costs. Used on bulky head and neck, thoracic, abdominal, pelvic, brain and appendage cancers, it overcomes hypoxic radioresistance and produces remarkable rates of tumor clearance. It is easily combined with photon radiotherapy and in proper schedules and doses, it can control advanced but still localized regional cancers to produce tumor cure. It will clear the local manifestations of recurrent or metastatic tumors or advanced stages of primary tumors and therefore in conjunction with other adjuvant therapies offers much more effective tumor control and palliation than present conventional therapy. (Auth.)
International Nuclear Information System (INIS)
Nahavandi, N.; Minuchehr, A.; Zolfaghari, A.; Abbasi, M.
2015-01-01
Highlights: • Powerful hp-SEM refinement approach for P N neutron transport equation has been presented. • The method provides great geometrical flexibility and lower computational cost. • There is a capability of using arbitrary high order and non uniform meshes. • Both posteriori and priori local error estimation approaches have been employed. • High accurate results are compared against other common adaptive and uniform grids. - Abstract: In this work we presented the adaptive hp-SEM approach which is obtained from the incorporation of Spectral Element Method (SEM) and adaptive hp refinement. The SEM nodal discretization and hp adaptive grid-refinement for even-parity Boltzmann neutron transport equation creates powerful grid refinement approach with high accuracy solutions. In this regard a computer code has been developed to solve multi-group neutron transport equation in one-dimensional geometry using even-parity transport theory. The spatial dependence of flux has been developed via SEM method with Lobatto orthogonal polynomial. Two commonly error estimation approaches, the posteriori and the priori has been implemented. The incorporation of SEM nodal discretization method and adaptive hp grid refinement leads to high accurate solutions. Coarser meshes efficiency and significant reduction of computer program runtime in comparison with other common refining methods and uniform meshing approaches is tested along several well-known transport benchmarks
Removal, transportation and disposal of the Millstone 2 neutron thermal shield
International Nuclear Information System (INIS)
Snedeker, D.F.; Thomas, L.S.; Schmoker, D.S.; Cade, M.S.
1985-01-01
Some PWR reactors equipped with neutron thermal shields (NTS) have experienced severe neutron shield degradation to the extent that removal and disposal of these shields has become necessary. Due to the relative size and activation levels of the thermal shield, disposal techniques, remote material handling and transportation equipment must be carefully evaluated to minimize plant down time and maintain disposal costs at a minimum. This paper describes the techniques, equipment and methodology employed in the removal, transportation and disposal of the NTS at the Millstone 2 Nuclear Generating Station, a PWR facility owned and operated by Northeast Utilities of Hartford, CT. Specific areas addressed include: (1) remote underwater equipment and tooling for use in segmenting and loading the thermal shield in a disposal liner; (2) adaptation of the General Electric IF-300 Irradiated Fuel Cask for transportation of the NTS for disposal; (3) equipment and techniques used for cask handling and liner burial at the Low Level Radioactive Waste (LLRW) disposal facility
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)
Development of a CAD-based neutron transport code with the method of characteristics
International Nuclear Information System (INIS)
Chen Zhenping; Wang Dianxi; He Tao; Wang Guozhong; Zheng Huaqing
2012-01-01
The main problem determining whether the method of characteristics (MOC) can be used in complicated and highly heterogeneous geometry is how to combine an effective geometry processing method with MOC. In this study, a new idea making use of MCAM, which is a Mutlti-Calculation Automatic Modeling for Neutronics and Radiation Transport program developed by FDS Team, for geometry description and ray tracing of particle transport was brought forward to solve the geometry problem mentioned above. Based on the theory and approach as the foregoing statement, a two dimensional neutron transport code was developed which had been integrated into VisualBUS, developed by FDS Team. Several benchmarks were used to verify the validity of the code and the numerical results were coincident with the reference values very well, which indicated the accuracy and feasibility of the method and the MOC code. (authors)
Energy Technology Data Exchange (ETDEWEB)
Saha Ray, S., E-mail: santanusaharay@yahoo.com; Patra, A.
2014-10-15
Highlights: • A stationary transport equation has been solved using the technique of Haar wavelet collocation method. • This paper intends to provide the great utility of Haar wavelets to nuclear science problem. • In the present paper, two-dimensional Haar wavelets are applied. • The proposed method is mathematically very simple, easy and fast. - Abstract: In this paper the numerical solution for the fractional order stationary neutron transport equation is presented using Haar wavelet Collocation Method (HWCM). Haar wavelet collocation method is efficient and powerful in solving wide class of linear and nonlinear differential equations. This paper intends to provide an application of Haar wavelets to nuclear science problems. This paper describes the application of Haar wavelets for the numerical solution of fractional order stationary neutron transport equation in homogeneous medium with isotropic scattering. The proposed method is mathematically very simple, easy and fast. To demonstrate about the efficiency and applicability of the method, two test problems are discussed.
International Nuclear Information System (INIS)
Chen, G.S.
1997-01-01
We apply and compare the preconditioned generalized conjugate gradient methods to solve the linear system equation that arises in the two-dimensional neutron and photon transport equation in this paper. Several subroutines are developed on the basis of preconditioned generalized conjugate gradient methods for time-independent, two-dimensional neutron and photon transport equation in the transport theory. These generalized conjugate gradient methods are used. TFQMR (transpose free quasi-minimal residual algorithm), CGS (conjuage gradient square algorithm), Bi-CGSTAB (bi-conjugate gradient stabilized algorithm) and QMRCGSTAB (quasi-minimal residual variant of bi-conjugate gradient stabilized algorithm). These sub-routines are connected to computer program DORT. Several problems are tested on a personal computer with Intel Pentium CPU. (author)
Summary of dynamic analyses of the advanced neutron source reactor inner control rods
International Nuclear Information System (INIS)
Hendrich, W.R.
1995-08-01
A summary of the structural dynamic analyses that were instrumental in providing design guidance to the Advanced Neutron source (ANS) inner control element system is presented in this report. The structural analyses and the functional constraints that required certain performance parameters were combined to shape and guide the design effort toward a prediction of successful and reliable control and scram operation to be provided by these inner control rods
Loss-of-coolant accident analyses of the Advanced Neutron Source Reactor
International Nuclear Information System (INIS)
Chen, N.C.J.; Yoder, G.L.; Wendel, M.W.
1991-01-01
Currently in the conceptual design stage, the Advanced Neutron Source Reactor (ANSR) will operate at a high heat flux, a high mass flux, an a high degree of coolant subcooling. Loss-of-coolant accident (LOCA) analyses using RELAP5 have been performed as part of an early evaluation of ANSR safety issues. This paper discusses the RELAP5 ANSR conceptual design system model and preliminary LOCA simulation results. Some previous studies were conducted for the preconceptual design. 12 refs., 7 figs
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.
The spectral element method for static neutron transport in AN approximation. Part I
International Nuclear Information System (INIS)
Barbarino, A.; Dulla, S.; Mund, E.H.; Ravetto, P.
2013-01-01
Highlights: ► Spectral elements methods (SEMs) are extended for the neutronics of nuclear reactor cores. ► The second-order, A N formulation of neutron trasport is adopted. ► Results for classical benchmark cases in 2D are presented and compared to finite elements. ► The advantages of SEM in terms of precision and convergence rate are illustrated. ► SEM consitutes a promising approach for the solution of neutron transport problems. - Abstract: Spectral elements methods provide very accurate solutions of elliptic problems. In this paper we apply the method to the A N (i.e. SP 2N−1 ) approximation of neutron transport. Numerical results for classical benchmark cases highlight its performance in comparison with finite element computations, in terms of accuracy per degree of freedom and convergence rate. All calculations presented in this paper refer to two-dimensional problems. The method can easily be extended to three-dimensional cases. The results illustrate promising features of the method for more complex transport problems
Asymptotic equivalence of neutron diffusion and transport in time-independent reactor systems
International Nuclear Information System (INIS)
Borysiewicz, M.; Mika, J.; Spiga, G.
1982-01-01
Presented in this paper is the asymptotic analysis of the time-independent neutron transport equation in the second-order variational formulation. The small parameter introduced into the equation is an estimate of the ratio of absorption and leakage to scattering in the system considered. When the ratio tends to zero, the weak solution to the transport problem tends to the weak solution of the diffusion problem, including properly defined boundary conditions. A formula for the diffusion coefficient different from that based on averaging the transport mean-free-path is derived
International Nuclear Information System (INIS)
Hoogenboom, J.E.
1981-01-01
An adjoint Monte Carlo technique is described for the solution of neutron transport problems. The optimum biasing function for a zero-variance collision estimator is derived. The optimum treatment of an analog of a non-velocity thermal group has also been derived. The method is extended to multiplying systems, especially for eigenfunction problems to enable the estimate of averages over the unknown fundamental neutron flux distribution. A versatile computer code, FOCUS, has been written, based on the described theory. Numerical examples are given for a shielding problem and a critical assembly, illustrating the performance of the FOCUS code. 19 refs
UN Method For The Critical Slab Problem In One-Speed Neutron Transport Theory
International Nuclear Information System (INIS)
Oeztuerk, Hakan; Guengoer, Sueleyman
2008-01-01
The Chebyshev polynomial approximation (U N method) is used to solve the critical slab problem in one-speed neutron transport theory using Marshak boundary condition. The isotropic scattering kernel with the combination of forward and backward scattering is chosen for the neutrons in a uniform finite slab. Numerical results obtained by the U N method are presented in the tables together with the results obtained by the well-known P N method for comparison. It is shown that the method converges rapidly with its easily executable equations.
