International Nuclear Information System (INIS)
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
Advanced method of solution of neutron transport equation in nuclear reactor cell - 361
International Nuclear Information System (INIS)
Method of solution of neutron transport integral equation has been developed. It is aimed into calculation analysis of neutron flux in nuclear reactor cell with complicated geometry and different boundary conditions. On this stage of nuclear reactor calculation it is important to take into account special futures of neutron flux behavior included anisotropy scattering. Modern computational strategy requires the ability to accurately solution of Boltzmann transport equation in the shortest possible time. This approach is based on neutron flux expansion with orthogonal polynomial system in every uniform mesh of the cell. As result of this approximation the system of linear integral equation is reduced to algebraic system with coefficients that are the six-fold integrals over the cell area in general case. In this paper formulae for calculation of these values are given. The algorithm of computer code for neutron flux calculation is described. The results obtained with general version of collision probabilities method code are given. The advantage of above described approach has been demonstrated. (authors)
Advanced Neutron Source (ANS) Project progress report
International Nuclear Information System (INIS)
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
Advanced Neutron Source (ANS) Project progress report
Energy Technology Data Exchange (ETDEWEB)
McBee, M.R.; Chance, C.M. (eds.) (Oak Ridge National Lab., TN (USA)); Selby, D.L.; Harrington, R.M.; Peretz, F.J. (Oak Ridge National Lab., TN (USA))
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 C research and development; facility concepts; design; and safety.
Recent advances in neutron tomography
International Nuclear Information System (INIS)
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
Advances in neutron tomography
Indian Academy of Sciences (India)
W Treimer
2008-11-01
In the last decade neutron radiography (NR) and tomography (NCT) have experienced a number of improvements, due to the well-known properties of neutrons interacting with matter, i.e. the low attenuation by many materials, the strong attenuation by hydrogenous constituent in samples, the wavelength-dependent attenuation in the neighbourhood of Bragg edges and due to better 2D neutron detectors. So NR and NCT were improved by sophisticated techniques that are based on the attenuation of neutrons or on phase changes of the associated neutron waves if they pass through structured materials. Up to now the interaction of the neutron spin with magnetic fields in samples has not been applied to imaging techniques despite the fact that it was proposed many years ago. About ten years ago neutron depolarization as imaging signal for neutron radiography or tomography was demonstrated and in principle it works. Now one can present much improved test experiments using polarized neutrons for radiographic imaging. For this purpose the CONRAD instrument of the HMI was equipped with polarizing and analysing benders very similar to conventional scattering experiments using polarized neutrons. Magnetic fields in different coils and in samples (superconductors) at low temperatures could be visualized. In this lecture a summary about standard signals (attenuation) and the more `sophisticated' imaging signals as refraction, small angle scattering and polarized neutrons will be given.
Advances in neutron scattering research
International Nuclear Information System (INIS)
This issue of the Supplement to Journal of the Physical Society of Japan collects invited and contributed papers from the first International Symposium on Advanced Science Research (ASR-2000) 'Advances in Neutron Scattering Research'. The 182 of the presented papers are indexed individually. (J.P.N.)
ANEMONA: multiassembly neutron transport modeling
Energy Technology Data Exchange (ETDEWEB)
Jevremovic, T.; Ito, T. E-mail: t-itoh@nfi.co.jp; Inaba, Y
2002-11-01
A new feature of the general geometry neutron transport code, ANEMONA, the modeling of multi-assembly geometries in 2D, is developed and presented in this paper. The new module is called the ANEMULT code. In addition, the two acceleration techniques are added: (a) the ANEMONA's original geometry independent ray tracer (GIT), now utilizes the, so called, virtual bounding volume concept that importantly speeds up the ray tracing, and (b) the flux solver is accelerated using the Chebyshev polynomials. A whole core configuration run by ANEMULT is generated linking assemblies through the boundary edges' flux. All geometrical data are prepared in advance running the ANEMONA code (independently for geometrically different assemblies only). In this paper, two numerical benchmarks are presented: a single BWR MOX fuel assembly and a 6x6 assembly geometry (each assembly is of BWR 9x9 type). The results compared with the Monte Carlo code, GMVP, show a very good agreement.
3-D neutron transport benchmarks
International Nuclear Information System (INIS)
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 Keff, 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
Christl, Mark; Dobson, Chris; Norwood, Joseph; Kayatin, Matthew; Apple, Jeff; Gibson, Brian; Dietz, Kurt; Benson, Carl; Smith, Dennis; Howard, David; Rodriquez, Miguel; Watts, John; Sabra, Mohammed; Kuznetsov, Evgeny
2013-01-01
Energetic neutron measurements remain a challenge for space science investigations and radiation monitoring for human exploration beyond LEO. We are investigating a new composite scintillator design that uses Li6 glass scintillator embedded in a PVT block. A comparison between Li6 and Boron 10 loaded scintillators are being studied to assess the advantages and shortcomings of these two techniques. We present the details of the new Li6 design and results from the comparison of the B10 and Li6 techniques during exposures in a mixed radiation field produced by high energy protons interacting in a target material.
Recent Advances in Neutron Physics
Feshbach, Herman; Sheldon, Eric
1977-01-01
Discusses new studies in neutron physics within the last decade, such as ultracold neutrons, neutron bottles, resonance behavior, subthreshold fission, doubly radiative capture, and neutron stars. (MLH)
Linear stochastic neutron transport theory
International Nuclear Information System (INIS)
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)
International Nuclear Information System (INIS)
This paper presents a synthesis of the latest advances in the Verification and Validation (V and V) process of the new French (CEA) deterministic neutron transport code APOLLO3® developed within the framework of a common CEA, AREVA and EDF project. It focuses more precisely on the generic V and V of the main transport flux solvers of the code (namely IDT, Minaret, Pastis, TDT and Minos,) through 1D to 3D international benchmarks (ZPR-1D, Stepanek, C5G7, Takeda). Precise criteria have been defined to assess the quality of each solver by comparison with TRIPOLI4® multigroup Monte-Carlo calculations that have been performed for each configuration. We show that pure transport flux solvers (IDT, Minaret, Pastis and TDT-MOC) based on Sn , Pn and characteristics methods meet the keff target precision criteria (100 pcm) whereas SPn solver (Minos) give satisfactory results within reasonable computation time. The complementary of the APOLLO3® flux solvers set is globally highlighted. (author)
Advanced Neutron Source operating philosophy
International Nuclear Information System (INIS)
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
Coupled neutron transport for HZETRN
Energy Technology Data Exchange (ETDEWEB)
Slaba, T.C., E-mail: Tony.C.Slaba@nasa.go [Old Dominion University, Norfolk, VA 23505 (United States); Blattnig, S.R. [NASA Langley Research Center, Hampton, VA 23681 (United States); Aghara, S.K. [Prairie View A and M University, Prairie View, TX 77446 (United States); Townsend, L.W.; Handler, T. [University of Tennessee, Knoxville, TN 37996 (United States); Gabriel, T.A. [Scientific Investigation and Development, Knoxville, TN 37922 (United States); Pinsky, L.S.; Reddell, B. [University of Houston, Houston, TX 77204 (United States)
2010-02-15
Exposure estimates inside space vehicles, surface habitats, and high altitude aircrafts exposed to space radiation are highly influenced by secondary neutron production. The deterministic transport code HZETRN has been identified as a reliable and efficient tool for such studies, but improvements to the underlying transport models and numerical methods are still necessary. In this paper, the forward-backward (FB) and directionally coupled forward-backward (DC) neutron transport models are derived, numerical methods for the FB model are reviewed, and a computationally efficient numerical solution is presented for the DC model. Both models are compared to the Monte Carlo codes HETC-HEDS, FLUKA, and MCNPX, and the DC model is shown to agree closely with the Monte Carlo results. Finally, it is found in the development of either model that the decoupling of low energy neutrons from the light ion transport procedure adversely affects low energy light ion fluence spectra and exposure quantities. A first order correction is presented to resolve the problem, and it is shown to be both accurate and efficient.
Coupled Neutron Transport for HZETRN
Slaba, Tony C.; Blattnig, Steve R.
2009-01-01
Exposure estimates inside space vehicles, surface habitats, and high altitude aircrafts exposed to space radiation are highly influenced by secondary neutron production. The deterministic transport code HZETRN has been identified as a reliable and efficient tool for such studies, but improvements to the underlying transport models and numerical methods are still necessary. In this paper, the forward-backward (FB) and directionally coupled forward-backward (DC) neutron transport models are derived, numerical methods for the FB model are reviewed, and a computationally efficient numerical solution is presented for the DC model. Both models are compared to the Monte Carlo codes HETC-HEDS, FLUKA, and MCNPX, and the DC model is shown to agree closely with the Monte Carlo results. Finally, it is found in the development of either model that the decoupling of low energy neutrons from the light particle transport procedure adversely affects low energy light ion fluence spectra and exposure quantities. A first order correction is presented to resolve the problem, and it is shown to be both accurate and efficient.
Advanced Neutron Source (ANS) Project
International Nuclear Information System (INIS)
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
Neutron measurement by transportable spectrometer
International Nuclear Information System (INIS)
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 CUTTINGS TRANSPORT STUDY
Energy Technology Data Exchange (ETDEWEB)
Troy Reed; Stefan Miska; Nicholas Takach; Kaveh Ashenayi; Mark Pickell; Len Volk; Mike Volk; Evren Ozbayoglu; Lei Zhou
2002-04-30
This is the third quarterly progress report for Year 3 of the ACTS Project. It includes a review of progress made in: (1) Flow Loop construction and development and (2) research tasks during the period of time between Jan. 1, 2002 and Mar. 31, 2002. This report presents a review of progress on the following specific tasks: (a) Design and development of an Advanced Cuttings Transport Facility (Task 3: Addition of a Cuttings Injection/Separation System), (b) Research project (Task 6): ''Study of Cuttings Transport with Foam Under LPAT Conditions (Joint Project with TUDRP)'', (c) Research project (Task 9b): ''Study of Foam Flow Behavior Under EPET Conditions'', (d) Research project (Task 10): ''Study of Cuttings Transport with Aerated Mud Under Elevated Pressure and Temperature Conditions'', (e) Research on three instrumentation tasks to measure: Cuttings concentration and distribution in a flowing slurry (Task 11), Foam texture while transporting cuttings. (Task 12), and Viscosity of Foam under EPET (Task 9b); (f) 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); and (g) Activities towards technology transfer and developing contacts with Petroleum and service company members, and increasing the number of JIP members.
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; Affonso Lourenco; Evren Ozbayoglu; Lei Zhou
2002-01-30
This is the second quarterly progress report for Year 3 of the ACTS project. It includes a review of progress made in: (1) Flow Loop development and (2) research tasks during the period of time between Oct 1, 2001 and Dec. 31, 2001. This report presents a review of progress on the following specific tasks: (a) Design and development of an Advanced Cuttings Transport Facility (Task 3: Addition of a Cuttings Injection/Collection System), (b) Research project (Task 6): ''Study of Cuttings Transport with Foam Under LPAT Conditions (Joint Project with TUDRP)'', (c) Research project (Task 9): ''Study of Foam Flow Behavior Under EPET Conditions'', (d) Research project (Task 10): ''Study of Cuttings Transport with Aerated Mud Under Elevated Pressure and Temperature Conditions'', (e) Research on instrumentation tasks to measure: Cuttings concentration and distribution in a flowing slurry (Task 11), and Foam properties while transporting cuttings. (Task 12), (f) 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). (g) Activities towards technology transfer and developing contacts with Petroleum and service company members, and increasing the number of JIP members.
The advanced neutron source (ANS) project
International Nuclear Information System (INIS)
The Advanced Neutron Source (ANS) is a new user experimental facility for neutron research planned at Oak Ridge. The centerpiece of the facility will be a steady-state source of neutrons from a reactor of unprecedented flux. In addition, extensive and comprehensive equipment and facilities for neutron research will be included. The scientific fields to be served include neutron scattering with cold, thermal, and hot neutrons (the most important scientific justification for the project); engineering materials irradiation; isotope production (including transuranium isotopes); materials analysis; and nuclear science
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.
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
2000-10-30
This is the first 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 July 14, 2000 and September 30, 2000. This report presents information on the following specific tasks: (a) Progress in Advanced Cuttings Transport Facility design and development (Task 2), (b) Progress on research project (Task 8): ''Study of Flow of Synthetic Drilling Fluids Under Elevated Pressure and Temperature Conditions'', (c) Progress on research project (Task 6): ''Study of Cuttings Transport with Foam Under LPAT Conditions (Joint Project with TUDRP)'', (d) Progress on research project (Task 7): ''Study of Cuttings Transport with Aerated Muds Under LPAT Conditions (Joint Project with TUDRP)'', (e) Progress on research project (Task 9): ''Study of Foam Flow Behavior Under EPET Conditions'', (f) Initiate research on project (Task 10): ''Study of Cuttings Transport with Aerated Mud Under Elevated Pressure and Temperature Conditions'', (g) Progress on instrumentation tasks to measure: Cuttings concentration and distribution (Tasks 11), and Foam properties (Task 12), (h) Initiate a comprehensive safety review of all flow-loop components and operational procedures. Since the previous Task 1 has been completed, we will now designate this new task as: (Task 1S). (i) Activities towards technology transfer and developing contacts with Petroleum and service company members, and increasing the number of JIP members.
Advanced Neutron Source (ANS) Project Progress report, FY 1991
International Nuclear Information System (INIS)
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
Advanced Neutron Source (ANS) Project Progress report, FY 1991
Energy Technology Data Exchange (ETDEWEB)
Campbell, J.H. (ed.) (Oak Ridge National Lab., TN (United States)); Selby, D.L.; Harrington, R.M. (Oak Ridge National Lab., TN (United States)); Thompson, P.B. (Martin Marietta Energy Systems, Inc., (United States). Engineering Division)
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 C Research and Development; Design; and Safety.
Advanced Neutron Source (ANS) Project Progress report, FY 1991
Energy Technology Data Exchange (ETDEWEB)
Campbell, J.H. [ed.] [Oak Ridge National Lab., TN (United States); Selby, D.L.; Harrington, R.M. [Oak Ridge National Lab., TN (United States); Thompson, P.B. [Martin Marietta Energy Systems, Inc., (United States). Engineering Division
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 & C Research and Development; Design; and Safety.
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
Neutron transport with periodic boundary conditions
Energy Technology Data Exchange (ETDEWEB)
Angelescu, N.; Marinescu, N.; Protopopescu, V.
1976-01-01
The initial value problem for monoenergetic neutron transport in homogeneous nonmultiplying, nonabsorbing medium with isotropic scattering and periodic boundary conditions. One completely determines the structure of the spectrum of the transport operator both in plane and parallelepipedic geometries.
Some improved methods in neutron transport theory
International Nuclear Information System (INIS)
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
Neutron transport equation - indications on homogenization and neutron diffusion
International Nuclear Information System (INIS)
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
An advanced neutron radiography system
International Nuclear Information System (INIS)
The Stationary Neutron Radiography System (SNRS) nuclear reactor and radiography systems and their performance are described. The primary mission of the SNRS is to conduct neutron radiographic inspections of aircraft components to detect corrosion and moisture. Preliminary measurements indicate that the facility is capable of producing high quality real-time and film radiography. The reactor is capable of providing various additional services including sample irradiations, nuclear harness testing, in-core irradiations, in-core pneumatic rabbit system irradiations, neutron activation analysis, and pulse and square wave operation. 2 refs
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.
Development of advanced neutron beam technology
International Nuclear Information System (INIS)
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
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.
Advances in neutron based bulk explosive detection
International Nuclear Information System (INIS)
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
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.
Neutron stars - cooling and transport
Potekhin, A Y; Page, Dany
2015-01-01
Observations of thermal radiation from neutron stars can potentially provide information about the states of supranuclear matter in the interiors of these stars with the aid of the theory of neutron-star thermal evolution. We review the basics of this theory for isolated neutron stars with strong magnetic fields, including most relevant thermodynamic and kinetic properties in the stellar core, crust, and blanketing envelopes.
Development of transient neutron transport calculation code
International Nuclear Information System (INIS)
A transient neutron transport code for time-dependent analyses of neutronics systems, named DOT4-T, has been developed. The code is based on the Discrete Ordinates code DOT4.2, which solves the steady-state neutron transport equation in two dimensions. For the discretization of time variable, a direct method, the fully implicit and unconditionally stable time integration scheme, has been employed. The resulting code has been tested using several one-dimensional and two-dimensional benchmark problems, and the results obtained with DOT4-T shows very satisfactory agreement with the benchmark problem results. (authors)
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 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)
Troy Reed; Stefan Miska; Nicholas Takach; Kaveh Ashenayi; Mark Pickell; Len Volk; Mike Volk; Lei Zhou; Zhu Chen; Crystal Redden; Aimee Washington
2003-04-30
Experiments on the flow loop are continuing. Improvements to the software for data acquisition are being made as additional experience with three-phase flow is gained. Modifications are being made to the Cuttings Injection System in order to improve control and the precision of cuttings injection. The design details for a drill-pipe Rotation System have been completed. A US Patent was filed on October 28, 2002 for a new design for an instrument that can generate a variety of foams under elevated pressures and temperatures and then transfer the test foam to a viscometer for measurements of viscosity. Theoretical analyses of cuttings transport phenomena based on a layered model is under development. Calibrations of two nuclear densitometers have been completed. Baseline tests have been run to determine wall roughness in the 4 different tests sections (i.e. 2-in, 3-in, 4-in pipes and 5.76-in by 3.5-in annulus) of the flow loop. Tests have also been conducted with aerated fluids at EPET conditions. Preliminary experiments on the two candidate aqueous foam formulations were conducted which included rheological tests of the base fluid and foam stability reports. These were conducted after acceptance of the proposal on the Study of Cuttings Transport with Foam Under Elevated Pressure and Elevated Temperature Conditions. Preparation of a test matrix for cuttings-transport experiments with foam in the ACTF is also under way. A controller for instrumentation to measure cuttings concentration and distribution has been designed that can control four transceivers at a time. A prototype of the control circuit board was built and tested. Tests showed that there was a problem with radiated noise. AN improved circuit board was designed and sent to an external expert to verify the new design. The new board is being fabricated and will first be tested with static water and gravel in an annulus at elevated temperatures. A series of viscometer tests to measure foam properties have
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
A New Monte Carlo Neutron Transport Code at UNIST
International Nuclear Information System (INIS)
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
Onsager equations and time dependent neutron transport
International Nuclear Information System (INIS)
The diffusion of neutrons following an abrupt, localized temperature fluctuation can be conducted in the framework of Onsager-type transport equations. Considering Onsager equations as a generalized Fick's law, time-dependent particle and energy 'generalized diffusion equations' can be obtained. Aim of the present paper is to obtain the time-dependent diffusion Onsager-type equations for the diffusion of neutrons and to apply them to simple trial cases to gain a feeling for their behaviour. (author)
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.
Study of a transportable neutron radiography system
International Nuclear Information System (INIS)
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 107 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 102 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 102 n.cm-2.s-1 and 2,65 x 102 n.cm-2.s-1, respectively. (author). 30 refs, 39 figs, 9 tabs
ALADIN - Advanced Laue Diffraction Instruments using Neutrons
International Nuclear Information System (INIS)
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)
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 6Li and 7Li 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 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.
Hydrogen transport studies using neutron radiography
International Nuclear Information System (INIS)
Neutron cross-sections and their angular and energy-dependence as characteristics of neutron interaction with hydrogen isotopes and compounds are presented. It is shown how deuteration and different molecular modifications (e.g. ortho and parahydrogen) affect the cross-sections and hence the beam attenuation. A comparison of neutron radiographic methods with other neutron techniques used for hydrogen detection is made and the necessary formalism to describe diffusion processes is given. The results obtained by neutron radiography on the measurement of hydrogen motion in various substances are reviewed, in particular diffusion measurements made on liquids (water, liquid hydrogen and methanol) and of hydrogen in metals (β-titanium, vanadium, niobium and tantalum). Finally, neutron-radiographic measurements of water transport in concrete and of carburetor icing are discussed. The advantages of the high detection efficiency of hydrogen by neutron radiography and the integral sample scan technique are simultaneously used for such measurements. Some typical results of this detection method in the field of physical and applied research are shown. (author)
Applications of the advanced neutron source reactor
International Nuclear Information System (INIS)
When the technique of neutron scattering was pioneered at the X-10 graphite reactor at Oak Ridge National Laboratory about 50 years ago, it was used to study certain important, but fairly esoteric, properties of crystals. From this modest beginning, neutron scattering has become a major tool in every branch of science, from the astrophysics of the early universe to human biology, and in many important industrial and engineering applications. In a typical modern research reactor it is not unusual to find one instrument studying new polymeric materials, while its neighbor is measuring residual stress in a jet turbine, sometimes with the jet operating. Most of this development has taken place outside of the United States, primarily in Western Europe, Japan and Russia, and it is generally recognized that we are a decade behind our competitors in this important field. The Advanced Neutron Source (ANS), planned to become operational as a user-facility at Oak Ridge at the end of this decade, will regain our leadership in neutron-based research and will be a major center for attracting new students into science. This paper discusses some of the research and development applications of the ANS, with an emphasis on applied materials science and engineering
Advanced Neutron Source: Plant Design Requirements
Energy Technology Data Exchange (ETDEWEB)
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.
Advanced Neutron Sources: Plant Design Requirements
Energy Technology Data Exchange (ETDEWEB)
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{sub 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)
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
Advanced Neutron Sources: Plant Design Requirements
International Nuclear Information System (INIS)
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-MWth, 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
Solving the equation of neutron transport
International Nuclear Information System (INIS)
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.
Advancement of German Neutron Spectrometers Relocation Project in 2008
Institute of Scientific and Technical Information of China (English)
2008-01-01
<正>Neutron scattering technique is going on in Neutron Scattering Laboratory (NSL) of China Institute of Atomic Energy (CIAE) based on China Advanced Research Reactor (CARR), which will be hopefully
Advanced neutron source materials surveillance program
International Nuclear Information System (INIS)
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
An Improved Neutron Transport Algorithm for HZETRN
Slaba, Tony C.; Blattnig, Steve R.; Clowdsley, Martha S.; Walker, Steven A.; Badavi, Francis F.
2010-01-01
Long term human presence in space requires the inclusion of radiation constraints in mission planning and the design of shielding materials, structures, and vehicles. In this paper, the numerical error associated with energy discretization in HZETRN is addressed. An inadequate numerical integration scheme in the transport algorithm is shown to produce large errors in the low energy portion of the neutron and light ion fluence spectra. It is further shown that the errors result from the narrow energy domain of the neutron elastic cross section spectral distributions, and that an extremely fine energy grid is required to resolve the problem under the current formulation. Two numerical methods are developed to provide adequate resolution in the energy domain and more accurately resolve the neutron elastic interactions. Convergence testing is completed by running the code for various environments and shielding materials with various energy grids to ensure stability of the newly implemented method.
Multi-group neutron transport theory
International Nuclear Information System (INIS)
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)
Neutron transports in diffusing and thermalising media
International Nuclear Information System (INIS)
Neutron transports in different diffusing and thermalising media were studied within one dimensional theory. Macroscopic cross section libraries for each medium or region were generated by one dimensional models that represent the geometry of the surrounding regions. Few group total and angular fluxes are computed. Especially, determination of angular fluxes at some points and directions are focused on. The results are compared with other computed and experimental values
AGENT code - neutron transport benchmark examples
International Nuclear Information System (INIS)
The paper focuses on description of representative benchmark problems to demonstrate the versatility and accuracy of the AGENT (Arbitrary Geometry Neutron Transport) code. AGENT couples the method of characteristics and R-functions allowing true modeling of complex geometries. AGENT is optimized for robustness, accuracy, and computational efficiency for 2-D assembly configurations. The robustness of R-function based geometry generator is achieved through the hierarchical union of the simple primitives into more complex shapes. The accuracy is comparable to Monte Carlo codes and is obtained by following neutron propagation through true geometries. The computational efficiency is maintained through a set of acceleration techniques introduced in all important calculation levels. The selected assembly benchmark problems discussed in this paper are: the complex hexagonal modular high-temperature gas-cooled reactor, the Purdue University reactor and the well known C5G7 benchmark model. (author)
A status report on the advanced neutron source project
International Nuclear Information System (INIS)
Design work on the Advanced Neutron Source facilities has progressed significantly, with cost saving changes to the buildings and other systems. The cold source design has advanced considerably, and in addition design work has been initiated on the hot neutron source and on a positron source. (J.P.N.)
GEANT 4 simulation of neutron transport and scattering in media
International Nuclear Information System (INIS)
GEANT 4 simulation toolkit and PhysList QGSP BIC HP for simulate neutron transport and scattering was used. Primary neutron spectrum was modeled similar spectrum of 239Pu - Be(alpha, n) neutron source. Spectra of neutron passing through the material and scattered were obtained. Number of thermal neutrons after passing various materials were calculated. Detector-dosimeter MKS-01R was used for measurements of the experimental thermal neutron flux from 239Pu - Be(alpha, n) neutron source. Satisfactory agreement between calculations and experiment was obtained.
Parallel Deterministic Neutron Transport with AMR
Energy Technology Data Exchange (ETDEWEB)
Clouse, C
2005-03-25
AMTRAN, a one, two and three dimensional Sn neutron transport code with adaptive mesh refinement (AMR) has been parallelized with MPI over spatial domains and energy groups and with threads over angles. Block refined AMR is used with linear finite element representations for the fluxes, which are node centered. AMR requirements are determined by minimum mean free path calculations throughout the problem and can provide an order of magnitude or more reduction in zoning requirements for the same level of accuracy, compared to a uniformly zoned problem.
Toward whole-core neutron transport without spatial homogenization
International Nuclear Information System (INIS)
, 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)
Coupling of neutron transport equations. First results
International Nuclear Information System (INIS)
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
Vector processing of the neutron transport codes
International Nuclear Information System (INIS)
One of the large computations in JAERI is the neutron transport ones used for reactor shielding and criticality analyses. The adaptability of vector processings has been investigated on the neutron transport codes under the assumption of future use of super-computer. Five codes have been tested. They are DOT3.5, TWOTRAN and ANISN based on finite difference method, and PALLAS-2DCY and BERMUDA on the direct integration method. It has been found that the gain from vectorization depends upon the numerical methods, geometries, and problems types to be solved. That is, the direct integration is rather suited for vector processing. But in the conventional finite difference method, the difference equation has an unvectorizable recurrence form in (r, z) and (r, -)-geometries. But by altering the interative process, the equation can be vectorized and some gains have been found to be achieved in a criticality problem. For each code, described are some views on vectorization, program restructurings, speedup ratio on F75 APU, numerical studies on the interative process, and so forth. (author)
Asymptotic Behaviour of Neutron Transport Processes
International Nuclear Information System (INIS)
reactor corresponds to strong mixing in the sense of ergodic theory; we define a reactor as critical if for all f and all g, positive almost everywhere, a positive limit (Ttf, g) exists for t --> ∞. This definition corresponds to the Fermi experiment. Boundedness of Tt can be demonstrated. Finally an attempt is made to define the mean entropy of a neutron transport process. (author)
Advanced spallation neutron sources for condensed matter research
International Nuclear Information System (INIS)
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)
Neutron transport on the connection machine
International Nuclear Information System (INIS)
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
ANL--LASL workshop on advanced neutron detection systems
International Nuclear Information System (INIS)
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
Advances in neutron radiography - applications and systems
International Nuclear Information System (INIS)
The performance of the neutron radiography as a technique of nondestructive materials testing was determined comprehensively in the aerospace sector, electrical engineering/electronics, mechanical engineering, constructional engineering and material engineering. Potential applications showed up in particular during the maintenance and inspection of aerospace equipment, the testing of airplane turbine blades and of pyrotechnical elements as well as in the case of the manufacturing control of fiber reinforced composites and ceramics. In order to enable an industrial utilization of the neutron radiography a high-performance, flexible and mobile neutron radiography system is being developed by the IABG in a EUREKA project with the partners SODERN (France), SENER (Spain) and LTV (USA) on the basis of a neutron generator newly designed by SODERN. The first prototype of this neutron generator, built-in into the IABG neutron radiography system procured within the framework of the project, is at present being tested. (orig.)
Advanced Neutron Source radiological design criteria
International Nuclear Information System (INIS)
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
Advanced Neutron Source radiological design criteria
Energy Technology Data Exchange (ETDEWEB)
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.
Recent advances in X-ray and neutron interferometry
International Nuclear Information System (INIS)
Since their advent interferometry with X-rays and neutrons have been developed steadily. A number of excellent reviews is covering the development up to about five years ago. Advances since then are treated in this review. Topics included are: Understanding of angstrom wave interferometers, theory of operation, types, contrast, complementarity, strategies and refinement of measurement, nonlinear Fizeau effect with neutrons, action of gravity and inertia of neutron phase, interferometers with separated crystals, interferometer combining X-ray and optical operation, interferometer combining X-ray and neutron operation. (orig.)
Advanced neutron source three-element-core fuel grading
International Nuclear Information System (INIS)
The proposed advanced neutron source (ANS) neutron research facility's purpose is to provide unprecedented experimental capabilities in the areas of neutron scattering, materials research, and isotope production. The primary goals of the ANS project are to obtain neutron flux levels that are 5 to 10 times larger than any current existing facility and to provide isotope irradiation facilities that are at least as good as the High-Flux Isotope Reactor at Oak Ridge National Laboratory. The design changes in the ANS are described
Enhancing the detector for advanced neutron capture experiments
International Nuclear Information System (INIS)
The Detector for Advanced Neutron Capture Experiments (DANCE) has been used for extensive studies of neutron capture, gamma decay, photon strength functions, and prompt and delayed fission-gamma emission. Despite these successes, the potential measurements have been limited by the data acquisition hardware. We report on a major upgrade of the DANCE data acquisition that simultaneously enables strait-forward coupling to auxiliary detectors, including high-resolution high-purity germanium detectors and neutron tagging array. The upgrade will enhance the time domain accessible for time-of-flight neutron measurements as well as improve the resolution in the DANCE barium fluoride crystals for photons
Generic programming for deterministic neutron transport codes
International Nuclear Information System (INIS)
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)
Trial fabrication of beryllides as advanced neutron multiplier
International Nuclear Information System (INIS)
Beryllium metal is considered as the neutron multiplier in the pebble bed blanket. On the other hand, advanced neutron multipliers with lower swelling and higher stability at high temperature are desired in pebble bed blankets for DEMO. Beryllium intermetallic compounds (beryllides) are the most promising advanced neutron multipliers. However, beryllides are too brittle to allow production of pebbles. Establishing fabrication techniques for beryllides is a key issue of advanced neutron multiplier development. In the previous study, it was clear that the intermetallic compound beryllides of Be-Ti can be directly synthesized by the plasma sintering method. In this study, it reports on the trial fabrication results of beryllides synthetic such as Be-V and Be-Nb using plasma sintering method for applicability evaluation of beryllide synthesis. The formation of Be-V and Be-Nb intermetallics was identified using a mixture of Be and V or Be and Nb particles for the plasma sintering method.
Advanced digital detectors for neutron imaging.
Energy Technology Data Exchange (ETDEWEB)
Doty, F. Patrick
2003-12-01
Neutron interrogation provides unique information valuable for Nonproliferation & Materials Control and other important applications including medicine, airport security, protein crystallography, and corrosion detection. Neutrons probe deep inside massive objects to detect small defects and chemical composition, even through high atomic number materials such as lead. However, current detectors are bulky gas-filled tubes or scintillator/PM tubes, which severely limit many applications. Therefore this project was undertaken to develop new semiconductor radiation detection materials to develop the first direct digital imaging detectors for neutrons. The approach relied on new discovery and characterization of new solid-state sensor materials which convert neutrons directly to electronic signals via reactions BlO(n,a)Li7 and Li6(n,a)T.
Advanced neutron diagnostics for ITER fusion experiments
International Nuclear Information System (INIS)
The diagnostics functions of neutron measurements are reviewed as well as the roles played by neutron yield monitors, cameras and spectrometers. The importance of recent developments in neutron emission spectroscopy (NES) diagnostics is emphasized. Results are presented from 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 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
Energy Technology Data Exchange (ETDEWEB)
D. W. Nigg; J. K. Hartwell; J. R. Venhuizen; C. A. Wemple; R. Risler; G. E. Laramore; W. Sauerwein; G. Hudepohl; A. Lennox
2006-06-01
The Idaho National Laboratory (INL), the University of Washington (UW) Neutron Therapy Center, the University of Essen (Germany) Neutron Therapy Clinic, and the Northern Illinois University(NIU) Institute for Neutron Therapy at Fermilab have been collaborating in the development of fast-neutron therapy (FNT) with concurrent neutron capture (NCT) augmentation [1,2]. As part of this effort, we have conducted measurements to produce suitable benchmark data as an aid in validation of advanced three-dimensional treatment planning methodologies required for successful administration of FNT/NCT. Free-beam spectral measurements as well as phantom measurements with Lucite{trademark} cylinders using thermal, resonance, and threshold activation foil techniques have now been completed at all three clinical accelerator facilities. The same protocol was used for all measurements to facilitate intercomparison of data. The results will be useful for further detailed characterization of the neutron beams of interest as well as for validation of various charged particle and neutron transport codes and methodologies for FNT/NCT computational dosimetry, such as MCNP [3], LAHET [4], and MINERVA [5].
Concise four-vector scheme for neutron transport calculations
International Nuclear Information System (INIS)
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
Design of a transportable high efficiency fast neutron spectrometer
Roecker, C.; Bernstein, A.; Bowden, N. S.; Cabrera-Palmer, B.; Dazeley, S.; Gerling, M.; Marleau, P.; Sweany, M. D.; Vetter, K.
2016-08-01
A transportable fast neutron detection system has been designed and constructed for measuring neutron energy spectra and flux ranging from tens to hundreds of MeV. The transportability of the spectrometer reduces the detector-related systematic bias between different neutron spectra and flux measurements, which allows for the comparison of measurements above or below ground. The spectrometer will measure neutron fluxes that are of prohibitively low intensity compared to the site-specific background rates targeted by other transportable fast neutron detection systems. To measure low intensity high-energy neutron fluxes, a conventional capture-gating technique is used for measuring neutron energies above 20 MeV and a novel multiplicity technique is used for measuring neutron energies above 100 MeV. The spectrometer is composed of two Gd containing plastic scintillator detectors arranged around a lead spallation target. To calibrate and characterize the position dependent response of the spectrometer, a Monte Carlo model was developed and used in conjunction with experimental data from gamma ray sources. Multiplicity event identification algorithms were developed and used with a Cf-252 neutron multiplicity source to validate the Monte Carlo model Gd concentration and secondary neutron capture efficiency. The validated Monte Carlo model was used to predict an effective area for the multiplicity and capture gating analyses. For incident neutron energies between 100 MeV and 1000 MeV with an isotropic angular distribution, the multiplicity analysis predicted an effective area of 500 cm2 rising to 5000 cm2. For neutron energies above 20 MeV, the capture-gating analysis predicted an effective area between 1800 cm2 and 2500 cm2. The multiplicity mode was found to be sensitive to the incident neutron angular distribution.
Advances in imaging with thermal neutrons
International Nuclear Information System (INIS)
Experiments have been conducted using a modern high-resolution 3He two-dimensional position-sensitive detection chamber combined with coded apertures to produce images by means of thermal neutrons. These images are comparable to those produced by gamma ray imaging, but with some important differences. The detector is much less sensitive to the fast neutrons than to the thermalized component. Therefore, assuming that the neutron source has a fission spectrum, the brightest regions in an image represent moderating material in close proximity to the source, rather than the source itself. Earlier experiments have shown that useful contrast can be produced with thermal neutrons using thin masks made of metallic Cd sheet, but the resolution in those experiments was detector-limited at a few centimeters per pixel. The newer detector can resolve a line image with a fwhm resolution of about 1 mm. The technique could in principle be used in re-entry vehicle on-site inspections to count multiple nuclear warheads. Thermal neutrons carry no detailed spectral information, so their detection should not be as intrusive as gamma ray imaging. This technique can be used in nuclear materials management and arms control
A transport optics for pulsed ultracold neutron sources
International Nuclear Information System (INIS)
High-density ultracold neutron (UCN) is commonly desired for the improvement of the experimental sensitivity to measure the electric dipole moment of neutrons. We discuss a method to suppress the decrease of the UCN density in transporting UCNs to the spatially separated storage volume by changing the UCN velocity synchronizing to the UCN time-of-flight.
A study of a transportable thermal neutron radiography unit based on a compact RFI linac
International Nuclear Information System (INIS)
A transportable thermal neutron radiography system, incorporating a compact proton accelerator as neutron source has been simulated using the MCNP4B code. The neutron source will be produced via the 7Li(p,n)7Be reactions by a 2.5 MeV, 10 mA proton beam into a thick lithium target. Variable values for the collimator ratio were calculated. Thermal neutron radiography parameters are comparable to the research nuclear reactors. Sapphire filter was treated in order to improve the results. Simple and advanced neutron shielding materials considered which was further enhanced with layers of bismuth. The system was compatible with the European Union Directive on 'Restriction of Hazardous Substances' (RoHS) 2002/95/EC, hence excluding the use of cadmium and lead. (author)
UPWIND DISCONTINUOUS GALERKIN METHODS FOR TWO DIMENSIONAL NEUTRON TRANSPORT EQUATIONS
Institute of Scientific and Technical Information of China (English)
袁光伟; 沈智军; 闫伟
2003-01-01
In this paper the upwind discontinuous Galerkin methods with triangle meshes for two dimensional neutron transport equations will be studied.The stability for both of the semi-discrete and full-discrete method will be proved.
