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Sample records for advanced neutron transport

  1. A Complex-Geometry Validation Experiment for Advanced Neutron Transport Codes

    Energy Technology Data Exchange (ETDEWEB)

    David W. Nigg; Anthony W. LaPorta; Joseph W. Nielsen; James Parry; Mark D. DeHart; Samuel E. Bays; William F. Skerjanc

    2013-11-01

    The Idaho National Laboratory (INL) has initiated a focused effort to upgrade legacy computational reactor physics software tools and protocols used for support of core fuel management and experiment management in the Advanced Test Reactor (ATR) and its companion critical facility (ATRC) at the INL.. This will be accomplished through the introduction of modern high-fidelity computational software and protocols, with appropriate new Verification and Validation (V&V) protocols, over the next 12-18 months. Stochastic and deterministic transport theory based reactor physics codes and nuclear data packages that support this effort include MCNP5[1], SCALE/KENO6[2], HELIOS[3], SCALE/NEWT[2], and ATTILA[4]. Furthermore, a capability for sensitivity analysis and uncertainty quantification based on the TSUNAMI[5] system has also been implemented. Finally, we are also evaluating the Serpent[6] and MC21[7] codes, as additional verification tools in the near term as well as for possible applications to full three-dimensional Monte Carlo based fuel management modeling in the longer term. On the experimental side, several new benchmark-quality code validation measurements based on neutron activation spectrometry have been conducted using the ATRC. Results for the first four experiments, focused on neutron spectrum measurements within the Northwest Large In-Pile Tube (NW LIPT) and in the core fuel elements surrounding the NW LIPT and the diametrically opposite Southeast IPT have been reported [8,9]. A fifth, very recent, experiment focused on detailed measurements of the element-to-element core power distribution is summarized here and examples of the use of the measured data for validation of corresponding MCNP5, HELIOS, NEWT, and Serpent computational models using modern least-square adjustment methods are provided.

  2. Innovative and Advanced Coupled Neutron Transport and Thermal Hydraulic Method (Tool) for the Design, Analysis and Optimization of VHTR/NGNP Prismatic Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Rahnema, Farzad; Garimeela, Srinivas; Ougouag, Abderrafi; Zhang, Dingkang

    2013-11-29

    This project will develop a 3D, advanced coarse mesh transport method (COMET-Hex) for steady- state and transient analyses in advanced very high-temperature reactors (VHTRs). The project will lead to a coupled neutronics and thermal hydraulic (T/H) core simulation tool with fuel depletion capability. The computational tool will be developed in hexagonal geometry, based solely on transport theory without (spatial) homogenization in complicated 3D geometries. In addition to the hexagonal geometry extension, collaborators will concurrently develop three additional capabilities to increase the code’s versatility as an advanced and robust core simulator for VHTRs. First, the project team will develop and implement a depletion method within the core simulator. Second, the team will develop an elementary (proof-of-concept) 1D time-dependent transport method for efficient transient analyses. The third capability will be a thermal hydraulic method coupled to the neutronics transport module for VHTRs. Current advancements in reactor core design are pushing VHTRs toward greater core and fuel heterogeneity to pursue higher burn-ups, efficiently transmute used fuel, maximize energy production, and improve plant economics and safety. As a result, an accurate and efficient neutron transport, with capabilities to treat heterogeneous burnable poison effects, is highly desirable for predicting VHTR neutronics performance. This research project’s primary objective is to advance the state of the art for reactor analysis.

  3. 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

  4. 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.

  5. 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

  6. 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.

  7. 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.

  8. Advanced Neutron Spectrometer

    Science.gov (United States)

    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.

  9. 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.

  10. Coupled Neutron Transport for HZETRN

    Science.gov (United States)

    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.

  11. 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

  12. 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.

  13. 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.

  14. 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

  15. 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.

  16. (International Collaboration on Advanced Neutron Sources)

    Energy Technology Data Exchange (ETDEWEB)

    Hayter, J.B.

    1990-11-08

    The International Collaboration on Advanced Neutron Sources was started about a decade ago with the purpose of sharing information throughout the global neutron community. The collaboration has been extremely successful in optimizing the use of resources, and the discussions are open and detailed, with reasons for failure shared as well as reasons for success. Although the meetings have become increasingly oriented toward pulsed neutron sources, many of the neutron instrumentation techniques, such as the development of better monochromators, fast response detectors and various data analysis methods, are highly relevant to the Advanced Neutron Source (ANS). I presented one paper on the ANS, and another on the neutron optical polarizer design work which won a 1989 R D-100 Award. I also gained some valuable design ideas, in particular for the ANS hot source, in discussions with individual researchers from Canada, Western Europe, and Japan.

  17. 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

  18. 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

    2003-01-30

    This is the second quarterly progress report for Year-4 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 October 1, 2002 and December 30, 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, Task 4: Addition of a Pipe Rotation System. (b) New research project (Task 9b): ''Development of a Foam Generator/Viscometer for Elevated Pressure and Elevated Temperature (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) New Research project (Task 13): ''Study of Cuttings Transport with Foam under Elevated Pressure and Temperature Conditions''. (g) 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). (h) Activities towards technology transfer and developing contacts with Petroleum and service company members, and increasing the number of JIP members.

  19. 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.

  20. Considerations in the design of an improved transportable neutron spectrometer

    CERN Document Server

    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.

  1. 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

    2002-10-30

    This is the first quarterly progress report for Year-4 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 July 1, 2002 and Sept. 30, 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, Task 4: Addition of a Pipe Rotation System, (b) New Research project (Task 9b): ''Development of a Foam Generator/Viscometer for Elevated Pressure and Elevated Temperature (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), 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). (g) Activities towards technology transfer and developing contacts with Petroleum and service company members, and increasing the number of JIP members.

  2. ADVANCED CUTTINGS TRANSPORT STUDY

    Energy Technology Data Exchange (ETDEWEB)

    Ergun Kuru; Stefan Miska; Nicholas Takach; Kaveh Ashenayi; Gerald Kane; Len Volk; Mark Pickell; Evren Ozbayoglu; Barkim Demirdal; Paco Vieira; Affonso Lourenco

    1999-10-15

    This report includes a review of the progress made in ACTF Flow Loop development and research during 90 days pre-award period (May 15-July 14, 1999) and the following three months after the project approval date (July15-October 15, 1999) The report presents information on the following specific subjects; (a) Progress in Advanced Cuttings Transport Facility design and development, (b) Progress report on the research project ''Study of Flow of Synthetic Drilling Fluids Under Elevated Pressure and Temperature Conditions'', (c) Progress report on the research project ''Study of Cuttings Transport with Foam Under LPAT Conditions (Joint Project with TUDRP)'', (d) Progress report on the research project ''Study of Cuttings Transport with Aerated Muds Under LPAT Conditions (Joint Project with TUDRP)'', (e) Progress report on the research project ''Study of Foam Flow Behavior Under EPET Conditions'', (f) Progress report on the instrumentation tasks (Tasks 11 and 12) (g) Activities towards technology transfer and developing contacts with oil and service company members.

  3. 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.

  4. Neutron stars - cooling and transport

    CERN Document Server

    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.

  5. 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

  6. 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.

  7. 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

  8. 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)

  9. 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)

  10. 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)

  11. 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.

  12. 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.

  13. 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

    2003-07-30

    This Quarter has been divided between running experiments and the installation of the drill-pipe rotation system. In addition, valves and piping were relocated, and three viewports were installed. Detailed design work is proceeding on a system to elevate the drill-string section. Design of the first prototype version of a Foam Generator has been finalized, and fabrication is underway. This will be used to determine the relationship between surface roughness and ''slip'' of foams at solid boundaries. Additional cups and rotors are being machined with different surface roughness. Some experiments on cuttings transport with aerated fluids have been conducted at EPET. Theoretical modeling of cuttings transport with aerated fluids is proceeding. The development of theoretical models to predict frictional pressure losses of flowing foam is in progress. The new board design for instrumentation to measure cuttings concentration is now functioning with an acceptable noise level. The ultrasonic sensors are stable up to 190 F. Static tests with sand in an annulus indicate that the system is able to distinguish between different sand concentrations. Viscometer tests with foam, generated by the Dynamic Test Facility (DTF), are continuing.

  14. 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

  15. An Improved Neutron Transport Algorithm for HZETRN

    Science.gov (United States)

    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.

  16. ADVANCED CUTTINGS TRANSPORT STUDY

    Energy Technology Data Exchange (ETDEWEB)

    Stefan Miska; Nicholas Takach; Kaveh Ashenayi

    2004-01-31

    Final design of the mast was completed (Task 5). The mast is consisting of two welded plate girders, set next to each other, and spaced 14-inches apart. Fabrication of the boom will be completed in two parts solely for ease of transportation. The end pivot connection will be made through a single 2-inch diameter x 4 feet-8 inch long 316 SS bar. During installation, hard piping make-ups using Chiksan joints will connect the annular section and 4-inch return line to allow full movement of the mast from horizontal to vertical. Additionally, flexible hoses and piping will be installed to isolate both towers from piping loads and allow recycling operations respectively. Calibration of the prototype Foam Generator Cell has been completed and experiments are now being conducted. We were able to generate up to 95% quality foam. Work is currently underway to attach the Thermo-Haake RS300 viscometer and install a view port with a microscope to measure foam bubble size and bubble size distribution. Foam rheology tests (Task 13) were carried out to evaluate the rheological properties of the proposed foam formulation. After successful completion of the first foam test, two sets of rheological tests were conducted at different foam flow rates while keeping other parameters constant (100 psig, 70F, 80% quality). The results from these tests are generally in agreement with the previous foam tests done previously during Task 9. However, an unanticipated observation during these tests was that in both cases, the frictional pressure drop in 2 inch pipe was lower than that in the 3 inch and 4 inch pipes. We also conducted the first foam cuttings transport test during this quarter. Experiments on aerated fluids without cuttings have been completed in ACTF (Task 10). Gas and liquid were injected at different flow rates. Two different sets of experiments were carried out, where the only difference was the temperature. Another set of tests was performed, which covered a wide range of

  17. 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.

  18. 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

  19. 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

  20. 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)

  1. Advanced Transport Operating Systems Program

    Science.gov (United States)

    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.

  2. Detector for advanced neutron capture experiments at LANSCE

    Energy Technology Data Exchange (ETDEWEB)

    Ullmann, J. L. (John L.); Reifarth, R. (Rene); Haight, Robert C.; Hunt, L. F. (Lloyd F.); O' Donnell, J. M.; Bredeweg, T. A. (Todd A); Wilhelmy, J. B. (Jerry B.); Fowler, Malcolm M.; Vieira, D. J. (David J.); Wouters, J. M. (Jan Marc); Strottman, D.; Kaeppeler, F. (Franz K.); Heil, M.; Chamberlin, E. P. (Edwin P.)

    2002-01-01

    The Detector for Advanced Neutron Capture Experiments (DANCE) is a 159-element 4x barium fluoride array designed to study neutron capture on small quantities, 1 mg or less, of radioactive nuclides. It is being built on a 20 m neutron flight path which views the 'upper tier' water moderator at the Manuel J. Lujan Jr. Neutron Scattering Center at the Los Alamos Neutron Science Center. The detector design is based on Monte Carlo calculations which have suggested ways to minimize backgrounds due to neutron scattering events. A data acquisition system based on fast transient digitizers is bcing implemented

  3. 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.

  4. 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.)

  5. 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)

  6. 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

  7. 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)

  8. Research on advanced transportation systems

    Science.gov (United States)

    Nagai, Hirokazu; Hashimoto, Ryouhei; Nosaka, Masataka; Koyari, Yukio; Yamada, Yoshio; Noda, Keiichirou; Shinohara, Suetsugu; Itou, Tetsuichi; Etou, Takao; Kaneko, Yutaka

    1992-08-01

    An overview of the researches on advanced space transportation systems is presented. Conceptual study is conducted on fly back boosters with expendable upper stage rocket systems assuming a launch capacity of 30 tons and returning to the launch site by the boosters, and prospect of their feasibility is obtained. Reviews are conducted on subjects as follows: (1) trial production of 10 tons sub scale engines for the purpose of acquiring hardware data and picking up technical problems for full scale 100 tons thrust engines using hydrocarbon fuels; (2) development techniques for advanced liquid propulsion systems from the aspects of development schedule, cost; (3) review of conventional technologies, and common use of component; (4) oxidant switching propulsion systems focusing on feasibility of Liquefied Air Cycle Engine (LACE) and Compressed Air Cycle Engine (CACE); (5) present status of slosh hydrogen manufacturing, storage, and handling; (6) construction of small high speed dynamometer for promoting research on mini pump development; (7) hybrid solid boosters under research all over the world as low-cost and clean propulsion systems; and (8) high performance solid propellant for upper stage and lower stage propulsion systems.

  9. 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

  10. 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.

  11. Design of a transportable high efficiency fast neutron spectrometer

    Science.gov (United States)

    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.

  12. 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)

  13. 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.

  14. 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.

  15. DIANE: Advanced system for mobile neutron radiology

    Science.gov (United States)

    Dance, W. E.; Huriet, J. R.; Cluzeau, S.; Mast, H.-U.; Albisu, F.

    1989-04-01

    Development of a new neutron radiology system, DIANE, is underway which will provide a ten-fold improvement in image-acquisition speed over presently operating mobile systems, insuring greater inspection throughput for production applications. Based on a 10 12 n/s sealed-tube (D-T) neutron generator under development by Sodern, on LTV's neutron moderator/collimator and electronic imaging systems and on robotic and safety systems being developed by IABG and Sener, the DIANE concept is that of a complete facility for on-site neutron radiography or radioscopy. The LTV components, which provide film or electronic imaging, including digital processing of 12-bit images, have been demonstrated in three basic systems now operating with Kaman A-711 neutron generators, including one operating in IABG's facilities. Sodern has fabricated a prototype neutron generator tube, the TN 46, for emission of 10 11 n/s over 1000 to 1500 hours, at 250 kV and 2 mA in the ion beam.

  16. 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.

  17. The neutron texture diffractometer at the China Advanced Research Reactor

    Science.gov (United States)

    Li, Mei-Juan; Liu, Xiao-Long; Liu, Yun-Tao; Tian, Geng-Fang; Gao, Jian-Bo; Yu, Zhou-Xiang; Li, Yu-Qing; Wu, Li-Qi; Yang, Lin-Feng; Sun, Kai; Wang, Hong-Li; Santisteban, J. r.; Chen, Dong-Feng

    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)

  18. 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)

  19. 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.

  20. 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.)

  1. Enhancing the Detector for Advanced Neutron Capture Experiments

    OpenAIRE

    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...

  2. An Advanced Neutron Spectrometer for Future Manned Exploration Missions

    Science.gov (United States)

    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

  3. Optics for Advanced Neutron Imaging and Scattering

    Energy Technology Data Exchange (ETDEWEB)

    Moncton, David E. [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Khaykovich, Boris [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States)

    2016-03-30

    During the report period, we continued the work as outlined in the original proposal. We have analyzed potential optical designs of Wolter mirrors for the neutron-imaging instrument VENUS, which is under construction at SNS. In parallel, we have conducted the initial polarized imaging experiment at Helmholtz Zentrum, Berlin, one of very few of currently available polarized-imaging facilities worldwide.

  4. Transport coefficients in superfluid neutron stars

    CERN Document Server

    Tolos, Laura; Sarkar, Sreemoyee; Tarrus, Jaume

    2014-01-01

    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.

  5. 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.

  6. On generating neutron transport tables with the NJOY system

    Energy Technology Data Exchange (ETDEWEB)

    Caldeira, Alexandre D.; Claro, Luiz H., E-mail: alexdc@ieav.cta.br, E-mail: luizhenu@ieav.cta.br [Instituto de Estudos Avancados (IEAv), Sao Jose dos Campos, SP (Brazil)

    2013-07-01

    Incorrect values for the product of the average number of neutrons released per fission and the fission microscopic cross-section were detected in several energy groups of a neutron transport table generated with the most updated version of the NJOY system. It was verified that the problem persists when older versions of this system are utilized. Although this problem exists for, at least, ten years, it is still an open question. (author)

  7. 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

  8. Guideline of Monte Carlo calculation. Neutron/gamma ray transport simulation by Monte Carlo method

    CERN Document Server

    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.

  9. 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)

  10. Neutron transport study of a beam port based dynamic neutron radiography facility

    Science.gov (United States)

    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

  11. 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

  12. TRIPOLI-3: a neutron/photon Monte Carlo transport code

    Energy Technology Data Exchange (ETDEWEB)

    Nimal, J.C.; Vergnaud, T. [Commissariat a l' Energie Atomique, Gif-sur-Yvette (France). Service d' Etudes de Reacteurs et de Mathematiques Appliquees

    2001-07-01

    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.)

  13. Hardening neutron spectrum for advanced actinide transmutation experiments in the ATR.

    Science.gov (United States)

    Chang, G S; Ambrosek, R G

    2005-01-01

    The most effective method for transmuting long-lived isotopes contained in spent nuclear fuel into shorter-lived fission products is in a fast neutron spectrum reactor. In the absence of a fast test reactor in the United States, initial irradiation testing of candidate fuels can be performed in a thermal test reactor that has been modified to produce a test region with a hardened neutron spectrum. Such a test facility, with a spectrum similar but somewhat softer than that of the liquid-metal fast breeder reactor (LMFBR), has been constructed in the INEEL's Advanced Test Reactor (ATR). The radial fission power distribution of the actinide fuel pin, which is an important parameter in fission gas release modelling, needs to be accurately predicted and the hardened neutron spectrum in the ATR and the LMFBR fast neutron spectrum is compared. The comparison analyses in this study are performed using MCWO, a well-developed tool that couples the Monte Carlo transport code MCNP with the isotope depletion and build-up code ORIGEN-2. MCWO analysis yields time-dependent and neutron-spectrum-dependent minor actinide and Pu concentrations and detailed radial fission power profile calculations for a typical fast reactor (LMFBR) neutron spectrum and the hardened neutron spectrum test region in the ATR. The MCWO-calculated results indicate that the cadmium basket used in the advanced fuel test assembly in the ATR can effectively depress the linear heat generation rate in the experimental fuels and harden the neutron spectrum in the test region.

  14. BNL Activities in Advanced Neutron Source Development: Past and Present

    Energy Technology Data Exchange (ETDEWEB)

    Hastings, J.B.; Ludewig, H.; Montanez, P.; Todosow, M.; Smith, G.C.; Larese, J.Z.

    1998-06-14

    Brookhaven National Laboratory has been involved in advanced neutron sources almost from its inception in 1947. These efforts have mainly focused on steady state reactors beginning with the construction of the first research reactor for neutron beams, the Brookhaven Graphite Research Reactor. This was followed by the High Flux Beam Reactor that has served as the design standard for all the subsequent high flux reactors constructed worldwide. In parallel with the reactor developments BNL has focused on the construction and use of high energy proton accelerators. The first machine to operate over 1 GeV in the world was the Cosmotron. The machine that followed this, the AGS, is still operating and is the highest intensity proton machine in the world and has nucleated an international collaboration investigating liquid metal targets for next generation pulsed spallation sources. Early work using the Cosmotron focused on spallation product studies for both light and heavy elements into the several GeV proton energy region. These original studies are still important today. In this report we discuss the facilities and activities at BNL focused on advanced neutron sources. BNL is involved in the proton source for the Spallation Neutron source, spectrometer development at LANSCE, target studies using the AGS and state-of-the-art neutron detector development.

  15. BNL ACTIVITIES IN ADVANCED NEUTRON SOURCE DEVELOPMENT: PAST AND PRESENT

    Energy Technology Data Exchange (ETDEWEB)

    HASTINGS,J.B.; LUDEWIG,H.; MONTANEZ,P.; TODOSOW,M.; SMITH,G.C.; LARESE,J.Z.

    1998-06-14

    Brookhaven National Laboratory has been involved in advanced neutron sources almost from its inception in 1947. These efforts have mainly focused on steady state reactors beginning with the construction of the first research reactor for neutron beams, the Brookhaven Graphite Research Reactor. This was followed by the High Flux Beam Reactor that has served as the design standard for all the subsequent high flux reactors constructed worldwide. In parallel with the reactor developments BNL has focused on the construction and use of high energy proton accelerators. The first machine to operate over 1 GeV in the world was the Cosmotron. The machine that followed this, the AGS, is still operating and is the highest intensity proton machine in the world and has nucleated an international collaboration investigating liquid metal targets for next generation pulsed spallation sources. Early work using the Cosmotron focused on spallation product studies for both light and heavy elements into the several GeV proton energy region. These original studies are still important today. In the sections below the authors discuss the facilities and activities at BNL focused on advanced neutron sources. BNL is involved in the proton source for the Spallation Neutron source, spectrometer development at LANSCE, target studies using the AGS and state-of-the-art neutron detector development.

  16. Analytical three-dimensional neutron transport benchmarks for verification of nuclear engineering codes. Final report

    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

  17. Analytical three-dimensional neutron transport benchmarks for verification of nuclear engineering codes. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Ganapol, B.D.; Kornreich, D.E. [Univ. of Arizona, Tucson, AZ (United States). Dept. of Nuclear Engineering

    1997-07-01

    Because of the requirement of accountability and quality control in the scientific world, a demand for high-quality analytical benchmark calculations has arisen in the neutron transport community. The intent of these benchmarks is to provide a numerical standard to which production neutron transport codes may be compared in order to verify proper operation. The overall investigation as modified in the second year renewal application includes the following three primary tasks. Task 1 on two dimensional neutron transport is divided into (a) single medium searchlight problem (SLP) and (b) two-adjacent half-space SLP. Task 2 on three-dimensional neutron transport covers (a) point source in arbitrary geometry, (b) single medium SLP, and (c) two-adjacent half-space SLP. Task 3 on code verification, includes deterministic and probabilistic codes. The primary aim of the proposed investigation was to provide a suite of comprehensive two- and three-dimensional analytical benchmarks for neutron transport theory applications. This objective has been achieved. The suite of benchmarks in infinite media and the three-dimensional SLP are a relatively comprehensive set of one-group benchmarks for isotropically scattering media. Because of time and resource limitations, the extensions of the benchmarks to include multi-group and anisotropic scattering are not included here. Presently, however, enormous advances in the solution for the planar Green`s function in an anisotropically scattering medium have been made and will eventually be implemented in the two- and three-dimensional solutions considered under this grant. Of particular note in this work are the numerical results for the three-dimensional SLP, which have never before been presented. The results presented were made possible only because of the tremendous advances in computing power that have occurred during the past decade.

  18. 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)

  19. 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

  20. Thermal and transport properties of the neutron star inner crust

    CERN Document Server

    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.

  1. Thermophysical properties of saturated light and heavy water for Advanced Neutron Source applications

    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.

  2. Thermophysical properties of saturated light and heavy water for advanced neutron source applications

    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.

  3. 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.

  4. 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.

  5. 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.

  6. 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

  7. Recent advances in neutral particle transport methods and codes

    Energy Technology Data Exchange (ETDEWEB)

    Azmy, Y.Y.

    1996-06-01

    An overview of ORNL`s three-dimensional neutral particle transport code, TORT, is presented. Special features of the code that make it invaluable for large applications are summarized for the prospective user. Advanced capabilities currently under development and installation in the production release of TORT are discussed; they include: multitasking on Cray platforms running the UNICOS operating system; Adjacent cell Preconditioning acceleration scheme; and graphics codes for displaying computed quantities such as the flux. Further developments for TORT and its companion codes to enhance its present capabilities, as well as expand its range of applications are disucssed. Speculation on the next generation of neutron particle transport codes at ORNL, especially regarding unstructured grids and high order spatial approximations, are also mentioned.

  8. 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)

  9. 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.

  10. 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

  11. Neutron shielding evaluation for a small fuel transport case

    CERN Document Server

    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).

  12. 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

  13. Neutron imaging of ion transport in mesoporous carbon materials.

    Science.gov (United States)

    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

  14. 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)

  15. 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.

  16. Experimental characterization of the Advanced Liquid Hydrogen Cold Neutron Source spectrum of the NBSR reactor at the NIST Center for Neutron Research

    Science.gov (United States)

    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.

  17. 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

  18. Electron transport through nuclear pasta in magnetized neutron stars

    CERN Document Server

    Yakovlev, D G

    2015-01-01

    We present a simple model for electron transport in a possible layer of exotic nuclear clusters (in the so called nuclear pasta layer) between the crust and liquid core of a strongly magnetized neutron star. The electron transport there can be strongly anisotropic and gyrotropic. The anisotropy is produced by different electron effective collision frequencies along and across local symmetry axis in domains of exotic ordered nuclear clusters and by complicated effects of the magnetic field. We also calculate averaged kinetic coefficients in case local domains are freely oriented. Possible applications of the obtained results and open problems are outlined.

  19. 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)

  20. 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.

  1. Conceptual study of advanced PWR core design. Development of advanced PWR core neutronics analysis system

    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)

  2. Validation of multigroup neutron cross sections for the Advanced Neutron Source against the FOEHN critical experimental measurements

    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

  3. Global shielding analysis for the three-element core advanced neutron source reactor under normal operating conditions

    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.

  4. Beam-transport optimization for cold-neutron spectrometer

    Directory of Open Access Journals (Sweden)

    Nakajima Kenji

    2015-01-01

    Full Text Available We report the design of the beam-transport system (especially the vertical geometry for a cold-neutron disk-chopper spectrometer AMATERAS at J-PARC. Based on the elliptical shape, which is one of the most effective geometries for a ballistic mirror, the design was optimized to obtain, at the sample position, a neutron beam with high flux without serious degrading in divergence and spacial homogeneity within the boundary conditions required from actual spectrometer construction. The optimum focal point was examined. An ideal elliptical shape was modified to reduce its height without serious loss of transmission. The final result was adapted to the construction requirements of AMATERAS. Although the ideas studied in this paper are considered for the AMATERAS case, they can be useful also to other spectrometers in similar situations.

  5. 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

  6. 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.

  7. Optimized design of the nuclear fuel rod transport container used for non-destructive testing with neutron radiography

    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)

  8. Combining measurements and 3D neutron transport calculations. A powerful tool in detailed neutron dosimetry and damage analysis

    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

  9. Validation of multigroup neutron cross sections and calculational methods for the advanced neutron source against the FOEHN critical experiments measurements

    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%.

  10. Neutron Transport Models and Methods for HZETRN and Coupling to Low Energy Light Ion Transport

    Science.gov (United States)

    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.

  11. Subroutines to Simulate Fission Neutrons for Monte Carlo Transport Codes

    CERN Document Server

    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.

  12. 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

  13. Advancements in the development of a directional-position sensing fast neutron detector using acoustically tensioned metastable fluids

    Science.gov (United States)

    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

  14. Advancements in the development of a directional-position sensing fast neutron detector using acoustically tensioned metastable fluids

    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

  15. FOCUS, Neutron Transport System for Complex Geometry Reactor Core and Shielding Problems by Monte-Carlo

    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

  16. Advanced thermal-hydraulic and neutronic codes: current and future applications. Summary and conclusions

    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

  17. Proceedings of the workshop on advanced thermal-hydraulic and neutronic codes: current and future applications

    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

  18. Advanced transport systems analysis, modeling, and evaluation of performances

    CERN Document Server

    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...

  19. Establishment and Verification of MCNP Neutron Transport Model About Tianwan Nuclear Power Plant

    Institute of Scientific and Technical Information of China (English)

    ZHOU; Qi

    2012-01-01

    <正>In order to calculating the neutron flux in the surveillance box and reactor pressure vessel of the Tianwan NPP, we need to build up the neutron transport model by using the Monte Carlo code MCNP. The core of the NPP is very complicated for modeling so we put forward some assumptions that can simplify the neutron transport model. A lot of calculation works have been done to prove that the assumptions are right and suitable.

  20. Gamma transport and diffusion calculation capability coupled with neutron transport simulation in KARMA 1.2

    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

  1. 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)

  2. 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)

  3. 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

  4. 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.

  5. Proceedings of the fifteenth meeting of the international collaboration on advanced neutron sources (ICANS-XV). Advanced neutron sources towards the next century

    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.)

  6. Proceedings of the fifteenth meeting of the international collaboration on advanced neutron sources (ICANS-XV). Advanced neutron sources towards the next century

    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.)

  7. 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.

  8. Containment performance analyses for the Advanced Neutron Source Reactor at the Oak Ridge National Laboratory

    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.

  9. 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

  10. 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

  11. 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

  12. 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

  13. Subroutines to Simulate Fission Neutrons for Monte Carlo Transport Codes

    OpenAIRE

    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...

  14. Advanced Monte Carlo procedure for the IFMIF d-Li neutron source term based on evaluated cross section data

    CERN Document Server

    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.

  15. 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

  16. Improved Convergence Rate of Multi-Group Scattering Moment Tallies for Monte Carlo Neutron Transport Codes

    Science.gov (United States)

    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

  17. Recent advances in understanding hepatic drug transport

    Science.gov (United States)

    Stieger, Bruno; Hagenbuch, Bruno

    2016-01-01

    Cells need to strictly control their internal milieu, a function which is performed by the plasma membrane. Selective passage of molecules across the plasma membrane is controlled by transport proteins. As the liver is the central organ for drug metabolism, hepatocytes are equipped with numerous drug transporters expressed at the plasma membrane. Drug disposition includes absorption, distribution, metabolism, and elimination of a drug and hence multiple passages of drugs and their metabolites across membranes. Consequently, understanding the exact mechanisms of drug transporters is essential both in drug development and in drug therapy. While many drug transporters are expressed in hepatocytes, and some of them are well characterized, several transporters have only recently been identified as new drug transporters. Novel powerful tools to deorphanize (drug) transporters are being applied and show promising results. Although a large set of tools are available for studying transport in vitro and in isolated cells, tools for studying transport in living organisms, including humans, are evolving now and rely predominantly on imaging techniques, e.g. positron emission tomography. Imaging is an area which, certainly in the near future, will provide important insights into "transporters at work" in vivo. PMID:27781095

  18. 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

  19. 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.

  20. MCNP, a general Monte Carlo code for neutron and photon transport: a summary

    International Nuclear Information System (INIS)

    The general-purpose Monte Carlo code MCNP can be used for neutron, photon, or coupled neutron-photon transport, including the capability to calculate eigenvalues for critical systems. The code treats an arbitrary three-dimensional configuration of materials in geometric cells bounded by first- and second-degree surfaces and some special fourth-degree surfaces

  1. 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

  2. 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)

  3. 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.)

  4. 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

  5. Terrestrial neutron-induced soft errors in advanced memory devices

    CERN Document Server

    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

  6. Neutron production and time resolution of a new class moderator for pulsed neutron diffraction. Measurements and transport calculations

    International Nuclear Information System (INIS)

    Measurements of neutron pulse time-width and intensity have been carried out on grids of small moderators placed side by side and decoupled by cadmium strips; a moderator concept introduced by the authors through previous publications. Transport calculations are based on the standard reactor code DOT 3.5 with the ENDF-B IV nuclear data library. (orig.)

  7. PHISICS multi-group transport neutronic capabilities for RELAP5

    Energy Technology Data Exchange (ETDEWEB)

    Epiney, A.; Rabiti, C.; Alfonsi, A.; Wang, Y.; Cogliati, J.; Strydom, G. [Idaho National Laboratory (INL), 2525 N. Fremont Ave., Idaho Falls, ID 83402 (United States)

    2012-07-01

    PHISICS is a neutronic code system currently under development at INL. Its goal is to provide state of the art simulation capability to reactor designers. This paper reports on the effort of coupling this package to the thermal hydraulic system code RELAP5. This will enable full prismatic core and system modeling and the possibility to model coupled (thermal-hydraulics and neutronics) problems with more options for 3D neutron kinetics, compared to the existing diffusion theory neutron kinetics module in RELAP5 (NESTLE). The paper describes the capabilities of the coupling and illustrates them with a set of sample problems. (authors)

  8. Transport calculation of thermal and cold neutrons using NMTC/JAERI-MCNP4A code system

    Energy Technology Data Exchange (ETDEWEB)

    Iga, Kiminori [Kyushu Univ., Fukuoka (Japan); Takada, Hiroshi; Nagao, Tadashi

    1998-01-01

    In order to investigate the applicability of the NMTC/JAERI-MCNP4A code system to the neutronics design study in the neutron science research project of JAERI, transport calculations of thermal and cold neutrons are performed with the code system on a spallation neutron source composed of light water cooled tantalum target with a moderator and a reflector system. The following neutronic characteristics are studied in the calculation : the variation of the intensity of neutrons emitted from a light water moderator or a liquid hydrogen with/without the B{sub 4}C decoupler, which are installed to produce sharp pulse, and that dependent on the position of external source neutrons in the tantalum target. The calculated neutron energy spectra are reproduced well by the semi-empirical formula with the parameter values reliable in physical meanings. It is found to be necessary to employ proper importance sampling technique in the statistics. It is confirmed from this work that the NMTC/JAERI-MCNP4A code system is applicable to the neutronics design study of spallation neutron sources proposed for the neutron science research project. (author)

  9. VVER-440 Ex-Core Neutron Transport Calculations by MCNP-5 Code and Comparison with Experiment

    Energy Technology Data Exchange (ETDEWEB)

    Borodkin, Pavel; Khrennikov, Nikolay [Scientific and Engineering Centre for Nuclear and Radiation Safety (SEC NRS) Malaya Krasnoselskaya ul., 2/8, bld. 5, 107140 Moscow (Russian Federation)

    2008-07-01

    Ex-core neutron transport calculations are needed to evaluate radiation loading parameters (neutron fluence, fluence rate and spectra) on the in-vessel equipment, reactor pressure vessel (RPV) and support constructions of VVER type reactors. Due to these parameters are used for reactor equipment life-time assessment, neutron transport calculations should be carried out by precise and reliable calculation methods. In case of RPVs, especially, of first generation VVER-440s, the neutron fluence plays a key role in the prediction of RPV lifetime. Main part of VVER ex-core neutron transport calculations are performed by deterministic and Monte-Carlo methods. This paper deals with precise calculations of the Russian first generation VVER-440 by MCNP-5 code. The purpose of this work was an application of this code for expert calculations, verification of results by comparison with deterministic calculations and validation by neutron activation measured data. Deterministic discrete ordinates DORT code, widely used for RPV neutron dosimetry and many times tested by experiments, was used for comparison analyses. Ex-vessel neutron activation measurements at the VVER-440 NPP have provided space (in azimuth and height directions) and neutron energy (different activation reactions) distributions data for experimental (E) validation of calculated results. Calculational intercomparison (DORT vs. MCNP-5) and comparison with measured values (MCNP-5 and DORT vs. E) have shown agreement within 10-15% for different space points and reaction rates. The paper submits a discussion of results and makes conclusions about practice use of MCNP-5 code for ex-core neutron transport calculations in expert analysis. (authors)

  10. Neutron secondary-particle production cross sections and their incorporation into Monte-Carlo transport codes

    International Nuclear Information System (INIS)

    Realistic simulations of the passage of fast neutrons through tissue require a large quantity of cross-sectional data. What are needed are differential (in particle type, energy and angle) cross sections. A computer code is described which produces such spectra for neutrons above ∼14 MeV incident on light nuclei such as carbon and oxygen. Comparisons have been made with experimental measurements of double-differential secondary charged-particle production on carbon and oxygen at energies from 27 to 60 MeV; they indicate that the model is adequate in this energy range. In order to utilize fully the results of these calculations, they should be incorporated into a neutron transport code. This requires defining a generalized format for describing charged-particle production, putting the calculated results in this format, interfacing the neutron transport code with these data, and charged-particle transport. The design and development of such a program is described. 13 refs., 3 figs

  11. Cosmic ray heliospheric transport study with neutron monitor data

    Science.gov (United States)

    Ahluwalia, H. S.; Ygbuhay, R. C.; Modzelewska, R.; Dorman, L. I.; Alania, M. V.

    2015-10-01

    Determining transport coefficients for galactic cosmic ray (GCR) propagation in the turbulent interplanetary magnetic field (IMF) poses a fundamental challenge in modeling cosmic ray modulation processes. GCR scattering in the solar wind involves wave-particle interaction, the waves being Alfven waves which propagate along the ambient field (B). Empirical values at 1 AU are determined for the components of the diffusion tensor for GCR propagation in the heliosphere using neutron monitor (NM) data. At high rigidities, particle density gradients and mean free paths at 1 AU in B can only be computed from the solar diurnal anisotropy (SDA) represented by a vector A (components Ar, Aϕ, and Aθ) in a heliospherical polar coordinate system. Long-term changes in SDA components of NMs (with long track record and the median rigidity of response Rm ~ 20 GV) are used to compute yearly values of the transport coefficients for 1963-2013. We confirm the previously reported result that the product of the parallel (to B) mean free path (λ||) and radial density gradient (Gr) computed from NM data exhibits a weak Schwabe cycle (11y) but strong Hale magnetic cycle (22y) dependence. Its value is most depressed in solar activity minima for positive (p) polarity intervals (solar magnetic field in the Northern Hemisphere points outward from the Sun) when GCRs drift from the polar regions toward the helioequatorial plane and out along the heliospheric current sheet (HCS), setting up a symmetric gradient Gθs pointing away from HCS. Gr drives all SDA components and λ|| Gr contributes to the diffusive component (Ad) of the ecliptic plane anisotropy (A). GCR transport is commonly discussed in terms of an isotropic hard sphere scattering (also known as billiard-ball scattering) in the solar wind plasma. We use it with a flat HCS model and the Ahluwalia-Dorman master equations to compute the coefficients α (=λ⊥/λ∥) and ωτ (a measure of turbulence in the solar wind) and transport

  12. 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.

  13. Advanced in the neutron feedback ICF reactor concept

    International Nuclear Information System (INIS)

    Results are reviewed and updated from an earlier design study of a novel nuclear-pumped flashlamp laser (NP-FL) inertial fusion energy (IFE) power reactor based on the neutron feedback concept for IFE. This concept includes nuclear pumping of the laser flashlamp, a D-T seeded D-3He target and magnetic protection of the first wall of the reactor chamber coupled with direct conversion of deflected charged particles. Advantages include an increased overall plant efficiency due to improved energy coupling via neutron feedback, increased thermal-to-electric energy conversion efficiency, and lower neutron activation and waste. These factors are reflected in a driver energy of 5 MJ and a target gain of only 50 for a 53 % efficient 1000-MWe power plant operating at 6 Hz, novel components involved. However, they require further technological development. Consequently, the NP-FL plant appears to provide a very attractive 'second-generation' IFE reactor. (authors)

  14. Final LDRD report : advanced plastic scintillators for neutron detection.

    Energy Technology Data Exchange (ETDEWEB)

    Vance, Andrew L.; Mascarenhas, Nicholas; O' Bryan, Greg; Mrowka, Stanley

    2010-09-01

    This report summarizes the results of a one-year, feasibility-scale LDRD project that was conducted with the goal of developing new plastic scintillators capable of pulse shape discrimination (PSD) for neutron detection. Copolymers composed of matrix materials such as poly(methyl methacrylate) (PMMA) and blocks containing trans-stilbene (tSB) as the scintillator component were prepared and tested for gamma/neutron response. Block copolymer synthesis utilizing tSBMA proved unsuccessful so random copolymers containing up to 30% tSB were prepared. These copolymers were found to function as scintillators upon exposure to gamma radiation; however, they did not exhibit PSD when exposed to a neutron source. This project, while falling short of its ultimate goal, demonstrated the possible utility of single-component, undoped plastics as scintillators for applications that do not require PSD.

  15. Neutron interaction and their transport with bulk materials

    Energy Technology Data Exchange (ETDEWEB)

    Rani, Esther Kalpana, E-mail: esther.kalpanarani@gmail.com [Department of Physics JNT University, Nachupally, Karimnagar, Telangana, 500055 (India); Radhika, K., E-mail: radhikanit@gmail.com [Department of Humanities and Applied Sciences, Talla Padmavathi College of Engineering, Warangal, Telangana, 506004 (India)

    2015-05-15

    In the current paper an attempt was made to study and provide fundamental information about neutron interactions that are important to nuclear material measurements. The application of this study is explained about macroscopic interactions with bulk compound materials through a program in DEV C++ language which is done by enabling interaction of neutrons in nature. The output of the entire process depends upon the random number (i.e., incident neutron number), thickness of the material and mean free path as input parameters. Further the current study emphasizes on the usage of materials in shielding.

  16. Spin diffusive modes and thermal transport in neutron star crusts

    CERN Document Server

    Sedrakian, Armen

    2015-01-01

    In this contribution we first review a method for obtaining the collective modes of pair-correlated neutron matter as found in a neutron star inner crust. We discuss two classes of modes corresponding to density and spin perturbations with energy spectra $\\omega = \\omega_0 + \\alpha q^2$, where $\\omega_0 = 2\\Delta$ is the threshold frequency and $\\Delta$ is the gap in the neutron fluid spectrum. For characteristic values of Landau parameters in neutron star crusts the exitonic density modes have $\\alpha 0$ and they exist above $\\omega_0$ which implies that these modes are damped. As an application of these findings we compute the thermal conductivity due to spin diffusive modes and show that it scales as $T^{1/2} \\exp(-2\\omega_0/T)$ in the case where their two-by-two scattering cross-section is weakly dependent on temperature.

  17. Simulation Research on Neutron Guide System CNGC for China Advanced Research Reactor

    Institute of Scientific and Technical Information of China (English)

    WEI; Guo-hai; HAN; Song-bai; HE; Lin-feng; WANG; Yu; WANG; Hong-li; LIU; Yun-tao; CHEN; Dong-feng; ZHAO; Zhi-xiang

    2012-01-01

    <正>The out-pile section of the neutron guide CNGC at CARR (China Advanced Research Reactor) was designed by Monte Carlo simulation with VITESS. The out-pile section of CNGC will be spitted to CNGC-S and CNGC-N, the cold neutron imaging facility and small angle neutron scattering facility will be installed at the end of guides respectively. XRD patterns of Bi1-xLaxFe1-yScyO3 were shown in Fig. 1.

  18. Detailed flux calculations for the conceptual design of the Advanced Neutron Source Reactor

    International Nuclear Information System (INIS)

    A detailed MCNP model of the Advanced Neutron Source Reactor has been developed. All reactor components inside the reflector tank were included, and all components were highly segmented. Neutron and photon multigroup flux spectra have been calculated for each segment in the model, and thermal-to-fast neutron flux ratios were determined for each component segment. Axial profiles of the spectra are provided for all components of the reactor. Individual segment statistical uncertainties were limited wherever possible, and the group fluxes for all important reflector components have a standard deviation below 10%

  19. New generation of cryogen free advanced superconducting magnets for neutron scattering experiments

    Science.gov (United States)

    Kirichek, O.; Brown, J.; Adroja, D. T.; Manuel, P.; Kouzmenko, G.; Bewley, R. I.; Wotherspoon, R.

    2012-12-01

    Recent advances in superconducting technology and cryocooler refrigeration have resulted in a new generation of advanced superconducting magnets for neutron beam applications. These magnets have outstanding parameters such as high homogeneity and stability at highest magnetic fields possible, a reasonably small stray field, low neutron scattering background and larger exposure to neutron detectors. At the same time the pulse tube refrigeration technology provides a complete re-condensing regime which allows to minimise the requirements for cryogens without introducing additional noise and mechanical vibrations. The magnets can be used with dilution refrigerator insert which expands the temperature range from 20mK to 300K. Here we are going to present design, test results and the operational data of the 14T magnet for neutron diffraction and the 9T wide angle chopper magnet for neutron spectroscopy developed by Oxford Instruments in collaboration with ISIS neutron source. First scientific results obtained from the neutron scattering experiments with these magnets are also going to be discussed.