The neutron transport with general boundary conditions (II)
International Nuclear Information System (INIS)
Boulanouar, Mohamed
2012-01-01
This Note deals with the one-dimensional transport operator, on an unbounded domain, endowed with general boundary conditions. We show the generation of a strongly continuous semigroup and we study its spectral properties. In particular, we prove the existence of a leading eigenvalue. (author)
Perturbative transport experiments in JET Advanced Tokamak plasmas
Energy Technology Data Exchange (ETDEWEB)
Mantica, P.; Gorini, G.; Sozzi, C. [Istituto di Fisica del Plasma, EURATOM-ENEA-CNR Association, Milan (Italy); Imbeaux, F.; Sarazin, Y.; Garbet, X. [Association Euratom-CEA, St. Paul-lez-Durance Cedex (France); Kinsey, J. [Lehigh Univ., Bethlehem, Pennsylvania (United States); Budny, R. [Princeton Plasma Physics Lab, New Jersey (United States); Coffey, I.; Parail, V.; Walden, A. [Euratom/UKAEA Fusion Association, Abingdon, Oxon (United Kingdom); Dux, R. [Max-Planck-Institut fuer Plasmaphysik, Garching (Germany); Garzotti, L. [Istituto Gas Ionizzati, Padova (Italy); Ingesson, C. [FOM-Instituut voor Plasmafysica, Nieuwegein (Netherlands); Kissick, M. [University of California, Los Angeles (United States)
2003-07-01
Perturbative transport experiments have been performed in JET Advanced Tokamak plasmas either in conditions of fully developed Internal Transport Barrier (ITB) or during a phase where an ITB was not observed. Transient peripheral cooling was induced by either Laser Ablation or Shallow Pellet Injection and the ensuing travelling cold pulse was used to probe the plasma transport in the electron and, for the first time, also in the ion channel. Cold pulses travelling through ITBs are observed to erode the ITB outer part, but, if the inner ITB portion survives, it strongly damps the propagating wave. The result is discussed in the context of proposed possible pictures for ITB formation. In the absence of an ITB, the cold pulse shows a fast propagation in the outer plasma half, which is consistent with a region of stiff transport, while in the inner half it slows down but shows the peculiar feature of amplitude growing while propagating. The data are powerful tests for the validation of theoretical transport models. (author)
International Nuclear Information System (INIS)
Kling, A.; Barao, F.J.C.; Nakagawa, M.; Tavora, L.
2001-01-01
The following topics were dealt with: Electron and photon interactions and transport mechanisms, random number generation, applications in medical physisc, microdosimetry, track structure, radiobiological modeling, Monte Carlo method in radiotherapy, dosimetry, and medical accelerator simulation, neutron transport, high-energy hadron transport. (HSI)
International Nuclear Information System (INIS)
Khayat, M.I.; March-Leuba, J.
1993-01-01
This paper documents some of the optimization studies performed to maximize the performance of the engineered safety features and scram systems to mitigate the consequences of large breaks in the primary cooling system of the Advanced Neutron Source (ANS) Reactor
Heilbronn, Lawrence H.; Townsend, Lawrence W.; Braley, G. Scott; Iwata, Yoshiyuki; Iwase, Hiroshi; Nakamura, Takashi; Ronningen, Reginald M.; Cucinotta, Francis A.
2003-01-01
For humans engaged in long-duration missions in deep space or near-Earth orbit, the risk from exposure to galactic and solar cosmic rays is an important factor in the design of spacecraft, spacesuits, and planetary bases. As cosmic rays are transported through shielding materials and human tissue components, a secondary radiation field is produced. Neutrons are an important component of that secondary field, especially in thickly-shielded environments. Calculations predict that 50% of the dose-equivalent in a lunar or Martian base comes from neutrons, and a recent workshop held at the Johnson Space Center concluded that as much as 30% of the dose in the International Space Station may come from secondary neutrons. Accelerator facilities provide a means for measuring the effectiveness of various materials in their ability to limit neutron production, using beams and energies that are present in cosmic radiation. The nearly limitless range of beams, energies, and target materials that are present in space, however, means that accelerator-based experiments will not provide a complete database of cross sections and thick-target yields that are necessary to plan and design long-duration missions. As such, accurate nuclear models of neutron production are needed, as well as data sets that can be used to compare with, and verify, the predictions from such models. Improvements in a model of secondary neutron production from heavy-ion interactions are presented here, along with the results from recent accelerator-based measurements of neutron-production cross sections. An analytical knockout-ablation model capable of predicting neutron production from high-energy hadron-hadron interactions (both nucleon-nucleus and nucleus-nucleus collisions) has been previously developed. In the knockout stage, the collision between two nuclei result in the emission of one or more nucleons from the projectile and/or target. The resulting projectile and target remnants, referred to as
Advanced Air Transportation Technologies Project, Final Document Collection
Mogford, Richard H.; Wold, Sheryl (Editor)
2008-01-01
This CD ROM contains a compilation of the final documents of the Advanced Air Transportation Technologies (AAIT) project, which was an eight-year (1996 to 2004), $400M project managed by the Airspace Systems Program office, which was part of the Aeronautics Research Mission Directorate at NASA Headquarters. AAIT focused on developing advanced automation tools and air traffic management concepts that would help improve the efficiency of the National Airspace System, while maintaining or enhancing safety. The documents contained in the CD are final reports on AAIT tasks that serve to document the project's accomplishments over its eight-year term. Documents include information on: Advanced Air Transportation Technologies, Autonomous Operations Planner, Collaborative Arrival Planner, Distributed Air/Ground Traffic Management Concept Elements 5, 6, & 11, Direct-To, Direct-To Technology Transfer, Expedite Departure Path, En Route Data Exchange, Final Approach Spacing Tool - (Active and Passive), Multi-Center Traffic Management Advisor, Multi Center Traffic Management Advisor Technology Transfer, Surface Movement Advisor, Surface Management System, Surface Management System Technology Transfer and Traffic Flow Management Research & Development.
Fixed Wing Project: Technologies for Advanced Air Transports
Del Rosario, Ruben; Koudelka, John M.; Wahls, Richard A.; Madavan, Nateri
2014-01-01
The NASA Fundamental Aeronautics Fixed Wing (FW) Project addresses the comprehensive challenge of enabling revolutionary energy efficiency improvements in subsonic transport aircraft combined with dramatic reductions in harmful emissions and perceived noise to facilitate sustained growth of the air transportation system. Advanced technologies and the development of unconventional aircraft systems offer the potential to achieve these improvements. Multidisciplinary advances are required in aerodynamic efficiency to reduce drag, structural efficiency to reduce aircraft empty weight, and propulsive and thermal efficiency to reduce thrust-specific energy consumption (TSEC) for overall system benefit. Additionally, advances are required to reduce perceived noise without adversely affecting drag, weight, or TSEC, and to reduce harmful emissions without adversely affecting energy efficiency or noise.The presentation will highlight the Fixed Wing project vision of revolutionary systems and technologies needed to achieve these challenging goals. Specifically, the primary focus of the FW Project is on the N+3 generation; that is, vehicles that are three generations beyond the current state of the art, requiring mature technology solutions in the 2025-30 timeframe.
Neutron shielding studies on an advanced molten salt fast reactor design
International Nuclear Information System (INIS)
Merk, Bruno; Konheiser, Jörg
2014-01-01
Highlights: • Material damage due to irradiation has already been discovered at the MSRE. • Neutronic analysis of MSFR with curved blanket wall geometry. • Neutron fluence limit at the wall of the outer vessel can be kept for 80 years. • Shielded MSFR core will be of same dimension than a SFR core. - Abstract: The molten salt reactor technology has gained some new interest. In contrast to the historic molten salt reactors, the current projects are based on designing a molten salt fast reactor. Thus the shielding becomes significantly more challenging than in historic concepts. One very interesting and innovative result of the most recent EURATOM project on molten salt reactors – EVOL – is the fluid flow optimized design of the inner reactor vessel using curved blanket walls. The developed structure leads to a very uniform flow distribution. The design avoids all internal structures. Based on this new geometry a model for neutron physics calculation is presented. The major steps are: the modeling of the curved geometry in the unstructured mesh neutron transport code HELIOS and the determination of the real neutron flux and power distribution for this new geometry. The developed model is then used for the determination of the neutron fluence distribution in the inner and outer wall of the system. Based on these results an optimized shielding strategy is developed for the molten salt fast reactor to keep the fluence in the safety related outer vessel below expected limit values. A lifetime of 80 years can be assured, but the size of the core/blanket system will be comparable to a sodium cooled fast reactor. The HELIOS results are verified against Monte-Carlo calculations with very satisfactory agreement for a deep penetration problem
An integral equation arising in two group neutron transport theory
International Nuclear Information System (INIS)
Cassell, J S; Williams, M M R
2003-01-01
An integral equation describing the fuel distribution necessary to maintain a flat flux in a nuclear reactor in two group transport theory is reduced to the solution of a singular integral equation. The formalism developed enables the physical aspects of the problem to be better understood and its relationship with the corresponding diffusion theory model is highlighted. The integral equation is solved by reducing it to a non-singular Fredholm equation which is then evaluated numerically
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
Methodology for coupling computational fluid dynamics and integral transport neutronics
International Nuclear Information System (INIS)
Thomas, J. W.; Zhong, Z.; Sofu, T.; Downar, T. J.
2004-01-01
The CFD code STAR-CD was coupled to the integral transport code DeCART in order to provide high-fidelity, full physics reactor simulations. An interface program was developed to perform the tasks of mapping the STAR-CD mesh to the DeCART mesh, managing all communication between STAR-CD and DeCART, and monitoring the convergence of the coupled calculations. The interface software was validated by comparing coupled calculation results with those obtained using an independently developed interface program. An investigation into the convergence characteristics of coupled calculations was performed using several test models on a multiprocessor LINUX cluster. The results indicate that the optimal convergence of the coupled field calculation depends on several factors, to include the tolerance of the STAR-CD solution and the number of DeCART transport sweeps performed before exchanging data between codes. Results for a 3D, multi-assembly PWR problem on 12 PEs of the LINUX cluster indicate the best performance is achieved when the STAR-CD tolerance and number of DeCART transport sweeps are chosen such that the two fields converge at approximately the same rate. (authors)
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)
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)
International Nuclear Information System (INIS)
Dura, Joseph A.; Pierce, Donald J.; Majkrzak, Charles F.; Maliszewskyj, Nicholas C.; McGillivray, Duncan J.; Loesche, Mathias; O'Donovan, Kevin V.; Mihailescu, Mihaela; Perez-Salas, Ursula; Worcester, David L.; White, Stephen H.