Advancement of neutron radiography technique in JRR-3M
International Nuclear Information System (INIS)
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)
Advanced neutron instrumentation at FRM-II
International Nuclear Information System (INIS)
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)
Advanced neutron diagnostics for ITER fusion experiments
International Nuclear Information System (INIS)
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)
Advances in neutron radiography at UJV
International Nuclear Information System (INIS)
A brief description is given of the development of neutron radiography and of planned development of neutron sources, imaging methods, evaluation methods and instrumentation. Experimental equipment and the application fields are described. The method is used in the metrology of fuel elements, for the study of the penetration of aggressive substances into building materials, for the diagnosis of bone tumors between surgeries, in archaeology, in crack detection of glued joints of honeycombed structures and in imaging the crystalline structure of castings of nickel-based superalloys. (J.P.)
Recent advances in neutron capture therapy (NCT)
Energy Technology Data Exchange (ETDEWEB)
Fairchild, R.G.
1985-01-01
The application of the /sup 10/B(n,..cap alpha..)/sup 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.
The neutron texture diffractometer at the China Advanced Research Reactor
Mei-Juan, Li; Xiao-Long, Liu; Yun-Tao, Liu; Geng-Fang, Tian; Jian-Bo, Gao; Zhou-Xiang, Yu; Yu-Qing, Li; Li-Qi, Wu; Lin-Feng, Yang; Kai, Sun; Hong-Li, Wang; R. Santisteban, J.; Dong-Feng, Chen
2016-03-01
The first neutron texture diffractometer in China has been built at the China Advanced Research Reactor, due to strong demand for texture measurement with neutrons from the domestic user community. This neutron texture diffractometer has high neutron intensity, moderate resolution and is mainly applied to study texture in commonly used industrial materials and engineering components. In this paper, the design and characteristics of this instrument are described. The results for calibration with neutrons and quantitative texture analysis of zirconium alloy plate are presented. The comparison of texture measurements with the results obtained in HIPPO at LANSCE and Kowari at ANSTO illustrates the reliability of the texture diffractometer. Supported by National Nature Science Foundation of China (11105231, 11205248, 51327902) and International Atomic Energy Agency-TC program (CPR0012)
Deterministic adjoint transport applications for He-3 neutron detector design
International Nuclear Information System (INIS)
This work focuses on the determination of predicted neutron detector response accomplished using neutron importance derived from an adjoint discrete ordinates (SN) transport calculation. A hypothetical detector apparatus, intended to detect fast neutrons, was modeled using He-3 tubes with graphite moderation using the PENTRANTM 3-D multi-group discrete ordinates parallel transport code system. The detector geometry was modeled using z-axis symmetry and discretized into 30,280 3-D Cartesian cells. The material spatial mesh was generated using the PENMSHTM code in the PENTRAN system. The 47-group BUGLE-96 neutron cross section library was used for construction of macroscopic neutron cross sections. Results from an S8 angular quadrature using P3 anisotropy are presented. An adjoint transport source was established in the model using group dependent He-3 response cross sections. Each He-3 tube contained an adjoint source aliased to group He-3 absorption cross sections to permit assessment of detector performance. The spectrally dependent detector response from neutron capture in He-3 tubes from an arbitrary source can, therefore, be readily determined. This response comes from the complete integral of the actual source strength weighted by the adjoint function at the source location for any source distribution scenario. For selected neutron energies, an equivalent forward MCNP Monte Carlo model was used to demonstrate good agreement with the detector response determined from the adjoint calculation. The graphite used in this design has a large impact on detector performance due to the increasing sensitivity inherent in He-3 gas as neutrons thermalize. Computational adjoint results presented here predict a fast neutron detector design that yields efficiencies between 30 and 50% for neutron energies below 3 keV, and up to 30% efficiencies for neutron energies between 3 keV and 1 MeV. Overall, the methodology applied here highlights the elegant nature of an adjoint
Enhancing the Detector for Advanced Neutron Capture Experiments
Couture A.; Mosby S.; Baramsai B.; Bredeweg T. A.; Jandel M.; Macon K.; O’Donnell J.M.; Rusev G.; Taddeucci T. N; Ullmann J.L.; Walker C.L.
2015-01-01
The Detector for Advanced Neutron Capture Experiments (DANCE) has been used for extensive studies of neutron capture, gamma decay, photon strength functions, and prompt and delayed fission-gamma emission. Despite these successes, the potential measurements have been limited by the data acquisition hardware. We report on a major upgrade of the DANCE data acquisition that simultaneously enables strait-forward coupling to auxiliary detectors, including high-resolution high-purity germanium detec...
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.; Norwood, Joseph K.; Merril, Garrick W.; Watts, John W.; Sabra, Mohammad S.; Smith, Dennis A.; Rodriquez-Otero, Miguel A.
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
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.
Scattered Neutron Tomography Based on A Neutron Transport Inverse Problem
Energy Technology Data Exchange (ETDEWEB)
William Charlton
2007-07-01
Neutron radiography and computed tomography are commonly used techniques to non-destructively examine materials. Tomography refers to the cross-sectional imaging of an object from either transmission or reflection data collected by illuminating the object from many different directions.
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.
Advanced neutron diagnostics for the Nova laser facility
International Nuclear Information System (INIS)
The authors report on recent work addressing advanced neutron diagnostics to be implemented on the Nova laser facility. The goals of these instruments are to measure the following properties of laser fusion targets: compressed fuel areal-density (Rho-R), time-duration, and spatial extent of the neutron emission. The authors will discuss the use of a noval time-of-flight system, radiochemical techniques, and the use of plastic track detectors to measure the compressed Rho-R. The authors will present the design of two proposed instruments to measure the burn time; one uses a sandwich of thin layers of plastic scintillator and uranium coupled to a streak camera while the other design makes use of a neutron sensitive transmission line. Finally, the authors will discuss methods capable of obtaining neutron images of the compressed pellet core
Advances in associated-particle sealed-tube neutron probe diagnostics for substance detection
International Nuclear Information System (INIS)
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
Advanced Neutronics Tools for BWR Design Calculations
International Nuclear Information System (INIS)
This paper summarizes the developments implemented in the new APOLLO2.8 neutronics tool to meet the required target accuracy in LWR applications, particularly void effects and pin-by-pin power map in BWRs. The Method Of Characteristics was developed to allow efficient LWR assembly calculations in 2D-exact heterogeneous geometry; resonant reaction calculation was improved by the optimized SHEM-281 group mesh, which avoids resonance self-shielding approximation below 23 eV, and the new space-dependent method for resonant mixture that accounts for resonance overlapping. Furthermore, a new library CEA2005, processed from JEFF3.1 evaluations involving feedback from Critical Experiments and LWR P.I.E, is used. The specific '2005-2007 BWR Plan' settled to demonstrate the validation/qualification of this neutronics tool is described. Some results from the validation process are presented: the comparison of APOLLO2.8 results to reference Monte Carlo TRIPOLI4 results on specific BWR benchmarks emphasizes the ability of the deterministic tool to calculate BWR assembly multiplication factor within 200 pcm accuracy for void fraction varying from 0 to 100%. The qualification process against the BASALA mock-up experiment stresses APOLLO2.8/CEA2005 performances: pin-by-pin power is always predicted within 2% accuracy, reactivity worth of B4C or Hf cruciform control blade, as well as Gd pins, is predicted within 1.2% accuracy. (authors)
Advanced neutronics tools for BWR design calculations
International Nuclear Information System (INIS)
This paper summarizes the developments implemented in the new APOLLO2.8 neutronics tool to meet the required target accuracy in LWR applications, particularly void effects and pin-by-pin power map in BWRs. The Method of Characteristics was developed to allow efficient LWR assembly calculations in 2D-exact heterogeneous geometry; resonant reaction calculation was improved by the optimized SHEM-281 group mesh, which avoids resonance self-shielding approximation below 23 eV, and the new space-dependent method for resonant mixture that accounts for resonance overlapping. Furthermore, a new library CEA2005, processed from JEFF3.1 evaluations involving feedback from Critical Experiments and LWR P.I.E, is used. The specific '2005-2007 BWR Plan' settled to demonstrate the validation/qualification of this neutronics tool is described. Some results from the validation process are presented: the comparison of APOLLO2.8 results to reference Monte Carlo TRIPOLI4 results on specific BWR benchmarks emphasizes the ability of the deterministic tool to calculate BWR assembly multiplication factor within 200 pcm accuracy for void fraction varying from 0 to 100%. The qualification process against the BASALA mock-up experiment stresses APOLLO2.8/CEA2005 performances: pin-by-pin power is always predicted within 2% accuracy, reactivity worth of B4C or Hf cruciform control blade, as well as Gd pins, is predicted within 1.2% accuracy
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.
Neutronic challenges of advanced boiling water reactor designs
International Nuclear Information System (INIS)
The advancement of Boiling Water Reactor technology has been under investigation at the Center for Advance Nuclear Energy Systems at MIT. The advanced concepts under study provide economic incentives through enabling further power uprates (i.e. increasing vessel power density) or better fuel cycle uranium utilization. The challenges in modeling of three advanced concepts with focus on neutronics are presented. First, the Helical Cruciform Fuel rod has been used in some Russian reactors, and studied at MIT for uprating the power in LWRs through increased heat transfer area per unit core volume. The HCF design requires high fidelity 3D tools to assess its reactor physics behavior as well as thermal and fuel performance. Second, an advanced core design, the BWR-HD, was found to promise 65% higher power density over existing BWRs, while using current licensing tools and existing technology. Its larger assembly size requires stronger coupling between neutronics and thermal hydraulics compared to the current practice. Third is the reduced moderation BWRs, which had been proposed in Japan to enable breeding and burning of fuel as an alternative to sodium fast reactors. Such technology suffers from stronger sensitivity of its neutronics to the void fraction than the traditional BWRs, thus requiring exact modeling of the core conditions such as bypass voiding, to correctly characterize its performance. (author)
Solution of modified neutron transport equation in plane geometry
International Nuclear Information System (INIS)
Neutron transport equation was formulated for universal anisotropic scattering function with integration over variable μ carried out segment (0,1) instead of segment (-1,1). A modified system of DPN equations was derived and solved by applying flux expansion in double Legendre polynomials over variable μ. As an example, case of neutron isotropic scattering was treated in detail and Green functions for infinitive medium were computed. The application of the eighth order analytical approximation achieved the accuracy to the unit on the sixth significant digit in the whole range of parameter c, angle cosine μ and distances x up ten optical lengths from the neutron source. 13 refs., 5 tabs
A Monte Carlo Green's function method for three-dimensional neutron transport
International Nuclear Information System (INIS)
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
Review of the Advanced Neutron Source (ANS) materials irradiation facilities
International Nuclear Information System (INIS)
The purpose of the workshop was to document as accurately as possible the present and future needs for neutron irradiation capacity and facilities as related to the design of the Advanced Neutron Source (ANS) which will be the next generation steady-state research reactor. The report provides the findings and recommendations of the working group. After introductory and background information is presented, the discussion includes the status of the ANS design, in particular in-core materials irradiation facilities design and important experimental parameters. The summary of workshop discussions describes a survey of irradiation-effects research community and opportunities for ex-core irradiation facilities. 20 refs., 2 figs., 4 tabs
Calculated characteristics of subcritical assembly with anisotropic transport of neutrons
Energy Technology Data Exchange (ETDEWEB)
Gorin, N.V.; Lipilina, E.N.; Lyutov, V.D.; Saukov, A.I. [Zababakhin Russian Federal Nuclear Center - All-Russian Scientific Researching Institute of Technical Physics (Russian Federation)
2003-07-01
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{sup 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)
TRIPOLI-3: a neutron/photon Monte Carlo transport code
International Nuclear Information System (INIS)
The present version of TRIPOLI-3 solves the transport equation for coupled neutron and gamma ray problems in three dimensional geometries by using the Monte Carlo method. This code is devoted both to shielding and criticality problems. The most important feature for particle transport equation solving is the fine treatment of the physical phenomena and sophisticated biasing technics useful for deep penetrations. The code is used either for shielding design studies or for reference and benchmark to validate cross sections. Neutronic studies are essentially cell or small core calculations and criticality problems. TRIPOLI-3 has been used as reference method, for example, for resonance self shielding qualification. (orig.)
International Nuclear Information System (INIS)
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
Energy Technology Data Exchange (ETDEWEB)
Crabtree, A.; Siman-Tov, M.
1993-05-01
The Advanced Neutron Source is an experimental facility being developed by Oak Ridge National Laboratory. As a new nuclear fission research reactor of unprecedented flux, the Advanced Neutron Source Reactor will provide the most intense steady-state beams of neutrons in the world. The high heat fluxes generated in the reactor [303 MW(t) with an average power density of 4.5 MW/L] will be accommodated by a flow of heavy water through the core at high velocities. In support of this experimental and analytical effort, a reliable, highly accurate, and uniform source of thermodynamic and transport property correlations for saturated light and heavy water were developed. In order to attain high accuracy in the correlations, the range of these correlations was limited to the proposed Advanced Neutron Source Reactor`s nominal operating conditions. The temperature and corresponding saturation pressure ranges used for light water were 20--300{degrees}C and 0.0025--8.5 MPa, respectively, while those for heavy water were 50--250{degrees}C and 0.012--3.9 MPa. Deviations between the correlation predictions and data from the various sources did not exceed 1.0%. Light water vapor density was the only exception, with an error of 1.76%. The physical property package consists of analytical correlations, SAS codes, and FORTRAN subroutines incorporating these correlations, as well as an interactive, easy-to-use program entitled QuikProp.
Energy Technology Data Exchange (ETDEWEB)
Crabtree, A.; Siman-Tov, M.
1993-05-01
The Advanced Neutron Source is an experimental facility being developed by Oak Ridge National Laboratory. As a new nuclear fission research reactor of unprecedented flux, the Advanced Neutron Source Reactor will provide the most intense steady-state beams of neutrons in the world. The high heat fluxes generated in the reactor [303 MW(t) with an average power density of 4.5 MW/L] will be accommodated by a flow of heavy water through the core at high velocities. In support of this experimental and analytical effort, a reliable, highly accurate, and uniform source of thermodynamic and transport property correlations for saturated light and heavy water were developed. In order to attain high accuracy in the correlations, the range of these correlations was limited to the proposed Advanced Neutron Source Reactor's nominal operating conditions. The temperature and corresponding saturation pressure ranges used for light water were 20--300[degrees]C and 0.0025--8.5 MPa, respectively, while those for heavy water were 50--250[degrees]C and 0.012--3.9 MPa. Deviations between the correlation predictions and data from the various sources did not exceed 1.0%. Light water vapor density was the only exception, with an error of 1.76%. The physical property package consists of analytical correlations, SAS codes, and FORTRAN subroutines incorporating these correlations, as well as an interactive, easy-to-use program entitled QuikProp.
Neutron transport study of a beam port based dynamic neutron radiography facility
Khaial, Anas M.
Neutron radiography has the ability to differentiate between gas and liquid in two-phase flow due both to the density difference and the high neutron scattering probability of hydrogen. Previous studies have used dynamic neutron radiography -- in both real-time and high-speed -- for air-water, steam-water and gas-liquid metal two-phase flow measurements. Radiography with thermal neutrons is straightforward and efficient as thermal neutrons are easier to detect with relatively higher efficiency and can be easily extracted from nuclear reactor beam ports. The quality of images obtained using neutron radiography and the imaging speed depend on the neutron beam intensity at the imaging plane. A high quality neutron beam, with thermal neutron intensity greater than 3.0x 10 6 n/cm2-s and a collimation ratio greater than 100 at the imaging plane, is required for effective dynamic neutron radiography up to 2000 frames per second. The primary objectives of this work are: (1) to optimize a neutron radiography facility for dynamic neutron radiography applications and (2) to investigate a new technique for three-dimensional neutron radiography using information obtained from neutron scattering. In this work, neutron transport analysis and experimental validation of a dynamic neutron radiography facility is studied with consideration of real-time and high-speed neutron radiography requirements. A beam port based dynamic neutron radiography facility, for a target thermal neutron flux of 1.0x107 n/cm2-s, has been analyzed, constructed and experimentally verified at the McMaster Nuclear Reactor. The neutron source strength at the beam tube entrance is evaluated experimentally by measuring the thermal and fast neutron fluxes using copper activation flux-mapping technique. The development of different facility components, such as beam tube liner, gamma ray filter, beam shutter and biological shield, is achieved analytically using neutron attenuation and divergence theories. Monte
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.
Scientific opportunities with advanced facilities for neutron scattering
Energy Technology Data Exchange (ETDEWEB)
Lander, G.H.; Emery, V.J. (eds.)
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/sup 15/ n cm/sup -2/s/sup -1/ steady state source or a 10/sup 17/ n cm/sup -2/s/sup -1/ peak flux spallation source; and (3) the findings of this Workshop are consistent with the recommendations of the Major Materials Facilities Committee.
Optimization study of a transportable neutron radiography unit based on a compact neutron generator
Energy Technology Data Exchange (ETDEWEB)
Fantidis, J.G. [Laboratory of Nuclear Technology, School of Engineering, ' Democritus' University of Thrace, Xanthi (Greece); Nicolaou, G.E., E-mail: nicolaou@ee.duth.g [Laboratory of Nuclear Technology, School of Engineering, ' Democritus' University of Thrace, Xanthi (Greece); Tsagas, N.F. [Laboratory of Nuclear Technology, School of Engineering, ' Democritus' University of Thrace, Xanthi (Greece)
2010-06-21
A transportable fast and thermal neutron radiography system, incorporating a compact DD neutron generator, has been simulated using the MCNPX code. The materials considered were compatible with the European Union Directive on 'Restriction of Hazardous Substances'(RoHS) 2002/95/EC, hence excluding the use of cadmium and lead. Appropriate collimators were simulated for each of the radiography modes. With suitable aperture and collimator designs, it was possible to optimize the parameters for both fast and thermal neutron radiographies, for a wide range of values of the collimator ratio. The system simulated allows different object sizes to be studied with a wide range of radiography parameters.
Optimization study of a transportable neutron radiography unit based on a compact neutron generator
International Nuclear Information System (INIS)
A transportable fast and thermal neutron radiography system, incorporating a compact DD neutron generator, has been simulated using the MCNPX code. The materials considered were compatible with the European Union Directive on 'Restriction of Hazardous Substances'(RoHS) 2002/95/EC, hence excluding the use of cadmium and lead. Appropriate collimators were simulated for each of the radiography modes. With suitable aperture and collimator designs, it was possible to optimize the parameters for both fast and thermal neutron radiographies, for a wide range of values of the collimator ratio. The system simulated allows different object sizes to be studied with a wide range of radiography parameters.
Neutron transport model based on the transmission probability method
International Nuclear Information System (INIS)
Highlights: • One hexagonal assembly is divided into 6 triangular prisms in order to get accurate flux distributions. • Transmission probability method is applied to solve the integral neutron transport equation. • The neutron flux and source are expanded spatially by a set of second order orthogonal polynomials. • The neutron flux at the interface is approximated with simplified P1 approximation. - Abstract: A new project has been started recently at KIT to develop a code able to treat hexagonal-z geometries with low density regions. The mathematical method chosen for that purpose is the Transmission Probability Method (TPM) for solving the integral neutron transport equation. In this model, one hexagonal prism is divided into six or more triangular prisms in order to get accurate flux distributions. Within each triangular prism, the neutron source is assumed to be isotropic, the scalar flux and source being approximated in space with a set of second order orthogonal polynomials. The neutron flux at the interfaces is constant in space and approximated with the simplified P1 approximation in angle. A new code, TPM-HEXZ, based on the described model is developed and some benchmarks are used to verify the code, the results are in good agreement with reference ones
The advanced neutron source research and development plan
Energy Technology Data Exchange (ETDEWEB)
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 {center_dot} 10{sup 19} {center_dot} m{sup -2} {center_dot} s{sup -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&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&D activity is not intended to produce new concepts to allow the project to meet minimum performance goals. Instead, the R&D program will focus on the four objectives described.
The advanced neutron source research and development plan
International Nuclear Information System (INIS)
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 · 1019 · 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
Thermal and transport properties of the neutron star inner crust
Page, Dany
2012-01-01
We review the nuclear and condensed matter physics underlying the thermal and transport properties of the neutron star inner crust. These properties play a key role in interpreting transient phenomena such as thermal relaxation in accreting neutron stars, superbursts, and magnetar flares. We emphasize simplifications that occur at low temperature where the inner crust can be described in terms of electrons and collective excitations. The heat conductivity and heat capacity of the solid and superfluid phase of matter is discussed in detail and we emphasize its role in interpreting observations of neutron stars in soft X-ray transients. We highlight recent theoretical and observational results, and identify future work needed to better understand a host of transient phenomena in neutron stars.
A new DPN formulation of neutron transport equation
International Nuclear Information System (INIS)
Neutron transport equation where integration over variable μ was carried out in segment [0,1] instead of segment [-1,1] was formulated for anisotropic scattering function. A new system of DPN equations is obtained by applying flux expansion in double Legendre polynomial over variable μ. This procedure enables an approximate analytical solution of transport equation with high accuracy, even in low order approximation. (author). 6 refs., 2 tabs
New developments in differencing the spherical geometry neutron transport equation
International Nuclear Information System (INIS)
Early differencing methods due to Carlson, Lathrop, and others have continued to be used to approximate the spherical geometry neutron transport equations. Nonphysical depressions in the scalar flux profiles continue to cause problems when these early techniques are used. Recent developments, however, provide better understanding of the behavior of these methods and have led to a simple approach to improve numerical solutions
STABILITY OF P2 METHODS FOR NEUTRON TRANSPORT EQUATIONS
Institute of Scientific and Technical Information of China (English)
袁光伟; 沈智军; 沈隆钧; 周毓麟
2002-01-01
In this paper the P2 approximation to the one-group planar neutron transport theory is discussed. The stability of the solutions for P2 equations with general boundary conditions, including the Marshak boundary condition, is proved. Moreover,the stability of the up-wind difference scheme for the P2 equation is demonstrated.
Neutron transport calculations of some fast critical assemblies
International Nuclear Information System (INIS)
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
Projection and conservation methods for neutron transport
International Nuclear Information System (INIS)
The solution of problems for large three-dimensional systems by conventional finite element methods is slow, even with the super-computer such as the CRAY. Projection and conservation methods can be used in conjunction to synthesis from a crude approximation a succession of more and more accurate approximations. The conservation method uses an extremum principle with two trial functions; but only one of these, the frame trial function, has to satisfy continuity conditions. When optimised the two trial functions ensure the satisfaction of the neutron conservation condition for each element. Having found a frame trial function the other trial function can be determined element by element. It is then transformed to provide another frame trial function. Extrapolation of these frame functions yields an improved frame trial function to initiate a fresh cycle of approximation. (author). 5 refs., 2 figs., 1 tab
The Advanced Neutron Source research and development plan
International Nuclear Information System (INIS)
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 1019 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
Optimization of a neutron detector design using adjoint transport simulation
Energy Technology Data Exchange (ETDEWEB)
Yi, C.; Manalo, K.; Huang, M.; Chin, M.; Edgar, C.; Applegate, S.; Sjoden, G. [Georgia Inst. of Technology, Gilhouse Boggs Bldg., 770 State St, Atlanta, GA 30332-0745 (United States)
2012-07-01
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)
Graphical User Interface for Simplified Neutron Transport Calculations
Energy Technology Data Exchange (ETDEWEB)
Schwarz, Randolph; Carter, Leland L
2011-07-18
A number of codes perform simple photon physics calculations. The nuclear industry is lacking in similar tools to perform simplified neutron physics shielding calculations. With the increased importance of performing neutron calculations for homeland security applications and defense nuclear nonproliferation tasks, having an efficient method for performing simple neutron transport calculations becomes increasingly important. Codes such as Monte Carlo N-particle (MCNP) can perform the transport calculations; however, the technical details in setting up, running, and interpreting the required simulations are quite complex and typically go beyond the abilities of most users who need a simple answer to a neutron transport calculation. The work documented in this report resulted in the development of the NucWiz program, which can create an MCNP input file for a set of simple geometries, source, and detector configurations. The user selects source, shield, and tally configurations from a set of pre-defined lists, and the software creates a complete MCNP input file that can be optionally run and the results viewed inside NucWiz.
A deterministic method for transient, three-dimensional neutron transport
International Nuclear Information System (INIS)
A deterministic method for solving the time-dependent, three-dimensional Boltzmann transport equation with explicit representation of delayed neutrons has been developed and evaluated. The methodology used in this study for the time variable of the neutron flux is known as the improved quasi-static (IQS) method. The position, energy, and angle-dependent neutron flux is computed deterministically by using the three-dimensional discrete ordinates code TORT. This paper briefly describes the methodology and selected results. The code developed at the University of Tennessee based on this methodology is called TDTORT. TDTORT can be used to model transients involving voided and/or strongly absorbing regions that require transport theory for accuracy. This code can also be used to model either small high-leakage systems, such as space reactors, or asymmetric control rod movements. TDTORT can model step, ramp, step followed by another step, and step followed by ramp type perturbations. It can also model columnwise rod movement. A special case of columnwise rod movement in a three-dimensional model of a boiling water reactor (BWR) with simple adiabatic feedback is also included. TDTORT is verified through several transient one-dimensional, two-dimensional, and three-dimensional benchmark problems. The results show that the transport methodology and corresponding code developed in this work have sufficient accuracy and speed for computing the dynamic behavior of complex multi-dimensional neutronic systems
Coupled neutron and photon cross sections for transport calculations
International Nuclear Information System (INIS)
A compact set of multigroup cross sections and transfer tables for use in neutron and photon transport calculations was prepared from ENDF/B-IV using the NJOY processing system. The library includes prompt and steady-state coupled sets for neutrons and photons in FIDO format, prompt and steady-state fission spectra (chi vectors) for the fissionable isotopes, and a table of useful response functions including heating and gas production. These multigroup constants should be useful for a wide variety of problems where self-shielding is not important. 15 references
Development of Library Processing System for Neutron Transport Calculation
Energy Technology Data Exchange (ETDEWEB)
Song, J. S.; Park, S. Y.; Kim, H. Y. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)] (and others)
2008-12-15
A system for library generation was developed for the lattice neutron transport program for pressurized water reactor core analysis. The system extracts multi energy group nuclear data for requested nuclides from ENDF/B whose data are based on continuous energy, generates hydrogen equivalent factor and resonance integral table as functions of temperature and background cross section for resonance nuclides, generates subgroup data for the lattice program to treat resonance exactly as possible, and generates multi-group neutron library file including nuclide depletion data for use of the lattice program.
Neutron shielding evaluation for a small fuel transport case
Coeck, M; Vanhavere, F
2002-01-01
We investigated the effectiveness of a small neutron shield configuration for the transportation of fresh MOX fuel rods in an experimental facility, this in order to reduce the dose received by the personnel. Monte Carlo simulations using the Tripoli and MCNP4B code were applied. Different configurations were studied, starting from the bare fuel rod positioned on an iron plate up to a fuel rod covered by a box-shaped shield made of different materials such as polyethylene, polyethylene with boron and polyethylene with a cadmium layer. We compared the neutron spectra for the different cases and calculated the corresponding ambient equivalent dose rate H*(10).
Construction and adjustment of neutron texture diffractometer at China advanced research reactor
International Nuclear Information System (INIS)
The neutron texture diffractometer is one of important and commonly used neutron instruments in the international neutron scattering laboratories. Under the demands of texture measurement with neutrons from domestic user community, the neutron texture diffractometer has been built at China Advanced Research Reactor (CARR). Currently, the preliminary adjustment and calibration with neutrons for this instrument has been finished. In this paper, the measurement principle and advantages of neutron texture diffractometer were briefly introduced. The key components and detailed characteristics for neutron texture diffractometer at CARR and the corresponding results of calibration and performance test were also presented. (authors)
Advanced Neutron Source: Plant Design Requirements. Revision 4
Energy Technology Data Exchange (ETDEWEB)
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.
The Advanced Neutron Source Facility: A new user facility for neutron research
International Nuclear Information System (INIS)
The Advanced Neutron Source (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 and for associated research space and equipment. The ANS will be open for use by scientists from universities, industry, and other federal laboratories. The ANS will be built around a new research reactor of unprecedented flux; that is, it will produce the most intense continuous beams of neutrons in the world. The goal is to reach a thermal neutron flux for beam experiments of 5 /times/ 1019 to 10 /times/ 1019 neutrons/(m2/center dot/s/sup /minus/1/). By combining the higher source flux with improved experimental facilities, the ANS will surpass current US high flux reactors---the High Flux Isotope Reactor (HFIR) at ORNL and the High Flux Beam Reactor (HFBR) at Brookhaven National Laboratory---by a factor of 10 to 20. The safety analysis of the ANS facility will include a complete probabilistic risk assessment (PRA), which will provide a systematic assessment of dependencies among systems at the malfunctions. For the current generation of nuclear power plants that have recently undergone the licensing review process, PRA has been used an an analysis tool after completion of the plant designs. For the ANS Project, the PRA effort has already begun, before the facility conceptual design. This allows safety insights from the PRA to be incorporated into the evolving plant design. 4 refs., 6 figs
International Nuclear Information System (INIS)
Due to a need for security screening instruments capable of detecting explosives and nuclear materials there is growing interest in neutron generator systems suitable for field use for applications broadly referred to as active neutron interrogation (ANI). Over the past two years Thermo Electron Corporation has developed a suite of different compact accelerator neutron generator products specifically designed for ANI field work to meet this demand. These systems incorporate hermetically-sealed particle accelerator tubes designed to produce fast neutrons using either the deuterium-deuterium (En = 2.5 MeV) or deuterium-tritium (En = 14.1 MeV) fusion reactions. Employing next-generation features including advanced sealed-tube accelerator designs, all-digital control electronics and innovative housing configurations these systems are suitable for many different uses. A compact system weighing less than 14 kg (MP 320) with a lifetime exceeding 1000 hours has been developed for portable applications. A system for fixed installations (P 325) has been developed with an operating life exceeding 4500 hours that incorporates specific serviceability features for permanent facilities with difficult-to-access shield blocks. For associated particle imaging (API) investigations a second-generation system (API 120) with an operating life of greater than 1000 hours has been developed for field use in which a high resolution fiberoptic imaging plate is specially configured to take advantage of a neutron point-source spot size of ∼2 mm. (author)
Advanced Neutron Source Reactor thermal analysis of fuel plate defects
International Nuclear Information System (INIS)
The Advanced Neutron Source Reactor (ANSR) is a research reactor designed to provide the highest continuous neutron beam intensity of any reactor in the world. The present technology for determining safe operations were developed for the High Flux Isotope Reactor (HFIR). These techniques are conservative and provide confidence in the safe operation of HFIR. However, the more intense requirements of ANSR necessitate the development of more accurate, but still conservative, techniques. This report details the development of a Local Analysis Technique (LAT) that provides an appropriate approach. Application of the LAT to two ANSR core designs are presented. New theories of the thermal and nuclear behavior of the U3Si2 fuel are utilized. The implications of lower fuel enrichment and of modifying the inspection procedures are also discussed. Development of the computer codes that enable the automate execution of the LAT is included
An adaptive finite element approach for neutron transport equation
International Nuclear Information System (INIS)
Highlights: → Using uniform grid solution gives high local residuals errors. → Element refinement in the region where the flux gradient is large improves accuracy of results. → It is not necessary to use high density element throughout problem domain. → The method provides great geometrical flexibility. → Implementation of different density of elements lowers computational cost. - Abstract: In this paper, we develop an adaptive element refinement strategy that progressively refines the elements in appropriate regions of domain to solve even-parity Boltzmann transport equation. A posteriori error approach has been used for checking the approximation solutions for various sizes of elements. The local balance of neutrons in elements is utilized as an error assessment. To implement the adaptive approach a new neutron transport code FEMPT, finite element modeling of particle transport, for arbitrary geometry has been developed. This code is based on even-parity spherical harmonics and finite element method. A variational formulation is implemented for the even-parity neutron transport equation for the general case of anisotropic scattering and sources. High order spherical harmonic functions expansion for angle and finite element method in space is used as trial function. This code can be used to solve the multi-group neutron transport equation in highly complex X-Y geometries with arbitrary boundary condition. Due to powerful element generator tools of FEMPT, the description of desired and complicated 2D geometry becomes quite convenient. The numerical results show that the locally adaptive element refinement approach enhances the accuracy of solution in comparison with uniform meshing approach.
Computational benchmarking of fast neutron transport throughout large water thicknesses
International Nuclear Information System (INIS)
Neutron dosimetry experiments seem to point our difficulties in the treatment of large water thickness like those encountered between the core baffle and the pressure vessel. This paper describes the theoretical benchmark undertaken by EDF, SCK/CEN and TRACTEBEL ENERGY ENGINEERING, concerning the transport of fast neutrons throughout a one meter cube of water, located after a U-235 fission sources plate. The results showed no major discrepancies between the calculations up to 50 cm from the source, accepting that a P3 development of the Legendre polynomials is necessary for the Sn calculations. The main differences occurred after 50 cm, reaching 20 % at the end of the water cube. This results lead us to consider an experimental benchmark, dedicated to the problem of fast neutron deep penetration in water, which has been launched at SCK/CEN. (authors)
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.
Energy Technology Data Exchange (ETDEWEB)
Penttila, S.I.; Fitzsimmons, M.R. [Los Alamos National Lab., NM (US); Delheij, P.J. [TRIUMF, Vancouver, British Columbia (Canada)] [and others
1998-12-01
The authors describe work on the development of polarized gaseous {sup 3}He cells, which are intended for use as neutron polarizers. Laser diode arrays polarize Rb vapor in a sample cell and the {sup 3}He is polarized via collisions. They describe development and tests of such a system at LANSCE.
Exact-to-precision generalized perturbation for neutron transport calculation
International Nuclear Information System (INIS)
This manuscript extends the exact-to-precision generalized perturbation theory (EPGPT), introduced previously, to neutron transport calculation whereby previous developments focused on neutron diffusion calculation only. The EPGPT collectively denotes new developments in generalized perturbation theory (GPT) that place premium on computational efficiency and defendable accuracy in order to render GPT a standard analysis tool in routine design and safety reactor calculations. EPGPT constructs a surrogate model with quantifiable accuracy which can replace the original neutron transport model for subsequent engineering analysis, e.g. functionalization of the homogenized few-group cross sections in terms of various core conditions, sensitivity analysis and uncertainty quantification. This is achieved by reducing the effective dimensionality of the state variable (i.e. neutron angular flux) by projection onto an active subspace. Confining the state variations to the active subspace allows one to construct a small number of what is referred to as the 'active' responses which are solely dependent on the physics model rather than on the responses of interest, the number of input parameters, or the number of points in the state phase space. (authors)
Neutron imaging of ion transport in mesoporous carbon materials.
Sharma, Ketki; Bilheux, Hassina Z; Walker, Lakeisha M H; Voisin, Sophie; Mayes, Richard T; Kiggans, Jim O; Yiacoumi, Sotira; DePaoli, David W; Dai, Sheng; Tsouris, Costas
2013-07-28
Neutron imaging is presented as a tool for quantifying the diffusion of ions inside porous materials, such as carbon electrodes used in the desalination process via capacitive deionization and in electrochemical energy-storage devices. Monolithic mesoporous carbon electrodes of ∼10 nm pore size were synthesized based on a soft-template method. The electrodes were used with an aqueous solution of gadolinium nitrate in an electrochemical flow-through cell designed for neutron imaging studies. Sequences of neutron images were obtained under various conditions of applied potential between the electrodes. The images revealed information on the direction and magnitude of ion transport within the electrodes. From the time-dependent concentration profiles inside the electrodes, the average value of the effective diffusion coefficient for gadolinium ions was estimated to be 2.09 ± 0.17 × 10(-11) m(2) s(-1) at 0 V and 1.42 ± 0.06 × 10(-10) m(2) s(-1) at 1.2 V. The values of the effective diffusion coefficient obtained from neutron imaging experiments can be used to evaluate model predictions of the ion transport rate in capacitive deionization and electrochemical energy-storage devices. PMID:23756558
KAMCCO, a reactor physics Monte Carlo neutron transport code
International Nuclear Information System (INIS)
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.)
Fast neutron transport through laminated iron-water shield
International Nuclear Information System (INIS)
Reaction rates were measured in a laminated iron-water shield by threshold detectors, from which the neutron spectra were obtained with the aid of the SAND-II code. The error analysis for the unfolding of the spectra proved that the spectra obtained satisfactorily in the energy range of 1 -- 10.5 MeV. One-dimensional calculations were made by the discrete ordinates transport codes ANISN-JR and PALLAS in a spherical geometry. Agreements within a factor of 1.6 for the spectra and 1.31 for the reaction rates were obtained between the measurements and calculations, though rather large discrepancies were found in the spectra at the energy range of 3 -- 7 MeV. All experimental data in absolute value and detailed specifications for source, detector and the experimental geometry are given for a fast neutron transport benchmark calculation. (author)
Neutron transport in WIMS by the characteristics method
International Nuclear Information System (INIS)
This report is the text of a Paper presented by the author at the American Nuclear Society meeting in San Diego, California in June 1993. It summarises the characteristics method known as CACTUS for solving the neutron transport equation, and describes its application to a benchmark problem with adjacent gadolinium pins. The new CACTUS options (a) to subdivide regions into computational meshes, and (b) to extend the method to allow for the spatial variation of source distributions are highlighted. (Author)
Deterministic methods to solve the integral transport equation in neutronic
International Nuclear Information System (INIS)
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
Neutron transport calculations using Quasi-Monte Carlo methods
Energy Technology Data Exchange (ETDEWEB)
Moskowitz, B.S.