  20. Implementation and Initial Testing of Advanced Processing and Analysis Algorithms for Correlated Neutron Counting

    Energy Technology Data Exchange (ETDEWEB)

    Santi, Peter Angelo [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Cutler, Theresa Elizabeth [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Favalli, Andrea [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Koehler, Katrina Elizabeth [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Henzl, Vladimir [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Henzlova, Daniela [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Parker, Robert Francis [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Croft, Stephen [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-12-01

    In order to improve the accuracy and capabilities of neutron multiplicity counting, additional quantifiable information is needed in order to address the assumptions that are present in the point model. Extracting and utilizing higher order moments (Quads and Pents) from the neutron pulse train represents the most direct way of extracting additional information from the measurement data to allow for an improved determination of the physical properties of the item of interest. The extraction of higher order moments from a neutron pulse train required the development of advanced dead time correction algorithms which could correct for dead time effects in all of the measurement moments in a self-consistent manner. In addition, advanced analysis algorithms have been developed to address specific assumptions that are made within the current analysis model, namely that all neutrons are created at a single point within the item of interest, and that all neutrons that are produced within an item are created with the same energy distribution. This report will discuss the current status of implementation and initial testing of the advanced dead time correction and analysis algorithms that have been developed in an attempt to utilize higher order moments to improve the capabilities of correlated neutron measurement techniques.

  1. Advanced Transport Systems Showcased in La Rochelle

    OpenAIRE

    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...

  2. Discrete ordinates method for three-dimensional neutron transport equation based on unstructured-meshes

    Institute of Scientific and Technical Information of China (English)

    2008-01-01

    A discrete ordinates method for a threedimensional first-order neutron transport equation based on unstructured-meshes that avoids the singularity of the second-order neutron transport equation in void regions was derived.The finite element variation equation was obtained using the least-squares method.A three-dimensional transport calculation code was developed.Both the triangular-z and the tetrahedron elements were included.The numerical results of some benchmark problems demonstrated that this method can solve neutron transport problems in unstructuredmeshes very well.For most problems,the error of the eigenvalue and the angular flux is less than 0.3% and 3.0% respectively.

  3. Advanced technologies for intelligent transportation systems

    CERN Document Server

    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...

  4. Neutron radiography as a non-destructive method for diagnosing neutron converters for advanced thermal neutron detectors

    Science.gov (United States)

    Muraro, A.; Albani, G.; Perelli Cippo, E.; Croci, G.; Angella, G.; Birch, J.; Cazzaniga, C.; Caniello, R.; Dell'Era, F.; Ghezzi, F.; Grosso, G.; Hall-Wilton, R.; Höglund, C.; Hultman, L.; Schimdt, S.; Robinson, L.; Rebai, M.; Salvato, G.; Tresoldi, D.; Vasi, C.; Tardocchi, M.

    2016-03-01

    Due to the well-known problem of 3He shortage, a series of different thermal neutron detectors alternative to helium tubes are being developed, with the goal to find valid candidates for detection systems for the future spallation neutron sources such as the European Spallation Source (ESS). A possible 3He-free detector candidate is a charged particle detector equipped with a three dimensional neutron converter cathode (3D-C). The 3D-C currently under development is composed by a series of alumina (Al2O3) lamellas coated by 1 μ m of 10B enriched boron carbide (B4C). In order to obtain a good characterization in terms of detector efficiency and uniformity it is crucial to know the thickness, the uniformity and the atomic composition of the B4C neutron converter coating. In this work a non-destructive technique for the characterization of the lamellas that will compose the 3D-C was performed using neutron radiography. The results of these measurements show that the lamellas that will be used have coating uniformity suitable for detector applications. This technique (compared with SEM, EDX, ERDA, XPS) has the advantage of being global (i.e. non point-like) and non-destructive, thus it is suitable as a check method for mass production of the 3D-C elements.

  5. Neutron radiography as a non-destructive method for diagnosing neutron converters for advanced thermal neutron detectors

    International Nuclear Information System (INIS)

    Due to the well-known problem of 3He shortage, a series of different thermal neutron detectors alternative to helium tubes are being developed, with the goal to find valid candidates for detection systems for the future spallation neutron sources such as the European Spallation Source (ESS). A possible 3He-free detector candidate is a charged particle detector equipped with a three dimensional neutron converter cathode (3D-C). The 3D-C currently under development is composed by a series of alumina (Al2O3) lamellas coated by 1 μ m of 10B enriched boron carbide (B4C). In order to obtain a good characterization in terms of detector efficiency and uniformity it is crucial to know the thickness, the uniformity and the atomic composition of the B4C neutron converter coating. In this work a non-destructive technique for the characterization of the lamellas that will compose the 3D-C was performed using neutron radiography. The results of these measurements show that the lamellas that will be used have coating uniformity suitable for detector applications. This technique (compared with SEM, EDX, ERDA, XPS) has the advantage of being global (i.e. non point-like) and non-destructive, thus it is suitable as a check method for mass production of the 3D-C elements

  6. The impact of emerging technologies on an advanced supersonic transport

    Science.gov (United States)

    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.

  7. Snow shielding factors for cosmogenic nuclide dating inferred from Monte Carlo neutron transport simulations

    Science.gov (United States)

    Zweck, Christopher; Zreda, Marek; Desilets, Darin

    2013-10-01

    Conventional formulations of changes in cosmogenic nuclide production rates with snow cover are based on a mass-shielding approach, which neglects the role of neutron moderation by hydrogen. This approach can produce erroneous correction factors and add to the uncertainty of the calculated cosmogenic exposure ages. We use a Monte Carlo particle transport model to simulate fluxes of secondary cosmic-ray neutrons near the surface of the Earth and vary surface snow depth to show changes in neutron fluxes above rock or soil surface. To correspond with shielding factors for spallation and low-energy neutron capture, neutron fluxes are partitioned into high-energy, epithermal and thermal components. The results suggest that high-energy neutrons are attenuated by snow cover at a significantly higher rate (shorter attenuation length) than indicated by the commonly-used mass-shielding formulation. As thermal and epithermal neutrons derive from the moderation of high-energy neutrons, the presence of a strong moderator such as hydrogen in snow increases the thermal neutron flux both within the snow layer and above it. This means that low-energy production rates are affected by snow cover in a manner inconsistent with the mass-shielding approach and those formulations cannot be used to compute snow correction factors for nuclides produced by thermal neutrons. Additionally, as above-ground low-energy neutron fluxes vary with snow cover as a result of reduced diffusion from the ground, low-energy neutron fluxes are affected by snow even if the snow is at some distance from the site where measurements are made.

  8. Recent advances in mass transport in materials

    CERN Document Server

    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

  9. In situ studies of mass transport in liquid alloys by means of neutron radiography.

    Science.gov (United States)

    Kargl, F; Engelhardt, M; Yang, F; Weis, H; Schmakat, P; Schillinger, B; Griesche, A; Meyer, A

    2011-06-29

    When in situ techniques became available in recent years this led to a breakthrough in accurately determining diffusion coefficients for liquid alloys. Here we discuss how neutron radiography can be used to measure chemical diffusion in a ternary AlCuAg alloy. Neutron radiography hereby gives complementary information to x-ray radiography used for measuring chemical diffusion and to quasielastic neutron scattering used mainly for determining self-diffusion. A novel Al(2)O(3) based furnace that enables one to study diffusion processes by means of neutron radiography is discussed. A chemical diffusion coefficient of Ag against Al around the eutectic composition Al(68.6)Cu(13.8)Ag(17.6) at.% was obtained. It is demonstrated that the in situ technique of neutron radiography is a powerful means to study mass transport properties in situ in binary and ternary alloys that show poor x-ray contrast. PMID:21654050

  10. Neutron cross-section probability tables in TRIPOLI-3 Monte Carlo transport code

    Energy Technology Data Exchange (ETDEWEB)

    Zheng, S.H.; Vergnaud, T.; Nimal, J.C. [Commissariat a l`Energie Atomique, Gif-sur-Yvette (France). Lab. d`Etudes de Protection et de Probabilite

    1998-03-01

    Neutron transport calculations need an accurate treatment of cross sections. Two methods (multi-group and pointwise) are usually used. A third one, the probability table (PT) method, has been developed to produce a set of cross-section libraries, well adapted to describe the neutron interaction in the unresolved resonance energy range. Its advantage is to present properly the neutron cross-section fluctuation within a given energy group, allowing correct calculation of the self-shielding effect. Also, this PT cross-section representation is suitable for simulation of neutron propagation by the Monte Carlo method. The implementation of PTs in the TRIPOLI-3 three-dimensional general Monte Carlo transport code, developed at Commissariat a l`Energie Atomique, and several validation calculations are presented. The PT method is proved to be valid not only in the unresolved resonance range but also in all the other energy ranges.

  11. Advances in passive neutron instruments for safeguards use

    International Nuclear Information System (INIS)

    Passive neutron and other nondestructive assay techniques have been used extensively by the International Atomic Energy Agency to verify plutonium metal, powder, mixed oxide, pellets, rods, assemblies, scrap, and liquids. Normally, the coincidence counting rate is used to measure the 240Pu-effective mass and gamma-ray spectrometry or mass spectrometry is used to verify the plutonium isotopic ratios. During the past few years, the passive neutron detectors have been installed in plants and operated in the unattended/continuous mode. These radiation data with time continuity have made it possible to use the totals counting rate to monitor the movement of nuclear material. Monte Carlo computer codes have been used to optimize the detector designs for specific applications. The inventory sample counter (INVS-III) has been designed to have a higher efficiency (43%) and a larger uniform counting volume than the original INVS. Data analyses techniques have been developed, including the ''known alpha'' and ''known multiplication'' methods that depend on the sample. For scrap and other impure or poorly characterized samples, we have developed multiplicity counting, initially implemented in the plutonium scrap multiplicity counter. For large waste containers such as 200-L drums, we have developed the add-a-source technique to give accurate corrections for the waste-matrix materials. This paper summarizes recent developments in the design and application of passive neutron assay systems

  12. Advances in passive neutron instruments for safeguards use

    Energy Technology Data Exchange (ETDEWEB)

    Menlove, H.O.; Krick, M.S.; Langner, D.G.; Miller, M.C.; Stewart, J.E.

    1994-02-01

    Passive neutron and other nondestructive assay techniques have been used extensively by the International Atomic Energy Agency to verify plutonium metal, powder, mixed oxide, pellets, rods, assemblies, scrap, and liquids. Normally, the coincidence counting rate is used to measure the {sup 240}Pu-effective mass and gamma-ray spectrometry or mass spectrometry is used to verify the plutonium isotopic ratios. During the past few years, the passive neutron detectors have been installed in plants and operated in the unattended/continuous mode. These radiation data with time continuity have made it possible to use the totals counting rate to monitor the movement of nuclear material. Monte Carlo computer codes have been used to optimize the detector designs for specific applications. The inventory sample counter (INVS-III) has been designed to have a higher efficiency (43%) and a larger uniform counting volume than the original INVS. Data analyses techniques have been developed, including the ``known alpha`` and ``known multiplication`` methods that depend on the sample. For scrap and other impure or poorly characterized samples, we have developed multiplicity counting, initially implemented in the plutonium scrap multiplicity counter. For large waste containers such as 200-L drums, we have developed the add-a-source technique to give accurate corrections for the waste-matrix materials. This paper summarizes recent developments in the design and application of passive neutron assay systems.

  13. Recent advances and future trends in neutron resonance spectroscopy

    International Nuclear Information System (INIS)

    Neutron resonance spectroscopy contributes primarily to two areas of nuclear physics: 1.) in medium weight and heavy nuclei with a high level density it tests their statistical properties, and 2.) in nuclei with a sufficiently low level density, i.e. light nuclei (A ≤ 50) and nuclei around /sup 208/Pb, it investigates nuclear structure at several MeV excitation energy. In the first field, recent years have seen growing knowledge and understanding of nuclear level densities and their spin and parity dependence. Several questions basic to the statistical properties of nuclei, although extensively studied in the past, are still open: the statistical distribution of partial widths; possible narrow energy variations of the average partial widths; and correlations between partial widths for different reaction channels. The major progress has occurred and will continue to take place in the field of light nuclei: improved resolution of neutron time-of-flight spectrometers yields detailed resonance data over an extended energy range, and model calculations become possible which will allow detailed comparison to experimental data. The main subjects of interest are the distributions of neutron, as well as radiative strengths and their interpretation in terms of nuclear structure

  14. Recent advances and future trends in neutron resonance spectroscopy

    International Nuclear Information System (INIS)

    Neutron resonance spectroscopy contributes primarily to two areas of nuclear physics: In medium weight and heavy nuclei with a high level density it tests their statistical properties; in nuclei with a sufficiently low level density, i.e. light nuclei (A 208Pb, it investigates nuclear structure at several MeV excitation energy. In the first field, recent years have seen growing knowledge and understanding of nuclear level densities and their spin and parity dependence. Several questions basic to the statistical properties of nuclei, although extensively studied in the past, are still open: the statistical distribution of partial widths; possible narrow energy variations of the average partial widths; and correlations between partial widths for different reaction channels. The major progress has occured and will continue to take place in the field of light nuclei: Improved resolution of neutron time-of-flight spectrometers yields detailed resonance data over an extended energy range, and model calculations become possible which will allow detailed comparison to experimental data. The main subjects of interest are the distributions of neutron- as well as radiative strengths and their interpretation in terms of nuclear structure. (author)

  15. Computational benchmarking of fast neutron transport throughout large water thicknesses; Benchmark theorique du transport de neutrons rapides a travers de larges epaisseurs d`eau

    Energy Technology Data Exchange (ETDEWEB)

    Risch, P.; Dekens, O.; Ait Abderrahim, H. [SCK-CEN, Fuel Research Department, (Belgium); Wouters, R. de [Tractebel, Energy Engineering, (Belgium)

    1997-10-01

    Neutron dosimetry experiments seem to point our difficulties in the treatment of large water thickness like those encountered between the core baffle and the pressure vessel. This paper describes the theoretical benchmark undertaken by EDF, SCK/CEN and TRACTEBEL ENERGY ENGINEERING, concerning the transport of fast neutrons throughout a one meter cube of water, located after a U-235 fission sources plate. The results showed no major discrepancies between the calculations up to 50 cm from the source, accepting that a P3 development of the Legendre polynomials is necessary for the Sn calculations. The main differences occurred after 50 cm, reaching 20 % at the end of the water cube. This results lead us to consider an experimental benchmark, dedicated to the problem of fast neutron deep penetration in water, which has been launched at SCK/CEN. (authors). 7 refs.

  16. 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.

  17. ORNL contributions to the Advanced Neutron Source (ANS) Project for October 1986-March 1987

    International Nuclear Information System (INIS)

    The Advanced Neutron Source (ANS) Facility - formerly called the Center for Neutron Research - will provide the world's best facilities for the study of neutron scattering. The ANS high power density reactor will be fueled with uranium silicide and cooled, moderated, and reflected by D2O. Peak neutron fluxes in the reflector are expected to be 5 to 10 x 1019 neutrons per square meter with a power level between 270 MW and 300 MW. This report describes the status of technical work at ORNL on the ANS Project during the first half of FY 1987. The scope of this report includes Research and Development Tasks; Safety Tasks; Conceptual Design Tasks; and Project Support. The last two areas were only initiated as separate activities during this reporting period. Technical highlights include a better understanding of the relationship among neutron flux, core power, and core volume; preconceptual design work on a cold source for use in a very high gamma and neutron flux environment; identification of the major applicable safety rules and guidelines; and establishment of initial functional objectives for the containment structure

  18. Computational Transport Modeling of High-Energy Neutrons Found in the Space Environment

    Science.gov (United States)

    Cox, Brad; Theriot, Corey A.; Rohde, Larry H.; Wu, Honglu

    2012-01-01

    The high charge and high energy (HZE) particle radiation environment in space interacts with spacecraft materials and the human body to create a population of neutrons encompassing a broad kinetic energy spectrum. As an HZE ion penetrates matter, there is an increasing chance of fragmentation as penetration depth increases. When an ion fragments, secondary neutrons are released with velocities up to that of the primary ion, giving some neutrons very long penetration ranges. These secondary neutrons have a high relative biological effectiveness, are difficult to effectively shield, and can cause more biological damage than the primary ions in some scenarios. Ground-based irradiation experiments that simulate the space radiation environment must account for this spectrum of neutrons. Using the Particle and Heavy Ion Transport Code System (PHITS), it is possible to simulate a neutron environment that is characteristic of that found in spaceflight. Considering neutron dosimetry, the focus lies on the broad spectrum of recoil protons that are produced in biological targets. In a biological target, dose at a certain penetration depth is primarily dependent upon recoil proton tracks. The PHITS code can be used to simulate a broad-energy neutron spectrum traversing biological targets, and it account for the recoil particle population. This project focuses on modeling a neutron beamline irradiation scenario for determining dose at increasing depth in water targets. Energy-deposition events and particle fluence can be simulated by establishing cross-sectional scoring routines at different depths in a target. This type of model is useful for correlating theoretical data with actual beamline radiobiology experiments. Other work exposed human fibroblast cells to a high-energy neutron source to study micronuclei induction in cells at increasing depth behind water shielding. Those findings provide supporting data describing dose vs. depth across a water-equivalent medium. This

  19. Modeling heat generation and flow in the Advanced Neutron Source Corrosion Test Loop specimen

    International Nuclear Information System (INIS)

    A finite difference computer code HEATING5 was used to model heat generation and flow in a typical experiment envisioned for the Advanced Neutron Source Corrosion Test Loop. The electrical resistivity and thermal conductivity of the test specimen were allowed to vary with local temperature, and the corrosion layer thickness was assigned along the length of the specimen in the manner predicted by the Griess Correlation. The computer solved the two-dimensional transport problem for a given total power dissipated in the specimen and stipulated coolant temperatures and water-side heat-transfer coefficients. The computed specimen temperatures were compared with those calculated on the basis of approximate analytical equations involving the total power dissipation and the assignment of the physical properties based on temperatures at single axial points on the specimen. The comparisons indicate that when temperature variations are large along the axis of the specimen, the variation in local heat flux should not be overlooked when using approximate equations or models. The approximate equations are most accurate near the center of the specimen where the heat flux remains closest to the average value, and in that region the calculated quantities agree closely with the results of the computer code. 4 figs., 1 tab

  20. Improved Algorithms and Coupled Neutron-Photon Transport for Auto-Importance Sampling Method

    CERN Document Server

    Wang, Xin; Qiu, Rui; Li, Chun-Yan; Liang, Man-Chun; Zhang, Hui; Li, Jun-Li

    2016-01-01

    Auto-Importance Sampling (AIS) method is a Monte Carlo variance reduction technique proposed by Tsinghua University for deep penetration problem, which can improve computational efficiency significantly without pre-calculations for importance distribution. However AIS method is only validated with several basic deep penetration problems of simple geometries and cannot be used for coupled neutron-photon transport. This paper firstly presented the latest algorithm improvements for AIS method including particle transport, fictitious particles creation and adjustment, fictitious surface geometry, random number allocation and calculation of estimated relative error, which made AIS method applicable to complicated deep penetration problem. Then, a coupled Neutron-Photon Auto-Importance Sampling (NP-AIS) method was proposed to apply AIS method with the improved algorithms in coupled neutron-photon Monte Carlo transport. Finally, the NUREG/CR-6115 PWR benchmark model was calculated with the method of geometry splitti...

  1. Least-squares finite element discretizations of neutron transport equations in 3 dimensions

    Energy Technology Data Exchange (ETDEWEB)

    Manteuffel, T.A [Univ. of Colorado, Boulder, CO (United States); Ressel, K.J. [Interdisciplinary Project Center for Supercomputing, Zurich (Switzerland); Starkes, G. [Universtaet Karlsruhe (Germany)

    1996-12-31

    The least-squares finite element framework to the neutron transport equation introduced in is based on the minimization of a least-squares functional applied to the properly scaled neutron transport equation. Here we report on some practical aspects of this approach for neutron transport calculations in three space dimensions. The systems of partial differential equations resulting from a P{sub 1} and P{sub 2} approximation of the angular dependence are derived. In the diffusive limit, the system is essentially a Poisson equation for zeroth moment and has a divergence structure for the set of moments of order 1. One of the key features of the least-squares approach is that it produces a posteriori error bounds. We report on the numerical results obtained for the minimum of the least-squares functional augmented by an additional boundary term using trilinear finite elements on a uniform tesselation into cubes.

  2. Assessment of the impact of advanced air-transport technology

    Science.gov (United States)

    Maxwell, R. L.; Dickinson, L. V., Jr.

    1981-01-01

    The long term prospects for commercial supersonic transportation appear attractive enough to keep supersonic research active and reasonably healthy. On the other hand, the uncertainties surrounding an advanced supersonic transport, (AST) specifically fuel price, fuel availability and noise, are too significant to warrant an accelerated research and development program until they are better resolved. It is estimated that an AST could capture about $50 billion (1979 dollars) of the potential $150 billion in sales up to the year 2010.

  3. Advances in thermal hydraulic and neutronic simulation for reactor analysis and safety

    Energy Technology Data Exchange (ETDEWEB)

    Tentner, A.M.; Blomquist, R.N.; Canfield, T.R.; Ewing, T.F.; Garner, P.L.; Gelbard, E.M.; Gross, K.C.; Minkoff, M.; Valentin, R.A.

    1993-03-01

    This paper describes several large-scale computational models developed at Argonne National Laboratory for the simulation and analysis of thermal-hydraulic and neutronic events in nuclear reactors and nuclear power plants. The impact of advanced parallel computing technologies on these computational models is emphasized.

  4. An examination of the elastic structural response of the Advanced Neutron Source fuel plates

    Energy Technology Data Exchange (ETDEWEB)

    Swinson, W.F.; Luttrell, C.R.; Yahr, G.T.

    1994-09-01

    Procedures for evaluating the elastic structural response of the Advanced Neutron Source (ANS) fuel plates to coolant flow and to temperature variations are presented in this report. Calculations are made that predict the maximum deflection and the maximum stress for a representative plate from the upper and from the lower fuel elements.

  5. Detecting binary neutron star systems with spin in advanced gravitational-wave detectors

    CERN Document Server

    Brown, Duncan A; Lundgren, Andrew; Nitz, Alexander H

    2012-01-01

    The detection of gravitational waves from binary neutron stars is a major goal of the gravitational-wave observatories Advanced LIGO and Advanced Virgo. Previous searches for binary neutron stars with LIGO and Virgo neglected the component stars' angular momentum (spin). We demonstrate that neglecting spin in matched-filter searches causes advanced detectors to lose more than 3% of the possible signal-to-noise ratio for 59% (6%) of sources, assuming that neutron star dimensionless spins, $cJ/GM^2$, are uniformly distributed with magnitudes between 0 and 0.4 (0.05) and that the neutron stars have isotropically distributed spin orientations. We present a new method of constructing filter banks for advanced-detector searches, which can create template banks of signals with non-zero spins that are (anti-)aligned with the orbital angular momentum. We show that this search loses more than 3% of the maximium signal-to-noise for only 9% (0.2%) of BNS sources with dimensionless spins between 0 and 0.4 (0.05) and isotr...

  6. Monte Carlo simulations to advance characterisation of landmines by pulsed fast/thermal neutron analysis

    NARCIS (Netherlands)

    Maucec, M.; Rigollet, C.

    2004-01-01

    The performance of a detection system based on the pulsed fast/thermal neutron analysis technique was assessed using Monte Carlo simulations. The aim was to develop and implement simulation methods, to support and advance the data analysis techniques of the characteristic gamma-ray spectra, potentia

  7. Flexible polyvinyl chloride neutron guides for transporting ultracold and very cold neutrons

    Energy Technology Data Exchange (ETDEWEB)

    Arzumanov, S. S., E-mail: sarzumanov@yandex.ru; Bondarenko, L. N. [Russian Research Center Kurchatov Institute (Russian Federation); Geltenbort, P. [Institut Laue-Langevin (France); Morozov, V. I. [Russian Research Center Kurchatov Institute (Russian Federation); Nesvizhevsky, V. V. [Institut Laue-Langevin (France); Panin, Yu. N.; Strepetov, A. N.; Chuvilin, D. Yu. [Russian Research Center Kurchatov Institute (Russian Federation)

    2011-12-15

    The transmission of ultracold neutrons (UCNs) through flexible polyvinyl chloride (PVC) tubes with lengths of up to 3 m and an internal diameter of 6-8 mm has been studied. High UCN transmission is found even for arbitrarily bent tubes (single bend, double bend, triple bend, figure eight, etc.). The transmission can be improved significantly by coating the inner surface of the tube with a thin layer of liquid fluorine polymer. The prospects of these neutron guides in fundamental and applied research are discussed.

  8. Flexible polyvinyl chloride neutron guides for transporting ultracold and very cold neutrons

    International Nuclear Information System (INIS)

    The transmission of ultracold neutrons (UCNs) through flexible polyvinyl chloride (PVC) tubes with lengths of up to 3 m and an internal diameter of 6–8 mm has been studied. High UCN transmission is found even for arbitrarily bent tubes (single bend, double bend, triple bend, figure eight, etc.). The transmission can be improved significantly by coating the inner surface of the tube with a thin layer of liquid fluorine polymer. The prospects of these neutron guides in fundamental and applied research are discussed.

  9. RADTRAN 4.0: Advanced computer code for transportation risk assessment

    International Nuclear Information System (INIS)

    RADTRAN 4.0 is a computer code for transportation risk assessment developed by Sandia National Laboratories for the US Department of Energy. While retaining the most useful and time-proven features of its predecessors, RADTRAN 4.0 incorporates significant advances over the earlier versions. The most useful new features are: improved route-specific analysis capability, internal radionuclide data library, improved logic for analysis of multiple-radionuclide packages such as spent fuel, separate treatment of gamma and neutron components of Transport Index (TI), and increased number of accident-severity categories. In this paper, each of these features will be described, and, where appropriate, potential applications will be discussed. 11 refs

  10. Neutron Transport Methods for Accelerator-Driven Systems

    Energy Technology Data Exchange (ETDEWEB)

    Nicholas Tsoulfanidis; Elmer Lewis

    2005-02-09

    The objective of this project has been to develop computational methods that will enable more effective analysis of Accelerator Driven Systems (ADS). The work is centered at the University of Missouri at Rolla, with a subcontract at Northwestern University, and close cooperation with the Nuclear Engineering Division at Argonne National Laboratory. The work has fallen into three categories. First, the treatment of the source for neutrons originating from the spallation target which drives the neutronics calculations of the ADS. Second, the generalization of the nodal variational method to treat the R-Z geometry configurations frequently needed for scoping calculations in Accelerator Driven Systems. Third, the treatment of void regions within variational nodal methods as needed to treat the accelerator beam tube.

  11. Neutron absorber plate and radioactive material transportation cask

    International Nuclear Information System (INIS)

    Aluminum alloy flame-coating layers are formed at the outer surface of a neutron absorber plate in order to prevent corrosion due to potential difference. However, pin holes of micron order are sometimes formed on the flame-coating membranes, which are hard to be found by usual inspection. Then, ferrous flame-coating membranes are formed at the outer surface of boron carbide and aluminum alloy flame-coating membranes are formed at the outer surface thereof. The outer surface of a boron carbide plate is coated with the ferrous flame-coating membranes instead of being coated with an external plate made of neutron cells, and an aluminum alloy flame-coating membranes or mixed flame-coating layers of aluminum oxide and titania are coated thereover in order to prevent rusts. Whether the pin holes are present or not can be confirmed easily by a ferroxyl test. If there are pin holes, flame-coating is applied again to form complete membranes. Then, since it is no more necessary to fix a neutron absorbing cell at the outer surface of a fuel cell by means of welding, production cost can be reduced. (N.H.)

  12. Measuring neutron star tidal deformability with Advanced LIGO: a Bayesian analysis of neutron star - black hole binary observations

    CERN Document Server

    Kumar, Prayush; Pfeiffer, Harald P

    2016-01-01

    The discovery of gravitational waves (GW) by Advanced LIGO has ushered us into an era of observational GW astrophysics. Compact binaries remain the primary target sources for LIGO, of which neutron star-black hole (NSBH) binaries form an important subset. GWs from NSBH sources carry signatures of (a) the tidal distortion of the neutron star by its companion black hole during inspiral, and (b) its potential tidal disruption near merger. In this paper, we present a Bayesian study of the measurability of neutron star tidal deformability $\\Lambda_\\mathrm{NS}\\propto (R/M)^{5}$ using observation(s) of inspiral-merger GW signals from disruptive NSBH coalescences, taking into account the crucial effect of black hole spins. First, we find that if non-tidal templates are used to estimate source parameters for an NSBH signal, the bias introduced in the estimation of non-tidal physical parameters will only be significant for loud signals with signal-to-noise ratios greater than $30$. For similarly loud signals, we also f...

  13. Advances in Atomic Data for Neutron-Capture Elements

    CERN Document Server

    Sterling, N C; Esteves, D A; Stancil, P C; Kilcoyne, A L D; Bilodeau, R C; Aguilar, A

    2011-01-01

    Neutron(n)-capture elements (atomic number Z>30), which can be produced in planetary nebula (PN) progenitor stars via s-process nucleosynthesis, have been detected in nearly 100 PNe. This demonstrates that nebular spectroscopy is a potentially powerful tool for studying the production and chemical evolution of trans-iron elements. However, significant challenges must be addressed before this goal can be achieved. One of the most substantial hurdles is the lack of atomic data for n-capture elements, particularly that needed to solve for their ionization equilibrium (and hence to convert ionic abundances to elemental abundances). To address this need, we have computed photoionization cross sections and radiative and dielectronic recombination rate coefficients for the first six ions of Se and Kr. The calculations were benchmarked against experimental photoionization cross section measurements. In addition, we computed charge transfer (CT) rate coefficients for ions of six n-capture elements. These efforts will ...

  14. Description of a neutron field perturbed by a probe using coupled Monte Carlo and discrete ordinates radiation transport calculations

    International Nuclear Information System (INIS)

    This work concerns calculation of a neutron response, caused by a neutron field perturbed by materials surrounding the source or the detector. Solution of a problem is obtained using coupling of the Monte Carlo radiation transport computation for the perturbed region and the discrete ordinates transport computation for the unperturbed system. (author). 62 refs

  15. Quantifying moisture transport in cementitious materials using neutron radiography

    Science.gov (United States)

    Lucero, Catherine L.

    A portion of the concrete pavements in the US have recently been observed to have premature joint deterioration. This damage is caused in part by the ingress of fluids, like water, salt water, or deicing salts. The ingress of these fluids can damage concrete when they freeze and expand or can react with the cementitious matrix causing damage. To determine the quality of concrete for assessing potential service life it is often necessary to measure the rate of fluid ingress, or sorptivity. Neutron imaging is a powerful method for quantifying fluid penetration since it can describe where water has penetrated, how quickly it has penetrated and the volume of water in the concrete or mortar. Neutrons are sensitive to light atoms such as hydrogen and thus clearly detect water at high spatial and temporal resolution. It can be used to detect small changes in moisture content and is ideal for monitoring wetting and drying in mortar exposed to various fluids. This study aimed at developing a method to accurately estimate moisture content in mortar. The common practice is to image the material dry as a reference before exposing to fluid and normalizing subsequent images to the reference. The volume of water can then be computed using the Beer-Lambert law. This method can be limiting because it requires exact image alignment between the reference image and all subsequent images. A model of neutron attenuation in a multi-phase cementitious composite was developed to be used in cases where a reference image is not available. The attenuation coefficients for water, un-hydrated cement, and sand were directly calculated from the neutron images. The attenuation coefficient for the hydration products was then back-calculated. The model can estimate the degree of saturation in a mortar with known mixture proportions without using a reference image for calculation. Absorption in mortars exposed to various fluids (i.e., deionized water and calcium chloride solutions) were investigated

  16. Thermal-hydraulic studies of the Advanced Neutron Source cold source

    International Nuclear Information System (INIS)

    The Advanced Neutron Source (ANS), in its conceptual design phase at Oak Ridge National Laboratory, was to be a user-oriented neutron research facility producing the most intense steady-state flux of thermal and cold neutrons in the world. Among its many scientific applications, the production of cold neutrons was a significant research mission for the ANS. The cold neutrons come from two independent cold sources positioned near the reactor core. Contained by an aluminum alloy vessel, each cold source is a 410-mm-diam sphere of liquid deuterium that functions both as a neutron moderator and a cryogenic coolant. With nuclear heating of the containment vessel and internal baffling, steady-state operation requires close control of the liquid deuterium flow near the vessel's inner surface. Preliminary thermal-hydraulic analyses supporting the cold source design were performed with heat conduction simulations of the vessel walls and multidimensional computational fluid dynamics simulations of the liquid deuterium flow and heat transfer. This report presents the starting phase of a challenging program and describes the cold source conceptual design, the thermal-hydraulic feasibility studies of the containment vessel, and the future computational and experimental studies that were planned to verify the final design

  17. Development of Advanced Multi-Modality Radiation Treatment Planning Software for Neutron Radiotherapy and Beyond

    Energy Technology Data Exchange (ETDEWEB)

    Nigg, D; Wessol, D; Wemple, C; Harkin, G; Hartmann-Siantar, C

    2002-08-20

    The Idaho National Engineering and Environmental Laboratory (INEEL) has long been active in development of advanced Monte-Carlo based computational dosimetry and treatment planning methods and software for advanced radiotherapy, with a particular focus on Neutron Capture Therapy (NCT) and, to a somewhat lesser extent, Fast-Neutron Therapy. The most recent INEEL software system of this type is known as SERA, Simulation Environment for Radiotherapy Applications. As a logical next step in the development of modern radiotherapy planning tools to support the most advanced research, INEEL and Lawrence Livermore National Laboratory (LLNL), the developers of the PEREGRTNE computational engine for radiotherapy treatment planning applications, have recently launched a new project to collaborate in the development of a ''next-generation'' multi-modality treatment planning software system that will be useful for all modern forms of radiotherapy.

  18. Auto MOC-A 2D neutron transport code for arbitrary geometry based on the method of characteristics and customization of AutoCAD

    International Nuclear Information System (INIS)

    A new 2D neutron transport code AutoMOC for arbitrary geometry has been developed. This code is based on the method of characteristics (MOCs) and the customization of AutoCAD. The MOC solves the neutron transport equation along characteristic lines. It is independent of the geometric shape of boundaries and regions. So theoretically, this method can be used to solve the neutron transport equation in highly complex geometries. However, it is important to describe the geometry and calculate intersection points of each characteristic line with every boundary and region in advance. In complex geometries, due to the complications of treating the arbitrary domain, the selection of geometric shapes and efficiency of ray tracing are generally limited. The geometry treatment through the customization of AutoCAD, a widely used computer-aided design software package, is given in this paper. Thanks to the powerful capability of AutoCAD, the description of arbitrary geometry becomes quite convenient. Moreover, with the language Visual Basic for Applications (VBAs), AutoCAD can be customized to carry out the ray tracing procedure with a high flexibility in geometry. The numerical results show that AutoMOC can solve 2D neutron transport problems in a complex geometry accurately and effectively

  19. Measurements of neutron cross sections for advanced nuclear energy systems at n_TOF (CERN

    Directory of Open Access Journals (Sweden)

    Barbagallo M.

    2014-03-01

    Full Text Available The n_TOF facility operates at CERN with the aim of addressing the request of high accuracy nuclear data for advanced nuclear energy systems as well as for nuclear astrophysics. Thanks to the features of the neutron beam, important results have been obtained on neutron induced fission and capture cross sections of U, Pu and minor actinides. Recently the construction of another beam line has started; the new line will be complementary to the first one, allowing to further extend the experimental program foreseen for next measurement campaigns.

  20. Asymptotic Analysis of Time-Dependent Neutron Transport Coupled with Isotopic Depletion and Radioactive Decay

    Energy Technology Data Exchange (ETDEWEB)

    Brantley, P S

    2006-09-27

    We describe an asymptotic analysis of the coupled nonlinear system of equations describing time-dependent three-dimensional monoenergetic neutron transport and isotopic depletion and radioactive decay. The classic asymptotic diffusion scaling of Larsen and Keller [1], along with a consistent small scaling of the terms describing the radioactive decay of isotopes, is applied to this coupled nonlinear system of equations in a medium of specified initial isotopic composition. The analysis demonstrates that to leading order the neutron transport equation limits to the standard time-dependent neutron diffusion equation with macroscopic cross sections whose number densities are determined by the standard system of ordinary differential equations, the so-called Bateman equations, describing the temporal evolution of the nuclide number densities.

  1. Advances in neutron capture therapy 2006. Proceedings of 12th international congress on neutron capture therapy

    International Nuclear Information System (INIS)

    The Twelfth International Congress on Neutron Capture Therapy (ICNCT-12) is being held from October 9th to 13th, 2006 at the Kagawa International Congress Hall in Takamatsu, Kagawa, Japan. The main theme of the congress is From the past to the Future'. Five symposiums were organized to accommodate all the contributions from the international scientific committees of the International Society for Neutron Capture Therapy (ISNCT), and two symposiums were added to balance the number of fields of specialties. The seven symposiums for ICNCT-12 are as follows: 1) Clinical Results of BNCT for Brain Tumors, 2) Dosimetry, 3) Treatment Planning system, 4) Drug Delivery System, 5) Biomedical and General Matters, 6) BNCT Systems using Accelerators, 7) New Applications and Protocols for BNCT. There are a total of 195 presentations in this congress: 3 special lectures, 34 symposium presentations, 10 presentations in two special sessions from the recipients of the Ralph G. Fairchild Award, 70 presentations in the oral parallel sessions and 78 presentations in the poster sessions. A compilation of 169 papers are published in this proceedings. The 165 of the presented papers are indexed individually. (J.P.N.)

  2. General Design for CARR Neutron Guide System

    Institute of Scientific and Technical Information of China (English)

    2011-01-01

    A neutron guide system has been designed and partly installed at the China Advanced Research Reactor (CARR) to transport cold neutrons from the cold neutron source (CNS) to several instruments,which are situated in a separate guide hall of 30 m×60 m.

  3. In situ neutron depth profiling: A powerful method to probe lithium transport in micro-batteries

    NARCIS (Netherlands)

    Oudenhoven, J.F.M.; Labohm, F.; Mulder, M.; Niessen, R.A.H.; Mulder, F.M.; Notten, P.H.L.

    2011-01-01

    In situ neutron depth profiling (NDP) offers the possibility to observe lithium transport inside micro-batteries during battery operation. It is demonstrated that NDP results are consistent with the results of electrochemical measurements, and that the use of an enriched6LiCoO2 cathode offers more i

  4. The neutron transport code DTF-Traca users manual and input data

    Energy Technology Data Exchange (ETDEWEB)

    Ahnert, C.

    1979-07-01

    This is a users manual of the neutron transport code DTF-TRACA, which is a version of the original DTF-IV with some modifications made at JEN. A detailed input data descriptions is given. The new options developed at JEN are included too. (Author) 18 refs.

  5. Existence result for the kinetic neutron transport problem with a general albedo boundary condition

    International Nuclear Information System (INIS)

    We present an existence result for the kinetic neutron transport equation with a general albedo boundary condition. The proof is constructive in the sense that we build a sequence that converges to the solution of the problem by iterating on the albedo term. Both nonhomogeneous and albedo boundary conditions are studied. (authors)

  6. Two-group neutron transport theory in adjacent space with lineary anisotropic scattering

    International Nuclear Information System (INIS)

    A solution method for two-group neutron transport theory with anisotropic scattering is introduced by the combination of case method (expansion method of self singular function) and the invariant imbedding (invariance principle). The numerical results for the Milne problem in light water and borated water is presented to demonstrate the avalibility of the method

  7. The adjoint neutron transport equation and the statistical approach for its solution

    CERN Document Server

    Saracco, Paolo; Ravetto, Piero

    2016-01-01

    The adjoint equation was introduced in the early days of neutron transport and its solution, the neutron importance, has ben used for several applications in neutronics. The work presents at first a critical review of the adjoint neutron transport equation. Afterwards, the adjont model is constructed for a reference physical situation, for which an analytical approach is viable, i.e. an infinite homogeneous scattering medium. This problem leads to an equation that is the adjoint of the slowing-down equation that is well-known in nuclear reactor physics. A general closed-form analytical solution to such adjoint equation is obtained by a procedure that can be used also to derive the classical Placzek functions. This solution constitutes a benchmark for any statistical or numerical approach to the adjoint equation. A sampling technique to evaluate the adjoint flux for the transport equation is then proposed and physically interpreted as a transport model for pseudo-particles. This can be done by introducing appr...

  8. Sensitive and transportable gadolinium-core plastic scintillator sphere for neutron detection and counting

    Science.gov (United States)

    Dumazert, Jonathan; Coulon, Romain; Carrel, Frédérick; Corre, Gwenolé; Normand, Stéphane; Méchin, Laurence; Hamel, Matthieu

    2016-08-01

    Neutron detection forms a critical branch of nuclear-related issues, currently driven by the search for competitive alternative technologies to neutron counters based on the helium-3 isotope. The deployment of plastic scintillators shows a high potential for efficient detectors, safer and more reliable than liquids, more easily scalable and cost-effective than inorganic. In the meantime, natural gadolinium, through its 155 and mostly 157 isotopes, presents an exceptionally high interaction probability with thermal neutrons. This paper introduces a dual system including a metal gadolinium core inserted at the center of a high-scale plastic scintillator sphere. Incident fast neutrons are thermalized by the scintillator shell and then may be captured with a significant probability by gadolinium 155 and 157 nuclei in the core. The deposition of a sufficient fraction of the capture high-energy prompt gamma signature inside the scintillator shell will then allow discrimination from background radiations by energy threshold, and therefore neutron detection. The scaling of the system with the Monte Carlo MCNPX2.7 code was carried out according to a tradeoff between the moderation of incident fast neutrons and the probability of slow neutron capture by a moderate-cost metal gadolinium core. Based on the parameters extracted from simulation, a first laboratory prototype for the assessment of the detection method principle has been synthetized. The robustness and sensitivity of the neutron detection principle are then assessed by counting measurement experiments. Experimental results confirm the potential for a stable, highly sensitive, transportable and cost-efficient neutron detector and orientate future investigation toward promising axes.

  9. Advanced Engineering Environments for Space Transportation System Development

    Science.gov (United States)

    Thomas, L. Dale; Smith, Charles A.; Beveridge, James

    2000-01-01

    There are significant challenges facing today's launch vehicle industry. Global competition, more complex products, geographically-distributed design teams, demands for lower cost, higher reliability and safer vehicles, and the need to incorporate the latest technologies quicker, all face the developer of a space transportation system. Within NASA, multiple technology development and demonstration projects are underway toward the objectives of safe, reliable, and affordable access to space. New information technologies offer promising opportunities to develop advanced engineering environments to meet these challenges. Significant advances in the state-of-the-art of aerospace engineering practice are envisioned in the areas of engineering design and analytical tools, cost and risk tools, collaborative engineering, and high-fidelity simulations early in the development cycle. At the Marshall Space Flight Center, work has begun on development of an advanced engineering environment specifically to support the design, modeling, and analysis of space transportation systems. This paper will give an overview of the challenges of developing space transportation systems in today's environment and subsequently discuss the advanced engineering environment and its anticipated benefits.

  10. Electromagnetic Signatures of Neutron Star Mergers in the Advanced LIGO Era

    CERN Document Server

    Fernández, Rodrigo

    2015-01-01

    The mergers of binaries containing neutron stars and stellar-mass black holes are the most promising sources for direct detection in gravitational waves by the interferometers Advanced LIGO and Virgo over the next few years. The concurrent detection of electromagnetic emission from these events would greatly enhance the scientific return of these discoveries. Here we review the state of the art in modeling the electromagnetic signal of neutron star binary mergers across different phases of the merger and multiple wavelengths. We focus on those observables which provide the most sensitive diagnostics of the merger physics and the contribution to the synthesis of rapid neutron capture ($r$-process) elements in the Galaxy. We also outline expected future developments on the observational and theoretical sides of this rapidly evolving field.