2006-01-01
An elastic neutron scattering instrument, the advanced neutron diffractometer/reflectometer (AND/R), has recently been commissioned at the National Institute of Standards and Technology Center for Neutron Research. The AND/R is the centerpiece of the Cold Neutrons for Biology and Technology partnership, which is dedicated to the structural characterization of thin films and multilayers of biological interest. The instrument is capable of measuring both specular and nonspecular reflectivity, as well as crystalline or semicrystalline diffraction at wave-vector transfers up to approximately 2.20 A -1 . A detailed description of this flexible instrument and its performance characteristics in various operating modes are given
Evaluation of undeveloped rocket engine cycle applications to advanced transportation
1990-01-01
Undeveloped pump-fed, liquid propellant rocket engine cycles were assessed and evaluated for application to Next Manned Transportation System (NMTS) vehicles, which would include the evolving Space Transportation System (STS Evolution), the Personnel Launch System (PLS), and the Advanced Manned Launch System (AMLS). Undeveloped engine cycles selected for further analysis had potential for increased reliability, more maintainability, reduced cost, and improved (or possibly level) performance when compared to the existing SSME and proposed STME engines. The split expander (SX) cycle, the full flow staged combustion (FFSC) cycle, and a hybrid version of the FFSC, which has a LOX expander drive for the LOX pump, were selected for definition and analysis. Technology requirements and issues were identified and analyses of vehicle systems weight deltas using the SX and FFSC cycles in AMLS vehicles were performed. A strawman schedule and cost estimate for FFSC subsystem technology developments and integrated engine system demonstration was also provided.
Advanced physical protection systems for facilities and transportation
International Nuclear Information System (INIS)
Jones, O.E.
1976-01-01
Sandia Laboratories is developing advanced physical protection safeguards in order to improve the security of special nuclear materials, facilities, and transportation. Computer models are being used to assess the cost-effectiveness of alternative systems for protecting facilities against external attack which may include internal assistance, and against internal theft or sabotage. Physical protection elements such as admittance controls, portals and detectors, perimeter and interior intrusion alarms, fixed and remotely activated barriers, and secure communications are being evaluated, adapted, and where required, developed. New facilities safeguards concepts which involve ''control loops'' between physical protection and materials control elements are being evolved jointly between Sandia Laboratories and Los Alamos Scientific Laboratory. Special vehicles and digital communications equipment have been developed for the ERDA safe-secure transportation system. The current status and direction of these activities are surveyed
Advanced physical protection systems for facilities and transportation
International Nuclear Information System (INIS)
Jones, O.E.
1976-01-01
Sandia Laboratories is developing advanced physical protection safeguards in order to improve the security of special nuclear materials, facilities, and transportation. Computer models are being used to assess the cost-effectiveness of alternative systems for protecting facilities against external attack which may include internal assistance, and against internal theft or sabotage. Physical protection elements such as admittance controls, portals and detectors, perimeter and interior intrusion alarms, fixed and remotely-activated barriers, and secure communications are being evaluated, adapted, and where required, developed. New facilities safeguards concepts which involve (control loops) between physical protection and materials control elements are being evolved jointly between Sandia Laboratories and Los Alamos Scientific Laboratory. Special vehicles and digital communications equipment have been developed for the ERDA safe-secure transportation system. The current status and direction of these activities are surveyed
JB-300: An advanced medium size transport for 2005
Debrouwer, Giles; Graham, Katherine; Ison, Jim; Juarez, Vince; Moskalik, Steve; Pankonin, Jon; Weinstein, Arnold
1993-01-01
In the fall of 1992, the TAC Team was presented with a Request for Proposal (PFP) for a mid-size (250-350 passenger) commercial transport. The aircraft was to be extremely competitive in the areas of passenger comfort, performance, and economic aspects. Through the use of supercritical airfoils, a technologically advanced Very High By-pass Ratio (VHBR) turbofan engine, a low overall drag configuration, a comparable interior layout, and mild use of composites, the JB-300 offers an economically viable choice to the airlines. The cents per passenger mile of the JB-300 is 1.76, which is considerably lower than current aircraft in the same range. Overall, the JB-300 is a technologically advanced aircraft, which will meet the demands of the 21st century.
International Nuclear Information System (INIS)
Ford, W.E. III; Arwood, J.W.; Greene, N.M.; Petrie, L.M.; Primm, R.T. III; Waddell, M.W.; Webster, C.C.; Westfall, R.M.; Wright, R.Q.
1987-01-01
Multigroup P3 neutron, P0-P3 secondary gamma ray production (SGRP), and P6 gamma ray interaction (GRI) cross section libraries have been generated to support design work on the Advanced Neutron Source (ANS) reactor. The libraries, designated ANSL-V (Advanced Neutron Source Cross-Section Libraries), are data bases in a format suitable for subsequent generation of problem dependent cross sections. The ANSL-V libraries are available on magnetic tape from the Radiation Shielding Information Center at Oak Ridge National Laboratory
Neutron scattering and muon spin rotation as probes of light interstitial transport
International Nuclear Information System (INIS)
Brown, D.W.
1985-01-01
The transport of light interstitials, specifically of hydrogen isotopes and the positive muon, is studied with the help of microscopic transport models. The principal observables are the differential neutron scattering cross section of the hydrogen isotopes and the muon spin rotation signal of the positive muon. The transport feature of primary interest is coherence arising as a result of persistence of quantum mechanical phase memory. Evaluation of observables is based on the generalized master equation, or alternatively, the stochastic Liouville equation. The latter is applied to obtain the neutron scattering lineshapes for local tunneling systems as well as for extended Bravais and non-Bravais lattices. It is found that the usual form of the stochastic Liouville equation does not address adequately transport among non-degenerate site-states. An appropriate modification is suggested and employed to obtain scattering lineshapes applicable to recent experiments on impurity-trapped hydrogen. The muon spin rotation signal is formulated under the assumption that spin interactions constitute a negligible source of scattering for muon transport. The depolarization function is evaluated for the cases of local tunneling systems and simple models of spatially extended transport. The former addresses consequences of coherence and both address the consequences of the spatial extent of the muon wavefunction. It is found that the depolarization function is sensitive to the wave function extent, and the detail attributable to it is characterized
Multi-group transport methods for high-resolution neutron activation analysis
International Nuclear Information System (INIS)
Burns, K. A.; Smith, L. E.; Gesh, C. J.; Shaver, M. W.
2009-01-01
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. In these applications, high-resolution gamma-ray spectrometers are used to preserve as much information as possible about the emitted photon flux, which consists of both continuum and characteristic gamma rays with discrete energies. Monte Carlo transport is the most commonly used modeling tool for this type of problem, but computational times for many problems can be prohibitive. This work explores the use of multi-group deterministic methods for the simulation of neutron activation problems. Central to this work is the development of a method for generating multi-group neutron-photon cross-sections in a way that separates the discrete and continuum photon emissions so that the key signatures in neutron activation analysis (i.e., the characteristic line energies) are preserved. The mechanics of the cross-section preparation method are described and contrasted with standard neutron-gamma cross-section sets. These custom cross-sections are then applied to several benchmark problems. Multi-group results for neutron and photon flux are compared to MCNP results. Finally, calculated responses of high-resolution spectrometers are compared. Preliminary findings show promising results when compared to MCNP. A detailed discussion of the potential benefits and shortcomings of the multi-group-based approach, in terms of accuracy, and computational efficiency, is provided. (authors)
Anisotropic kernel p(μ → μ') for transport calculations of elastically scattered neutrons
International Nuclear Information System (INIS)
Stevenson, B.
1985-01-01
Literature in the area of anisotropic neutron scattering is by no means lacking. Attention, however, is usually devoted to solution of some particular neutron transport problem and the model employed is at best approximate. The present approach to the problem in general is classically exact and may be of some particular value to individuals seeking exact numerical results in transport calculations. For attempts neutrons originally directed toward the unit vector Omega, it attempts the evaluation of p(theta'), defined such that p(theta') d theta' is that fraction of scattered neutrons that emerges in the vicinity of a cone i.e., having been scattered to between angles theta' and theta' + d theta' with the axis of preferred orientation i; Omega makes an angle theta with i. The relative simplicity of the final form of the solution for hydrogen, in spite of the complicated nature of the limits involved, is a trade-off that truly is not necessary. The exact general solution presented here in integral form, has exceedingly simple limits, i.e., 0 ≤ theta' ≤ π regardless of the material involved; but the form of the final solution is extraordinarily complicated
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)
Neutron transport by collision probability method in complicated geometries
International Nuclear Information System (INIS)
Constantin, Marin
2000-01-01
For the first flight collision probability (FFCP) method a rapidly increasing of the memory requirements and execution time with the number of discrete regions occurs. Generally, the use of the method is restricted at cell/supercell level. However, the amazing developments both in computer hardware and computer architecture allow a real extending of the problems' domain and a more detailed treatment of the geometry. Two ways are discussed into the paper: the direct design of new codes and the improving of the mainframe old versions. The author's experience is focused on the performances' improving of the 3D integral transport code PIJXYZ (from an old version to a modern one) and on the design and developing of the 2D transport code CP 2 D in the last years. In the first case an optimization process have been performed before the parallelization. In the second a modular design and the newest techniques (factorization of the geometry, the macrobands method, the mobile set of chords, the automatic calculation of the integration error, optimal algorithms for the innermost programming level, the mixed method for tracking process and CPs calculation, etc.) were adopted. In both cases the parallelization uses a PCs network system. Some short examples for CP 2 D and PIJXYZ calculation are presented: reactivity void effect in typical CANDU cells using a multistratified coolant model, a problem of some adjacent fuel assemblies, CANDU reactivity devices 3D simulation. (author)
Resolution of the neutron transport equation by a three-dimensional least square method
International Nuclear Information System (INIS)
Varin, Elisabeth
2001-01-01
The knowledge of space and time distribution of neutrons with a certain energy or speed allows the exploitation and control of a nuclear reactor and the assessment of the irradiation dose about an irradiated nuclear fuel storage site. The neutron density is described by a transport equation. The objective of this research thesis is to develop a software for the resolution of this stationary equation in a three-dimensional Cartesian domain by means of a deterministic method. After a presentation of the transport equation, the author gives an overview of the different deterministic resolution approaches, identifies their benefits and drawbacks, and discusses the choice of the Ressel method. The least square method is precisely described and then applied. Numerical benchmarks are reported for validation purposes
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
A coordinate transform method for one-speed neutron transport in composite slabs
International Nuclear Information System (INIS)
Haidar, N.H.S.