1997-07-01
This paper examines the use of quasirandom sequences of points in place of pseudorandom points in Monte Carlo neutron transport calculations. For two simple demonstration problems, the root mean square error, computed over a set of repeated runs, is found to be significantly less when quasirandom sequences are used ({open_quotes}Quasi-Monte Carlo Method{close_quotes}) than when a standard Monte Carlo calculation is performed using only pseudorandom points.
Energy Technology Data Exchange (ETDEWEB)
Kim, Chang Hyo; Kim, Seung Cho; Kim, Taek Kyum; Cho, Jin Young; Lee, Hyun Cheol; Lee, Jung Hun; Jung, Gu Young [Seoul National University, Seoul (Korea, Republic of)
1995-08-01
The neutronics design system of the advanced PWR consists of (i) hexagonal cell and fuel assembly code for generation of homogenized few-group cross sections and (ii) global core neutronics analysis code for computations of steady-state pin-wise or assembly-wise core power distribution, core reactivity with fuel burnup, control rod worth and reactivity coefficients, transient core power, etc.. The major research target of the first year is to establish the numerical method and solution of multi-group diffusion equations for neutronics code development. Specifically, the following studies are planned; (i) Formulation of various numerical methods such as finite element method(FEM), analytical nodal method(ANM), analytic function expansion nodal(AFEN) method, polynomial expansion nodal(PEN) method that can be applicable for the hexagonal core geometry. (ii) Comparative evaluation of the numerical effectiveness of these methods based on numerical solutions to various hexagonal core neutronics benchmark problems. Results are follows: (i) Formulation of numerical solutions to multi-group diffusion equations based on numerical methods. (ii) Numerical computations by above methods for the hexagonal neutronics benchmark problems such as -VVER-1000 Problem Without Reflector -VVER-440 Problem I With Reflector -Modified IAEA PWR Problem Without Reflector -Modified IAEA PWR Problem With Reflector -ANL Large Heavy Water Reactor Problem -Small HTGR Problem -VVER-440 Problem II With Reactor (iii) Comparative evaluation on the numerical effectiveness of various numerical methods. (iv) Development of HEXFEM code, a multi-dimensional hexagonal core neutronics analysis code based on FEM. In the target year of this research, the spatial neutronics analysis code for hexagonal core geometry(called NEMSNAP-H temporarily) will be completed. Combination of NEMSNAP-H with hexagonal cell and assembly code will then equip us with hexagonal core neutronics design system. (Abstract Truncated)
Transport modeling and advanced computer techniques
International Nuclear Information System (INIS)
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
Advanced Neutron Source (ANS) Project. Progress report FY 1993
Energy Technology Data Exchange (ETDEWEB)
Campbell, J.H. [ed.; 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). Engineering Div.
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.
Accuracy preserving surrogate for neutron transport calculations
International Nuclear Information System (INIS)
Recent advances in reduced order modeling and exact-to-precision generalized perturbation theory are combined in a novel algorithm that constructs a surrogate model for the Boltzmann equation, commonly used in assembly calculations to functionalize the few-group cross-sections in terms of the various assembly types, depletion characteristics, and thermal-hydraulics conditions. First, the algorithm employs reduced order modeling to determine the dominant input parameters, aggregated in the so-called active subspace, using a random sample of first-order derivatives calculated using an adjoint model. Next, exact-to-precision generalized perturbation theory identifies an active subspace for the state solution (i.e., angular flux) and constructs a surrogate model that is parameterized over the active subspace of the input parameters. This approach is shown to significantly reduce computational time needed for the analysis of a large number of model variations, while meeting the user-defined accuracy requirements. Numerical experiments are employed to demonstrate the mechanics and application of the proposed approach to assembly calculations commonly used in reactor physics analysis. (author)
Transport of D-D fusion neutrons in thick concrete
International Nuclear Information System (INIS)
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
MAGGENTA: Multiassembly General Geometry Neutron Transport Theory Code
International Nuclear Information System (INIS)
MAGGENTA solves the multigroup steady-state neutron integral transport equation in arbitrary two-dimensional multi-assembly geometries that can be described by combinatorial geometry. Given transport corrected macroscopic cross sections, MAGGENTA solves an eigenvalue problem and calculates the volumetric flux and incoming/outgoing current distributions. MAGGENTA utilizes the p4 Parallel Programming System on a network of workstations or other supercomputers to solve large multi-assembly problems. The solver is optimized for vectro processing on vector machines. A graphical interface has been developed to simplify the assembly layout construction and processor assignments
Solution of neutron transport equation by Method of Characteristics
International Nuclear Information System (INIS)
Highlights: • A neutron transport theory code, based on Method of Characteristics (MOC), is developed. • The code is able to simulate square, circular, hexagonal geometries and their combinations. • Delaunay triangulation together with Bower–Watson algorithm is used for mesh generation. • The code is benchmarked against different geometry and boundary conditions. • Results corroborate well with the results available in literature. - Abstract: A computer code based on Method of Characteristics (MOC) is developed to solve neutron transport equation for mainly assembly level lattice calculation with reflective and periodic boundary conditions and to some extent core level calculation with vacuum boundary condition. The code is able to simulate square, circular and hexagonal geometries and their combinations. Delaunay triangulation together with the Bower–Watson algorithm is used to divide the problem geometry into triangular meshes. Ray tracing technique is developed to draw characteristics lines along different directions over the geometry and the transport equation is solved over these lines to obtain neutron flux distribution and multiplication factor for the geometry. A number of benchmark problems available in literature are analyzed to demonstrate the capability and validity of the code
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 Neutron Source (ANS) Project progress report, FY 1994
International Nuclear Information System (INIS)
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
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.
Energy Technology Data Exchange (ETDEWEB)
Slater, C.O.; Bucholz, J.A.
1995-08-01
Two-dimensional discrete ordinates radiation transport calculations were performed for a model of the three-element core Advanced Neutron Source reactor design under normal operating conditions. The core consists of two concentric upper elements and a lower element radially centered in the annulus between the upper elements. The initial radiation transport calculations were performed with the DORT two-dimensional discrete ordinates radiation transport code using the 39-neutron-group/44-gamma-ray-group ANSL-V cross-section library, an S{sub 6} quadrature, and a P{sub 1} Legendre polynomial expansion of the cross sections to determine the fission neutron source distribution in the core fuel elements. These calculations were limited to neutron groups only. The final radiation transport calculations, also performed with DORT using the 39-neutron-group/44-gamma-ray-group ANSL-V cross-section library, an S{sub l0} quadrature, and a P{sub 3} Legendre polynomial expansion of the cross sections, produced neutron and gamma-ray fluxes over the full extent of the geometry model. Responses (or activities) at various locations in the model were then obtained by folding the appropriate response functions with the fluxes at those locations. Some comparisons were made with VENTURE-calculated (diffusion theory) 20-group neutron fluxes that were summed into four broad groups. Tne results were in reasonably good agreement when the effects of photoneutrons were not included, thus verifying the physics model upon which the shielding model was based. Photoneutrons increased the fast-neutron flux levels deep within the D{sub 2}0 several orders of magnitude. Results are presented as tables of activity values for selected radial and axial traverses, plots of the radial and axial traverse data, and activity contours superimposed on the calculational geometry model.
International Nuclear Information System (INIS)
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 transport benchmark examples with web-based AGENT
International Nuclear Information System (INIS)
The AGENT (Arbitrary GEometry Neutron Transport) an open-architecture reactor modeling tool is deterministic neutron transport code for two- or three-dimensional heterogeneous neutronic design and analysis of the whole reactor cores regardless of geometry types and material configurations. The AGENT neutron transport methodology is applicable to all generations of nuclear power and research reactors. It combines three theories: (1) mathematical theory of R-functions that is used to generate real three-dimensional geometries of square or hexagonal heterogeneous geometries, (2) the x-y method of characteristics (MOC) used to solve isotropic neutron transport in non-homogenized 2D reactor slices, and (3) the one-dimensional diffusion theory or MOC theory used to couple the x-y and z neutron tracks through the transverse leakage and angular mesh-wise flux values. The R-function geometrical module allows a sequential building of the layers of geometry and automatic submeshing based on the network of geometric domain functions. The simplicity of geometry description and selection of parameters for accurate treatment of neutron propagation is achieved through the Boolean algebraic hierarchically organized simple primitives into complex domains (both being represented with corresponding domain functions). AGENT methodologies and numerical solutions are applicable in validating neutronic analysis for GenIV reactor designs while the effect of double heterogeneity in very high temperature reactors (VHTRs) is under development. The accuracy is comparable to Monte Carlo codes and is obtained by following neutron propagation through real geometrical domains that does not require homogenization or simplifications. The efficiency is maintained through set of acceleration techniques introduced at all important calculation levels. The flux solution incorporates power iteration with two different acceleration techniques: coarse mesh rebalancing (CMR) and coarse mesh finite difference
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.
Direct measurement of lithium transport in graphite electrodes using neutrons
International Nuclear Information System (INIS)
Highlights: ► Spatiotemporal measurements of lithium through the electrode thickness were quantified with high resolution neutron imaging. ► A nonuniform lithium distribution was observed early in the first intercalation cycle but relaxed as the electrode filled with lithium. ► Through-plane transport resistance in the bulk of the graphite composite electrode was measured. ► The distribution of lost capacity associated with trapped lithium was quantified and linked to regions with low intercalation rates. - Abstract: Lithium intercalation into graphite electrodes is widely studied, but few direct in situ diagnostic methods exist. Such diagnostic methods are desired to probe the influence of factors such as charge rate, electrode structure and solid electrolyte interphase layer transport resistance as they relate to lithium-ion battery performance and durability. In this work, we present a continuous measurement of through-plane lithium distributions in a composite graphite/lithium metal electrochemical cell. Capacity change in a thick graphite electrode was measured during several charge/discharge cycles with high resolution (14 μm) neutron imaging. A custom test fixture and a method for quantifying lithium are described. The measured lithium distribution within the graphite electrode is given as a function of state of charge. Bulk transport resistance is considered by comparing intercalation rates through the thickness of the electrode near the separator and current collector. The residual lithium content associated with irreversible capacity loss that results from cycling is also measured.
Treatment experience: locally advanced sarcomas with 15 MeV fast neutrons
International Nuclear Information System (INIS)
Experience with ten evaluable osseous sarcomas and ten evaluable advanced soft tissue sarcomas treated with neutrons of a mean neutron energy of 15 MeV are described. Neutron irradiation with or without conventional megavoltage radiotherapy is an effective modality in the treatment of these patients. No correlation between response rate and grade or whether fast neutrons alone or combined with megavoltage radiotherapy was noted. Those patients receiving a neutron dose of 2195 neutron plus gamma rads or greater all had a complete response
International Nuclear Information System (INIS)
Background: Working under extreme conditions, nuclear fuel rods, the key component of nuclear plants and reactors, are easy to be broken. In order to be safe in operation, lots of testing methods on the fuel rods have to be carried out from fabrication to operation. Purpose: Neutron radiography is a unique non-destructive testing technique which can be used to test samples with radioactivity. As the essential equipment, the nuclear fuel rod transport container has to shield the radioactivity of fuel rod and control its movement during testing and transporting. Methods: The shielding simulation of the transport container was performed using the MCNP4C code, which is a general purpose Monte Carlo code for calculating the time dependent multi-group energy transport equation for neutrons, photons and electrons in three dimensional geometries. Results: The material and dimension of the transport container used for neutron radiography testing fuel rods at Chinese Advanced Research Reactor (CARR) were optimally designed by MCNP, and the mechanical devices used to control fuel rods' movement were also described. Conclusion: The 2-m long fuel rod can be tested at CARR's neutron radiography facility (under construction) with this transport container. (authors)
BERMUDA-2DN: a two-dimensional neutron transport code
International Nuclear Information System (INIS)
A two-dimensional neutron transport code BERMUDA-2DN has been developed from the one-dimensional code PALLAS-TS (BERMUDA-1DN). The purpose of the present code is to analyze the fusion blanket neutronics experiments for plane or cylindrical assemblies, and to establish a basis of an accurate shielding analysis system for fusion and fission reactors. The time-independent transport equation is solved for two-dimensional, cylindrical, multi-regional geometry using the direct integration method in a multigroup model. In addition, group-angle transfer matrices are accurately obtained from the double-differential cross section data, without the Legendre polynomial expansion, but with the energy and scattering angle correlation. As to group constants, user is able to choose a 120-group or a 46-group library. For angular discrete ordinates, a set of 40 points is fixed over the hemisphere drawn by unit direction vectors. Not only latitudes but also longitudes (as the boundaries of the angular regions on the unit sphere) are taken into account for the calculation of the group-angle transfer matrices. For the fixed point source located at the origin of (r,z) coordinates, the uncollided flux is obtained at each spatial mesh point using the usual point kernel. The transport equation is solved for the first collision source from the uncollided flux plus the slowing down source from upper groups. Thus, the angular flux distribution is obtained as the sum of the solution and the uncollided flux values. At an intense D-T neutron source FNS, measurements were performed on the angular dependence of leakage spectra from Li2O slab assemblies. The present code has been tested by analyzing the measured spectra. The results have shown to represent fairly well the observed values. (author)
A transportable neutron radiography system based on a SbBe neutron source
Energy Technology Data Exchange (ETDEWEB)
Fantidis, J.G. [Laboratory of Nuclear Technology, School of Engineering, ' Democritus' University of Thrace, Xanthi (Greece); Nicolaou, G.E. [Laboratory of Nuclear Technology, School of Engineering, ' Democritus' University of Thrace, Xanthi (Greece)], E-mail: nicolaou@ee.duth.gr; Tsagas, N.F. [Laboratory of Nuclear Technology, School of Engineering, ' Democritus' University of Thrace, Xanthi (Greece)
2009-07-21
A transportable neutron radiography system, incorporating a SbBe neutron source, has been simulated using the MCNPX code. Design provisions have allowed two radiography systems to be utilised using the same SbBe neutron source. In this respect, neutron radiographies can be carried out using the photoneutrons produced when the {sup 124}Sb is surrounded by the Be target. Alternatively, {gamma}-radiography can be utilised with the photons from the {sup 124}Sb with the target removed. Appropriate collimators were simulated for each of the radiography modes. Apart from Be, the materials considered were compatible with the European Union Directive on 'Restriction of Hazardous Substances' (RoHS) 2002/95/EC, hence excluding the use of cadmium and lead. Bismuth was chosen as the material for {gamma}-radiation shielding and the proposed system allowed a maximum activity of the {sup 124}Sb up to 1.85x10{sup 13} Bq. The system simulated allows different object sizes to be studied with a wide range of radiography parameters.
A transportable neutron radiography system based on a SbBe neutron source
International Nuclear Information System (INIS)
A transportable neutron radiography system, incorporating a SbBe neutron source, has been simulated using the MCNPX code. Design provisions have allowed two radiography systems to be utilised using the same SbBe neutron source. In this respect, neutron radiographies can be carried out using the photoneutrons produced when the 124Sb is surrounded by the Be target. Alternatively, γ-radiography can be utilised with the photons from the 124Sb with the target removed. Appropriate collimators were simulated for each of the radiography modes. Apart from Be, the materials considered were compatible with the European Union Directive on 'Restriction of Hazardous Substances' (RoHS) 2002/95/EC, hence excluding the use of cadmium and lead. Bismuth was chosen as the material for γ-radiation shielding and the proposed system allowed a maximum activity of the 124Sb up to 1.85x1013 Bq. The system simulated allows different object sizes to be studied with a wide range of radiography parameters.
Current status of the PSG Monte Carlo neutron transport code
International Nuclear Information System (INIS)
PSG is a new Monte Carlo neutron transport code, developed at the Technical Research Centre of Finland (VTT). The code is mainly intended for fuel assembly-level reactor physics calculations, such as group constant generation for deterministic reactor simulator codes. This paper presents the current status of the project and the essential capabilities of the code. Although the main application of PSG is in lattice calculations, the geometry is not restricted in two dimensions. This paper presents the validation of PSG against the experimental results of the three-dimensional MOX fuelled VENUS-2 reactor dosimetry benchmark. (authors)
MINARET: Towards a time-dependent neutron transport parallel solver
International Nuclear Information System (INIS)
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)
Fabrication of beryllide pebble as advanced neutron multiplier
International Nuclear Information System (INIS)
Highlights: • A new beryllide granulation process that combined process with a plasma sintering method for electrode fabrication and a rotating electrode method (REM) for granulation was suggested. • The beryllide electrode fabrication process was investigated for mass production. • As optimized beryllide electrode indicated higher ductility and was sintered at a lower temperature for a shorter time. • It appears to be more able to not only withstand the thermal shock from arc-discharge during granulation but also produce beryllide pebbles on a large scale. • These optimization results can reduce the time for electrode fabrication by 40%, they suggest the possibility of great reductions in time and cost for mass production of beryllide pebbles. - Abstract: Fusion reactors require advanced neutron multipliers with great stability at high temperatures. Beryllium intermetallic compounds, called beryllides such as Be12Ti, are the most promising materials for use as advanced neutron multipliers. However, few studies have been conducted on the development of mass production methods for beryllide pebbles. A granulation process for beryllide needs to have both low cost and high efficiency. To fabricate beryllide pebbles, a new granulation process is established in this research by combining a plasma sintering method for beryllide synthesis and a rotating electrode method using a plasma-sintered electrode for granulation. The fabrication process of the beryllide electrode is investigated and optimized for mass production. The optimized beryllide electrode exhibits higher ductility and can be sintered at a lower temperature for a shorter time, indicating that it is more suitable not only for withstanding the thermal shock from arc-discharge during granulation but also for producing the beryllide pebbles on a large scale. Accordingly, because these optimization results can reduce the time required for electrode fabrication by 40%, they suggest the possibility of
On eigenvalue problems of the one-speed neutron transport equation for isotropic scattering
International Nuclear Information System (INIS)
Highlights: • We consider isoperimetric inequalities for the one-speed neutron transport equation. • A ball will be a minimizer domain of the first eigenvalue. • We prove the Rayleigh–Faber–Krahn inequality for the neutron transport equation. - Abstract: In this paper, we consider eigenvalue problems of the one-speed neutron transport equation with isotropic scattering in a steady state and prove the Rayleigh–Faber–Krahn type inequality for the first eigenvalue
Energy Technology Data Exchange (ETDEWEB)
Smith, L.A.; Gallmeier, F.X. [Oak Ridge Institute for Science and Energy, TN (United States); Gehin, J.C. [Oak Ridge National Lab., TN (United States)] [and others
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 {approx} 13%, while the average differences are < 8%.
Structures of the fractional spaces generated by the difference neutron transport operator
International Nuclear Information System (INIS)
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
International Nuclear Information System (INIS)
It is shown that the combination of 3D neutron transport calculations and the results from activation foil measurements at a limited number of locations in a materials testing irradiation experiment can provide information at any position in the experiment for detailed neutron dosimetry and damage analysis. 4 refs
Benchmarking of neutron production of heavy-ion transport codes
International Nuclear Information System (INIS)
Document available in abstract form only, full text of document follows: Accurate prediction of radiation fields generated by heavy ion interactions is important in medical applications, space missions, and in design and operation of rare isotope research facilities. In recent years, several well-established computer codes in widespread use for particle and radiation transport calculations have been equipped with the capability to simulate heavy ion transport and interactions. To assess and validate these capabilities, we performed simulations of a series of benchmark-quality heavy ion experiments with the computer codes FLUKA, MARS15, MCNPX, and PHITS. We focus on the comparisons of secondary neutron production. Results are encouraging; however, further improvements in models and codes and additional benchmarking are required. (authors)
Neutron Transport Models and Methods for HZETRN and Coupling to Low Energy Light Ion Transport
Blattnig, S.R.; Slaba, T.C.; Heinbockel, J.H.
2008-01-01
Exposure estimates inside space vehicles, surface habitats, and high altitude aircraft exposed to space radiation are highly influenced by secondary neutron production. The deterministic transport code HZETRN has been identified as a reliable and efficient tool for such studies, but improvements to the underlying transport models and numerical methods are still necessary. In this paper, the forward-backward (FB) and directionally coupled forward-backward (DC) neutron transport models are derived, numerical methods for the FB model are reviewed, and a computationally efficient numerical solution is presented for the DC model. Both models are compared to the Monte Carlo codes HETCHEDS and FLUKA, and the DC model is shown to agree closely with the Monte Carlo results. Finally, it is found in the development of either model that the decoupling of low energy neutrons from the light ion (A<4) transport procedure adversely affects low energy light ion fluence spectra and exposure quantities. A first order correction is presented to resolve the problem, and it is shown to be both accurate and efficient.
Approximation theory and homogenization for neutron transport processes
International Nuclear Information System (INIS)
In practical calculation of reactor systems homogenization is performed by some techniques mostly based on intuition and there is no uniquely accepted approach to this problem. In the first part of the paper an attempt is made to formulate mathematical basis of homogenization for the neutron diffusion and transport equations using recent developments in this field. The boundary value problems for both equations for non smooth H - periodic coefficients are related to appropriate variational problems stated in terms of bilinear forms. The behaviour of the solutions for H → 0 is investigated under various assumptions concerning a limit process to get the coefficients of homogenized equations. In the second part of the paper the assymptotic equivalence of the neutron diffusion to the transport equation is studied. The relation between homogenization procedures for both equations is also examined. As an example, the deriviation of the equations of homogenization in the case of hexagonal geometry typical for V.V.E.R. reactor is given. The obtained formulae for so called effective diffusion coefficient are analyzed for various types of lattices. (author)
Discontinuous finite element formulations for neutron transport in spherical geometry
International Nuclear Information System (INIS)
Highlights: • We developed linear and quadratic discontinuous finite element methods in sphere. • We found that quadratic discontinuous finite element method is the best method. • Quadratic method has the desired convergence properties. • Smallest L2 error norms are obtained in scalar fluxes if quadratic method is used. - Abstract: We have developed the linear and quadratic Galerkin discontinuous finite element methods for the solution of both time-independent and time-dependent spherical geometry neutron transport problems. Discrete ordinates method is used for the angular discretization while the implicit method is utilized for temporal discretization in time-dependent problems. In order to assess the relative performance of the newly developed linear and quadratic discontinuous finite element spatial differencing methods relative to the previously developed linear discontinuous finite element and diamond difference discretizations, a computer code is developed and numerical solutions of the neutron transport equation for some benchmark problems are obtained. These numerical applications reveal that the newly developed quadratic discontinuous finite element method produces the most accurate results while the newly developed linear discontinuous finite element method follows as the second best discontinuous finite element method
PALLAS-TS: a one-dimensional neutron transport code for analyzing fusion blanket neutronics
International Nuclear Information System (INIS)
The one-dimensional neutron transport code PALLAS-TS has been developed for solving the transport equation by direct numerical integration method. Group-transference kernels are accurately obtained from the double-differential cross section data using the energy and scattering angle correlation relation for elastic and inelastic (discrete levels) scattering. In addition, a usual multigroup model is adopted in calculation of spatial and angular flux distribution so as to make it possible to use iteration technique with neutron rebalancing in each group. This code uses a 120-group data library for 29 nuclides prepared temporarily by processing the ENDF/B-IV file, though the nuclear data file available now is incomplete for accounting fully the anisotropy of scattering. Results of test calculation for a 4-region system consisting of lithium and carbon were compared with the P5-S8 calculations by the ANISN code. The present code is the first trial of incorporating the multigroup to the direct integration method for solving the transport equation. It is observed that computing time by this code is shorter than that of the usual S sub(n) method by a factor of 2 or 3. (author)
Archambault, Brian C.; Webster, Jeffrey A.; Grimes, Thomas F.; Fischer, Kevin F.; Hagen, Alex R.; Taleyakhan, Rusi P.
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 (239Pu-Be and 252Cf). 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 3He sensor replacement was performed utilizing MCNP-PoliMi simulations, the results of which
International Nuclear Information System (INIS)
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 (239Pu–Be and 252Cf). 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 3He sensor replacement was performed utilizing MCNP-PoliMi simulations, the results of
International Nuclear Information System (INIS)
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)
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
Subroutines to Simulate Fission Neutrons for Monte Carlo Transport Codes
Lestone, J P
2014-01-01
Fortran subroutines have been written to simulate the production of fission neutrons from the spontaneous fission of 252Cf and 240Pu, and from the thermal neutron induced fission of 239Pu and 235U. The names of these four subroutines are getnv252, getnv240, getnv239, and getnv235, respectively. These subroutines reproduce measured first, second, and third moments of the neutron multiplicity distributions, measured neutron-fission correlation data for the spontaneous fission of 252Cf, and measured neutron-neutron correlation data for both the spontaneous fission of 252Cf and the thermal neutron induced fission of 235U. The codes presented here can be used to study the possible uses of neutron-neutron correlations in the area of transparency measurements and the uses of neutron-neutron correlations in coincidence neutron imaging.
Conventional and advanced containers for LPG transport
Energy Technology Data Exchange (ETDEWEB)
Hausen, J.
1982-04-08
For the purpose of storage and transport, natural gas, petroleum gas, and chemical gases, must be liquefied. They are either transported in pressure or cooling vessels or in a combined type of vessel. Membrane tanks and solid tanks have been developed for LNG transport. These tanks are made of aluminium alloys or nickel steels. The production expenditure of the present systems is high. Savings may be possible by using plastics. Investigations have already shown good results.
International Nuclear Information System (INIS)
Highlights: ► We have extended the KAERI library generation system to include gamma cross section generation capability. ► A gamma transport/diffusion calculation module has been implemented in KARMA 1.2. ► The computational results for benchmark problems show that the gamma library and gamma simulation in KARMA are reasonable. - Abstract: KAERI has developed a lattice transport calculation code KARMA (Kernel Analyzer by Ray-tracing Method for fuel Assembly) and its library generation system. Recently, the library generation system has been extended to include a gamma cross section generation capability and a gamma transport/diffusion calculation module has been implemented in KARMA 1.2. The method of characteristics for the neutron transport calculation to estimate eigenvalue has been utilized to predict gamma flux distribution and energy deposition. In addition, the coarse mesh finite difference method with diffusion approximation has also been utilized to estimate gamma flux distribution and energy depositions for each coarse mesh with homogenized pins as a computationally efficient alternative. This paper describes the procedure to generate neutron induced gamma production and gamma cross section data, and the methods to predict gamma flux distribution, gamma energy deposition and gamma smeared pin power distribution. The computational results for benchmark problems show that the gamma library and gamma simulation in KARMA are reasonable. And it is noted that gamma smeared power distributions predicted by coarse mesh diffusion calculation are very accurate compared to the results of transport calculation
International Nuclear Information System (INIS)
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
Transport calculations for a 14.8 MeV neutron beam in a water phantom
International Nuclear Information System (INIS)
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
Advanced Modeling of Prompt Fission Neutrons and Gamma Rays
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Kawano T.
2010-03-01
Full Text Available Prompt fission neutrons and gamma rays are computed using a Monte Carlo treatment of the statistical evaporation of the excited primary fission fragments. The assumption of two fragments in thermal equilibrium at the time of neutron emission is addressed by studying the neutron multiplicity as a function of fragment mass. Results for the neutron-induced fission of 235U are discussed, for incident neutron energies from 0.5 to 5.5 MeV. Recent experimental data on the fission fragment yields as a function of mass and total kinetic energy are used as input data.
Advanced Modeling of Prompt Fission Neutrons and Gamma Rays
Kawano T; Talou P.
2010-01-01
Prompt fission neutrons and gamma rays are computed using a Monte Carlo treatment of the statistical evaporation of the excited primary fission fragments. The assumption of two fragments in thermal equilibrium at the time of neutron emission is addressed by studying the neutron multiplicity as a function of fragment mass. Results for the neutron-induced fission of 235U are discussed, for incident neutron energies from 0.5 to 5.5 MeV. Recent experimental data on the fission fragment yield...
Birman-Schwinger principle and Nelkin conjecture theory of neutron transport
Stepin, S A
2001-01-01
The work is dedicated to studying the spectral properties of the operator model, appearing in the neutron transport theory. The operator L under the consideration, corresponds to the Boltzmann linearized equation, describing the neutron transport in the uniform medium with the isotopic distribution of the scattered neutrons. The effective evaluation of the number of the operator L eigenvalues, confirming and quantitatively supplementing the Nelkin hypothesis, is obtained
Conceptual design of a high-intensity positron source for the Advanced Neutron Source
International Nuclear Information System (INIS)
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 64Cu 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
Conceptual design of a high-intensity positron source for the Advanced Neutron Source
Energy Technology Data Exchange (ETDEWEB)
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 {sup 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.
Parallel computing for homogeneous diffusion and transport equations in neutronics
International Nuclear Information System (INIS)
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)
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)
International Nuclear Information System (INIS)
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)
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)
Energy Technology Data Exchange (ETDEWEB)
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.
Advanced neutron source reactor probabilistic flow blockage assessment
International Nuclear Information System (INIS)
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)
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
Reactor installation and maintenance for the Advanced Neutron Source
International Nuclear Information System (INIS)
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 Reactor zoning, shielding, and radiological optimization guide
International Nuclear Information System (INIS)
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
Assessment of the roles of the Advanced Neutron Source Operators
Energy Technology Data Exchange (ETDEWEB)
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.
Advanced neutron source corrosion test-loop facility
International Nuclear Information System (INIS)
The reference core for the advanced neutron source (ANS) will have a configuration similar to the present High-Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory: simply, an array of aluminum-alloy-clad fuel plates immersed in rapidly flowing water. The high thermal conductivity of the aluminum combined with the high heat transfer coefficient governing heat flow from the plate to the water keep the fuel inside the plates at an acceptable temperature. Unfortunately, the exposed of aluminum under these conditions leads to the formation of a thin layer of oxide (boehmite) that separates the fuel plates from the coolant water. The boehmite film has very poor thermal conductivity, and the heat flux that must cross this film can cause excessive heating of the fuel during the lifetime of the core. A test loop has been built to determine experimentally the rate of corrosion product formation on the aluminum cladding at the higher heat fluxes. Preliminary experiments have been completed that illustrate the capabilities of the loop system and the general behavior of an aluminum specimen exposed to large heat fluxes and coolant velocities. This same facility will be used for thermal-hydraulic testing; however, modifications will be necessary because of higher heat fluxes, temperatures, and pressures. Currently, the design is for light water testing; heavy water tests will be conducted in the future, which will require additional modifications
Advanced neutron source final preconceptual reference core design
International Nuclear Information System (INIS)
The preconceptual design phase of the Advanced Neutron Source (ANS) Project ended with the selection of a reference reactor core that will be used to begin conceptual design work. The new reference core consists of two involute fuel elements, of different diameters, aligned axially with a small axial gap between them. The use of different element diameters permits a separate flow of coolant to be provided for each one, thus enhancing the heat removal capability and increasing the thermal-hydraulic margins. The improved cooling allows the elements to be relatively long and thin, so self-shielding is reduced and an acceptable core life can be achieved with a relatively small loading of highly enriched uranium silicide fuel clad in aluminium. The new reference design has a fueled volume 67.4 L, each element having a heated length of 474 mm and a radial fuel thickness of 66 mm. The end-of-cycle peak thermal flux in the large heavy-water reflector tank around the core is estimated to be in the range of 0.8 to 1.0 x 1020 m-2 · s-1. 7 refs., 23 figs., 15 tabs
Advanced neutron source design: burnout heat flux correlation development
International Nuclear Information System (INIS)
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/m2, respectively. Design changes under consideration may lower these values to ∼ 9 and 17 MW/m1. 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
Neutron transport and Montecarlo method: analysis and revision
International Nuclear Information System (INIS)
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)
Tracking soil transport to sugarcane industry using neutron activation analysis
International Nuclear Information System (INIS)
Soil as mineral impurity in sugarcane loads impacts the Brazilian sugar-ethanol industry with rising production and maintenance costs as well as decreased productivity. The mechanical harvesting of sugarcane was conceived as a technology with potential to increase the raw material quality thereby has been gradually replacing manual harvesting throughout the country. Instrumental neutron activation analysis was applied for determination of soil tracers in order to compare the performance of both harvesting systems in terms of mineral impurities. There were no significant differences in the amount of soil transported to sugarcane industry despite the technological progress aggregated to mechanical harvesting. However, for both harvesting systems there were significant differences on the amount of such mineral impurity between clay and sandy soils. (author)
Approximate solution to neutron transport equation with linear anisotropic scattering
International Nuclear Information System (INIS)
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)
Massively parallel performance of neutron transport response matrix algorithms
International Nuclear Information System (INIS)
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 P3 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.)
International Nuclear Information System (INIS)
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.)
International Nuclear Information System (INIS)
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
International Nuclear Information System (INIS)
This paper describes work done at the Oak Ridge National Laboratory (ORNL) for evaluating the potential and resulting consequences of a hypothetical criticality accident during refueling of the 330-MW Advanced Neutron Source (ANS) research reactor. The development of an analytical capability is described. Modeling and problem formulation were conducted using concepts of reactor neutronic theory for determining power level escalation, coupled with ORIGEN and MELCOR code simulations for radionuclide buildup and containment transport Gaussian plume transport modeling was done for determining off-site radiological consequences. Nuances associated with modeling this blast-type scenario are described. Analysis results for ANS containment response under a variety of postulated scenarios and containment failure modes are presented. It is demonstrated that individuals at the reactor site boundary will not receive doses beyond regulatory limits for any of the containment configurations studied
Parallel processing of neutron transport in fuel assembly calculation
International Nuclear Information System (INIS)
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
Experimental software design of neutron texture diffractometer at China advanced research reactor
International Nuclear Information System (INIS)
The experimental software of the neutron texture diffractometer at China Advanced Research Reactor (CARR) was designed. Based on the principle of texture measurement by neutron diffraction and the motion control and data acquisition system of the diffractometer, the functions needed for texture measurement were proposed. Then the flow charts of these functions were described in detail and realized by Python language in Linux system. The experimental software for CARR neutron texture diffractometer has been successfully accomplished. (authors)
Chemical Kinetic Modeling of Advanced Transportation Fuels
Energy Technology Data Exchange (ETDEWEB)
PItz, W J; Westbrook, C K; Herbinet, O
2009-01-20
Development of detailed chemical kinetic models for advanced petroleum-based and nonpetroleum based fuels is a difficult challenge because of the hundreds to thousands of different components in these fuels and because some of these fuels contain components that have not been considered in the past. It is important to develop detailed chemical kinetic models for these fuels since the models can be put into engine simulation codes used for optimizing engine design for maximum efficiency and minimal pollutant emissions. For example, these chemistry-enabled engine codes can be used to optimize combustion chamber shape and fuel injection timing. They also allow insight into how the composition of advanced petroleum-based and non-petroleum based fuels affect engine performance characteristics. Additionally, chemical kinetic models can be used separately to interpret important in-cylinder experimental data and gain insight into advanced engine combustion processes such as HCCI and lean burn engines. The objectives are: (1) Develop detailed chemical kinetic reaction models for components of advanced petroleum-based and non-petroleum based fuels. These fuels models include components from vegetable-oil-derived biodiesel, oil-sand derived fuel, alcohol fuels and other advanced bio-based and alternative fuels. (2) Develop detailed chemical kinetic reaction models for mixtures of non-petroleum and petroleum-based components to represent real fuels and lead to efficient reduced combustion models needed for engine modeling codes. (3) Characterize the role of fuel composition on efficiency and pollutant emissions from practical automotive engines.