  11. Post-merger evolution of a neutron star-black hole binary with neutrino transport

    CERN Document Server

    Foucart, Francois; Roberts, Luke; Duez, Matthew D; Haas, Roland; Kidder, Lawrence E; Ott, Christian D; Pfeiffer, Harald P; Scheel, Mark A; Szilagyi, Bela

    2015-01-01

    We present a first simulation of the post-merger evolution of a black hole-neutron star binary in full general relativity using an energy-integrated general relativistic truncated moment formalism for neutrino transport. We describe our implementation of the moment formalism and important tests of our code, before studying the formation phase of a disk after a black hole-neutron star merger. We use as initial data an existing general relativistic simulation of the merger of a neutron star of 1.4 solar mass with a black hole of 7 solar mass and dimensionless spin a/M=0.8. Comparing with a simpler leakage scheme for the treatment of the neutrinos, we find noticeable differences in the neutron to proton ratio in and around the disk, and in the neutrino luminosity. We find that the electron neutrino luminosity is much lower in the transport simulations, and that the remnant is less neutron-rich. The spatial distribution of the neutrinos is significantly affected by relativistic effects. Over the short timescale e...

  12. Applications of advanced transport aircraft in developing countries

    Science.gov (United States)

    Gobetz, F. W.; Assarabowski, R. J.; Leshane, A. A.

    1978-01-01

    Four representative market scenarios were studied to evaluate the relative performance of air-and surface-based transportation systems in meeting the needs of two developing contries, Brazil and Indonesia, which were selected for detailed case studies. The market scenarios were: remote mining, low-density transport, tropical forestry, and large cargo aircraft serving processing centers in resource-rich, remote areas. The long-term potential of various aircraft types, together with fleet requirements and necessary technology advances, is determined for each application.

  13. Advanced Transport Operating System (ATOPS) control display unit software description

    Science.gov (United States)

    Slominski, Christopher J.; Parks, Mark A.; Debure, Kelly R.; Heaphy, William J.

    1992-01-01

    The software created for the Control Display Units (CDUs), used for the Advanced Transport Operating Systems (ATOPS) project, on the Transport Systems Research Vehicle (TSRV) is described. Module descriptions are presented in a standardized format which contains module purpose, calling sequence, a detailed description, and global references. The global reference section includes subroutines, functions, and common variables referenced by a particular module. The CDUs, one for the pilot and one for the copilot, are used for flight management purposes. Operations performed with the CDU affects the aircraft's guidance, navigation, and display software.

  14. The use of neutron generators for the detection of illicit materials in the sea transportation system

    International Nuclear Information System (INIS)

    In today's society acts of terrorism must involve in some stages the illicit trafficking either of explosives, chemical agents and/or nuclear materials. Therefore society must rely on an anti-trafficking infrastructure which encompasses responsible authorities, field personnel and adequate instrumental networks. Modern inspection systems for personnel, parcel, vehicle and cargo, as noninvasive imaging techniques, are based on the use of nuclear analytical methods. The inspection systems make use of penetrating radiation (neutrons, gamma and x-rays) in a scanning geometry, with the detection of radiation either transmitted or produced in the interrogated object. Explosives and chemical agent detection systems are based on the fact that the problem of identification can be reduced to the measurement of elemental concentrations. Different nuclear analytical techniques could be used for this purpose; however the use of neutrons has some specific advantages due to the high penetrability in large payloads. Of special interest is the design and use of a transportable neutron system coupled to a gamma-ray radiographic device for inspecting large containers searching for contraband, explosives, weapons etc. The use of neutron induced reactions for non-destructive bulk elemental analysis is well documented. All neutrons, in particular fast neutrons, are well suited to explore large volume samples because of their high penetration in bulk material. Fast neutrons can be produced efficiently and economically by natural radioactive sources, small accelerators or compact electronic neutron generators, making possible the use of neutron based techniques in field applications. Gamma-rays produced by irradiating the sample with neutrons gives the elemental composition of the material, moreover, knowing the nuclear cross-sections and estimating the absorption factors in the different materials, it is possible to perform a quantitative analysis of elements in the sample even in depth

  15. Numerical solution of neutron transport equations in discrete ordinates and slab geometry

    International Nuclear Information System (INIS)

    An unified formalism to solve numerically, between other equation, the neutron transport in discrete ordinates, slab geometry, several energy groups and independents of time, has been developed recently. Such a formalism cover some of the conventional schemes as diamond difference, (WDD) characteristic step (SC) lineal characteristic (LC), quadratic characteristic (QC) and lineal discontinuous. Unified formation gives before hand the convergence order of the previously selected scheme. In fact it allows besides to generate a big amount of numerical schemes, with which is also possible to solve numerical equations as soon as neutron transport. The essential purpose of this work was to solve the neutron transport equations in slab geometry and discrete ordinates considering several energy groups without to take under advisement time dependence based in the above mentioned unified formalism. To reach this purpose it was necesary to design a computer code with the name TNOD1 (Neutron transport in discrete ordinates and 1 dimension) which includes each one of the schemes already pointed out. there exist two numerical schemes, also recently developed, quadratic continuous (QC) and cubic continuous (CN), although covered by unified formalism, it has been possible to include them inside this computer code without make substantial changes in its structure. In chapter I, derivative of neutron transport equation independent of time is taken, for angular flux, including boundary conditions and discontinuity. In chapter II the neutron transport equations are obtained in multigroups, independents of time, for approximation of discrete ordinates. Description of theory related with unified formalism and its relationship with mentioned discretization schemes is presented in chapter III. Chapter IV describes the computer code developed and finally, in chapter V different numerical results obtained with TNOD1 program are shown. In Appendix A theorems and mathematical arguments used

  16. Development and benchmarking of higher energy neutron transport data libraries

    International Nuclear Information System (INIS)

    Neutron cross-section evaluations covering the energy range from 10/sup /minus/11/ to 100 MeV have been prepared for several materials. The principal method used to generate this data base has employed statistical-preequilibrium nuclear models, sophisticated phase shift analyses, and R-matrix techniques. The library takes advantage of formats developed for Version 6 of the Evaluated Nuclear Data File, ENDF. Methods to efficiently utilize the ENDF/B-VI representation of this library in the MCNP Monte Carlo code have been developed. MCNP results using the new library have been compared with calculated results using codes or data based upon intranuclear cascade models. 7 refs., 8 figs

  17. Transport simulation and image reconstruction for fast-neutron detection of explosives and narcotics

    Energy Technology Data Exchange (ETDEWEB)

    Micklich, B.J.; Fink, C.L.; Sagalovsky, L.

    1995-07-01

    Fast-neutron inspection techniques show considerable promise for explosive and narcotics detection. A key advantage of using fast neutrons is their sensitivity to low-Z elements (carbon, nitrogen, and oxygen), which are the primary constituents of these materials. We are currently investigating two interrogation methods in detail: Fast-Neutron Transmission Spectroscopy (FNTS) and Pulsed Fast-Neutron Analysis (PFNA). FNTS is being studied for explosives and narcotics detection in luggage and small containers for which the transmission ratio is greater than about 0.01. The Monte-Carlo radiation transport code MCNP is being used to simulate neutron transmission through a series of phantoms for a few (3-5) projection angles and modest (2 cm) resolution. Areal densities along projection rays are unfolded from the transmission data. Elemental abundances are obtained for individual voxels by tomographic reconstruction, and these reconstructed elemental images are combined to provide indications of the presence or absence of explosives or narcotics. PFNA techniques are being investigated for detection of narcotics in cargo containers because of the good penetration of the fast neutrons and the low attenuation of the resulting high-energy gamma-ray signatures. Analytic models and Monte-Carlo simulations are being used to explore the range of capabilities of PFNA techniques and to provide insight into systems engineering issues. Results of studies from both FNTS and PFNA techniques are presented.

  18. Scientific Advancements and Technological Developments of High P-T Neutron Diffraction at LANSCE, Los Alamos

    Science.gov (United States)

    Zhao, Y.; Daemen, L. L.; Zhang, J.

    2003-12-01

    In-situ high P-T neutron diffraction experiments provide unique opportunities to study the crystal structure, hydrogen bonding, magnetism, and thermal parameters of light elements (eg. H, Li, B) and heavy elements (eg. Ta, U, Pu,), that are virtually impossible to determine with x-ray diffraction techniques. For example, thermoelasticity and Debye-Waller factor as function of pressure and temperature can be derived using in-situ high P-T neutron diffraction techniques. These applications can also be extended to a much broader spectrum of scientific problems. For instance, puzzles in Earth science such as the carbon cycle and the role of hydrous minerals for water exchange between lithosphere and biosphere can be directly addressed. Moreover, by introducing in-situ shear, texture of metals and minerals accompanied with phase transitions at high P-T conditions can also be studied by high P-T neutron diffraction. We have successfully conducted high P-T neutron diffraction experiments at LANSCE and achieved simultaneous high pressures and temperatures of 10 GPa and 1500 K. With an average 3-6 hours of data collection, the diffraction data are of sufficiently high quality for the determination of structural parameters and thermal vibrations. We have studied hydrous mineral (MgOD), perovskite (K.15,Na.85)MgF3, clathrate hydrates (CH4-, CO2-, and H2-), metals (Mo, Al, Zr), and amorphous materials (carbon black, BMG). The aim of our research is to accurately map bond lengths, bond angles, neighboring atomic environments, and phase stability in P-T-X space. Studies based on high-pressure neutron diffraction are important for multi-disciplinary science and we welcome researchers from all fields to use this advanced technique. We have developed a 500-ton toroidal press, TAP-98, to conduct simultaneous high P-T neutron diffraction experiments inside of HIPPO (High-Pressure and Preferred-Orientation diffractometer). We have also developed a large gem-crystal anvil cell, ZAP-01

  19. Ageing of a neutron shielding used in transport/storage casks

    Energy Technology Data Exchange (ETDEWEB)

    Nizeyiman, Fidele; Alami, Aatif; Issard, Herve; Bellenger, Veronique [TN International, 1 rue des herons, Montigny le Bretonneux, 78054 Saint Quentin en Yvelines (France); Laboratoire PIMM, Arts and Metiers ParisTech, 151 Bd de l' Hopital, 75013 Paris (France)

    2012-07-11

    In radioactive materials transport/storage casks, a mineral-filled vinylester composite is used for neutron shielding which relies on its hydrogen and boron atoms content. During cask service life, this composite is mainly subjected to three types of ageing: hydrothermal ageing, thermal oxidation and neutron irradiation. The aim of this study is to investigate the effect of hydrothermal ageing on the properties and chemical composition of this polymer composite. At high temperature (120 Degree-Sign C and 140 Degree-Sign C), the main consequence is the strong decrease of mechanical properties induced by the filler/matrix debonding.

  20. Development of a 1D neutron transport code employing the method of characteristics

    International Nuclear Information System (INIS)

    To investigate the 2D/1D fusion core analysis method, a 1D neutron transport problem solver, PEACH-ID, is developed. It is a code of method of characteristics (MOC), both the usual fiat-source step characteristics (SC) scheme and linear source (LS) approximation scheme are adopted for tracking calculation along the neutron flying trajectory. Exponential function interpolation table and fission source extrapolation are adopted as two major methods to accelerate the computational process. Numerical results demonstrate that PEACH-1D is accurate and efficient, and the proposed LS scheme is able to handle quite larger mesh division and deserves much more application in the MOC codes. (authors)

  1. Advanced Transport Operating System (ATOPS) utility library software description

    Science.gov (United States)

    Clinedinst, Winston C.; Slominski, Christopher J.; Dickson, Richard W.; Wolverton, David A.

    1993-01-01

    The individual software processes used in the flight computers on-board the Advanced Transport Operating System (ATOPS) aircraft have many common functional elements. A library of commonly used software modules was created for general uses among the processes. The library includes modules for mathematical computations, data formatting, system database interfacing, and condition handling. The modules available in the library and their associated calling requirements are described.

  2. Advanced Reactors Thermal Energy Transport for Process Industries

    Energy Technology Data Exchange (ETDEWEB)

    P. Sabharwall; S.J. Yoon; M.G. McKellar; C. Stoots; George Griffith

    2014-07-01

    The operation temperature of advanced nuclear reactors is generally higher than commercial light water reactors and thermal energy from advanced nuclear reactor can be used for various purposes such as liquid fuel production, district heating, desalination, hydrogen production, and other process heat applications, etc. Some of the major technology challenges that must be overcome before the advanced reactors could be licensed on the reactor side are qualification of next generation of nuclear fuel, materials that can withstand higher temperature, improvement in power cycle thermal efficiency by going to combined cycles, SCO2 cycles, successful demonstration of advanced compact heat exchangers in the prototypical conditions, and from the process side application the challenge is to transport the thermal energy from the reactor to the process plant with maximum efficiency (i.e., with minimum temperature drop). The main focus of this study is on doing a parametric study of efficient heat transport system, with different coolants (mainly, water, He, and molten salts) to determine maximum possible distance that can be achieved.

  3. Comparison of neutronic transport equation resolution nodal methods

    International Nuclear Information System (INIS)

    In this work, some transport equation resolution nodal methods are comparatively studied: the constant-constant (CC), linear-nodal (LN) and the constant-quadratic (CQ). A nodal scheme equivalent to finite differences has been used for its programming, permitting its inclusion in existing codes. Some bidimensional problems have been solved, showing that linear-nodal (LN) are, in general, obtained with accuracy in CPU shorter times. (Author)

  4. Upper limits on the rates of binary neutron star and neutron-star--black-hole mergers from Advanced LIGO's first observing run

    CERN Document Server

    Abbott, B P; Abbott, T D; Abernathy, M R; Acernese, F; Ackley, K; Adams, C; Adams, T; Addesso, P; Adhikari, R X; Adya, V B; Affeldt, C; Agathos, M; Agatsuma, K; Aggarwal, N; Aguiar, O D; Aiello, L; Ain, A; Ajith, P; Allen, B; Allocca, A; Altin, P A; Anderson, S B; Anderson, W G; Arai, K; Araya, M C; Arceneaux, C C; Areeda, J S; Arnaud, N; Arun, K G; Ascenzi, S; Ashton, G; Ast, M; Aston, S M; Astone, P; Aufmuth, P; Aulbert, C; Babak, S; Bacon, P; Bader, M K M; Baker, P T; Baldaccini, F; Ballardin, G; Ballmer, S W; Barayoga, J C; Barclay, S E; Barish, B C; Barker, D; Barone, F; Barr, B; Barsotti, L; Barsuglia, M; Barta, D; Bartlett, J; Bartos, I; Bassiri, R; Basti, A; Batch, J C; Baune, C; Bavigadda, V; Bazzan, M; Bejger, M; Bell, A S; Berger, B K; Bergmann, G; Berry, C P L; Bersanetti, D; Bertolini, A; Betzwieser, J; Bhagwat, S; Bhandare, R; Bilenko, I A; Billingsley, G; Birch, J; Birney, R; Biscans, S; Bisht, A; Bitossi, M; Biwer, C; Bizouard, M A; Blackburn, J K; Blair, C D; Blair, D G; Blair, R M; Bloemen, S; Bock, O; Boer, M; Bogaert, G; Bogan, C; Bohe, A; Bond, C; Bondu, F; Bonnand, R; Boom, B A; Bork, R; Boschi, V; Bose, S; Bouffanais, Y; Bozzi, A; Bradaschia, C; Brady, P R; Braginsky, V B; Branchesi, M; Brau, J E; Briant, T; Brillet, A; Brinkmann, M; Brisson, V; Brockill, P; Broida, J E; Brooks, A F; Brown, D A; Brown, D D; Brown, N M; Brunett, S; Buchanan, C C; Buikema, A; Bulik, T; Bulten, H J; Buonanno, A; Buskulic, D; Buy, C; Byer, R L; Cabero, M; Cadonati, L; Cagnoli, G; Cahillane, C; Bustillo, J Calder'on; Callister, T; Calloni, E; Camp, J B; Cannon, K C; Cao, J; Capano, C D; Capocasa, E; Carbognani, F; Caride, S; Diaz, J Casanueva; Casentini, C; Caudill, S; Cavagli`a, M; Cavalier, F; Cavalieri, R; Cella, G; Cepeda, C B; Baiardi, L Cerboni; Cerretani, G; Cesarini, E; Chamberlin, S J; Chan, M; Chao, S; Charlton, P; Chassande-Mottin, E; Cheeseboro, B D; Chen, H Y; Chen, Y; Cheng, C; Chincarini, A; Chiummo, A; Cho, H S; Cho, M; Chow, J H; Christensen, N; Chu, Q; 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Drever, R W P; Driggers, J C; Ducrot, M; Dwyer, S E; Edo, T B; Edwards, M C; Effler, A; Eggenstein, H -B; Ehrens, P; Eichholz, J; Eikenberry, S S; Engels, W; Essick, R C; Etzel, T; Evans, M; Evans, T M; Everett, R; Factourovich, M; Fafone, V; Fair, H; Fairhurst, S; Fan, X; Fang, Q; Farinon, S; Farr, B; Farr, W M; Favata, M; Fays, M; Fehrmann, H; Fejer, M M; Fenyvesi, E; Ferrante, I; Ferreira, E C; Ferrini, F; Fidecaro, F; Fiori, I; Fiorucci, D; Fisher, R P; Flaminio, R; Fletcher, M; Fournier, J -D; Frasca, S; Frasconi, F; Frei, Z; Freise, A; Frey, R; Frey, V; Fritschel, P; Frolov, V V; Fulda, P; Fyffe, M; Gabbard, H A G; Gair, J R; Gammaitoni, L; Gaonkar, S G; Garufi, F; Gaur, G; Gehrels, N; Gemme, G; Geng, P; Genin, E; Gennai, A; George, J; Gergely, L; Germain, V; Ghosh, Abhirup; Ghosh, Archisman; Ghosh, S; Giaime, J A; Giardina, K D; Giazotto, A; Gill, K; Glaefke, A; Goetz, E; Goetz, R; Gondan, L; Gonz'alez, G; Castro, J M Gonzalez; Gopakumar, A; Gordon, N A; Gorodetsky, M L; Gossan, S E; Gosselin, M; Gouaty, R; Grado, A; Graef, C; Graff, P B; Granata, M; Grant, A; Gras, S; Gray, C; Greco, G; Green, A C; Groot, P; Grote, H; Grunewald, S; Guidi, G M; Guo, X; Gupta, A; Gupta, M K; Gushwa, K E; Gustafson, E K; Gustafson, R; Hacker, J J; Hall, B R; Hall, E D; Hammond, G; Haney, M; Hanke, M M; Hanks, J; Hanna, C; Hannam, M D; Hanson, J; Hardwick, T; Harms, J; Harry, G M; Harry, I W; Hart, M J; Hartman, M T; Haster, C -J; Haughian, K; Heidmann, A; Heintze, M C; Heitmann, H; Hello, P; Hemming, G; Hendry, M; Heng, I S; Hennig, J; Henry, J; Heptonstall, A W; Heurs, M; Hild, S; Hoak, D; Hofman, D; Holt, K; Holz, D E; Hopkins, P; Hough, J; Houston, E A; Howell, E J; Hu, Y M; Huang, S; Huerta, E A; Huet, D; Hughey, B; Husa, S; Huttner, S H; Huynh-Dinh, T; Indik, N; Ingram, D R; Inta, R; Isa, H N; Isac, J -M; Isi, M; Isogai, T; Iyer, B R; Izumi, K; Jacqmin, T; Jang, H; Jani, K; Jaranowski, P; Jawahar, S; Jian, L; Jim'enez-Forteza, F; Johnson, W W; Jones, D I; Jones, R; Jonker, R J G; Ju, L; K, Haris; Kalaghatgi, C V; Kalogera, V; Kandhasamy, S; Kang, G; Kanner, J B; Kapadia, S J; Karki, S; Karvinen, K S; Kasprzack, M; Katsavounidis, E; Katzman, W; Kaufer, S; Kaur, T; Kawabe, K; K'ef'elian, F; Kehl, M S; Keitel, D; Kelley, D B; Kells, W; Kennedy, R; Key, J S; Khalili, F Y; Khan, I; Khan, S; Khan, Z; Khazanov, E A; Kijbunchoo, N; Kim, Chi-Woong; Kim, Chunglee; Kim, J; Kim, K; Kim, N; Kim, W; Kim, Y -M; Kimbrell, S J; King, E J; King, P J; Kissel, J S; Klein, B; Kleybolte, L; Klimenko, S; Koehlenbeck, S M; Koley, S; Kondrashov, V; Kontos, A; Korobko, M; Korth, W Z; Kowalska, I; Kozak, D B; Kringel, V; Krishnan, B; Kr'olak, A; Krueger, C; Kuehn, G; Kumar, P; Kumar, R; Kuo, L; Kutynia, A; Lackey, B D; Landry, M; Lange, J; Lantz, B; Lasky, P D; Laxen, M; Lazzarini, A; Lazzaro, C; Leaci, P; Leavey, S; Lebigot, E O; Lee, C H; Lee, H K; Lee, H M; Lee, K; Lenon, A; Leonardi, M; Leong, J R; Leroy, N; Letendre, N; Levin, Y; Lewis, J B; Li, T G F; Libson, A; Littenberg, T B; Lockerbie, N A; Lombardi, A L; London, L T; Lord, J E; Lorenzini, M; Loriette, V; Lormand, M; Losurdo, G; Lough, J D; L"uck, H; Lundgren, A P; Lynch, R; Ma, Y; Machenschalk, B; MacInnis, M; Macleod, D M; Magana-Sandoval, F; Zertuche, L Magana; Magee, R M; Majorana, E; Maksimovic, I; Malvezzi, V; Man, N; Mandic, V; Mangano, V; Mansell, G L; Manske, M; Mantovani, M; Marchesoni, F; Marion, F; M'arka, S; M'arka, Z; Markosyan, A S; Maros, E; Martelli, F; Martellini, L; Martin, I W; Martynov, D V; Marx, J N; Mason, K; Masserot, A; Massinger, T J; Masso-Reid, M; Mastrogiovanni, S; Matichard, F; Matone, L; Mavalvala, N; Mazumder, N; McCarthy, R; McClelland, D E; McCormick, S; McGuire, S C; McIntyre, G; McIver, J; McManus, D J; McRae, T; McWilliams, S T; Meacher, D; Meadors, G D; Meidam, J; Melatos, A; Mendell, G; Mercer, R A; Merilh, E L; Merzougui, M; Meshkov, S; Messenger, C; Messick, C; Metzdorff, R; Meyers, P M; Mezzani, F; Miao, H; Michel, C; Middleton, H; Mikhailov, E E; Milano, L; Miller, A L; Miller, A; Miller, B B; Miller, J; Millhouse, M; Minenkov, Y; Ming, J; Mirshekari, S; Mishra, C; Mitra, S; Mitrofanov, V P; Mitselmakher, G; Mittleman, R; Moggi, A; Mohan, M; Mohapatra, S R P; Montani, M; Moore, B C; Moore, C J; Moraru, D; Moreno, G; Morriss, S R; Mossavi, K; Mours, B; Mow-Lowry, C M; Mueller, G; Muir, A W; Mukherjee, Arunava; Mukherjee, D; Mukherjee, S; Mukund, N; Mullavey, A; Munch, J; Murphy, D J; Murray, P G; Mytidis, A; Nardecchia, I; Naticchioni, L; Nayak, R K; Nedkova, K; Nelemans, G; Nelson, T J N; Neri, M; Neunzert, A; Newton, G; Nguyen, T T; Nielsen, A B; Nissanke, S; Nitz, A; Nocera, F; Nolting, D; Normandin, M E N; Nuttall, L K; Oberling, J; Ochsner, E; O'Dell, J; Oelker, E; Ogin, G H; Oh, J J; Oh, S H; Ohme, F; Oliver, M; Oppermann, P; Oram, Richard J; O'Reilly, B; O'Shaughnessy, R; Ottaway, D J; Overmier, H; Owen, B J; Pai, A; Pai, S A; Palamos, J R; Palashov, O; Palomba, C; Pal-Singh, A; Pan, H; Pankow, C; Pannarale, F; Pant, B C; Paoletti, F; Paoli, A; Papa, M A; 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R"udiger, A; Ruggi, P; Ryan, K; Sachdev, S; Sadecki, T; Sadeghian, L; Sakellariadou, M; Salconi, L; Saleem, M; Salemi, F; Samajdar, A; Sammut, L; Sanchez, E J; Sandberg, V; Sandeen, B; Sanders, J R; Sassolas, B; Sathyaprakash, B S; Saulson, P R; Sauter, O E S; Savage, R L; Sawadsky, A; Schale, P; Schilling, R; Schmidt, J; Schmidt, P; Schnabel, R; Schofield, R M S; Sch"onbeck, A; Schreiber, E; Schuette, D; Schutz, B F; Scott, J; Scott, S M; Sellers, D; Sengupta, A S; Sentenac, D; Sequino, V; Sergeev, A; Setyawati, Y; Shaddock, D A; Shaffer, T; Shahriar, M S; Shaltev, M; Shapiro, B; Shawhan, P; Sheperd, A; Shoemaker, D H; Shoemaker, D M; Siellez, K; Siemens, X; Sieniawska, M; Sigg, D; Silva, A D; Singer, A; Singer, L P; Singh, A; Singh, R; Singhal, A; Sintes, A M; Slagmolen, B J J; Smith, J R; Smith, N D; Smith, R J E; Son, E J; Sorazu, B; Sorrentino, F; Souradeep, T; Srivastava, A K; Staley, A; Steinke, M; Steinlechner, J; Steinlechner, S; Steinmeyer, D; Stephens, B C; Stone, R; Strain, K A; Straniero, N; Stratta, G; Strauss, N A; Strigin, S; Sturani, R; Stuver, A L; Summerscales, T Z; Sun, L; Sunil, S; Sutton, P J; Swinkels, B L; Szczepa'nczyk, M J; Tacca, M; Talukder, D; Tanner, D B; T'apai, M; Tarabrin, S P; Taracchini, A; Taylor, R; Theeg, T; Thirugnanasambandam, M P; Thomas, E G; Thomas, M; Thomas, P; Thorne, K A; Thrane, E; Tiwari, S; Tiwari, V; Tokmakov, K V; Toland, K; Tomlinson, C; Tonelli, M; Tornasi, Z; Torres, C V; Torrie, C I; T"oyr"a, D; Travasso, F; Traylor, G; Trifir`o, D; Tringali, M C; Trozzo, L; Tse, M; Turconi, M; Tuyenbayev, D; Ugolini, D; Unnikrishnan, C S; Urban, A L; Usman, S A; Vahlbruch, H; Vajente, G; Valdes, G; van Bakel, N; van Beuzekom, M; Brand, J F J van den; Broeck, C Van Den; Vander-Hyde, D C; van der Schaaf, L; van Heijningen, J V; van Veggel, A A; Vardaro, M; Vass, S; Vas'uth, M; Vaulin, R; Vecchio, A; Vedovato, G; Veitch, J; Veitch, P J; Venkateswara, K; Verkindt, D; Vetrano, F; Vicer'e, A; Vinciguerra, S; Vine, D J; Vinet, J -Y; Vitale, S; Vo, T; Vocca, H; Vorvick, C; Voss, D V; Vousden, W D; Vyatchanin, S P; Wade, A R; Wade, L E; Wade, M; Walker, M; Wallace, L; Walsh, S; Wang, G; Wang, H; Wang, M; Wang, X; Wang, Y; Ward, R L; Warner, J; Was, M; Weaver, B; Wei, L -W; Weinert, M; Weinstein, A J; Weiss, R; Wen, L; Wessels, P; Westphal, T; Wette, K; Whelan, J T; Whiting, B F; Williams, R D; Williamson, A R; Willis, J L; Willke, B; Wimmer, M H; Winkler, W; Wipf, C C; Wittel, H; Woan, G; Woehler, J; Worden, J; Wright, J L; Wu, D S; Wu, G; Yablon, J; Yam, W; Yamamoto, H; Yancey, C C; Yu, H; Yvert, M; zny, A Zadro; Zangrando, L; Zanolin, M; Zendri, J -P; Zevin, M; Zhang, L; Zhang, M; Zhang, Y; Zhao, C; Zhou, M; Zhou, Z; Zhu, X J; Zucker, M E; Zuraw, S E; Zweizig, J

    2016-01-01

    We report here the non-detection of gravitational waves from the merger of binary neutron star systems and neutron-star--black-hole systems during the first observing run of Advanced LIGO. In particular we searched for gravitational wave signals from binary neutron star systems with component masses $\\in [1,3] M_{\\odot}$ and component dimensionless spins $< 0.05$. We also searched for neutron-star--black-hole systems with the same neutron star parameters, black hole mass $\\in [2,99] M_{\\odot}$ and no restriction on the black hole spin magnitude. We assess the sensitivity of the two LIGO detectors to these systems, and find that they could have detected the merger of binary neutron star systems with component mass distributions of $1.35\\pm0.13 M_{\\odot}$ at a volume-weighted average distance of $\\sim$ 70Mpc, and for neutron-star--black-hole systems with neutron star masses of $1.4M_\\odot$ and black hole masses of at least $5M_\\odot$, a volume-weighted average distance of at least $\\sim$ 110Mpc. From this we...

  5. Reference neutron transport calculation note for Korea nuclear power plants with 3-loop PWR reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Byung Cheol; Chang, Ki Oak

    1997-05-01

    Reactor pressure vessel (RPV) steels are subjected to neutron irradiation at a temperature of about 290 deg C. This radiation exposure alters the mechanical properties, leading to a shift of the brittle-to-ductile transition temperature toward higher temperatures and to a diminution of the rupture energy as determined by Charpy V-notch tests. This radiation embrittlement is one of the important aging factors of nuclear power plants. U.S. NRC recommended the basic requirements for the determination of the pressure vessel fluence by regulatory guide DG-1025 in order to reduce the uncertainty in the determination of neutron fluence calculation and measurements. The determination of the pressure vessel fluence is based on both calculations and measurements. The fluence prediction is made with a calculation and the measurements are used to qualify the calculational methodology. Because of the importance and the difficulty of these calculations, the method`s qualification by comparison to measurement must be made to ensure a reliable and accurate vessel fluence determination. This reference calculation note is to provide a series of forward and adjoint neutron transport calculations for use in the evaluation of neutron dosimetry from surveillance capsule irradiations at 3-loop PWR reactor as well as for use in the determination of the neutron exposure of the reactor vessel wall in accordance with U.S Regulatory Guide DG-1025 requirements. The calculations of the pressure vessel fluence consist of the following steps; (1) Determination of the geometrical and material input data, (2) Determination of the core neutron source, and (3) Propagation of the neutron fluence from the core to the vessel and into the cavity. (author). 12 tabs., 3 figs., 7 refs.

  6. Neutron- and proton-induced evaluated transport library up to 150 MeV

    International Nuclear Information System (INIS)

    A new evaluated nuclear data library has been created. The library consists of two sub-libraries for neutron and proton incident particles. The neutron sub-library contains nuclear data for transport, heating and shielding applications for 242 nuclides with atomic numbers ranging from 8 to 82 in the energy region of primary neutrons from 10-5 eV to 150 MeV. Data below 20 MeV are taken mainly from ENDF/B-VI (revision 8) and for some nuclides, from the JENDL-3.3 and JEFF-3.0 libraries. The proton sub-library should contain data for the same range of target nuclides and energies. Proton-induced evaluated cross-section files are available for 15 nuclides at the moment. The evaluation of emitted particle energy and angular distributions at energies above 20 MeV (for incident neutrons) and above the reaction threshold (for incident protons) was performed with the help of the ALICE/ASH code and the analysis of available experimental data. The total cross-sections, elastic cross-sections and elastic scattering angular distributions were calculated with the help of the coupled channel model. The results of the calculation were adjusted to the data from ENDF/B-VI, JENDL-3.3, or JEFF-3.0 at the neutron energy equal to 20 MeV. The library is written in ENDF-6 format using the MF=3/MT=5 and MF=6/MT=5 representations

  7. Collocation method for the solution of the neutron transport equation with both symmetric and asymmetric scattering

    Energy Technology Data Exchange (ETDEWEB)

    Morel, J.E.

    1981-01-01

    A collocation method is developed for the solution of the one-dimensional neutron transport equation in slab geometry with both symmetric and polarly asymmetric scattering. For the symmetric scattering case, it is found that the collocation method offers a combination of some of the best characteristics of the finite-element and discrete-ordinates methods. For the asymmetric scattering case, it is found that the computational cost of cross-section data processing under the collocation approach can be significantly less than that associated with the discrete-ordinates approach. A general diffusion equation treating both symmetric and asymmetric scattering is developed and used in a synthetic acceleration algorithm to accelerate the iterative convergence of collocation solutions. It is shown that a certain type of asymmetric scattering can radically alter the asymptotic behavior of the transport solution and is mathematically equivalent within the diffusion approximation to particle transport under the influence of an electric field. The method is easily extended to other geometries and higher dimensions. Applications exist in the areas of neutron transport with highly anisotropic scattering (such as that associated with hydrogenous media), charged-particle transport, and particle transport in controlled-fusion plasmas. 23 figures, 6 tables.

  8. A new paradigm for whole core neutron transport without homogenization

    International Nuclear Information System (INIS)

    A new paradigm is introduced which allows the performance of whole core transport calculations without lattice homogenization. Quasi-reflected interface conditions are formulated to partially decouple periodic lattice effects from the pin-cell to pin-cell flux variation in the finite sub-element form of the variational nodal code VARIANT. With fuel-coolant homogenization eliminated, the interface variables that couple pin-cell sized nodes are divided into low-order and high-order spherical harmonic terms. Reflected interface conditions are subsequently applied to the high-order terms to remove them from the system of unknowns. Combined with an integral transport treatment within the node, the new approach dramatically reduces both the formation time and the size of the response matrices and leads to sharply reduced memory and CPU requirements. The method is applied to the two-dimensional C5-G7 problem, an OECD/NEA PWR benchmark containing MOX and UO2 fuel assemblies. Results indicate the new approach results in very little loss of accuracy relative to the corresponding full spherical harmonics expansions while reducing CPU times by well over an order of magnitude. (authors)

  9. Monte Carlo method for neutron transport calculations in graphics processing units (GPUs)

    International Nuclear Information System (INIS)

    Monte Carlo simulation is well suited for solving the Boltzmann neutron transport equation in an inhomogeneous media for complicated geometries. However, routine applications require the computation time to be reduced to hours and even minutes in a desktop PC. The interest in adopting Graphics Processing Units (GPUs) for Monte Carlo acceleration is rapidly growing. This is due to the massive parallelism provided by the latest GPU technologies which is the most promising solution to the challenge of performing full-size reactor core analysis on a routine basis. In this study, Monte Carlo codes for a fixed-source neutron transport problem were developed for GPU environments in order to evaluate issues associated with computational speedup using GPUs. Results obtained in this work suggest that a speedup of several orders of magnitude is possible using the state-of-the-art GPU technologies. (author)

  10. Numerical evaluation of the light transport properties of alternative He-3 neutron detectors using ceramic scintillators

    Energy Technology Data Exchange (ETDEWEB)

    Ohzu, A., E-mail: ohzu.akira@jaea.go.jp [Nuclear Science and Engineering Center, Japan Atomic Energy Agency, Tokai, Naka, Ibaraki 319-1195 (Japan); Takase, M.; Haruyama, M.; Kurata, N.; Kobayashi, N.; Kureta, M. [Nuclear Science and Engineering Center, Japan Atomic Energy Agency, Tokai, Naka, Ibaraki 319-1195 (Japan); Nakamura, T.; Toh, K.; Sakasai, K.; Suzuki, H.; Soyama, K. [J-PARC, Japan Atomic Energy Agency, Tokai, Naka, Ibaraki 319-1195 (Japan); Seya, M. [Integrated Support Center for Nuclear Nonproliferation and Nuclear Security, Japan Atomic Energy Agency, Tokai, Naka, Ibaraki 319-1195 (Japan)

    2015-10-21

    The light transport properties of scintillator light inside alternative He-3 neutron detectors using scintillator sheets have been investigated by a ray-tracing simulation code. The detector consists of a light-reflecting tube, a thin rectangular ceramic scintillator sheet laminated on a glass plate, and two photo-multiplier tubes (PMTs) mounted at both ends of the detector tube. The flashes of light induced on the surface of the scintillator sheet via nuclear interaction between the scintillator and neutrons are detected by the two PMTs. The light output at both ends of various detectors in which the scintillator sheets are installed with several different arrangements were examined and evaluated in comparison with experimental results. The results derived from the simulation reveal that the light transport property is strongly dependent on the arrangement of the scintillator sheet inside the tube and the shape of the tube.

  11. Advanced Air Transportation Technologies Project, Final Document Collection

    Science.gov (United States)

    Mogford, Richard H.; Wold, Sheryl (Editor)

    2008-01-01

    This CD ROM contains a compilation of the final documents of the Advanced Air Transportation Technologies (AAIT) project, which was an eight-year (1996 to 2004), $400M project managed by the Airspace Systems Program office, which was part of the Aeronautics Research Mission Directorate at NASA Headquarters. AAIT focused on developing advanced automation tools and air traffic management concepts that would help improve the efficiency of the National Airspace System, while maintaining or enhancing safety. The documents contained in the CD are final reports on AAIT tasks that serve to document the project's accomplishments over its eight-year term. Documents include information on: Advanced Air Transportation Technologies, Autonomous Operations Planner, Collaborative Arrival Planner, Distributed Air/Ground Traffic Management Concept Elements 5, 6, & 11, Direct-To, Direct-To Technology Transfer, Expedite Departure Path, En Route Data Exchange, Final Approach Spacing Tool - (Active and Passive), Multi-Center Traffic Management Advisor, Multi Center Traffic Management Advisor Technology Transfer, Surface Movement Advisor, Surface Management System, Surface Management System Technology Transfer and Traffic Flow Management Research & Development.

  12. Neutron transport solver parallelization using a Domain Decomposition method

    International Nuclear Information System (INIS)

    A domain decomposition (DD) method is investigated for the parallel solution of the second-order even-parity form of the time-independent Boltzmann transport equation. The spatial discretization is performed using finite elements, and the angular discretization using spherical harmonic expansions (PN method). The main idea developed here is due to P.L. Lions. It consists in having sub-domains exchanging not only interface point flux values, but also interface flux 'derivative' values. (The word 'derivative' is here used with quotes, because in the case considered here, it in fact consists in the Ω.∇ operator, with Ω the angular variable vector and ∇ the spatial gradient operator.) A parameter α is introduced, as proportionality coefficient between point flux and 'derivative' values. This parameter can be tuned - so far heuristically - to optimize the method. (authors)

  13. Deterministic methods to solve the differential transport equation in neutronic

    International Nuclear Information System (INIS)

    We present a synthesis of the methods used to solve the integro-differential form of the transport equation. This form is used above all to achieve whole core calculations in 2 and 3D. First, we discretize the equation in energy and it leads us to an one group energy equation. For each of these groups, the scope of the calculation is so big that we have to calculate our solution on homogenized cells. On this homogenized medium, we describe different angular and spatial discretizations with acceleration methods. Finally we select some promising schemes to test: - SN Linear Nodal method with a Diffusion Synthetic Acceleration method; - Variational Nodal Method. These methods could be compared with more classical ones. To say, finite element or finite difference methods. (author). 65 refs., 3 annexes

  14. New contributions to neutron stochastic transport theory in the time and in the frequency domain

    International Nuclear Information System (INIS)

    The authors generalize the stochastic transport theory of Pal and Munoz-Cobo and Perez methodology, to include the delayed neutron effects. They apply this theory to interpret several experiments measuring the cross power spectral density G12(W), G13(W), G23(W) of three detectors 1, 2 and 3, located in and out of a tank containing a UO2F2 solution in water. (Auth.)

  15. Instrumentation development for magneto-transport and neutron scattering measurements at high pressure and low temperature

    OpenAIRE

    Wang, Weiwei

    2013-01-01

    High pressure, high magnetic field and low temperature techniques are required to investigate magnetic transitions and quantum critical behaviour in different ferromagnetic materials to elucidate how novel forms of superconductivity and other new states are brought about. In this project, several instruments for magneto-transport and neutron scattering measurements have been designed and built. They include inserts for a dilution refrigerator and pressure cells for resistivity,...

  16. Systems guide to MCNP (Monte Carlo Neutron and Photon Transport Code)

    International Nuclear Information System (INIS)

    The subject of this report is the implementation of the Los Alamos National Laboratory Monte Carlo Neutron and Photon Transport Code - Version 3 (MCNP) on the different types of computer systems, especially the IBM MVS system. The report supplements the documentation of the RSIC computer code package CCC-200/MCNP. Details of the procedure to follow in executing MCNP on the IBM computers, either in batch mode or interactive mode, are provided

  17. Normal and adjoint integral and integrodifferential neutron transport equations. Pt. 2

    International Nuclear Information System (INIS)

    Using the simplifying hypotheses of the integrodifferential Boltzmann equations of neutron transport, given in JEN 334 report, several integral equations, and theirs adjoint ones, are obtained. Relations between the different normal and adjoint eigenfunctions are established and, in particular, proceeding from the integrodifferential Boltzmann equation it's found out the relation between the solutions of the adjoint equation of its integral one, and the solutions of the integral equation of its adjoint one (author)

  18. Modular, object-oriented redesign of a large-scale Monte Carlo neutron transport program

    International Nuclear Information System (INIS)

    This paper describes the modular, object-oriented redesign of a large-scale Monte Carlo neutron transport program. This effort represents a complete 'white sheet of paper' rewrite of the code. In this paper, the motivation driving this project, the design objectives for the new version of the program, and the design choices and their consequences will be discussed. The design itself will also be described, including the important subsystems as well as the key classes within those subsystems

  19. From the similarities between neutrons and radon to advanced radon-detection and improved cold fusion neutron-measurements

    Science.gov (United States)

    Tommasino, L.; Espinosa, G.

    2014-07-01

    Neutrons and radon are both ubiquitous in the earth's crust. The neutrons of terrestrial origin are strongly related to radon since they originate mainly from the interactions between the alpha particles from the decays of radioactive-gas (namely Radon and Thoron) and the light nuclei. Since the early studies in the field of neutrons, the radon gas was used to produce neutrons by (α, n) reactions in beryllium. Another important similarity between radon and neutrons is that they can be detected only through the radiations produced respectively by decays or by nuclear reactions. These charged particles from the two distinct nuclear processes are often the same (namely alpha-particles). A typical neutron detector is based on a radiator facing a alpha-particle detector, such as in the case of a neutron film badge. Based on the similarity between neutrons and radon, a film badge for radon has been recently proposed. The radon film badge, in addition to be similar, may be even identical to the neutron film badge. For these reasons, neutron measurements can be easily affected by the presence of unpredictable large radon concentration. In several cold fusion experiments, the CR-39 plastic films (typically used in radon and neutron film-badges), have been the detectors of choice for measuring neutrons. In this paper, attempts will be made to prove that most of these neutron-measurements might have been affected by the presence of large radon concentrations.

  20. Advanced LIGO Constraints on Neutron Star Mergers and R-Process Sites

    CERN Document Server

    Côté, Benoit; Fryer, Chris L; Ritter, Christian; Paul, Adam; Wehmeyer, Benjamin; O'Shea, Brian W

    2016-01-01

    The role of compact binary mergers as the main production site of r-process elements is investigated by combining stellar abundances of Eu observed in the Milky Way, galactic chemical evolution (GCE) simulations, binary population synthesis models, and Advanced LIGO gravitational wave measurements. We compiled and reviewed seven recent GCE studies to extract the frequency of neutron star - neutron star (NS-NS) mergers that is needed in order to reproduce the observed [Eu/Fe] vs [Fe/H] relationship. We used our simple chemical evolution code to explore the impact of different analytical delay-time distribution (DTD) functions for NS-NS mergers. We then combined our metallicity-dependent population synthesis models with our chemical evolution code to bring their predictions, for both NS-NS mergers and black hole - neutron star mergers, into a GCE context. Finally, we convolved our results with the cosmic star formation history to provide a direct comparison with current and upcoming Advanced LIGO measurements. ...