1988-01-01
The optical path transformation is applied to reduce the one-speed neutron transport equation for a class of composite subcritical slabs to single-region problems. The class idealises, within the uncertainty of the one-speed model, a variety of practical situations such as U-D 2 O-C-Zr-Pb or Pu-U-Na-Fe symmetric reactor assemblies; which may possibly contain a symmetrically anisotropic neutron source. A closed form double series solution, which turns out to be quite convenient for design and optimisation purposes, has been obtained, in terms of discontinuous functions for the multi-regional angular flux by application of a double finite Legendre transform. Disadvantage factor evaluations for a U-C lattice cell resulting from a low-order P 0 P 1 approximation of this method are found to be in full agreement with hybrid diffusion-transport estimates. (author)
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.
International Nuclear Information System (INIS)
Bosevski, T.
1971-01-01
The polynomial interpolation of neutron flux between the chosen space and energy variables enabled transformation of the integral transport equation into a system of linear equations with constant coefficients. Solutions of this system are the needed values of flux for chosen values of space and energy variables. The proposed improved method for solving the neutron transport problem including the mathematical formalism is simple and efficient since the number of needed input data is decreased both in treating the spatial and energy variables. Mathematical method based on this approach gives more stable solutions with significantly decreased probability of numerical errors. Computer code based on the proposed method was used for calculations of one heavy water and one light water reactor cell, and the results were compared to results of other very precise calculations. The proposed method was better concerning convergence rate, decreased computing time and needed computer memory. Discretization of variables enabled direct comparison of theoretical and experimental results
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)
International Nuclear Information System (INIS)
Asadzadeh, M.; Thevenot, L.
2010-01-01
The objective of this paper is to give a mathematical framework for a fully discrete numerical approach for the study of the neutron transport equation in a cylindrical domain (container model,). More specifically, we consider the discontinuous Galerkin (D G) finite element method for spatial approximation of the mono-energetic, critical neutron transport equation in an infinite cylindrical domain ??in R3 with a polygonal convex cross-section ? The velocity discretization relies on a special quadrature rule developed to give optimal estimates in discrete ordinate parameters compatible with the quasi-uniform spatial mesh. We use interpolation spaces and derive optimal error estimates, up to maximal available regularity, for the fully discrete scalar flux. Finally we employ a duality argument and prove superconvergence estimates for the critical eigenvalue.
Neutron transport solver parallelization using a Domain Decomposition method
International Nuclear Information System (INIS)
Van Criekingen, S.; Nataf, F.; Have, P.
2008-01-01
A domain decomposition (DD) method is investigated for the parallel solution of the second-order even-parity form of the time-independent Boltzmann transport equation. The spatial discretization is performed using finite elements, and the angular discretization using spherical harmonic expansions (P N method). The main idea developed here is due to P.L. Lions. It consists in having sub-domains exchanging not only interface point flux values, but also interface flux 'derivative' values. (The word 'derivative' is here used with quotes, because in the case considered here, it in fact consists in the Ω.∇ operator, with Ω the angular variable vector and ∇ the spatial gradient operator.) A parameter α is introduced, as proportionality coefficient between point flux and 'derivative' values. This parameter can be tuned - so far heuristically - to optimize the method. (authors)
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)
Systems guide to MCNP (Monte Carlo Neutron and Photon Transport Code)
International Nuclear Information System (INIS)
Kirk, B.L.; West, J.T.
1984-06-01
The subject of this report is the implementation of the Los Alamos National Laboratory Monte Carlo Neutron and Photon Transport Code - Version 3 (MCNP) on the different types of computer systems, especially the IBM MVS system. The report supplements the documentation of the RSIC computer code package CCC-200/MCNP. Details of the procedure to follow in executing MCNP on the IBM computers, either in batch mode or interactive mode, are provided
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
Boron neutron capture therapy for advanced and/or recurrent cancers in the oral cavity
International Nuclear Information System (INIS)
Ariyoshi, Yasunori; Shimahara, Masashi; Kimura, Yoshihiro; Miyatake, Shin-ichi; Kuroiwa, Toshihiko; Nagata, Kenji; Sakurai, Yoshinori; Maruhashi, Akira; Ono, Koji
2006-01-01
This preliminary study of 5 patients with advanced and/or recurrent cancer in the oral cavity was performed to evaluate the effectiveness of Boron Neutron Capture Therapy (BNCT). The patients received therapy with the 10 B-carrier p-boronophenylalanine (BPA) with or without borocaptate sodium (BSH) and irradiation thereafter with epithermal neutrons. All underwent 18 F-BPA PET studies before receiving BNCT to determine the accumulation ratios of BPA in tumor and normal tissues. The tumor mass was decreased in size and at minimum a transient partial response was achieved in all cases, though rapid tumor re-growth was observed in 2. Although tentative clinical responses and improvements in quality of life were recognized, obliteration of the tumor was not obtained in any of the cases. Additional studies are required to determine the utility and indication of BNCT for oral cancer. (author)
Force analysis of the advanced neutron source control rod drive latch mechanism
International Nuclear Information System (INIS)
Damiano, B.
1989-01-01
The Advanced Neutron Source reactor (ANS), a proposed Department of Energy research reactor currently undergoing conceptual design at the Oak Ridge National Laboratory (ORNL), will generate a thermal neutron flux approximating 10 30 M -2 emdash S -1 . The compact core necessary to produce this flux provides little space for the shim safety control rods, which are located in the central annulus of the core. Without proper control rod drive design, the control rod drive magnets (which hold the control rod latch in a ready-to-scram position) may be unable to support the required load due to their restricted size. This paper describes the force analysis performed on the control rod latch mechanism to determine the fraction of control rod weight transferred to the drive magnet. This information will be useful during latch, control rod drive and magnet design. 5 refs., 12 figs
Post-merger evolution of a neutron star-black hole binary with neutrino transport
Foucart, Francois; O'Connor, Evan; Roberts, Luke; Duez, Matthew D.; Haas, Roland; Kidder, Lawrence E.; Ott, Christian D.; Pfeiffer, Harald P.; Scheel, Mark A.; Szilagyi, Bela
2015-06-01
We present a first simulation of the post-merger evolution of a black hole-neutron star binary in full general relativity using an energy-integrated general-relativistic truncated moment formalism for neutrino transport. We describe our implementation of the moment formalism and important tests of our code, before studying the formation phase of an accretion disk after a black hole-neutron star merger. We use as initial data an existing general-relativistic simulation of the merger of a neutron star of mass 1.4 M⊙ with a black hole of mass 7 M⊙ and dimensionless spin χBH=0.8 . Comparing with a simpler leakage scheme for the treatment of the neutrinos, we find noticeable differences in the neutron-to-proton ratio in and around the disk, and in the neutrino luminosity. We find that the electron neutrino luminosity is much lower in the transport simulations, and that both the disk and the disk outflows are less neutron rich. The spatial distribution of the neutrinos is significantly affected by relativistic effects, due to large velocities and curvature in the regions of strongest emission. Over the short time scale evolved, we do not observe purely neutrino-driven outflows. However, a small amount of material (3 ×10-4M⊙ ) is ejected in the polar region during the circularization of the disk. Most of that material is ejected early in the formation of the disk, and is fairly neutron rich (electron fraction Ye˜0.15 - 0.25 ). Through r-process nucleosynthesis, that material should produce high-opacity lanthanides in the polar region, and could thus affect the light curve of radioactively powered electromagnetic transients. We also show that by the end of the simulation, while the bulk of the disk remains neutron rich (Ye˜0.15 - 0.2 and decreasing), its outer layers have a higher electron fraction: 10% of the remaining mass has Ye>0.3 . As that material would be the first to be unbound by disk outflows on longer time scales, and as composition evolution is
Application of advanced handling techniques to transportation cask design
International Nuclear Information System (INIS)
Bennett, P.C.
1992-01-01
Sandia National Laboratories supports the US Department of Energy (DOE) Office of Civilian Radioactive Waste Management (OCRWM) applying technology to the safe transport of nuclear waste. Part of that development effort includes investigation of advanced handling technologies for automation of cask operations at nuclear waste receiving facilities. Although low radiation levels are expected near transport cask surfaces, cumulative occupational exposure at a receiving facility can be significant. Remote automated cask handling has the potential to reduce both the occupational exposure and the time necessary to process a cask. Thus, automated handling is consistent with DOE efforts to reduce the lifecycle costs of the waste disposal system and to maintain public and occupational radiological risks as low as reasonably achievable. This paper describes the development of advanced handling laboratory mock-ups and demonstrations for spent fuel casks. Utilizing the control enhancements described below, demonstrations have been carried out including cask location and identification, contact and non-contact surveys, impact limiter removal, tiedown release, uprighting, swing-free movement, gas sampling, and lid removal operations. Manually controlled movement around a cask under off-normal conditions has also been demonstrated
Solution to the monoenergetic time-dependent neutron transport equation with a time-varying source
International Nuclear Information System (INIS)
Ganapol, B.D.
1986-01-01
Even though fundamental time-dependent neutron transport problems have existed since the inception of neutron transport theory, it has only been recently that a reliable numerical solution to one of the basic problems has been obtained. Experience in generating numerical solutions to time-dependent transport equations has indicated that the multiple collision formulation is the most versatile numerical technique for model problems. The formulation coupled with a moment reconstruction of each collided flux component has led to benchmark-quality (four- to five-digit accuracy) numerical evaluation of the neutron flux in plane infinite geometry for any degree of scattering anisotropy and for both pulsed isotropic and beam sources. As will be shown in this presentation, this solution can serve as a Green's function, thus extending the previous results to more complicated source situations. Here we will be concerned with a time-varying source at the center of an infinite medium. If accurate, such solutions have both pedagogical and practical uses as benchmarks against which other more approximate solutions designed for a wider class of problems can be compared
Synergism of the method of characteristics and CAD technology for neutron transport calculation
International Nuclear Information System (INIS)
Chen, Z.; Wang, D.; He, T.; Wang, G.; Zheng, H.