International Nuclear Information System (INIS)
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
Design and manufacture of neutron time of flight spectrometer on China Advanced Research Reactor
International Nuclear Information System (INIS)
The cold or thermal neutron energy spectra on China Advanced Research Reactor (CARR) could be directly measured by neutron time of flight spectrometer. Spectrometer structure and selected parameters of its key components were introduced. The impact of chopper slit and flux limit slit on neutron counts and pulse width was analyzed. The formulas of neutron counts and pulse width which were dependent on neutron wavelength were acquired. According to neutron energy spectrum measurement requirement for high fluence rate neutron beam, low-sensitivity detector, detector flux limit slit and multi-channel scaler for data acquisition were selected. These would ensure that the count loss rate was less than 0.5%. Electronics framework of detection system was designed and the total resolution time was 22.15-29.46 μs. (authors)
Angular dependent rebalance method for solving the neutron transport equation
International Nuclear Information System (INIS)
The behavior of neutrons in a medium is described mathematically by the Boltzmann transport equation. But the equation cannot be solved analytically even in one-dimensional geomerties. Therefore, for most realistic neutron transport problems and all production transport codes, the transport equation is numerically solved through discretization of the variables. To solve the discretized transport equation, the most widely used method is a form of Von Neumann's series solution referred to as iteration on the scattering source. It is simply called as the scattering source iteration (SI) method. However, it is well known that the scattering source iteration method converges arbitrary slowly for highly scattering dominant problems. Hence, many techniques for accelerating the scattering source iteration have been developed. Typically, the acceleration method consists of two equations. The first is the higher-order equation that is the general discretized transport equation and the second is the lower-order equation that improves the result of the higher-order equation. The most popular lower-order equation is the diffusion equation that is derived based on consistency with the higher-order equation. This type of methods are called as the diffusion synthetic acceleration method (DSA). Although this type of methods works very effectively, it is very difficult to devise diffusion acceleration equations that are both effective at reducing iteration counts and easy to solve computationally. Also, implementing the DSA method in an existing transport code usually requires a significant effort. The difficulty in solving the diffusion equation relative to that of the transport equation increases with additional spatial dimensions. This further complicates the task of devising efficient DSA methods for multidimensional problems. Also, development of new transport methods requires a complicated effort in deriving DSA equations or may be impossible to derive DSA equations. The
Neutron transport validation of variational nodal subelement methods
International Nuclear Information System (INIS)
The properties of whole-core neutron transport computations are discussed and the shortcomings of present methods resulting from spatial homogenization at the fuel-pin cell and the fuel assembly levels examined. To eliminate spatial homogenization errors the variational nodal method implemented in the VARIANT code is generalized to perform full core transport calculations without spatial homogenization of cross sections at either the fuel-pin cell or fuel assembly level. The node size is chosen to correspond to one fuel-pin cell in the radial plane. Each node is divided into triangular finite subelements, with the interior spatial flux distribution represented by continuous, piecewise linear trial functions. The step change in the cross sections at the fuel-coolant interface can thus be represented explicitly in global calculations while retaining the full spherical harmonics capability of VARIANT. The resulting method is applied to a two-dimensional seven-group representation of a LWR containing MOX fuel assemblies. The accuracy of eigenvalues and peak pin powers and the CPU times are examined for various space-angle approximations. Monte Carlo reference solutions provide a basis for assessment. (author)
Experimental validation of a coupled neutron-photon inverse radiation transport solver
International Nuclear Information System (INIS)
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.
Flow blockage analysis for the advanced neutron source reactor
International Nuclear Information System (INIS)
The Advanced Neutron Source (ANS) reactor was designed to provide a research tool with capabilities beyond those of any existing reactors. One portion of its state-of-the-art design required high-speed fluid flow through narrow channels between the fuel plates in the core. Experience with previous reactors has shown that fuel plate damage can occur when debris becomes lodged at the entrance to these channels. Such debris disrupts the fluid flow to the plate surfaces and can prevent adequate cooling of the fuel. Preliminary ANS designs addressed this issue by providing an unheated entrance length for each fuel plate so that any flow disruption would recover, thus providing adequate heat removal from the downstream, heated portions of the fuel plates. As part of the safety analysis, the adequacy of this unheated entrance length was assessed using both analytical models and experimental measurements. The Flow Blockage Test Facility (FBTF) was designed and built to conduct experiments in an environment closely matching the ANS channel geometry. The FBTF permitted careful measurements of both heat transfer and hydraulic parameters. In addition to these experimental efforts, a thin, rectangular channel was modeled using the Fluent computational fluid dynamics computer code. The numerical results were compared with the experimental data to benchmark the hydrodynamics of the model. After this comparison, the model was extended to include those elements of the safety analysis that were difficult to measure experimentally. These elements included the high wall heat flux pattern and variable fluid properties. The results were used to determine the relationship between potential blockage sizes and the unheated entrance length required
Advanced Neutron Source reactor control and plant protection systems design
International Nuclear Information System (INIS)
This paper describes the reactor control and plant protection systems' conceptual design of the Advanced Neutron Source (ANS). The Plant Instrumentation, Control, and Data Systems and the Reactor Instrumentation and Control System of the ANS are planned as an integrated digital system with a hierarchical, distributed control structure of qualified redundant subsystems and a hybrid digital/analog protection system to achieve the necessary fast response for critical parameters. Data networks transfer information between systems for control, display, and recording. Protection is accomplished by the rapid insertion of negative reactivity with control rods or other reactivity mechanisms to shut down the fission process and reduce heat generation in the fuel. The shutdown system is designed for high functional reliability by use of conservative design features and a high degree of redundance and independence to guard against single failures. Two independent reactivity control systems of different design principles are provided, and each system has multiple independent rods or subsystems to provide appropriate margin for malfunctions such as stuck rods or other single failures. Each system is capable of maintaining the reactor in a cold shutdown condition independently of the functioning of the other system. A highly reliable, redundant channel control system is used not only to achieve high availability of the reactor, but also to reduce challenges to the protection system by maintaining important plant parameters within appropriate limits. The control system has a number of contingency features to maintain acceptable, off-normal conditions in spite of limited control or plant component failures thereby further reducing protection system challenges
Advanced Transport Systems Showcased in La Rochelle
Alessandrini, Adriano; Parent, Michel; Holguin, Carlos
2011-01-01
International audience CityMobil project, a large integrated project co-funded by DG RESEARCH of the European Commission, organized in La Rochelle an advanced city car showcase in which it gave to the citizens the possibility to ride driverless vehicles. 256 users where interviewed. Responses where very positive with all indicators passing the threshold of positive acceptance; only the perception of safety was on the threshold but not above. Such positive response of the citizens to the ne...
Improved Computational Neutronics Methods And Validation Protocols For The Advanced Test Reactor
International Nuclear Information System (INIS)
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.
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...
Advanced neutron diagnostics for the Nova laser facility
International Nuclear Information System (INIS)
Implosion experiments performed on Nova are expected to produce an increased yield of thermonuclear neutrons compared with that of earlier ICF experiments. This yield will make feasible a number of neutron-based measurements heretofore not possible. Laser fusion neutron diagnostics can be divided into two categories: invasive and noninvasive. Invasive techniques require the placement of a tracer material in an interesting region of the target to be activated by the thermonuclear neutrons. Noninvasive techniques involve the energy, spatial, or temporal analysis of the neutrons emitted from the target. After examining a host of diagnostic options from both categories for Nova, the authors decided to pursue both techniques. Ideas for some diagnostic systems are described
The impact of emerging technologies on an advanced supersonic transport
Driver, C.; Maglieri, D. J.
1986-01-01
The effects of advances in propulsion systems, structure and materials, aerodynamics, and systems on the design and development of supersonic transport aircraft are analyzed. Efficient propulsion systems with variable-cycle engines provide the basis for improved propulsion systems; the propulsion efficienies of supersonic and subsonic engines are compared. Material advances consist of long-life damage-tolerant structures, advanced material development, aeroelastic tailoring, and low-cost fabrication. Improvements in the areas of aerodynamics and systems are examined. The environmental problems caused by engine emissions, airport noise, and sonic boom are studied. The characteristics of the aircraft designed to include these technical advances are described.
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.
Subroutines to Simulate Fission Neutrons for Monte Carlo Transport Codes
Lestone, J. P.
2014-01-01
Fortran subroutines have been written to simulate the production of fission neutrons from the spontaneous fission of 252Cf and 240Pu, and from the thermal neutron induced fission of 239Pu and 235U. The names of these four subroutines are getnv252, getnv240, getnv239, and getnv235, respectively. These subroutines reproduce measured first, second, and third moments of the neutron multiplicity distributions, measured neutron-fission correlation data for the spontaneous fission of 252Cf, and meas...
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
Advanced lithium battery chemistries for sustainable transportation
Monaco, Simone
2014-01-01
The specific energy of lithium-ion batteries (LIBs) is today 200 Wh/kg, a value not sufficient to power fully electric vehicles with a driving range of 400 km which requires a battery pack of 90 kWh. To deliver such energy the battery weight should be higher than 400 kg and the corresponding increase of vehicle mass would narrow the driving range to 280 km. Two main strategies are pursued to improve the energy of the rechargeable lithium batteries up to the transportation targets. The first i...
Advanced modeling of prompt fission neutrons and gamma rays
International Nuclear Information System (INIS)
Prompt fission neutrons and gamma rays are computed using a Monte Carlo treatment of the statistical evaporation of the excited primary fission fragments. The assumption of two fragments in thermal equilibrium at the time of neutron emission is addressed by studying the neutron multiplicity as a function of fragment mass. Results for the neutron-induced fission of 235U are discussed, for incident neutron energies from 0.5 to 5.5 MeV. Recent experimental data on the fission fragment yields as a function of mass and total kinetic energy are used as input data. Monte-Carlo calculations allow the exploration of physical observables beyond average quantities. A new parameter RT has been introduced: RT=Tl/Th where Tl and Th are the temperatures in the light and heavy fragments. The average neutron multiplicity computed as a function of the fragment mass agrees best with the experimental data (with En=5.5 MeV) when RT=1 which can be understood as follows: as the incident neutron energy increases, the role of shell effects diminishes and the ratio of collective energies stored in the light and heavy fragment tends toward 1
Design of the cold neutron triple-axis spectrometer at the China Advanced Research Reactor
Cheng, P.; Zhang, Hongxia; Bao, W.; Schneidewind, A.; Link, P.; Grünwald, A. T. D.; Georgii, R.; Hao, L. J.; Liu, Y. T.
2016-06-01
The design of the first cold neutron triple-axis spectrometer at the China Advanced Research Reactor is presented. Based on the Monte Carlo simulations using neutron ray-tracing program McStas, the parameters of major neutron optics in this instrument are optimized. The neutron flux at sample position is estimated to be 5.6 ×107 n/cm2/s at neutron incident energy Ei=5 meV when the reactor operates normally at the designed 60 MW power. The performances of several neutron supermirror polarizing devices are compared and their critical parameters are optimized for this spectrometer. The polarization analysis will be realized with a flexible switch from the unpolarized experimental mode.
Advanced Neutron Source (ANS) Project: Annual report, April 1987--March 1988
Energy Technology Data Exchange (ETDEWEB)
Selby, D.L.; Harrington, R.M.; Peretz, F.J.; McBee, M.R. (comp.)
1989-02-01
The Advanced Neutron Source (ANS) Project (formerly called the Center for Neutron Research) will provide the world's best facilities for the study of neutron scattering. The ANS high-power density reactor will be fueled with uranium silicide and cooled, moderated, and reflected by deuterium oxide. Peak neutron fluxes in the reflector are expected to be 5 to 10 x 10/sup 19/ neutrons/center dot/m/sup -2//center dot/s/sup -1/ with a power level between 270 and 300 MW. This report describes the status of technical work funded through the ANS Project during the period April 1987 through March 1988. Earlier work is described in Center for Neutron Research Project Status Report and other Oak Ridge National Laboratory reports. 22 refs., 57 figs., 23 tabs.
Advanced Neutron Source (ANS) Project: Annual report, April 1987--March 1988
International Nuclear Information System (INIS)
The Advanced Neutron Source (ANS) Project (formerly called the Center for Neutron Research) will provide the world's best facilities for the study of neutron scattering. The ANS high-power density reactor will be fueled with uranium silicide and cooled, moderated, and reflected by deuterium oxide. Peak neutron fluxes in the reflector are expected to be 5 to 10 x 1019 neutrons/center dot/m-2/center dot/s-1 with a power level between 270 and 300 MW. This report describes the status of technical work funded through the ANS Project during the period April 1987 through March 1988. Earlier work is described in Center for Neutron Research Project Status Report and other Oak Ridge National Laboratory reports. 22 refs., 57 figs., 23 tabs
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
Nelson, Adam
Multi-group scattering moment matrices are critical to the solution of the multi-group form of the neutron transport equation, as they are responsible for describing the change in direction and energy of neutrons. These matrices, however, are difficult to correctly calculate from the measured nuclear data with both deterministic and stochastic methods. Calculating these parameters when using deterministic methods requires a set of assumptions which do not hold true in all conditions. These quantities can be calculated accurately with stochastic methods, however doing so is computationally expensive due to the poor efficiency of tallying scattering moment matrices. This work presents an improved method of obtaining multi-group scattering moment matrices from a Monte Carlo neutron transport code. This improved method of tallying the scattering moment matrices is based on recognizing that all of the outgoing particle information is known a priori and can be taken advantage of to increase the tallying efficiency (therefore reducing the uncertainty) of the stochastically integrated tallies. In this scheme, the complete outgoing probability distribution is tallied, supplying every one of the scattering moment matrices elements with its share of data. In addition to reducing the uncertainty, this method allows for the use of a track-length estimation process potentially offering even further improvement to the tallying efficiency. Unfortunately, to produce the needed distributions, the probability functions themselves must undergo an integration over the outgoing energy and scattering angle dimensions. This integration is too costly to perform during the Monte Carlo simulation itself and therefore must be performed in advance by way of a pre-processing code. The new method increases the information obtained from tally events and therefore has a significantly higher efficiency than the currently used techniques. The improved method has been implemented in a code system
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.
Identification of materials by an advanced neutronic method
International Nuclear Information System (INIS)
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)
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.)
Cooperative learning of neutron diffusion and transport theories
International Nuclear Information System (INIS)
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 spectrum obtained with Monte Carlo and transport theory
International Nuclear Information System (INIS)
The development of the computer, resulting in increasing memory capacity and processing speed, has enabled the application of Monte Carlo method to estimate the fluxes in thousands of fine bin energy structure. Usually the MC calculation is made using continuous energy nuclear data and exact geometry. Self shielding and interference of nuclides resonances are properly considered. Therefore, the fluxes obtained by this method may be a good estimation of the neutron energy distribution (spectrum) for the problem. In an early work it was proposed to use these fluxes as weighting spectrum to generate multigroup cross section for fast reactor analysis using deterministic codes. This non-traditional use of MC calculation needs a validation to gain confidence in the results. The work presented here is the validation start step of this scheme. The spectra of the JOYO first core fuel assembly MK-I and the benchmark Godiva were calculated using the tally flux estimator of the MCNP code and compared with the reference. Also, the two problems were solved with the multigroup transport theory code XSDRN of the AMPX system using the 171 energy groups VITAMIN-C library. The spectra differences arising from the utilization of these codes, the influence of evaluated data file and the application to fast reactor calculation are discussed. (author)
Adjacent-cell preconditioners for accelerating multidimensional neutron transport methods
International Nuclear Information System (INIS)
The Adjacent-cell Preconditioner (AP) is derived for accelerating generic fixed-weight, Weighted Diamond Difference (WDD) neutron transport methods in multidimensional Cartesian geometry. The AP is determined by requiring: (a) the eigenvalue of the combined mesh sweep-AP iterations to vanish in the vicinity of the origin in Fourier space; and (b) the diagonal and off-diagonal elements of the preconditioner to satisfy a diffusion-like condition. The spectra of the resulting iterations for a wide range of problem parameters exhibit a spectral radius smaller than .25, that vanishes implying immediate convergence for very large computational cells. More importantly, unlike other unconditionally stable acceleration schemes, the AP is cell-centered and its spectral radius remains small when the cell aspect ratio approaches 0 or ∞. Testing of the AP and comparison of its rate of convergence to the standard Source Iterations (SI) for Burre's Suite of Test Problems (BSTeP) demonstrates its high efficiency in reducing the number of iterations required to achieve convergence, especially for optically thick cells where acceleration is most needed
Assessment of the impact of advanced air-transport technology
Maxwell, R. L.; Dickinson, L. V., Jr.
1981-01-01
The long term prospects for commercial supersonic transportation appear attractive enough to keep supersonic research active and reasonably healthy. On the other hand, the uncertainties surrounding an advanced supersonic transport, (AST) specifically fuel price, fuel availability and noise, are too significant to warrant an accelerated research and development program until they are better resolved. It is estimated that an AST could capture about $50 billion (1979 dollars) of the potential $150 billion in sales up to the year 2010.
Transport calculation of neutron flux distribution in reflector of PW reactor
International Nuclear Information System (INIS)
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)
MCNP, a general Monte Carlo code for neutron and photon transport: a summary
International Nuclear Information System (INIS)
The general-purpose Monte Carlo code MCNP can 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
Calculation of neutron transport in plane geometry by invariant imbedding method
International Nuclear Information System (INIS)
A practical combination of invariant imbedding and transfer matrix methods was displayed in this paper. A very simple scheme for neutron transport analysis was obtained for slab materials and some results of numerical calculations are presented. (author)
International Nuclear Information System (INIS)
The LTSN formulation is extended to the neutron transport equation in slab geometry is anisotropic scattering of second order and one group of energy. The procedure consists in the LTSN matrix decomposition. Numerical results are presented. (author)
International Nuclear Information System (INIS)
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, He4 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/cm2 of aluminum and 100 gm/cm2 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
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)
International Nuclear Information System (INIS)
Measurements of neutron pulse time-width and intensity have been carried out on grids of small moderators placed side by side and decoupled by cadmium strips; a moderator concept introduced by the authors through previous publications. Transport calculations are based on the standard reactor code DOT 3.5 with the ENDF-B IV nuclear data library. (orig.)
Advanced in the neutron feedback ICF reactor concept
International Nuclear Information System (INIS)
Results are reviewed and updated from an earlier design study of a novel nuclear-pumped flashlamp laser (NP-FL) inertial fusion energy (IFE) power reactor based on the neutron feedback concept for IFE. This concept includes nuclear pumping of the laser flashlamp, a D-T seeded D-3He target and magnetic protection of the first wall of the reactor chamber coupled with direct conversion of deflected charged particles. Advantages include an increased overall plant efficiency due to improved energy coupling via neutron feedback, increased thermal-to-electric energy conversion efficiency, and lower neutron activation and waste. These factors are reflected in a driver energy of 5 MJ and a target gain of only 50 for a 53 % efficient 1000-MWe power plant operating at 6 Hz, novel components involved. However, they require further technological development. Consequently, the NP-FL plant appears to provide a very attractive 'second-generation' IFE reactor. (authors)
Advancements in Tumor Targeting Strategies for Boron Neutron Capture Therapy.
Luderer, Micah John; de la Puente, Pilar; Azab, Abdel Kareem
2015-09-01
Boron neutron capture therapy (BNCT) is a promising cancer therapy modality that utilizes the nuclear capture reaction of epithermal neutrons by boron-10 resulting in a localized nuclear fission reaction and subsequent cell death. Since cellular destruction is limited to approximately the diameter of a single cell, primarily only cells in the neutron field with significant boron accumulation will be damaged. However, the emergence of BNCT as a prominent therapy has in large part been hindered by a paucity of tumor selective boron containing agents. While L-boronophenylalanine and sodium borocaptate are the most commonly investigated clinical agents, new agents are desperately needed due to their suboptimal tumor selectivity. This review will highlight the various strategies to improve tumor boron delivery including: nucleoside and carbohydrate analogs, unnatural amino acids, porphyrins, antibody-dendrimer conjugates, cationic polymers, cell-membrane penetrating peptides, liposomes and nanoparticles. PMID:26033767
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)
On the neutron-noise transmission studies for non-multiplying media using transport theory
International Nuclear Information System (INIS)
This paper reports the results of our investigations on the neutron-noise transmission characteristics of non-multiplying media using transport theory. The study has been carried out systematically by first considering the infinite medium case for monoenergetic neutrons and then extending it to the finite media, multigroup and anisotropic scattering cases. The results are particularly related with the problems and prospects of the neutron-noise studies by excore detectors in fast reactors and would be particularly useful in developing the technology of malfunction detection by neutron-noise methods. (author)
Thomas, Justin W.
2006-12-01
The Numerical Nuclear Reactor (NNR) is a code suite that is being developed to provide high-fidelity multi-physics capability for the analysis of light water nuclear reactors. The focus of the work here is to extend the capability of the NNR by incorporation of the neutronics module, DeCART, for Boiling Water Reactor (BWR) applications. The DeCART code has been coupled to the NNR fluid mechanics and heat transfer module STAR-CD for light water reactor applications. The coupling has been accomplished via an interface program, which is responsible for mapping the STAR-CD and DeCART meshes, managing communication, and monitoring convergence. DeCART obtains the solution of the 3-D Boltzmann transport equation by performing a series of 2-D modular ray tracing-based method of characteristics problems that are coupled within the framework of 3-D coarse-mesh finite difference. The relatively complex geometry and increased axial heterogeneity found in BWRs are beyond the modeling capability of the original version of DeCART. In this work, DeCART is extended in three primary areas. First, the geometric capability is generalized by extending the modular ray tracing scheme and permitting an unstructured mesh in the global finite difference kernel. Second, numerical instabilities, which arose as a result of the severe axial heterogeneity found in BWR cores, have been resolved. Third, an advanced nodal method has been implemented to improve the accuracy of the axial flux distribution. In this semi-analytic nodal method, the analytic solution to the transverse-integrated neutron diffusion equation is obtained, where the nonhomogeneous neutron source was first approximated by a quartic polynomial. The successful completion of these three tasks has allowed the application of the coupled DeCART/STAR-CD code to practical BWR problems.
International Nuclear Information System (INIS)
Monte Carlo method is widely used for solving neutron transport equation. Basically Monte Carlo method treats continuous angle, space and energy. It gives very accurate solution when enough many particle histories are used, but it takes too long computation time. To reduce computation time, discrete Monte Carlo method was proposed. It is called Discrete Transport Monte Carlo (DTMC) method. It uses discrete space but continuous angle in mono energy one dimension problem and uses lump, linear-discontinuous (LLD) equation to make probabilities of leakage, scattering, and absorption. LLD may cause negative angular fluxes in highly scattering problem, so two scatter variance reduction method is applied to DTMC and shows very accurate solution in various problems. In transport Monte Carlo calculation, the particle history does not end for scattering event. So it also takes much computation time in highly scattering problem. To further reduce computation time, Discrete Diffusion Monte Carlo (DDMC) method is implemented. DDMC uses diffusion equation to make probabilities and has no scattering events. So DDMC takes very short computation time comparing with DTMC and shows very well-agreed results with cell-centered diffusion results. It is known that diffusion result may not be good in boundaries. So in hybrid method of DTMC and DDMC, boundary regions are calculated by DTMC and the other regions are calculated by DDMC. In this thesis, DTMC, DDMC and hybrid methods and their results of several problems are presented. The results show that DDMC and DTMC are well agreed with deterministic diffusion and transport results, respectively. The hybrid method shows transport-like results in problems where diffusion results are poor. The computation time of hybrid method is between DDMC and DTMC, as expected
Analytical benchmarks for nuclear engineering applications. Case studies in neutron transport theory
International Nuclear Information System (INIS)
Preservation of know-how in the nuclear field is promoted through the activities of the OECD Nuclear Energy Agency Data Bank. One area of importance concerns methods for solving radiation transport problems, especially with regard to neutrons. This handbook (in the form of a case study), prepared by Barry D Ganapol, is the result of such an initiative. It is a compilation of solutions to the transport equation for which analytical representations can be found. It is designed for educational use in courses on analytical transport methods and numerical methods with application to reactor physics. In addition, it contains elements for the continuous improvement of transport methods and for computer code verification. The areas of neutron slowing down, thermalization and one-, two- and three-dimensional neutron transport theory are covered. A series of training courses, based on this compilation of solutions has recently begun. (author)
Reactor physics analyses of the advanced neutron source three-element core
International Nuclear Information System (INIS)
A reactor physics analysis was performed for the Advanced Neutron Source reactor with a three-element core configuration. The analysis was performed with a two-dimensional r-z 20-energy-group finite-difference diffusion theory model of the 17-d fuel cycle. The model included equivalent r-z geometry representations of the central control rods, the irradiation and production targets, and reflector components. Calculated quantities include fuel cycle parameters, fuel element power distributions, unperturbed neutron fluxes in the reflector and target regions, reactivity perturbations, and neutron kinetics parameters
Simulation Research on Neutron Guide System CNGC for China Advanced Research Reactor
Institute of Scientific and Technical Information of China (English)
WEI; Guo-hai; HAN; Song-bai; HE; Lin-feng; WANG; Yu; WANG; Hong-li; LIU; Yun-tao; CHEN; Dong-feng; ZHAO; Zhi-xiang
2012-01-01
<正>The out-pile section of the neutron guide CNGC at CARR (China Advanced Research Reactor) was designed by Monte Carlo simulation with VITESS. The out-pile section of CNGC will be spitted to CNGC-S and CNGC-N, the cold neutron imaging facility and small angle neutron scattering facility will be installed at the end of guides respectively. XRD patterns of Bi1-xLaxFe1-yScyO3 were shown in Fig. 1.
Detailed flux calculations for the conceptual design of the Advanced Neutron Source Reactor
International Nuclear Information System (INIS)
A detailed MCNP model of the Advanced Neutron Source Reactor has been developed. All reactor components inside the reflector tank were included, and all components were highly segmented. Neutron and photon multigroup flux spectra have been calculated for each segment in the model, and thermal-to-fast neutron flux ratios were determined for each component segment. Axial profiles of the spectra are provided for all components of the reactor. Individual segment statistical uncertainties were limited wherever possible, and the group fluxes for all important reflector components have a standard deviation below 10%
Cosmic ray heliospheric transport study with neutron monitor data
Ahluwalia, H. S.; Ygbuhay, R. C.; Modzelewska, R.; Dorman, L. I.; Alania, M. V.
2015-10-01
Determining transport coefficients for galactic cosmic ray (GCR) propagation in the turbulent interplanetary magnetic field (IMF) poses a fundamental challenge in modeling cosmic ray modulation processes. GCR scattering in the solar wind involves wave-particle interaction, the waves being Alfven waves which propagate along the ambient field (B). Empirical values at 1 AU are determined for the components of the diffusion tensor for GCR propagation in the heliosphere using neutron monitor (NM) data. At high rigidities, particle density gradients and mean free paths at 1 AU in B can only be computed from the solar diurnal anisotropy (SDA) represented by a vector A (components Ar, Aϕ, and Aθ) in a heliospherical polar coordinate system. Long-term changes in SDA components of NMs (with long track record and the median rigidity of response Rm ~ 20 GV) are used to compute yearly values of the transport coefficients for 1963-2013. We confirm the previously reported result that the product of the parallel (to B) mean free path (λ||) and radial density gradient (Gr) computed from NM data exhibits a weak Schwabe cycle (11y) but strong Hale magnetic cycle (22y) dependence. Its value is most depressed in solar activity minima for positive (p) polarity intervals (solar magnetic field in the Northern Hemisphere points outward from the Sun) when GCRs drift from the polar regions toward the helioequatorial plane and out along the heliospheric current sheet (HCS), setting up a symmetric gradient Gθs pointing away from HCS. Gr drives all SDA components and λ|| Gr contributes to the diffusive component (Ad) of the ecliptic plane anisotropy (A). GCR transport is commonly discussed in terms of an isotropic hard sphere scattering (also known as billiard-ball scattering) in the solar wind plasma. We use it with a flat HCS model and the Ahluwalia-Dorman master equations to compute the coefficients α (=λ⊥/λ∥) and ωτ (a measure of turbulence in the solar wind) and transport
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.
RADTRAN 4.0: Advanced computer code for transportation risk assessment
International Nuclear Information System (INIS)
RADTRAN 4.0 is a computer code for transportation risk assessment developed by Sandia National Laboratories for the US Department of Energy. While retaining the most useful and time-proven features of its predecessors, RADTRAN 4.0 incorporates significant advances over the earlier versions. The most useful new features are: improved route-specific analysis capability, internal radionuclide data library, improved logic for analysis of multiple-radionuclide packages such as spent fuel, separate treatment of gamma and neutron components of Transport Index (TI), and increased number of accident-severity categories. In this paper, each of these features will be described, and, where appropriate, potential applications will be discussed. 11 refs
VVER-440 Ex-Core Neutron Transport Calculations by MCNP-5 Code and Comparison with Experiment
International Nuclear Information System (INIS)
Ex-core neutron transport calculations are needed to evaluate radiation loading parameters (neutron fluence, fluence rate and spectra) on the in-vessel equipment, reactor pressure vessel (RPV) and support constructions of VVER type reactors. Due to these parameters are used for reactor equipment life-time assessment, neutron transport calculations should be carried out by precise and reliable calculation methods. In case of RPVs, especially, of first generation VVER-440s, the neutron fluence plays a key role in the prediction of RPV lifetime. Main part of VVER ex-core neutron transport calculations are performed by deterministic and Monte-Carlo methods. This paper deals with precise calculations of the Russian first generation VVER-440 by MCNP-5 code. The purpose of this work was an application of this code for expert calculations, verification of results by comparison with deterministic calculations and validation by neutron activation measured data. Deterministic discrete ordinates DORT code, widely used for RPV neutron dosimetry and many times tested by experiments, was used for comparison analyses. Ex-vessel neutron activation measurements at the VVER-440 NPP have provided space (in azimuth and height directions) and neutron energy (different activation reactions) distributions data for experimental (E) validation of calculated results. Calculational intercomparison (DORT vs. MCNP-5) and comparison with measured values (MCNP-5 and DORT vs. E) have shown agreement within 10-15% for different space points and reaction rates. The paper submits a discussion of results and makes conclusions about practice use of MCNP-5 code for ex-core neutron transport calculations in expert analysis. (authors)
International Nuclear Information System (INIS)
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
Spin diffusive modes and thermal transport in neutron star crusts
Sedrakian, Armen
2015-01-01
In this contribution we first review a method for obtaining the collective modes of pair-correlated neutron matter as found in a neutron star inner crust. We discuss two classes of modes corresponding to density and spin perturbations with energy spectra $\\omega = \\omega_0 + \\alpha q^2$, where $\\omega_0 = 2\\Delta$ is the threshold frequency and $\\Delta$ is the gap in the neutron fluid spectrum. For characteristic values of Landau parameters in neutron star crusts the exitonic density modes have $\\alpha 0$ and they exist above $\\omega_0$ which implies that these modes are damped. As an application of these findings we compute the thermal conductivity due to spin diffusive modes and show that it scales as $T^{1/2} \\exp(-2\\omega_0/T)$ in the case where their two-by-two scattering cross-section is weakly dependent on temperature.
Advances in passive neutron instruments for safeguards use
International Nuclear Information System (INIS)
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 240Pu-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
Advances in passive neutron instruments for safeguards use
Energy Technology Data Exchange (ETDEWEB)
Menlove, H.O.; Krick, M.S.; Langner, D.G.; Miller, M.C.; Stewart, J.E.
1994-02-01
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 {sup 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.
Advances in detectors for single crystal neutron diffraction
International Nuclear Information System (INIS)
Parallax error severely limits the use of gas detectors for neutron and X-ray single-crystal diffractometry. This is particularly the case in Neutron Macromolecule Crystallography (NMC), where a large solid angle is needed. High-resolution cylindrical detectors based on Neutron Image Plates provide a large angular coverage but they are gamma sensitive and do not allow time-resolved measurements. We describe the principle of a gas counter, the multi-blade Microstrip Gas Chamber (MSGC), which has the potential to design parallax-free cylindrical detectors. It contains radial MSGC plates, providing a partitioning of the gas volume and a segmentation of the signal readout. The axial coordinate is measured by reading out the charge signal on the anodes and the trans-axial coordinate is deduced from the electron drift distance measured optically using the time difference between the primary and the avalanche light. Expected characteristics are a sub-millimeter position resolution, a sub-microsecond time resolution, and a global counting rate greater than 107s-1
Advances in detectors for single crystal neutron diffraction
Energy Technology Data Exchange (ETDEWEB)
Buffet, J.C. [Institut Laue Langevin, 6 rue Jules Horowitz, 38042 Grenoble (France); Clergeau, J.F. [Institut Laue Langevin, 6 rue Jules Horowitz, 38042 Grenoble (France); Cooper, R.G. [SNS, Oak Ridge National Laboratory, Oak Ridge, TN 37821 (United States); Darpentigny, J. [Laboratoire Leon Brillouin, CEA-CNRS-Saclay, 91191 Gif-sur-Yvette (France); De Laulany, A. [Laboratoire Leon Brillouin, CEA-CNRS-Saclay, 91191 Gif-sur-Yvette (France); Fermon, C. [Laboratoire Leon Brillouin, CEA-CNRS-Saclay, 91191 Gif-sur-Yvette (France); Fetal, S. [LIP-Coimbra, 3004-516 Coimbra (Portugal); Fraga, F. [LIP-Coimbra, 3004-516 Coimbra (Portugal); Guerard, B. [Institut Laue Langevin, 6 rue Jules Horowitz, 38042 Grenoble (France)]. E-mail: guerard@ill.fr; Kampmann, R. [GKSS Forschungszentrum, Max Planck Strasse 1, 21502 Geesthacht (Germany); Kastenmueller, A. [Physik Department, Tchnische Universitaet Muenchen, 85748 Garchin (Germany); Mc Intyre, G.J. [Institut Laue Langevin, 6 rue Jules Horowitz, 38042 Grenoble (France); Manzin, G.; Meilleur, F.; Millier, F. [Institut Laue Langevin, 6 rue Jules Horowitz, 38042 Grenoble (France); Rhodes, N. [Rutherford Appleton Laboratory, ISIS facility, CCLRC, Chilton, Didcot, Oxfordshire OX11 0QX (United Kingdom); Rosta, L. [BNC-Research Institute for Solid State Physics and Optics, 1525 Budapest, POB 49 (Hungary); Schooneveld, E. [Rutherford Appleton Laboratory, ISIS facility, CCLRC, Chilton, Didcot, Oxfordshire OX11 0QX (United Kingdom); Smith, G.C. [Brookhaven National Laboratory, Upron, NY 11973-5001 (United States); Takahashi, H. [School of Engineering, The University of Tokyo, 7-3-1 Hongo, Bunkyo-ku, Tokyo 113-8656 (Japan); Van Esch, P.; Van Vuure, T.L. [Institut Laue Langevin, 6 rue Jules Horowitz, 38042 Grenoble (France); Zeitelhack, K. [Physik Department, Tchnische Universitaet Muenchen, 85748 Garchin (Germany)
2005-12-01
Parallax error severely limits the use of gas detectors for neutron and X-ray single-crystal diffractometry. This is particularly the case in Neutron Macromolecule Crystallography (NMC), where a large solid angle is needed. High-resolution cylindrical detectors based on Neutron Image Plates provide a large angular coverage but they are gamma sensitive and do not allow time-resolved measurements. We describe the principle of a gas counter, the multi-blade Microstrip Gas Chamber (MSGC), which has the potential to design parallax-free cylindrical detectors. It contains radial MSGC plates, providing a partitioning of the gas volume and a segmentation of the signal readout. The axial coordinate is measured by reading out the charge signal on the anodes and the trans-axial coordinate is deduced from the electron drift distance measured optically using the time difference between the primary and the avalanche light. Expected characteristics are a sub-millimeter position resolution, a sub-microsecond time resolution, and a global counting rate greater than 10{sup 7}s{sup -1}.
Recent advances and future trends in neutron resonance spectroscopy
International Nuclear Information System (INIS)
Neutron resonance spectroscopy contributes primarily to two areas of nuclear physics: 1.) in medium weight and heavy nuclei with a high level density it tests their statistical properties, and 2.) in nuclei with a sufficiently low level density, i.e. light nuclei (A ≤ 50) and nuclei around /sup 208/Pb, it investigates nuclear structure at several MeV excitation energy. In the first field, recent years have seen growing knowledge and understanding of nuclear level densities and their spin and parity dependence. Several questions basic to the statistical properties of nuclei, although extensively studied in the past, are still open: the statistical distribution of partial widths; possible narrow energy variations of the average partial widths; and correlations between partial widths for different reaction channels. The major progress has occurred and will continue to take place in the field of light nuclei: improved resolution of neutron time-of-flight spectrometers yields detailed resonance data over an extended energy range, and model calculations become possible which will allow detailed comparison to experimental data. The main subjects of interest are the distributions of neutron, as well as radiative strengths and their interpretation in terms of nuclear structure
Recent advances and future trends in neutron resonance spectroscopy
International Nuclear Information System (INIS)
Neutron resonance spectroscopy contributes primarily to two areas of nuclear physics: In medium weight and heavy nuclei with a high level density it tests their statistical properties; in nuclei with a sufficiently low level density, i.e. light nuclei (A 208Pb, it investigates nuclear structure at several MeV excitation energy. In the first field, recent years have seen growing knowledge and understanding of nuclear level densities and their spin and parity dependence. Several questions basic to the statistical properties of nuclei, although extensively studied in the past, are still open: the statistical distribution of partial widths; possible narrow energy variations of the average partial widths; and correlations between partial widths for different reaction channels. The major progress has occured and will continue to take place in the field of light nuclei: Improved resolution of neutron time-of-flight spectrometers yields detailed resonance data over an extended energy range, and model calculations become possible which will allow detailed comparison to experimental data. The main subjects of interest are the distributions of neutron- as well as radiative strengths and their interpretation in terms of nuclear structure. (author)
Symmetry and Solution of Neutron Transport Equations in Nonhomogeneous Media
2014-01-01
We propose the group-theoretical approach which enables one to generate solutions of equations of mathematical physics in nonhomogeneous media from solutions of the same problem in a homogeneous medium. The efficiency of this method is illustrated with examples of thermal neutron diffusion problems. Such problems appear in neutron physics and nuclear geophysics. The method is also applicable to nonstationary and nonintegrable in quadratures differential equations.