  1. Numerical research on the anisotropic transport of thermal neutron in heterogeneous porous media with micron X-ray computed tomography

    OpenAIRE

    Yong Wang; Wenzheng Yue; Mo Zhang

    2016-01-01

    The anisotropic transport of thermal neutron in heterogeneous porous media is of great research interests in many fields. In this paper, it is the first time that a new model based on micron X-ray computed tomography (CT) has been proposed to simultaneously consider both the separation of matrix and pore and the distribution of mineral components. We apply the Monte Carlo method to simulate thermal neutrons transporting through the model along different directions, and meanwhile detect those ...

  2. Early Advanced LIGO binary neutron-star sky localization and parameter estimation

    CERN Document Server

    Berry, C P L; Farr, W M; Haster, C-J; Mandel, I; Middleton, H; Singer, L P; Urban, A L; Vecchio, A; Vitale, S; Cannon, K; Graff, P B; Hanna, C; Mohapatra, S; Pankow, C; Price, L R; Sidery, T; Veitch, J

    2016-01-01

    2015 will see the first observations of Advanced LIGO and the start of the gravitational-wave (GW) advanced-detector era. One of the most promising sources for ground-based GW detectors are binary neutron-star (BNS) coalescences. In order to use any detections for astrophysics, we must understand the capabilities of our parameter-estimation analysis. By simulating the GWs from an astrophysically motivated population of BNSs, we examine the accuracy of parameter inferences in the early advanced-detector era. We find that sky location, which is important for electromagnetic follow-up, can be determined rapidly (~5 s), but that sky areas may be hundreds of square degrees. The degeneracy between component mass and spin means there is significant uncertainty for measurements of the individual masses and spins; however, the chirp mass is well measured (typically better than 0.1%).

  3. Neutron and gamma ray transport calculations in shielding system

    Energy Technology Data Exchange (ETDEWEB)

    Masukawa, Fumihiro; Sakamoto, Hiroki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1998-03-01

    In the shields for radiation in nuclear facilities, the penetrating holes of various kinds and irregular shapes are made for the reasons of operation, control and others. These penetrating holes and gaps are filled with air or the substances with relatively small shielding performance, and radiation flows out through them, which is called streaming. As the calculation techniques for the shielding design or analysis related to the streaming problem, there are the calculations by simplified evaluation, transport calculation and Monte Carlo method. In this report, the example of calculation by Monte Carlo method which is represented by MCNP code is discussed. A number of variance reduction techniques which seem effective for the analysis of streaming problem were tried. As to the investigation of the applicability of MCNP code to streaming analysis, the object of analysis which are the concrete walls without hole and with horizontal hole, oblique hole and bent oblique hole, the analysis procedure, the composition of concrete, and the conversion coefficient of dose equivalent, and the results of analysis are reported. As for variance reduction technique, cell importance was adopted. (K.I.)

  4. Boron neutron capture therapy for advanced and/or recurrent cancers in the oral cavity

    International Nuclear Information System (INIS)

    This preliminary study of 5 patients with advanced and/or recurrent cancer in the oral cavity was performed to evaluate the effectiveness of Boron Neutron Capture Therapy (BNCT). The patients received therapy with the 10B-carrier p-boronophenylalanine (BPA) with or without borocaptate sodium (BSH) and irradiation thereafter with epithermal neutrons. All underwent 18F-BPA PET studies before receiving BNCT to determine the accumulation ratios of BPA in tumor and normal tissues. The tumor mass was decreased in size and at minimum a transient partial response was achieved in all cases, though rapid tumor re-growth was observed in 2. Although tentative clinical responses and improvements in quality of life were recognized, obliteration of the tumor was not obtained in any of the cases. Additional studies are required to determine the utility and indication of BNCT for oral cancer. (author)

  5. A portable, parallel, object-oriented Monte Carlo neutron transport code in C++

    Energy Technology Data Exchange (ETDEWEB)

    Lee, S.R.; Cummings, J.C. [Los Alamos National Lab., NM (United States); Nolen, S.D. [Texas A and M Univ., College Station, TX (United States)]|[Los Alamos National Lab., NM (United States)

    1997-05-01

    We have developed a multi-group Monte Carlo neutron transport code using C++ and the Parallel Object-Oriented Methods and Applications (POOMA) class library. This transport code, called MC++, currently computes k and {alpha}-eigenvalues and is portable to and runs parallel on a wide variety of platforms, including MPPs, clustered SMPs, and individual workstations. It contains appropriate classes and abstractions for particle transport and, through the use of POOMA, for portable parallelism. Current capabilities of MC++ are discussed, along with physics and performance results on a variety of hardware, including all Accelerated Strategic Computing Initiative (ASCI) hardware. Current parallel performance indicates the ability to compute {alpha}-eigenvalues in seconds to minutes rather than hours to days. Future plans and the implementation of a general transport physics framework are also discussed.

  6. The Saudi experience with neutron therapy in locally advanced head and neck cancers

    International Nuclear Information System (INIS)

    The neutron therapy program at King Faisal Specialist Hospital and Research Center conducted a phase II study to evaluate the toxicity and efficacy of neutrons against conventional external megavoltage irradiation in patients with locally advanced head and neck malignancy. One hundred and nineteen patients were allocated to receive either photons (46/119) or neutrons (73/119). Radiation effects were scored according to the EORTC/RTOG criteria; data was collected weekly during treatment, once a month for the first year and at 6-month intervals subsequently. While acute effects were scored in all patients, only 59 were evaluable for late effects and locoregional control. A composite of the average reaction results were obtained using this information, to compare them in time, for acute and late effects in both arms of the study. The maximum acute reactions in the two groups were similar. In the majority of the patients (80 %) acute skin and mucosal reactions occurred during the last week of treatment. The changes in the subcutaneous tissues and salivary glands became clinically apparent at 3 months or later. Salivary gland toxicity was more severe in the photon arm and the difference was statistically significant at 3 months (P + 0.04) but this was lost at 12 months. Late effects for skin and subcutaneous tissues were significantly more severe in the neutron arm with P values of 0.04 and 0.01 respectively. Three patients in the neutron arm died of grade 4 radiation complications. The local control and survival were similar in both groups. (author)

  7. Phase 1 environmental report for the Advanced Neutron Source at Oak Ridge National Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Blasing, T.J.; Brown, R.A.; Cada, G.F.; Easterly, C.; Feldman, D.L.; Hagan, C.W.; Harrington, R.M.; Johnson, R.O.; Ketelle, R.H.; Kroodsma, R.L.; McCold, L.N.; Reich, W.J.; Scofield, P.A.; Socolof, M.L.; Taleyarkhan, R.P.; Van Dyke, J.W.

    1992-02-01

    The US Department of Energy (DOE) has proposed the construction and operation of the Advanced Neutron Source (ANS), a 330-MW(f) reactor, at Oak Ridge National Laboratory (ORNL) to support neutron scattering and nuclear physics experiments. ANS would provide a steady-state source of neutrons that are thermalized to produce sources of hot, cold, and very coal neutrons. The use of these neutrons in ANS experiment facilities would be an essential component of national research efforts in basic materials science. Additionally, ANS capabilities would include production of transplutonium isotopes, irradiation of potential fusion and fission reactor materials, activation analysis, and production of medical and industrial isotopes such as {sup 252}Cf. Although ANS would not require licensing by the US Nuclear Regulatory Commission (NRC), DOE regards the design, construction, and operation of ANS as activities that would produce a licensable facility; that is, DOE is following the regulatory guidelines that NRC would apply if NRC were licensing the facility. Those guidelines include instructions for the preparation of an environmental report (ER), a compilation of available data and preliminary analyses regarding the environmental impacts of nuclear facility construction and operation. The ER, described and outlined in NRC Regulatory Guide 4.2, serves as a background document to facilitate the preparation of environmental impact statements (EISs). Using Regulatory Guide 4.2 as a model, this ANS ER provides analyses and information specific to the ANS site and area that can be adopted (and modified, if necessary) for the ANS EIS. The ER is being prepared in two phases. Phase 1 ER includes many of the data and analyses needed to prepare the EIS but does not include data or analyses of alternate sites or alternate technologies. Phase 2 ER will include the additional data and analyses stipulated by Regulatory Guide 4.2.

  8. Phase 1 environmental report for the Advanced Neutron Source at Oak Ridge National Laboratory

    International Nuclear Information System (INIS)

    The US Department of Energy (DOE) has proposed the construction and operation of the Advanced Neutron Source (ANS), a 330-MW(f) reactor, at Oak Ridge National Laboratory (ORNL) to support neutron scattering and nuclear physics experiments. ANS would provide a steady-state source of neutrons that are thermalized to produce sources of hot, cold, and very coal neutrons. The use of these neutrons in ANS experiment facilities would be an essential component of national research efforts in basic materials science. Additionally, ANS capabilities would include production of transplutonium isotopes, irradiation of potential fusion and fission reactor materials, activation analysis, and production of medical and industrial isotopes such as 252Cf. Although ANS would not require licensing by the US Nuclear Regulatory Commission (NRC), DOE regards the design, construction, and operation of ANS as activities that would produce a licensable facility; that is, DOE is following the regulatory guidelines that NRC would apply if NRC were licensing the facility. Those guidelines include instructions for the preparation of an environmental report (ER), a compilation of available data and preliminary analyses regarding the environmental impacts of nuclear facility construction and operation. The ER, described and outlined in NRC Regulatory Guide 4.2, serves as a background document to facilitate the preparation of environmental impact statements (EISs). Using Regulatory Guide 4.2 as a model, this ANS ER provides analyses and information specific to the ANS site and area that can be adopted (and modified, if necessary) for the ANS EIS. The ER is being prepared in two phases. Phase 1 ER includes many of the data and analyses needed to prepare the EIS but does not include data or analyses of alternate sites or alternate technologies. Phase 2 ER will include the additional data and analyses stipulated by Regulatory Guide 4.2

  9. Modelling of neutron and photon transport in iron and concrete radiation shieldings by the Monte Carlo method - Version 2

    CERN Document Server

    Žukauskaite, A; Plukiene, R; Plukis, A

    2007-01-01

    Particle accelerators and other high energy facilities produce penetrating ionizing radiation (neutrons and γ-rays) that must be shielded. The objective of this work was to model photon and neutron transport in various materials, usually used as shielding, such as concrete, iron or graphite. Monte Carlo method allows obtaining answers by simulating individual particles and recording some aspects of their average behavior. In this work several nuclear experiments were modeled: AVF 65 – γ-ray beams (1-10 MeV), HIMAC and ISIS-800 – high energy neutrons (20-800 MeV) transport in iron and concrete. The results were then compared with experimental data.

  10. Advances in implosion physics, alternative targets design, and neutron effects on heavy ion fusion reactors

    International Nuclear Information System (INIS)

    The coupling of a new radiation transport (RT) solver with an existing multimaterial fluid dynamics code (ARWEN) using Adaptive Mesh Refinement named DAFNE, has been completed. In addition, improvements were made to ARWEN in order to work properly with the RT code, and to make it user-friendlier, including new treatment of Equations of State, and graphical tools for visualization. The evaluation of the code has been performed, comparing it with other existing RT codes (including the one used in DAFNE, but in the single-grid version). These comparisons consist in problems with real input parameters (mainly opacities and geometry parameters). Important advances in Atomic Physics, Opacity calculations and NLTE atomic physics calculations, with participation in significant experiments in this area, have been obtained. Early published calculations showed that a DTx fuel with a small tritium initial content (xe and to enhance radiation losses, reducing the plasma temperature, Ti. The neutron activation of all natural elements in First Structural Wall (FSW) component of an Inertial Fusion Energy (IFE) reactor for waste management, and the analysis of activation of target debris in NIF-type facilities has been completed. Using an original efficient modeling for pulse activation, the FSW behavior in inertial fusion has been studied. A radiological dose library coupled to the ACAB code is being generated for assessing impact of environmental releases, and atmospheric dispersion analysis from HIF reactors indicate the uncertainty in tritium release parameters. The first recognition of recombination barriers in SiC, modify the understanding of the calculation of displacement per atom, dpa, to quantify the collisional damage. An important analysis has been the confirmation, using Molecular Dynamics (MD) with an astonishing agreement, of the experimental evidence of low-temperature amorphization by damage accumulation in SiC, which could modify extensively its viability as a

  11. A stochastic transport theory of neutron and photon coupled fields neutron and photon counting statistics in nuclear assemblies

    International Nuclear Information System (INIS)

    The behavior of neutrons and gamma rays in a nuclear reactor or configuration of fissile material can be represented as a stochastic process. The observation of this stochastic process is usually achieved by measuring the fluctuations of the neutron and gamma ray population on the system. The general theory of the stochastic neutron field has been developed to a high degree. However, the theory of the stochastic nature of the gamma rays and neutrons couples the two processes. The generalized probability balances are developed from which the first and higher moments of the neutron and gamma rays fields are obtained. The paper also provides a description of the probability generating functions for both photon and neutron detectors that are the foundations for measurements of the fluctuations. The formalism developed in this paper for the representation of the statistical descriptors of the neutron-photon coupled field is applicable for many neutron noise analysis measurements

  12. Follow-up fuel plate stability experiments and analyses for the Advanced Neutron Source

    Energy Technology Data Exchange (ETDEWEB)

    Swinson, W.F.; Battiste, R.L.; Luttrell, C.R.; Yahr, G.T.

    1993-11-01

    The reactor for the planned Advanced Neutron Source uses closely spaced plates cooled by heavy water flowing through narrow channels. Two sets of tests were performed on the upper and lower fuel plates for the structural response of the fuel plates to the required high coolant flow velocities. This report contains the data from the second round of tests. Results and conclusions from all of the tests are also included in this report. The tests were done using light water on full-scale epoxy models, and through model theory, the results were related to the prototype plates, which are aluminum-clad aluminum/uranium silicide involute-shaped plates.

  13. THE COMMISSIONING PLAN FOR THE SPALLATION NEUTRON SOURCE RING AND TRANSPORT LINES.

    Energy Technology Data Exchange (ETDEWEB)

    RAPARIA,D.BLASKIEWICZ,M.LEE,Y.Y.WEI,J.ET AL.

    2004-03-10

    The Spallation Neutron Source (SNS) accelerator systems will provide a 1 GeV, 1.44 MW proton beam to a liquid mercury target for neutron production. In order to satisfy the accelerator systems' portion of the Critical Decision 4 (CD-4) commissioning goal (which marks the completion of the construction phase of the project), a beam pulse with intensity greater than 1 x 10{sup 13} protons must be accumulated in the ring, extracted in a single turn and delivered to the target. A commissioning plan has been formulated for bringing into operation and establishing nominal operating conditions for the various ring and transport line subsystems as well as for establishing beam conditions and parameters which meet the commissioning goal.

  14. Transport equations and linear response of superfluid Fermi mixtures in neutron stars

    CERN Document Server

    Gusakov, M E

    2010-01-01

    We study transport properties of a strongly interacting superfluid mixture of two Fermi-liquids. A typical example of such matter is the neutron-proton liquid in the cores of neutron stars. To describe the mixture, we employ the Landau theory of Fermi-liquids, generalized to allow for the effects of superfluidity. We formulate the kinetic equation and analyze linear response of the system to vector (e.g., electromagnetic) perturbation. In particular, we calculate the transverse and longitudinal polarization functions for both liquid components. We demonstrate, that they can be expressed through the Landau parameters of the mixture and polarization functions of noninteracting matter (when the Landau quasiparticle interaction is neglected). Our results can be used, e.g., for studies of the kinetic coefficients and low-frequency long-wavelength collective modes in superfluid Fermi-mixtures.

  15. MCNP: a general Monte Carlo code for neutron and photon transport

    International Nuclear Information System (INIS)

    The general-purpose Monte Carlo code MCNP ca be used for neutron, photon, or coupled neutron-photon transport, including the capability to calculate eigenvalues for critical systems. The code treats an arbitrary three-dimensional configuration of materials in geometric cells bounded by first- and second-degree surfaces and some special fourth-degree surfaces (elliptical tori). Pointwise cross-section data are used. For neutrons, all reactions given in a particular cross-section evaluation are accounted for. Thermal neutrons are described by both the free-gas and S(α,β) models. For photons, the code takes account of incoherent and coherent scattering, the possibility of fluorescent emission following photoelectric absorption, and absorption in pair production with local emission of annihilation radiation. MCNP includes an elaborate, interactive plotting capability that allows the user to view his input geometry to help check for setup errors. Standard features which are available to improve computational efficiency include geometry splitting and Russian roulette, weight cutoff with Russian roulette, correlated sampling, analog capture or capture by weight reduction, the exponential transformation, energy splitting, forced collisions in designated cells, flux estimates at point or ring detectors, deterministically transporting pseudo-particles to designated regions, track-length estimators, source biasing, and several parameter cutoffs. Extensive summary information is provided to help the user better understand the physics and Monte Carlo simulation of his problem. The standard, user-defined output of MCNP includes two-way current as a function of direction across any set of surfaces or surface segments in the problem. Flux across any set of surfaces or surface segments is available. 58 figures, 28 tables

  16. Initial global 2-D shielding analysis for the Advanced Neutron Source core and reflector

    Energy Technology Data Exchange (ETDEWEB)

    Bucholz, J.A.

    1995-08-01

    This document describes the initial global 2-D shielding analyses for the Advanced Neutron Source (ANS) reactor, the D{sub 2}O reflector, the reflector vessel, and the first 200 mm of light water beyond the reflector vessel. Flux files generated here will later serve as source terms in subsequent shielding analyses. In addition to reporting fluxes and other data at key points of interest, a major objective of this report was to document how these analyses were performed, the phenomena that were included, and checks that were made to verify that these phenomena were properly modeled. In these shielding analyses, the fixed neutron source distribution in the core was based on the `lifetime-averaged` spatial power distribution. Secondary gamma production cross sections in the fuel were modified so as to account intrinsically for delayed fission gammas in the fuel as well as prompt fission gammas. In and near the fuel, this increased the low-energy gamma fluxes by 50 to 250%, but out near the reflector vessel, these same fluxes changed by only a few percent. Sensitivity studies with respect to mesh size were performed, and a new 2-D mesh distribution developed after some problems were discovered with respect to the use of numerous elongated mesh cells in the reflector. All of the shielding analyses were performed sing the ANSL-V 39n/44g coupled library with 25 thermal neutron groups in order to obtain a rigorous representation of the thermal neutron spectrum throughout the reflector. Because of upscatter in the heavy water, convergence was very slow. Ultimately, the fission cross section in the various materials had to be artificially modified in order to solve this fixed source problem as an eigenvalue problem and invoke the Vondy error-mode extrapolation technique which greatly accelerated convergence in the large 2-D RZ DORT analyses. While this was quite effective, 150 outer iterations (over energy) were still required.

  17. Dosimetric evaluation of neutron capture therapy for local advanced breast cancer

    Energy Technology Data Exchange (ETDEWEB)

    Yanagie, H. [Department of Nuclear Engineering and Management, Graduate School of Engineering, University of Tokyo, Tokyo (Japan); Cooperative Unit of Medicine and Engineering, University of Tokyo Hospital, Tokyo (Japan)], E-mail: yanagie@n.t.u-tokyo.ac.jp; Kumada, H. [Japan Atomic Research Institute, Ibaraki (Japan); Sakurai, Y. [Research Reactor Institute, Kyoto University, Osaka (Japan); Nakamura, T. [Japan Atomic Research Institute, Ibaraki (Japan); Department of Nuclear Physics, Ibaraki University, Ibaraki (Japan); Furuya, Y. [Department of Surgery, Satukidai Hospital, Chiba (Japan); Sugiyama, H. [Cooperative Unit of Medicine and Engineering, University of Tokyo Hospital, Tokyo (Japan); Ono, K. [Research Reactor Institute, Kyoto University, Osaka (Japan); Takamoto, S. [Cooperative Unit of Medicine and Engineering, University of Tokyo Hospital, Tokyo (Japan); Department of Cardiac Surgery, University of Tokyo Hospital, Tokyo (Japan); Eriguchi, M. [Cooperative Unit of Medicine and Engineering, University of Tokyo Hospital, Tokyo (Japan); Department of Microbiology, Syowa University School of Pharmaceutical Sciences, Tokyo (Japan); Takahashi, H. [Department of Nuclear Engineering and Management, Graduate School of Engineering, University of Tokyo, Tokyo (Japan); Cooperative Unit of Medicine and Engineering, University of Tokyo Hospital, Tokyo (Japan)

    2009-07-15

    Local recurrence breast cancer is one of the most difficult conditions to cure and there is a need for new therapy. If sufficient boron compound can be targeted to the tumor, boron neutron capture therapy (BNCT) can be applied to local recurrent breast cancer. In this study, we performed a preliminary dosimetry with a phantom model of the mammary gland at Kyoto University Research Reactor (KUR), and a feasibility dosimetry with JAERI Computational Dosimetry System (JCDS) at JRR4 reactor of Japan Atomic Research Institute. We performed preliminary dosimetry of a phantom model of the mammary gland with thermal neutron irradiation (OO-0011 mode) on LiF collimation at KUR. The thermal neutron flux was 5.16 E+08 cm{sup -2} s{sup -1} at the surface of phantom. The blood boron concentration is estimated to be 30 ppm; tumor boron concentration is also estimated to be 90 ppm according to tumor/blood ratio 3 and skin/blood ratio 1.2. Tumor RBE dose is estimated to be 47 Gy/h, and skin RBE dose is 12.4 Gy/h. In case of advanced breast cancer, we performed the feasibility estimation of 3D construction of tumor according to the MRI imaging of a patient with epithermal neutron mode at JRR4. The blood boron concentration (ppm) and tumor/normal tissue ratio are estimated to be 24 and 3.5, respectively. Skin RBE dose is restricted to 10 Gy/h, the maximum tumor RBE dose, minimum tumor RBE dose, and mean tumor RBE dose are 42.2, 11.3, and 28.9 Gy-Eq, respectively, in half hour irradiation. In this study, we showed the possibility to apply BNCT to local recurrent breast cancer. We can irradiate tumors selectively and as safely as possible, reducing the effects on neighboring healthy tissues.

  18. Monte Carlo transport simulation for a long counter neutron detector employed as a cosmic rays induced neutron monitor at ground level

    Energy Technology Data Exchange (ETDEWEB)

    Pazianotto, Mauricio Tizziani; Carlson, Brett Vern [Instituto Tecnologico de Aeronautica (ITA), Sao Jose dos Campos, SP (Brazil); Federico, Claudio Antonio; Goncalez, Odair Lelis [Centro Tecnico Aeroespacial (CTA), Sao Jose dos Campos, SP (Brazil). Instituto de Estudos Avancados

    2011-07-01

    Full text: Great effort is required to understand better the cosmic radiation (CR) dose received by sensitive equipment, on-board computers and aircraft crew members at Brazil airspace, because there is a large area of South America and Brazil subject to the South Atlantic Anomaly (SAA). High energy neutrons are produced by interactions between primary cosmic ray and atmospheric atoms, and also undergo moderation resulting in a wider spectrum of energy ranging from thermal energies (0:025eV ) to energies of several hundreds of MeV. Measurements of the cosmic radiation dose on-board aircrafts need to be followed with an integral flow monitor on the ground level in order to register CR intensity variations during the measurements. The Long Counter (LC) neutron detector was designed as a directional neutron flux meter standard because it presents fairly constant response for energy under 10MeV. However we would like to use it as a ground based neutron monitor for cosmic ray induced neutron spectrum (CRINS) that presents an isotropic fluency and a wider spectrum of energy. The LC was modeled and tested using a Monte Carlo transport simulation for irradiations with known neutron sources ({sup 241}Am-Be and {sup 251}Cf) as a benchmark. Using this geometric model its efficiency was calculated to CRINS isotropic flux, introducing high energy neutron interactions models. The objective of this work is to present the model for simulation of the isotropic neutron source employing the MCNPX code (Monte Carlo N-Particle eXtended) and then access the LC efficiency to compare it with experimental results for cosmic ray neutrons measures on ground level. (author)

  19. Neutronics methods, models, and applications at the Idaho National Engineering Laboratory for the advanced neutron source reactor three-element core design

    International Nuclear Information System (INIS)

    A summary of the methods and models used to perform neutronics analyses on the Advanced Neutron Source reactor three-element core design is presented. The applications of the neutral particle Monte Carlo code MCNP are detailed, as well as the expansion of the static role of MCNP to analysis of fuel cycle depletion calculations. Results to date of these applications are presented also. A summary of the calculations not yet performed is also given to provide a open-quotes to-doclose quotes list if the project is resurrected

  20. GNES-R: Global nuclear energy simulator for reactors task 1: High-fidelity neutron transport

    International Nuclear Information System (INIS)

    A multi-laboratory, multi-university collaboration has formed to advance the state-of-the-art in high-fidelity, coupled-physics simulation of nuclear energy systems. We are embarking on the first-phase in the development of a new suite of simulation tools dedicated to the advancement of nuclear science and engineering technologies. We seek to develop and demonstrate a new generation of multi-physics simulation tools that will explore the scientific phenomena of tightly coupled physics parameters within nuclear systems, support the design and licensing of advanced nuclear reactors, and provide benchmark quality solutions for code validation. In this paper, we have presented the general scope of the collaborative project and discuss the specific challenges of high-fidelity neutronics for nuclear reactor simulation and the inroads we have made along this path. The high-performance computing neutronics code system utilizes the latest version of SCALE to generate accurate, problem-dependent cross sections, which are used in NEWTRNX - a new 3-D, general-geometry, discrete-ordinates solver based on the Slice-Balance Approach. The Global Nuclear Energy Simulator for Reactors (GNES-R) team is embarking on a long-term simulation development project that encompasses multiple laboratories and universities for the expansion of high-fidelity coupled-physics simulation of nuclear energy systems. (authors)

  1. 3-D Deep Penetration Neutron Imaging of Thick Absorgin and Diffusive Objects Using Transport Theory

    Energy Technology Data Exchange (ETDEWEB)

    Ragusa, Jean; Bangerth, Wolfgang

    2011-08-01

    here explores the inverse problem of optical tomography applied to heterogeneous domains. The neutral particle transport equation was used as the forward model for how neutral particles stream through and interact within these heterogeneous domains. A constrained optimization technique that uses Newtons method served as the basis of the inverse problem. Optical tomography aims at reconstructing the material properties using (a) illuminating sources and (b) detector readings. However, accurate simulations for radiation transport require that the particle (gamma and/or neutron) energy be appropriate discretize in the multigroup approximation. This, in turns, yields optical tomography problems where the number of unknowns grows (1) about quadratically with respect to the number of energy groups, G, (notably to reconstruct the scattering matrix) and (2) linearly with respect to the number of unknown material regions. As pointed out, a promising approach could rely on algorithms to appropriately select a material type per material zone rather than G2 values. This approach, though promising, still requires further investigation: (a) when switching from cross-section values unknowns to material type indices (discrete integer unknowns), integer programming techniques are needed since derivative information is no longer available; and (b) the issue of selecting the initial material zoning remains. The work reported here proposes an approach to solve the latter item, whereby a material zoning is proposed using one-group or few-groups transport approximations. The capabilities and limitations of the presented method were explored; they are briefly summarized next and later described in fuller details in the Appendices. The major factors that influenced the ability of the optimization method to reconstruct the cross sections of these domains included the locations of the sources used to illuminate the domains, the number of separate experiments used in the reconstruction, the

  2. An advanced control system for a next generation transport aircraft

    Science.gov (United States)

    Rising, J. J.; Davis, W. J; Grantham, W. D.

    1983-01-01

    The use of modern control theory to develop a high-authority stability and control system for the next generation transport aircraft is described with examples taken from work performed on an advanced pitch active control system (PACS). The PACS was configured to have short-period and phugoid modes frequency and damping characteristics within the shaded S-plane areas, column force gradients with set bounds and with constant slope, and a blended normal-acceleration/pitch rate time history response to a step command. Details of the control law, feedback loop, and modal control syntheses are explored, as are compensation for the feedback gain, the deletion of the velocity signal, and the feed-forward compensation. Scheduling of the primary and secondary gains are discussed, together with control law mechanization, flying qualities analyses, and application on the L-1011 aircraft.

  3. Preliminary study on CAD-based method of characteristics for neutron transport calculation

    CERN Document Server

    Chen, Zhen-Ping; Sun, Guang-Yao; Song, Jing; Hao, Li-Juan; Hu, Li-Qin; Wu, Yi-Can

    2013-01-01

    The method of characteristics (MOC) is widely used for neutron transport calculation in recent decades. However, the key problem determining whether MOC can be applied in highly heterogeneous geometry is how to combine an effective geometry modeling method with it. Most of the existing MOC codes conventionally describe the geometry model just by lines and arcs with extensive input data. Thus they have difficulty in geometry modeling and ray tracing for complicated geometries. In this study, a new method making use of a CAD-based automatic modeling tool MCAM which is a CAD/Image-based Automatic Modeling Program for Neutronics and Radiation Transport developed by FDS Team in China was introduced for geometry modeling and ray tracing of particle transport to remove those limitations. The diamond -difference scheme was applied to MOC to reduce the spatial discretization errors of the flat flux approximation. Based on MCAM and MOC, a new MOC code was developed and integrated into SuperMC system, whic h is a Super ...

  4. Advanced sources and optical components for the McStas neutron scattering instrument simulation package

    DEFF Research Database (Denmark)

    Farhi, E.; Monzat, C.; Arnerin, R.;

    2014-01-01

    We present new McStas components Virtual_mcnp_input and Virtual_tripoli4_input, Virtual_mcnp_output and Virtual_tripoli_output to be used as interface for the MCNP and Tripoli neutron transport codes. Similarly, the new Lens component can be used to describe any refracting material set-up, includ......We present new McStas components Virtual_mcnp_input and Virtual_tripoli4_input, Virtual_mcnp_output and Virtual_tripoli_output to be used as interface for the MCNP and Tripoli neutron transport codes. Similarly, the new Lens component can be used to describe any refracting material set...... be enclosed in a scattering material in order to model the absorption and scattering in the detector housing, prior to the actual detection. An extended model of the IN5b time-of-flight spectrometer at the Institut Laue Langevin is used to simulate vanadium and powder diffractograms, making use of the gas...

  5. Application of synthetic diffusion method in the numerical solution of the equations of neutron transport in slab geometry

    International Nuclear Information System (INIS)

    One of the main problems in reactor physics is to determine the neutron distribution in reactor core, since knowing that, it is possible to calculate the rapidity of occurrence of different nuclear reaction inside the reactor core. Within different theories existing in nuclear reactor physics, is neutron transport the one in which equation who govern the exact behavior of neutronic distribution are developed even inside the proper neutron transport theory, there exist different methods of solution which are approximations to exact solution; still more, with the purpose to reach a more precise solution, the majority of methods have been approached to the obtention of solutions in numerical form with the aim of take the advantages of modern computers, and for this reason a great deal of effort is dedicated to numerical solution of the equations of neutron transport. In agreement with the above mentioned, in this work has been developed a computer program which uses a relatively new techniques known as 'acceleration of synthetic diffusion' which has been applied to solve the neutron transport equation with 'classical schemes of spatial integration' obtaining results with a smaller quantity of interactions, if they compare to done without using such equation (Author)

  6. Domain decomposition and terabyte tallies with the OpenMC Monte Carlo neutron transport code

    International Nuclear Information System (INIS)

    Memory limitations are a key obstacle to applying Monte Carlo neutron transport methods to high-fidelity full-core reactor analysis. Billions of unique regions are needed to carry out full-core depletion and fuel performance analyses, equating to terabytes of memory for isotopic abundances and tally scores - far more than can fit on a single computational node in modern architectures. This work introduces an implementation of domain decomposition that addresses this problem, demonstrating excellent scaling up to a 2.39TB mesh-tally distributed across 512 compute nodes running a full-core reactor benchmark on the Mira Blue Gene/Q supercomputer at Argonne National Laboratory. (author)

  7. Solution of the neutron transport problem with anisotropic scattering in cylindrical geometry by the decomposition method

    Energy Technology Data Exchange (ETDEWEB)

    Goncalves, G.A. [UFRGS, Departamento de Engenharia Nuclear, Av. Osvaldo Aranha 99, 4o andar, Porto Alegre, RS 90046-900 (Brazil); Bogado Leite, S.Q. [Comissao Nacional de Energia Nuclear, Coordenacao Geral de Reatores e Ciclo do Combustivel, Rua General Severiano, 90, Rio de Janeiro, RJ 22294-900 (Brazil)], E-mail: bogado@cnen.gov.br; Vilhena, M.T. de [UFRGS, Departamento de Matematica Aplicada, Av. Bento Goncalves, 9500, Porto Alegre, RS 91509-900 (Brazil)

    2009-01-15

    An analytical solution has been obtained for the one-speed stationary neutron transport problem, in an infinitely long cylinder with anisotropic scattering by the decomposition method. Series expansions of the angular flux distribution are proposed in terms of suitably constructed functions, recursively obtainable from the isotropic solution, to take into account anisotropy. As for the isotropic problem, an accurate closed-form solution was chosen for the problem with internal source and constant incident radiation, obtained from an integral transformation technique and the F{sub N} method.

  8. Advanced fuel cells for transportation applications. Final report

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-02-10

    This Research and Development (R and D) contract was directed at developing an advanced technology compressor/expander for supplying compressed air to Proton Exchange Membrane (PEM) fuel cells in transportation applications. The objective of this project was to develop a low-cost high-efficiency long-life lubrication-free integrated compressor/expander utilizing scroll technology. The goal of this compressor/expander was to be capable of providing compressed air over the flow and pressure ranges required for the operation of 50 kW PEM fuel cells in transportation applications. The desired ranges of flow, pressure, and other performance parameters were outlined in a set of guidelines provided by DOE. The project consisted of the design, fabrication, and test of a prototype compressor/expander module. The scroll CEM development program summarized in this report has been very successful, demonstrating that scroll technology is a leading candidate for automotive fuel cell compressor/expanders. The objectives of the program are: develop an integrated scroll CEM; demonstrate efficiency and capacity goals; demonstrate manufacturability and cost goals; and evaluate operating envelope. In summary, while the scroll CEM program did not demonstrate a level of performance as high as the DOE guidelines in all cases, it did meet the overriding objectives of the program. A fully-integrated, low-cost CEM was developed that demonstrated high efficiency and reliable operation throughout the test program. 26 figs., 13 tabs.

  9. Consumer Views on Transportation and Advanced Vehicle Technologies

    Energy Technology Data Exchange (ETDEWEB)

    Singer, Mark [National Renewable Energy Lab. (NREL), Golden, CO (United States)

    2015-09-01

    Vehicle manufacturers, U.S. Department of Energy laboratories, universities, private researchers, and organizations from countries around the globe are pursuing advanced vehicle technologies that aim to reduce gasoline and diesel consumption. This report details study findings of broad American public sentiments toward issues surrounding advanced vehicle technologies and is supported by the U.S. Department of Energy Vehicle Technology Office (VTO) in alignment with its mission to develop and deploy these technologies to improve energy security, increase mobility flexibility, reduce transportation costs, and increase environmental sustainability. Understanding and tracking consumer sentiments can influence the prioritization of development efforts by identifying barriers to and opportunities for broad acceptance of new technologies. Predicting consumer behavior toward developing technologies and products is inherently inexact. A person's stated preference given in an interview about a hypothetical setting may not match the preference that is demonstrated in an actual situation. This difference makes tracking actual consumer actions ultimately more valuable in understanding potential behavior. However, when developing technologies are not yet available and actual behaviors cannot be tracked, stated preferences provide some insight into how consumers may react in new circumstances. In this context this report provides an additional source to validate data and a new resource when no data are available. This report covers study data captured from December 2005 through June 2015 relevant to VTO research efforts at the time of the studies. Broadly the report covers respondent sentiments about vehicle fuel economy, future vehicle technology alternatives, ethanol as a vehicle fuel, plug-in electric vehicles, and willingness to pay for vehicle efficiency. This report represents a renewed effort to publicize study findings and make consumer sentiment data available to

  10. The EU power plant conceptual study - neutronic design analyses for near term and advanced reactor models

    International Nuclear Information System (INIS)

    A power plant conceptual study (PPCS) has been conducted in the framework of the European fusion programme with the main objective to demonstrate the safety and environmental advantages and the economic viability of fusion power. Power plant models with limited (''near term concepts'') and advanced plasma physics and technological extrapolations (''advanced concepts'') were considered. Two near term plant models were selected, one employing a water cooled lithium-lead (WCLL), and the other one a helium cooled pebble bed (HCPB) blanket. Two variants were also considered for the advanced power plant models, one adopting a liquid metal blanket with a self-cooled lithium-lead breeder zone and a helium cooled steel structure (''dual coolant lithium lead'', DCLL), and the other one a self-cooled lithium-lead (SCLL) blanket with SiCf/SiC composite as structural material. This report provides a detailed documentation of the neutronics design analyses performed as part of the PPCS study for both the near term and advanced power plant models. Main issues are the assessment of the tritium breeding capability, the evaluation of the nuclear power generation and its spatial distribution, and the assessment and optimisation of the shielding performance. The analyses were based on three-dimensional Monte Carlo calculations with the MCNP code using suitable torus sector models developed for the different PPCS plant variants. (orig.)

  11. Neutronic Analysis of Advanced SFR Burner Cores using Deep-Burn PWR Spent Fuel TRU Feed

    International Nuclear Information System (INIS)

    In this work, an advanced sodium-cooled fast TRU (Transuranics) burner core using deep-burn TRU feed composition discharged from small LWR cores was neutronically analyzed to show the effects of deeply burned TRU feed composition on the performances of sodium-cooled fast burner core. We consider a nuclear park that is comprised of the commercial PWRs, small PWRs of 100MWe for TRU deep burning using FCM (Fully Ceramic Micro-encapsulated) fuels and advanced sodium-cooled fast burners for their synergistic combination for effective TRU burning. In the small PWR core having long cycle length of 4.0 EFPYs, deep burning of TRU up to 35% is achieved with FCM fuel pins whose TRISO particle fuels contain TRUs in their central kernel. In this paper, we analyzed the performances of the advanced SFR burner cores using TRU feeds discharged from the small long cycle PWR deep-burn cores. Also, we analyzed the effect of cooling time for the TRU feeds on the SFR burner core. The results showed that the TRU feed composition from FCM fuel pins of the small long cycle PWR core can be effectively used into the advanced SFR burner core by significantly reducing the burnup reactivity swing which reduces smaller number of control rod assemblies to satisfy all the conditions for the self controllability than the TRU feed composition discharged from the typical PWR cores

  12. Evaluation of HYLIFE-II and Sombrero using 175- and 566-group neutron transport and activation cross sections

    Energy Technology Data Exchange (ETDEWEB)

    Latkowski, Jeffery F. E-mail: latkowski1@llnl.gov; Cullen, Dermott E.; Sanz, Javier

    2000-11-01

    Recent modifications to the TART Monte Carlo neutron and photon transport code allow enable calculation of 566-group neutron spectra. This expanded group structure represents a significant improvement over the 50- and 175-group structures that have been previously available. To support use of this new capability, neutron activation cross-section libraries have been created in the 175- and 566-group structures starting from the FENDL/A-2.0 pointwise data. Neutron spectra have been calculated for the first walls of the HYLIFE-II and Sombrero inertial fusion energy power plant designs and have been used in subsequent neutron activation calculations. The results obtained using the two different group structures are compared with each other as well as to those obtained using a 175-group version of the EAF3.1 activation cross-section library.

  13. Modeling of neutron and photon transport in iron and concrete radiation shields by using Monte Carlo method

    CERN Document Server

    Žukauskaitėa, A; Plukienė, R; Ridikas, D

    2007-01-01

    Particle accelerators and other high energy facilities produce penetrating ionizing radiation (neutrons and γ-rays) that must be shielded. The objective of this work was to model photon and neutron transport in various materials, usually used as shielding, such as concrete, iron or graphite. Monte Carlo method allows obtaining answers by simulating individual particles and recording some aspects of their average behavior. In this work several nuclear experiments were modeled: AVF 65 (AVF cyclotron of Research Center of Nuclear Physics, Osaka University, Japan) – γ-ray beams (1-10 MeV), HIMAC (heavy-ion synchrotron of the National Institute of Radiological Sciences in Chiba, Japan) and ISIS-800 (ISIS intensive spallation neutron source facility of the Rutherford Appleton laboratory, UK) – high energy neutron (20-800 MeV) transport in iron and concrete. The calculation results were then compared with experimental data.compared with experimental data.

  14. Automatic braking system modification for the Advanced Transport Operating Systems (ATOPS) Transportation Systems Research Vehicle (TSRV)

    Science.gov (United States)

    Coogan, J. J.

    1986-01-01

    Modifications were designed for the B-737-100 Research Aircraft autobrake system hardware of the Advanced Transport Operating Systems (ATOPS) Program at Langley Research Center. These modifications will allow the on-board flight control computer to control the aircraft deceleration after landing to a continuously variable level for the purpose of executing automatic high speed turn-offs from the runway. A bread board version of the proposed modifications was built and tested in simulated stopping conditions. Test results, for various aircraft weights, turnoff speed, winds, and runway conditions show that the turnoff speeds are achieved generally with errors less than 1 ft/sec.

  15. Validation and benchmarking of calculation methods for photon and neutron transport at cask configurations

    International Nuclear Information System (INIS)

    The reliability of calculation tools to evaluate and calculate dose rates appearing behind multi-layered shields is important with regard to the certification of transport and storage casks. Actual benchmark databases like SINBAD do not offer such configurations because they were developed for reactor and accelerator purposes. Due to this, a bench-mark-suite based on own experiments that contain dose rates measured in different distances and levels from a transport and storage cask and on a public benchmark to validate Monte-Carlo-transport-codes has been developed. The analysed and summarised experiments include a 60Co point-source located in a cylindrical cask, a 252Cf line-source shielded by iron and polyethylene (PE) and a bare 252Cf source moderated by PE in a concrete-labyrinth with different inserted shielding materials to quantify neutron streaming effects on measured dose rates. In detail not only MCNPTM (version 5.1.6) but also MAVRIC, included in the SCALE 6.1 package, have been compared for photon and neutron transport. Aiming at low deviations between calculation and measurement requires precise source term specification and exact measurements of the dose rates which have been evaluated carefully including known uncertainties. In MAVRIC different source-descriptions with respect to the group-structure of the nuclear data library are analysed for the calculation of gamma dose rates because the energy lines of 60Co can only be modelled in groups. In total the comparison shows that MCNPTM fits very wall to the measurements within up to two standard deviations and that MAVRIC behaves similarly under the prerequisite that the source-model can be optimized. (author)

  16. McSUB, a Monte Carlo Library for neutron transport in two different media for the neutron energy range 0.1-20 MeV

    International Nuclear Information System (INIS)

    Library McSUB is a package of easy-to-use subroutines and functions treating neutron transport in two different kind of media by Monte Carlo calculations. The first medium, D0, contains deuterium and natural carbon while the second medium, D1, contains hydrogen and natural carbon. In the neutron energy interval 0.1-20 MeV eight different kinds of interactions are considered: Elastic and (n,2n) interactions with deuterium, elastic interactions with hydrogen and elastic and inelastic interactions with natural carbon. The inelastic interaction with carbon are subdivided into four different interaction classes, one for each excited state of the recoiled carbon nucleus. The neutron cross sections and Legendre coefficients (expressing differential cross sections) have been supplied by NEA Data Bank in France. (author)

  17. Advanced Neutron Source enrichment study. Volume 2: Appendices -- Final report, Revision 12/94

    International Nuclear Information System (INIS)

    A study has been performed of the impact on performance of using low enriched uranium (20% 235U) or medium enriched uranium (35% 235U) as an alternative fuel for the Advanced Neutron Source, which is currently designed to use uranium enriched to 93% 235U. Higher fuel densities and larger volume cores were evaluated at the lower enrichments in terms of impact on neutron flux, safety, safeguards, technical feasibility, and cost. The feasibility of fabricating uranium silicide fuel at increasing material density was specifically addressed by a panel of international experts on research reactor fuels. The most viable alternative designs for the reactor at lower enrichments were identified and discussed. Several sensitivity analyses were performed to gain an understanding of the performance of the reactor at parametric values of power, fuel density, core volume, and enrichment that were interpolations between the boundary values imposed on the study or extrapolations from known technology. Volume 2 of this report contains 26 appendices containing results, meeting minutes, and fuel panel presentations. There are 26 appendices in this volume

  18. NATO Advanced Study Institute on Chemical Crystallography with Pulsed Neutrons and Synchrotron X-Rays

    CERN Document Server

    Jeffrey, George

    1988-01-01

    X-ray and neutron crystallography have played an increasingly impor­ tant role in the chemical and biochemical sciences over the past fifty years. The principal obstacles in this methodology, the phase problem and com­ puting, have been overcome. The former by the methods developed in the 1960's and just recognised by the 1985 Chemistry Nobel Prize award to Karle and Hauptman, the latter by the dramatic advances that have taken place in computer technology in the past twenty years. Within the last decade, two new radiation sources have been added to the crystallographer's tools. One is synchrotron X-rays and the other is spallation neutrons. Both have much more powerful fluxes than the pre­ vious sources and they are pulsed rather than continuos. New techniques are necessary to fully exploit the intense continuos radiation spectrum and its pulsed property. Both radiations are only available from particular National Laboratories on a guest-user basis for scientists outside these Na­ tional Laboratories. Hi...