2013-01-01
The method of characteristics (MOC) is a very popular methodology in neutron transport calculation and numerical simulation in recent decades for its unique advantages. One of the key problems determining whether the MOC can be applied in complicated and highly heterogeneous geometry is how to combine an effective geometry processing method with MOC. Most of the existing MOC codes describe the geometry by lines and arcs with extensive input data, such as circles, ellipses, regular polygons and combination of them. Thus they have difficulty in geometry modeling, background meshing and ray tracing for complicated geometry domains. In this study, a new idea making use of a CAD solid modeler MCAM which is a CAD/Image-based Automatic Modeling Program for Neutronics and Radiation Transport developed by FDS Team in China was introduced for geometry modeling and ray tracing of particle transport to remove these geometrical limitations mentioned above. The diamond-difference scheme was applied to MOC to reduce the spatial discretization error of the flat flux approximation in theory. Based on MCAM and MOC, a new MOC code was developed and integrated into SuperMC system, which is a Super Multi-function Computational system for neutronics and radiation simulation. The numerical testing results demonstrated the feasibility and effectiveness of the new idea for geometry treatment in SuperMC. (authors)
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
Criticality problems for slabs and spheres in energy dependent neutron transport theory
International Nuclear Information System (INIS)
Victory, H.D. Jr.
1980-01-01
The steady-state equation for energy-dependent neutron transport in isotropically scattering slabs and spheres is formulated as an integral equation. The Perron-Frobenius-Jentzsch theory of positive operators is used to analyze criticality problems for transport in slab and spherical media consisting of core and reflector. In addition, with an adroit selection of diffusion-like solutions, this theory is used to obtain an expression relating the critical radius of a homogeneous sphere to a parameter characterizing fission production. 21 refs
Neutron transport study based on assembly modular ray tracing MOC method
International Nuclear Information System (INIS)
Tian Chao; Zheng Youqi; Li Yunzhao; Li Shuo; Chai Xiaoming
2015-01-01
It is difficulty for the MOC method based on Cell Modular Ray Tracing to deal with the irregular geometry such as the water gap between the PWR lattices. Hence, the neutron transport code NECP-Medlar based on Assembly Modular Ray Tracing is developed. CMFD method is used to accelerate the transport calculation. The numerical results of the 2D C5G7 benchmark and typical PWR lattice prove that NECP-Medlar has an excellent performance in terms of accuracy and efficiency. Besides, NECP-Medlar can describe clearly the flux distribution of the lattice with water gap. (authors)
Phase 1 environmental report for the Advanced Neutron Source at Oak Ridge National Laboratory
International Nuclear Information System (INIS)
Blasing, T.J.; Brown, R.A.; Cada, G.F.; Easterly, C.; Feldman, D.L.; Hagan, C.W.; Harrington, R.M.; Johnson, R.O.; Ketelle, R.H.; Kroodsma, R.L.; McCold, L.N.; Reich, W.J.; Scofield, P.A.; Socolof, M.L.; Taleyarkhan, R.P.; Van Dyke, J.W.
1992-02-01
The US Department of Energy (DOE) has proposed the construction and operation of the Advanced Neutron Source (ANS), a 330-MW(f) reactor, at Oak Ridge National Laboratory (ORNL) to support neutron scattering and nuclear physics experiments. ANS would provide a steady-state source of neutrons that are thermalized to produce sources of hot, cold, and very coal neutrons. The use of these neutrons in ANS experiment facilities would be an essential component of national research efforts in basic materials science. Additionally, ANS capabilities would include production of transplutonium isotopes, irradiation of potential fusion and fission reactor materials, activation analysis, and production of medical and industrial isotopes such as 252 Cf. Although ANS would not require licensing by the US Nuclear Regulatory Commission (NRC), DOE regards the design, construction, and operation of ANS as activities that would produce a licensable facility; that is, DOE is following the regulatory guidelines that NRC would apply if NRC were licensing the facility. Those guidelines include instructions for the preparation of an environmental report (ER), a compilation of available data and preliminary analyses regarding the environmental impacts of nuclear facility construction and operation. The ER, described and outlined in NRC Regulatory Guide 4.2, serves as a background document to facilitate the preparation of environmental impact statements (EISs). Using Regulatory Guide 4.2 as a model, this ANS ER provides analyses and information specific to the ANS site and area that can be adopted (and modified, if necessary) for the ANS EIS. The ER is being prepared in two phases. Phase 1 ER includes many of the data and analyses needed to prepare the EIS but does not include data or analyses of alternate sites or alternate technologies. Phase 2 ER will include the additional data and analyses stipulated by Regulatory Guide 4.2
Phase 1 environmental report for the Advanced Neutron Source at Oak Ridge National Laboratory
Energy Technology Data Exchange (ETDEWEB)
Blasing, T.J.; Brown, R.A.; Cada, G.F.; Easterly, C.; Feldman, D.L.; Hagan, C.W.; Harrington, R.M.; Johnson, R.O.; Ketelle, R.H.; Kroodsma, R.L.; McCold, L.N.; Reich, W.J.; Scofield, P.A.; Socolof, M.L.; Taleyarkhan, R.P.; Van Dyke, J.W.
1992-02-01
The US Department of Energy (DOE) has proposed the construction and operation of the Advanced Neutron Source (ANS), a 330-MW(f) reactor, at Oak Ridge National Laboratory (ORNL) to support neutron scattering and nuclear physics experiments. ANS would provide a steady-state source of neutrons that are thermalized to produce sources of hot, cold, and very coal neutrons. The use of these neutrons in ANS experiment facilities would be an essential component of national research efforts in basic materials science. Additionally, ANS capabilities would include production of transplutonium isotopes, irradiation of potential fusion and fission reactor materials, activation analysis, and production of medical and industrial isotopes such as {sup 252}Cf. Although ANS would not require licensing by the US Nuclear Regulatory Commission (NRC), DOE regards the design, construction, and operation of ANS as activities that would produce a licensable facility; that is, DOE is following the regulatory guidelines that NRC would apply if NRC were licensing the facility. Those guidelines include instructions for the preparation of an environmental report (ER), a compilation of available data and preliminary analyses regarding the environmental impacts of nuclear facility construction and operation. The ER, described and outlined in NRC Regulatory Guide 4.2, serves as a background document to facilitate the preparation of environmental impact statements (EISs). Using Regulatory Guide 4.2 as a model, this ANS ER provides analyses and information specific to the ANS site and area that can be adopted (and modified, if necessary) for the ANS EIS. The ER is being prepared in two phases. Phase 1 ER includes many of the data and analyses needed to prepare the EIS but does not include data or analyses of alternate sites or alternate technologies. Phase 2 ER will include the additional data and analyses stipulated by Regulatory Guide 4.2.
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.)
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
Velarde, G.; Perlado, J. M.; Alonso, E.; Alonso, M.; Domínguez, E.; Rubiano, J. G.; Gil, J. M.; Gómez del Rio, J.; Lodi, D.; Malerba, L.; Marian, J.; Martel, P.; Martínez-Val, J. M.; Mínguez, E.; Piera, M.; Ogando, F.; Reyes, S.; Salvador, M.; Sanz, J.; Sauvan, P.; Velarde, M.; Velarde, P.
2001-05-01
The coupling of a new radiation transport (RT) solver with an existing multimaterial fluid dynamics code (ARWEN) using Adaptive Mesh Refinement named DAFNE, has been completed. In addition, improvements were made to ARWEN in order to work properly with the RT code, and to make it user-friendlier, including new treatment of Equations of State, and graphical tools for visualization. The evaluation of the code has been performed, comparing it with other existing RT codes (including the one used in DAFNE, but in the single-grid version). These comparisons consist in problems with real input parameters (mainly opacities and geometry parameters). Important advances in Atomic Physics, Opacity calculations and NLTE atomic physics calculations, with participation in significant experiments in this area, have been obtained. Early published calculations showed that a DT x fuel with a small tritium initial content ( x<3%) could work in a catalytic regime in Inertial Fusion Targets, at very high burning temperatures (≫100 keV). Otherwise, the cross-section of DT remains much higher than that of DD and no internal breeding of tritium can take place. Improvements in the calculation model allow to properly simulate the effect of inverse Compton scattering which tends to lower Te and to enhance radiation losses, reducing the plasma temperature, Ti. The neutron activation of all natural elements in First Structural Wall (FSW) component of an Inertial Fusion Energy (IFE) reactor for waste management, and the analysis of activation of target debris in NIF-type facilities has been completed. Using an original efficient modeling for pulse activation, the FSW behavior in inertial fusion has been studied. A radiological dose library coupled to the ACAB code is being generated for assessing impact of environmental releases, and atmospheric dispersion analysis from HIF reactors indicate the uncertainty in tritium release parameters. The first recognition of recombination barriers in Si
Energy Technology Data Exchange (ETDEWEB)
Velarde, G.; Perlado, J.M. E-mail: mperlado@denim.upm.es; Alonso, E.; Alonso, M.; Dominguez, E.; Rubiano, J.G.; Gil, J.M.; Gomez del Rio, J.; Lodi, D.; Malerba, L.; Marian, J.; Martel, P.; Martinez-Val, J.M.; Minguez, E.; Piera, M.; Ogando, F.; Reyes, S.; Salvador, M.; Sanz, J.; Sauvan, P.; Velarde, M.; Velarde, P
2001-05-21
The coupling of a new radiation transport (RT) solver with an existing multimaterial fluid dynamics code (ARWEN) using Adaptive Mesh Refinement named DAFNE, has been completed. In addition, improvements were made to ARWEN in order to work properly with the RT code, and to make it user-friendlier, including new treatment of Equations of State, and graphical tools for visualization. The evaluation of the code has been performed, comparing it with other existing RT codes (including the one used in DAFNE, but in the single-grid version). These comparisons consist in problems with real input parameters (mainly opacities and geometry parameters). Important advances in Atomic Physics, Opacity calculations and NLTE atomic physics calculations, with participation in significant experiments in this area, have been obtained. Early published calculations showed that a DT{sub x} fuel with a small tritium initial content (x<3%) could work in a catalytic regime in Inertial Fusion Targets, at very high burning temperatures ({>=}100 keV). Otherwise, the cross-section of DT remains much higher than that of DD and no internal breeding of tritium can take place. Improvements in the calculation model allow to properly simulate the effect of inverse Compton scattering which tends to lower T{sub e} and to enhance radiation losses, reducing the plasma temperature, T{sub i}. The neutron activation of all natural elements in First Structural Wall (FSW) component of an Inertial Fusion Energy (IFE) reactor for waste management, and the analysis of activation of target debris in NIF-type facilities has been completed. Using an original efficient modeling for pulse activation, the FSW behavior in inertial fusion has been studied. A radiological dose library coupled to the ACAB code is being generated for assessing impact of environmental releases, and atmospheric dispersion analysis from HIF reactors indicate the uncertainty in tritium release parameters. The first recognition of recombination
Some results on the neutron transport and the coupling of equations
International Nuclear Information System (INIS)
Bal, G.