Resolution of the neutron transport equation by massively parallel computer in the Cronos code
International Nuclear Information System (INIS)
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)
In 1953, the first of what was supposed to be two volumes on neutron transport theory was published. The monograph, entitled open-quotes Introduction to the Theory of Neutron Diffusionclose quotes by Case et al., appeared as a Los Alamos National Laboratory report and was to be followed by a second volume, which never appeared as intended because of the death of Placzek. Instead, Case and Zweifel collaborated on the now classic work entitled Linear Transport Theory 2 in which the underlying mathematical theory of linear transport was presented. The initial monograph, however, represented the coming of age of neutron transport theory, which had its roots in radiative transfer and kinetic theory. In addition, it provided the first benchmark results along with the mathematical development for several fundamental neutron transport problems. In particular, one-dimensional infinite medium Green's functions for the monoenergetic transport equation in plane and spherical geometries were considered complete with numerical results to be used as standards to guide code development for applications. Unfortunately, because of the limited computational resources of the day, some numerical results were incorrect. Also, only conventional mathematics and numerical methods were used because the transport theorists of the day were just becoming acquainted with more modern mathematical approaches. In this paper, Green's function solution is revisited in light of modern numerical benchmarking methods with an emphasis on evaluation rather than theoretical results. The primary motivation for considering the Green's function at this time is its emerging use in solving finite and heterogeneous media transport problems
Application of neutron/gamma transport codes for the design of explosive detection systems
International Nuclear Information System (INIS)
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
Institute of Scientific and Technical Information of China (English)
2008-01-01
A discrete ordinates method for a threedimensional first-order neutron transport equation based on unstructured-meshes that avoids the singularity of the second-order neutron transport equation in void regions was derived.The finite element variation equation was obtained using the least-squares method.A three-dimensional transport calculation code was developed.Both the triangular-z and the tetrahedron elements were included.The numerical results of some benchmark problems demonstrated that this method can solve neutron transport problems in unstructuredmeshes very well.For most problems,the error of the eigenvalue and the angular flux is less than 0.3% and 3.0% respectively.
International Nuclear Information System (INIS)
Due to the well-known problem of 3He shortage, a series of different thermal neutron detectors alternative to helium tubes are being developed, with the goal to find valid candidates for detection systems for the future spallation neutron sources such as the European Spallation Source (ESS). A possible 3He-free detector candidate is a charged particle detector equipped with a three dimensional neutron converter cathode (3D-C). The 3D-C currently under development is composed by a series of alumina (Al2O3) lamellas coated by 1 μ m of 10B enriched boron carbide (B4C). In order to obtain a good characterization in terms of detector efficiency and uniformity it is crucial to know the thickness, the uniformity and the atomic composition of the B4C neutron converter coating. In this work a non-destructive technique for the characterization of the lamellas that will compose the 3D-C was performed using neutron radiography. The results of these measurements show that the lamellas that will be used have coating uniformity suitable for detector applications. This technique (compared with SEM, EDX, ERDA, XPS) has the advantage of being global (i.e. non point-like) and non-destructive, thus it is suitable as a check method for mass production of the 3D-C elements
Muraro, A.; Albani, G.; Perelli Cippo, E.; Croci, G.; Angella, G.; Birch, J.; Cazzaniga, C.; Caniello, R.; Dell'Era, F.; Ghezzi, F.; Grosso, G.; Hall-Wilton, R.; Höglund, C.; Hultman, L.; Schimdt, S.; Robinson, L.; Rebai, M.; Salvato, G.; Tresoldi, D.; Vasi, C.; Tardocchi, M.
2016-03-01
Due to the well-known problem of 3He shortage, a series of different thermal neutron detectors alternative to helium tubes are being developed, with the goal to find valid candidates for detection systems for the future spallation neutron sources such as the European Spallation Source (ESS). A possible 3He-free detector candidate is a charged particle detector equipped with a three dimensional neutron converter cathode (3D-C). The 3D-C currently under development is composed by a series of alumina (Al2O3) lamellas coated by 1 μ m of 10B enriched boron carbide (B4C). In order to obtain a good characterization in terms of detector efficiency and uniformity it is crucial to know the thickness, the uniformity and the atomic composition of the B4C neutron converter coating. In this work a non-destructive technique for the characterization of the lamellas that will compose the 3D-C was performed using neutron radiography. The results of these measurements show that the lamellas that will be used have coating uniformity suitable for detector applications. This technique (compared with SEM, EDX, ERDA, XPS) has the advantage of being global (i.e. non point-like) and non-destructive, thus it is suitable as a check method for mass production of the 3D-C elements.
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.
Transport synthetic acceleration scheme for multi-dimensional neutron transport problems
International Nuclear Information System (INIS)
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)
TDTORT: Time-Dependent, 3-D, Discrete Ordinates, Neutron Transport Code System with Delayed Neutrons
International Nuclear Information System (INIS)
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
Modeling heat generation and flow in the Advanced Neutron Source Corrosion Test Loop specimen
International Nuclear Information System (INIS)
A finite difference computer code HEATING5 was used to model heat generation and flow in a typical experiment envisioned for the Advanced Neutron Source Corrosion Test Loop. The electrical resistivity and thermal conductivity of the test specimen were allowed to vary with local temperature, and the corrosion layer thickness was assigned along the length of the specimen in the manner predicted by the Griess Correlation. The computer solved the two-dimensional transport problem for a given total power dissipated in the specimen and stipulated coolant temperatures and water-side heat-transfer coefficients. The computed specimen temperatures were compared with those calculated on the basis of approximate analytical equations involving the total power dissipation and the assignment of the physical properties based on temperatures at single axial points on the specimen. The comparisons indicate that when temperature variations are large along the axis of the specimen, the variation in local heat flux should not be overlooked when using approximate equations or models. The approximate equations are most accurate near the center of the specimen where the heat flux remains closest to the average value, and in that region the calculated quantities agree closely with the results of the computer code. 4 figs., 1 tab
The advanced neutron source--designing to meet the needs of the user community
International Nuclear Information System (INIS)
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
Key nuclear data measurements for advanced fission energy and white neutron source at CSNS
International Nuclear Information System (INIS)
The key nuclear data for advanced fission energy are important in designing advanced nuclear reactors and facilities for nuclear-waste transmutation. Because the present nuclear data library is limited by experimental condition and energy range, the precision of some nuclear data is low, even some nuclear data are blank. In this paper, the status of the nuclear data and white neutron sources were presented. The back-streaming neutron beam at China Spallation Neutron Source (CSNS) has very wide energy spectrum (0.01 eV-200 MeV) and excellent time structure. From the simulation results, it's obtained that the uncollimated neutron fluence rate is around 9.3 × 106 cm-2 · s-1 within the given energy range at 80 m away from the target, which accounts for about 53% of the total neutrons. The time resolution of 0.3%-0.9%, which is important for the Time-of-Flight method, can be obtained for both the parasite operation mode with two proton bunches and the dedicated operation mode with a single proton bunch. CSNS white neutron source will be a good facility for nuclear data measurement. (authors)
Performance of the advanced cold neutron source and optics upgrades at the NIST Research Reactor
International Nuclear Information System (INIS)
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
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.
Neutron reaction cross section data for advanced nuclear applications
International Nuclear Information System (INIS)
Full text of publication follows: Worldwide major research efforts are currently being carried out in order to develop a new concept of nuclear power generation, so-called accelerator driven systems (ADS) for energy production and transmutation of radioactive nuclear waste. A suggested approach is the energy amplifier (EA), which is a sub-critical reactor using a powerful proton accelerator and a spallation reaction as neutron source. Since the EA is based on the thorium-uranium fuel cycle, where the natural resources of the main fuel thorium are estimated to last for hundred thousands of years, it is considered to provide clean and almost inexhaustible nuclear energy. Apart from necessary new technical developments, the realization of these concepts depends strongly on the availability of accurate nuclear reaction data. In particular, precise knowledge about cross sections for fission, neutron capture and scattering is required for the nuclides involved in the Th-U fuel cycle. Among the first priority isotopes the IAEA had pointed out 231Pa and 233Pa. The latter one, 233Pa, is of specific interest, since it plays an important role as an intermediate nucleus in the formation of the fissile 233U from the fertile 232Th. With its half life of 27.0 days for β-decay, 233Pa is not a 'long-lived' nucleus, but it still requires careful attention in the design and operation of thorium-fueled reactors. When a thorium-fueled reactor is stopped, the present amount of 233Pa will continue to decay into 233U, leading to an increase in reactivity, which may even cause criticality. This mechanism is known as 'protactinium effect' and is proportional to the power level of the reactor. Also the precise knowledge of the fission cross section of 231Pa (above 1 b for fast neutrons) is essential for simulations of the balance of nuclei in and, thus, the reactivity behavior of the reactor. We present recent cross section data from direct, energy resolved measurements of the neutron
Experimental validation of a coupled neutron-photon inverse radiation transport solver
International Nuclear Information System (INIS)
Forward radiation transport is the problem of calculating the radiation field given a description of the radiation source and transport medium. In contrast, inverse transport is the problem of inferring the configuration of the radiation source and transport medium from measurements of the radiation field. As such, the identification and characterization of special nuclear materials (SNM) is a problem of inverse radiation transport, and numerous techniques to solve this problem have been previously developed. The authors have developed a solver based on nonlinear regression applied to deterministic coupled neutron-photon transport calculations. The subject of this paper is the experimental validation of that solver. This paper describes a series of experiments conducted with a 4.5-kg sphere of alpha-phase, weapons-grade plutonium. The source was measured in six different configurations: bare, and reflected by high-density polyethylene (HDPE) spherical shells with total thicknesses of 1.27, 2.54, 3.81, 7.62, and 15.24 cm. Neutron and photon emissions from the source were measured using three instruments: a gross neutron counter, a portable neutron multiplicity counter, and a high-resolution gamma spectrometer. These measurements were used as input to the inverse radiation transport solver to characterize the solver's ability to correctly infer the configuration of the source from its measured signatures.
ORNL contributions to the Advanced Neutron Source (ANS) Project for October 1986-March 1987
International Nuclear Information System (INIS)
The Advanced Neutron Source (ANS) Facility - formerly called the Center for Neutron Research - will provide the world's best facilities for the study of neutron scattering. The ANS high power density reactor will be fueled with uranium silicide and cooled, moderated, and reflected by D2O. Peak neutron fluxes in the reflector are expected to be 5 to 10 x 1019 neutrons per square meter with a power level between 270 MW and 300 MW. This report describes the status of technical work at ORNL on the ANS Project during the first half of FY 1987. The scope of this report includes Research and Development Tasks; Safety Tasks; Conceptual Design Tasks; and Project Support. The last two areas were only initiated as separate activities during this reporting period. Technical highlights include a better understanding of the relationship among neutron flux, core power, and core volume; preconceptual design work on a cold source for use in a very high gamma and neutron flux environment; identification of the major applicable safety rules and guidelines; and establishment of initial functional objectives for the containment structure
Detecting binary neutron star systems with spin in advanced gravitational-wave detectors
Brown, Duncan A; Lundgren, Andrew; Nitz, Alexander H
2012-01-01
The detection of gravitational waves from binary neutron stars is a major goal of the gravitational-wave observatories Advanced LIGO and Advanced Virgo. Previous searches for binary neutron stars with LIGO and Virgo neglected the component stars' angular momentum (spin). We demonstrate that neglecting spin in matched-filter searches causes advanced detectors to lose more than 3% of the possible signal-to-noise ratio for 59% (6%) of sources, assuming that neutron star dimensionless spins, $cJ/GM^2$, are uniformly distributed with magnitudes between 0 and 0.4 (0.05) and that the neutron stars have isotropically distributed spin orientations. We present a new method of constructing filter banks for advanced-detector searches, which can create template banks of signals with non-zero spins that are (anti-)aligned with the orbital angular momentum. We show that this search loses more than 3% of the maximium signal-to-noise for only 9% (0.2%) of BNS sources with dimensionless spins between 0 and 0.4 (0.05) and isotr...
Advanced Reactors Thermal Energy Transport for Process Industries
Energy Technology Data Exchange (ETDEWEB)
P. Sabharwall; S.J. Yoon; M.G. McKellar; C. Stoots; George Griffith
2014-07-01
The operation temperature of advanced nuclear reactors is generally higher than commercial light water reactors and thermal energy from advanced nuclear reactor can be used for various purposes such as liquid fuel production, district heating, desalination, hydrogen production, and other process heat applications, etc. Some of the major technology challenges that must be overcome before the advanced reactors could be licensed on the reactor side are qualification of next generation of nuclear fuel, materials that can withstand higher temperature, improvement in power cycle thermal efficiency by going to combined cycles, SCO2 cycles, successful demonstration of advanced compact heat exchangers in the prototypical conditions, and from the process side application the challenge is to transport the thermal energy from the reactor to the process plant with maximum efficiency (i.e., with minimum temperature drop). The main focus of this study is on doing a parametric study of efficient heat transport system, with different coolants (mainly, water, He, and molten salts) to determine maximum possible distance that can be achieved.
In situ studies of mass transport in liquid alloys by means of neutron radiography.
Kargl, F; Engelhardt, M; Yang, F; Weis, H; Schmakat, P; Schillinger, B; Griesche, A; Meyer, A
2011-06-29
When in situ techniques became available in recent years this led to a breakthrough in accurately determining diffusion coefficients for liquid alloys. Here we discuss how neutron radiography can be used to measure chemical diffusion in a ternary AlCuAg alloy. Neutron radiography hereby gives complementary information to x-ray radiography used for measuring chemical diffusion and to quasielastic neutron scattering used mainly for determining self-diffusion. A novel Al(2)O(3) based furnace that enables one to study diffusion processes by means of neutron radiography is discussed. A chemical diffusion coefficient of Ag against Al around the eutectic composition Al(68.6)Cu(13.8)Ag(17.6) at.% was obtained. It is demonstrated that the in situ technique of neutron radiography is a powerful means to study mass transport properties in situ in binary and ternary alloys that show poor x-ray contrast. PMID:21654050
Transport modeling and gyrokinetic analysis of advanced high performance discharges
International Nuclear Information System (INIS)
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
Energy Technology Data Exchange (ETDEWEB)
Risch, P.; Dekens, O.; Ait Abderrahim, H. [SCK-CEN, Fuel Research Department, (Belgium); Wouters, R. de [Tractebel, Energy Engineering, (Belgium)
1997-10-01
Neutron dosimetry experiments seem to point our difficulties in the treatment of large water thickness like those encountered between the core baffle and the pressure vessel. This paper describes the theoretical benchmark undertaken by EDF, SCK/CEN and TRACTEBEL ENERGY ENGINEERING, concerning the transport of fast neutrons throughout a one meter cube of water, located after a U-235 fission sources plate. The results showed no major discrepancies between the calculations up to 50 cm from the source, accepting that a P3 development of the Legendre polynomials is necessary for the Sn calculations. The main differences occurred after 50 cm, reaching 20 % at the end of the water cube. This results lead us to consider an experimental benchmark, dedicated to the problem of fast neutron deep penetration in water, which has been launched at SCK/CEN. (authors). 7 refs.
Development of deterministic transport methods for low energy neutrons for shielding in space
Ganapol, Barry
1993-01-01
Transport of low energy neutrons associated with the galactic cosmic ray cascade is analyzed in this dissertation. A benchmark quality analytical algorithm is demonstrated for use with BRYNTRN, a computer program written by the High Energy Physics Division of NASA Langley Research Center, which is used to design and analyze shielding against the radiation created by the cascade. BRYNTRN uses numerical methods to solve the integral transport equations for baryons with the straight-ahead approximation, and numerical and empirical methods to generate the interaction probabilities. The straight-ahead approximation is adequate for charged particles, but not for neutrons. As NASA Langley improves BRYNTRN to include low energy neutrons, a benchmark quality solution is needed for comparison. The neutron transport algorithm demonstrated in this dissertation uses the closed-form Green's function solution to the galactic cosmic ray cascade transport equations to generate a source of neutrons. A basis function expansion for finite heterogeneous and semi-infinite homogeneous slabs with multiple energy groups and isotropic scattering is used to generate neutron fluxes resulting from the cascade. This method, called the FN method, is used to solve the neutral particle linear Boltzmann transport equation. As a demonstration of the algorithm coded in the programs MGSLAB and MGSEMI, neutron and ion fluxes are shown for a beam of fluorine ions at 1000 MeV per nucleon incident on semi-infinite and finite aluminum slabs. Also, to demonstrate that the shielding effectiveness against the radiation from the galactic cosmic ray cascade is not directly proportional to shield thickness, a graph of transmitted total neutron scalar flux versus slab thickness is shown. A simple model based on the nuclear liquid drop assumption is used to generate cross sections for the galactic cosmic ray cascade. The ENDF/B V database is used to generate the total and scattering cross sections for neutrons in
International Nuclear Information System (INIS)
This paper discusses salient aspects of severe accident related phenomenological considerations, scoping studies, and mitigative design features being studied for incorporation into a high-power research reactor plant. Key results of scoping studies on steam explosions, recriticality, core-concrete interactions, and containment transport are highlighted. Evolving design features of the containment are described. Containment response calculations for a site-suitability basis transient are presented that demonstrate acceptable source term values and superior containment performance. Oak Ridge National Laboratory's (ORNL) Advanced Neutron Source (ANS) will be a new user facility for all kinds of neutron research, centered around a research reactor of unprecedented neutron beam flux. A defense-in-depth philosophy has been adopted. In response to this commitment, ANS Project management initiated severe accident analysis and related technology development early-on in the design phase itself. This was done to aid in designing a sufficiently robust containment for retention and controlled release of radionuclides in the event of such an accident. It also provides a means for satisfying on- and off-site regulatory requirements, accident-related dose exposures, and containment response and source-term best-estimate analyses for level-2 and -3 Probabilistic Risk Analysis (PRAs) that will be produced. Moreover, it will provide the best possible understanding of the ANS under severe accident conditions and consequently provide insights for the development of strategies and design philosophies for accident mitigation, management, and emergency preparedness efforts
Energy Technology Data Exchange (ETDEWEB)
Poveschenko, T.; Poveschenko, O. [RRC Kurchatov Inst., Kurchatov square, 1, 123182, Moscow (Russian Federation)
2012-07-01
This paper presents the new approach to creation of geometrical module for nuclear reactor neutron transport computer simulation analysis so called the differential cross method. It is elaborated for detecting boards between physical zones. It is proposed to use GMSH open source mesh editor extended by some features: a special option and a special kind of mesh (cubic background mesh).This method is aimed into Monte Carlo Method as well as for deterministic neutron transport methods. Special attention is attended for reactor core composed of a set of material zones with complicate geometrical boundaries. The idea of this approach is described. In general case method works for 3-D space. Algorithm of creation of the geometrical module is given. 2-D neutron transport benchmark-test for RBMK reactor cluster cell is described. It demonstrates the ability of this approach to provide flexible definition of geometrical meshing with preservation of curved surface or any level of heterogeneity. (authors)
International Nuclear Information System (INIS)
This paper presents the new approach to creation of geometrical module for nuclear reactor neutron transport computer simulation analysis so called the differential cross method. It is elaborated for detecting boards between physical zones. It is proposed to use GMSH open source mesh editor extended by some features: a special option and a special kind of mesh (cubic background mesh).This method is aimed into Monte Carlo Method as well as for deterministic neutron transport methods. Special attention is attended for reactor core composed of a set of material zones with complicate geometrical boundaries. The idea of this approach is described. In general case method works for 3-D space. Algorithm of creation of the geometrical module is given. 2-D neutron transport benchmark-test for RBMK reactor cluster cell is described. It demonstrates the ability of this approach to provide flexible definition of geometrical meshing with preservation of curved surface or any level of heterogeneity. (authors)
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.
Improved Algorithms and Coupled Neutron-Photon Transport for Auto-Importance Sampling Method
Wang, Xin; Qiu, Rui; Li, Chun-Yan; Liang, Man-Chun; Zhang, Hui; Li, Jun-Li
2016-01-01
Auto-Importance Sampling (AIS) method is a Monte Carlo variance reduction technique proposed by Tsinghua University for deep penetration problem, which can improve computational efficiency significantly without pre-calculations for importance distribution. However AIS method is only validated with several basic deep penetration problems of simple geometries and cannot be used for coupled neutron-photon transport. This paper firstly presented the latest algorithm improvements for AIS method including particle transport, fictitious particles creation and adjustment, fictitious surface geometry, random number allocation and calculation of estimated relative error, which made AIS method applicable to complicated deep penetration problem. Then, a coupled Neutron-Photon Auto-Importance Sampling (NP-AIS) method was proposed to apply AIS method with the improved algorithms in coupled neutron-photon Monte Carlo transport. Finally, the NUREG/CR-6115 PWR benchmark model was calculated with the method of geometry splitti...
Analytical benchmarks for nuclear engineering applications. Case studies in neutron transport theory
International Nuclear Information System (INIS)
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)
Computational Transport Modeling of High-Energy Neutrons Found in the Space Environment
Cox, Brad; Theriot, Corey A.; Rohde, Larry H.; Wu, Honglu
2012-01-01
The high charge and high energy (HZE) particle radiation environment in space interacts with spacecraft materials and the human body to create a population of neutrons encompassing a broad kinetic energy spectrum. As an HZE ion penetrates matter, there is an increasing chance of fragmentation as penetration depth increases. When an ion fragments, secondary neutrons are released with velocities up to that of the primary ion, giving some neutrons very long penetration ranges. These secondary neutrons have a high relative biological effectiveness, are difficult to effectively shield, and can cause more biological damage than the primary ions in some scenarios. Ground-based irradiation experiments that simulate the space radiation environment must account for this spectrum of neutrons. Using the Particle and Heavy Ion Transport Code System (PHITS), it is possible to simulate a neutron environment that is characteristic of that found in spaceflight. Considering neutron dosimetry, the focus lies on the broad spectrum of recoil protons that are produced in biological targets. In a biological target, dose at a certain penetration depth is primarily dependent upon recoil proton tracks. The PHITS code can be used to simulate a broad-energy neutron spectrum traversing biological targets, and it account for the recoil particle population. This project focuses on modeling a neutron beamline irradiation scenario for determining dose at increasing depth in water targets. Energy-deposition events and particle fluence can be simulated by establishing cross-sectional scoring routines at different depths in a target. This type of model is useful for correlating theoretical data with actual beamline radiobiology experiments. Other work exposed human fibroblast cells to a high-energy neutron source to study micronuclei induction in cells at increasing depth behind water shielding. Those findings provide supporting data describing dose vs. depth across a water-equivalent medium. This
International Nuclear Information System (INIS)
Originally the work was oriented towards two main topics: a) difference and integral methods in neutron transport theory. Two computers were used for numerical calculations GIER and CYBER-72. During the first year the main effort was shifted towards basic theoretical investigations. At the first step the ANIS code was adopted and later modified to check various finite difference approaches against each other. Then the general finite element method and the singular perturbation method were developed. The analysis of singularities of the one-dimensional neutron transport equation in spherical geometry has been done and presented. Later the same analysis for the case of cylindrical symmetry has been carried out. The second and the third year programme included the following topics: 1) finite difference methods in stationary neutron transport theory; 2)mathematical fundamentals of approximate methods for solving the transport equation; 3) singular perturbation method for the time-dependent transport equation; 4) investigation of various iterative procedures in reactor calculations. This investigation will help to better understanding of the mathematical basis for existing and developed numerical methods resulting in more effective algorithms for reactor computer codes
Energy Technology Data Exchange (ETDEWEB)
Nigg, D; Wessol, D; Wemple, C; Harkin, G; Hartmann-Siantar, C
2002-08-20
The Idaho National Engineering and Environmental Laboratory (INEEL) has long been active in development of advanced Monte-Carlo based computational dosimetry and treatment planning methods and software for advanced radiotherapy, with a particular focus on Neutron Capture Therapy (NCT) and, to a somewhat lesser extent, Fast-Neutron Therapy. The most recent INEEL software system of this type is known as SERA, Simulation Environment for Radiotherapy Applications. As a logical next step in the development of modern radiotherapy planning tools to support the most advanced research, INEEL and Lawrence Livermore National Laboratory (LLNL), the developers of the PEREGRTNE computational engine for radiotherapy treatment planning applications, have recently launched a new project to collaborate in the development of a ''next-generation'' multi-modality treatment planning software system that will be useful for all modern forms of radiotherapy.
Possibility of neutron transport cross section measurement in a sphere surrounded by moderation
International Nuclear Information System (INIS)
The possibility of an estimation of the neutron macroscopic transport cross section for a medium with known adsorption cross section is presented. A two-region spherical system is used with the sample of interest as the inner sphere. The fundamental decay constant of the thermal neutron flux is calculated on the basis of diffusion theory for such a system as a function of the dimensions of the external sphere and/or the macroscopic absorption cross section of the inner medium. The influence of the diffusion cooling coefficient and the hydrogen content in the inner sphere on the transport cross section estimation is discussed. (author)
Multigroup neutron transport equation in the diffusion and P1 approximation
International Nuclear Information System (INIS)
Investigations of the properties of the multigroup transport operator, width and without delayed neutrons in the diffusion and P1 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 P1 approximation possesses a complete set of generalized eigenvectors. A formal solution to the initial value problem is also given. (author)
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.
Thermal-hydraulic studies of the Advanced Neutron Source cold source
International Nuclear Information System (INIS)
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
Measurements of neutron cross sections for advanced nuclear energy systems at n_TOF (CERN
Directory of Open Access Journals (Sweden)
Barbagallo M.
2014-03-01
Full Text Available The n_TOF facility operates at CERN with the aim of addressing the request of high accuracy nuclear data for advanced nuclear energy systems as well as for nuclear astrophysics. Thanks to the features of the neutron beam, important results have been obtained on neutron induced fission and capture cross sections of U, Pu and minor actinides. Recently the construction of another beam line has started; the new line will be complementary to the first one, allowing to further extend the experimental program foreseen for next measurement campaigns.
Effects of fuel particle size distributions on neutron transport in stochastic media
International Nuclear Information System (INIS)
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 (keff) 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
International Nuclear Information System (INIS)
Highlights: → Meyer's sub-space iteration method is used to evaluate dominant prompt time eigenvalues of neutron transport equation. → Mono-energetic 1-D benchmark problems are analysed. → The method is found to correctly compute complex eigenvalues also. - Abstract: The aim of this paper is to explore the use of Meyer's sub-space iteration (SSI) method for the evaluation of dominant prompt time-eigenvalues of the neutron transport equation. The integro-differential form of the transport equation is considered. The SSI method is known to be an efficient technique to find the dominant eigenvalues of a non-symmetric matrix. It has been earlier used for eigenvalue problems in neutron diffusion theory. However, it does not seem to be tried in the transport theory case. Here, the use of SSI has been tested in transport theory for some 1-D mono-energetic homogeneous and heterogeneous benchmark problems. The space variable is discretised by finite differencing while neutron directions are discretised by discrete ordinates (Sn-) method. The SSI method needs frequent multiplication of the relevant matrix operator with vectors. As known from earlier works in this area, this can be achieved in terms of external source calculations for which a 1-D programme was developed and used. With the availability of more versatile Sn-method codes, it may perhaps be possible to extend use of SSI to more realistic cases.
Flexible polyvinyl chloride neutron guides for transporting ultracold and very cold neutrons
Energy Technology Data Exchange (ETDEWEB)
Arzumanov, S. S., E-mail: sarzumanov@yandex.ru; Bondarenko, L. N. [Russian Research Center Kurchatov Institute (Russian Federation); Geltenbort, P. [Institut Laue-Langevin (France); Morozov, V. I. [Russian Research Center Kurchatov Institute (Russian Federation); Nesvizhevsky, V. V. [Institut Laue-Langevin (France); Panin, Yu. N.; Strepetov, A. N.; Chuvilin, D. Yu. [Russian Research Center Kurchatov Institute (Russian Federation)
2011-12-15
The transmission of ultracold neutrons (UCNs) through flexible polyvinyl chloride (PVC) tubes with lengths of up to 3 m and an internal diameter of 6-8 mm has been studied. High UCN transmission is found even for arbitrarily bent tubes (single bend, double bend, triple bend, figure eight, etc.). The transmission can be improved significantly by coating the inner surface of the tube with a thin layer of liquid fluorine polymer. The prospects of these neutron guides in fundamental and applied research are discussed.
Flexible polyvinyl chloride neutron guides for transporting ultracold and very cold neutrons
International Nuclear Information System (INIS)
The transmission of ultracold neutrons (UCNs) through flexible polyvinyl chloride (PVC) tubes with lengths of up to 3 m and an internal diameter of 6–8 mm has been studied. High UCN transmission is found even for arbitrarily bent tubes (single bend, double bend, triple bend, figure eight, etc.). The transmission can be improved significantly by coating the inner surface of the tube with a thin layer of liquid fluorine polymer. The prospects of these neutron guides in fundamental and applied research are discussed.
PELAN - a transportable, neutron-based UXO identification technique
International Nuclear Information System (INIS)
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)
Neutron absorber plate and radioactive material transportation cask
International Nuclear Information System (INIS)
Aluminum alloy flame-coating layers are formed at the outer surface of a neutron absorber plate in order to prevent corrosion due to potential difference. However, pin holes of micron order are sometimes formed on the flame-coating membranes, which are hard to be found by usual inspection. Then, ferrous flame-coating membranes are formed at the outer surface of boron carbide and aluminum alloy flame-coating membranes are formed at the outer surface thereof. The outer surface of a boron carbide plate is coated with the ferrous flame-coating membranes instead of being coated with an external plate made of neutron cells, and an aluminum alloy flame-coating membranes or mixed flame-coating layers of aluminum oxide and titania are coated thereover in order to prevent rusts. Whether the pin holes are present or not can be confirmed easily by a ferroxyl test. If there are pin holes, flame-coating is applied again to form complete membranes. Then, since it is no more necessary to fix a neutron absorbing cell at the outer surface of a fuel cell by means of welding, production cost can be reduced. (N.H.)
International Nuclear Information System (INIS)
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)
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
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)
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)
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
International Nuclear Information System (INIS)
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
Monte Carlo simulations of the particle transport in semiconductor detectors of fast neutrons
International Nuclear Information System (INIS)
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 239Pu–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
International Nuclear Information System (INIS)
For the treatment of anisotropic elastic neutron scattering in Ssub(N) reactor calculations extended transport approximations are widely used, which in the simplest case describe the elastic anisotropy by the mean elastic-scattering cosine anti μ in the transport cross section Σsub(tr) = Σsub(t) - anti μΣsub(s). In the present paper this approximation is improved by higher-order transport approximations with transport cross sections that consistently take into account anisotropic neutron inscattering. The quality of different weighting procedures for the generation of anisotropic group constants in the resonance region is assessed. Elastic anisotropy increasing with neutron energy on one hand and weighting functions with resonance structure up to about 3 MeV on the other hand are connected by the use of numerically advantageous energy-dependent higher-order transport approximations. With the application of the usual heavy material weighting procedure a consistent transition from the structural-material resonance region to the heavy-material resonance region is achieved. It is shown: in a fine group structure of 208 energy groups the macroscopic shape of the weighting functions may be neglected, this shape however, is important in case of collapsing to coarse groups in different spatial zones. For the critical assembly ZPRIII-56B the above-mentioned methods together with consistently improved transport cross sections of the KFKINR group constant set yield ksub(eff) = 1.0066. The prediction of directional neutron spectra in a small lithium sphere with a 14 MeV neutron source is successful within an accuracy of 20% with respect to experimental measurements. (orig.)
On the Spectrum of Neutron Transport Equations with Reflecting Boundary Conditions
Song, Degong
2000-01-01
This dissertation is devoted to investigating the time dependent neutron transport equations with reflecting boundary conditions. Two typical geometries --- slab geometry and spherical geometry --- are considered in the setting of L^p including L^1. Some aspects of the spectral properties of the transport operator A and the strongly continuous semigroup T(t) generated by A are studied. It is shown under fairly general assumptions that the accumulation points of { m Pas...
One-speed neutron transport eigenvalues review and some new results
International Nuclear Information System (INIS)
The eigenvalue problem for one-speed neutron transport in stationary and time-dependent systems is stated. Various types of anisotropic scattering and boundary conditions are discussed. The calculation methods used for solving the transport equation in its differential or integral form are described. A review is given of the work done at our institute on homogenous and heterogenous systems. Preliminary results from some recent investigations are also presented. (44 refs.)
MC++: A parallel, portable, Monte Carlo neutron transport code in C++
International Nuclear Information System (INIS)
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
International Nuclear Information System (INIS)
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
DIAMANT2 - A multigroup neutron transport program for triangular and hexagonal geometry
International Nuclear Information System (INIS)
DIAMANT2 evolved out of the DIAMANT-code. DIAMANT2 solves the multigroup neutron transport equation in planar geometry using the Ssub(N) method. Spatial discretization is accomplished by taking finite differences on a meshgrid composed of equilateral triangles. This report contains a detailed documentation of the program and the input description. (orig./HJ)
Quadrature sums of highest algebraic degree of precision for neutron transport integrals
International Nuclear Information System (INIS)
A Gaussian-type quadrature formula for neutron transport integrals is here re-established according to the orthogonal-polynomial method. Limit properties of the asymptotic and nonasymptotic parts of the quadrature sum are also obtained, together with another quadrature formula of the highest algebraic degree of precision, for purely scattering media. Numerical-application examples are given in the appendix. (author)
Existence result for the kinetic neutron transport problem with a general albedo boundary condition
International Nuclear Information System (INIS)
We present an existence result for the kinetic neutron transport equation with a general albedo boundary condition. The proof is constructive in the sense that we build a sequence that converges to the solution of the problem by iterating on the albedo term. Both nonhomogeneous and albedo boundary conditions are studied. (authors)
Two-group neutron transport theory in adjacent space with lineary anisotropic scattering
International Nuclear Information System (INIS)
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
Development of a 3D neutron transport code and benchmark tests
International Nuclear Information System (INIS)
Results are reported of NEACRP '3D Neutron Transport Benchmarks' proposed from Osaka UNiversity, and of recent progress in the development of a 3D neutron transport code. Takeda et al. proposed four problems to NEACRP as 3D neutron transport benchmarks, and 22 results from 20 organizations were submitted. A variety of methods have been used, such as the Monte Carlo, Sn, Pn, synthetic, and nodal method. The results for k-eff, control-rod worths, and region-averaged fluxes are summarized with the conclusions that (1) in XYZ geometry the Sn method with n=8 shows a good agreement with the Monte-Carlo method, and gives even better results in some cases, (2) the Pn method has significant spatial mesh effects, and (3) the Sn method is not satisfactory in hexagonal-Z geometry, and improvements in accuracy are desirable. Improvement of a 3D neutron transport code is in progress to resolve the problem in the hexagonal-Z geometry by considering new diamond difference schemes and an improved coarse-mesh method, and also by applying the nodal method. (author)
The neutron transport code DTF-Traca users manual and input data
International Nuclear Information System (INIS)
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
Monte Carlo neutron transport simulation of the Ghana Research Reactor-1
International Nuclear Information System (INIS)
Stochastic Monte Carlo neutron particle transport methods have been applied to successfully model in 3-D, the HEU-fueled Ghana Research Reactor-1 (GHARR-1), a commercial version of the Miniature Neutron Source Reactor (MNSR) using the MCNP version 4c3 particle transport code. The preliminary multigroup neutronic criticality calculations yielded a keff is contained in 1.00449 with a corresponding cold clean excess reactivity of 4.47mk (447pcm) compared with experimental values of keff is contained in 1.00402 and excess reactivity of 4.00mk (400pcm). The Monte Carlo simulations also show comparable results in the neutron fluxes in the HEU core and some regions of interest. The observed trends in the radial and axial flux distributions in the core, beryllium annular reflector and the water region in the top shim reflector tray were reproduced, indicating consistency of the results, accuracy of the model, precision of the MCNP transport code and the comparability of the Monte Carlo simulations. The results further illustrate the close agreement between stochastic transport theory and the experimental measurements conducted during off-site zero power cold tests. (author)
Remarkable moments in the history of neutron transport Monte Carlo methods
International Nuclear Information System (INIS)
I highlight a few results from the past of the neutron and photon transport Monte Carlo methods which have caused me a great pleasure for their ingenuity and wittiness and which certainly merit to be remembered even when tricky methods are not needed anymore. (orig.)
Fast neutron radiotherapy for advanced malignant tumors of the paranasal sinuses
International Nuclear Information System (INIS)
The radiotherapy of advanced and low grade tumors of the internal nose and paranasal sinuses is still a problem. The results of neutron therapy of 20 patients with advanced and low grade tumors are presented. These patients were treated between July 1977 and October 1988. The mean age of patients is 53 years. Ten patients had adenoid cystic carcinomas, three patients squamous cell carcinomas, one patient a mucinous cell carcinoma, two patients a low grade adenocarcinoma and four patients low grade sarcomas. Mean follow-up is 16 months. Local control could be achieved in eleven cases. The results show that adenoid cystic carcinomas, low grade adenocarcinomas and sarcomas benefit from neutron therapy. (orig.)