  19. Steady-state thermal-hydraulic design analysis of the Advanced Neutron Source reactor

    International Nuclear Information System (INIS)

    The Advanced Neutron Source (ANS) is a research reactor that is planned for construction at Oak Ridge National Laboratory. This reactor will be a user facility with the major objective of providing the highest continuous neutron beam intensities of any reactor in the world. Additional objectives for the facility include providing materials irradiation facilities and isotope production facilities as good as, or better than, those in the High Flux Isotope Reactor. To achieve these objectives, the reactor design uses highly subcooled heavy water as both coolant and moderator. Two separate core halves of 67.6-L total volume operate at an average power density of 4.5 MW(t)/L, and the coolant flows upward through the core at 25 m/s. Operating pressure is 3.1 MPa at the core inlet with a 1.4-MPa pressure drop through the core region. Finally, in order to make the resources available for experimentation, the fuel is designed to provide a 17-d fuel cycle with an additional 4 d planned in each cycle for the refueling process. This report examines the codes and models used to develop the thermal-hydraulic design for ANS, as well as the correlations and physical data; evaluates thermal-hydraulic uncertainties; reports on thermal-hydraulic design and safety analysis; describes experimentation in support of the ANS reactor design and safety analysis; and provides an overview of the experimental plan

  20. Review and Assessment of Neutron Cross Section and Nubar Covariances for Advanced Reactor Systems

    Energy Technology Data Exchange (ETDEWEB)

    Maslov,V.M.; Oblozinsky, P.; Herman, M.

    2008-12-01

    In January 2007, the National Nuclear Data Center (NNDC) produced a set of preliminary neutron covariance data for the international project 'Nuclear Data Needs for Advanced Reactor Systems'. The project was sponsored by the OECD Nuclear Energy Agency (NEA), Paris, under the Subgroup 26 of the International Working Party on Evaluation Cooperation (WPEC). These preliminary covariances are described in two recent BNL reports. The NNDC used a simplified version of the method developed by BNL and LANL that combines the recent Atlas of Neutron Resonances, the nuclear reaction model code EMPIRE and the Bayesian code KALMAN with the experimental data used as guidance. There are numerous issues involved in these estimates of covariances and it was decided to perform an independent review and assessment of these results so that better covariances can be produced for the revised version in future. Reviewed and assessed are uncertainties for fission, capture, elastic scattering, inelastic scattering and (n,2n) cross sections as well as prompt nubars for 15 minor actinides ({sup 233,234,236}U, {sup 237}Np, {sup 238,240,241,242}Pu, {sup 241,242m,243}Am and {sup 242,243,244,245}Cm) and 4 major actinides ({sup 232}Th, {sup 235,238}U and {sup 239}Pu). We examined available evaluations, performed comparison with experimental data, taken into account uncertainties in model parameterization and made use state-of-the-art nuclear reaction theory to produce the uncertainty assessment.

  1. The use of albedo neutron dosemeters for the measurement of low doses in mixed photon neutron radiation fields at transport casks for high active waste

    International Nuclear Information System (INIS)

    Radiation exposure of the police forces accompanying transports of spent fuel elements and high-active waste form reprocessing (HAW) is determined by means of albedo dosemeters. The official dosimetry services use this type of dosemeter to mesure the personal dose in mixed gamma/neutron radiation fields above all for routine monitoring of workers occupationally exposed to radiation. The present report describes the detailed set-up and functioning of the albedo dosemeter, the process of obtaining the photon and neutron personal dose from the detector indications as well as the determination of the detection limit of the total personal dose of the albedo dosemeter according to the methods specified in the valid standards. Determination of the detection limit is based on the experience gained during previous transports, on measurements performed at transport casks, on results of type tests at PTB (Federal Physical and Technical Authority), on the measurement uncertainties obtained from the annual intercomparison measurements of the PTB as well as on the test irradiation specially performed in the range of small neutron and photon doses under laboratory conditions. For the dosimetry systems of the dosimetry services and the specific transport conditions, a reference level of 100 μSv was specified with regard to the dose detection limit. (orig.)

  2. Practical application of passive safety features for the advanced neutron source cooling system

    International Nuclear Information System (INIS)

    The results of a conceptual design study leading to the definition of a reference design for the Advanced Neutron Source (ANS) heavy water cooling system are presented. The objective of this study was to define a cooling system that not only met the ANS goals for operating parameters, reliability, availability, and maintainability, but also used inherent, passive, and diverse features and characteristics to satisfy the ANS internal events core melt goal of -5/yr. The approach taken in this study was to define a cooling system configuration having the minimum basic components and characteristics to satisfy the requirements for normal operation, and then to add only those features necessary to meet the requirements for all emergency design-basis events

  3. Report of the advanced neutron source (ANS) aluminum cladding corrosion workshop

    International Nuclear Information System (INIS)

    The Advanced Neutron Source (ANS) Corrosion Workshop on aluminum cladding corrosion in reactor environments is summarized. The Workshop was held to examine the aluminum cladding oxidation studies being conducted in support of the ANS design. This report was written principally to provide a record of the ideas and judgments expressed by the workshop attendees. The ANS operating heat flux is significantly higher than that in existing reactors, and early experiments indicate that there may be an aluminum cladding oxidation problem unique to higher heat fluxes or associated cladding temperatures that, if not solved, may limit the operation of the ANS to unacceptably low power levels. A brief description of the information presented by each speaker is included along with a compilation of the most significant ideas and recommended research areas. The appendixes contain a copy of the workshop agenda and a list of attendees

  4. TRANSX-2.15, Neutron Gamma Particle Transport Tables from MATXS Format Cross-Sections

    International Nuclear Information System (INIS)

    1 - Description of program or function: TRANSX is a computer code that reads nuclear data from a library in MATXS format and produces transport tables compatible with many discrete-ordinates (SN) and diffusion codes. Tables can be produced for neutron, photon, charged-particle, or coupled transport. Options include adjoint tables, mixtures, homogeneous or heterogeneous self-shielding, group collapse, homogenization, thermal up-scatter, prompt or steady-state fission, transport corrections, elastic removal corrections, and flexible response function edits. 2 - Method of solution: TRANSX reads through the materials in a MATXS library and accumulates the cross sections into a transport table using the user's mix instructions. At the same time, response function edit cross sections are accumulated using the user's edit instructions. They can thus be any linear combination of the cross sections available in the library. When the table is complete, it is written out in the desired format. Output options include DTF-style card images, FIDO, ISOTXS, and the binary group-ordered GOXS format. Self-shielding is handled using the background cross section method. Heterogeneity options include homogeneous mixtures, escape using mean chord, lattices of cylinders by the Bell or Sauer approximations, and reflected or periodic slab cell by the bell or E3 approximations. 3 - Restrictions on the complexity of the problem: Only narrow- resonance self-shielding is available in this version. This may affect accuracy for thermal problems

  5. NATO Advanced Study Institute on Chemical Transport in Melasomatic Processes

    CERN Document Server

    1987-01-01

    As indicated on the title page, this book is an outgrowth of the NATO Advanced Study Institute (ASI) on Chemical Transport in Metasomatic Processes, which was held in Greece, June 3-16, 1985. The ASI consisted of five days of invited lectures, poster sessions, and discussion at the Club Poseidon near Loutraki, Corinthia, followed by a two-day field trip in Corinthia and Attica. The second week of the ASI consisted of an excursion aboard M/S Zeus, M/Y Dimitrios II, and the M/S Irini to four of the Cycladic Islands to visit, study, and sample outstanding exposures of metasomatic activity on Syros, Siphnos, Seriphos, and Naxos. Nine­ teen invited lectures and 10 session chairmen/discussion leaders participated in the ASI, which was attended by a total of 92 professional scientists and graduate stu­ dents from 15 countries. Seventeen of the invited lectures and the Field Excursion Guide are included in this volume, together with 10 papers and six abstracts representing contributed poster sessions. Although more...

  6. Advanced transportation system studies. Alternate propulsion subsystem concepts: Propulsion database

    Science.gov (United States)

    Levack, Daniel

    1993-01-01

    The Advanced Transportation System Studies alternate propulsion subsystem concepts propulsion database interim report is presented. The objective of the database development task is to produce a propulsion database which is easy to use and modify while also being comprehensive in the level of detail available. The database is to be available on the Macintosh computer system. The task is to extend across all three years of the contract. Consequently, a significant fraction of the effort in this first year of the task was devoted to the development of the database structure to ensure a robust base for the following years' efforts. Nonetheless, significant point design propulsion system descriptions and parametric models were also produced. Each of the two propulsion databases, parametric propulsion database and propulsion system database, are described. The descriptions include a user's guide to each code, write-ups for models used, and sample output. The parametric database has models for LOX/H2 and LOX/RP liquid engines, solid rocket boosters using three different propellants, a hybrid rocket booster, and a NERVA derived nuclear thermal rocket engine.

  7. Comparison of preconditioned generalized conjugate gradient methods to two-dimensional neutron and photon transport equation

    International Nuclear Information System (INIS)

    We apply and compare the preconditioned generalized conjugate gradient methods to solve the linear system equation that arises in the two-dimensional neutron and photon transport equation in this paper. Several subroutines are developed on the basis of preconditioned generalized conjugate gradient methods for time-independent, two-dimensional neutron and photon transport equation in the transport theory. These generalized conjugate gradient methods are used: TFQMR (transpose free quasi-minimal residual algorithm) CGS (conjugate gradient square algorithm), Bi-CGSTAB (bi-conjugate gradient stabilized algorithm) and QMRCGSTAB (quasi-minimal residual variant of bi-conjugate gradient stabilized algorithm). These subroutines are connected to computer program DORT. Several problems are tested on a personal computer with Intel Pentium CPU. The reasons to choose the generalized conjugate gradient methods are that the methods have better residual (equivalent to error) control procedures in the computation and have better convergent rate. The pointwise incomplete LU factorization ILU, modified pointwise incomplete LU factorization MILU, block incomplete factorization BILU and modified blockwise incomplete LU factorization MBILU are the preconditioning techniques used in the several testing problems. In Bi-CGSTAB, CGS, TFQMR and QMRCGSTAB method, we find that either CGS or Bi-CGSTAB method combined with preconditioner MBILU is the most efficient algorithm in these methods in the several testing problems. The numerical solution of flux by preconditioned CGS and Bi-CGSTAB methods has the same result as those from Cray computer, obtained by either the point successive relaxation method or the line successive relaxation method combined with Gaussian elimination

  8. MONTE CARLO NEUTRINO TRANSPORT THROUGH REMNANT DISKS FROM NEUTRON STAR MERGERS

    Energy Technology Data Exchange (ETDEWEB)

    Richers, Sherwood; Ott, Christian D. [TAPIR, Mailcode 350-17, Walter Burke Institute for Theoretical Physics, California Institute of Technology, Pasadena, CA 91125 (United States); Kasen, Daniel; Fernández, Rodrigo [Department of Astronomy and Theoretical Astrophysics Center, University of California, Berkeley, CA 94720 (United States); O’Connor, Evan [Department of Physics, Campus Code 8202, North Carolina State University, Raleigh, NC 27695 (United States)

    2015-11-01

    We present Sedonu, a new open source, steady-state, special relativistic Monte Carlo (MC) neutrino transport code, available at bitbucket.org/srichers/sedonu. The code calculates the energy- and angle-dependent neutrino distribution function on fluid backgrounds of any number of spatial dimensions, calculates the rates of change of fluid internal energy and electron fraction, and solves for the equilibrium fluid temperature and electron fraction. We apply this method to snapshots from two-dimensional simulations of accretion disks left behind by binary neutron star mergers, varying the input physics and comparing to the results obtained with a leakage scheme for the cases of a central black hole and a central hypermassive neutron star. Neutrinos are guided away from the densest regions of the disk and escape preferentially around 45° from the equatorial plane. Neutrino heating is strengthened by MC transport a few scale heights above the disk midplane near the innermost stable circular orbit, potentially leading to a stronger neutrino-driven wind. Neutrino cooling in the dense midplane of the disk is stronger when using MC transport, leading to a globally higher cooling rate by a factor of a few and a larger leptonization rate by an order of magnitude. We calculate neutrino pair annihilation rates and estimate that an energy of 2.8 × 10{sup 46} erg is deposited within 45° of the symmetry axis over 300 ms when a central BH is present. Similarly, 1.9 × 10{sup 48} erg is deposited over 3 s when an HMNS sits at the center, but neither estimate is likely to be sufficient to drive a gamma-ray burst jet.

  9. Parameter estimation for binary neutron-star coalescences with realistic noise during the Advanced LIGO era

    CERN Document Server

    Berry, Christopher P L; Middleton, Hannah; Singer, Leo P; Urban, Alex L; Vecchio, Alberto; Vitale, Salvatore; Cannon, Kipp; Farr, Ben; Farr, Will M; Graff, Philip B; Hanna, Chad; Haster, Carl-Johan; Mohapatra, Satya; Pankow, Chris; Price, Larry R; Sidery, Trevor; Veitch, John

    2014-01-01

    Advanced ground-based gravitational-wave (GW) detectors begin operation imminently. Their intended goal is not only to make the first direct detection of GWs, but also to make inferences about the source systems. Binary neutron-star mergers are among the most promising sources. We investigate the performance of the parameter-estimation pipeline that will be used during the first observing run of the Advanced Laser Interferometer Gravitational-wave Observatory (aLIGO) in 2015: we concentrate on the ability to reconstruct the source location on the sky, but also consider the ability to measure masses and the distance. Accurate, rapid sky-localization is necessary to alert electromagnetic (EM) observatories so that they can perform follow-up searches for counterpart transient events. We consider parameter-estimation accuracy in the presence of realistic, non-Gaussian noise. We find that the character of the noise makes negligible difference to the parameter-estimation performance. The source luminosity distance ...

  10. 6.3 MeV fast neutrons in the treatment of patients with locally advanced and locally recurrent breast cancer

    Science.gov (United States)

    Velikaya, V. V.; Musabaeva, L. I.; Lisin, V. A.; Startseva, Zh. A.

    2016-08-01

    The study included 135 breast cancer patients (70 patients with locally recurrent breast cancer and 65 patients with locally advanced breast cancer with unfavorable prognostic factors) who received the neutron therapy alone or in combination with the photon therapy. The neutron therapy was shown to be effective in multimodality treatment of patients with locally advanced and locally recurrent breast cancer. The 8-year survival rate in patients without repeated breast cancer recurrence was 87.6 ± 8.7% after the neutron and neutron-photon therapy and 54.3 ± 9.2% after the electron beam therapy.

  11. GPU-based high performance Monte Carlo simulation in neutron transport

    Energy Technology Data Exchange (ETDEWEB)

    Heimlich, Adino; Mol, Antonio C.A.; Pereira, Claudio M.N.A. [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil). Lab. de Inteligencia Artificial Aplicada], e-mail: cmnap@ien.gov.br

    2009-07-01

    Graphics Processing Units (GPU) are high performance co-processors intended, originally, to improve the use and quality of computer graphics applications. Since researchers and practitioners realized the potential of using GPU for general purpose, their application has been extended to other fields out of computer graphics scope. The main objective of this work is to evaluate the impact of using GPU in neutron transport simulation by Monte Carlo method. To accomplish that, GPU- and CPU-based (single and multicore) approaches were developed and applied to a simple, but time-consuming problem. Comparisons demonstrated that the GPU-based approach is about 15 times faster than a parallel 8-core CPU-based approach also developed in this work. (author)

  12. NUMERICAL SIMULATION OF SINGLE-GROUP, STEADY STATE AND ISOTROPIC NEUTRON TRANSPORT EQUATION IN DIFFUSIVE REGIMES

    Institute of Scientific and Technical Information of China (English)

    2006-01-01

    The single-group,steadystate,isotropic for mofthe neutron transport equationis given by[1]Ω·+σtI-σsPψ(x,Ω)=q(x,Ω)(x,Ω)∈D×Sψ(x,Ω)=g(x,Ω)x∈Din={x∈D,γ(x)·Ω<0(1)whereσtis the total cross section,σSis the scatteringcross section,andψ(x,Ω)is the angular flux to bedeter mined for all pointsx∈D,D Rn(n=2,3)and all possible travel directionsΩ,ΩS(Sis a u-nit disk or a unit sphere),γ(x)denotes the out wardunit nor mal atx∈D,Idenotes the identity opera-tor,the operatorPis defined by[Pψ](x)=∫Sψ(x,Ω)dΩ(2)Whenσt→∞,andσσ...

  13. Annealing study of electron transport in slightly neutron-irradiated graphite

    International Nuclear Information System (INIS)

    A study has been made on the recovery of transport properties of synthetic single-crystal graphite irradiated with fast neutrons to a total dose of 3 x 1036 nvt. Isochronal annealing was conducted stepwise between 100 0C and 1600 0C. An analysis based on the rigid two-band model gives a consistent account of the diffusion thermoelectricity and the galvanomagnetic effect. In addition to the major normal recovery between 200 0C and 300 0C, a reverse recovery in the carrier constitution is found to take place in the annealing range from 300 0C to 350 0C. This seems explicable by considering that divacancies are intermediately transformed to single vacancies during their annihilation process. The phonon-drag thermoelectric power vs. acceptor concentration relationship implies that positive holes are more strongly scattered than electrons by radiation-induced vacancies negatively charged by accepting electrons within the basal plane. (author)

  14. Milne problem in two adjacent half-spaces in neutron transport theory in two energy groups

    International Nuclear Information System (INIS)

    Case's method, combined with invariance principle, is used to obtain exact solutions of neutron transport problems in two adjacent half-spaces, in the two-group isotropic scattering model. The continuity condition and the invariance principle are used to obtain a set of coupled regular integral equations for the angular distribution at the interface. The expansion coefficients can be obtained from the solutions of these integral equations using the orthogonality properties of the eigen functions. Numerical results are presented for the Milne and the Constant Source problems for pure and borated light water media. The results show the feasibility of the proposed method to provide exact numerical results which can be used as standards of comparison for various approximate methods

  15. The solution of the multigroup neutron transport equation using spherical harmonics

    International Nuclear Information System (INIS)

    A solution of the multi-group neutron transport equation in up to three space dimensions is presented. The flux is expanded in a series of unnormalised spherical harmonics. Using the various recurrence formulae a linked set of first order differential equations is obtained for the moments psisup(g)sub(lm)(r), γsup(g)sub(lm)(r). Terms with odd l are eliminated resulting in a second order system which is solved by two methods. The first is a finite difference formulation using an iterative procedure, secondly, in XYZ and XY geometry a finite element solution is given. Results for a test problem using both methods are exhibited and compared. (orig./RW)

  16. Design of the low energy beam transport line for the China spallation neutron source

    Institute of Scientific and Technical Information of China (English)

    LI Jin-Hai; OUYANG Hua-Fu; FU Shi-Nian; ZHANG Sua-Shun; HE Wei

    2008-01-01

    The design of the China Spallation Neutron Source (CSNS) low-energy beam transport (LEBT) line, which locates between the ion source and the radio-frequency quadrupole (RFQ), has been completed with the TRACE3D code. The design aims at perfect matching, primary chopping, a small emittance growth and sufficient space for beam diagnostics. The line consists of three solenoids, three vacuum chambers, two steering magnets and a pre-chopper. The total length of LEBT is about 1.74 m. This LEBT is designed to transfer 20 mA of H-pulsed beam from the ion source to the RFQ. An induction cavity is adopted as the pre-chopper.The electrostatic octupole steerer is discussed as a candidate. A four-quadrant aperture for beam scraping and beam position monitoring is designed.

  17. First studies of electron transport along small gas gaps of novel foil radiation converters for fast-neutron detectors

    OpenAIRE

    Cortesi, M.; Zboray, R.; Adams, R.; Dangendorf, V.; Breskin, A.; Mayer, S.; Hoedlmoser, H.; Prasser, H.-M.

    2012-01-01

    Novel high efficiency fast-neutron detectors were suggested for fan-beam tomography applications. They combine multi-layer polymer converters in gas medium, coupled to thick gaseous electron multipliers (THGEM). In this work we discuss the results of a systematic study of the electron transport inside a narrow gap between successive converter foils, which affects the performance of the detector, both in terms of detection efficiency and localization properties. The efficiency of transporting ...

  18. Experimental Software Design of the Neutron Texture Diffractometer at China Advanced Research Reactor

    Institute of Scientific and Technical Information of China (English)

    2011-01-01

    Neutron scattering lab is building our country's first neutron texture diffractometer, which will be used for the texture measurement and analysis in the materials science and engineering applications. The sample table and its measurement and control

  19. High Neutron Fluence Survivability Testing of Advanced Fiber Bragg Grating Sensors

    International Nuclear Information System (INIS)

    The motivation for the reported research was to support NASA space nuclear power initiatives through the development of advanced fiber optic sensors for space-based nuclear power applications. The purpose of the high-neutron fluence testing was to demonstrate the survivability of fiber Bragg grating (FBG) sensors in a fission reactor environment. 520 FBGs were installed in the Ford reactor at the University of Michigan. The reactor was operated for 1012 effective full power hours resulting in a maximum neutron fluence of approximately 5x1019 n/cm2, and a maximum gamma dose of 2x103 MGy gamma. This work is significant in that, to the knowledge of the authors, the exposure levels obtained are approximately 1000 times higher than for any previously published experiment. Four different fiber compositions were evaluated. An 87% survival rate was observed for fiber Bragg gratings located at the fuel centerline. Optical Frequency Domain Reflectometry (OFDR), originally developed at the NASA Langley Research Center, can be used to interrogate several thousand low-reflectivity FBG strain and/or temperature sensors along a single optical fiber. A key advantage of the OFDR sensor technology for space nuclear power is the extremely low mass of the sensor, which consists of only a silica fiber 125μm in diameter. The sensors produced using this technology will fill applications in nuclear power for current reactor plants, emerging Generation-IV reactors, and for space nuclear power. The reported research was conducted by Luna Innovations and was funded through a Small Business Innovative Research (SBIR) contract with the NASA Glenn Research Center

  20. 3-D Deep Penetration Neutron Imaging of Thick Absorgin and Diffusive Objects Using Transport Theory

    Energy Technology Data Exchange (ETDEWEB)

    Ragusa, Jean; Bangerth, Wolfgang

    2011-08-01

    here explores the inverse problem of optical tomography applied to heterogeneous domains. The neutral particle transport equation was used as the forward model for how neutral particles stream through and interact within these heterogeneous domains. A constrained optimization technique that uses Newtons method served as the basis of the inverse problem. Optical tomography aims at reconstructing the material properties using (a) illuminating sources and (b) detector readings. However, accurate simulations for radiation transport require that the particle (gamma and/or neutron) energy be appropriate discretize in the multigroup approximation. This, in turns, yields optical tomography problems where the number of unknowns grows (1) about quadratically with respect to the number of energy groups, G, (notably to reconstruct the scattering matrix) and (2) linearly with respect to the number of unknown material regions. As pointed out, a promising approach could rely on algorithms to appropriately select a material type per material zone rather than G2 values. This approach, though promising, still requires further investigation: (a) when switching from cross-section values unknowns to material type indices (discrete integer unknowns), integer programming techniques are needed since derivative information is no longer available; and (b) the issue of selecting the initial material zoning remains. The work reported here proposes an approach to solve the latter item, whereby a material zoning is proposed using one-group or few-groups transport approximations. The capabilities and limitations of the presented method were explored; they are briefly summarized next and later described in fuller details in the Appendices. The major factors that influenced the ability of the optimization method to reconstruct the cross sections of these domains included the locations of the sources used to illuminate the domains, the number of separate experiments used in the reconstruction, the

  1. Neutron radiation therapy: application of advanced technology to the treatment of cancer

    CERN Document Server

    Maughan, R L; Kota, C; Burmeister, J; Porter, A T; Forman, J D; Blosser, H G; Blosser, E; Blosser, G

    1999-01-01

    The design and construction of a unique superconducting cyclotron for use in fast neutron radiation therapy is described. The clinical results obtained in the treatment of adenocarcinoma of the prostate with this accelerator are presented. Future use of the boron neutron capture reaction as a means of enhancing fast neutron therapy in the treatment of patients with brain tumors (glioblastoma multiforme) is also discussed.

  2. Diagnosing Implosion Performance at the National Ignition Facility by Means of Advanced Neutron-Spectrometry and Neutron-Imaging Techniques

    International Nuclear Information System (INIS)

    Full text: Proper assembly of capsule mass, as manifested through the evolution of fuel areal density, is essential for achieving hot-spot ignition planned at the National Ignition Facility (NIF). Experimental information about areal density and areal-density asymmetries, hot-spot ion temperature (Ti) and yield (Yn) are therefore critical for understanding the assembly of the fuel. To obtain this information, a suite of neutron Time-of-Flight (nTOF) spectrometers and a Magnetic Recoil Spectrometer (MRS) has been commissioned and extensively used on the NIF for measurements of the neutron spectrum in the energy range from 1.5 to 20 MeV. This range covers all essential details of the neutron spectrum, allowing for the determination of areal density, Yn, and Ti. The spectrometers are fielded at different locations around the implosion for directional measurements of the neutron spectrum, also allowing for determination of areal-density asymmetries and possible kinetic effects. The data obtained from these diagnostics have been essential to the progress of the National Ignition Campaign (NIC), indicating that the implosion performance, characterized by the Experimental Ignition Threshold Factor (ITFx), has improved about two orders of magnitude since the first cryogenic shot taken in September 2010. Areal-density values greater than 1 g/cm2 are now readily achieved. By combining the areal-density data with information about the spatial extent of the high-density region obtained from Neutron Imaging System (NIS), it has been demonstrated that densities above 500 g/cc and pressure-time (Pτ) products in excess of 10 atm s have been achieved, which are according to HYDRA simulations about a factor of three from ignition conditions. (author)

  3. Development of new multigrid schemes for the method of characteristics in neutron transport theory

    International Nuclear Information System (INIS)

    This dissertation is based upon our doctoral research that dealt with the conception and development of new non-linear multigrid techniques for the Method of the Characteristics (MOC) within the TDT code. Here we focus upon a two-level scheme consisting of a fine level on which the neutron transport equation is iteratively solved using the MOC algorithm, and a coarse level defined by a more coarsely discretized phase space on which a low-order problem is considered. The solution of this problem is then used in order to correct the angular flux moments resulting from the previous transport iteration. A flux-volume homogenization procedure is employed to evaluate the coarse-level material properties after each transport iteration. This entails the non-linearity of the methods. According to the Generalised Equivalence Theory (GET), additional degrees of freedom are introduced for the low-order problem so that the convergence of the acceleration scheme can be ensured. We present two classes of non-linear methods: transport-like methods and discussion-like methods. Transport-like methods consider a homogenized low-order transport problem on the coarse level. This problem is iteratively solved using the same MOC algorithm as for the transport problem on the fine level. Discontinuity factors are then employed, per region or per surface, in order to reconstruct the currents evaluated by the low-order operator, which ensure the convergence of the acceleration scheme. On the other hand, discussion-like methods consider a low-order problem inspired by diffusion. We studied the non-linear Coarse Mesh Finite Difference (CMFD) method, already present in literature, in the perspective of integrating it into TDT code. Then, we developed a new non-linear method on the model of CMFD. From the latter, we borrowed the idea to establish a simple relation between currents and fluxes in order to obtain a problem involving only coarse fluxes. Finally, those non-linear methods have been

  4. Numerical research on the anisotropic transport of thermal neutron in heterogeneous porous media with micron X-ray computed tomography

    Science.gov (United States)

    Wang, Yong; Yue, Wenzheng; Zhang, Mo

    2016-06-01

    The anisotropic transport of thermal neutron in heterogeneous porous media is of great research interests in many fields. In this paper, it is the first time that a new model based on micron X-ray computed tomography (CT) has been proposed to simultaneously consider both the separation of matrix and pore and the distribution of mineral components. We apply the Monte Carlo method to simulate thermal neutrons transporting through the model along different directions, and meanwhile detect those unreacted thermal neutrons by an array detector on the other side of the model. Therefore, the anisotropy of pore structure can be imaged by the amount of received thermal neutrons, due to the difference of rock matrix and pore-filling fluids in the macroscopic reaction cross section (MRCS). The new model has been verified by the consistent between the simulated data and the pore distribution from X-ray CT. The results show that the evaluation of porosity can be affected by the anisotropy of media. Based on the research, a new formula is developed to describe the correlation between the resolution of array detectors and the quality of imaging. The formula can be further used to analyze the critical resolution and the suitable number of thermal neutrons emitted in each simulation. Unconventionally, we find that a higher resolution cannot always lead to a better image.

  5. An overview of the Boltzmann transport equation solution for neutrons, photons and electrons in cartesian geometry

    International Nuclear Information System (INIS)

    Questions regarding accuracy and efficiency of deterministic transport methods are still on our mind today, even with modern supercomputers. The most versatile and widely used deterministic methods are the PN approximation, the SN method (discrete ordinates method) and their variants. In the discrete ordinates (SN) formulations of the transport equation, it is assumed that the linearized Boltzmann equation only holds for a set of distinct numerical values of the direction-of-motion variables. In this work, looking forward to confirm the capabilities of deterministic methods in obtaining accurate results, we present a general overview of deterministic methods to solve the Boltzmann transport equation for neutral and charged particles. First, we describe a review in the Laplace transform technique applied to SN two dimensional transport equation in a rectangular domain considering Compton scattering. Next, we solved the Fokker-Planck (FP) equation, an alternative approach for the Boltzmann transport equation, assuming a monoenergetic electron beam in a rectangular domain. The main idea relies on applying the PN approximation, a recent advance in the class of deterministic methods, in the angular variable, to the two dimensional Fokker-Planck equation and then applying the Laplace Transform in the spatial x-variable. Numerical results are given to illustrate the accuracy of deterministic methods presented. (author)

  6. Boron neutron capture therapy applied to advanced breast cancers: Engineering simulation and feasibility study of the radiation treatment protocol

    Science.gov (United States)

    Sztejnberg Goncalves-Carralves, Manuel Leonardo

    This dissertation describes a novel Boron Neutron Capture Therapy (BNCT) application for the treatment of human epidermal growth factor receptor type 2 positive (HER2+) breast cancers. The original contribution of the dissertation is the development of the engineering simulation and the feasibility study of the radiation treatment protocol for this novel combination of BNCT and HER2+ breast cancer treatment. This new concept of BNCT, representing a radiation binary targeted treatment, consists of the combination of two approaches never used in a synergism before. This combination may offer realistic hope for relapsed and/or metastasized breast cancers. This treatment assumes that the boronated anti-HER2 monoclonal antibodies (MABs) are administrated to the patient and accumulate preferentially in the tumor. Then the tumor is destroyed when is exposed to neutron irradiation. Since the use of anti-HER2 MABs yields good and promising results, the proposed concept is expected to amplify the known effect and be considered as a possible additional treatment approach to the most severe breast cancers for patients with metastasized cancer for which the current protocol is not successful and for patients refusing to have the standard treatment protocol. This dissertation makes an original contribution with an integral numerical approach and proves feasible the combination of the aforementioned therapy and disease. With these goals, the dissertation describes the theoretical analysis of the proposed concept providing an integral engineering simulation study of the treatment protocol. An extensive analysis of the potential limitations, capabilities and optimization factors are well studied using simplified models, models based on real CT patients' images, cellular models, and Monte Carlo (MCNP5/X) transport codes. One of the outcomes of the integral dosimetry assessment originally developed for the proposed treatment of advanced breast cancers is the implementation of BNCT

  7. Structural advances for the major facilitator superfamily (MFS) transporters.

    Science.gov (United States)

    Yan, Nieng

    2013-03-01

    The major facilitator superfamily (MFS) is one of the largest groups of secondary active transporters conserved from bacteria to humans. MFS proteins selectively transport a wide spectrum of substrates across biomembranes and play a pivotal role in multiple physiological processes. Despite intense investigation, only seven MFS proteins from six subfamilies have been structurally elucidated. These structures were captured in distinct states during a transport cycle involving alternating access to binding sites from either side of the membrane. This review discusses recent progress in MFS structure analysis and focuses on the molecular basis for substrate binding, co-transport coupling, and alternating access.

  8. The TORT three-dimensional discrete ordinates neutron/photon transport code (TORT version 3)

    Energy Technology Data Exchange (ETDEWEB)

    Rhoades, W.A.; Simpson, D.B.

    1997-10-01

    TORT calculates the flux or fluence of neutrons and/or photons throughout three-dimensional systems due to particles incident upon the system`s external boundaries, due to fixed internal sources, or due to sources generated by interaction with the system materials. The transport process is represented by the Boltzman transport equation. The method of discrete ordinates is used to treat the directional variable, and a multigroup formulation treats the energy dependence. Anisotropic scattering is treated using a Legendre expansion. Various methods are used to treat spatial dependence, including nodal and characteristic procedures that have been especially adapted to resist numerical distortion. A method of body overlay assists in material zone specification, or the specification can be generated by an external code supplied by the user. Several special features are designed to concentrate machine resources where they are most needed. The directional quadrature and Legendre expansion can vary with energy group. A discontinuous mesh capability has been shown to reduce the size of large problems by a factor of roughly three in some cases. The emphasis in this code is a robust, adaptable application of time-tested methods, together with a few well-tested extensions.

  9. MCNP: a general Monte Carlo code for neutron and photon transport

    International Nuclear Information System (INIS)

    MCNP is a very general Monte Carlo neutron photon transport code system with approximately 250 person years of Group X-6 code development invested. It is extremely portable, user-oriented, and a true production code as it is used about 60 Cray hours per month by about 150 Los Alamos users. It has as its data base the best cross-section evaluations available. MCNP contains state-of-the-art traditional and adaptive Monte Carlo techniques to be applied to the solution of an ever-increasing number of problems. Excellent user-oriented documentation is available for all facets of the MCNP code system. Many useful and important variants of MCNP exist for special applications. The Radiation Shielding Information Center (RSIC) in Oak Ridge, Tennessee is the contact point for worldwide MCNP code and documentation distribution. A much improved MCNP Version 3A will be available in the fall of 1985, along with new and improved documentation. Future directions in MCNP development will change the meaning of MCNP to Monte Carlo N Particle where N particle varieties will be transported

  10. Lattice design of medium energy beam transport line for n spallation neutron source

    International Nuclear Information System (INIS)

    A 1 GeV H- injector linac is being designed at RRCAT for the proposed Indian Spallation Neutron Source (ISNS). The front-end of the injector linac will consist of Radiofrequency Quadrupole (RFQ) linac, which will accelerate the H- beam from 50 keV to 3 MeV. The beam will be further accelerated in superconducting Single Spoke Resonators (SSRs). A Medium Energy Beam Transport (MEBT) line will be used to transport the beam from the exit of RFQ to the input of SSR. The main purpose of MEBT is to carry out beam matching from RFQ to SSR, and beam chopping. In this paper, we describe the optimization criteria for the lattice design of MEBT. The optimized lattice element parameters are presented for zero and full (15 mA) current case. Beam dynamics studies have been carried out using an envelope tracing code Trace-3D. Required beam deflection angle due to the chopper housed inside the MEBT for beam chopping has also been estimated. (author)

  11. Preparation of a one-curie 171Tm target for the Detector for Advanced Neutron Capture Experiments (DANCE)

    Energy Technology Data Exchange (ETDEWEB)

    Schwantes, Jon M.; Taylor, Wayne A.; Rundberg, Robert S.; Vieira, David J.

    2008-05-15

    Roughly one curie of 171Tm (t1/2=1.92a) has been produced and purified for the purpose of making a nuclear target for the first measurements of its neutron capture cross section. Target preparation consisted of three key steps: (1) material production; (2) separation and purification; and (3) electrodeposition onto a suitable backing material. Approximately 1.5 mg of the target material (at the time of separation) was produced by irradiating roughly 250 mg of its stable enriched 170Er lanthanide neighbor with neutrons at the ILL reactor in France. This production method resulted in a “difficult-to-separate” 1:167 mixture of near-neighboring lanthanides, Tm and Er. Separation and purification was accomplished using high-performance liquid chromatorgraphy (HPLC), with a proprietary cation exchange column (Dionex, CS-3) and alpha-hydroxyisobutyric acid (a-HIB) eluent. This technique yielded a final product of ~95% purity with respect to Tm. A portion (20 ug) of the Tm was electrodeposited on thin Be foil and delivered to the Los Alamos Neutron Science CEnter (LANSCE) for preliminary analysis of its neutron capture cross section using the Detector for Advanced Neutron Capture Experiments (DANCE). This paper discusses the major hurdles associated with the separation and purification step including, scale-up issues related to the use of HPLC for material separation and purification of the target material from a-HIB and 4-(2-pyridylazo)resorcinol (PAR) colorant.

  12. Preparation of a one-curie 171Tm target for the detector for advanced neutron capture experiments (DANCE)

    International Nuclear Information System (INIS)

    Approximately one curie of 171Tm (T1/2 = 1.92a) has been produced and purified for the purpose of making a nuclear target for the first measurements of its neutron capture cross section. Target preparation consisted of three key steps: (1) material production; (2) separation and purification; and (3) electrodeposition onto a suitable backing material. Approximately 1.5 mg of the target material (at the time of separation) was produced by irradiating ca. 250 mg of its stable enriched 170Er lanthanide neighbour with neutrons at the ILL reactor in France. This production method resulted in a 'difficult-to-separate' 1:167 mixture of near-neighboring lanthanides, Tm and Er. Separation and purification was accomplished using high-performance liquid chromatography (HPLC), with a proprietary cation-exchange column (Dionex, CS-3) and alphahydroxyisobutyric acid (α-HIB) eluent. This technique yielded a final product of ∼95% purity with respect to Tm. A portion (20 μg) of the Tm was electrodeposited onto thin Be foil and delivered to the Los Alamos Neutron Science Center (LANSCE) for preliminary analysis of its neutron capture cross section using the Detector for Advanced Neutron Capture Experiments (DANCE). This paper discusses the major hurdles associated with the separation and purification step, including scale-up issues related to the use of HPLC for material separation and purification of the target material from α-HIB and 4-(2-pyridylazo)resorcinol (PAR) colorant. (author)

  13. Boron neutron capture therapy for advanced salivary gland carcinoma in head and neck

    International Nuclear Information System (INIS)

    Boron neutron capture therapy (BNCT) is a among the radiation treatments known to have a selective lethal effect on tumor cells. This study summarizes the tumor responses and the acute and late adverse effects of BNCT in the treatment of patients with both recurrent and newly diagnosed T4 salivary gland carcinoma. Two patients with recurrent cancer and 3 with newly diagnosed T4 advanced malignancy were registered between October 2003 and September 2007, with the approval of the medical ethics committees of Kawasaki Medical School and Kyoto University. BNCT was performed, in a single fraction using an epithermal beam, at Japan Research Reactor 4. All patients achieved a complete response within 6 months of treatment. The median duration of the complete response was 24.0 months; the median overall survival time was 32.0 months. Three of the 5 patients are still alive; the other 2 died of distant metastatic disease. Open biopsy of the parotid gland after BNCT was performed in 1 patient and revealed no residual viable cancer cells and no serious damage to the normal glandular system. Although mild alopecia, xerostomia, and fatigue occurred in all patients, there were no severe adverse effects of grade 3 or greater. Our preliminary results demonstrate that BNCT is a potential curative therapy for patients with salivary gland carcinoma. The treatment does not cause any serious adverse effects, and may be used regardless of whether the primary tumor has been previously treated. (author)

  14. Oxidation behavior of plasma sintered beryllium-titanium intermetallic compounds as an advanced neutron multiplier

    Science.gov (United States)

    Kim, Jae-Hwan; Nakamichi, Masaru

    2013-07-01

    Beryllium intermetallic compounds (beryllides) such as Be12Ti are very promising candidates for advanced neutron multiplier materials in a demonstration fusion power reactor (DEMO). However, beryllides are too brittle to be fabricated either into pebble-type or rod-type shapes via conventional methods (i.e. arc melting and hot isostatic pressing). We have proposed a plasma sintering technique as a new method for beryllide fabrication, and our studies on the properties of plasma sintered beryllides are ongoing. In the present work, the oxidation properties of plasma sintered beryllides were investigated at 1273 K for 24 h in a dry air atmosphere to evaluate the high temperature properties of this material. Thermal gravimetry measurements indicate that specimens with larger fractions of Be12Ti phase corresponding to samples that have been sintered for longer time periods, exhibit superior oxidation properties. Our evaluation of the oxidation behavior of each phase in our beryllide samples is as follows: Be12Ti and Be17Ti2 both have good oxidation resistance, owing to the formation of dense and protective scales, while the Be and Be2Ti phases are mainly responsible for thermal-gravimetry (TG) weight gains, which is indicative of severe oxidation. We attribute the degradation in oxidation resistance specifically to Be2Ti that transforms into TiO2, and also find this phase to be the cause of deterioration in the mechanical properties of samples, owing to cracks near Be2Ti phase conglomerates.

  15. Preliminary fracture analysis of the core pressure boundary tube for the Advanced Neutron Source Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Schulz, K.C. [Univ. of Turabo, Gurabo, Puerto (Puerto Rico). College of Engineering; Yahr, G.T. [Oak Ridge National Lab., TN (United States)

    1995-08-01

    The outer core pressure boundary tube (CPBT) of the Advanced neutron Source (ANS) reactor being designed at Oak Ridge National Laboratory is currently specified as being composed of 6061-T6 aluminum. ASME Boiler and Pressure Vessel Code fracture analysis rules for nuclear components are based on the use of ferritic steels; the expressions, tables, charts and equations were all developed from tests and analyses conducted for ferritic steels. Because of the nature of the Code, design with thin aluminum requires analytical approaches that do not directly follow the Code. The intent of this report is to present a methodology comparable to the ASME Code for ensuring the prevention of nonductile fracture of the CPBT in the ANS reactor. 6061-T6 aluminum is known to be a relatively brittle material; the linear elastic fracture mechanics (LEFM) approach is utilized to determine allowable flaw sizes for the CPBT. A J-analysis following the procedure developed by the Electric Power Research Institute was conducted as a check; the results matched those for the LEFM analysis for the cases analyzed. Since 6061-T6 is known to embrittle when irradiated, the reduction in K{sub Q} due to irradiation is considered in the analysis. In anticipation of probable requirements regarding maximum allowable flaw size, a survey of nondestructive inspection capabilities is also presented. A discussion of probabilistic fracture mechanics approaches, principally Monte Carlo techniques, is included in this report as an introduction to what quantifying the probability of nonductile failure of the CPBT may entail.