1997-01-01
Neutron transport in nuclear reactors is well modeled by the linear Boltzmann transport equation. Its resolution is relatively easy but very expensive. To achieve whole core calculations, one has to consider simpler models, such as diffusion or homogeneous transport equations. However, the solutions may become inaccurate in particular situations (as accidents for instance). That is the reason why we wish to solve the equations on small area accurately and more coarsely on the remaining part of the core. It is than necessary to introduce some links between different discretizations or modelizations. In this note, we give some results on the coupling of different discretizations of all degrees of freedom of the integral-differential neutron transport equation (two degrees for the angular variable, on for the energy component, and two or three degrees for spatial position respectively in 2D (cylindrical symmetry) and 3D). Two chapters are devoted to the coupling of discrete ordinates methods (for angular discretization). The first one is theoretical and shows the well posing of the coupled problem, whereas the second one deals with numerical applications of practical interest (the results have been obtained from the neutron transport code developed at the R and D, which has been modified for introducing the coupling). Next, we present the nodal scheme RTN0, used for the spatial discretization. We show well posing results for the non-coupled and the coupled problems. At the end, we deal with the coupling of energy discretizations for the multigroup equations obtained by homogenization. Some theoretical results of the discretization of the velocity variable (well-posing of problems), which do not deal directly with the purposes of coupling, are presented in the annexes. (author)
Andreasen, M.; Looms, M. C.; Bogena, H. R.; Desilets, D.; Zreda, M. G.; Sonnenborg, T. O.; Jensen, K. H.
2014-12-01
The water stored in the various compartments of the terrestrial ecosystem (in snow, canopy interception, soil and litter) controls the exchange of the water and energy between the land surface and the atmosphere. Therefore, measurements of the water stored within these pools are critical for the prediction of e.g. evapotranspiration and groundwater recharge. The detection of cosmic-ray neutron intensity is a novel non-invasive method for the quantification of continuous intermediate-scale soil moisture. The footprint of the cosmic-ray neutron probe is a hemisphere of a few hectometers and subsurface depths of 10-70 cm depending on wetness. The cosmic-ray neutron method offers measurements at a scale between the point-scale measurements and large-scale satellite retrievals. The cosmic-ray neutron intensity is inversely correlated to the hydrogen stored within the footprint. Overall soil moisture represents the largest pool of hydrogen and changes in the soil moisture clearly affect the cosmic-ray neutron signal. However, the neutron intensity is also sensitive to variations of hydrogen in snow, canopy interception and biomass offering the potential to determine water content in such pools from the signal. In this study we tested the potential of determining canopy interception and biomass using cosmic-ray neutron intensity measurements within the framework of the Danish Hydrologic Observatory (HOBE) and the Terrestrial Environmental Observatories (TERENO). Continuous measurements at the ground and the canopy level, along with profile measurements were conducted at towers at forest field sites. Field experiments, including shielding the cosmic-ray neutron probes with cadmium foil (to remove lower-energy neutrons) and measuring reference intensity rates at complete water saturated conditions (on the sea close to the HOBE site), were further conducted to obtain an increased understanding of the physics controlling the cosmic-ray neutron transport and the equipment used
Advanced LIGO constraints on neutron star mergers and r-process sites
International Nuclear Information System (INIS)
Côté, Benoit; Belczynski, Krzysztof; Fryer, Chris L.; Ritter, Christian
2017-01-01
The role of compact binary mergers as the main production site of r-process elements is investigated by combining stellar abundances of Eu observed in the Milky Way, galactic chemical evolution (GCE) simulations, and binary population synthesis models, and gravitational wave measurements from Advanced LIGO. We compiled and reviewed seven recent GCE studies to extract the frequency of neutron star–neutron star (NS–NS) mergers that is needed in order to reproduce the observed [Eu/Fe] versus [Fe/H] relationship. We used our simple chemical evolution code to explore the impact of different analytical delay-time distribution functions for NS–NS mergers. We then combined our metallicity-dependent population synthesis models with our chemical evolution code to bring their predictions, for both NS–NS mergers and black hole–neutron star mergers, into a GCE context. Finally, we convolved our results with the cosmic star formation history to provide a direct comparison with current and upcoming Advanced LIGO measurements. When assuming that NS–NS mergers are the exclusive r-process sites, and that the ejected r-process mass per merger event is 0.01 M_⊙, the number of NS–NS mergers needed in GCE studies is about 10 times larger than what is predicted by standard population synthesis models. Here, these two distinct fields can only be consistent with each other when assuming optimistic rates, massive NS–NS merger ejecta, and low Fe yields for massive stars. For now, population synthesis models and GCE simulations are in agreement with the current upper limit (O1) established by Advanced LIGO during their first run of observations. Upcoming measurements will provide an important constraint on the actual local NS–NS merger rate, will provide valuable insights on the plausibility of the GCE requirement, and will help to define whether or not compact binary mergers can be the dominant source of r-process elements in the universe.
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.
Advanced Neutron Source Dynamic Model (ANSDM) code description and user guide
International Nuclear Information System (INIS)
March-Leuba, J.
1995-08-01
A mathematical model is designed that simulates the dynamic behavior of the Advanced Neutron Source (ANS) reactor. Its main objective is to model important characteristics of the ANS systems as they are being designed, updated, and employed; its primary design goal, to aid in the development of safety and control features. During the simulations the model is also found to aid in making design decisions for thermal-hydraulic systems. Model components, empirical correlations, and model parameters are discussed; sample procedures are also given. Modifications are cited, and significant development and application efforts are noted focusing on examination of instrumentation required during and after accidents to ensure adequate monitoring during transient conditions
Circular arc fuel plate stability experiments and analyses for the advanced neutron source
International Nuclear Information System (INIS)
Swinson, W.F.; Battiste, R.L.; Yahr, G.T.
1995-08-01
The thin fuel plates planned for the Advanced Neutron Source are to be cooled by forcing heavy water at high velocity, 25 m/s, through thin cooling channels on each side of each plate. Because the potential for structural failure of the plates is a design concern, considerable effort has been expended in assessing this potential. As part of this effort, experimental flow tests and analyses to evaluate the structural response of circular arc plates have been conducted, and the results are given in this report
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
Calculation of neutron spectra for a 252Cf transport cask using ANISN running on a PC
International Nuclear Information System (INIS)
West, L.; Akin, B.P.; Lemley, E.C.
1995-01-01
Neutron spectra have been calculated using the ANISN one-dimensional discrete ordinates code for the case of a 152 Cf source in a transport cask of a particular design. All computations were done on personal computers (PCs) (mostly 486 models) with the ANISN-ORNL (486 version) computer code. With a source of 252 Cf fission neutrons, the neutron flux spectrum in the cask cannot be characterized as open-quotes moderated.close quotes Concern about an appropriate choice for the cross-section data set has led to a comparison, for this application, of three different cross-section libraries: DABL, HILO, and BUGLE-80. Although the cross-section sets were not originally designed for PC use, the libraries have been successfully employed for PC computations. Work with yet another data library, BUGLE-93, is incomplete at this stage. From neutron flux spectra on the surface of the cask, personnel dosimetric quantities (such as dose equivalent) have been determined for the DABL, HILO, and BUGLE-80 ANISN calculations
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)
Takahashi, Akito; Yamamoto, Junji; Ebisuya, Mituo; Sumita, Kenji
1979-01-01
A new method for calculating the anisotropic neutron transport is proposed for the angular spectral analysis of D-T fusion reactor neutronics. The method is based on the transport equation with new type of anisotropic scattering kernels formulated by a single function I sub(i) (μ', μ) instead of polynomial expansion, for instance, Legendre polynomials. In the calculation of angular flux spectra by using scattering kernels with the Legendre polynomial expansion, we often observe the oscillation with negative flux. But in principle this oscillation disappears by this new method. In this work, we discussed anisotropic scattering kernels of the elastic scattering and the inelastic scatterings which excite discrete energy levels. The other scatterings were included in isotropic scattering kernels. An approximation method, with use of the first collision source written by the I sub(i) (μ', μ) function, was introduced to attenuate the ''oscillations'' when we are obliged to use the scattering kernels with the Legendre polynomial expansion. Calculated results with this approximation showed remarkable improvement for the analysis of the angular flux spectra in a slab system of lithium metal with the D-T neutron source. (author)
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)
Energy Technology Data Exchange (ETDEWEB)
Bal, G.
1995-07-01
To achieve whole core calculations of the neutron transport equation, we have to follow this 2 step method: space and energy homogenization of the assemblies; resolution of the homogenized equation on the whole core. However, this is no more valid when accidents occur (for instance depressurization causing locally strong heterogeneous media). One solution consists then in coupling two kinds of resolutions: a fine computation on the damaged cell (fine mesh, high number of energy groups) coupled with a coarse one everywhere else. We only deal here with steady state solutions (which already live in 6D spaces). We present here two such methods: The coupling by transmission of homogenized sections and the coupling by transmission of boundary conditions. To understand what this coupling is, we first restrict ourselves to 1D with respect to space in one energy group. The first two chapters deal with a recall of basic properties of the neutron transport equation. We give at chapter 3 some indications of the behaviour of the flux with respect to the cross sections. We present at chapter 4 some couplings and give some properties. Chapter 5 is devoted to a presentation of some numerical applications. (author). 9 refs., 7 figs.
International Nuclear Information System (INIS)
Khayat, M.I.; March-Leuba, J.