Thirty meters small angle neutron scattering instrument at China advanced research reactor
International Nuclear Information System (INIS)
A high resolution 30 m small angle neutron scattering (SANS) instrument has been constructed by the Institute of Chemistry, Chinese Academy of Sciences (ICCAS), and installed at China Advanced Research Reactor (CARR). It is equipped with a mechanical velocity selector, pinhole (including multi-pinhole) collimation system, sample chamber, and high resolution two dimensional 3He position sensitive neutron detector. The flexible variations of incident neutron wavelength, source to sample distance, sample to detector distance and the presence of neutron focusing lenses enable a wide Q range from 0.001 Å−1 to 0.5 Å−1 in reciprocal space and to optimize the resolution required. The instrument is the first SANS instrument in China, and can be widely used for the structure characterization of various materials, as well as kinetic and dynamic observation during external stimulation. The design and characteristics of the instrument are presented in the manuscript. -- Highlights: • The first small angle neutron scattering instrument in China is developed. • It is equipped with a pinhole, multi-pinhole and focusing lenses. • It provides flexible variations of source-sample and sample-detector distances. • A wide Q ranges from 0.001 Å−1 to 0.5 Å−1 in reciprocal space
Dumazert, Jonathan; Coulon, Romain; Carrel, Frédérick; Corre, Gwenolé; Normand, Stéphane; Méchin, Laurence; Hamel, Matthieu
2016-08-01
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.
Summary of dynamic analyses of the advanced neutron source reactor inner control rods
International Nuclear Information System (INIS)
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
J-PARC designated as the 'specific advanced neutron beam facility'
International Nuclear Information System (INIS)
Two years ago, the Japanese Government applied in 'Law for the Promotion of Public Utilization of the Specific Advanced Large Research Facilities' to J-PARC for its secure operation regardless of owner's financial conditions. Under this Law, the Comprehensive Research Organization for Science and Society (CROSS) has been awarded the 'Registered Institution for Facilities Use Promotion' to conduct users program for the public beam lines at the Japanese Spallation Neutron Source. (author)
General Design for CARR Neutron Guide System
Institute of Scientific and Technical Information of China (English)
无
2011-01-01
A neutron guide system has been designed and partly installed at the China Advanced Research Reactor (CARR) to transport cold neutrons from the cold neutron source (CNS) to several instruments,which are situated in a separate guide hall of 30 m×60 m.
Development and benchmarking of higher energy neutron transport data libraries
International Nuclear Information System (INIS)
Neutron cross-section evaluations covering the energy range from 10/sup /minus/11/ to 100 MeV have been prepared for several materials. The principal method used to generate this data base has employed statistical-preequilibrium nuclear models, sophisticated phase shift analyses, and R-matrix techniques. The library takes advantage of formats developed for Version 6 of the Evaluated Nuclear Data File, ENDF. Methods to efficiently utilize the ENDF/B-VI representation of this library in the MCNP Monte Carlo code have been developed. MCNP results using the new library have been compared with calculated results using codes or data based upon intranuclear cascade models. 7 refs., 8 figs
International Nuclear Information System (INIS)
A design-oriented three dimensional stochastic neutron transport code was made with a new lattice model for shielding analysis. The basic assumption of this lattice model is that neutron motion may be sampled at predetermined points. A medium is considered to be filled with a cubic lattice. The number of allowed directions of motion is revised from 26 in the old lattice model to 98 in the new one. By using the lattice model, a computer code named DIMOS has been developed on the basis of a stochastic approach. In addition, this code has an option coupling the two-dimensional discrete ordinates code PALLAS with the DIMOS code. In order to demonstrate the ability of this code, three neutron streaming problems were calculated with the option of coupling in the DIMOS: a cylindrical air duct in water, a straight annular duct in an unsymmetrical configuration and an annular duct with the one bend. Results obtained are in good agreement experimental ones. (author)
Numerical solution of neutron transport equations in discrete ordinates and slab geometry
International Nuclear Information System (INIS)
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
Post-merger evolution of a neutron star-black hole binary with neutrino transport
Foucart, Francois; Roberts, Luke; Duez, Matthew D; Haas, Roland; Kidder, Lawrence E; Ott, Christian D; Pfeiffer, Harald P; Scheel, Mark A; Szilagyi, Bela
2015-01-01
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 a 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 1.4 solar mass with a black hole of 7 solar mass and dimensionless spin a/M=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 the remnant is less neutron-rich. The spatial distribution of the neutrinos is significantly affected by relativistic effects. Over the short timescale e...
International Nuclear Information System (INIS)
Highlights: • Powerful hp-SEM refinement approach for PN 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
Development of a CAD-based neutron transport code with the method of characteristics
International Nuclear Information System (INIS)
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)
International Nuclear Information System (INIS)
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
Radiative or neutron transport modeling using a lattice Boltzmann equation framework
Bindra, H.; Patil, D. V.
2012-07-01
In this paper, the lattice Boltzmann equation (LBE)-based framework is used to obtain the solution for the linear radiative or neutron transport equation. The LBE framework is devised for the integrodifferential forms of these equations which arise due to the inclusion of the scattering terms. The interparticle collisions are neglected, hence omitting the nonlinear collision term. Furthermore, typical representative examples for one-dimensional or two-dimensional geometries and inclusion or exclusion of the scattering term (isotropic and anisotropic) in the Boltzmann transport equation are illustrated to prove the validity of the method. It has been shown that the solution from the LBE methodology is equivalent to the well-known Pn and Sn methods. This suggests that the LBE can potentially provide a more convenient and easy approach to solve the physical problems of neutron and radiation transport.
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.
Transport simulation and image reconstruction for fast-neutron detection of explosives and narcotics
Energy Technology Data Exchange (ETDEWEB)
Micklich, B.J.; Fink, C.L.; Sagalovsky, L.
1995-07-01
Fast-neutron inspection techniques show considerable promise for explosive and narcotics detection. A key advantage of using fast neutrons is their sensitivity to low-Z elements (carbon, nitrogen, and oxygen), which are the primary constituents of these materials. We are currently investigating two interrogation methods in detail: Fast-Neutron Transmission Spectroscopy (FNTS) and Pulsed Fast-Neutron Analysis (PFNA). FNTS is being studied for explosives and narcotics detection in luggage and small containers for which the transmission ratio is greater than about 0.01. The Monte-Carlo radiation transport code MCNP is being used to simulate neutron transmission through a series of phantoms for a few (3-5) projection angles and modest (2 cm) resolution. Areal densities along projection rays are unfolded from the transmission data. Elemental abundances are obtained for individual voxels by tomographic reconstruction, and these reconstructed elemental images are combined to provide indications of the presence or absence of explosives or narcotics. PFNA techniques are being investigated for detection of narcotics in cargo containers because of the good penetration of the fast neutrons and the low attenuation of the resulting high-energy gamma-ray signatures. Analytic models and Monte-Carlo simulations are being used to explore the range of capabilities of PFNA techniques and to provide insight into systems engineering issues. Results of studies from both FNTS and PFNA techniques are presented.
International Nuclear Information System (INIS)
In today's society acts of terrorism must involve in some stages the illicit trafficking either of explosives, chemical agents and/or nuclear materials. Therefore society must rely on an anti-trafficking infrastructure which encompasses responsible authorities, field personnel and adequate instrumental networks. Modern inspection systems for personnel, parcel, vehicle and cargo, as noninvasive imaging techniques, are based on the use of nuclear analytical methods. The inspection systems make use of penetrating radiation (neutrons, gamma and x-rays) in a scanning geometry, with the detection of radiation either transmitted or produced in the interrogated object. Explosives and chemical agent detection systems are based on the fact that the problem of identification can be reduced to the measurement of elemental concentrations. Different nuclear analytical techniques could be used for this purpose; however the use of neutrons has some specific advantages due to the high penetrability in large payloads. Of special interest is the design and use of a transportable neutron system coupled to a gamma-ray radiographic device for inspecting large containers searching for contraband, explosives, weapons etc. The use of neutron induced reactions for non-destructive bulk elemental analysis is well documented. All neutrons, in particular fast neutrons, are well suited to explore large volume samples because of their high penetration in bulk material. Fast neutrons can be produced efficiently and economically by natural radioactive sources, small accelerators or compact electronic neutron generators, making possible the use of neutron based techniques in field applications. Gamma-rays produced by irradiating the sample with neutrons gives the elemental composition of the material, moreover, knowing the nuclear cross-sections and estimating the absorption factors in the different materials, it is possible to perform a quantitative analysis of elements in the sample even in depth
The spectral element method for static neutron transport in AN approximation. Part I
International Nuclear Information System (INIS)
Highlights: ► Spectral elements methods (SEMs) are extended for the neutronics of nuclear reactor cores. ► The second-order, AN 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 AN (i.e. SP2N−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
Development of a 1D neutron transport code employing the method of characteristics
International Nuclear Information System (INIS)
To investigate the 2D/1D fusion core analysis method, a 1D neutron transport problem solver, PEACH-ID, is developed. It is a code of method of characteristics (MOC), both the usual fiat-source step characteristics (SC) scheme and linear source (LS) approximation scheme are adopted for tracking calculation along the neutron flying trajectory. Exponential function interpolation table and fission source extrapolation are adopted as two major methods to accelerate the computational process. Numerical results demonstrate that PEACH-1D is accurate and efficient, and the proposed LS scheme is able to handle quite larger mesh division and deserves much more application in the MOC codes. (authors)
Interface software package for generating the source for neutron transport discrete ordinates code
International Nuclear Information System (INIS)
The safe operation of the reactors imposes heightened requirements toward the quality of the calculated values of the irradiation for the life-time limit assessment of the reactor vessel. The organisation of the calculations has to assure maximum authenticity of the input data, possibility for control and revision of the initial conditions. That's why the whole calculating process has to be computerised. This work presents the software package by means of which the distribution of the primary neutron source in the reactor core, calculated with the,help of diffusion codes is transformed to the source suitable for the codes calculating the neutron transport out of the core by discrete ordinates method
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 concept proof-of-principle demonstration: Switchable radioactive neutron source
International Nuclear Information System (INIS)
An advanced concept proof-of-principle demonstration was successfully performed to show the feasibility of a practical switchable radioactive neutron source (SRNS) that can be switched on and off like an accelerator, but without requiring accelerator equipment such as high voltage supply, control unit, etc. This source concept would provide a highly portable neutron source for field radiation measurement applications. Such a source would require minimal, if any, shielding when not in use. The SRNS, previously patented by Argonne staff, provides a means of constructing the alpha-emitting and light-element components of a radioactive neutron source, in such a fashion that these two components can brought together to turn the source ''on'' and then be separated to turn the source ''off''. An SRNS could be used for such field applications as active neutron interrogation of objects to detect fissile materials or to measure their concentration; and to excite gamma-ray emission for detection of specific elements that indicate toxic chemicals, drugs, explosives, etc. The demonstration was performed using Pu-238 as the alpha emitter and Be as the light element, in an air-atmosphere glovebox having no atmosphere purification capability. A stable, thin film of Pu-238 oxide was deposited on a stainless steel planchet. The ''on'' output of the demonstration Pu-238 film was measured to be 2.5 x 106 neutrons/sec-gram of Pu-238. The measured ''off'' neutron rate was satisfactory, only about 5% of the ''on'' output, after two weeks of exposure to the glovebox atmosphere. After several weeks additional exposure, the ''off'' rate had increased to about 15%. This work demonstrates the feasibility of constructing practical, highly portable SRNS units with very low gamma-ray dose in the ''off'' position
Comparison of neutronic transport equation resolution nodal methods
International Nuclear Information System (INIS)
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)
An advanced control system for a next generation transport aircraft
Rising, J. J.; Davis, W. J; Grantham, W. D.
1983-01-01
The use of modern control theory to develop a high-authority stability and control system for the next generation transport aircraft is described with examples taken from work performed on an advanced pitch active control system (PACS). The PACS was configured to have short-period and phugoid modes frequency and damping characteristics within the shaded S-plane areas, column force gradients with set bounds and with constant slope, and a blended normal-acceleration/pitch rate time history response to a step command. Details of the control law, feedback loop, and modal control syntheses are explored, as are compensation for the feedback gain, the deletion of the velocity signal, and the feed-forward compensation. Scheduling of the primary and secondary gains are discussed, together with control law mechanization, flying qualities analyses, and application on the L-1011 aircraft.
Advanced Finite Element Discretizations for High-Energy Ion Transport
International Nuclear Information System (INIS)
The dominant continuous slowing-down energy loss process coupled with the small (but nonnegligible) straggling poses a significant challenge for deterministic numerical solution when incident beams are monoenergetic or have discontinuous energy spectra. Such spectra broaden very slowly with depth into the target material. Advanced space-energy discretization methods are consequently necessary to achieve numerical robustness. Finite element solutions to this problem were investigated using two general families of discontinuous trial functions, one linear and the other nonlinear. The two families were numerically tested, and results are shown for 1.7-GeV protons incident on a W target. Results from quadratic and exponential-quadratic discontinuous trial functions are in excellent agreement with Monte Carlo results. It is found that very high order finite element schemes are necessary for monoenergetic charged-particle beam transport
Methodology for coupling computational fluid dynamics and integral transport neutronics
International Nuclear Information System (INIS)
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)
Engine Concept Study for an Advanced Single-Aisle Transport
Guynn, Mark D.; Berton, Jeffrey J.; Fisher, Kenneth L.; Haller, William J.; Tong, Michael; Thurman, Douglas R.
2009-01-01
The desire for higher engine efficiency has resulted in the evolution of aircraft gas turbine engines from turbojets, to low bypass ratio, first generation turbofans, to today's high bypass ratio turbofans. Although increased bypass ratio has clear benefits in terms of propulsion system metrics such as specific fuel consumption, these benefits may not translate into aircraft system level benefits due to integration penalties. In this study, the design trade space for advanced turbofan engines applied to a single aisle transport (737/A320 class aircraft) is explored. The benefits of increased bypass ratio and associated enabling technologies such as geared fan drive are found to depend on the primary metrics of interest. For example, bypass ratios at which mission fuel consumption is minimized may not require geared fan technology. However, geared fan drive does enable higher bypass ratio designs which result in lower noise. The results of this study indicate the potential for the advanced aircraft to realize substantial improvements in fuel efficiency, emissions, and noise compared to the current vehicles in this size class.
Assessment of the performance of the spectral element method applied to neutron transport problems
International Nuclear Information System (INIS)
Highlights: • The spectral element method (SEM) is applied to various transport models. • The results allow to assess the performance of SEM when applied to neutron transport problems in reactor physics. • The method is validated against benchmark results and manufactured solutions. • The results presented prove the effectiveness of the method and the high level of accuracy that can be attained. - Abstract: The spectral element method can be used to deal with the spatial operators of neutron transport problems with high efficiency, as shown recently in the framework of the second-order AN transport approximation. The results highlight interesting computational features and show the appeal of the scheme for reactor physics applications. In this paper we investigate the numerical performance of the method in detail. In order to carry out an accurate monitoring of the error behavior to levels close to numerical round-off, we use benchmark problems with known analytical solutions, or with manufactured solutions. Manufactured solutions can easily be obtained for source-injected problems, by tailoring the external neutron source and the boundary conditions to a pre-established analytical solution for a given system. The results presented prove the effectiveness of the method and the high level of accuracy that can be attained
A time-dependent neutron transport model and its coupling to thermal-hydraulics
International Nuclear Information System (INIS)
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)
Numerical solution of the neutron transport equation using cellular neural networks
International Nuclear Information System (INIS)
Various methods have been used for solving the neutron transport equation in the past, and a number of computer codes have been developed based on these solution methods. This paper describes a novel method for the solution of the steady-state and time-dependent neutron transport equation using the duality between neutronic parameters in the method of characteristic (MOC) and the electrical parameters in the cellular neural networks (CNN). The relevant electrical circuit can be simulated by professional electrical circuit simulator software, HSPICE. This software is used for numerical solution of the transport equation only by preparation of appropriate inputs. This method does not need inner and outer iterations, which is a necessary step in the other deterministic methods. One of the main applications of the proposed method may be the development of a new hardware by VLSI technology for online spatio-temporal calculations of the transport equation for nuclear reactor core. The accuracy and capability of this method are examined in a 2D steady-state problem for a BWR fuel assembly, and a 2D time-dependent TWIGL seed/blanket problem
Neutron measurements for advanced nuclear systems: The n{sub T}OF project at CERN
Energy Technology Data Exchange (ETDEWEB)
Colonna, N., E-mail: nicola.colonna@ba.infn.it [Istituto Nazionale di Fisica Nucleare, Bari (Italy); Andriamonje, S. [European Organization for Nuclear Research (CERN), Geneva (Switzerland); Andrzejewski, J. [Uniwersytet Lodzki, Lodz (Poland); Audouin, L. [Centre National de la Recherche Scientifique/IN2P3, IPN, Orsay (France); Barbagallo, M. [Istituto Nazionale di Fisica Nucleare, Bari (Italy); Becares, V. [Centro de Investigaciones Energeticas Medioambientales y Technologicas (CIEMAT), Madrid (Spain); Becvar, F. [Charles University, Prague (Czech Republic); Belloni, F. [Commissariat a l' Energie Atomique (CEA) Saclay - Irfu, Gif-sur-Yvette (France); Berthier, B. [Centre National de la Recherche Scientifique/IN2P3, IPN, Orsay (France); Berthoumieux, E. [Commissariat a l' Energie Atomique (CEA) Saclay - Irfu, Gif-sur-Yvette (France); Brugger, M.; Calviani, M. [European Organization for Nuclear Research (CERN), Geneva (Switzerland); Calvino, F. [Universitat Politecnica de Catalunya, Barcelona (Spain); Cano-Ott, D. [Centro de Investigaciones Energeticas Medioambientales y Technologicas (CIEMAT), Madrid (Spain); Carrapico, C. [Instituto Tecnologico e Nuclear (ITN), Lisbon (Portugal); Cennini, P.; Cerutti, F.; Chiaveri, E.; Chin, M. [European Organization for Nuclear Research (CERN), Geneva (Switzerland); Cortes, G. [Universitat Politecnica de Catalunya, Barcelona (Spain); and others
2011-12-15
A few years ago, the neutron time-of-flight facility n{sub T}OF was built at CERN to address some of the urgent needs of high-accuracy nuclear data for Accelerator Driven Systems and other advanced nuclear energy systems, as well as for nuclear astrophysics and fundamental nuclear physics. Thanks to the characteristics of the neutron beam, and to state-of-the-art detection and acquisition systems, high quality neutron cross-section data have been obtained for a variety of isotopes, many of which radioactive. Following an important upgrade of the spallation target and of the experimental area, a new measurement campaign has started last year. After a brief review of the most important results obtained so far at n{sub T}OF, the new features of the facility are presented, together with the first results on the commissioning of the neutron beam. The plans for future measurements, in particular related to nuclear technology are finally discussed.
Advanced fuel cells for transportation applications. Final report
Energy Technology Data Exchange (ETDEWEB)
NONE
1998-02-10
This Research and Development (R and D) contract was directed at developing an advanced technology compressor/expander for supplying compressed air to Proton Exchange Membrane (PEM) fuel cells in transportation applications. The objective of this project was to develop a low-cost high-efficiency long-life lubrication-free integrated compressor/expander utilizing scroll technology. The goal of this compressor/expander was to be capable of providing compressed air over the flow and pressure ranges required for the operation of 50 kW PEM fuel cells in transportation applications. The desired ranges of flow, pressure, and other performance parameters were outlined in a set of guidelines provided by DOE. The project consisted of the design, fabrication, and test of a prototype compressor/expander module. The scroll CEM development program summarized in this report has been very successful, demonstrating that scroll technology is a leading candidate for automotive fuel cell compressor/expanders. The objectives of the program are: develop an integrated scroll CEM; demonstrate efficiency and capacity goals; demonstrate manufacturability and cost goals; and evaluate operating envelope. In summary, while the scroll CEM program did not demonstrate a level of performance as high as the DOE guidelines in all cases, it did meet the overriding objectives of the program. A fully-integrated, low-cost CEM was developed that demonstrated high efficiency and reliable operation throughout the test program. 26 figs., 13 tabs.
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
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R"udiger, A; Ruggi, P; Ryan, K; Sachdev, 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; Schofield, R M S; Sch"onbeck, A; Schreiber, E; Schuette, D; Schutz, B F; Scott, J; Scott, S M; Sellers, D; Sengupta, A S; Sentenac, D; Sequino, V; Sergeev, A; Setyawati, Y; Shaddock, D A; Shaffer, T; Shahriar, M S; Shaltev, M; Shapiro, B; Shawhan, P; Sheperd, A; Shoemaker, D H; Shoemaker, D M; Siellez, K; Siemens, X; Sieniawska, M; Sigg, D; Silva, A D; Singer, A; Singer, L P; Singh, A; Singh, R; Singhal, A; Sintes, A M; Slagmolen, B J J; Smith, J R; 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, R; Strain, K A; Straniero, N; Stratta, G; Strauss, N A; Strigin, S; Sturani, R; Stuver, A L; Summerscales, T Z; Sun, L; Sunil, S; Sutton, P J; Swinkels, B L; Szczepa'nczyk, M J; Tacca, M; Talukder, D; Tanner, D B; T'apai, M; Tarabrin, S P; Taracchini, A; Taylor, R; Theeg, T; Thirugnanasambandam, M P; Thomas, E G; Thomas, M; Thomas, P; Thorne, K A; Thrane, E; Tiwari, S; Tiwari, V; Tokmakov, K V; Toland, K; Tomlinson, C; Tonelli, M; Tornasi, Z; Torres, C V; Torrie, C I; T"oyr"a, D; Travasso, F; Traylor, G; Trifir`o, 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; Brand, J F J van den; Broeck, C Van Den; Vander-Hyde, D C; van der Schaaf, L; van Heijningen, J V; van Veggel, A A; Vardaro, M; Vass, S; Vas'uth, M; Vaulin, R; Vecchio, A; Vedovato, G; Veitch, J; Veitch, P J; Venkateswara, K; Verkindt, D; Vetrano, F; Vicer'e, 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, M; 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; Wessels, P; Westphal, T; Wette, K; Whelan, J T; Whiting, B F; Williams, R D; 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; Yablon, J; Yam, W; Yamamoto, H; Yancey, C C; Yu, H; Yvert, M; zny, A Zadro; 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 Advanced LIGO. In particular we searched for gravitational wave signals from binary neutron star systems with component masses $\\in [1,3] M_{\\odot}$ and component dimensionless spins $< 0.05$. We also searched for neutron-star--black-hole systems with the same neutron star parameters, black hole mass $\\in [2,99] M_{\\odot}$ and no restriction on the black hole spin magnitude. We assess the sensitivity of the two LIGO detectors to these systems, and find that they could have detected the merger of binary neutron star systems with component mass distributions of $1.35\\pm0.13 M_{\\odot}$ at a volume-weighted average distance of $\\sim$ 70Mpc, and for neutron-star--black-hole systems with neutron star masses of $1.4M_\\odot$ and black hole masses of at least $5M_\\odot$, a volume-weighted average distance of at least $\\sim$ 110Mpc. From this we...
Neutron transport through semi-infinite continuous stochastic media using Gaussian statistics
International Nuclear Information System (INIS)
The stationary solution of the one-speed neutron transport equation in a semi-infinite stochastic medium with linear anisotropic scattering is considered. The cross-section function of the medium is assumed to be a continuous random function of position with fluctuations about the mean taken as Gaussian distributed. The joint probability distribution function of these Gaussian random variables is used to calculate the ensemble-averaged quantities, such as radiant neutron energy and net neutron flux, for an arbitrary correlation function. The problem is solved at first in the deterministic case, then the solution is averaged using Gaussian joint probability distribution function. A modified Gaussian probability distribution function is also used to average the solution. Numerical results are given for the sake of comparison
Benchmark calculations of neutron dose rates at transport and storage casks
International Nuclear Information System (INIS)
The application of numerical calculations methods for demonstration of sufficient radiation shielding of radioactive waste transport and storage casks requires a validation based on appropriate measurements of gamma and neutron sources. The results of the comparison of measured data and calculations using the Monte Carlo program MCNP show deviations dependent on the loading of the cask within the standard deviation which is dominated by the measuring method. Considering the neutrons scattered at the salt MCNP (in case of disposal in the salt) tends to underestimate the nominal values, but still within the double standard deviation. This accuracy is not reached with MAVRIC. Based on AHE (active handling experiments) data benchmark calculations were performed that can be used as reference value. The total accuracy results from the accuracy of the source term and the measurement of the neutron dose rate with a deviation of 15%.
International Nuclear Information System (INIS)
A Monte Carlo tool RSMC (Reaction Sequence Monte Carlo) was developed to simulate deuteron/triton transportation and reaction coupled problem. The 'Forced particle production' variance reduction technique was used to improve the simulation speed, which made the secondary product play a major role. The mono-energy 14 MeV fusion neutron source was employed as a validation. Then the thermal-to-fusion neutron convertor was studied with our tool. Moreover, an in-core conversion efficiency measurement experiment was performed with 6LiD and 6LiH converters. Threshold activation foils was used to indicate the fast and fusion neutron flux. Besides, two other pivotal parameters were calculated theoretically. Finally, the conversion efficiency of 6LiD is obtained as 1.97x10-4, which matches well with the theoretical result. (authors)
International Nuclear Information System (INIS)
The neutron transport in a solid cylinder that contains an inhomogeneous medium with anisotropic scattering is considered. The medium has diffuse reflecting boundary with an external incidence and contains an internal neutron source. This general problem can be solved in terms of the solution of the corresponding source-free problem with transparent boundary and isotropic external incidence. The source-free problem is solved in its integral form using the variationai method. Trial functions of the solution are assumed in terms of the asymptotic solution of the Pomraning-Eddington approximation of the source-free problem. The variational technique is used to determine the constants of the trial functions. The partial flux at the boundary, the neutron density and the net flux of the general problem are calculated
Early Advanced LIGO binary neutron-star sky localization and parameter estimation
Berry, C P L; Farr, W M; Haster, C-J; Mandel, I; Middleton, H; Singer, L P; Urban, A L; Vecchio, A; Vitale, S; Cannon, K; Graff, P B; Hanna, C; Mohapatra, S; Pankow, C; Price, L R; Sidery, T; Veitch, J
2016-01-01
2015 will see the first observations of Advanced LIGO and the start of the gravitational-wave (GW) advanced-detector era. One of the most promising sources for ground-based GW detectors are binary neutron-star (BNS) coalescences. In order to use any detections for astrophysics, we must understand the capabilities of our parameter-estimation analysis. By simulating the GWs from an astrophysically motivated population of BNSs, we examine the accuracy of parameter inferences in the early advanced-detector era. We find that sky location, which is important for electromagnetic follow-up, can be determined rapidly (~5 s), but that sky areas may be hundreds of square degrees. The degeneracy between component mass and spin means there is significant uncertainty for measurements of the individual masses and spins; however, the chirp mass is well measured (typically better than 0.1%).
NOTRAN/3D, 3-D Neutron Transport in X-Y-Z Geometry by Discrete Nodal Transport Method
International Nuclear Information System (INIS)
1 - Description of program or function: NOTRAN/3D solves the neutron transport equation in three-dimensional XYZ geometry by the discrete nodal transport method. Source and eigenvalue problems can be solved. The input format for cross sections is the same as for ANISN. Multigroup cross section libraries such as DLC-37 and DLC75/BUGLE-80 can be used. 2 - Method of solution: NOTRAN/3D uses the discrete nodal transport method. Anisotropic scattering is treated using Legendre expansion. The order of interior flux approximation is two. Plane or linear leakage approximation of surface flux is used. 3 - Restrictions on the complexity of the problem: Maximum order of: anisotropic scattering = 3; material compositions = 20; energy groups = 2; angular quadrature = 8; zones = 20. When coarse-mesh re-balancing is used, the maximum number of course meshes is 5 in each direction. If computer memory permits, some arrays can be enlarged to reduce the above restrictions
A new paradigm for whole core neutron transport without homogenization
International Nuclear Information System (INIS)
A new paradigm is introduced which allows the performance of whole core transport calculations without lattice homogenization. Quasi-reflected interface conditions are formulated to partially decouple periodic lattice effects from the pin-cell to pin-cell flux variation in the finite sub-element form of the variational nodal code VARIANT. With fuel-coolant homogenization eliminated, the interface variables that couple pin-cell sized nodes are divided into low-order and high-order spherical harmonic terms. Reflected interface conditions are subsequently applied to the high-order terms to remove them from the system of unknowns. Combined with an integral transport treatment within the node, the new approach dramatically reduces both the formation time and the size of the response matrices and leads to sharply reduced memory and CPU requirements. The method is applied to the two-dimensional C5-G7 problem, an OECD/NEA PWR benchmark containing MOX and UO2 fuel assemblies. Results indicate the new approach results in very little loss of accuracy relative to the corresponding full spherical harmonics expansions while reducing CPU times by well over an order of magnitude. (authors)
Neutron transport by collision probability method in complicated geometries
Energy Technology Data Exchange (ETDEWEB)
Constantin, Marin [Institute for Nuclear Research, Pitesti (Romania)
2000-09-01
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{sub 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{sub 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)
Energy Technology Data Exchange (ETDEWEB)
Morel, J.E.
1981-01-01
A collocation method is developed for the solution of the one-dimensional neutron transport equation in slab geometry with both symmetric and polarly asymmetric scattering. For the symmetric scattering case, it is found that the collocation method offers a combination of some of the best characteristics of the finite-element and discrete-ordinates methods. For the asymmetric scattering case, it is found that the computational cost of cross-section data processing under the collocation approach can be significantly less than that associated with the discrete-ordinates approach. A general diffusion equation treating both symmetric and asymmetric scattering is developed and used in a synthetic acceleration algorithm to accelerate the iterative convergence of collocation solutions. It is shown that a certain type of asymmetric scattering can radically alter the asymptotic behavior of the transport solution and is mathematically equivalent within the diffusion approximation to particle transport under the influence of an electric field. The method is easily extended to other geometries and higher dimensions. Applications exist in the areas of neutron transport with highly anisotropic scattering (such as that associated with hydrogenous media), charged-particle transport, and particle transport in controlled-fusion plasmas. 23 figures, 6 tables.
International Nuclear Information System (INIS)
1 - Nature of physical problem solved: TDA (Time-Dependent ANISN) solves the one-dimensional time- dependent Boltzmann transport equation for neutrons and/or gamma- rays in slab, sphere or cylindrical geometries. Delayed neutron and other time-dependent effects are not considered in the present version. A choice of two types of sources and one initial condition specification is given (A. Space and energy distributed source with a step function time distribution. B. Analytical first collision source). 2 - Method of solution: TDA is based on the steady-state SN code ANISN for reasons of stability and generality. The weighted difference equations are used. 3 - Restrictions on the complexity of the problem: Limited only by available main storage
Boron neutron capture therapy for advanced and/or recurrent cancers in the oral cavity
International Nuclear Information System (INIS)
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 10B-carrier p-boronophenylalanine (BPA) with or without borocaptate sodium (BSH) and irradiation thereafter with epithermal neutrons. All underwent 18F-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)
Anisotropic kernel p(μ → μ') for transport calculations of elastically scattered neutrons
International Nuclear Information System (INIS)
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
High performance parallel Monte Carlo transport computations for ITER fusion neutronics applications
International Nuclear Information System (INIS)
Large scale neutronics calculations for radiation safety and machine reliability are required to support design activities for the ITER fusion reactor which is currently in phase of construction. Its large size and complexity of diagnostics, control and heating systems and ports, and also channel penetrations inside the thick blanket shielding surrounding the 14 MeV D-T neutron source are essential challenges for neutronics calculations. In the ITER tokamak geometry, the Monte Carlo (MC) method is the preferred one for radiation transport calculations. This method allows describing neutrons interactions with matter by tracking individual particle histories. The precision of the MC method depends on number of sampled particles according to statistical laws and on systematic uncertainties introduced by modeling assumptions. Due to the independence of particle histories, their tracks can be processed in parallel. Parallel computations on high performance cluster computers substantially increase number of sampled particles and therefore allow reaching the desired statistical precision of the MC results. Use of CAD-based approach with high spatial resolution improves systematic adequacy of the MC geometry modeling. These achievements are demonstrated on radiation transport calculations for designing the Blanket Shield Module and Auxiliary Shield of the ITER Electron Cyclotron Heating (ECH) upper launcher. (author)
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.
Neutron transport solver parallelization using a Domain Decomposition method
International Nuclear Information System (INIS)
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 (PN 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)
Deterministic methods to solve the differential transport equation in neutronic
International Nuclear Information System (INIS)
We present a synthesis of the methods used to solve the integro-differential form of the transport equation. This form is used above all to achieve whole core calculations in 2 and 3D. First, we discretize the equation in energy and it leads us to an one group energy equation. For each of these groups, the scope of the calculation is so big that we have to calculate our solution on homogenized cells. On this homogenized medium, we describe different angular and spatial discretizations with acceleration methods. Finally we select some promising schemes to test: - SN Linear Nodal method with a Diffusion Synthetic Acceleration method; - Variational Nodal Method. These methods could be compared with more classical ones. To say, finite element or finite difference methods. (author). 65 refs., 3 annexes
Neutron- and proton-induced evaluated transport library up to 150 MeV
International Nuclear Information System (INIS)
A new evaluated nuclear data library has been created. The library consists of two sub-libraries for neutron and proton incident particles. The neutron sub-library contains nuclear data for transport, heating and shielding applications for 242 nuclides with atomic numbers ranging from 8 to 82 in the energy region of primary neutrons from 10-5 eV to 150 MeV. Data below 20 MeV are taken mainly from ENDF/B-VI (revision 8) and for some nuclides, from the JENDL-3.3 and JEFF-3.0 libraries. The proton sub-library should contain data for the same range of target nuclides and energies. Proton-induced evaluated cross-section files are available for 15 nuclides at the moment. The evaluation of emitted particle energy and angular distributions at energies above 20 MeV (for incident neutrons) and above the reaction threshold (for incident protons) was performed with the help of the ALICE/ASH code and the analysis of available experimental data. The total cross-sections, elastic cross-sections and elastic scattering angular distributions were calculated with the help of the coupled channel model. The results of the calculation were adjusted to the data from ENDF/B-VI, JENDL-3.3, or JEFF-3.0 at the neutron energy equal to 20 MeV. The library is written in ENDF-6 format using the MF=3/MT=5 and MF=6/MT=5 representations
Transport coefficients of relativistic nuclear and neutron matter with in-medium effects
International Nuclear Information System (INIS)
The transport coefficients (thermal conductivity, shear and bulk viscosities) of symmetric nuclear matter and neutron matter are calculated in the Walecka model with a Boltzmann-Uehling-Uhlenbeck (BUU) collision term by means of a Chapman-Enskog expansion in first order. The order of magnitude of the influence of in-medium effects induced by the presence of the mean σ and ω fields on these coefficients is evaluated. It is found that the transport coefficients can be modified by a factor up to 4 in the range of interest for heavy-ion collisions hydrodynamics or for neutron stars. The results obtained from the BUU collision term are in agreement with those of a previous calculation (R. Hakim and L. Mornas, Phys. Rev. C47 (1993) 2846) in the relaxation-time approximation. ((orig.))
Monte Carlo method for neutron transport calculations in graphics processing units (GPUs)
International Nuclear Information System (INIS)
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)
Monte Carlo Neutrino Transport Through Remnant Disks from Neutron Star Mergers
Richers, S; O'Connor, Evan; Fernandez, Rodrigo; Ott, Christian
2015-01-01
We present Sedonu, a new open source, steady-state, special relativistic Monte Carlo (MC) neutrino transport code, available at bitbucket.org/srichers/sedonu. The code calculates the energy- and angle-dependent neutrino distribution function on fluid backgrounds of any number of spatial dimensions, calculates the rates of change of fluid internal energy and electron fraction, and solves for the equilibrium fluid temperature and electron fraction. We apply this method to snapshots from two dimensional simulations of accretion disks left behind by binary neutron star mergers, varying the input physics and comparing to the results obtained with a leakage scheme for the case of a central black hole and a central hypermassive neutron star. Neutrinos are guided away from the densest regions of the disk and escape preferentially around 45 degrees from the equatorial plane. Neutrino heating is strengthened by MC transport a few scale heights above the disk midplane near the innermost stable circular orbit, potentiall...
International Nuclear Information System (INIS)
The light transport properties of scintillator light inside alternative He-3 neutron detectors using scintillator sheets have been investigated by a ray-tracing simulation code. The detector consists of a light-reflecting tube, a thin rectangular ceramic scintillator sheet laminated on a glass plate, and two photo-multiplier tubes (PMTs) mounted at both ends of the detector tube. The flashes of light induced on the surface of the scintillator sheet via nuclear interaction between the scintillator and neutrons are detected by the two PMTs. The light output at both ends of various detectors in which the scintillator sheets are installed with several different arrangements were examined and evaluated in comparison with experimental results. The results derived from the simulation reveal that the light transport property is strongly dependent on the arrangement of the scintillator sheet inside the tube and the shape of the tube
Energy Technology Data Exchange (ETDEWEB)
Ohzu, A., E-mail: ohzu.akira@jaea.go.jp [Nuclear Science and Engineering Center, Japan Atomic Energy Agency, Tokai, Naka, Ibaraki 319-1195 (Japan); Takase, M.; Haruyama, M.; Kurata, N.; Kobayashi, N.; Kureta, M. [Nuclear Science and Engineering Center, Japan Atomic Energy Agency, Tokai, Naka, Ibaraki 319-1195 (Japan); Nakamura, T.; Toh, K.; Sakasai, K.; Suzuki, H.; Soyama, K. [J-PARC, Japan Atomic Energy Agency, Tokai, Naka, Ibaraki 319-1195 (Japan); Seya, M. [Integrated Support Center for Nuclear Nonproliferation and Nuclear Security, Japan Atomic Energy Agency, Tokai, Naka, Ibaraki 319-1195 (Japan)
2015-10-21
The light transport properties of scintillator light inside alternative He-3 neutron detectors using scintillator sheets have been investigated by a ray-tracing simulation code. The detector consists of a light-reflecting tube, a thin rectangular ceramic scintillator sheet laminated on a glass plate, and two photo-multiplier tubes (PMTs) mounted at both ends of the detector tube. The flashes of light induced on the surface of the scintillator sheet via nuclear interaction between the scintillator and neutrons are detected by the two PMTs. The light output at both ends of various detectors in which the scintillator sheets are installed with several different arrangements were examined and evaluated in comparison with experimental results. The results derived from the simulation reveal that the light transport property is strongly dependent on the arrangement of the scintillator sheet inside the tube and the shape of the tube.