  16. Boron neutron capture therapy outcomes for advanced or recurrent head and neck cancer

    International Nuclear Information System (INIS)

    We retrospectively review outcomes of applying boron neutron capture therapy (BNCT) to unresectable advanced or recurrent head and neck cancers. Patients who were treated with BNCT for either local recurrent or newly diagnosed unresectable head or neck cancers between December 2001 and September 2007 were included. Clinicopathological characteristics and clinical outcomes were retrieved from hospital records. Either a combination of borocaptate sodium and boronophenylalanine (BPA) or BPA alone were used as boron compounds. In all the treatment cases, the dose constraint was set to deliver a dose <10–12 Gy-eq to the skin or oral mucosa. There was a patient cohort of 62, with a median follow-up of 18.7 months (range, 0.7–40.8). A total of 87 BNCT procedures were performed. The overall response rate was 58% within 6 months after BNCT. The median survival time was 10.1 months from the time of BNCT. The 1- and 2-year overall survival (OS) rates were 43.1% and 24.2%, respectively. The major acute Grade 3 or 4 toxicities were hyperamylasemia (38.6%), fatigue (6.5%), mucositis/stomatitis (9.7%) and pain (9.7%), all of which were manageable. Three patients died of treatment-related toxicity. Three patients experienced carotid artery hemorrhage, two of whom had coexistent infection of the carotid artery. This study confirmed the feasibility of our dose-estimation method and that controlled trials are warranted. (author)

  17. Induction heating of a spherical aluminum moderator vessel for the Advanced Neutron Source (ANS)

    International Nuclear Information System (INIS)

    This task was to identify and design a heating system to apply 15 kW of heat to a cold source vessel to simulate the Advanced Neutron Source reactor. This research project aims at the analysis of the induction heating of a spherical aluminum moderator vessel. Computer modeling is presented for the design and analysis of the induction heating system. The objective is to apply 15 kW of heat as uniformly as possible to the outer wall of a 410 mm diameter sphere of thickness 1.5 mm. The report also aims at the analysis of a system model which is simulated using the Eddycuff electromagnetic software. The computer model is built with the finite element analysis software Patran. The induction heating system analysis shows that the predicted performance is in close agreement with the computer simulated data. Hardware constraints such as power supplies and matching load are also analyzed in terms of performance and cost. Physical modeling is also suggested, in which the coil and the workpiece are scaled down

  18. Report of the Advanced Neutron Source (ANS) safety workshop, Knoxville, Tennessee, October 25--26, 1988

    International Nuclear Information System (INIS)

    On October 25--26, 1988, about 60 people took part in an Advanced Neutron Source (ANS) Safety Workshop, organized in cooperation with the Oak Ridge Operations (ORO) Office of the Department of Energy (DOE) and held in Knoxville, Tennessee. After a plenary session at which ANS Project staff presented status reports on the ANS design, research and development (R and D), and safety analysis efforts, the workshop broke into three working groups, each covering a different topic: Environmental and Waste Management, Applicable Regulatory Safety Criteria and Goals, and Reactor Concepts. Each group was asked to review the Project's approach to safety-related issues and to provide guidance on future reactor safety needs or directions for the Project. With the help of able chairmen, assisted by reporters and secretarial support, the working groups were extremely successful. Draft reports from each group were prepared before the workshop closed, and the major findings of each group were presented for review and discussion by the entire workshop attendance. This report contains the final version of the group reports, incorporating the results of the overall review by all the workshop participants

  19. Detecting gravitational waves from mountains on neutron stars in the Advanced Detector Era

    CERN Document Server

    Haskell, Brynmor; Patruno, Alessandro; Oppenoorth, Manuel; Melatos, Andrew; Lasky, Paul

    2015-01-01

    Rapidly rotating Neutron Stars (NSs) in Low Mass X-ray Binaries (LMXBs) are thought to be interesting sources of Gravitational Waves (GWs) for current and next generation ground based detectors, such as Advanced LIGO and the Einstein Telescope. The main reason is that many of the NS in these systems appear to be spinning well below their Keplerian breakup frequency, and it has been suggested that torques associated with GW emission may be setting the observed spin period. This assumption has been used extensively in the literature to assess the strength of the likely gravitational wave signal. There is now, however, a significant amount of theoretical and observation work that suggests that this may not be the case, and that GW emission is unlikely to be setting the spin equilibrium period in many systems. In this paper we take a different starting point and predict the GW signal strength for two physical mechanisms that are likely to be at work in LMXBs: crustal mountains due to thermal asymmetries and magne...

  20. Neutron transport measurements and source monitor development for uranium borehole logging sonde. Final report

    International Nuclear Information System (INIS)

    The first part discusses the design and results of a 252Cf-neutron and a 14-MeV neutron benchmark experiment for verifying previously developed theoretical methods for use in the design of a neutron logging tool for uranium borehole exploration. The second part discusses the Science Applications, Inc. (SAI) development of a fast fission monitor for measuring 14-MeV neutrons for a D-T borehole sonde in a high neutron moderating and absorbing environment. The 14-MeV pulsed neutron monitor was used in carrying out the 14-MeV neutron benchmark experiment, and a variation on the method of quantitatively counting many events in one burst of pileup counts, developed for the 14-MeV-pulsed neutron monitor, was employed in successfully counting the epithermal neutrons produced by a short (2 μs) burst of 14-MeV neutrons. Thus, the development of the 14-MeV-neutron monitor and the measurements with 14-MeV neutrons were intimately related

  1. A Novel Algorithm for Solving the Multidimensional Neutron Transport Equation on Massively Parallel Architectures

    Energy Technology Data Exchange (ETDEWEB)

    Azmy, Yousry

    2014-06-10

    We employ the Integral Transport Matrix Method (ITMM) as the kernel of new parallel solution methods for the discrete ordinates approximation of the within-group neutron transport equation. The ITMM abandons the repetitive mesh sweeps of the traditional source iterations (SI) scheme in favor of constructing stored operators that account for the direct coupling factors among all the cells' fluxes and between the cells' and boundary surfaces' fluxes. The main goals of this work are to develop the algorithms that construct these operators and employ them in the solution process, determine the most suitable way to parallelize the entire procedure, and evaluate the behavior and parallel performance of the developed methods with increasing number of processes, P. The fastest observed parallel solution method, Parallel Gauss-Seidel (PGS), was used in a weak scaling comparison with the PARTISN transport code, which uses the source iteration (SI) scheme parallelized with the Koch-baker-Alcouffe (KBA) method. Compared to the state-of-the-art SI-KBA with diffusion synthetic acceleration (DSA), this new method- even without acceleration/preconditioning-is completitive for optically thick problems as P is increased to the tens of thousands range. For the most optically thick cells tested, PGS reduced execution time by an approximate factor of three for problems with more than 130 million computational cells on P = 32,768. Moreover, the SI-DSA execution times's trend rises generally more steeply with increasing P than the PGS trend. Furthermore, the PGS method outperforms SI for the periodic heterogeneous layers (PHL) configuration problems. The PGS method outperforms SI and SI-DSA on as few as P = 16 for PHL problems and reduces execution time by a factor of ten or more for all problems considered with more than 2 million computational cells on P = 4.096.

  2. Analysis and development of spatial hp-refinement methods for solving the neutron transport equation

    International Nuclear Information System (INIS)

    The different neutronic parameters have to be calculated with a higher accuracy in order to design the 4. generation reactor cores. As memory storage and computation time are limited, adaptive methods are a solution to solve the neutron transport equation. The neutronic flux, solution of this equation, depends on the energy, angle and space. The different variables are successively discretized. The energy with a multigroup approach, considering the different quantities to be constant on each group, the angle by a collocation method called SN approximation. Once the energy and angle variable are discretized, a system of spatially-dependent hyperbolic equations has to be solved. Discontinuous finite elements are used to make possible the development of hp-refinement methods. Thus, the accuracy of the solution can be improved by spatial refinement (h-refinement), consisting into subdividing a cell into sub-cells, or by order refinement (p-refinement), by increasing the order of the polynomial basis. In this thesis, the properties of this methods are analyzed showing the importance of the regularity of the solution to choose the type of refinement. Thus, two error estimators are used to lead the refinement process. Whereas the first one requires high regularity hypothesis (analytical solution), the second one supposes only the minimal hypothesis required for the solution to exist. The comparison of both estimators is done on benchmarks where the analytic solution is known by the method of manufactured solutions. Thus, the behaviour of the solution as a regard of the regularity can be studied. It leads to a hp-refinement method using the two estimators. Then, a comparison is done with other existing methods on simplified but also realistic benchmarks coming from nuclear cores. These adaptive methods considerably reduces the computational cost and memory footprint. To further improve these two points, an approach with energy-dependent meshes is proposed. Actually, as the

  3. Summary of the FY 2005 Batteries for Advanced Transportation Technologies (BATT) research program annual review

    Energy Technology Data Exchange (ETDEWEB)

    None, None

    2005-08-01

    This document presents a summary of the evaluation and comments provided by the review panel for the FY 2005 Department of Energy (DOE) Batteries for Advanced Transportation Technologies (BATT) program annual review.

  4. The ion beam sputtering facility at KURRI: Coatings for advanced neutron optical devices

    Science.gov (United States)

    Hino, Masahiro; Oda, Tatsuro; Kitaguchi, Masaaki; Yamada, Norifumi L.; Tasaki, Seiji; Kawabata, Yuji

    2015-10-01

    We describe a film coating facility for the development of multilayer mirrors for use in neutron optical devices that handle slow neutron beams. Recently, we succeeded in fabricating a large neutron supermirror with high reflectivity using an ion beam sputtering system (KUR-IBS), as well as all neutron supermirrors in two neutron guide tubes at BL06 at J-PARC/MLF. We also realized a large flexible self-standing m=5 NiC/Ti supermirror and very small d-spacing (d=1.65 nm) multilayer sheets. In this paper, we present an overview of the performance and utility of non-magnetic neutron multilayer mirrors fabricated with the KUR-IBS

  5. The ion beam sputtering facility at KURRI: Coatings for advanced neutron optical devices

    Energy Technology Data Exchange (ETDEWEB)

    Hino, Masahiro, E-mail: hino@rri.kyoto-u.ac.jp [Research Reactor Institute, Kyoto university, Kumatori, Osaka 590-0494 (Japan); Oda, Tatsuro [Department of Nuclear Engineering, Kyoto University, Kyoto 615-8540 (Japan); Kitaguchi, Masaaki [Center for Experimental Studies, KMI, Nagoya University, Nagoya 464-8602 (Japan); Yamada, Norifumi L. [Neutron Science Laboratory, High Energy Accelerator Research Organization, 203-1 Shirakata, Tokai, Ibaraki 319-1106 (Japan); Tasaki, Seiji [Department of Nuclear Engineering, Kyoto University, Kyoto 615-8540 (Japan); Kawabata, Yuji [Research Reactor Institute, Kyoto university, Kumatori, Osaka 590-0494 (Japan)

    2015-10-11

    We describe a film coating facility for the development of multilayer mirrors for use in neutron optical devices that handle slow neutron beams. Recently, we succeeded in fabricating a large neutron supermirror with high reflectivity using an ion beam sputtering system (KUR-IBS), as well as all neutron supermirrors in two neutron guide tubes at BL06 at J-PARC/MLF. We also realized a large flexible self-standing m=5 NiC/Ti supermirror and very small d-spacing (d=1.65 nm) multilayer sheets. In this paper, we present an overview of the performance and utility of non-magnetic neutron multilayer mirrors fabricated with the KUR-IBS.

  6. Neutron scattering studies of structure, hydrothermal stability and transport in porous silica catalyst supports

    Science.gov (United States)

    Pollock, Rachel A.

    Mesoporous materials are interesting as catalyst supports, because molecules can move efficiently in and out of the pore network, but they must be stable in water if they are to be used for the production of biofuels. Before investigating hydrothermal stability and transport properties, the pore structure of SBA-15 was characterized using small angle neutron scattering (SANS) and non-local density functional theory (NLDFT) analysis of nitrogen sorption isotherms. A new Contrast Matching SANS method, using a range of probe molecules to directly probe the micropore size, gave a pore size distribution onset of 6 ± 0.2 Å, consistent with cylindrical pores formed from polymer template strands that unravel into the silica matrix. Diffraction intensity analysis of SANS measurements, combined with pore size distributions calculated from NLDFT, showed that the secondary pores are distributed relatively uniformly throughout the silica framework. The hydrothermal stability of SBA-15 was evaluated using a post-calcination hydrothermal treatment in both liquid and vapor phase water. The results were consistent with a degradation mechanism in which silica dissolves from regions of small positive curvature, e.g. near the entrance to the secondary pores, and is re-deposited deeper into the framework. Under water treatment at 115 °C, the mesopore diameter increases and the intra-wall void fraction decreases significantly. The behavior is similar for steam treatment, but occurs more slowly, suggesting that transport is faster when condensation occurs in the pores. Quasielastic neutron scattering (QENS) measurements of methane in SBA-15 probed the rotational and translational motion as a function of temperature and loading. A qualitative analysis of the QENS data suggested that for the initial dose of methane at 100 K, the self diffusion constant is similar in magnitude to literature values for methane in ZSM-5 and Y-zeolite, showing that the secondary pores trap methane and limit

  7. Graphite/Polyimide Composites. [conference on Composites for Advanced Space Transportation Systems

    Science.gov (United States)

    Dexter, H. B. (Editor); Davis, J. G., Jr. (Editor)

    1979-01-01

    Technology developed under the Composites for Advanced Space Transportation System Project is reported. Specific topics covered include fabrication, adhesives, test methods, structural integrity, design and analysis, advanced technology developments, high temperature polymer research, and the state of the art of graphite/polyimide composites.

  8. The coupling of the Star-Cd software to a whole-core neutron transport code Decart for PWR applications

    International Nuclear Information System (INIS)

    As part of a U.S.- Korea collaborative U.S. Department of Energy INERI project, a comprehensive high-fidelity reactor-core modeling capability is being developed for detailed analysis of existing and advanced PWR reactor designs. An essential element of the project has been the development of an interface between the computational fluid dynamics (CFD) module, STAR-CD, and the neutronics module, DeCART. Since the computational mesh for CFD and neutronics calculations are generally different, the capability to average and decompose data on these different meshes has been an important part of code coupling activities. An averaging process has been developed to extract neutronics zone temperatures in the fuel and coolant and to generate appropriate multi group cross sections and densities. Similar procedures have also been established to map the power distribution from the neutronics zones to the mesh structure used in the CFD module. Since MPI is used as the parallel model in STAR-CD and conflicts arise during initiation of a second level of MPI, the interface developed here is based on using TCP/IP protocol sockets to establish communication between the CFD and neutronics modules. Preliminary coupled calculations have been performed for PWR fuel assembly size problems and converged solutions have been achieved for a series of steady-state problems ranging from a single pin to a 1/8 model of a 17 x 17 PWR fuel assembly. (authors)

  9. Monte Carlo 2000 Conference : Advanced Monte Carlo for Radiation Physics, Particle Transport Simulation and Applications

    CERN Document Server

    Baräo, Fernando; Nakagawa, Masayuki; Távora, Luis; Vaz, Pedro

    2001-01-01

    This book focusses on the state of the art of Monte Carlo methods in radiation physics and particle transport simulation and applications, the latter involving in particular, the use and development of electron--gamma, neutron--gamma and hadronic codes. Besides the basic theory and the methods employed, special attention is paid to algorithm development for modeling, and the analysis of experiments and measurements in a variety of fields ranging from particle to medical physics.

  10. Proceedings of the 1986 workshop on advanced time-of-flight neutron powder diffraction

    International Nuclear Information System (INIS)

    This report contains abstracts of talks and summaries of discussions from a small workshop held to discuss the future of time-of-flight neutron powder diffraction and its implementation at the Los Alamos Neutron Scattering Center. 47 refs., 3 figs

  11. Design of Stopper of Prompt Gamma Neutron Activation Analysis Facility at China Advanced Research Reactor

    Institute of Scientific and Technical Information of China (English)

    2011-01-01

    The PGNAA facility consists of the filtered collimated neutron beam, the shielding of the whole facility, the control system, the detecting equipment and the data acquisition and analysis system. The neutron beam is filtered by a mono-crystalline bismuth filter,

  12. Advanced broad-band solid-state supermirror polarizers for cold neutrons

    CERN Document Server

    Petukhov, A K; Bigault, T; Courtois, P; Jullien, D; Soldner, T

    2016-01-01

    An ideal solid-state supermirror (SM) neutron polarizer assumes total reflection of neutrons from the SM coating for one spin-component and total absorption for the other, thus providing a perfectly polarized neutron beam at the exit. However, in practice, the substrate's neutron-nucleai optical potential does not match perfectly that for spin-down neutrons in the SM. For a positive step in the optical potential (as in a Fe/SiN(x) SM on Si substrate), this mismatch results in spin-independent total reflection for neutrons with small momentum transfer Q, limiting the useful neutron bandwidth in the low-Q region. To overcome this limitation, we propose to replace Si single-crystal substrates by media with higher optical potential than that for spin-down neutrons in the SM ferromagnetic layers. We found single-crystal sapphire and single-crystal quartz as good candidates for solid-state Fe/SiN(x) SM polarizers. To verify this idea, we coated a thick plate of single-crystal sapphire with a m=2.4 Fe/SiN(x) SM. At ...

  13. Neutron cross-sections for advanced nuclear systems: the n_TOF project at CERN

    Science.gov (United States)

    Barbagallo, M.; Mastromarco, M.; Colonna, N.; Altstadt, S.; Andrzejewski, J.; Audouin, L.; Bécares, V.; Bečvář, F.; Belloni, F.; Berthoumieux, E.; Billowes, J.; Bosnar, D.; Brugger, M.; Calviani, M.; Calviño, F.; Cano-Ott, D.; Carrapiço, C.; Cerutti, F.; Chiaveri, E.; Chin, M.; Cortés, G.; Cortés-Giraldo, M. A.; Diakaki, M.; Domingo-Pardo, C.; Duran, I.; Dressler, R.; Eleftheriadis, C.; Ferrari, A.; Fraval, K.; Ganesan, S.; García, A. R.; Giubrone, G.; Gonçalves, I. F.; González-Romero, E.; Griesmayer, E.; Guerrero, C.; Gunsing, F.; Hernández-Prieto, A.; Jenkins, D. G.; Jericha, E.; Kadi, Y.; Käppeler, F.; Karadimos, D.; Kivel, N.; Koehler, P.; Krtička, M.; Kroll, J.; Lampoudis, C.; Langer, C.; Leal-Cidoncha, E.; Lederer, C.; Leeb, H.; Leong, L. S.; Losito, R.; Manousos, A.; Marganiec, J.; Martínez, T.; Massimi, C.; Mastinu, P. F.; Mendoza, E.; Mengoni, A.; Milazzo, P. M.; Mingrone, F.; Mirea, M.; Mondalaers, W.; Paradela, C.; Pavlik, A.; Perkowski, J.; Plompen, A.; Praena, J.; Quesada, J. M.; Rauscher, T.; Reifarth, R.; Riego, A.; Rubbia, C.; Sabaté-Gilarte, M.; Sarmento, R.; Saxena, A.; Schillebeeckx, P.; Schmidt, S.; Schumann, D.; Tagliente, G.; Tain, J. L.; Tarrío, D.; Tassan-Got, L.; Tsinganis, A.; Valenta, S.; Vannini, G.; Variale, V.; Vaz, P.; Ventura, A.; Vermeulen, M. J.; Vlachoudis, V.; Vlastou, R.; Wallner, A.; Ware, T.; Weigand, M.; Weiß, C.; Wright, T.; Žugec, P.

    2014-12-01

    The study of neutron-induced reactions is of high relevance in a wide variety of fields, ranging from stellar nucleosynthesis and fundamental nuclear physics to applications of nuclear technology. In nuclear energy, high accuracy neutron data are needed for the development of Generation IV fast reactors and accelerator driven systems, these last aimed specifically at nuclear waste incineration, as well as for research on innovative fuel cycles. In this context, a high luminosity Neutron Time Of Flight facility, n_TOF, is operating at CERN since more than a decade, with the aim of providing new, high accuracy and high resolution neutron cross-sections. Thanks to the features of the neutron beam, a rich experimental program relevant to nuclear technology has been carried out so far. The program will be further expanded in the near future, thanks in particular to a new high-flux experimental area, now under construction.

  14. Experimental tests of an advanced proton-to-neutron converter at ISOLDE-CERN

    CERN Document Server

    Gottberg, A; Luis, R; Ramos, J P; Seiffert, C; Cimmino, S; Marzari, S; Crepieux, B; Manea, V; Wolf, R N; Wienholtz, F; Kreim, S; Fedosseev, V N; Marsh, B A; Rothe, S; Vaz, P; Marques, J G; Stora, T

    2014-01-01

    The suppression of isobaric contaminations is of growing importance for many scientific programs using radioactive isotopes produced at isotope separation on-line (ISOL) facilities, such as ISOLDE-CERN. A solid tungsten proton-to-neutron converter has been used for ten years to produce neutron-rich fission fragments from an UC x target while suppressing the production of neutron-deficient isobaric contaminants. The remaining contamination is mainly produced by primary protons that are scattered by the heavy neutron converter and finally impinge on the UC x target itself. Therefore, the knowledge of the energy-dependant cross-sections of proton and neutron induced fission events is crucial in order to evaluate future converter concepts.

  15. Neutron cross-sections for advanced nuclear systems: the n_TOF project at CERN

    Directory of Open Access Journals (Sweden)

    Barbagallo M.

    2014-01-01

    Full Text Available The study of neutron-induced reactions is of high relevance in a wide variety of fields, ranging from stellar nucleosynthesis and fundamental nuclear physics to applications of nuclear technology. In nuclear energy, high accuracy neutron data are needed for the development of Generation IV fast reactors and accelerator driven systems, these last aimed specifically at nuclear waste incineration, as well as for research on innovative fuel cycles. In this context, a high luminosity Neutron Time Of Flight facility, n_TOF, is operating at CERN since more than a decade, with the aim of providing new, high accuracy and high resolution neutron cross-sections. Thanks to the features of the neutron beam, a rich experimental program relevant to nuclear technology has been carried out so far. The program will be further expanded in the near future, thanks in particular to a new high-flux experimental area, now under construction.

  16. Recent advances in the understanding of the interaction of antidepressant drugs with serotonin and norepinephrine transporters

    DEFF Research Database (Denmark)

    Andersen, Jacob; Kristensen, Anders Skov; Bang-Andersen, Benny;

    2009-01-01

    The biogenic monoamine transporters are integral membrane proteins that perform active transport of extracellular dopamine, serotonin and norepinephrine into cells. These transporters are targets for therapeutic agents such as antidepressants, as well as addictive substances such as cocaine...... and amphetamine. Seminal advances in the understanding of the structure and function of this transporter family have recently been accomplished by structural studies of a bacterial transporter, as well as medicinal chemistry and pharmacological studies of mammalian transporters. This feature article focuses...... on antidepressant drugs that act on the serotonin and/or the norepinephrine transporters. Specifically, we focus on structure-activity relationships of these drugs with emphasis on relationships between their molecular properties and the current knowledge of transporter structure....

  17. Improvements in Neutronics/Thermal-Hydraulics Coupling in Two-Phase Flow Systems Using Stochastic-Mixture Transport Models

    CERN Document Server

    Palmer, T S

    2003-01-01

    In this NEER project, researchers from Oregon State University have investigated the limitations of the treatment of two-phase coolants as a homogeneous mixture in neutron transport calculations. Improved methods of calculating the neutron distribution in binary stochastic mixtures have been developed over the past 10-15 years and are readily available in the transport literature. These methods are computationally more expensive than the homogeneous (or atomic mix) models, but can give much more accurate estimates of ensemble average fluxes and reaction rates provided statistical descriptions of the distributions of the two materials are know. A thorough review of the two-phase flow literature has been completed and the relevant mixture distributions have been identified. Using these distributions, we have performed Monte Carlo criticality calculations of fuel assemblies to assess the accuracy of the atomic mix approximation when compared to a resolved treatment of the two-phase coolant. To understand the ben...

  18. Transport and mixing of r-process elements in neutron star binary merger blast waves

    CERN Document Server

    Montes, Gabriela; Naiman, Jill; Shen, Sijing; Lee, William H

    2016-01-01

    The r-process nuclei are robustly synthesized in the material ejected during a neutron star binary merger (NSBM), as tidal torques transport angular momentum and energy through the outer Lagrange point in the form of a vast tidal tail. If NSBM are indeed solely responsible for the solar system r- process abundances, a galaxy like our own would require to host a few NSBM per million years, with each event ejecting, on average, about 5x10^{-2} M_sun of r-process material. Because the ejecta velocities in the tidal tail are significantly larger than in ordinary supernovae, NSBM deposit a comparable amount of energy into the interstellar medium (ISM). In contrast to extensive efforts studying spherical models for supernova remnant evolution, calculations quantifying the impact of NSBM ejecta in the ISM have been lacking. To better understand their evolution in a cosmological context, we perform a suite of three-dimensional hydrodynamic simulations with optically-thin radiative cooling of isolated NSBM ejecta expa...

  19. A time-dependent neutron transport method of characteristics formulation with time derivative propagation

    Science.gov (United States)

    Hoffman, Adam J.; Lee, John C.

    2016-02-01

    A new time-dependent Method of Characteristics (MOC) formulation for nuclear reactor kinetics was developed utilizing angular flux time-derivative propagation. This method avoids the requirement of storing the angular flux at previous points in time to represent a discretized time derivative; instead, an equation for the angular flux time derivative along 1D spatial characteristics is derived and solved concurrently with the 1D transport characteristic equation. This approach allows the angular flux time derivative to be recast principally in terms of the neutron source time derivatives, which are approximated to high-order accuracy using the backward differentiation formula (BDF). This approach, called Source Derivative Propagation (SDP), drastically reduces the memory requirements of time-dependent MOC relative to methods that require storing the angular flux. An SDP method was developed for 2D and 3D applications and implemented in the computer code DeCART in 2D. DeCART was used to model two reactor transient benchmarks: a modified TWIGL problem and a C5G7 transient. The SDP method accurately and efficiently replicated the solution of the conventional time-dependent MOC method using two orders of magnitude less memory.

  20. Inversion of Source and Transport Parameters of Relativistic SEPs from Neutron Monitor Data

    Science.gov (United States)

    Agueda, Neus; Bütikofer, Rolf; Vainio, Rami; Heber, Bernd; Afanasiev, Alexander; Malandraki, Olga E.

    2016-04-01

    We present a new methodology to study the release processes of relativistic solar energetic particles (SEPs) based on the direct inversion of Ground Level Enhancements (GLEs) observed by the worldwide network of neutron monitors (NMs). The new approach makes use of several models, including: the propagation of relativistic SEPs from the Sun to the Earth, their transport in the Earth's magnetosphere and atmosphere, as well as the detection of the nucleon component of the secondary cosmic rays by ground based NMs. The combination of these models allows us to compute the expected ground-level NM counting rates for a series of instantaneous releases from the Sun. The amplitudes of the source components are then inferred by fitting the NM observations with the modeled NM counting rate increases. Within the HESPERIA project, we will develop the first software package for the direct inversion of GLEs and we will make it freely available for the solar and heliospheric communities. Acknowledgement: This work has received funding from the European Union's Horizon 2020 research and innovation programme under grant agreement No 637324.

  1. Resonance self-shielding methodology of new neutron transport code STREAM

    International Nuclear Information System (INIS)

    This paper reports on the development and verification of three new resonance self-shielding methods. The verifications were performed using the new neutron transport code, STREAM. The new methodologies encompass the extension of energy range for resonance treatment, the development of optimum rational approximation, and the application of resonance treatment to isotopes in the cladding region. (1) The extended resonance energy range treatment has been developed to treat the resonances below 4 eV of three resonance isotopes and shows significant improvements in the accuracy of effective cross sections (XSs) in that energy range. (2) The optimum rational approximation can eliminate the geometric limitations of the conventional approach of equivalence theory and can also improve the accuracy of fuel escape probability. (3) The cladding resonance treatment method makes it possible to treat resonances in cladding material which have not been treated explicitly in the conventional methods. These three new methods have been implemented in the new lattice physics code STREAM and the improvement in the accuracy of effective XSs is demonstrated through detailed verification calculations. (author)

  2. RELAP5 analyses of two hypothetical flow reversal events for the advanced neutron source reactor

    Energy Technology Data Exchange (ETDEWEB)

    Chen, N.C.J.; Wendel, M.W.; Yoder, G.L. Jr. [Oak Ridge National Lab., TN (United States)

    1995-09-01

    This paper presents RELAP5 results of two hypothetical, low flow transients analyzed as part of the Advanced Neutron Source Reactor safety program. The reactor design features four independent coolant loops (three active and one in standby), each containing a main curculation pump (with battery powered pony motor), heat exchanger, an accumulator, and a check valve. The first transient assumes one of these pumps fails, and additionally, that the check valve in that loop remains stuck in the open position. This accident is considered extremely unlikely. Flow reverses in this loop, reducing the core flow because much of the coolant is diverted from the intact loops back through the failed loop. The second transient examines a 102-mm-diam instantaneous pipe break near the core inlet (the worst break location). A break is assumed to occur 90 s after a total loss-of-offsite power. Core flow reversal occurs because accumulator injection overpowers the diminishing pump flow. Safety margins are evaluated against four thermal limits: T{sub wall}=T{sub sat}, incipient boiling, onset of significant void, and critical heat flux. For the first transient, the results show that these limits are not exceeded (at a 95% non-exceedance probability level) if the pony motor battery lasts 30 minutes (the present design value). For the second transient, the results show that the closest approach of the fuel surface temperature to the local saturation temperature during core flow reversal is about 39{degrees}C. Therefore the fuel remains cool during this transient. Although this work is done specifically for the ANSR geometry and operating conditions, the general conclusions may be applicable to other highly subcooled reactor systems.

  3. Update to advanced neutron source steady-state thermal-hydraulic report

    Energy Technology Data Exchange (ETDEWEB)

    Yoder, G.L.; Carbajo, J.J.; Morris, D.G.; Nelson, W.R.

    1996-05-01

    This report is intended to be a supplement to ORNL/TM-12398, Steady-State Thermal-Hydraulic Design Analysis of the Advanced Neutron Source Reactor. It updates the core thermal-hydrualic design to the latest three-element configuration and also provides the most recent information on the thermal-hydraulic statistical uncertainty analysis. In addition, it includes calculations of beam tube cooling and control rod lift forces, which were not addressed in the initial report. This report describes work that is a snapshot in time as it stood at the end of the project. The three-element core calculations include a description of changes made to the overall coolant system; however, most of the analysis is focused on fuel loading thermal-hydraulic calculations. This analysis uses updated uncertainty values and indicates that a two-dimensional fuel grading in the three-element core would still be necessary to meet the desired operating and safety criteria. Analysis of cooling in the reflector tank examines various cooling options for the reflector tank components. This work investigated multiple forced convection designs as well as natural convection cooling requirements. Lift forces on the inner control rods caused by the upward coolant flow were also examined. Initial control rod designs were such that a sheared control rod would tend to lift because of flow forces. Design changes were recommended that would eliminate this issue. They included geometry changes to the inner control rod cooling channels, changes to the orificing in the central hole region, and reduction of inner control rod coolant velocity.

  4. A 10(9) neutrons/pulse transportable pulsed D-D neutron source based on flexible head plasma focus unit.

    Science.gov (United States)

    Niranjan, Ram; Rout, R K; Srivastava, R; Kaushik, T C; Gupta, Satish C

    2016-03-01

    A 17 kJ transportable plasma focus (PF) device with flexible transmission lines is developed and is characterized. Six custom made capacitors are used for the capacitor bank (CB). The common high voltage plate of the CB is fixed to a centrally triggered spark gap switch. The output of the switch is coupled to the PF head through forty-eight 5 m long RG213 cables. The CB has a quarter time-period of 4 μs and an estimated current of 506 kA is delivered to the PF device at 17 kJ (60 μF, 24 kV) energy. The average neutron yield measured using silver activation detector in the radial direction is (7.1 ± 1.4) × 10(8) neutrons/shot over 4π sr at 5 mbar optimum D2 pressure. The average neutron yield is more in the axial direction with an anisotropy factor of 1.33 ± 0.18. The average neutron energies estimated in the axial as well as in the radial directions are (2.90 ± 0.20) MeV and (2.58 ± 0.20) MeV, respectively. The flexibility of the PF head makes it useful for many applications where the source orientation and the location are important factors. The influence of electromagnetic interferences from the CB as well as from the spark gap on applications area can be avoided by putting a suitable barrier between the bank and the PF head. PMID:27036774

  5. TRIPOLI capabilities proved by a set of solved problems. [Monte Carlo neutron and gamma ray transport code

    Energy Technology Data Exchange (ETDEWEB)

    Vergnaud, T.; Nimal, J.C. (CEA Centre d' Etudes Nucleaires de Saclay, 91 - Gif-sur-Yvette (France))

    1990-01-01

    The three-dimensional polycinetic Monte Carlo particle transport code TRIPOLI has been under development in the French Shielding Laboratory at Saclay since 1965. TRIPOLI-1 began to run in 1970 and became TRIPOLI-2 in 1978: since then its capabilities have been improved and many studies have been performed. TRIPOLI can treat stationary or time dependent problems in shielding and in neutronics. Some examples of solved problems are presented to demonstrate the many possibilities of the system. (author).

  6. Citymobil, advanced transport for the urban environment : an update

    NARCIS (Netherlands)

    Dijke, J.P. van; Schijndel-de Nooij, M. van

    2012-01-01

    CityMobil was an integrated project in the Sixth Framework Program of the European Union. The objective of the project was to achieve more effective organization of urban transport for more rational use of motorized traffic with less congestion and pollution, safer driving, a higher quality of livin

  7. Neutron transport in hexagonal reactor cores modeled by trigonal-geometry diffusion and simplified P{sub 3} nodal methods

    Energy Technology Data Exchange (ETDEWEB)

    Duerigen, Susan

    2013-05-15

    The superior advantage of a nodal method for reactor cores with hexagonal fuel assemblies discretized as cells consisting of equilateral triangles is its mesh refinement capability. In this thesis, a diffusion and a simplified P{sub 3} (or SP{sub 3}) neutron transport nodal method are developed based on trigonal geometry. Both models are implemented in the reactor dynamics code DYN3D. As yet, no other well-established nodal core analysis code comprises an SP{sub 3} transport theory model based on trigonal meshes. The development of two methods based on different neutron transport approximations but using identical underlying spatial trigonal discretization allows a profound comparative analysis of both methods with regard to their mathematical derivations, nodal expansion approaches, solution procedures, and their physical performance. The developed nodal approaches can be regarded as a hybrid NEM/AFEN form. They are based on the transverse-integration procedure, which renders them computationally efficient, and they use a combination of polynomial and exponential functions to represent the neutron flux moments of the SP{sub 3} and diffusion equations, which guarantees high accuracy. The SP{sub 3} equations are derived in within-group form thus being of diffusion type. On this basis, the conventional diffusion solver structure can be retained also for the solution of the SP{sub 3} transport problem. The verification analysis provides proof of the methodological reliability of both trigonal DYN3D models. By means of diverse hexagonal academic benchmark and realistic detailed-geometry full-transport-theory problems, the superiority of the SP{sub 3} transport over the diffusion model is demonstrated in cases with pronounced anisotropy effects, which is, e.g., highly relevant to the modeling of fuel assemblies comprising absorber material.

  8. Neutron transport in hexagonal reactor cores modeled by trigonal-geometry diffusion and simplified P3 nodal methods

    International Nuclear Information System (INIS)

    The superior advantage of a nodal method for reactor cores with hexagonal fuel assemblies discretized as cells consisting of equilateral triangles is its mesh refinement capability. In this thesis, a diffusion and a simplified P3 (or SP3) neutron transport nodal method are developed based on trigonal geometry. Both models are implemented in the reactor dynamics code DYN3D. As yet, no other well-established nodal core analysis code comprises an SP3 transport theory model based on trigonal meshes. The development of two methods based on different neutron transport approximations but using identical underlying spatial trigonal discretization allows a profound comparative analysis of both methods with regard to their mathematical derivations, nodal expansion approaches, solution procedures, and their physical performance. The developed nodal approaches can be regarded as a hybrid NEM/AFEN form. They are based on the transverse-integration procedure, which renders them computationally efficient, and they use a combination of polynomial and exponential functions to represent the neutron flux moments of the SP3 and diffusion equations, which guarantees high accuracy. The SP3 equations are derived in within-group form thus being of diffusion type. On this basis, the conventional diffusion solver structure can be retained also for the solution of the SP3 transport problem. The verification analysis provides proof of the methodological reliability of both trigonal DYN3D models. By means of diverse hexagonal academic benchmark and realistic detailed-geometry full-transport-theory problems, the superiority of the SP3 transport over the diffusion model is demonstrated in cases with pronounced anisotropy effects, which is, e.g., highly relevant to the modeling of fuel assemblies comprising absorber material.

  9. Proceedings of the 182nd basic science seminar (The workshop on neutron structural biology ) 'New frontiers of structural biology advanced by solution scattering'

    Energy Technology Data Exchange (ETDEWEB)

    Fujiwara, Satoru (ed.) [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2001-03-01

    182nd advanced science seminar (the workshop on neutron structural biology) was held in February 9-10, 2000 at Tokai. Thirty-six participants from universities, research institutes, and private companies took part in the workshop, and total of 24 lectures were given. This proceedings collects abstracts, the figures and tables, which the speakers used in their lectures. The proceedings contains two reviews from the point of view of x-ray and neutron scatterings, and six subjects (21 papers) including neutron and x-ray scattering in the era of structure genomics, structural changes detected with solution scattering, a new way in structural biology opened by neutron crystallography and neutron scattering, x-ray sources and detectors, simulation and solution scattering, and neutron sources and detectors. (Kazumata, Y.)

  10. Quantitative analysis of saline transport and crystallization damage in porous limestone visualized by neutron and X-ray imaging

    OpenAIRE

    Derluyn, Hannelore; Griffa, Michele; Mannes, David; Jerjen, Iwan; Dewanckele, Jan; Vontobel, Peter; Derome, Dominique; Cnudde, Veerle; Lehmann, Eberhard; Carmeliet, Jan

    2011-01-01

    This article presents coupled data on saline transport, pore filling due to salt crystallization and resulting salt damage in Savonnières limestone. This is achieved by combining the non-destructive techniques of neutron radiography - for transport imaging - and X-ray microtomography - for pore structure and fracture visualization – applied to the same sample when subjecting it to consecutive wetting-drying cycles. Capillary uptake of water, 1.4m sodium sulfate and 5.8m sodium chloride soluti...

  11. Passive and Active Fast-Neutron Imaging in Support of Advanced Fuel Cycle Initiative Safeguards Campaign

    Energy Technology Data Exchange (ETDEWEB)

    Blackston, Matthew A [ORNL; Hausladen, Paul [ORNL

    2010-04-01

    Results from safeguards-related passive and active coded-aperture fast-neutron imaging measurements of plutonium and highly enriched uranium (HEU) material configurations performed at Idaho National Laboratory s Zero Power Physics Reactor facility are presented. The imaging measurements indicate that it is feasible to use fast neutron imaging in a variety of safeguards-related tasks, such as monitoring storage, evaluating holdup deposits in situ, or identifying individual leached hulls still containing fuel. The present work also presents the first demonstration of imaging of differential die away fast neutrons.

  12. APOLLO-2: An advanced transport code for LWRs

    International Nuclear Information System (INIS)

    APOLLO-2 is a fully modular code in which each module corresponds to a specific task: access to the cross-sections libraries, creation of isotopes medium or mixtures, geometry definition, self-shielding calculations, computation of multigroup collision probabilities, flux solver, depletion calculations, transport-transport or transport-diffusion equivalence process, SN calculations, etc... Modules communicate exclusively by ''objects'' containing structured data, these objects are identified and handled by user's given names. Among the major improvements offered by APOLLO-2 the modelization of the self-shielding: it is possible now to deal with a great precision, checked versus Montecarlo calculations, a fuel rod divided into several concentric rings. So the total production of Plutonium is quite better estimated than before and its radial distribution may be predicted also with a good accuracy. Thanks to the versatility of the code some reference calculations and routine ones may be compared easily because only one parameter is changed; for example the self-shielding approximations are modified, the libraries or the flux solver being exactly the same. Other interesting features have been introduced in APOLLO-2: the new isotopes JEF.2 are available in 99 and 172 energy groups libraries, the surface leakage model improves the calculation of the control rod efficiency, the flux-current method allows faster calculations, the possibility of an automatic convergence checking during the depletion calculations coupled with fully automatic corrections, heterogeneous diffusion coefficients used for voiding analysis. 17 refs, 1 tab

  13. Monte Carlo calculation of ''skyshine'' neutron dose from ALS [Advanced Light Source

    International Nuclear Information System (INIS)

    This report discusses the following topics on ''skyshine'' neutron dose from ALS: Sources of radiation; ALS modeling for skyshine calculations; MORSE Monte-Carlo; Implementation of MORSE; Results of skyshine calculations from storage ring; and Comparison of MORSE shielding calculations

  14. Prospects For High Frequency Burst Searches Following Binary Neutron Star Coalescence With Advanced Gravitational Wave Detectors

    CERN Document Server

    Clark, J; Cadonati, L; Janka, H -T; Pankow, C; Stergioulas, N

    2014-01-01

    The equation of state plays a critical role in the physics of the merger of two neutron stars. Recent numerical simulations with microphysical equation of state suggest the outcome of such events depends on the mass of the neutron stars. For less massive systems, simulations favor the formation of a hypermassive, quasi-stable neutron star, whose oscillations produce a short, high frequency burst of gravitational radiation. Its dominant frequency content is tightly correlated with the radius of the neutron star, and its measurement can be used to constrain the supranuclear equation of state. In contrast, the merger of higher mass systems results in prompt gravitational collapse to a black hole. We have developed an algorithm which combines waveform reconstruction from a morphology-independent search for gravitational wave transients with Bayesian model selection, to discriminate between post-merger scenarios and accurately measure the dominant oscillation frequency. We demonstrate the efficacy of the method us...

  15. Analytical solution of neutron transport equation in an annular reactor with a rotating pulsed source; Resolucao analitica da equacao de transporte de neutrons em um reator anelar com fonte pulsada rotativa

    Energy Technology Data Exchange (ETDEWEB)

    Teixeira, Paulo Cleber Mendonca

    2002-12-01

    In this study, an analytical solution of the neutron transport equation in an annular reactor is presented with a short and rotating neutron source of the type S(x) {delta} (x- Vt), where V is the speed of annular pulsed reactor. The study is an extension of a previous study by Williams [12] carried out with a pulsed source of the type S(x) {delta} (t). In the new concept of annular pulsed reactor designed to produce continuous high flux, the core consists of a subcritical annular geometry pulsed by a rotating modulator, producing local super prompt critical condition, thereby giving origin to a rotating neutron pulse. An analytical solution is obtained by opening up of the annular geometry and applying one energy group transport theory in one dimension using applied mathematical techniques of Laplace transform and Complex Variables. The general solution for the flux consists of a fundamental mode, a finite number of harmonics and a transient integral. A condition which limits the number of harmonics depending upon the circumference of the annular geometry has been obtained. Inverse Laplace transform technique is used to analyse instability condition in annular reactor core. A regenerator parameter in conjunction with perimeter of the ring and nuclear properties is used to obtain stable and unstable harmonics and to verify if these exist. It is found that the solution does not present instability in the conditions stated in the new concept of annular pulsed reactor. (author)

  16. Advances in carbon dioxide compression and pipeline transportation processes

    CERN Document Server

    Witkowski, Andrzej; Majkut, Mirosław; Rulik, Sebastian; Stolecka, Katarzyna

    2015-01-01

    Providing a comprehensive analysis of CO2 compression, transportation processes and safety issues for post combustion CO2 capture applications for a 900 MW pulverized hard coal-fired power plant, this book assesses techniques for boosting the pressure of CO2 to pipeline pressure values with a minimal amount of energy. Four different types of compressors are examined in detail: a conventional multistage centrifugal compressor, integrally geared centrifugal compressor, supersonic shock wave compressor, and pump machines. The study demonstrates that the total compression power is closely related

  17. Advanced transportation concept for round-trip space travel

    Science.gov (United States)

    Yen, Chen-Wan L.