1993-01-01
This paper documents some of the optimization studies performed to maximize the performance of the engineered safety features and scram systems to mitigate the consequences of large breaks in the primary cooling system of the advanced neutron source (ANS) reactor. The ANS is a new basic and applied research facility based on a powerful steady-state research reactor that provides beams of neutrons for measurements and experiments in the field of material science and engineering, biology, chemistry, material analysis, and nuclear science. To achieve the high neutron fluxes for these state-of-the-art experiments, the ANS design has a very high power density core (330 MW fission with an active volume of 67.6 ell) surrounded by a large heavy-water reflector, where most neutrons are moderated. This design maximizes the number of neutrons available for experiments but results in a low heat capacity core that creates unique challenges to the design of the plant protection system
Janic, M.
2016-01-01
Multidimensional examination of performances of the future advanced ETT Evacuated Tube Transport) system operated by TRM (TransRapidMaglev); assessment of the ETT TRM system contribution to sustainability of the future transport sector through its completion with APT (Air Passenger Transport) system
International Nuclear Information System (INIS)
Daniehl', A.V.; Dushin, V.N.
1987-01-01
The methods for simulation of neutron transport with Z<20 MeV used in the SITHA (simulation transport hadron) program, the original library of group microconstants (175 groups) with subgroup description of resonance range and a set of programs for its creation are described. The results of a number of integral experiments are discussed
An advanced hadron facility: A combined kaon factory and cold-neutron source
International Nuclear Information System (INIS)
Thiessen, H.A.
1987-01-01
A design concept is presented for an advanced hadron facility consisting of a combined kaon factory and second generation spallation source. Our proposed facility consists of a 1.2 GeV superconducting H - linac to bring the LAMPF energy up to 2 GeV, a multi-ring 2 GeV compressor, a shared cold-neutron and stopped-pion neutrino source, a 60 GeV 25 μAmp 6 Hz proton synchrotron, and kaon and proton experimental areas. We discuss the considerations which led to this design concept. We summarize recent results of r and d work on components for rapid-cycling synchrotrons. Finally, we mention briefly a pion linac, which may be a good way to gain experience with superconducting cavities if advanced hadron facility funding is delayed
Initial global 2-D shielding analysis for the Advanced Neutron Source core and reflector
Energy Technology Data Exchange (ETDEWEB)
Bucholz, J.A.
1995-08-01
This document describes the initial global 2-D shielding analyses for the Advanced Neutron Source (ANS) reactor, the D{sub 2}O reflector, the reflector vessel, and the first 200 mm of light water beyond the reflector vessel. Flux files generated here will later serve as source terms in subsequent shielding analyses. In addition to reporting fluxes and other data at key points of interest, a major objective of this report was to document how these analyses were performed, the phenomena that were included, and checks that were made to verify that these phenomena were properly modeled. In these shielding analyses, the fixed neutron source distribution in the core was based on the `lifetime-averaged` spatial power distribution. Secondary gamma production cross sections in the fuel were modified so as to account intrinsically for delayed fission gammas in the fuel as well as prompt fission gammas. In and near the fuel, this increased the low-energy gamma fluxes by 50 to 250%, but out near the reflector vessel, these same fluxes changed by only a few percent. Sensitivity studies with respect to mesh size were performed, and a new 2-D mesh distribution developed after some problems were discovered with respect to the use of numerous elongated mesh cells in the reflector. All of the shielding analyses were performed sing the ANSL-V 39n/44g coupled library with 25 thermal neutron groups in order to obtain a rigorous representation of the thermal neutron spectrum throughout the reflector. Because of upscatter in the heavy water, convergence was very slow. Ultimately, the fission cross section in the various materials had to be artificially modified in order to solve this fixed source problem as an eigenvalue problem and invoke the Vondy error-mode extrapolation technique which greatly accelerated convergence in the large 2-D RZ DORT analyses. While this was quite effective, 150 outer iterations (over energy) were still required.
Initial global 2-D shielding analysis for the Advanced Neutron Source core and reflector
International Nuclear Information System (INIS)
Bucholz, J.A.
1995-08-01
This document describes the initial global 2-D shielding analyses for the Advanced Neutron Source (ANS) reactor, the D 2 O reflector, the reflector vessel, and the first 200 mm of light water beyond the reflector vessel. Flux files generated here will later serve as source terms in subsequent shielding analyses. In addition to reporting fluxes and other data at key points of interest, a major objective of this report was to document how these analyses were performed, the phenomena that were included, and checks that were made to verify that these phenomena were properly modeled. In these shielding analyses, the fixed neutron source distribution in the core was based on the 'lifetime-averaged' spatial power distribution. Secondary gamma production cross sections in the fuel were modified so as to account intrinsically for delayed fission gammas in the fuel as well as prompt fission gammas. In and near the fuel, this increased the low-energy gamma fluxes by 50 to 250%, but out near the reflector vessel, these same fluxes changed by only a few percent. Sensitivity studies with respect to mesh size were performed, and a new 2-D mesh distribution developed after some problems were discovered with respect to the use of numerous elongated mesh cells in the reflector. All of the shielding analyses were performed sing the ANSL-V 39n/44g coupled library with 25 thermal neutron groups in order to obtain a rigorous representation of the thermal neutron spectrum throughout the reflector. Because of upscatter in the heavy water, convergence was very slow. Ultimately, the fission cross section in the various materials had to be artificially modified in order to solve this fixed source problem as an eigenvalue problem and invoke the Vondy error-mode extrapolation technique which greatly accelerated convergence in the large 2-D RZ DORT analyses. While this was quite effective, 150 outer iterations (over energy) were still required
Advances in 14 MeV neutron activation analysis by means of a new intense neutron source
International Nuclear Information System (INIS)
Pepelnik, R.; Fanger, H.-U.; Michaelis, W.; Anders, B.
1982-01-01
A new intense 14 MeV neutron generator with cylindrical acceleration structure has been put in operation at the GKSS Research Center Geesthacht. The sealed neutron tube is combined with a fast pneumatic rabbit system with particular capabilities for neutron activation analysis involving short-lived reaction products. The sample transfer time is less than 140 ms. The maximum neutron flux available for activation is 5.2x10 10 n/cm 2 s. Theoretical sensitivity predictions made in a previous study have been verified for some important trace elements. As a first application, samples of freeze-dried suspended matter and fishes of the Elbe river were analyzed. (author)
Neutron and Gamma Shielding Evaluation for KN-12 Spent Nuclear Fuel Transport Cask
Energy Technology Data Exchange (ETDEWEB)
Cho, I. J.; Min, D. K.; Lee, J. C.; You, G. S.; Yoon, J. S. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Chang, G. H.; Jeong, Y. C.; Ko, Y. W. [Korea Hydro and Nuclear Power Co., LTD., Kori (Korea, Republic of)
2007-07-01
The CASTOR KN-12 is designed to transport 12 intact PWR spent fuel assemblies for dry and wet transportation conditions. The overall cask length is 480.1 cm with a wall thickness 37.5 cm. Shield for the KN-12 is maintained by the thick walled cask body and the lid. For neutron shielding, polyethylene rods (PE) are arranged in longitudinal boreholes in the vessel wall and PE-plates are inserted between the cask lid and lid side shock absorber and between the cask bottom and bottom steel plate. The shielding evaluation of the cask has been performed with MCNP to confirm the shielding integrity of cask for pre-service inspection of transport cask.
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.
International Nuclear Information System (INIS)
West, C.D.
1992-01-01
The International Group on Research Reactors (IGORR) was formed in 1990 to facilitate the sharing of knowledge and experience among those institutions and individuals who are actively working to design, build, and promote new research reactors or to make significant upgrades to existing facilities. The Advanced Neutron Source Project expects to complete conceptual design in mid-1992. In the present design concept, the neutron source is a heavy-water-cooled, moderated, and reflected reactor of about 350 MW(f) power. (author)
Advancement in tritium transport simulations for solid breeding blanket system
Energy Technology Data Exchange (ETDEWEB)
Ying, Alice, E-mail: ying@fusion.ucla.edu [Mechanical and Aerospace Engineering Department, UCLA, Los Angeles, CA 90095 (United States); Zhang, Hongjie [Mechanical and Aerospace Engineering Department, UCLA, Los Angeles, CA 90095 (United States); Merrill, Brad J. [Idaho National Laboratory, Idaho Falls, ID 83415 (United States); Ahn, Mu-Young [National Fusion Research Institute, Daejeon (Korea, Republic of)
2016-11-01
In this paper, advancement on tritium transport simulations was demonstrated for a solid breeder blanket HCCR TBS, where multi-physics and detailed engineering descriptions are considered using a commercial simulation code. The physics involved includes compressible purge gas fluid flow, heat transfer, chemical reaction, isotope swamping effect, and tritium isotopes mass transport. The strategy adopted here is to develop numerical procedures and techniques that allow critical details of material, geometric and operational heterogeneity in a most complete engineering description of the TBS being incorporated into the simulation. Our application focuses on the transient assessment in view of ITER being pulsed operations. An immediate advantage is a more realistic predictive and design analysis tool accounting pulsed operations induced temperature variations which impact helium purge gas flow as well as Q{sub 2} composition concentration time and space evolutions in the breeding regions. This affords a more accurate prediction of tritium permeation into the He coolant by accounting correct temperature and partial pressure effects and realistic diffusion paths. The analysis also shows that by introducing by-pass line to accommodate ITER pulsed operations in the TES loop allows tritium extraction design being more cost effective.
International Nuclear Information System (INIS)
Cao Liangzhi; Wu Hongchun; Zheng Youqi
2008-01-01
Daubechies' wavelet expansion is introduced to discretize the angular variables of the neutron transport equation when the neutron angular flux varies very acutely with the angular directions. An improvement is made by coupling one-dimensional wavelet expansion and discrete ordinate method to make two-dimensional angular discretization efficient and stable. The angular domain is divided into several subdomains for treating the vacuum boundary condition exactly in the unstructured geometry. A set of wavelet equations coupled with each other is obtained in each subdomain. An iterative method is utilized to decouple the wavelet moments. The numerical results of several benchmark problems demonstrate that the wavelet expansion method can provide more accurate results by lower-order expansion than other angular discretization methods
International Nuclear Information System (INIS)
Kobayashi, Keisuke
1977-01-01
A method of solution of a monoenergetic neutron transport equation in P sub(L) approximation is presented for x-y and x-y-z geometries using the finite Fourier transformation. A reactor system is assumed to consist of multiregions in each of which the nuclear cross sections are spatially constant. Since the unknown functions of this method are the spherical harmonics components of the neutron angular flux at the material boundaries alone, the three- and two-dimensional equations are reduced to two- and one-dimensional equations, respectively. The present approach therefore gives fewer unknowns than in the usual series expansion method or in the finite difference method. Some numerical examples are shown for the criticality problem. (auth.)