International Nuclear Information System (INIS)
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
A transportable fast neutron and dual gamma-ray system for the detection of illicit materials
International Nuclear Information System (INIS)
A transportable FNGR radiography system has been simulated using the MCNPX Monte Carlo code. The system is envisaged to be applied to the material characterisation of a suspicious bulky object, in view of identifying illegal materials. The system combines a neutron and two gamma-ray sources achieving characterisation of the material of the object through two ratios, namely 137Cs/DD and 60Co/DD. Hence, the system discriminates materials of similar or even the same of either of the two ratios. The proposed unit complies with radiation protection requirements achieving a safe occupational environment. - Highlights: → Transportable radiography system. → Neutron- and dual energy photon-beams available. → Discrimination of materials. → Detection of illicit materials.
A transportable fast neutron and dual gamma-ray system for the detection of illicit materials
Energy Technology Data Exchange (ETDEWEB)
Fantidis, J.G. [Laboratory of Nuclear Technology, School of Engineering, ' Democritus' University of Thrace, Vas. Sofias 12, Xanthi 67100 (Greece); Nicolaou, G.E., E-mail: nicolaou@ee.duth.gr [Laboratory of Nuclear Technology, School of Engineering, ' Democritus' University of Thrace, Vas. Sofias 12, Xanthi 67100 (Greece)
2011-08-21
A transportable FNGR radiography system has been simulated using the MCNPX Monte Carlo code. The system is envisaged to be applied to the material characterisation of a suspicious bulky object, in view of identifying illegal materials. The system combines a neutron and two gamma-ray sources achieving characterisation of the material of the object through two ratios, namely {sup 137}Cs/DD and {sup 60}Co/DD. Hence, the system discriminates materials of similar or even the same of either of the two ratios. The proposed unit complies with radiation protection requirements achieving a safe occupational environment. - Highlights: > Transportable radiography system. > Neutron- and dual energy photon-beams available. > Discrimination of materials. > Detection of illicit materials.
The Saudi experience with neutron therapy in locally advanced head and neck cancers
International Nuclear Information System (INIS)
The neutron therapy program at King Faisal Specialist Hospital and Research Center conducted a phase II study to evaluate the toxicity and efficacy of neutrons against conventional external megavoltage irradiation in patients with locally advanced head and neck malignancy. One hundred and nineteen patients were allocated to receive either photons (46/119) or neutrons (73/119). Radiation effects were scored according to the EORTC/RTOG criteria; data was collected weekly during treatment, once a month for the first year and at 6-month intervals subsequently. While acute effects were scored in all patients, only 59 were evaluable for late effects and locoregional control. A composite of the average reaction results were obtained using this information, to compare them in time, for acute and late effects in both arms of the study. The maximum acute reactions in the two groups were similar. In the majority of the patients (80 %) acute skin and mucosal reactions occurred during the last week of treatment. The changes in the subcutaneous tissues and salivary glands became clinically apparent at 3 months or later. Salivary gland toxicity was more severe in the photon arm and the difference was statistically significant at 3 months (P + 0.04) but this was lost at 12 months. Late effects for skin and subcutaneous tissues were significantly more severe in the neutron arm with P values of 0.04 and 0.01 respectively. Three patients in the neutron arm died of grade 4 radiation complications. The local control and survival were similar in both groups. (author)
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.
Phase 1 environmental report for the Advanced Neutron Source at Oak Ridge National Laboratory
International Nuclear Information System (INIS)
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 252Cf. 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
International Nuclear Information System (INIS)
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 DTx fuel with a small tritium initial content (xe 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 SiC, modify the understanding of the calculation of displacement per atom, dpa, to quantify the collisional damage. An important analysis has been the confirmation, using Molecular Dynamics (MD) with an astonishing agreement, of the experimental evidence of low-temperature amorphization by damage accumulation in SiC, which could modify extensively its viability as a
New contributions to neutron stochastic transport theory in the time and in the frequency domain
International Nuclear Information System (INIS)
The authors generalize the stochastic transport theory of Pal and Munoz-Cobo and Perez methodology, to include the delayed neutron effects. They apply this theory to interpret several experiments measuring the cross power spectral density G12(W), G13(W), G23(W) of three detectors 1, 2 and 3, located in and out of a tank containing a UO2F2 solution in water. (Auth.)
Transient Neutron Transport in Semi-Infinite Media for Pure-Triplet Scattering
International Nuclear Information System (INIS)
The transient neutron transport in a semi-infinite medium with pure-triplet scattering is presented. Case s eigenfunctions for this problem can be obtained for this high order anisotropic scattering and orthogonality relations of these eigenfunctions can be derived mathematically. The reflectivity at the boundary, radiant energy and net heat flux are computed for specular-reflecting boundary with angular-dependent externally-incident flux. For the sake of comparison, we use two different weight functions in our calculations
Wang, Weiwei
2013-01-01
High pressure, high magnetic field and low temperature techniques are required to investigate magnetic transitions and quantum critical behaviour in different ferromagnetic materials to elucidate how novel forms of superconductivity and other new states are brought about. In this project, several instruments for magneto-transport and neutron scattering measurements have been designed and built. They include inserts for a dilution refrigerator and pressure cells for resistivity,...
Neutron, electron and photon transport in ICF tragets in direct and fast ignition
A. Parvazian; A. Okhovat
2005-01-01
Fusion energy due to inertial confinement has progressed in the last few decades. In order to increase energy efficiency in this method various designs have been presented. The standard scheme for direct ignition and fast ignition fuel targets are considered. Neutrons, electrons and photons transport in targets containing different combinations of Li and Be are calculated in both direct and fast ignition schemes. To compress spherical multilayer targets having fuel in the central part, they a...
Normal and adjoint integral and integrodifferential neutron transport equations. Pt. 2
International Nuclear Information System (INIS)
Using the simplifying hypotheses of the integrodifferential Boltzmann equations of neutron transport, given in JEN 334 report, several integral equations, and theirs adjoint ones, are obtained. Relations between the different normal and adjoint eigenfunctions are established and, in particular, proceeding from the integrodifferential Boltzmann equation it's found out the relation between the solutions of the adjoint equation of its integral one, and the solutions of the integral equation of its adjoint one (author)
International Nuclear Information System (INIS)
Typical half-space problems in two-group neutron transport theory are solved numerically using the singular-eigenfunction-expansion technique, considering isotropic-and linearly anisotropic scattering. Numerical results are reported for the Albedo, Milne and Constant-Source problems in a half-space pure light-water medium using isotropic scattering data set of Metacalf and Zweifel and considering various degrees of anisotropy
International Nuclear Information System (INIS)
The first-order neutron transport equation is solved by the least-squares finite element method based on the discrete ordinates discretization. The angular dependent rebalance (ADR) acceleration arithmetic and its extrapolate method are given. The numerical results of some benchmark problems demonstrate that the arithmetic can shorten the CPU time to 34% ∼ 50% and it is effective even for the strong scattering problem. (authors)
Analysis of two different benchmark problems using one-dimensional neutron transport theory code
International Nuclear Information System (INIS)
This paper focuses on the application of method of characteristics (MOC) for the solution of neutron transport equation in one-dimensional geometries. The paper discusses the results obtained for two different benchmark problems. The results compared well with the benchmark results. An interesting result is that, in case of MOC the unphysical flux dip at the centre of sphere (commonly found with SN - method) is absent. (author)
Modular, object-oriented redesign of a large-scale Monte Carlo neutron transport program
International Nuclear Information System (INIS)
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
International Nuclear Information System (INIS)
A new coarse-mesh rebalance method is developed and tested to accelerate one-dimensional discrete ordinates neutron transport equation. The method is based on the use of angular dependent rebalance factors. Unlike the original Coarse-Mesh Rebalance method, Fourier analysis and numerical results show that this Angular Dependent Coarse-Mesh Rebalance(ADCMR) method is unconditionally stable for any optical thickness and that the acceleration effect is significant
Systems guide to MCNP (Monte Carlo Neutron and Photon Transport Code)
International Nuclear Information System (INIS)
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
Energy Technology Data Exchange (ETDEWEB)
Kara, Ayahn [Sinop Univ. (Turkey). Dept. of Nuclear Energy Engineering; Anli, Fikret [Univ. Kahramanmaras (Turkey). Dept of Physics
2015-03-15
PN approximation is known as a proper method to solve neutron transport equation when literature is taken into consideration. Generally, conventional scattering function is used to solve criticality and diffusion problems in Legendre polynomial approximation. In this study, instead of conventional scattering function, Henyey-Greenstein (HG) and Anli-Gungor phase functions (AG) are used in slab geometry transport equation and some critical thicknesses of the slab are calculated as an application with Legendre polynomial (PN) approximation and Marshak boundary condition. Results obtained from HG and AG scattering functions are compared and the correlations and discrepancies between the two functions are presented in the tables.
Hybrid method of deterministic and probabilistic approaches for multigroup neutron transport problem
International Nuclear Information System (INIS)
A hybrid method of deterministic and probabilistic methods is proposed to solve Boltzmann transport equation. The new method uses a deterministic method, Method of Characteristics (MOC), for the fast and thermal neutron energy ranges and a probabilistic method, Monte Carlo (MC), for the intermediate resonance energy range. The hybrid method, in case of continuous energy problem, will be able to take advantage of fast MOC calculation and accurate resonance self shielding treatment of MC method. As a proof of principle, this paper presents the hybrid methodology applied to a multigroup form of Boltzmann transport equation and confirms that the hybrid method can produce consistent results with MC and MOC methods. (authors)
Synergism of the method of characteristics and CAD technology for neutron transport calculation
International Nuclear Information System (INIS)
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)
International Nuclear Information System (INIS)
A new generation transport in TNR is characterized by substantially higher level of requirements on the accuracy of cross-sections (1--3% for total and 10--20% for double differential ones). This is true in the first place for the breeder and coolant materials in the blanket as well as for shielding materials. However, the experience of integral experiments calculation analysis convinces that the increase in accuracy of single microscopic data ensures not always in achievement of sufficient accuracy of single microscopic data ensures not always in achievement of sufficient accuracy in an integral result in the calculation with concrete nuclide data. The natural criterion of the data quality is an integral experiment which enables for the data producer to obtain information on advantages and disadvantages of a file, and for the user -- to determine the level of confidence in his calculation results. This paper provides a short review of experimental and calculated works on testing neutronic data for neutron multiplying materials, such as 238U, Th, Be and Pb
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)
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
Transport Advances in Disposable Bioreactors for Liver Tissue Engineering
Catapano, Gerardo; Patzer, John F.; Gerlach, Jörg Christian
Acute liver failure (ALF) is a devastating diagnosis with an overall survival of approximately 60%. Liver transplantation is the therapy of choice for ALF patients but is limited by the scarce availability of donor organs. The prognosis of ALF patients may improve if essential liver functions are restored during liver failure by means of auxiliary methods because liver tissue has the capability to regenerate and heal. Bioartificial liver (BAL) approaches use liver tissue or cells to provide ALF patients with liver-specific metabolism and synthesis products necessary to relieve some of the symptoms and to promote liver tissue regeneration. The most promising BAL treatments are based on the culture of tissue engineered (TE) liver constructs, with mature liver cells or cells that may differentiate into hepatocytes to perform liver-specific functions, in disposable continuous-flow bioreactors. In fact, adult hepatocytes perform all essential liver functions. Clinical evaluations of the proposed BALs show that they are safe but have not clearly proven the efficacy of treatment as compared to standard supportive treatments. Ambiguous clinical results, the time loss of cellular activity during treatment, and the presence of a necrotic core in the cell compartment of many bioreactors suggest that improvement of transport of nutrients, and metabolic wastes and products to or from the cells in the bioreactor is critical for the development of therapeutically effective BALs. In this chapter, advanced strategies that have been proposed over to improve mass transport in the bioreactors at the core of a BAL for the treatment of ALF patients are reviewed.
Follow-up fuel plate stability experiments and analyses for the Advanced Neutron Source
Energy Technology Data Exchange (ETDEWEB)
Swinson, W.F.; Battiste, R.L.; Luttrell, C.R.; Yahr, G.T.
1993-11-01
The reactor for the planned Advanced Neutron Source uses closely spaced plates cooled by heavy water flowing through narrow channels. Two sets of tests were performed on the upper and lower fuel plates for the structural response of the fuel plates to the required high coolant flow velocities. This report contains the data from the second round of tests. Results and conclusions from all of the tests are also included in this report. The tests were done using light water on full-scale epoxy models, and through model theory, the results were related to the prototype plates, which are aluminum-clad aluminum/uranium silicide involute-shaped plates.
Yong Wang; Wenzheng Yue; Mo Zhang
2016-01-01
The anisotropic transport of thermal neutron in heterogeneous porous media is of great research interests in many fields. In this paper, it is the first time that a new model based on micron X-ray computed tomography (CT) has been proposed to simultaneously consider both the separation of matrix and pore and the distribution of mineral components. We apply the Monte Carlo method to simulate thermal neutrons transporting through the model along different directions, and meanwhile detect those ...
Some results on the neutron transport and the coupling of equations
International Nuclear Information System (INIS)
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)
Advanced transportation system studies. Alternate propulsion subsystem concepts: Propulsion database
Levack, Daniel
1993-01-01
The Advanced Transportation System Studies alternate propulsion subsystem concepts propulsion database interim report is presented. The objective of the database development task is to produce a propulsion database which is easy to use and modify while also being comprehensive in the level of detail available. The database is to be available on the Macintosh computer system. The task is to extend across all three years of the contract. Consequently, a significant fraction of the effort in this first year of the task was devoted to the development of the database structure to ensure a robust base for the following years' efforts. Nonetheless, significant point design propulsion system descriptions and parametric models were also produced. Each of the two propulsion databases, parametric propulsion database and propulsion system database, are described. The descriptions include a user's guide to each code, write-ups for models used, and sample output. The parametric database has models for LOX/H2 and LOX/RP liquid engines, solid rocket boosters using three different propellants, a hybrid rocket booster, and a NERVA derived nuclear thermal rocket engine.
NATO Advanced Study Institute on Chemical Transport in Melasomatic Processes
1987-01-01
As indicated on the title page, this book is an outgrowth of the NATO Advanced Study Institute (ASI) on Chemical Transport in Metasomatic Processes, which was held in Greece, June 3-16, 1985. The ASI consisted of five days of invited lectures, poster sessions, and discussion at the Club Poseidon near Loutraki, Corinthia, followed by a two-day field trip in Corinthia and Attica. The second week of the ASI consisted of an excursion aboard M/S Zeus, M/Y Dimitrios II, and the M/S Irini to four of the Cycladic Islands to visit, study, and sample outstanding exposures of metasomatic activity on Syros, Siphnos, Seriphos, and Naxos. Nine teen invited lectures and 10 session chairmen/discussion leaders participated in the ASI, which was attended by a total of 92 professional scientists and graduate stu dents from 15 countries. Seventeen of the invited lectures and the Field Excursion Guide are included in this volume, together with 10 papers and six abstracts representing contributed poster sessions. Although more...
Advances in explosives analysis--part II: photon and neutron methods.
Brown, Kathryn E; Greenfield, Margo T; McGrane, Shawn D; Moore, David S
2016-01-01
The number and capability of explosives detection and analysis methods have increased dramatically since publication of the Analytical and Bioanalytical Chemistry special issue devoted to Explosives Analysis [Moore DS, Goodpaster JV, Anal Bioanal Chem 395:245-246, 2009]. Here we review and critically evaluate the latest (the past five years) important advances in explosives detection, with details of the improvements over previous methods, and suggest possible avenues towards further advances in, e.g., stand-off distance, detection limit, selectivity, and penetration through camouflage or packaging. The review consists of two parts. Part I discussed methods based on animals, chemicals (including colorimetry, molecularly imprinted polymers, electrochemistry, and immunochemistry), ions (both ion-mobility spectrometry and mass spectrometry), and mechanical devices. This part, Part II, will review methods based on photons, from very energetic photons including X-rays and gamma rays down to the terahertz range, and neutrons. PMID:26446898
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.
Photon neutron mixed-beam radiotherapy of locally advanced prostate cancer
International Nuclear Information System (INIS)
Purpose: In this article we present the results of mixed-beam, photon/neutron radiation therapy in 45 patients with locally advanced, bulky, or postoperative recurrent prostate cancer treated at the University of Chicago between 1978 and 1991. Survival, disease-free survival, local control, and long-term complications are analyzed in detail. Methods and Materials: Between 1978 and 1991, 45 patients with locally advanced (> 5 cm Stage B2, Stage C, or Stage D1) prostate cancer underwent mixed-beam (photon/neutron) radiation therapy. Forty percent of the treatment was delivered with neutron irradiation at either the University of Chicago or Fermilab. Sixty percent of treatment was delivered with photons at the University of Chicago. Initially, the whole pelvis was irradiated to 50 photon Gy equivalent. This was followed by a boost to the prostate for an additional 20 photon Gy equivalent. Results: The median follow-up for patients in this series is 72 months. The overall 5-year actuarial survival was 72%, and the 5-year disease-free survival was 45%. Thus far, 18 patients have died. Eleven patients have died from prostate cancer and 7 from other medical illness. Twenty-seven patients are alive, and 12 of these patients have recurrent and or metastatic disease. The local control rate was 89% (40 out of 45). Histologic material was available on 18 patients following treatment (i.e., prostate biopsy in 16 patients and autopsy in 2 patients) and was negative for carcinoma in 13 (72%). Significant Grade 3-5 complications occurred in 36% (16 out of 45) of the patients treated with mixed-beam radiation therapy and were related to dose and beam quality. Factors related to survival, disease-free survival, local control, and complications are analyzed. Conclusions: The survival and local control results of mixed-beam radiation therapy at the University of Chicago appear to be superior to those series using photon radiation in patients with locally advanced prostate carcinoma
International Nuclear Information System (INIS)
This report documents the research and development goals reached within the reactor safety research project RS1503 ''Development and Application of Neutron Transport Methods and Uncertainty Analyses for Reactor Core Calculations''. The superordinate goal of the project is the development, validation, and application of neutron transport methods and uncertainty analyses for reactor core calculations. These calculation methods will mainly be applied to problems related to the core behaviour of light water reactors and innovative reactor concepts. The contributions of this project towards achieving this goal are the further development, validation, and application of deterministic and stochastic calculation programmes and of methods for uncertainty and sensitivity analyses, as well as the assessment of artificial neutral networks, for providing a complete nuclear calculation chain. This comprises processing nuclear basis data, creating multi-group data for diffusion and transport codes, obtaining reference solutions for stationary states with Monte Carlo codes, performing coupled 3D full core analyses in diffusion approximation and with other deterministic and also Monte Carlo transport codes, and implementing uncertainty and sensitivity analyses with the aim of propagating uncertainties through the whole calculation chain from fuel assembly, spectral and depletion calculations to coupled transient analyses. This calculation chain shall be applicable to light water reactors and also to innovative reactor concepts, and therefore has to be extensively validated with the help of benchmarks and critical experiments.
Development of tailorable advanced blanket insulation for advanced space transportation systems
Calamito, Dominic P.
1987-01-01
Two items of Tailorable Advanced Blanket Insulation (TABI) for Advanced Space Transportation Systems were produced. The first consisted of flat panels made from integrally woven, 3-D fluted core having parallel fabric faces and connecting ribs of Nicalon silicon carbide yarns. The triangular cross section of the flutes were filled with mandrels of processed Q-Fiber Felt. Forty panels were prepared with only minimal problems, mostly resulting from the unavailability of insulation with the proper density. Rigidizing the fluted fabric prior to inserting the insulation reduced the production time. The procedures for producing the fabric, insulation mandrels, and TABI panels are described. The second item was an effort to determine the feasibility of producing contoured TABI shapes from gores cut from flat, insulated fluted core panels. Two gores of integrally woven fluted core and single ply fabric (ICAS) were insulated and joined into a large spherical shape employing a tadpole insulator at the mating edges. The fluted core segment of each ICAS consisted of an Astroquartz face fabric and Nicalon face and rib fabrics, while the single ply fabric segment was Nicalon. Further development will be required. The success of fabricating this assembly indicates that this concept may be feasible for certain types of space insulation requirements. The procedures developed for weaving the ICAS, joining the gores, and coating certain areas of the fabrics are presented.
Advanced LWR concept with hard neutron spectrum (FLWR) for realizing flexible plutonium management
International Nuclear Information System (INIS)
An advanced LWR concept with hard neutron spectrum (FLWR) has been proposed in order to ensure sustainable energy supply in the future based on the well-experienced LWR technologies. The FLWR is essentially a BWR-type reactor, in which the moderation of neutron in the core is reduced by use of the hexagonal-shaped fuel assemblies with the triangular-tight-lattice fuel rod configuration. The core design concept of FLWR is to realize effective and flexible utilization of uranium and plutonium resources by two stages, corresponding to the advancement of the fuel cycle technologies and related infrastructures. The core in the first stage of FLWR aims at intensive utilization and preservation of plutonium based on the experiences of the current LWR and MOX utilization, and the one in the second stage realizes sustainable multiple plutonium recycling with a high conversion ratio over 1.0. The key point is that the first stage core can proceed to the second stage in the same reactor system during the reactor operation period. The present paper summarizes the recent core design studies of FLWR. (author)
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)
This document describes the initial global 2-D shielding analyses for the Advanced Neutron Source (ANS) reactor, the D2O 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
Time step control for solving the transient even-parity neutron transport equation
International Nuclear Information System (INIS)
Recently a discretization (in time) scheme for the transient even-parity neutron transport equation was successfully developed and implemented in the VARIANT/KIN3D code. Starting from this result we have first more thoroughly investigated the order of approximation, with respect to the time variable, arising from such a scheme especially having in mind the use of a variable time step. Based on the results of this analysis we have then introduced a time discretization scheme capable of keeping the mathematical structure of the equation also in the framework of a variable time step. Starting from this new scheme we have also introduced an automatic time step control option based on the estimation of the third order time derivative by comparing the backward and centered Euler schemes. In the paper, the developed theoretical background is presented and results for a small size fast reactor are shown. We have investigated the time step control option and find its behavior being in agreement with the physical phenomena. In particular the increase of the delayed neutron fraction changes the behavior in the expected manner. The time step control was, in fact, able to detect the presence of additional time scales introduced by two families of delayed neutrons in comparison with the case with almost no delayed neutrons. (authors)
International Nuclear Information System (INIS)
In this study, an analytical solution of the neutron transport equation in an annular reactor is presented with a short and rotating neutron source of the type S(x) δ (x- Vt), where V is the speed of annular pulsed reactor. The study is an extension of a previous study by Williams [12] carried out with a pulsed source of the type S(x) δ (t). In the new concept of annular pulsed reactor designed to produce continuous high flux, the core consists of a subcritical annular geometry pulsed by a rotating modulator, producing local super prompt critical condition, thereby giving origin to a rotating neutron pulse. An analytical solution is obtained by opening up of the annular geometry and applying one energy group transport theory in one dimension using applied mathematical techniques of Laplace transform and Complex Variables. The general solution for the flux consists of a fundamental mode, a finite number of harmonics and a transient integral. A condition which limits the number of harmonics depending upon the circumference of the annular geometry has been obtained. Inverse Laplace transform technique is used to analyse instability condition in annular reactor core. A regenerator parameter in conjunction with perimeter of the ring and nuclear properties is used to obtain stable and unstable harmonics and to verify if these exist. It is found that the solution does not present instability in the conditions stated in the new concept of annular pulsed reactor. (author)
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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)
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A power plant conceptual study (PPCS) has been conducted in the framework of the European fusion programme with the main objective to demonstrate the safety and environmental advantages and the economic viability of fusion power. Power plant models with limited (''near term concepts'') and advanced plasma physics and technological extrapolations (''advanced concepts'') were considered. Two near term plant models were selected, one employing a water cooled lithium-lead (WCLL), and the other one a helium cooled pebble bed (HCPB) blanket. Two variants were also considered for the advanced power plant models, one adopting a liquid metal blanket with a self-cooled lithium-lead breeder zone and a helium cooled steel structure (''dual coolant lithium lead'', DCLL), and the other one a self-cooled lithium-lead (SCLL) blanket with SiCf/SiC composite as structural material. This report provides a detailed documentation of the neutronics design analyses performed as part of the PPCS study for both the near term and advanced power plant models. Main issues are the assessment of the tritium breeding capability, the evaluation of the nuclear power generation and its spatial distribution, and the assessment and optimisation of the shielding performance. The analyses were based on three-dimensional Monte Carlo calculations with the MCNP code using suitable torus sector models developed for the different PPCS plant variants. (orig.)
Neutronic Analysis of Advanced SFR Burner Cores using Deep-Burn PWR Spent Fuel TRU Feed
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In this work, an advanced sodium-cooled fast TRU (Transuranics) burner core using deep-burn TRU feed composition discharged from small LWR cores was neutronically analyzed to show the effects of deeply burned TRU feed composition on the performances of sodium-cooled fast burner core. We consider a nuclear park that is comprised of the commercial PWRs, small PWRs of 100MWe for TRU deep burning using FCM (Fully Ceramic Micro-encapsulated) fuels and advanced sodium-cooled fast burners for their synergistic combination for effective TRU burning. In the small PWR core having long cycle length of 4.0 EFPYs, deep burning of TRU up to 35% is achieved with FCM fuel pins whose TRISO particle fuels contain TRUs in their central kernel. In this paper, we analyzed the performances of the advanced SFR burner cores using TRU feeds discharged from the small long cycle PWR deep-burn cores. Also, we analyzed the effect of cooling time for the TRU feeds on the SFR burner core. The results showed that the TRU feed composition from FCM fuel pins of the small long cycle PWR core can be effectively used into the advanced SFR burner core by significantly reducing the burnup reactivity swing which reduces smaller number of control rod assemblies to satisfy all the conditions for the self controllability than the TRU feed composition discharged from the typical PWR cores
Energy Technology Data Exchange (ETDEWEB)
Seubert, A.; Langenbuch, S.; Velkov, K.; Zwermann, W. [Gesellschaft fuer Anlagen- und Reaktorsicherheit mbH (GRS), Garching (Germany). Forschungsinstitute
2007-07-01
An overview is given of the recent progress at GRS concerning deterministic transport and Monte Carlo methods with thermal-hydraulic feedback. The development of the time-dependent 3D discrete ordinates transport code TORT-TD is described which has also been coupled with ATHLET. TORT-TD/ATHLET allows 3D pin-by-pin coupled analyses of transients using few energy groups and anisotropic scattering. As a step towards Monte Carlo steady-state calculations with nuclear point data and thermal-hydraulic feedback, MCNP has been prepared to incorporate thermal-hydraulic parameters. Results obtained for selected test cases demonstrate the applicability of deterministic and Monte Carlo neutron transport models coupled with thermo-fluiddynamics. (orig.)
International Nuclear Information System (INIS)
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
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The behavior of neutrons and gamma rays in a nuclear reactor or configuration of fissile material can be represented as a stochastic process. The observation of this stochastic process is usually achieved by measuring the fluctuations of the neutron and gamma ray population on the system. The general theory of the stochastic neutron field has been developed to a high degree. However, the theory of the stochastic nature of the gamma rays and neutrons couples the two processes. The generalized probability balances are developed from which the first and higher moments of the neutron and gamma rays fields are obtained. The paper also provides a description of the probability generating functions for both photon and neutron detectors that are the foundations for measurements of the fluctuations. The formalism developed in this paper for the representation of the statistical descriptors of the neutron-photon coupled field is applicable for many neutron noise analysis measurements
Application of ENDF nuclear data for testing a Monte-Carlo neutron and photon transport code
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A Monte-Carlo photon and neutron transport code was developed at OAEP. The code was written in C and C++ languages in an object-oriented programming style. Constructive solid geometry (CSG), rather than combinatorial, was used such that making its input file more readable and recognizable. As the first stage of code validation, data from some ENDF files, in the MCNP's specific format, were used and compared with experimental data. The neutron (from a 300 mCi Am/Be source) attenuation by water was chosen to compare the results. The agreement of the quantity 1/Σ among the calculation from SIPHON and MCNP, and the experiment - which are 10.39 cm, 9.71 cm and 10.25 cm respectively - was satisfactorily well within the experimental uncertainties. These results also agree with the 10.8 cm result of N.M., Mirza, et al. (author)
Transport equations and linear response of superfluid Fermi mixtures in neutron stars
Gusakov, M E
2010-01-01
We study transport properties of a strongly interacting superfluid mixture of two Fermi-liquids. A typical example of such matter is the neutron-proton liquid in the cores of neutron stars. To describe the mixture, we employ the Landau theory of Fermi-liquids, generalized to allow for the effects of superfluidity. We formulate the kinetic equation and analyze linear response of the system to vector (e.g., electromagnetic) perturbation. In particular, we calculate the transverse and longitudinal polarization functions for both liquid components. We demonstrate, that they can be expressed through the Landau parameters of the mixture and polarization functions of noninteracting matter (when the Landau quasiparticle interaction is neglected). Our results can be used, e.g., for studies of the kinetic coefficients and low-frequency long-wavelength collective modes in superfluid Fermi-mixtures.
GNES-R: Global nuclear energy simulator for reactors task 1: High-fidelity neutron transport
International Nuclear Information System (INIS)
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)
Energy Conversion Advanced Heat Transport Loop and Power Cycle
Energy Technology Data Exchange (ETDEWEB)
Oh, C. H.
2006-08-01
The Department of Energy and the Idaho National Laboratory are developing a Next Generation Nuclear Plant (NGNP) to serve as a demonstration of state-of-the-art nuclear technology. The purpose of the demonstration is two fold 1) efficient low cost energy generation and 2) hydrogen production. Although a next generation plant could be developed as a single-purpose facility, early designs are expected to be dual-purpose. While hydrogen production and advanced energy cycles are still in its early stages of development, research towards coupling a high temperature reactor, electrical generation and hydrogen production is under way. Many aspects of the NGNP must be researched and developed in order to make recommendations on the final design of the plant. Parameters such as working conditions, cycle components, working fluids, and power conversion unit configurations must be understood. Three configurations of the power conversion unit were demonstrated in this study. A three-shaft design with 3 turbines and 4 compressors, a combined cycle with a Brayton top cycle and a Rankine bottoming cycle, and a reheated cycle with 3 stages of reheat were investigated. An intermediate heat transport loop for transporting process heat to a High Temperature Steam Electrolysis (HTSE) hydrogen production plant was used. Helium, CO2, and an 80% nitrogen, 20% helium mixture (by weight) were studied to determine the best working fluid in terms cycle efficiency and development cost. In each of these configurations the relative component size were estimated for the different working fluids. The relative size of the turbomachinery was measured by comparing the power input/output of the component. For heat exchangers the volume was computed and compared. Parametric studies away from the baseline values of the three-shaft and combined cycles were performed to determine the effect of varying conditions in the cycle. This gives some insight into the sensitivity of these cycles to various
MCNP: a general Monte Carlo code for neutron and photon transport
International Nuclear Information System (INIS)
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
Basic parameter estimation of binary neutron star systems by the advanced LIGO/Vigro network
Energy Technology Data Exchange (ETDEWEB)
Rodriguez, Carl L.; Farr, Benjamin; Raymond, Vivien; Farr, Will M.; Littenberg, Tyson B.; Fazi, Diego; Kalogera, Vicky, E-mail: cr@u.northwestern.edu [Center for Interdisciplinary Exploration and Research in Astrophysics (CIERA) and Department of Physics and Astronomy, Northwestern University, 2145 Sheridan Road, Evanston, IL 60208 (United States)
2014-04-01
Within the next five years, it is expected that the Advanced LIGO/Virgo network will have reached a sensitivity sufficient to enable the routine detection of gravitational waves. Beyond the initial detection, the scientific promise of these instruments relies on the effectiveness of our physical parameter estimation capabilities. A major part of this effort has been toward the detection and characterization of gravitational waves from compact binary coalescence, e.g., the coalescence of binary neutron stars. While several previous studies have investigated the accuracy of parameter estimation with advanced detectors, the majority have relied on approximation techniques such as the Fisher Matrix which are insensitive to the non-Gaussian nature of the gravitational wave posterior distribution function. Here we report average statistical uncertainties that will be achievable for strong detection candidates (S/N = 20) over a comprehensive sample of source parameters. We use the Markov Chain Monte Carlo based parameter estimation software developed by the LIGO/Virgo Collaboration with the goal of updating the previously quoted Fisher Matrix bounds. We find the recovery of the individual masses to be fractionally within 9% (15%) at the 68% (95%) credible intervals for equal-mass systems, and within 1.9% (3.7%) for unequal-mass systems. We also find that the Advanced LIGO/Virgo network will constrain the locations of binary neutron star mergers to a median uncertainty of 5.1 deg{sup 2} (13.5 deg{sup 2}) on the sky. This region is improved to 2.3 deg{sup 2} (6 deg{sup 2}) with the addition of the proposed LIGO India detector to the network. We also report the average uncertainties on the luminosity distances and orbital inclinations of strong detections that can be achieved by different network configurations.
International Nuclear Information System (INIS)
1 - Nature of physical problem solved: The function of the AIRTRANS system is to calculate by Monte Carlo methods the radiation field produced by neutron and/or gamma-ray sources which are located in the atmosphere. The radiation field is expressed as the time - and energy-dependent flux at a maximum of 50 point detectors in the atmosphere. The system calculates un-collided fluxes analytically and collided fluxes by the 'once-more collided' flux-at-a-point technique. Energy-dependent response functions can be applied to the fluxes to obtain desired flux functionals, such as doses, at the detector point. AIRTRANS also can be employed to generate sources of secondary gamma radiation. 2 - Method of solution - Neutron interactions treated in the calculational scheme include elastic (isotropic and anisotropic) scattering, inelastic (discrete level and continuum) scattering, and absorption. Charged particle reactions, e.g, (n,p) are treated as absorptions. A built-in kernel option can be employed to take neutrons from the 150 keV to thermal energy, thus eliminating the need for particle tracking in this energy range. Another option used in conjunction with the neutron transport problem creates an 'interaction tape' which describes all the collision events that can lead to the production of secondary gamma-rays. This interaction tape subsequently can be used to generate a source of secondary gamma rays. The gamma-ray interactions considered include Compton scattering, pair production, and the photoelectric effect; the latter two processes are treated as absorption events. Incorporated in the system is an option to use a simple importance sampling technique for detectors that are many mean free paths from the source. In essence, particles which fly far from the source are split into fragments, the degree of fragmentation being proportional to the penetration distance from the source. Each fragment is tracked separately, thus increasing the percentage of computer time spent
Duffy, James B.
1993-01-01
The purpose of the Advanced Transportation System Study (ATSS) task area 1 study effort is to examine manned launch vehicle booster concepts and two-way cargo transfer and return vehicle concepts to determine which of the many proposed concepts best meets NASA's needs for two-way transportation to low earth orbit. The study identified specific configurations of the normally unmanned, expendable launch vehicles (such as the National Launch System family) necessary to fly manned payloads. These launch vehicle configurations were then analyzed to determine the integrated booster/spacecraft performance, operations, reliability, and cost characteristics for the payload delivery and return mission. Design impacts to the expendable launch vehicles which would be required to perform the manned payload delivery mission were also identified. These impacts included the implications of applying NASA's man-rating requirements, as well as any mission or payload unique impacts. The booster concepts evaluated included the National Launch System (NLS) family of expendable vehicles and several variations of the NLS reference configurations to deliver larger manned payload concepts (such as the crew logistics vehicle (CLV) proposed by NASA JSC). Advanced, clean sheet concepts such as an F-1A engine derived liquid rocket booster (LRB), the single stage to orbit rocket, and a NASP-derived aerospace plane were also included in the study effort. Existing expendable launch vehicles such as the Titan 4, Ariane 5, Energia, and Proton were also examined. Although several manned payload concepts were considered in the analyses, the reference manned payload was the NASA Langley Research Center's HL-20 version of the personnel launch system (PLS). A scaled up version of the PLS for combined crew/cargo delivery capability, the HL-42 configuration, was also included in the analyses of cargo transfer and return vehicle (CTRV) booster concepts. In addition to strictly manned payloads, two-way cargo
International Nuclear Information System (INIS)
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)
Hybrid parallel programming models for AMR neutron Monte-Carlo transport
International Nuclear Information System (INIS)
This paper deals with High Performance Computing (HPC) applied to neutron transport theory on complex geometries, thanks to both an Adaptive Mesh Refinement (AMR) algorithm and a Monte-Carlo (MC) solver. Several Parallelism models are presented and analyzed in this context, among them shared memory and distributed memory ones such as Domain Replication and Domain Decomposition, together with Hybrid strategies. The study is illustrated by weak and strong scalability tests on complex benchmarks on several thousands of cores thanks to the peta-flop supercomputer Tera100. (authors)
A vectorized Monte Carlo method with pseudo-scattering for neutron transport analysis
International Nuclear Information System (INIS)
A vectorized Monte Carlo method has been developed for the neutron transport analysis on the vector supercomputer HITAC S810. In this method, a multi-particle tracking algorithm is adopted and fundamental processing such as pseudo-random number generation is modified to use the vector processor effectively. The flight analysis of this method is characterized by the new algorithm with pseudo-scattering. This algorithm was verified by comparing its results with those of the conventional one. The method realized a speed-up of factor 10; about 7 times by vectorization and 1.5 times by the new algorithm for flight analysis
Benchmark results for the critical slab and sphere problem in one-speed neutron transport theory
International Nuclear Information System (INIS)
Research highlights: → The critical slab and sphere problem in neutron transport under Case eigenfunction formalism is considered. → These equations reduce to integral expressions involving X functions. → Gauss quadrature is not ideal but DE quadrature is well-suited. → Several fold decrease in computational effort with improved accuracy is realisable. - Abstract: In this paper benchmark numerical results for the one-speed criticality problem with isotropic scattering for the slab and sphere are reported. The Fredholm integral equations of the second kind based on the Case eigenfunction formalism are numerically solved by Neumann iterations with the Double Exponential quadrature.