    1988-01-01

    A departure from the conventional concept of round-trip space travel is introduced. It is shown that a substantial reduction in the initial load required of the Shuttle or other launch vehicle can be achieved by staging the ascent orbit and leaving fuel for the return trip at each stage of the orbit. Examples of round trips from a low-inclination LEO to a high-inclination LEO and from an LEO to a GEO are used to show the merits of the new concept. Potential problem areas and research needed for the development of an efficient space transportation network are discussed.

  18. Acoustic charge transport technology investigation for advanced development transponder

    Science.gov (United States)

    Kayalar, S.

    1993-01-01

    Acoustic charge transport (ACT) technology has provided a basis for a new family of analog signal processors, including a programmable transversal filter (PTF). Through monolithic integration of ACT delay lines with GaAs metal semiconductor field effect transistor (MESFET) digital memory and controllers, these devices significantly extend the performance of PTF's. This article introduces the basic operation of these devices and summarizes their present and future specifications. The production and testing of these devices indicate that this new technology is a promising one for future space applications.

  19. Design and layout decisions for refuelling system of advanced fast neutron reactor

    International Nuclear Information System (INIS)

    The experience in operation of BOR-60, BN-350 and BN-600 power units, as well as development of refuelling systems for BN-800 power unit, allows developing of refuelling system for BN-1200 advanced reactor of new generation. The refuelling system was developed on the basis of possible technical decisions aimed at improvement of safety and technical-and-economic indices. Structural layout of BN-1200 reactor refuelling system is given. Main differences in BN-1200 reactor refuelling system as compared with BN-800 reactor are given. Design features of refuelling equipment are: - BN-1200 reactor has a split large rotating plug to allow transporting of its components by railway with subsequent assembling at site; - the refuelling box is fabricated in the form of sectional parallelepiped to allow transporting of its components by railway with subsequent assembling at site; - one 'direct' refuelling mechanism and one cantilever' refuelling mechanism are used to refuel rarely replaced protection assemblies that allows reducing of overall dimensions of rotating plugs; - the vertical elevator is arranged on the oval plug installed on the reactor cover. The upper structure with elevator drive rotates together with the elevator plug under rotary drive located on the oval plug. The vertical elevator allows sufficient reduction of refuelling box; - the refuelling machine runs on straight-line rails. The vertical elevator, gas gate valve on reactor refuelling channel, non-use of spent FA drum and enhanced radiation protection on the column of refuelling box machine allows reduction of specific materials consumption of BN-1200 reactor refuelling system by more than 10 times as compared with BN-800 reactor. To verify refuelling equipment operability the following experiments are planned: - mastering of gripper design for 'direct' refuelling mechanism and refuelling machine; - mastering of 'cantilever' for refuelling mechanism; - mastering of fresh FA conveyor design. As for the

  20. Advanced rocket propulsion technology assessment for future space transportation

    Science.gov (United States)

    Wilhite, A. W.

    1982-01-01

    Single-stage and two-stage launch vehicles were evaluated for various levels of propulsion technology and payloads. The evaluation included tradeoffs between ascent flight performance and vehicle sizing that were driven by engine mass, specific impulse, and propellant requirements. Numerous mission, flight, and vehicle-related requirements and constraints were satisfied in the design process. The results showed that advanced technology had a large effect on reducing both single- and two-stage vehicle size. High-pressure hydrocarbon-fueled engines that were burned in parallel with two-position nozzle hydrogen-fueled engines reduced dry mass by 23% for the two-stage vehicle and 28% for the single-stage vehicle as compared to an all-hydrogen-fueled system. The dual-expander engine reduced single-stage vehicle dry mass by 41%. Using advanced technology, the single-stage vehicle became comparable in size and sensitivity to that of the two-stage vehicle for small payloads.

  1. Radiation transport and shielding information, computer codes, and nuclear data for use in CTR neutronics research and development

    International Nuclear Information System (INIS)

    The activities of the Radiation Shielding Information Center (RSIC) of the Oak Ridge National Laboratory are being utilized in support of fusion reactor technology. The major activities of RSIC include the operation of a computer-based information storage and retrieval system, the collection, packaging, and distribution of large computer codes, and the compilation and dissemination of processed and evaluated data libraries, with particular emphasis on neutron and gamma-ray cross-section data. The Center has acquired thirteen years of experience in serving fission reactor, weapons, and accelerator shielding research communities, and the extension of its technical base to fusion reactor research represents a logical progression. RSIC is currently working with fusion reactor researchers and contractors in computer code development to provide tested radiation transport and shielding codes and data library packages. Of significant interest to the CTR community are the 100 energy group neutron and 21 energy group gamma-ray coupled cross-section data package (DLC-37) for neutronics studies, a comprehensive 171 energy group neutron and 36 energy group gamma-ray coupled cross-section data base with retrieval programs, including resonance self-shielding, that are tailored to CTR application, and a data base for the generation of energy-dependent atomic displacement and gas production cross sections and heavy-particle-recoil spectra for estimating radiation damage to CTR structural components

  2. Development of a new neutron shielding material, TN trademark Resin Vyal for transport/storage casks for radioactive materials

    International Nuclear Information System (INIS)

    TN trademark Resin Vyal, a patent pending composite, is a new neutron shielding material developed by COGEMA LOGISTICS for transport/storage casks of radioactive materials (spent fuel, reprocessed fuel,..). This material is composed of a thermosetting resin (vinylester resin in solution of styrene) and two mineral fillers (alumine hydrate and zinc borate). Its shielding ability for neutron radiation is related to a high hydrogen content (for slowing down neutrons) and a high boron content (for absorbing neutrons). Source of hydrogen is organic matrix (resin) and alumine hydrate; source of boron is zinc borate. Atomic concentrations are equal to 5.1022 at/cm3 for hydrogen and 9.1020 at/cm3 for boron. Due to the flame retardant character of components, the final material has a good fire resistance (it is auto-extinguishable). Its density is equal to 1,8. The manufacturing process of these material is easy: it consists in mixing the fillers and pouring in-situ (in cask); so, the curing of this composite, which leads to a three-dimensional structure, takes place at ambient temperature. Temperature resistance of this material was evaluated by performing exposition tests of samples at different temperatures (150 C to 170 C). According to tests results, its maximal temperature of use was confirmed equal to 160 C

  3. Development of a new neutron shielding material, TN trademark Resin Vyal for transport/storage casks for radioactive materials

    Energy Technology Data Exchange (ETDEWEB)

    Abadie, P. [COGEMA Logistics (AREVA Group), Saint-Quentin-en-Yvelines (France)

    2004-07-01

    TN trademark Resin Vyal, a patent pending composite, is a new neutron shielding material developed by COGEMA LOGISTICS for transport/storage casks of radioactive materials (spent fuel, reprocessed fuel,..). This material is composed of a thermosetting resin (vinylester resin in solution of styrene) and two mineral fillers (alumine hydrate and zinc borate). Its shielding ability for neutron radiation is related to a high hydrogen content (for slowing down neutrons) and a high boron content (for absorbing neutrons). Source of hydrogen is organic matrix (resin) and alumine hydrate; source of boron is zinc borate. Atomic concentrations are equal to 5.10{sup 22} at/cm{sup 3} for hydrogen and 9.10{sup 20} at/cm{sup 3} for boron. Due to the flame retardant character of components, the final material has a good fire resistance (it is auto-extinguishable). Its density is equal to 1,8. The manufacturing process of these material is easy: it consists in mixing the fillers and pouring in-situ (in cask); so, the curing of this composite, which leads to a three-dimensional structure, takes place at ambient temperature. Temperature resistance of this material was evaluated by performing exposition tests of samples at different temperatures (150 C to 170 C). According to tests results, its maximal temperature of use was confirmed equal to 160 C.

  4. Advanced neutron source reactor thermal-hydraulic test loop facility description

    Energy Technology Data Exchange (ETDEWEB)

    Felde, D.K.; Farquharson, G.; Hardy, J.H.; King, J.F.; McFee, M.T.; Montgomery, B.H.; Pawel, R.E.; Power, B.H.; Shourbaji, A.A.; Siman-Tov, M.; Wood, R.J.; Yoder, G.L.

    1994-02-01

    The Thermal-Hydraulic Test Loop (THTL) is a facility for experiments constructed to support the development of the Advanced Neutron Source Reactor (ANSR) at Oak Ridge National Laboratory. The ANSR is both cooled and moderated by heavy water and uses uranium silicide fuel. The core is composed of two coaxial fuel-element annuli, each of different diameter. There are 684 parallel aluminum-clad fuel plates (252 in the inner-lower core and 432 in the outer-upper core) arranged in an involute geometry that effectively creates an array of thin rectangular flow channels. Both the fuel plates and the coolant channels are 1.27 mm thick, with a span of 87 mm (lower core), 70 mm (upper core), and 507-mm heated length. The coolant flows vertically upwards at a mass flux of 27 Mg/m{sup 2}s (inlet velocity of 25 m/s) with an inlet temperature of 45{degrees}C and inlet pressure of 3.2 MPa. The average and peak heat fluxes are approximately 6 and 12 MW/m{sup 2}, respectively. The availability of experimental data for both flow excursion (FE) and true critical heat flux (CHF) at the conditions applicable to the ANSR is very limited. The THTL was designed and built to simulate a full-length coolant subchannel of the core, allowing experimental determination of thermal limits under the expected ANSR thermal-hydraulic conditions. For these experimental studies, the involute-shaped fuel plates of the ANSR core with the narrow 1.27-mm flow gap are represented by a narrow rectangular channel. Tests in the THTL will provide both single- and two-phase thermal-hydraulic information. The specific phenomena that are to be examined are (1) single-phase heat-transfer coefficients and friction factors, (2) the point of incipient boiling, (3) nucleate boiling heat-transfer coefficients, (4) two-phase pressure-drop characteristics in the nucleate boiling regime, (5) flow instability limits, and (6) CHF limits.

  5. A 2D/1D coupling neutron transport method based on the matrix MOC and NEM methods

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, H.; Zheng, Y.; Wu, H.; Cao, L. [School of Nuclear Science and Technology, Xi' an Jiaotong University, No. 28, Xianning West Road, Xi' an, Shaanxi 710049 (China)

    2013-07-01

    A new 2D/1D coupling method based on the matrix MOC method (MMOC) and nodal expansion method (NEM) is proposed for solving the three-dimensional heterogeneous neutron transport problem. The MMOC method, used for radial two-dimensional calculation, constructs a response matrix between source and flux with only one sweep and then solves the linear system by using the restarted GMRES algorithm instead of the traditional trajectory sweeping process during within-group iteration for angular flux update. Long characteristics are generated by using the customization of commercial software AutoCAD. A one-dimensional diffusion calculation is carried out in the axial direction by employing the NEM method. The 2D and ID solutions are coupled through the transverse leakage items. The 3D CMFD method is used to ensure the global neutron balance and adjust the different convergence properties of the radial and axial solvers. A computational code is developed based on these theories. Two benchmarks are calculated to verify the coupling method and the code. It is observed that the corresponding numerical results agree well with references, which indicates that the new method is capable of solving the 3D heterogeneous neutron transport problem directly. (authors)

  6. A 2D/1D coupling neutron transport method based on the matrix MOC and NEM methods

    International Nuclear Information System (INIS)

    A new 2D/1D coupling method based on the matrix MOC method (MMOC) and nodal expansion method (NEM) is proposed for solving the three-dimensional heterogeneous neutron transport problem. The MMOC method, used for radial two-dimensional calculation, constructs a response matrix between source and flux with only one sweep and then solves the linear system by using the restarted GMRES algorithm instead of the traditional trajectory sweeping process during within-group iteration for angular flux update. Long characteristics are generated by using the customization of commercial software AutoCAD. A one-dimensional diffusion calculation is carried out in the axial direction by employing the NEM method. The 2D and ID solutions are coupled through the transverse leakage items. The 3D CMFD method is used to ensure the global neutron balance and adjust the different convergence properties of the radial and axial solvers. A computational code is developed based on these theories. Two benchmarks are calculated to verify the coupling method and the code. It is observed that the corresponding numerical results agree well with references, which indicates that the new method is capable of solving the 3D heterogeneous neutron transport problem directly. (authors)

  7. The Development of WARP - A Framework for Continuous Energy Monte Carlo Neutron Transport in General 3D Geometries on GPUs

    Science.gov (United States)

    Bergmann, Ryan

    Graphics processing units, or GPUs, have gradually increased in computational power from the small, job-specific boards of the early 1990s to the programmable powerhouses of today. Compared to more common central processing units, or CPUs, GPUs have a higher aggregate memory bandwidth, much higher floating-point operations per second (FLOPS), and lower energy consumption per FLOP. Because one of the main obstacles in exascale computing is power consumption, many new supercomputing platforms are gaining much of their computational capacity by incorporating GPUs into their compute nodes. Since CPU-optimized parallel algorithms are not directly portable to GPU architectures (or at least not without losing substantial performance), transport codes need to be rewritten to execute efficiently on GPUs. Unless this is done, reactor simulations cannot take full advantage of these new supercomputers. WARP, which can stand for ``Weaving All the Random Particles,'' is a three-dimensional (3D) continuous energy Monte Carlo neutron transport code developed in this work as to efficiently implement a continuous energy Monte Carlo neutron transport algorithm on a GPU. WARP accelerates Monte Carlo simulations while preserving the benefits of using the Monte Carlo Method, namely, very few physical and geometrical simplifications. WARP is able to calculate multiplication factors, flux tallies, and fission source distributions for time-independent problems, and can run in both criticality or fixed source modes. WARP can transport neutrons in unrestricted arrangements of parallelepipeds, hexagonal prisms, cylinders, and spheres. WARP uses an event-based algorithm, but with some important differences. Moving data is expensive, so WARP uses a remapping vector of pointer/index pairs to direct GPU threads to the data they need to access. The remapping vector is sorted by reaction type after every transport iteration using a high-efficiency parallel radix sort, which serves to keep the

  8. ADVANCED HYDROGEN TRANSPORT MEMBRANES FOR VISION 21 FOSSIL FUEL PLANTS

    Energy Technology Data Exchange (ETDEWEB)

    Shane E. Roark; Anthony F. Sammells; Richard Mackay; Scott R. Morrison; Sara L. Rolfe; U. Balachandran; Richard N. Kleiner; James E. Stephen; Frank E. Anderson; Shandra Ratnasamy; Jon P. Wagner; Clive Brereton

    2004-01-30

    The objective of this project is to develop an environmentally benign, inexpensive, and efficient method for separating hydrogen from gas mixtures produced during industrial processes, such as coal gasification. Currently, this project is focusing on four basic categories of dense membranes: (1) mixed conducting ceramic/ceramic composites, (2) mixed conducting ceramic/metal (cermet) composites, (3) cermets with hydrogen permeable metals, and (4) layered composites with hydrogen permeable alloys. The primary technical challenge in achieving the goals of this project will be to optimize membrane composition to enable practical hydrogen separation rates and chemical stability. Other key aspects of this developing technology include catalysis, ceramic processing methods, and separation unit design operating under high pressure. To achieve these technical goals, Eltron Research Inc. has organized a consortium consisting of CoorsTek, Sued Chemie, Inc. (SCI), Argonne National Laboratory (ANL), and NORAM. Hydrogen permeation rates in excess of 50 mL {center_dot} min{sup -1} {center_dot} cm{sup 2} at {approx}440 C were routinely achieved under less than optimal experimental conditions using a range of membrane compositions. Factors that limit the maximum permeation attainable were determined to be mass transport resistance of H{sub 2} to and from the membrane surface, as well as surface contamination. Mass transport resistance was partially overcome by increasing the feed and sweep gas flow rates to greater than five liters per minute. Under these experimental conditions, H2 permeation rates in excess of 350 mL {center_dot} min{sup -1} {center_dot} cm{sup 2} at {approx}440 C were attained. These results are presented in this report, in addition to progress with cermets, thin film fabrication, catalyst development, and H{sub 2} separation unit scale up.

  9. Visualization of root water uptake: quantification of deuterated water transport in roots using neutron radiography and numerical modeling.

    Science.gov (United States)

    Zarebanadkouki, Mohsen; Kroener, Eva; Kaestner, Anders; Carminati, Andrea

    2014-10-01

    Our understanding of soil and plant water relations is limited by the lack of experimental methods to measure water fluxes in soil and plants. Here, we describe a new method to noninvasively quantify water fluxes in roots. To this end, neutron radiography was used to trace the transport of deuterated water (D2O) into roots. The results showed that (1) the radial transport of D2O from soil to the roots depended similarly on diffusive and convective transport and (2) the axial transport of D2O along the root xylem was largely dominated by convection. To quantify the convective fluxes from the radiographs, we introduced a convection-diffusion model to simulate the D2O transport in roots. The model takes into account different pathways of water across the root tissue, the endodermis as a layer with distinct transport properties, and the axial transport of D2O in the xylem. The diffusion coefficients of the root tissues were inversely estimated by simulating the experiments at night under the assumption that the convective fluxes were negligible. Inverse modeling of the experiment at day gave the profile of water fluxes into the roots. For a 24-d-old lupine (Lupinus albus) grown in a soil with uniform water content, root water uptake was higher in the proximal parts of lateral roots and decreased toward the distal parts. The method allows the quantification of the root properties and the regions of root water uptake along the root systems. PMID:25189533

  10. Visualization of root water uptake: quantification of deuterated water transport in roots using neutron radiography and numerical modeling.

    Science.gov (United States)

    Zarebanadkouki, Mohsen; Kroener, Eva; Kaestner, Anders; Carminati, Andrea

    2014-10-01

    Our understanding of soil and plant water relations is limited by the lack of experimental methods to measure water fluxes in soil and plants. Here, we describe a new method to noninvasively quantify water fluxes in roots. To this end, neutron radiography was used to trace the transport of deuterated water (D2O) into roots. The results showed that (1) the radial transport of D2O from soil to the roots depended similarly on diffusive and convective transport and (2) the axial transport of D2O along the root xylem was largely dominated by convection. To quantify the convective fluxes from the radiographs, we introduced a convection-diffusion model to simulate the D2O transport in roots. The model takes into account different pathways of water across the root tissue, the endodermis as a layer with distinct transport properties, and the axial transport of D2O in the xylem. The diffusion coefficients of the root tissues were inversely estimated by simulating the experiments at night under the assumption that the convective fluxes were negligible. Inverse modeling of the experiment at day gave the profile of water fluxes into the roots. For a 24-d-old lupine (Lupinus albus) grown in a soil with uniform water content, root water uptake was higher in the proximal parts of lateral roots and decreased toward the distal parts. The method allows the quantification of the root properties and the regions of root water uptake along the root systems.

  11. Neutron shielding material

    International Nuclear Information System (INIS)

    From among the neutron shielding materials of the 'kobesh' series developed by Kobe Steel, Ltd. for transport and storage packagings, silicon rubber base type material has been tested for several items with a view to practical application and official authorization, and in order to determine its adaptability to actual vessels. Silicon rubber base type 'kobesh SR-T01' is a material in which, from among the silicone rubber based neutron shielding materials, the hydrogen content is highest and the boron content is most optimized. Its neutron shielding capability has been already described in the previous report (Taniuchi, 1986). The following tests were carried out to determine suitability for practical application; 1) Long-term thermal stability test 2) Pouring test on an actual-scale model 3) Fire test The experimental results showed that the silicone rubber based neutron shielding material has good neutron shielding capability and high long-term fire resistance, and that it can be applied to the advanced transport packaging. (author)

  12. PREFACE: Exploring surfaces and buried interfaces of functional materials by advanced x-ray and neutron techniques Exploring surfaces and buried interfaces of functional materials by advanced x-ray and neutron techniques

    Science.gov (United States)

    Sakurai, Kenji

    2010-12-01

    This special issue is devoted to describing recent applications of x-ray and neutron scattering techniques to the exploration of surfaces and buried interfaces of various functional materials. Unlike many other surface-sensitive methods, these techniques do not require ultra high vacuum, and therefore, a variety of real and complicated surfaces fall within the scope of analysis. It must be particularly emphasized that the techniques are capable of seeing even buried function interfaces as well as the surface. Furthermore, the information, which ranges from the atomic to mesoscopic scale, is highly quantitative and reproducible. The non-destructive nature of the techniques is another important advantage of using x-rays and neutrons, when compared with other atomic-scale analyses. This ensures that the same specimen can be measured by other techniques. Such features are fairly attractive when exploring multilayered materials with nanostructures (dots, tubes, wires, etc), which are finding applications in electronic, magnetic, optical and other devices. The Japan Applied Physics Society has established a group to develop the research field of studying buried function interfaces with x-rays and neutrons. As the methods can be applied to almost all types of materials, from semiconductor and electronic devices to soft materials, participants have fairly different backgrounds but share a common interest in state-of-the-art x-ray and neutron techniques and sophisticated applications. A series of workshops has been organized almost every year since 2001. Some international interactions have been continued intensively, although the community is part of a Japanese society. This special issue does not report the proceedings of the recent workshop, although all the authors are in some way involved in the activities of the above society. Initially, we intended to collect quite long overview papers, including the authors' latest and most important original results, as well as

  13. Radiation transport in earth for neutron and gamma ray point sources above an air-ground interface

    International Nuclear Information System (INIS)

    Two-dimensional discrete ordinates methods were used to calculate the instantaneous dose rate in silicon and neutron and gamma ray fluences as a function of depth in earth from point sources at various heights (1.0, 61.3, and 731.5 meters) above an air--ground interface. The radiation incident on the earth's surface was transported through an earth-only and an earth--concrete model containing 0.9 meters of borated concrete beginning 0.5 meters below the earth's surface to obtain fluence distributions to a depth of 3.0 meters. The inclusion of borated concrete did not significantly reduce the total instantaneous dose rate in silicon and, in all cases, the secondary gamma ray fluence and corresponding dose are substantially larger than the primary neutron fluence and corresponding dose for depths greater than 0.6 meter. 4 figures, 4 tables

  14. Neutron diffraction and electrical transport studies on the incommensurate magnetic phase transition in holmium at high pressures

    Energy Technology Data Exchange (ETDEWEB)

    Thomas, Sarah [University of Alabama, Birmingham; Uhoya, Walter [University of Alabama, Birmingham; Tsoi, Georgiy [University of Alabama, Birmingham; Wenger, Lowell E [University of Alabama, Birmingham; Vohra, Yogesh [University of Alabama, Birmingham; Chesnut, Gary Neal [University of Alabama, Birmingham; Weir, S. T. [Lawrence Livermore National Laboratory (LLNL); Tulk, Christopher A [ORNL; Moreira Dos Santos, Antonio F [ORNL

    2012-01-01

    Neutron diffraction and electrical transport measurements have been made on the heavy rare earth metal holmium at high pressures and low temperatures in order to elucidate its transition from a paramagnetic (PM) to a helical antiferromagnetic (AFM) ordered phase as a function of pressure. The electrical resistance measurements show a change in the resistance slope as the temperature is lowered through the antiferromagnetic Neel temperature. The temperature of this antiferromagnetic transition decreases from approximately 122 K at ambient pressure at a rate of -4.9 K GPa(-1) up to a pressure of 9 GPa, whereupon the PM-to-AFM transition vanishes for higher pressures. Neutron diffraction measurements as a function of pressure at 89 and 110 K confirm the incommensurate nature of the phase transition associated with the antiferromagnetic ordering of the magnetic moments in a helical arrangement and that the ordering occurs at similar pressures as determined from the resistance results for these temperatures.

  15. Direct utilization of information from nuclear data files in Monte Carlo simulation of neutron and photon transport

    International Nuclear Information System (INIS)

    Questions, related to Monte-Carlo method for solution of neutron and photon transport equation, are discussed in the work concerned. Problems dealing with direct utilization of information from evaluated nuclear data files in run-time calculations are considered. ENDF-6 format libraries have been used for calculations. Approaches provided by the rules of ENDF-6 files 2, 3-6, 12-15, 23, 27 and algorithms for reconstruction of resolved and unresolved resonance region cross sections under preset energy are described. The comparison results of calculations made by NJOY and GRUCON programs and computed cross sections data are represented. Test computation data of neutron leakage spectra for spherical benchmark-experiments are also represented. (authors)

  16. First studies of electron transport along small gas gaps of novel foil radiation converters for fast-neutron detectors

    CERN Document Server

    Cortesi, M; Adams, R; Dangendorf, V; Breskin, A; Mayer, S; Hoedlmoser, H; Prasser, H -M

    2012-01-01

    Novel high efficiency fast-neutron detectors were suggested for fan-beam tomography applications. They combine multi-layer polymer converters in gas medium, coupled to thick gaseous electron multipliers (THGEM). In this work we discuss the results of a systematic study of the electron transport inside a narrow gap between successive converter foils, which affects the performance of the detector, both in terms of detection efficiency and localization properties. The efficiency of transporting ionization electrons was measured along a 0.6 mm wide gas gap in 6 and 10 mm wide polymer converters Computer simulations provided conceptual understanding of the observations. For a drift lengths of 6 mm electrons were efficiently transported along the narrow gas gap, with minimal diffusion-induced losses; an average collection efficiency of 95% was achieved for the ionization electrons induced by a primary electron of a few keV initial energy. The 10 mm height converter yielded considerably lower efficiency due to elect...

  17. Recovery Act - Sustainable Transportation: Advanced Electric Drive Vehicle Education Program

    Energy Technology Data Exchange (ETDEWEB)

    Caille, Gary

    2013-12-13

    The collective goals of this effort include: 1) reach all facets of this society with education regarding electric vehicles (EV) and plug–in hybrid electric vehicles (PHEV), 2) prepare a workforce to service these advanced vehicles, 3) create web–based learning at an unparalleled level, 4) educate secondary school students to prepare for their future and 5) train the next generation of professional engineers regarding electric vehicles. The Team provided an integrated approach combining secondary schools, community colleges, four–year colleges and community outreach to provide a consistent message (Figure 1). Colorado State University Ventures (CSUV), as the prime contractor, plays a key program management and co–ordination role. CSUV is an affiliate of Colorado State University (CSU) and is a separate 501(c)(3) company. The Team consists of CSUV acting as the prime contractor subcontracted to Arapahoe Community College (ACC), CSU, Motion Reality Inc. (MRI), Georgia Institute of Technology (Georgia Tech) and Ricardo. Collaborators are Douglas County Educational Foundation/School District and Gooru (www.goorulearning.org), a nonprofit web–based learning resource and Google spin–off.

  18. Recent Improvements to an Advanced Atmospheric Transport Modeling System

    Energy Technology Data Exchange (ETDEWEB)

    Buckley, R. L.; Hunter, C. H.

    2005-10-24

    The Atmospheric Technologies Group (ATG) has developed an advanced atmospheric modeling capability using the Regional Atmospheric Modeling System (RAMS) and a stochastic Lagrangian particle dispersion model (LPDM) for operational use at the Savannah River Site (SRS). For local simulations concerning releases from the Central Savannah River Area (CSRA), RAMS is run in a nested grid configuration with horizontal grid spacing of 8 and 2 km for each grid, with 6-hr forecasts updated every 3 hours. An interface to allow for easy user access to LPDM had been generated, complete with post-processing results depicting surface concentration, deposition, and a variety of dose quantities. A prior weakness in this approach was that observations from the SRS tower network were only incorporated into the three-dimensional modeling effort during the initialization process. Thus, if the forecasted wind fields were in error, the resulting plume predictions would also be erroneous. To overcome this shortcoming, the procedure for generating RAMS wind fields and reading them into LPDM has been modified such that SRS wind measurements are blended with the predicted three-dimensional wind fields from RAMS using the Barnes technique. In particular, the horizontal components in RAMS are replaced with the observed values at a series of 8 towers that exist within the SRS boundary (covering {approx}300 km{sup 2}). Even though LPDM is currently configured to account only for radioactive releases, it was used in a recent chlorine gas release to generate plume concentrations based on unit releases from the site of a train accident in Graniteville, South Carolina. This information was useful to local responders as an indication of potential protective actions downwind of the release.

  19. Advanced High-Temperature, High-Pressure Transport Reactor Gasification

    Energy Technology Data Exchange (ETDEWEB)

    Michael L. Swanson

    2005-08-30

    The transport reactor development unit (TRDU) was modified to accommodate oxygen-blown operation in support of a Vision 21-type energy plex that could produce power, chemicals, and fuel. These modifications consisted of changing the loop seal design from a J-leg to an L-valve configuration, thereby increasing the mixing zone length and residence time. In addition, the standpipe, dipleg, and L-valve diameters were increased to reduce slugging caused by bubble formation in the lightly fluidized sections of the solid return legs. A seal pot was added to the bottom of the dipleg so that the level of solids in the standpipe could be operated independently of the dipleg return leg. A separate coal feed nozzle was added that could inject the coal upward into the outlet of the mixing zone, thereby precluding any chance of the fresh coal feed back-mixing into the oxidizing zone of the mixing zone; however, difficulties with this coal feed configuration led to a switch back to the original downward configuration. Instrumentation to measure and control the flow of oxygen and steam to the burner and mix zone ports was added to allow the TRDU to be operated under full oxygen-blown conditions. In total, ten test campaigns have been conducted under enriched-air or full oxygen-blown conditions. During these tests, 1515 hours of coal feed with 660 hours of air-blown gasification and 720 hours of enriched-air or oxygen-blown coal gasification were completed under this particular contract. During these tests, approximately 366 hours of operation with Wyodak, 123 hours with Navajo sub-bituminous coal, 143 hours with Illinois No. 6, 106 hours with SUFCo, 110 hours with Prater Creek, 48 hours with Calumet, and 134 hours with a Pittsburgh No. 8 bituminous coal were completed. In addition, 331 hours of operation on low-rank coals such as North Dakota lignite, Australian brown coal, and a 90:10 wt% mixture of lignite and wood waste were completed. Also included in these test campaigns was

  20. Neutrons, deuteration and synchrotron X-rays for the study of biology and advanced materials: A match made in atoms..

    International Nuclear Information System (INIS)

    Together, the Australian Synchrotron in Melbourne and the OPAL research reactor, at the Bragg Institute in Sydney represent Australia's largest ever investment in scientific infrastructure. Both facilities commenced operation in 2007, have passed through their infancy and adolescence to take their place amongst the rank of top-flight international user facilities. Far from middle-aged, these two vibrant landmark facilities (each with 10 operational beamlines) and along with the National Deuteration Facility at ANSTO have provided transformational research capabilities for the Australian scientific community. Although modest in size compared to the well-established international competition, both institutions are producing excellent amounts of high-quality research with the Bragg Institute and the Australian Synchrotron generating more than 200 and 450 peer-reviewed publications per annum respectively. At first glance both synchrotron and neutron sources show similar scientific profiles, encompassing an extremely wide range of disciplines: materials, chemistry, biology, condensed matter physics, nanotechnology, engineering, geosciences, archaeology and studies relating to cultural heritage. Common to both are advanced capabilities for the study of atomic and molecular structure, as well as operational studies of functional materials under a diverse range of extreme environments. A more forensic examination however reveals fundamental differences in their DNA. While the biological, pharmaceutical and medical research communities drive substantial capability development and research outcomes at the Australian Synchrotron, neutron scattering and molecular deuteration at the Bragg Institute provides a focus for studies in soft condensed matter, physical and inorganic chemistry, solid state physics and crystallography. Although their respective probes are generated from different parts of the atom and interact with matter in fundamentally different ways, my

  1. Advanced in-situ measurement of soil carbon content using inelastic neutron scattering

    Science.gov (United States)

    Measurement and mapping of natural and anthropogenic variations in soil carbon stores is a critical component of any soil resource evaluation process. Emerging modalities for soil carbon analysis in the field is the registration of gamma rays from soil under neutron irradiation. The inelastic neutro...

  2. Design and layout decision for refueling system of advanced fast neutron reactors

    International Nuclear Information System (INIS)

    Describes fast neutron reactor refueling features, BN-1200 power unit general data, its refueling system design concepts, individual refueling equipment purpose and designs, and required experimental studies to create it. Refueling equipment characteristics for BN-800 and BN-1200 reactors are compared. (author)

  3. The coupling of the neutron transport application RATTLESNAKE to the nuclear fuels performance application BISON under the MOOSE framework

    International Nuclear Information System (INIS)

    The MOOSE based reactor physics tool MAMMOTH provides the capability to seamlessly couple the neutron transport application RATTLESNAKE to the fuels performance application BISON to produce a higher fidelity tool for fuel performance simulations. The ultimate purpose of this coupling is to provide a tool with the predictive capabilities to gain new knowledge and help resolve fundamental questions in the fuel performance arena, i.e. high-burnup structures, pellet-cladding interaction, missing pellet surface, etc. RATTLESNAKE solves the self-adjoint angular flux transport equation, derived from the linearized Boltzmann transport equation, and provides a sub-pin level resolution of the multigroup neutron flux. BISON solves the coupled thermomechanical equations for the fuel on a sub-millimeter scale. The coupling within the MOOSE framework allows both applications to solve their respective systems on aligned and unaligned unstructured finite element meshes. MAMMOTH uses the power density calculated by RATTLESNAKE to compute the local burnup evolution. Subsequently, MAMMOTH transfers the power density and burnup distribution to BISON with the MOOSE Multiapp transfer system. BISON in turn is able to provide sub-pin level temperature for cross section feed back effects. Multiple depletion cases were run with one-way and two-way data transfer in MAMMOTH for RATTLESNAKE-BISON. The one-way eigenvalues obtained show good agreement with the reference values obtained from the lattice physics code DRAGON4 while the two-way eigenvalue show expected differences. The power distributions obtained are consistent with both DRAGON4 and the SERPENT Monte Carlo code. The one-way and two-way calculations produce power density results that are comparable with those of the internal, static, Lassmannstyle model in BISON. Differences in the power densities arise from the use of better neutron energy deposition parameters obtained from the DRAGON4 tabulations, and differences in the fuel

  4. Neutron cross-sections for advanced nuclear systems. The nTOF project at CERN

    International Nuclear Information System (INIS)

    In 2012, nuclear energy continued to play an important role in global electricity production. Despite a small reduction of the total generating nuclear power capacity after the accident at the Fukushima Daiichi nuclear power plant, a significant growth, between 35% and 100% by 2030, is foreseen in the use of nuclear energy worldwide. The knowledge of a wide variety of nuclear processes is a fundamental prerequisite in nuclear technology, as well as in other field of fundamental and applied Nuclear Physics. In particular, neutron-induced reactions play a key role in the operation of present nuclear reactors as well as in the design of future ones aiming at minimizing nuclear waste, such as Generation-IV reactors, ADS or reactors based on Th/U fuel cycle. The cross sections of a large number of neutron-induced reactions are requested with high accuracy to improve safety and efficiency of current reactors, and for the design of future generation systems. Since 2001 nTOF, an innovative neutron Time-Of-Flight facility, has been operating at CERN with the aim of addressing the needs of nuclear data for basic and applied nuclear Physics. An extensive program on both neutron induced fission and capture reactions has been carried out so far. Thanks to the well suited features of the nTOF neutron beam, such as the high instantaneous neutron flux, the high resolution and the wide energy range covered, from thermal to a few GeV, coupled with state-of-the-art detectors and data acquisition system, it has been possible to collect high accuracy and high resolution neutron cross-section data on a variety of isotopes, many of which radioactive. In particular, important results for nuclear technologies have been obtained on isotopes of U, Pu and minor actinides with long half life. Recently the construction of a new, high-flux measuring station has started. A 25 times higher fluence relative to the existing experimental area will allow to measure isotopes with short half life, as

  5. Recent advances in the study of H environments and behavior in minerals using neutron powder diffraction

    Science.gov (United States)

    Welch, M. D.

    2002-12-01

    It is now possible to probe the structural environments and behavior of H atoms directly in complex minerals such as amphiboles, micas, chlorites and humites using neutron powder diffraction, in some cases as a function of pressure and/or temperature. A combination of high neutron flux and increased detector sensitivity and size offers the chance to see details of H behaviour. In the last year or so the advent of new gasket designs for the Paris-Edinburgh pressure cell allow the use of ethanol/methanol (EtOD/MeOD) as a pressure medium, removing peak broadening arising from deviatoric stress that occurs above 3 GPa for the standard fluorinert pressure medium. Essentially hydrostatic conditions obtain with EtOD/MeOD to 8 GPa at 298 K. A further recent development has been the design of a high P-T module for use with the Paris-Edinburgh cell. These technological improvements in pressure-cell design now allow us to make meaningful correlations between OH vibrational spectra collected at high P and/or T and detailed structural information on H behaviour obtained from neutron diffraction under similar conditions. In this talk I shall discuss recent neutron diffraction experiments on the effect of pressure upon hydrogen bonding in deuterated chlorite to 5 GPa (298 K), and a high P-T study of hydrogen bonding in deuterated brucite to 7 GPa, 1100 K. These two studies illustrate how far high-pressure neutron diffraction has come in the last 5 years. Finally, I shall describe a neutron powder diffraction study (ambient conditions) of leucophoenicite, Mn7Si3O12(OH)2, a close structural analogue of Phase-B and Superhydrous-B: the structure of leucophoenicite is topologically identical to the hydrous sheet of Phase-B and similar to that of Superhydrous-B. For various reasons it was not possible to deuterate the sample. Nonetheless, the two distinct H atoms were approximately located in difference-Fourier maps and then refined isotropically. The H positions in Phase-B were only

  6. FEM-RZ, 2-D Multigroup Neutron Transport in R-Z Geometry, Eigenvalue and Fixed Source Problems

    International Nuclear Information System (INIS)

    1 - Nature of the physical problem solved: FEM-RZ is a computer program for solving multi-group neutron transport problems in two-dimensional cylindrical (r,z) geometry. It can solve not only eigenvalue problems but also other problems, such as fixed source problems. 2 - Method of solution: The method of higher order finite elements is used for the spatial variables. It is based on the discontinuous method with Galerkin-type scheme. The discrete ordinate Sn method is used for the angular variables. 3 - Restrictions on the complexity of the problem: No restrictions except for computer size

  7. Some Spectral Properties of the Transport Equation and Their Relevance to the Theory of Pulsed Neutron Experiments

    International Nuclear Information System (INIS)

    The pulsed neutron experiments rest, as is well known, on the assumption that the neutron transport equation, in a finite bare body, possesses at least one discrete eigenvalue. Nelkin and then Corngold pointed out that this assumption could no longer be true if the sample is very small. Their arguments, however, were based on certain simplifying assumptions about the scattering kernel or the spatial dependence of the neutron density, which could, in principle, greatly restrict the validity of their results when dealing with realistic problems. We have thus preferred to attack the problem from a quite general point of view, which resembles, in many aspects, the approach of Lehner and Wing’s paper, ''On the spectrum of an asymmetric operator arising in the transport theory of neutron'', relating to the one-velocity theory. We have considered the integro-differential transport equation in a finite homogeneous convex body of an arbitrary shape, surrounded by vacuum. The free gas-scattering kernel (averaged over the angles, so that the scattering is isotropic) has been adopted, and then the absorption has been assumed to follow the ''1/v'' law. Let then lim vΣs(v) = h0, where h0 results in a positive constant. v->0 The eigenvalue spectrum of the transport equation can be shown to have the following structure: the half-plane Re λ ≤ -h0 is occupied by a densely distributed spectrum (let us say ''continuous spectrum''), while the remaining half-plane Re λ > - h0 contains at most a finite number of real eigenvalues λi. It has been shown that the number of the discrete eigenvalues can reduce to zero if the body is small enough (of the order of a mean free path). Thus we see that Nelkin’s idea on the disappearing of the decay modes for samples of very small size is correct also within our general assumptions. As far as we know the experimental results do not seem to provide an unambiguous answer pro or contra this theoretical assertion. We are now developing a

  8. 脉冲中子-裂变中子探测铀黄饼的MCNP模拟%The Monte Carlo N particle transport code simulation of pulsed neutron-fission neutron uranium yellowcake exploration

    Institute of Scientific and Technical Information of China (English)

    张坤明; 张雄杰; 瞿金辉; 汤彬

    2015-01-01

    利用MCNP程序模拟研究脉冲中子-裂变中子探测铀黄饼,采用脉冲式中子源,利用氦三管中子探测器记录裂变中子,得到铀黄饼中的铀含量信息。通过对14 MeV脉冲中子源和产生的裂变中子在不同铀含量模型中的输运计算,分析了裂变中子与铀含量的关系。结果表明:利用裂变超热中子衰减时间谱,可以确定铀黄饼中的铀含量;通过对热中子衰减时间谱进行校正,可以提高铀黄饼中铀含量计算结果的准确度。%The Monte Carlo N particle transport code ( MCNP ) is used to simulate how to explore the uranium yel⁃lowcake by using the pulsed neutron⁃fission neutron ( PNFN) method. In order to obtain uranium yellowcake quan⁃titation, pulsed neutron source was used, prompt fission neutrons were detected by using the neutron detector. Un⁃der the condition of different uranium quantitation models, the transport of the 14 MeV pulsed neutron source and the released fission neutron were calculated. On the basis of these, the relationship between fission neutron and ura⁃nium quantitation was studied. The results show that using the epithermal neutron time decay spectrum, the urani⁃um yellowcake quantitation can be determined; the precision of the uranium yellowcake quantitation could be in⁃creased by the correction of thermal neutron time decay spectrum.

  9. 75 FR 38151 - Governors' Designees Receiving Advance Notification of Transportation of Nuclear Waste

    Science.gov (United States)

    2010-07-01

    ... From the Federal Register Online via the Government Publishing Office NUCLEAR REGULATORY COMMISSION Governors' Designees Receiving Advance Notification of Transportation of Nuclear Waste On January 6, 1982 (47 FR 596 and 47 FR 600), the U.S. Nuclear Regulatory Commission (NRC) published in the Federal Register final amendments to Title 10 of...