One-speed neutron transport in spheres with totally absorbing cores
International Nuclear Information System (INIS)
Sjoestrand, N.G.
1988-01-01
Stationary and time-dependent transport of neutrons of one speed has been studied in spheres with totally absorbing cores. For stationary, critical reactors the number of secondaries per collision has been calculated numerically for various inner and outer radii. In the time-dependent case, the decay constant has been calculated for spherical shells of different inner radii and thicknesses. For a fixed ratio between shell thickness and inner radius, the curve of the decay constant versus shell thickness crosses the Corngold limit in the same way as the curve for a homogeneous sphere. When the ratio goes to zero the curve approaches that for an infinite slab. The behaviour is discussed in view of a new result from collision theory, viz. that the following condition must be fulfilled for a body at the point where the decay constant curve crosses the Corngold limit: the average exit distance of the neutrons is equal to the mean free path for scattering
Consumer Views on Transportation and Advanced Vehicle Technologies
Energy Technology Data Exchange (ETDEWEB)
Singer, Mark [National Renewable Energy Lab. (NREL), Golden, CO (United States)
2015-09-01
Vehicle manufacturers, U.S. Department of Energy laboratories, universities, private researchers, and organizations from countries around the globe are pursuing advanced vehicle technologies that aim to reduce gasoline and diesel consumption. This report details study findings of broad American public sentiments toward issues surrounding advanced vehicle technologies and is supported by the U.S. Department of Energy Vehicle Technology Office (VTO) in alignment with its mission to develop and deploy these technologies to improve energy security, increase mobility flexibility, reduce transportation costs, and increase environmental sustainability. Understanding and tracking consumer sentiments can influence the prioritization of development efforts by identifying barriers to and opportunities for broad acceptance of new technologies. Predicting consumer behavior toward developing technologies and products is inherently inexact. A person's stated preference given in an interview about a hypothetical setting may not match the preference that is demonstrated in an actual situation. This difference makes tracking actual consumer actions ultimately more valuable in understanding potential behavior. However, when developing technologies are not yet available and actual behaviors cannot be tracked, stated preferences provide some insight into how consumers may react in new circumstances. In this context this report provides an additional source to validate data and a new resource when no data are available. This report covers study data captured from December 2005 through June 2015 relevant to VTO research efforts at the time of the studies. Broadly the report covers respondent sentiments about vehicle fuel economy, future vehicle technology alternatives, ethanol as a vehicle fuel, plug-in electric vehicles, and willingness to pay for vehicle efficiency. This report represents a renewed effort to publicize study findings and make consumer sentiment data available to
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
International Nuclear Information System (INIS)
Wemple, C.A.; Schnitzler, B.G.; Ryskamp, J.M.
1995-08-01
A summary of the methods and models used to perform neutronics analyses on the Advanced Neutron Source reactor three-element core design is presented. The applications of the neutral particle Monte Carlo code MCNP are detailed, as well as the expansion of the static role of MCNP to analysis of fuel cycle depletion calculations. Results to date of these applications are presented also. A summary of the calculations not yet performed is also given to provide a open-quotes to-doclose quotes list if the project is resurrected
DEFF Research Database (Denmark)
Andersen, Jacob; Kristensen, Anders Skov; Bang-Andersen, Benny
2009-01-01
The biogenic monoamine transporters are integral membrane proteins that perform active transport of extracellular dopamine, serotonin and norepinephrine into cells. These transporters are targets for therapeutic agents such as antidepressants, as well as addictive substances such as cocaine...... and amphetamine. Seminal advances in the understanding of the structure and function of this transporter family have recently been accomplished by structural studies of a bacterial transporter, as well as medicinal chemistry and pharmacological studies of mammalian transporters. This feature article focuses...
International Nuclear Information System (INIS)
Massimiliano, Rosa; Azmy, Y.Y.; Morel, J.E.
2005-01-01
solution of the neutron transport equation. (authors)
International Nuclear Information System (INIS)
Jones, D.B.
1986-01-01
EPRI-LATTICE is a multigroup neutron transport computer code for the analysis of light water reactor fuel assemblies. It can solve the two-dimensional neutron transport problem by two distinct methods: (a) the method of collision probabilities and (b) the method of discrete ordinates. The code was developed by S. Levy Inc. as an account of work sponsored by the Electric Power Research Institute (EPRI). The collision probabilities calculation in EPRI-LATTICE (L-CP) is based on the same methodology that exists in the lattice codes CPM-2 and EPRI-CPM. Certain extensions have been made to the data representations of the CPM programs to improve the overall accuracy of the calculation. The important extensions include unique representations of scattering matrices and fission fractions (chi) for each composition in the problem. A new capability specifically developed for the EPRI-LATTICE code is a discrete ordinates methodology. The discrete ordinates calculation in EPRI-LATTICE (L-SN) is based on the discrete S/sub n/ methodology that exists in the TWODANT program. In contrast to TWODANT, which utilizes synthetic diffusion acceleration and supports multiple geometries, only the transport equations are solved by L-SN and only the data representations for the two-dimensional geometry are treated
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
Coogan, J. J.
1986-01-01
Modifications were designed for the B-737-100 Research Aircraft autobrake system hardware of the Advanced Transport Operating Systems (ATOPS) Program at Langley Research Center. These modifications will allow the on-board flight control computer to control the aircraft deceleration after landing to a continuously variable level for the purpose of executing automatic high speed turn-offs from the runway. A bread board version of the proposed modifications was built and tested in simulated stopping conditions. Test results, for various aircraft weights, turnoff speed, winds, and runway conditions show that the turnoff speeds are achieved generally with errors less than 1 ft/sec.
Advances in methods for identification and characterization of plant transporter function
DEFF Research Database (Denmark)
Larsen, Bo; Xu, Deyang; Halkier, Barbara Ann
2017-01-01
Transport proteins are crucial for cellular function at all levels. Numerous importers and exporters facilitate transport of a diverse array of metabolites and ions intra- and intercellularly. Identification of transporter function is essential for understanding biological processes at both......-based approaches. In this review, we highlight examples that illustrate how new technology and tools have advanced identification and characterization of plant transporter functions....
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)
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)
SERA -- An advanced treatment planning system for neutron therapy and BNCT
International Nuclear Information System (INIS)
Nigg, D.W.; Wemple, C.A.; Wessol, D.E.; Wheeler, F.J.; Albright, C.; Cohen, M.; Frandsen, M.; Harkin, G.; Rossmeier, M.
1999-01-01
Detailed treatment planning calculations on a patient-specific basis are required for boron neutron capture therapy (BNCT). Two integrated treatment planning systems developed specifically for BNCT have been in clinical use in the United States over the past few years. The MacNCTPLAN BNCT treatment planning system is used in the clinical BNCT trials that are underway at the Massachusetts Institute of Technology. A second system, BNCT rtpe (BNCT radiation therapy planning environment), developed independently by the Idaho national Engineering and Environmental Laboratory (INEEL) in collaboration with Montana State University (MSU), is used for treatment planning in the current series of BNCT clinical trials for glioblastoma at Brookhaven National Laboratory (BNL). This latter system is also licensed for use at several other BNCT research facilities worldwide. Although the currently available BNCT planning systems have served their purpose well, they suffer from somewhat long computation times (2 to 3 CPU-hours or more per field) relative to standard photon therapy planning software. This is largely due to the need for explicit three-dimensional solutions to the relevant transport equations. The simplifying approximations that work well for photon transport computations are not generally applicable to neutron transport computations. Greater computational speeds for BNCT treatment planning must therefore generally be achieved through the application of improved numerical techniques rather than by simplification of the governing equations. Recent efforts at INEEL and MSU have been directed toward this goal. This has resulted in a new paradigm for this type of calculation and the subsequent creation of the new simulation environment for radiotherapy applications (SERA) treatment planning system for BNCT. SERA is currently in initial clinical testing in connection with the trials at BNL, and it is expected to replace the present BNCT rtpe system upon general release
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
Shternin, P. S.; Baldo, M.; Schulze, H.-J.
2017-12-01
Thermal conductivity and shear viscosity of npeµ matter in non-superfluid neutron star cores are considered in the framework of Brueckner-Hartree-Fock many-body theory. We extend our previous work (Shternin et al 2013 PRC 88 065803) by analysing different nucleon-nucleon potentials and different three-body forces. We find that the use of different potentials leads up to one order of magnitude variations in the values of the nucleon contribution to transport coefficients. The nucleon contribution dominates the thermal conductivity, but for all considered models the shear viscosity is dominated by leptons.
Comparison of two Ssub(infinity) methods for solving the neutron transport equation
International Nuclear Information System (INIS)
Mennig, J.; Brandt, D.; Haelg, W.
1978-01-01
A semianalytic method (S 0 sub(infinity)) is presented for solving the monoenergetic multi-region transport equation. This method is compared with results from S 1 sub(infinity)-theory given in the literature. Application of S 1 sub(infinity)-theory to reactor shields may lead to negative neutron fluxes and to flux oscillations. These unphysical effects are completely avoided by the new method. Numerical results demonstrate the limitations of S 1 sub(infinity) and confirm the numerical stability of (S 0 sub(infinity)). (Auth.)
Solution of the neutron transport equation by means of Hermite-Ssub(infinity)-theory
International Nuclear Information System (INIS)
Brandt, D.; Haelg, W.; Mennig, J.
1979-01-01
A stable numerical approximation Hsub(α)-Ssub(infinity) is obtained through the use of Hermite's method of order α(Hsub(α)) in the spatial integration of the ID neutron transport equation. The theory for α = 1 is applied to a one-group shielding problem. Numerical calculations show the new method to converge much faster than earlier versions of Ssub(infinity)-theory. Comparison of H 1 - Ssub(infinity) with the well-known Ssub(N)-code ANISN indicates a large gain in computing time for the former. (Auth.)