Simulation of neutron transport process, photons and charged particles within the Monte Carlo method
International Nuclear Information System (INIS)
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
Domain decomposition and terabyte tallies with the OpenMC Monte Carlo neutron transport code
International Nuclear Information System (INIS)
Memory limitations are a key obstacle to applying Monte Carlo neutron transport methods to high-fidelity full-core reactor analysis. Billions of unique regions are needed to carry out full-core depletion and fuel performance analyses, equating to terabytes of memory for isotopic abundances and tally scores - far more than can fit on a single computational node in modern architectures. This work introduces an implementation of domain decomposition that addresses this problem, demonstrating excellent scaling up to a 2.39TB mesh-tally distributed across 512 compute nodes running a full-core reactor benchmark on the Mira Blue Gene/Q supercomputer at Argonne National Laboratory. (author)
3-D Monte Carlo neutron-photon transport code JMCT and its algorithms
International Nuclear Information System (INIS)
JMCT Monte Carlo neutron and photon transport code has been developed which is based on the JCOGIN toolbox. JCOGIN includes the geometry operation, tally, the domain decomposition and the parallel computation about particle (MPI) and spatial domain (OpenMP) etc. The viewdata of CAD is equipped in JMCT preprocessor. The full-core pin-mode, which is from Chinese Qinshan-II nuclear power station, is design and simulated by JMCT. The detail pin-power distribution and keff results are shown in this paper. (author)
Adjacent-cell Preconditioners for solving optically thick neutron transport problems
International Nuclear Information System (INIS)
We develop, analyze, and test a new acceleration scheme for neutron transport methods, the Adjacent-cell Preconditioner (AP) that is particularly suited for solving optically thick problems. Our method goes beyond Diffusion Synthetic Acceleration (DSA) methods in that it's spectral radius vanishes with increasing cell thickness. In particular, for the ID case the AP method converges immediately, i.e. in one iteration, to 10-4 pointwise relative criterion in problems with dominant cell size of 10 mfp or thicker. Also the AP has a simple formalism and is cell-centered hence, multidimensional and high order extensions are easier to develop, and more efficient to implement
Development of calculational procedures for the neutron physics design of advanced reactors
International Nuclear Information System (INIS)
The Nuclear Reactor Center Karlsruhe has been involved with the development of Light Water Tight Lattice Reactors (LWTLR) since more than ten years. A considerable amount of thermohydraulic and nuclear physics code development has been performed during this time. The present paper describes the main aspects of the neutron physics calculational tools. From the neutron physics point of view, two different tasks have to be adapted for LWTLR calculations: determination of mean cross sections sets within the hexagonal fuel assemblies (FA); determination of the characteristics of LWR cores with hexagonal FA. All developments for the neutron physics design of LWTLR have been performed within the established system for Fast Breeder Reactor (FBR) calculations at KfK, KAPROS, using a various number of available options for FBR work. The present status of the calculational tools for LWTLR-investigations will be described, especially the features of a newly developed KAPROS procedure ARCOSI: Advanced Reactor COre SImulator, including: Preparation of the ARCOSI library HXSLIB, containing burnup dependent cross section sets for FA with control rod positions containing control rod material or waterholes and with borated water in the moderator region of the pin cells. Also, data for different coolant densities and pin cell temperatures may be processed; Simulation of equilibrium core calculations, including critical reactivity search by waterboration control and simple FA management. Three-dimensional full core calculations are performed with the KAPROS version of the hexagonal nodal code HEXNOD, developed by Wagner, KWU; Powerful interfaces for interactive graphical analysis of results. (author). 37 refs, 12 figs
Review and Assessment of Neutron Cross Section and Nubar Covariances for Advanced Reactor Systems
Energy Technology Data Exchange (ETDEWEB)
Maslov,V.M.; Oblozinsky, P.; Herman, M.
2008-12-01
In January 2007, the National Nuclear Data Center (NNDC) produced a set of preliminary neutron covariance data for the international project 'Nuclear Data Needs for Advanced Reactor Systems'. The project was sponsored by the OECD Nuclear Energy Agency (NEA), Paris, under the Subgroup 26 of the International Working Party on Evaluation Cooperation (WPEC). These preliminary covariances are described in two recent BNL reports. The NNDC used a simplified version of the method developed by BNL and LANL that combines the recent Atlas of Neutron Resonances, the nuclear reaction model code EMPIRE and the Bayesian code KALMAN with the experimental data used as guidance. There are numerous issues involved in these estimates of covariances and it was decided to perform an independent review and assessment of these results so that better covariances can be produced for the revised version in future. Reviewed and assessed are uncertainties for fission, capture, elastic scattering, inelastic scattering and (n,2n) cross sections as well as prompt nubars for 15 minor actinides ({sup 233,234,236}U, {sup 237}Np, {sup 238,240,241,242}Pu, {sup 241,242m,243}Am and {sup 242,243,244,245}Cm) and 4 major actinides ({sup 232}Th, {sup 235,238}U and {sup 239}Pu). We examined available evaluations, performed comparison with experimental data, taken into account uncertainties in model parameterization and made use state-of-the-art nuclear reaction theory to produce the uncertainty assessment.
Advanced Neutron Source project information management: A model for the future
International Nuclear Information System (INIS)
The Advanced Neutron Source (ANS) is a proposed new research facility that will provide steady-state beams of neutrons for experiments by more than 1,000 researchers per year in the fields of materials science and engineering, biology, chemistry, materials analysis, and nuclear science. The facility will also include irradiation capabilities to produce radioisotopes for medical applications, research, industry, and materials testing. This paper discusses the architecture and data flow used by the project, some quantitative examinations of potential cost savings and return on investment, and software applications used to generate and manage data across IBM-compatible personal computers, Macintosh, and Unix-based workstations. Personnel management aspects addressed include providing paper copy to users only when needed for adequate technical review, using graded approaches to providing support for numerous user-needed software applications, and implementing a phased approach to compliance with computer-aided acquisition and logistic support (CALS) standards that allows sufficient user flexibility for performing technical tasks while providing needed data sharing and integration
Advanced Neutron Source enrichment study. Volume 2: Appendices -- Final report, Revision 12/94
International Nuclear Information System (INIS)
A study has been performed of the impact on performance of using low enriched uranium (20% 235U) or medium enriched uranium (35% 235U) as an alternative fuel for the Advanced Neutron Source, which is currently designed to use uranium enriched to 93% 235U. 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. Volume 2 of this report contains 26 appendices containing results, meeting minutes, and fuel panel presentations. There are 26 appendices in this volume
Jeffrey, George
1988-01-01
X-ray and neutron crystallography have played an increasingly impor tant role in the chemical and biochemical sciences over the past fifty years. The principal obstacles in this methodology, the phase problem and com puting, have been overcome. The former by the methods developed in the 1960's and just recognised by the 1985 Chemistry Nobel Prize award to Karle and Hauptman, the latter by the dramatic advances that have taken place in computer technology in the past twenty years. Within the last decade, two new radiation sources have been added to the crystallographer's tools. One is synchrotron X-rays and the other is spallation neutrons. Both have much more powerful fluxes than the pre vious sources and they are pulsed rather than continuos. New techniques are necessary to fully exploit the intense continuos radiation spectrum and its pulsed property. Both radiations are only available from particular National Laboratories on a guest-user basis for scientists outside these Na tional Laboratories. Hi...
Steady-state thermal-hydraulic design analysis of the Advanced Neutron Source reactor
International Nuclear Information System (INIS)
The Advanced Neutron Source (ANS) is a research reactor that is planned for construction at Oak Ridge National Laboratory. This reactor will be a user facility with the major objective of providing the highest continuous neutron beam intensities of any reactor in the world. Additional objectives for the facility include providing materials irradiation facilities and isotope production facilities as good as, or better than, those in the High Flux Isotope Reactor. To achieve these objectives, the reactor design uses highly subcooled heavy water as both coolant and moderator. Two separate core halves of 67.6-L total volume operate at an average power density of 4.5 MW(t)/L, and the coolant flows upward through the core at 25 m/s. Operating pressure is 3.1 MPa at the core inlet with a 1.4-MPa pressure drop through the core region. Finally, in order to make the resources available for experimentation, the fuel is designed to provide a 17-d fuel cycle with an additional 4 d planned in each cycle for the refueling process. This report examines the codes and models used to develop the thermal-hydraulic design for ANS, as well as the correlations and physical data; evaluates thermal-hydraulic uncertainties; reports on thermal-hydraulic design and safety analysis; describes experimentation in support of the ANS reactor design and safety analysis; and provides an overview of the experimental plan
International Nuclear Information System (INIS)
A study has been performed to measure the potential of recriticality during hypothetical severe accident in Advanced Neutron Source (ANS). For the lumped debris configuration in the Reactor Coolant System (RCS), as found in the previous study, recriticality potential may be very low. However, if fuel debris is dispersed and mixed with heavy water in RCS, recriticality potential has been predicted to be substantial depending on thermal-hydraulic conditions surrounding fuel debris mixture. The recriticality potential in RCS is substantially reduced for the three element core design with 50% enrichment. Also, as observed in the previous study, strong dependencies of keff on key thermal hydraulic parameters are shown. Light water contamination is shown to provide a positive reactivity, and void formation due to boiling of mixed water provides enough negative reactivity and to bring the system down to subcritical. For criticality potential in the subpile room, the lumped debris configuration does not pose a concern. Dispersed configuration in light water pool of the subpile room is also unlikely to result in criticality. However, if the debris is dispersed in the pool that is mixed with heavy water, the results indicate that a substantial potential exists for the debris to reach the criticality. However, if prompt recriticality disperses the debris completely in the subpile room pool, subsequent recriticality may be prevented since neutron leakage effects become large enough
International Nuclear Information System (INIS)
The Advanced Neutron Source Reactor (ANSR) is currently being designed to become the world's highest-flux, steady-state, thermal neutron source for scientific experiments. Highly subcooled, heavy-water coolant flows vertically upward at a very high velocity of 25 m/s through parallel aluminum fuel-plates. The core has average and peak heat fluxes of 5.9 and 12 MW/m2, respectively. In this configuration, both flow excursion (FE) and true critical heat flux (CHF), represent potential thermal limitations. The availability of experimental data for both FE and true CHF at the conditions applicable to the ANSR is very limited. A Thermal Hydraulic Test Loop (THTL) facility was designed and built to simulate a full-length coolant subchannel of the core, allowing experimental determination of both thermal limits under the expected ANSR T/H conditions. A series of FE tests with water flowing vertically upward was completed over a nominal heat flux range of 6 to 14 MW/m2 and a corresponding velocity range of 8 to 21 m/s. Both the exit pressure (1.7 MPa) and inlet temperature (45 degrees C) were maintained constant for these tests, while the loop was operated in a ''stiff''(constant flow) mode. Limited experiments were also conducted at 12 MW/m2 using a ''soft'' mode (near constant pressure-drop) for actual FE burnout tests and using a ''stiff' mode for true CHF tests, to compare with the original FE experiments
Advanced neutron source project information management. A model for the future
International Nuclear Information System (INIS)
The Advanced Neutron Source (ANS) is a proposed new research facility that will provide steady-state beams of neutrons for experiments by more than 1000 researchers per year in the fields of materials science and engineering, biology, chemistry, materials analysis, and nuclear science. The facility will also include irradiation capabilities to produce radioisotopes for medical applications, research, industry, and materials testing. This paper discusses the architecture and data flow used by the project, some quantitative examinations of potential cost savings and return on investment and software applications used to generate and manage data across IBM-compatible personal computers, Macintosh, and Unix-based workstations. Personnel management aspects addressed include providing paper copy to users only when needed for adequate technical review, using graded approaches to providing support for numerous user-needed software applications, and implementing a phased approach to compliance with computer-aided acquisition and logistic support (CALS) standards that allows sufficient user flexibility for performing technical tasks while providing needed data sharing and integration
Advanced Neutron Source enrichment study -- Volume 1: Main report. Final report, Revision 12/94
International Nuclear Information System (INIS)
A study has been performed of the impact on performance of using low enriched uranium (20% 235U) or medium enriched uranium (35% 235U) as an alternative fuel for the Advanced Neutron Source, which is currently designed to use uranium enriched to 93% 235U. 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. Volume 2 of this report contains 26 appendices containing results, meeting minutes, and fuel panel presentations
International Nuclear Information System (INIS)
A Space-Point Energy-Group integral transport theory method (SPEG) is developed and applied to the local and global calculations of the Yugoslav RA reactor. Compared to other integral transport theory methods, the SPEG distinguishes by (1) the arbitrary order of the polynomial, (2) the effective determination of integral parameters through point flux values, (3) the use of neutron balance condition. as a posterior measure of the accuracy of the calculation and (4) the elimination of the subdivisions- into zones, in realistic cases. In addition, different direct (collision probability) and indirect (Monte Carlo) approaches to integral transport theory have been investigated and Some effective acceleration procedures introduced. The study was performed on three test problems in plane and cylindrical geometry, as well as on the nine-region cell of the RA reactor. In particular, the limitations of the integral transport theory including its non-applicability to optically large material regions and to global reactor calculations were examined. The proposed strictly multipoint approach, avoiding the subdivision into zones and groups, seems to provide a good starting point to overcome these limitations of the integral transport theory. (author)
Practical application of passive safety features for the advanced neutron source cooling system
International Nuclear Information System (INIS)
The results of a conceptual design study leading to the definition of a reference design for the Advanced Neutron Source (ANS) heavy water cooling system are presented. The objective of this study was to define a cooling system that not only met the ANS goals for operating parameters, reliability, availability, and maintainability, but also used inherent, passive, and diverse features and characteristics to satisfy the ANS internal events core melt goal of -5/yr. The approach taken in this study was to define a cooling system configuration having the minimum basic components and characteristics to satisfy the requirements for normal operation, and then to add only those features necessary to meet the requirements for all emergency design-basis events
Report of the advanced neutron source (ANS) aluminum cladding corrosion workshop
International Nuclear Information System (INIS)
The Advanced Neutron Source (ANS) Corrosion Workshop on aluminum cladding corrosion in reactor environments is summarized. The Workshop was held to examine the aluminum cladding oxidation studies being conducted in support of the ANS design. This report was written principally to provide a record of the ideas and judgments expressed by the workshop attendees. The ANS operating heat flux is significantly higher than that in existing reactors, and early experiments indicate that there may be an aluminum cladding oxidation problem unique to higher heat fluxes or associated cladding temperatures that, if not solved, may limit the operation of the ANS to unacceptably low power levels. A brief description of the information presented by each speaker is included along with a compilation of the most significant ideas and recommended research areas. The appendixes contain a copy of the workshop agenda and a list of attendees
International Nuclear Information System (INIS)
The High-flux Advanced Neutron Application Reactor (HANARO) has the equipment to treat the gaseous radioactive waste generated within itself but it does not have proper ways to treat the waste in either a liquid or solid form. For the last 5 years, every effort has been made to reduce the radioactive wastes in HANARO. Improvements of the equipment and operating procedures regarding the generation of radioactive waste in the field have resulted in an effective reduction of the radioactive waste. In addition, as an outgrowth of the research efforts in connection with the demand in the field, several new methods have been developed to effectively treat the radioactive waste. Efforts to reduce and treat the radioactive waste in HANARO will continue and that will contribute greatly to an improvement of the reliability and safety of HANARO. (authors)
Coarse mesh transport method for whole-core neutronics analysis in cylindrical geometry
International Nuclear Information System (INIS)
In this paper, a coarse mesh transport (Comet) method has been developed to predict neutronics behaviors in reactor cores with 2-dimensional cylindrical (r, θ) geometry. Based on neutron balance in each coarse mesh, the outgoing partial current (or flux) from a coarse mesh is calculated as the sum of the local responses to all incoming partial currents/fluxes via pre-computed incident response expansion functions. These local functions essentially represents the probabilities that the incoming neutrons and their progenies contributes to the quantities of interest such as outgoing partial currents/fluxes and fission densities. The Comet code was benchmark against the Monte Carlo Code MCNP in a test problem representative of a simplified Pebble Bed reactor core in (r, θ) geometry, consisting of an inner reflector, an annular fuel region, and a controlled outer reflector. With 4th order expansion in space and 2nd expansion order in the polar and azimuthal angles, the Comet calculation agrees very well with the MCNP reference solution. (Author)
International Nuclear Information System (INIS)
The stationary solution of the one-speed neutron transport equation in a semi infinite stochastic medium with pure-triplet scattering is considered. The cross section function of the medium is assumed to be a continuous random function of position, with fluctuations about the mean taken as Gaussian distributed. The joint probability distribution function of these Gaussian random variables is used to calculate the ensemble-averaged solution for an arbitrary correlation function. The problem is solved at first in the deterministic case, then the solution is averaged using Gaussian joint probability distribution function. Numerical results are given for the radiant neutron energy and net neutron flux. Keywords: Neutron transport, Stochastic media, Gaussian statistics, Pure triplet scattering
Žukauskaitėa, A; Plukienė, R; Ridikas, D
2007-01-01
Particle accelerators and other high energy facilities produce penetrating ionizing radiation (neutrons and γ-rays) that must be shielded. The objective of this work was to model photon and neutron transport in various materials, usually used as shielding, such as concrete, iron or graphite. Monte Carlo method allows obtaining answers by simulating individual particles and recording some aspects of their average behavior. In this work several nuclear experiments were modeled: AVF 65 (AVF cyclotron of Research Center of Nuclear Physics, Osaka University, Japan) – γ-ray beams (1-10 MeV), HIMAC (heavy-ion synchrotron of the National Institute of Radiological Sciences in Chiba, Japan) and ISIS-800 (ISIS intensive spallation neutron source facility of the Rutherford Appleton laboratory, UK) – high energy neutron (20-800 MeV) transport in iron and concrete. The calculation results were then compared with experimental data.compared with experimental data.
OMEGA, Subcritical and Critical Neutron Transport in General 3-D Geometry by Monte-Carlo
International Nuclear Information System (INIS)
1 - Description of problem or function: OMEGA is a Monte Carlo code for the solution of the stationary neutron transport equation with k-eff as the Eigenvalue. A three-dimensional geometry is permitted consisting of a very general arrangement of three basic shapes (columns with circular, rectangular, or hexagonal cross section with a finite height and different material layers along their axes). The main restriction is that all the basic shapes must have parallel axes. Most real arrangements of fissile material inside and outside a reactor (e.g., in a fuel storage or transport container) can be described without approximation. The main field of application is the estimation of criticality safety. Many years of experience and comparison with reference cases have shown that the code together with the built-in cross section libraries gives reliable results. The following results can be calculated: - the effective multiplication factor k-eff; - the flux distribution; - reaction rates; - spatially and energetically condensed cross sections for later use in a subsequent OMEGA run. A running job may be interrupted and continued later, possibly with an increased number of batches for an improved statistical accuracy. The geometry as well as the k-eff results may be visualized. The use of the code is demonstrated by many illustrating examples. 2 - Method of solution: The Monte Carlo method is used with neutrons starting from an initial source distribution. The histories of a generation (or batch) of neutrons are followed from collision to collision until the histories are terminated by capture, fission, or leakage. For the solution of the Eigenvalue problem, the starting positions of the neutrons for a given generation are determined by the fission points of the preceding generation. The summation of the results starts only after some initial generations when the spatial part of the fission source has converged. At present the code uses the BNAB-78 subgroup library of the
System of constants to calculate neutron transport with energy 10-2-4x108 eV
International Nuclear Information System (INIS)
A description of the library of nuclear data to calculate neutron transport in the energy range 10-2 eV-4x102 MeV (BND-400) is presented. The library contains a seven-group system of data for neutrons of E>10.5 MeV and a standard 26-group system for neutrons with E10.5 MeV, and those for matching with the file of data for neutrons with E<10.5 MeV are briefly described. In the BND-400 complex there are subroutines, which allow one to calculate the cross sections for neutron interaction with nuclei of matter with the help of various methods and models as well as to calculate group cross sections. It also provides output files in the form convenient for work. A brief instruction for BND-400 explotation on the computer BESM-6 is given
The impact of fuel particle size distribution on neutron transport in stochastic media
International Nuclear Information System (INIS)
This paper presents a study of the particle size distribution impact on neutron transport in three-dimensional stochastic media. 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. We construct 15 cases 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 (keff) and flux distribution along axial and radial directions are compared between different size distributions. At low packing fraction and low optical thickness, the size distribution has a significant impact on radiation transport in stochastic media, which can cause as high as ∼270 pcm difference in keff value and ∼2.6% relative error difference in peak flux. As the packing fraction and optical thickness increase, the impact gradually dissipates. (authors)
International Nuclear Information System (INIS)
The first-order neutron transport equation was solved by the least-squares finite element method based on the discrete ordinates discretization. For the traditional source iteration method is very slowly for the optically thick diffusive medium, sometime even divergent especially for the scattering ratio is close to unity, so the acceleration method should be proposed. There is only diffusive synthetical acceleration (DSA) for the discontinuous finite element method (DFEM) and almost no one for the least- squares finite element method. The additive angular dependent rebalance (AADR) acceleration arithmetic and its extrapolate method were given, in which the additive modification was used. It was applied to solve the transport equation with fixed source, fission source, in optically thick diffusive regions and with unstructured-mesh. The numerical results of benchmark problems demonstrate that the arithmetic can shorten the CPU time about 1.5-2 times and give high precise. (authors)
CACTUS, a characteristics solution to the neutron transport equations in complicated geometries
International Nuclear Information System (INIS)
CACTUS has been written to solve the multigroup neutron transport equation in a general two-dimensional geometry. The method is based upon a characteristics formulation for the problem in which the transport equation is integrated explicitly along straight line tracks that are suitably distributed throughout the problem. Source distributions and scattering are assumed to be isotropic, but the only restriction on geometry is that the outer boundary should be rectangular. Within this rectangular boundary the user is free to build his problem geometry using any combination of intersecting straight lines and circular arcs. The theory of the method is described, followed by some details of a coding, a sensitivity study on the number of tracks required to integrate fluxes in a particular problem, a user's guide, and a few test cases. (author)
On the use of the conjugate gradient method for the solution of the neutron transport equation
International Nuclear Information System (INIS)
The possibility of using the standard conjugate gradient (CG) method to directly solve the Sn equations based on the diamond difference scheme is studied for mono-energetic neutron transport problems with isotropic scattering. It is shown that such a direct use is possible for practical heterogeneous problems with a significant speed-up over the conventional source iteration (SI) method except for the problems that are prone to unphysical negative fluxes. Some recipes are suggested to make use of the CG-method even in those cases which need negative flux fix-up in the SI-method. The transport synthetic acceleration scheme, recently developed by Ramone [Nucl. Sci. Eng. 125 (1997) 257] and others, is shown to be useful in such cases. A symmetrisation scheme for the coefficient matrix has also been presented to enable the use of the CG-method. This scheme is compared with another approach of using weighted inner products
Deterministic numerical methods for eigenvalue problems in neutron linear transport theory
International Nuclear Information System (INIS)
Conventional deterministic numerical methods applied to eigenvalue problems in neutron transport theory in the discrete ordinates (SN) formulation are described. In order to architect an efficient time-dependent simulator for thermal reactor cores, for the cases where diffusion theory fails to give good results, the accuracy of a spectral nodal method applied to a simplified time-dependent transport model is investigated. Moreover, in order to generate the initial conditions for the one-dimensional kinetics problems, considered in the CINUNI section section of the simulator, a numerical method for SN eigenvalue problems with no spatial truncation errors is introduced. The convergence of the outer iterations with the Tchebycheff technique and implemented the option of albedo boundary conditions has been accelerated. These albedo boundary exactly substitute the top and bottom reflector regions. (author)
International Nuclear Information System (INIS)
The reliability of calculation tools to evaluate and calculate dose rates appearing behind multi-layered shields is important with regard to the certification of transport and storage casks. Actual benchmark databases like SINBAD do not offer such configurations because they were developed for reactor and accelerator purposes. Due to this, a bench-mark-suite based on own experiments that contain dose rates measured in different distances and levels from a transport and storage cask and on a public benchmark to validate Monte-Carlo-transport-codes has been developed. The analysed and summarised experiments include a 60Co point-source located in a cylindrical cask, a 252Cf line-source shielded by iron and polyethylene (PE) and a bare 252Cf source moderated by PE in a concrete-labyrinth with different inserted shielding materials to quantify neutron streaming effects on measured dose rates. In detail not only MCNPTM (version 5.1.6) but also MAVRIC, included in the SCALE 6.1 package, have been compared for photon and neutron transport. Aiming at low deviations between calculation and measurement requires precise source term specification and exact measurements of the dose rates which have been evaluated carefully including known uncertainties. In MAVRIC different source-descriptions with respect to the group-structure of the nuclear data library are analysed for the calculation of gamma dose rates because the energy lines of 60Co can only be modelled in groups. In total the comparison shows that MCNPTM fits very wall to the measurements within up to two standard deviations and that MAVRIC behaves similarly under the prerequisite that the source-model can be optimized. (author)
Large-scale Monte Carlo neutron transport calculations with thermal hydraulic feedback
International Nuclear Information System (INIS)
Highlights: • Method of internal coupling, based on dynamic material distribution, is presented. • The Wielandt shift method is implemented to accelerate Mote Carlo calculations. • The Uniform Fission Site method is introduced for tallies with large numbers of bins. • The stochastic approximation scheme is used to stabilize coupled code convergence. - Abstract: The Monte Carlo method provides the most accurate description of the particle transport problem. The criticality problem is simulated by following the histories of individual particles without approximating the energy, angle or the coordinate dependence. These calculations are usually done using homogeneous thermal hydraulic conditions. This is a very crude approximation in the general case. In this paper, the method of internal coupling between neutron transport and thermal hydraulics is presented. The method is based on dynamic material distribution, where coordinate dependent temperature and density information is supplied on the fly during the transport calculation. This method does not suffer from the deficiencies characteristic of the external coupling via the input files. In latter case, the geometry is split into multiple cells having distinct temperatures and densities to supply the feedback. The possibility to efficiently simulate large scale geometries at pin-by-pin and subchannel level resolution was investigated. The Wielandt shift method for reducing the dominance ratio of the system and accelerating the fission source convergence was implemented. During the coupled iteration a detailed distribution of the fission heat deposition is required by the thermal hydraulics calculation. Providing reasonable statistical uncertainties for tallies having large numbers of bins, is a complicated task. This problem was resolved by applying the Uniform Fission Site method. Previous investigations showed that the convergence of the coupled neutron transport/thermal hydraulics calculation is limited by
Advances in nuclear data and all-particle transport for radiation oncology
International Nuclear Information System (INIS)
Fast neutrons have been used to treat over 15,000 cancer patients worldwide and proton therapy is rapidly emerging as a treatment of choice for tumors around critical anatomical structures. Neutron therapy requires evaluated data to ∼70 MeV while proton therapy requires data to ∼250 MeV. Collaboration between Lawrence Livermore National Laboratory (LLNL) and the medical physics community has revealed limitations in nuclear cross section evaluations and radiation transport capabilities that have prevented neutron and proton radiation therapy centers from using Monte Carlo calculations to accurately predict dose in patients. These evaluations require energy- and angle-dependent cross sections for secondary neutrons, charged-particles and recoil nuclei. We are expanding the LLNL nuclear databases to higher energies for biologically important elements and have developed a three-dimensional, all-particle Monte Carlo radiation transport code that uses computer-assisted-tomography (CT) images as the input mesh. This code, called PEREGRINE calculates dose distributions in the human body and can be used as a tool to determine the dependence of dose on details of the evaluated nuclear data. In this paper, we will review the status of the nuclear data required for neutron and proton therapy, describe the capabilities of the PEREGRINE package, and show the effects of tissue inhomogeneities on dose distribution
TIMOC-72, 3-D Time-Dependent Homogeneous or Inhomogeneous Neutron Transport by Monte-Carlo
International Nuclear Information System (INIS)
1 - Nature of physical problem solved: TIMOC solves the energy and time dependent (or stationary) homogeneous or inhomogeneous neutron transport equation in three-dimensional geometries. The program can treat all commonly used scattering kernels, such as absorption, fission, isotropic and anisotropic elastic scattering, level excitation, the evaporation model, and the energy transfer matrix model, which includes (n,2n) reactions. The exchangeable geometry routines consist at present of (a) periodical multilayer slab, spherical and cylindrical lattices, (b) an elaborate three-dimensional cylindrical geometry which allows all kinds of subdivisions, (c) the very flexible O5R geometry routine which is able to describe any body combinations with surfaces of second order. The program samples the stationary or time-energy-region dependent fluxes as well as the transmission ratios between geometrical regions and the following integral quantities or eigenvalues, the leakage rate, the slowing down density, the production to source ratio, the multiplication factor based on flux and collision estimator, the mean production time, the mean destruction time, time distribution of production and destruction, the fission rates, the energy dependent absorption rates, the energy deposition due to elastic scattering for the different geometrical regions. 2 - Method of solution: TIMOC is a Monte Carlo program and uses several, partially optional variance reducing techniques, such as the method of expected values (weight factor), Russian roulette, the method of fractional generated neutrons, double sampling, semi-systematic sampling and the method of expected leakage probability. Within the neutron lifetime a discrete energy value is given after each collision process. The nuclear data input is however done by group averaged cross sections. The program can generate the neutron fluxes either resulting from an external source or in the form of fundamental mode distributions by a special
Agent code: Neutron transport benchmark example and extension to 3D lattice geometry
Directory of Open Access Journals (Sweden)
Hursin Mathieu
2005-01-01
Full Text Available The general methodology be hind 2D arbitrary geometry neutron transport AGENT code is the theory of R-functions, which al lows for simple modeling of complex geometries, and the method of characteristics, which solves the integral transport equation along characteristic neutron trajectories. This paper focuses on the extension of the methodology to ac count for 3D lattice geometries. Since the direct application of method of characteristics to 3D non-homogenized core con figuration may re quire a tremendous amount of memory and computing time, an alternative approximate solution based on coupling 2D method of characteristics and 1D diffusion solution is developed. The planar 2D method of characteristics and axial 1D diffusion solutions are coupled through the trans verse leak age. The use of a lower order 1D solution in the axial direction is justified by the fact that more heterogeneity in current PWR and BWR reactor cores occurs in the radial direction than in the axial one. In order to demonstrate the versatility and accuracy of the AGENT code, a 2D heterogeneous lattice problem, C5G7 is described in details. A theoretical description of the coupling methodology for 3D method of characteristics solution is followed by preliminary validation in comparison to the DeCART code.
International Nuclear Information System (INIS)
The purpose of this work is to solve the neutron transport equation in discrete-ordinates and X-Y geometry by developing and using the strong discontinuous and strong modified discontinuous nodal finite element schemes. The strong discontinuous and modified strong discontinuous nodal finite element schemes go from two to ten interpolation parameters per cell. They are describing giving a set Dc and polynomial space Sc corresponding for each scheme BDMO, RTO, BL, BDM1, HdV, BDFM1, RT1, BQ and BDM2. The solution is obtained solving the neutron transport equation moments for each nodal scheme by developing the basis functions defined by Pascal triangle and the Legendre moments giving in the polynomial space Sc and, finally, looking for the non singularity of the resulting linear system. The linear system is numerically solved using a computer program for each scheme mentioned . It uses the LU method and forward and backward substitution and makes a partition of the domain in cells. The source terms and angular flux are calculated, using the directions and weights associated to the SN approximation and solving the angular flux moments to find the effective multiplication constant. The programs are written in Fortran language, using the dynamic allocation of memory to increase efficiently the available memory of the computing equipment. (Author)
International Nuclear Information System (INIS)
The computer code block VENTURE, designed to solve multigroup neutronics problems with application of the finite-difference diffusion-theory approximation to neutron transport (or alternatively simple P1) in up to three-dimensional geometry is described. A variety of types of problems may be solved: the usual eigenvalue problem, a direct criticality search on the buckling, on a reciprocal velocity absorber (prompt mode), or on nuclide concentrations, or an indirect criticality search on nuclide concentrations, or on dimensions. First-order perturbation analysis capability is available at the macroscopic cross section level
International Nuclear Information System (INIS)
The report documents the computer code block VENTURE designed to solve multigroup neutronics problems with application of the finite-difference diffusion-theory approximation to neutron transport (or alternatively simple P1) in up to three-dimensional geometry. It uses and generates interface data files adopted in the cooperative effort sponsored by the Reactor Physics Branch of the Division of Reactor Research and Development of the Energy Research and Development Administration. Several different data handling procedures have been incorporated to provide considerable flexibility; it is possible to solve a wide variety of problems on a variety of computer configurations relatively efficiently
Energy Technology Data Exchange (ETDEWEB)
Vondy, D.R.; Fowler, T.B.; Cunningham, G.W.
1977-11-01
The report documents the computer code block VENTURE designed to solve multigroup neutronics problems with application of the finite-difference diffusion-theory approximation to neutron transport (or alternatively simple P/sub 1/) in up to three-dimensional geometry. It uses and generates interface data files adopted in the cooperative effort sponsored by the Reactor Physics Branch of the Division of Reactor Research and Development of the Energy Research and Development Administration. Several different data handling procedures have been incorporated to provide considerable flexibility; it is possible to solve a wide variety of problems on a variety of computer configurations relatively efficiently.
Energy Technology Data Exchange (ETDEWEB)
None, None
2005-08-01
This document presents a summary of the evaluation and comments provided by the review panel for the FY 2005 Department of Energy (DOE) Batteries for Advanced Transportation Technologies (BATT) program annual review.
Energy Technology Data Exchange (ETDEWEB)
Bernnat, W.; Keinert, J.; Mattes, M. [Inst. for Nuclear Energy and Energy Systems, Univ. of Stuttgart, Stuttgart (Germany)
2004-03-01
For the calculation of neutron spectra in cold and super thermal sources scattering laws for a variety of liquid and solid cyrogenic materials were evaluated and prepared for use in deterministic and Monte Carlo transport calculations. For moderator materials like liquid and solid H{sub 2}O, liquid He, liquid D{sub 2}O, liquid and solid H{sub 2} and D{sub 2}, solid CH{sub 4} and structure materials such as Al, Bi, Pb, ZrHx, and graphite scattering law data and cross sections are available. The evaluated data were validated by comparison with measured cross sections and comparison of measured and calculated neutron spectra as far as available. Further applications are the calculation of production and transport and storing of ultra cold neutrons (UCN) in different UCN sources. The data structures of the evaluated data are prepared for the common S{sub N}-transport codes and the Monte Carlo Code MCNP. (orig.)
International Nuclear Information System (INIS)
For the calculation of neutron spectra in cold and super thermal sources scattering laws for a variety of liquid and solid cyrogenic materials were evaluated and prepared for use in deterministic and Monte Carlo transport calculations. For moderator materials like liquid and solid H2O, liquid He, liquid D2O, liquid and solid H2 and D2, solid CH4 and structure materials such as Al, Bi, Pb, ZrHx, and graphite scattering law data and cross sections are available. The evaluated data were validated by comparison with measured cross sections and comparison of measured and calculated neutron spectra as far as available. Further applications are the calculation of production and transport and storing of ultra cold neutrons (UCN) in different UCN sources. The data structures of the evaluated data are prepared for the common SN-transport codes and the Monte Carlo Code MCNP. (orig.)
Advanced physics of electron transport in semiconductors and nanostructures
Fischetti, Massimo V
2016-01-01
This textbook is aimed at second-year graduate students in Physics, Electrical Engineering, or Materials Science. It presents a rigorous introduction to electronic transport in solids, especially at the nanometer scale. Understanding electronic transport in solids requires some basic knowledge of Hamiltonian Classical Mechanics, Quantum Mechanics, Condensed Matter Theory, and Statistical Mechanics. Hence, this book discusses those sub-topics which are required to deal with electronic transport in a single, self-contained course. This will be useful for students who intend to work in academia or the nano/ micro-electronics industry. Further topics covered include: the theory of energy bands in crystals, of second quantization and elementary excitations in solids, of the dielectric properties of semiconductors with an emphasis on dielectric screening and coupled interfacial modes, of electron scattering with phonons, plasmons, electrons and photons, of the derivation of transport equations in semiconduc...