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)