  10. CityMobil: advanced road transport for the urban environment. Final results

    NARCIS (Netherlands)

    Dijke, J.P. van

    2011-01-01

    CityMobil is an Integrated Project in the 6th Framework Programme of the European Union. The project addresses the topic “Advanced Road transport for the Urban Environment.” The project started in May 2006 and will run until December 2011. The project is carried out by a group of 29 partners led by

  11. Development and Implementation of Photonuclear Cross-Section Data for Mutually Coupled Neutron-Photon Transport Calculations in the Monte Carlo N-Particle (MCNP) Radiation Transport Code

    Energy Technology Data Exchange (ETDEWEB)

    Morgan C. White

    2000-07-01

    The fundamental motivation for the research presented in this dissertation was the need to development a more accurate prediction method for characterization of mixed radiation fields around medical electron accelerators (MEAs). Specifically, a model is developed for simulation of neutron and other particle production from photonuclear reactions and incorporated in the Monte Carlo N-Particle (MCNP) radiation transport code. This extension of the capability within the MCNP code provides for the more accurate assessment of the mixed radiation fields. The Nuclear Theory and Applications group of the Los Alamos National Laboratory has recently provided first-of-a-kind evaluated photonuclear data for a select group of isotopes. These data provide the reaction probabilities as functions of incident photon energy with angular and energy distribution information for all reaction products. The availability of these data is the cornerstone of the new methodology for state-of-the-art mutually coupled photon-neutron transport simulations. The dissertation includes details of the model development and implementation necessary to use the new photonuclear data within MCNP simulations. A new data format has been developed to include tabular photonuclear data. Data are processed from the Evaluated Nuclear Data Format (ENDF) to the new class ''u'' A Compact ENDF (ACE) format using a standalone processing code. MCNP modifications have been completed to enable Monte Carlo sampling of photonuclear reactions. Note that both neutron and gamma production are included in the present model. The new capability has been subjected to extensive verification and validation (V&V) testing. Verification testing has established the expected basic functionality. Two validation projects were undertaken. First, comparisons were made to benchmark data from literature. These calculations demonstrate the accuracy of the new data and transport routines to better than 25 percent. Second

  12. Development and Implementation of Photonuclear Cross-Section Data for Mutually Coupled Neutron-Photon Transport Calculations in the Monte Carlo N-Particle (MCNP) Radiation Transport Code

    International Nuclear Information System (INIS)

    The fundamental motivation for the research presented in this dissertation was the need to development a more accurate prediction method for characterization of mixed radiation fields around medical electron accelerators (MEAs). Specifically, a model is developed for simulation of neutron and other particle production from photonuclear reactions and incorporated in the Monte Carlo N-Particle (MCNP) radiation transport code. This extension of the capability within the MCNP code provides for the more accurate assessment of the mixed radiation fields. The Nuclear Theory and Applications group of the Los Alamos National Laboratory has recently provided first-of-a-kind evaluated photonuclear data for a select group of isotopes. These data provide the reaction probabilities as functions of incident photon energy with angular and energy distribution information for all reaction products. The availability of these data is the cornerstone of the new methodology for state-of-the-art mutually coupled photon-neutron transport simulations. The dissertation includes details of the model development and implementation necessary to use the new photonuclear data within MCNP simulations. A new data format has been developed to include tabular photonuclear data. Data are processed from the Evaluated Nuclear Data Format (ENDF) to the new class ''u'' A Compact ENDF (ACE) format using a standalone processing code. MCNP modifications have been completed to enable Monte Carlo sampling of photonuclear reactions. Note that both neutron and gamma production are included in the present model. The new capability has been subjected to extensive verification and validation (V and V) testing. Verification testing has established the expected basic functionality. Two validation projects were undertaken. First, comparisons were made to benchmark data from literature. These calculations demonstrate the accuracy of the new data and transport routines to better than 25 percent. Second, the ability to

  13. Transport analysis of measured neutron leakage spectra from spheres as tests of evaluated high energy cross sections

    Science.gov (United States)

    Bogart, D. D.; Shook, D. F.; Fieno, D.

    1973-01-01

    Integral tests of evaluated ENDF/B high-energy cross sections have been made by comparing measured and calculated neutron leakage flux spectra from spheres of various materials. An Am-Be (alpha,n) source was used to provide fast neutrons at the center of the test spheres of Be, CH2, Pb, Nb, Mo, Ta, and W. The absolute leakage flux spectra were measured in the energy range 0.5 to 12 MeV using a calibrated NE213 liquid scintillator neutron spectrometer. Absolute calculations of the spectra were made using version 3 ENDF/B cross sections and an S sub n discrete ordinates multigroup transport code. Generally excellent agreement was obtained for Be, CH2, Pb, and Mo, and good agreement was observed for Nb although discrepancies were observed for some energy ranges. Poor comparative results, obtained for Ta and W, are attributed to unsatisfactory nonelastic cross sections. The experimental sphere leakage flux spectra are tabulated and serve as possible benchmarks for these elements against which reevaluated cross sections may be tested.

  14. Heterogeneity of solid neutron-star matter: transport coefficients and neutrino emissivity

    CERN Document Server

    Jones, P B

    2004-01-01

    Calculations of weak-interaction transition rates and of nuclear formation enthalpies show that in isolated neutron stars, the solid phase, above the neutron-drip threshold, is amorphous and heterogeneous in nuclear charge. The neutrino emissivities obtained are very dependent on the effects of proton shell structure but may be several orders of magnitude larger than the electron bremsstrahlung neutrino-pair emissivity at temperatures of 10^9 K. In this phase, electrical and thermal conductivities are much smaller than for a homogeneous bcc lattice. In particular, the reduced electrical conductivity, which is also temperature-independent, must have significant consequences for the evolution of high-multipole magnetic fields in neutron stars.

  15. Neutron Albedo

    CERN Document Server

    Ignatovich, V K

    2005-01-01

    A new, algebraic, method is applied to calculation of neutron albedo from substance to check the claim that use of ultradispersive fuel and moderator of an active core can help to gain in size and mass of the reactor. In a model of isotropic distribution of incident and reflected neutrons it is shown that coherent scattering on separate grains in the case of thermal neutrons increases transport cross section negligibly, however it decreases albedo from a wall of finite thickness because of decrease of substance density. A visible increase of albedo takes place only for neutrons with wave length of the order of the size of a single grain.

  16. Evaluation of the response of a neutron detector of the Long-Counter type using a Monte Carlo transport simulation

    Energy Technology Data Exchange (ETDEWEB)

    Pazianotto, Mauricio Tizziani; Goncalez, Odair Lelis; Federico, Claudio Antonio [Centro Tecnico Aeroespacial (IEAv/CTA), Sao Jose dos Campos, SP (Brazil). Inst. de Estudos Avancados; Carlson, Brett Vern [Centro Tecnico Aeroespacial (ITA/CTA), Sao Jose dos Campos, SP (Brazil). Inst. Tecnologico de Aeronautica

    2010-07-01

    Full text: The Institute for Advanced Studies (IEAv) is developing activities to study the dose levels of ionizing radiation from cosmic rays (CR) received by aircraft crews, sensitive equipment (on-board computers, for example) and embedded electronics in Brazilian airspace. Neutrons generated by the interaction of CR with the atmosphere are the dominant particles in the dose accumulation in electronic circuits and aircraft crews at flight altitude. Their production has a very broad energy spectrum, ranging from thermal neutrons (0.025eV ) to neutrons of several hundreds of MeV , making their detection a very difficult process. To observe the temporal variation in flow during the measurements, a detector of the Long Counter (LC) type is being used. This detector is designed to measure the one-way flow of neutrons with constant response over a wide energy range (thermal to 20 MeV ). However, to measure cosmic rays, the flow of which is non-directional, the dependence of the response on the angle of incidence, as well as energy, should be properly investigated. The objective of this study is to assess the angular response of the neutron detector (Long Counter) using the code MCNP5 (Monte Carlo N-Particle) and to compare it with the experimental data previously obtained with a {sup 241}Am-Be source at a distance of 1.66 m from the geometric center of the detector, varying the angle of incidence from 00 to 3600 in intervals of 150. The simulation was performed by modeling in detail the structure and materials of the LC, as well as the experimental arrangement for irradiation. The results of the simulation present reasonable agreement with the experimental data. This agreement shows that the modeling of the geometry of the source-detector system is adequate. The next step is to develop a model of neutron detection for the higher energy present in cosmic radiation fields, for which the experimental calibration is not so easily achievable. (author)

  17. The effect of biological shielding on fast neutron and photon transport in the VVER-1000 mock-up model placed in the LR-0 reactor.

    Science.gov (United States)

    Košťál, Michal; Cvachovec, František; Milčák, Ján; Mravec, Filip

    2013-05-01

    The paper is intended to show the effect of a biological shielding simulator on fast neutron and photon transport in its vicinity. The fast neutron and photon fluxes were measured by means of scintillation spectroscopy using a 45×45 mm(2) and a 10×10 mm(2) cylindrical stilbene detector. The neutron spectrum was measured in the range of 0.6-10 MeV and the photon spectrum in 0.2-9 MeV. The results of the experiment are compared with calculations. The calculations were performed with various nuclear data libraries.

  18. Neutronics analyses for fast spectrum nuclear systems and scenario studies for advanced nuclear fuel cycles

    OpenAIRE

    Grasso, Giacomo

    2010-01-01

    The present PhD thesis summarizes the three-years study about the neutronic investigation of a new concept nuclear reactor aiming at the optimization and the sustainable management of nuclear fuel in a possible European scenario. A new generation nuclear reactor for the nuclear reinassance is indeed desired by the actual industrialized world, both for the solution of the energetic question arising from the continuously growing energy demand together with the corresponding reduction of oil ava...

  19. A least-squares finite-element S{sub n} method for solving first-order neutron transport equation

    Energy Technology Data Exchange (ETDEWEB)

    Ju Haitao [School of Energy and Power Engineering, Xi' an Jiaotong University, Xi' an 710049 (China)]. E-mail: jht0@hotmail.com; Wu Hongchun [School of Energy and Power Engineering, Xi' an Jiaotong University, Xi' an 710049 (China); Zhou Yongqiang [School of Energy and Power Engineering, Xi' an Jiaotong University, Xi' an 710049 (China); Cao Liangzhi [School of Energy and Power Engineering, Xi' an Jiaotong University, Xi' an 710049 (China); Yao Dong [Nuclear Power Institute of China, Chengdu 610041 (China); Xian, Chun-Yu [Nuclear Power Institute of China, Chengdu 610041 (China)

    2007-04-15

    A discrete ordinates finite-element method for solving the two-dimensional first-order neutron transport equation is derived using the least-squares variation. It avoids the singularity in void regions of the method derived from the second-order equation which contains the inversion of the cross-section. Different from using the standard Galerkin variation to the first-order equation, the least-squares variation results in a symmetric matrix, which can be solved easily and effectively. To eliminate the discontinuity of the angular flux on the vacuum boundary in the spherical harmonics method, the angle variable is discretized by the discrete ordinates method. A two-dimensional transport simulation code is developed and applied to some benchmark problems with unstructured geometry. The numerical results verified the validity of this method.

  20. Thermal Scattering Law Data: Implementation and Testing Using the Monte Carlo Neutron Transport Codes COG, MCNP and TART

    Energy Technology Data Exchange (ETDEWEB)

    Cullen, D E; Hansen, L F; Lent, E M; Plechaty, E F

    2003-05-17

    Recently we implemented the ENDF/B-VI thermal scattering law data in our neutron transport codes COG and TART. Our objective was to convert the existing ENDF/B data into double differential form in the Livermore ENDL format. This will allow us to use the ENDF/B data in any neutron transport code, be it a Monte Carlo, or deterministic code. This was approached as a multi-step project. The first step was to develop methods to directly use the thermal scattering law data in our Monte Carlo codes. The next step was to convert the data to double-differential form. The last step was to verify that the results obtained using the data directly are essentially the same as the results obtained using the double differential data. Part of the planned verification was intended to insure that the data as finally implemented in the COG and TART codes, gave the same answer as the well known MCNP code, which includes thermal scattering law data. Limitations in the treatment of thermal scattering law data in MCNP have been uncovered that prevented us from performing this part of our verification.

  1. The coupling of the neutron transport application RATTLESNAKE to the nuclear fuels performance application BISON under the MOOSE framework

    Energy Technology Data Exchange (ETDEWEB)

    Gleicher, Frederick N.; Williamson, Richard L.; Ortensi, Javier; Wang, Yaqi; Spencer, Benjamin W.; Novascone, Stephen R.; Hales, Jason D.; Martineau, Richard C.

    2014-10-01

    The MOOSE neutron transport application RATTLESNAKE was coupled to the fuels performance application BISON to provide a higher fidelity tool for fuel performance simulation. This project is motivated by the desire to couple a high fidelity core analysis program (based on the self-adjoint angular flux equations) to a high fidelity fuel performance program, both of which can simulate on unstructured meshes. RATTLESNAKE solves self-adjoint angular flux transport equation and provides a sub-pin level resolution of the multigroup neutron flux with resonance treatment during burnup or a fast transient. BISON solves the coupled thermomechanical equations for the fuel on a sub-millimeter scale. Both applications are able to solve their respective systems on aligned and unaligned unstructured finite element meshes. The power density and local burnup was transferred from RATTLESNAKE to BISON with the MOOSE Multiapp transfer system. Multiple depletion cases were run with one-way data transfer from RATTLESNAKE to BISON. The eigenvalues are shown to agree well with values obtained from the lattice physics code DRAGON. The one-way data transfer of power density is shown to agree with the power density obtained from an internal Lassman-style model in BISON.

  2. Application of the efficient consistent spatial homogenization method in neutron transport theory to a gas cooled thermal reactor problem

    International Nuclear Information System (INIS)

    In this paper, the accuracy and computational efficiency of the efficient consistent spatial homogenization method (ECSH) in neutron transport theory is assessed in a 1D benchmark problem characteristic of gas cooled thermal systems that are extremely challenging for conventional homogenization methods because of their longer neutron mean free path than water-based thermal reactors. The ECSH method is an extension of the consistent spatial homogenization method by using: (1) B-spline instead of Fourier series for the expansion of the spatial domain in the auxiliary cross section term and (2) a source iteration scheme instead of local fixed-source calculations in the re-homogenization procedure. Furthermore, the effect of the angular expansion order in the definition of the auxiliary cross section is studied. This method can be viewed as a significant improvement in accuracy of standard homogenization methods used for VHTR whole core analysis in which core environment effects are pronounced. It is shown that the ECSH method can reproduce the heterogeneous transport solution with up to 4 times faster computational speed, depending on the configuration of the control rods while maintaining reasonable accuracy and robust re-homogenization procedure. (author)

  3. TRIPOLI: a general Monte Carlo code, present state and future prospects. [Neutron and gamma ray transport

    Energy Technology Data Exchange (ETDEWEB)

    Nimal, J.C.; Vergnaud, T. (CEA Centre d' Etudes Nucleaires de Saclay, 91 - Gif-sur-Yvette (France))

    1990-01-01

    This paper describes the most important features of the Monte Carlo code TRIPOLI-2. This code solves the Boltzmann equation in three-dimensional geometries for coupled neutron and gamma rays problems. A particular emphasis is devoted to the biasing techniques, which are very important for deep penetration. Future developments in TRIPOLI are described in the conclusion. (author).

  4. Advances on the study of air pollution in Cordoba by neutron activation analysis

    International Nuclear Information System (INIS)

    Air pollution biomonitoring has been carried out in an area of 160.000 km2 by neutron activation analysis of lichen samples (Usnea sp. and Ramalina ecklonii) in the framework of a Co-ordinated Research Programme of the IAEA and an ARCAL Technical Co-operation Project. The samples were irradiated in the RA-3 reactor and after a decay time of 6, 12 and 30 days, 24 elements (As, Ba, Br, Ce, Co, Cr, Cs, Eu, Fe, Hf, La, Lu, Na, Nd, Rb, Sb, Sc, Sm, Ta, Tb, Th, U and Zn) were determined by gamma spectrometry. (author)

  5. Advanced liquid and solid extraction procedures for ultratrace determination of rhenium by radiochemical neutron activation analysis

    Science.gov (United States)

    Mizera, J.; Kučera, J.; Řanda, Z.; Lučaníková, M.

    2006-01-01

    Radiochemical neutron activation analysis (RNAA) procedures for determination of Re at the ultratrace level based on use of liquid-liquid extraction (LLE) and extraction chromatography (EXC) have been developed. Two different LLE procedures were used depending on the way of sample decomposition using either 2-butanone or tetraphenylarsonium chloride in CHCl3. EXC employed new solid extractant materials prepared by incorporation of the liquid trioctyl-methyl-ammonium chloride into an inert polyacrylonitrile matrix. The RNAA procedures presented have been compared and applied for Re determination in several biological and environmental reference materials.

  6. Analysis of the neutrons dispersion in a semi-infinite medium based in transport theory and the Monte Carlo method; Analisis de la dispersion de neutrones en un medio semi-infinito en base a teoria de transporte y el metodo de Monte Carlo

    Energy Technology Data Exchange (ETDEWEB)

    Arreola V, G. [IPN, Escuela Superior de Fisica y Matematicas, Posgrado en Ciencias Fisicomatematicas, area en Ingenieria Nuclear, Unidad Profesional Adolfo Lopez Mateos, Edificio 9, Col. San Pedro Zacatenco, 07730 Mexico D. F. (Mexico); Vazquez R, R.; Guzman A, J. R., E-mail: energia.arreola.uam@gmail.com [Universidad Autonoma Metropolitana, Unidad Iztapalapa, Area de Ingenieria en Recursos Energeticos, Av. San Rafael Atlixco 186, Col. Vicentina, 09340 Mexico D. F. (Mexico)

    2012-10-15

    In this work a comparative analysis of the results for the neutrons dispersion in a not multiplicative semi-infinite medium is presented. One of the frontiers of this medium is located in the origin of coordinates, where a neutrons source in beam form, i.e., {mu}{omicron}=1 is also. The neutrons dispersion is studied on the statistical method of Monte Carlo and through the unidimensional transport theory and for an energy group. The application of transport theory gives a semi-analytic solution for this problem while the statistical solution for the flow was obtained applying the MCNPX code. The dispersion in light water and heavy water was studied. A first remarkable result is that both methods locate the maximum of the neutrons distribution to less than two mean free trajectories of transport for heavy water, while for the light water is less than ten mean free trajectories of transport; the differences between both methods is major for the light water case. A second remarkable result is that the tendency of both distributions is similar in small mean free trajectories, while in big mean free trajectories the transport theory spreads to an asymptote value and the solution in base statistical method spreads to zero. The existence of a neutron current of low energy and toward the source is demonstrated, in contrary sense to the neutron current of high energy coming from the own source. (Author)

  7. Parallel computing for homogeneous diffusion and transport equations in neutronics; Calcul parallele pour les equations de diffusion et de transport homogenes en neutronique

    Energy Technology Data Exchange (ETDEWEB)

    Pinchedez, K

    1999-06-01

    Parallel computing meets the ever-increasing requirements for neutronic computer code speed and accuracy. In this work, two different approaches have been considered. We first parallelized the sequential algorithm used by the neutronics code CRONOS developed at the French Atomic Energy Commission. The algorithm computes the dominant eigenvalue associated with PN simplified transport equations by a mixed finite element method. Several parallel algorithms have been developed on distributed memory machines. The performances of the parallel algorithms have been studied experimentally by implementation on a T3D Cray and theoretically by complexity models. A comparison of various parallel algorithms has confirmed the chosen implementations. We next applied a domain sub-division technique to the two-group diffusion Eigen problem. In the modal synthesis-based method, the global spectrum is determined from the partial spectra associated with sub-domains. Then the Eigen problem is expanded on a family composed, on the one hand, from eigenfunctions associated with the sub-domains and, on the other hand, from functions corresponding to the contribution from the interface between the sub-domains. For a 2-D homogeneous core, this modal method has been validated and its accuracy has been measured. (author)

  8. Advanced Public Transportation Sytems; A Taxonomy, Commercial Availability And Deployment, Phase II

    OpenAIRE

    Khattak, Asad; Et. al.,

    1997-01-01

    This study explores the development and availability of Advanced Public Transportation Systems (APTS) technologies. The study refines a taxonomy of transit technologies and uses it to explore the availability of new technologies and their impacts in transit agencies. THe taxonomy is based on defining the features, functions and performance characteristics of transit technologies. Based on the taxonomy, three surveys of technology suppliers were conducted. Questions were related to technology ...

  9. System safety engineering in the development of advanced surface transportation vehicles

    Science.gov (United States)

    Arnzen, H. E.

    1971-01-01

    Applications of system safety engineering to the development of advanced surface transportation vehicles are described. As a pertinent example, the paper describes a safety engineering efforts tailored to the particular design and test requirements of the Tracked Air Cushion Research Vehicle (TACRV). The test results obtained from this unique research vehicle provide significant design data directly applicable to the development of future tracked air cushion vehicles that will carry passengers in comfort and safety at speeds up to 300 miles per hour.

  10. Fuel-cycle greenhouse gas emissions impacts of alternative transportation fuels and advanced vehicle technologies.

    Energy Technology Data Exchange (ETDEWEB)

    Wang, M. Q.

    1998-12-16

    At an international conference on global warming, held in Kyoto, Japan, in December 1997, the United States committed to reduce its greenhouse gas (GHG) emissions by 7% over its 1990 level by the year 2012. To help achieve that goal, transportation GHG emissions need to be reduced. Using Argonne's fuel-cycle model, I estimated GHG emissions reduction potentials of various near- and long-term transportation technologies. The estimated per-mile GHG emissions results show that alternative transportation fuels and advanced vehicle technologies can help significantly reduce transportation GHG emissions. Of the near-term technologies evaluated in this study, electric vehicles; hybrid electric vehicles; compression-ignition, direct-injection vehicles; and E85 flexible fuel vehicles can reduce fuel-cycle GHG emissions by more than 25%, on the fuel-cycle basis. Electric vehicles powered by electricity generated primarily from nuclear and renewable sources can reduce GHG emissions by 80%. Other alternative fuels, such as compressed natural gas and liquefied petroleum gas, offer limited, but positive, GHG emission reduction benefits. Among the long-term technologies evaluated in this study, conventional spark ignition and compression ignition engines powered by alternative fuels and gasoline- and diesel-powered advanced vehicles can reduce GHG emissions by 10% to 30%. Ethanol dedicated vehicles, electric vehicles, hybrid electric vehicles, and fuel-cell vehicles can reduce GHG emissions by over 40%. Spark ignition engines and fuel-cell vehicles powered by cellulosic ethanol and solar hydrogen (for fuel-cell vehicles only) can reduce GHG emissions by over 80%. In conclusion, both near- and long-term alternative fuels and advanced transportation technologies can play a role in reducing the United States GHG emissions.

  11. Diurnally modulating neutron flux in the Moon's high-latitudes: Evidence for transported hydrogen volatiles and/ or complex regolith compositions in topographic slopes

    Science.gov (United States)

    McClanahan, Timothy; Mirofanov, Igor; Boynton, William; Chin, Gordon; Livengood, Timothy; Su, Jiao Jang; Sagdeev, Raold; Parsons, Ann; Evans, Larry; Starr, Richard; Hamara, Dave; Bodnarik, Julia; Williams, Jeane-Pierre; Mazarico, Erwan; Litvak, Maxim; Sanin, Anton; Murray, Joseph

    2016-04-01

    We report evidence that the Moon's diurnally modulating neutron flux is being forced by a latitude dependent mix of 1) transient hydrogen-bearing volatiles near the surface in the upper latitudes and 2) regolith temperature variation in lower latitudes. In this study we investigate diurnally varying neutron flux measurements from the Lunar Reconnaissance Orbiter's (LRO) Lunar Exploration Neutron Detector's Collimated Sensor for Epithermal Neutrons (LEND CSETN) and surface temperature observations from the Diviner radiometer poleward of >±45°. Our presentation shows that the modulating neutron flux is not consistent with a regolith temperature control for latitudes >70°. The anticorrelation may be evidence for transported lunar hydrogen volatiles or highly non-uniform regolith compositional dynamics. Observational evidence is consistent with regolith temperature being the source of the neutron flux modulation in the northern mare (45° to 60°) and may be related to its mafic composition and fast neutron contributions. Predictions for hypothesized regolith temperature effects are evaluated using insolation inferred from the Lunar Observing Laser Altimeter (LOLA) topography.

  12. Determination of neutron flux distribution by using ANISN, a one-dimensional discrete S sub n ordinates transport code with anisotropic scattering

    Science.gov (United States)

    Ghorai, S. K.

    1983-01-01

    The purpose of this project was to use a one-dimensional discrete coordinates transport code called ANISN in order to determine the energy-angle-spatial distribution of neutrons in a 6-feet cube rock box which houses a D-T neutron generator at its center. The project was two-fold. The first phase of the project involved adaptation of the ANISN code written for an IBM 360/75/91 computer to the UNIVAC system at JSC. The second phase of the project was to use the code with proper geometry, source function and rock material composition in order to determine the neutron flux distribution around the rock box when a 14.1 MeV neutron generator placed at its center is activated.

  13. Fast Neutron Transport in the Biological Shielding Model and Other Regions of the VVER-1000 Mock-Up on the LR-0 Research Reactor

    Science.gov (United States)

    Košťál, Michal; Milčák, Ján; Cvachovec, František; Jánský, Bohumil; Rypar, Vojtěch; Juříček, Vlastimil; Novák, Evžen; Egorov, Alexander; Zaritskiy, Sergey

    2016-02-01

    A set of benchmark experiments was carried out in the full scale VVER-1000 mock-up on the reactor LR-0 in order to validate neutron transport calculation methodologies and to perform the optimization of the shape and locations of neutron flux operation monitors channels inside the shielding of the new VVER-1000 type reactors. Compared with previous experiments on the VVER-1000 mock-up on the reactor LR-0, the fast neutron spectra were measured in the extended neutron energy interval (0.1-10 MeV) and new calculations were carried out with the MCNPX code using various nuclear data libraries (ENDF/B VII.0, JEFF 3.1, JENDL 3.3, JENDL 4, ROSFOND 2009, and CENDL 3.1). Measurements and calculations were carried out at different points in the mock-up. The calculation and experimental data are compared.

  14. Fast Neutron Transport in the Biological Shielding Model and Other Regions of the VVER-1000 Mock-Up on the LR-0 Research Reactor

    Directory of Open Access Journals (Sweden)

    Košťál Michal

    2016-01-01

    Full Text Available A set of benchmark experiments was carried out in the full scale VVER-1000 mock-up on the reactor LR-0 in order to validate neutron transport calculation methodologies and to perform the optimization of the shape and locations of neutron flux operation monitors channels inside the shielding of the new VVER-1000 type reactors. Compared with previous experiments on the VVER-1000 mock-up on the reactor LR-0, the fast neutron spectra were measured in the extended neutron energy interval (0.1–10 MeV and new calculations were carried out with the MCNPX code using various nuclear data libraries (ENDF/B VII.0, JEFF 3.1, JENDL 3.3, JENDL 4, ROSFOND 2009, and CENDL 3.1. Measurements and calculations were carried out at different points in the mock-up. The calculation and experimental data are compared.

  15. VIM4.0, Stead-State 3-D Neutron Transport Using ENDF/B or Multigroup Cross Sections

    International Nuclear Information System (INIS)

    1 - Description of program or function: VIM solves the steady-state neutron or photon transport problem in any detailed three-dimensional geometry using either continuous energy-dependent ENDF nuclear data or multigroup cross sections. Neutron transport is carried out in a criticality mode, or in a fixed source mode (optionally incorporating subcritical multiplication). Photon transport is simulated in the fixed source mode. The geometry options are infinite medium, combinatorial geometry, and hexagonal or rectangular lattices of combinatorial geometry unit cells, and rectangular lattices of cells of assembled plates. Boundary conditions include vacuum, specular and white reflection, and periodic boundaries for reactor cell calculations. The VIM 4.0 distribution includes data from ENDF/B-IV, ENDF/B-V, ENDF/B-VI and JEF2.2. Binary sequential data libraries for use with the code system on IBM or Sun workstations are included. ASCII data libraries and a convenient means to convert them to binary on a target machine are included for users on other systems. In addition to be included in the RSICC distribution files, the VIM User Guide is available on the developer's web site http://www.ra.anl.gov/vimguide/. 2 - Methods:VIM uses standard Monte Carlo methods for particle tracking with several optional variance-reduction techniques. These include splitting/Russian roulette, non-terminating absorption with non-analog weight cutoff energy. The keff is determined by the optimum linear combinations of two of the three eigenvalue estimates - analog, collision, and track length. Resonance and smooth cross sections are specified pointwise with linear-linear interpolation, frequently with many thousands of energy points. Unresolved resonances are described by the probability table method, which allows the statistical nature of the evaluated resonance cross sections to be incorporated naturally into self-shielding. Neutron interactions are elastic, inelastic and thermal scattering

  16. Collisionality scaling of turbulence and transport in advanced inductive plasmas in DIII-D

    Science.gov (United States)

    Yan, Z.; McKee, G. R.; Petty, C.; Luce, T.; Chen, X.; Holland, C.; Rhodes, T.; Schmitz, L.; Wang, G.; Zeng, L.; Marinoni, A.; Solomon, W.; DIII-D Team

    2015-11-01

    The collisionality scaling of multiscale turbulence properties and thermal transport characteristics in high-beta, high confinement Advanced Inductive (AI) plasmas was determined via systematic dimensionless scaling experiments on DIII-D. Preliminary estimate indicates a weak collisionality dependence of energy confinement as v* varied by a factor of ~2. Electron density and scaled (~Bt2) temperature profiles are well matched in the scan. Interestingly, low-k density fluctuation amplitudes are observed to decrease at lower v* near ρ ~ 0 . 75 . Ion and electron thermal transport values, computed with ONETWO using experimentally measured profiles and sources, will be presented, along with multi-scale turbulence measurements obtained with various fluctuation diagnostics. Altering collisionality should change the relative contribution of different modes to transport.

  17. Recent advances in statistical methods for the estimation of sediment and nutrient transport in rivers

    Science.gov (United States)

    Colin, T. A.

    1995-07-01

    This paper reviews advances in methods for estimating fluvial transport of suspended sediment and nutrients. Research from the past four years, mostly dealing with estimating monthly and annual loads, is emphasized. However, because this topic has not appeared in previous IUGG reports, some research prior to 1990 is included. The motivation for studying sediment transport has shifted during the past few decades. In addition to its role in filling reservoirs and channels, sediment is increasingly recognized as an important part of fluvial ecosystems and estuarine wetlands. Many groups want information about sediment transport [Bollman, 1992]: Scientists trying to understand benthic biology and catchment hydrology; citizens and policy-makers concerned about environmental impacts (e.g. impacts of logging [Beschta, 1978] or snow-fences [Sturges, 1992]); government regulators considering the effectiveness of programs to protect in-stream habitat and downstream waterbodies; and resource managers seeking to restore wetlands.

  18. Overview of the projects recently developed by the advanced neutron environment team at the ILL

    Science.gov (United States)

    Bourgeat-Lami, Eric; Chapuis, Jean-François; Chastagnier, Jérémie; Demas, Steffen; Gonzales, Jean-Paul; Keay, Morley-Patrick; Laborier, Jean-Luc; Lelièvre-Berna, Eddy; Losserand, Olivier; Martin, Paul; Mélési, Louis; Petoukhov, Alexander; Pujol, Serge; Ragazzoni, Jean-Louis; Thomas, Frédéric; Tonon, Xavier

    2006-11-01

    Within the framework of the Millennium Programme, we have started the design and building of novel equipment with the aim at facilitating and diversifying the experimental conditions on Institut Laue-Langevin (ILL) and ILL-CRG instruments. We anticipate new devices for applying external parameters (pressure, temperature, magnetic or electric fields, etc.), handling the neutron beam polarisation (RF wide-band flipper, TOF-Cryopad, etc.) or carrying multi-task experiments. The facilities already in operation are briefly reviewed: 3 K cryogen-free cryostat hosting the Paris-Edinburgh pressure cell, 3 K pulse-tube top-loading cryostat with 700 K high-temperature insert, 2 K Joule-Thomson cryogen-free cryostat for Eulerian cradles, 20 mK dilution fridge for the recently acquired 15 T cryomagnet and a low-temperature gas-injection sample stick for Orange cryostats.

  19. Progress report on R and D results from the advanced neutron source

    International Nuclear Information System (INIS)

    This presentation consists of six parts describing the the following: Oxide Formation; U3Si2 Fuel Performance; Aluminum Irradiation Properties and Code Case; Fuel Plate Hydraulic Stability; Thermal-Hydraulic Test Loop designed to examine the CHF/flow instability limits and thermal hydraulic correlations of the ANS core; Cold Source Design Concept. HFIR results indicate good performance of U3Si2 for ANS conditions. Additional data from HFIR tests, RERTR fuel, and simulation experiments are expected to improve understanding of basic behavior. Further research plans for the cold neutron source are: Test the circulation system, Test beryllium fabricability and properties, Develop and test modified pressure-balanced cryostat, design, if possible, of safe continued operation of the reactor even if cold source refrigeration is lost

  20. Neutron diffraction and electrical transport studies on magnetic ordering in terbium at high pressures and low temperatures

    Energy Technology Data Exchange (ETDEWEB)

    Thomas, Sarah [University of Alabama, Birmingham; Montgomery, Jeffrey M [University of Alabama, Birmingham; Tsoi, Georgiy [University of Alabama, Birmingham; Vohra, Yogesh [University of Alabama, Birmingham; Chesnut, Gary Neal [University of Alabama, Birmingham; Weir, S. T. [Lawrence Livermore National Laboratory (LLNL); Tulk, Christopher A [ORNL; Moreira Dos Santos, Antonio F [ORNL

    2013-01-01

    Neutron diffraction and electrical transport measurements have been carried out on the heavy rare-earth metal terbium at high pressures and low temperatures in order to elucidate the onset of ferromagnetic (FM) order as a function of pressure. The electrical resistance measurements show a change in slope as the temperature is lowered through the FM Curie temperature. The temperature of this FM transition decreases at a rate of-16.7 K/GPa up to a pressure of 3.6 GPa, at which point the onset of FM order is suppressed. The neutron diffraction measurements as a function of pressure at temperatures ranging from 90 to 290 K confirm that the change of slope in the resistance is associated with the FM ordering, since this occurs at pressures similar to those determined from the resistance results at these temperatures. A disappearance of FM ordering was observed as the pressure is increased above 3.6 GPa and is correlated with the phase transition from the ambient hexagonal close packed structure to an -Sm-type structure at high pressures.

  1. ACCELERATING FUSION REACTOR NEUTRONICS MODELING BY AUTOMATIC COUPLING OF HYBRID MONTE CARLO/DETERMINISTIC TRANSPORT ON CAD GEOMETRY

    Energy Technology Data Exchange (ETDEWEB)

    Biondo, Elliott D [ORNL; Ibrahim, Ahmad M [ORNL; Mosher, Scott W [ORNL; Grove, Robert E [ORNL

    2015-01-01

    Detailed radiation transport calculations are necessary for many aspects of the design of fusion energy systems (FES) such as ensuring occupational safety, assessing the activation of system components for waste disposal, and maintaining cryogenic temperatures within superconducting magnets. Hybrid Monte Carlo (MC)/deterministic techniques are necessary for this analysis because FES are large, heavily shielded, and contain streaming paths that can only be resolved with MC. The tremendous complexity of FES necessitates the use of CAD geometry for design and analysis. Previous ITER analysis has required the translation of CAD geometry to MCNP5 form in order to use the AutomateD VAriaNce reducTion Generator (ADVANTG) for hybrid MC/deterministic transport. In this work, ADVANTG was modified to support CAD geometry, allowing hybrid (MC)/deterministic transport to be done automatically and eliminating the need for this translation step. This was done by adding a new ray tracing routine to ADVANTG for CAD geometries using the Direct Accelerated Geometry Monte Carlo (DAGMC) software library. This new capability is demonstrated with a prompt dose rate calculation for an ITER computational benchmark problem using both the Consistent Adjoint Driven Importance Sampling (CADIS) method an the Forward Weighted (FW)-CADIS method. The variance reduction parameters produced by ADVANTG are shown to be the same using CAD geometry and standard MCNP5 geometry. Significant speedups were observed for both neutrons (as high as a factor of 7.1) and photons (as high as a factor of 59.6).

  2. Using 2H labeling with neutron radiography for the study of solid polymer electrolyte water transport properties.

    Science.gov (United States)

    Boillat, P; Oberholzer, P; Seyfang, B C; Kästner, A; Perego, R; Scherer, G G; Lehmann, E H; Wokaun, A

    2011-06-15

    A method combining (2)H labeling of different sources of H atoms (hydrogen, water vapor) with neutron imaging for the analysis of transport parameters in the bulk and at the interfaces of Nafion polymer electrolyte membranes is proposed. The use of different isotope compositions in the steady state allows evaluation of the relation between bulk and interface transport parameters, but relies on literature data for evaluating absolute values. By using transients of isotope composition, absolute values of these parameters including the self-diffusion coefficient of H can be extracted, making this method an attractive alternative to self-diffusion measurements using nuclear magnetic resonance (NMR), allowing measurements in precisely controlled conditions in real fuel cell structures. First measurements were realized on samples with and without electrodes and we report values of the self-diffusion coefficient of the same order of magnitude as values measured using NMR, although with slightly higher numbers. In our particular case, lower interfacial exchange rates for water transport were observed for samples with an electrode. PMID:21613688

  3. Development of advanced materials for spallation neutron sources and radiation damage simulation based on multi-scale models

    Energy Technology Data Exchange (ETDEWEB)

    Kawai, Masayoshi, E-mail: masayoshi.kawai@kek.jp [High Energy Accelerator Research Organization (KEK), Tsukuba, Ibaraki 305-0801 (Japan); Kurishita, Hiroaki [Institute for Materials Research, Tohoku University, Oarai, Ibaraki 311-1313 (Japan); Kokawa, Hiroyuki [Graduate School of Engineering, Tohoku University, Aramaki Aza Aoba, Aoba-ku, Sendai, Miyagi 980-8579 (Japan); Watanabe, Seiichi; Sakaguchi, Norihito [School of Engineering, Hokkaido University, Kita-ku, Sapporo 060-8628 (Japan); Kikuchi, Kenji [Institute of Applied Beam Science, Ibaraki University, Tokai-mura, Naka-gun, Ibaraki 319-1195 (Japan); Saito, Shigeru [J-PARC, Japan Atomic Energy Agency (JAEA), Tokai-mura, Naka-gun, Ibaraki 319-1195 (Japan); Yoshiie, Toshimasa [Kyoto University Research Reactor Institute, Kumatori, Sennan-gun, Osaka 590-0494 (Japan); Iwase, Hiroshi [High Energy Accelerator Research Organization (KEK), Tsukuba, Ibaraki 305-0801 (Japan); Ito, Takahiro [Toyohashi University of Technology, Tenpaku-cho, Toyohasi-shi, Aichi 441-8580 (Japan); Hashimoto, Satoshi; Kaneko, Yoshihisa [Osaka City University, Sugimoto, Sumiyoshi-ku, Osaka 558-8585 (Japan); Futakawa, Masatoshi [J-PARC, Japan Atomic Energy Agency (JAEA), Tokai-mura, Naka-gun, Ibaraki 319-1195 (Japan); Ishino, Shiori [University of Tokyo, Hongo, Bunkyo-ku, Tokyo 113-8656 (Japan)

    2012-12-15

    This report describes the status review of the JSPS Grant Team to develop advanced materials for the spallation neutron sources and modeling of radiation damage. One of the advanced materials is a toughness enhanced, fine-grained tungsten material (W-TiC) having four-times larger fracture toughness than ordinary tungsten and appreciable RT ductility in the recrystallized state. The other is an intergranular crack (IGC)-resistant austenitic stainless steel which was processed by the grain-boundary engineering (GBE). The experimental results are devoted to corrosion in a lead-bismuth eutectic, arrest of corrosion of weld-decay, radiation damage and creep rupture as well as new technique of GBE using a laser and annealing procedure. New technique seems to be applicable to large or complicated-shaped components. A series of the multi-scale models is built up from nuclear reaction between incident particles and medium nuclei to material property change due to radiation damage. Sample calculation is made on 3 mm-thick nickel bombarded by 3 GeV protons.

  4. Development of advanced materials for spallation neutron sources and radiation damage simulation based on multi-scale models

    Science.gov (United States)

    Kawai, Masayoshi; Kurishita, Hiroaki; Kokawa, Hiroyuki; Watanabe, Seiichi; Sakaguchi, Norihito; Kikuchi, Kenji; Saito, Shigeru; Yoshiie, Toshimasa; Iwase, Hiroshi; Ito, Takahiro; Hashimoto, Satoshi; Kaneko, Yoshihisa; Futakawa, Masatoshi; Ishino, Shiori; JSPS Grant Team

    2012-12-01

    This report describes the status review of the JSPS Grant Team to develop advanced materials for the spallation neutron sources and modeling of radiation damage. One of the advanced materials is a toughness enhanced, fine-grained tungsten material (W-TiC) having four-times larger fracture toughness than ordinary tungsten and appreciable RT ductility in the recrystallized state. The other is an intergranular crack (IGC)-resistant austenitic stainless steel which was processed by the grain-boundary engineering (GBE). The experimental results are devoted to corrosion in a lead-bismuth eutectic, arrest of corrosion of weld-decay, radiation damage and creep rupture as well as new technique of GBE using a laser and annealing procedure. New technique seems to be applicable to large or complicated-shaped components. A series of the multi-scale models is built up from nuclear reaction between incident particles and medium nuclei to material property change due to radiation damage. Sample calculation is made on 3 mm-thick nickel bombarded by 3 GeV protons.

  5. Development of advanced materials for spallation neutron sources and radiation damage simulation based on multi-scale models

    International Nuclear Information System (INIS)

    This report describes the status review of the JSPS Grant Team to develop advanced materials for the spallation neutron sources and modeling of radiation damage. One of the advanced materials is a toughness enhanced, fine-grained tungsten material (W-TiC) having four-times larger fracture toughness than ordinary tungsten and appreciable RT ductility in the recrystallized state. The other is an intergranular crack (IGC)-resistant austenitic stainless steel which was processed by the grain-boundary engineering (GBE). The experimental results are devoted to corrosion in a lead–bismuth eutectic, arrest of corrosion of weld-decay, radiation damage and creep rupture as well as new technique of GBE using a laser and annealing procedure. New technique seems to be applicable to large or complicated-shaped components. A series of the multi-scale models is built up from nuclear reaction between incident particles and medium nuclei to material property change due to radiation damage. Sample calculation is made on 3 mm-thick nickel bombarded by 3 GeV protons.

  6. Users Manual for TART 2002: A Coupled Neutron-Photon 3-D, Combinatorial Geometry Time Dependent Monte Carlo Transport Code

    Energy Technology Data Exchange (ETDEWEB)

    Cullen, D E

    2003-06-06

    TART 2002 is a coupled neutron-photon, 3 Dimensional, combinatorial geometry, time dependent Monte Carlo radiation transport code. This code can run on any modern computer. It is a complete system to assist you with input preparation, running Monte Carlo calculations, and analysis of output results. TART 2002 is also incredibly FAST; if you have used similar codes, you will be amazed at how fast this code is compared to other similar codes. Use of the entire system can save you a great deal of time and energy. TART 2002 is distributed on CD. This CD contains on-line documentation for all codes included in the system, the codes configured to run on a variety of computers, and many example problems that you can use to familiarize yourself with the system. TART 2002 completely supersedes all older versions of TART, and it is strongly recommended that users only use the most recent version of TART 2002 and its data files.

  7. TART 2000: A Coupled Neutron-Photon, 3-D, Combinatorial Geometry, Time Dependent, Monte Carlo Transport Code

    Energy Technology Data Exchange (ETDEWEB)

    Cullen, D.E

    2000-11-22

    TART2000 is a coupled neutron-photon, 3 Dimensional, combinatorial geometry, time dependent Monte Carlo radiation transport code. This code can run on any modern computer. It is a complete system to assist you with input Preparation, running Monte Carlo calculations, and analysis of output results. TART2000 is also incredibly FAST; if you have used similar codes, you will be amazed at how fast this code is compared to other similar codes. Use of the entire system can save you a great deal of time and energy. TART2000 is distributed on CD. This CD contains on-line documentation for all codes included in the system, the codes configured to run on a variety of computers, and many example problems that you can use to familiarize yourself with the system. TART2000 completely supersedes all older versions of TART, and it is strongly recommended that users only use the most recent version of TART2000 and its data files.

  8. Calculations of reactivity based in the solution of the Neutron transport equation in X Y geometry and Lineal perturbation theory

    International Nuclear Information System (INIS)

    In our country, in last congresses, Gomez et al carried out reactivity calculations based on the solution of the diffusion equation for an energy group using nodal methods in one dimension and the TPL approach (Lineal Perturbation Theory). Later on, Mugica extended the application to the case of multigroup so much so much in one as in two dimensions (X Y geometry) with excellent results. Presently work is carried out similar calculations but this time based on the solution of the neutron transport equation in X Y geometry using nodal methods and again the TPL approximation. The idea is to provide a calculation method that allows to obtain in quick form the reactivity solving the direct problem as well as the enclosed problem of the not perturbed problem. A test problem for the one that results are provided for the effective multiplication factor is described and its are offered some conclusions. (Author)

  9. Performance tests on PNL's transportable neutron/gamma waste waste assay system

    International Nuclear Information System (INIS)

    Battelle Pacific Northwest Laboratory, in conjunction with Canberra Industries, has implemented a 55-gallon drum waste assay system. The single system unit consists of a combined segmented gamma assay system and a neutron assay system. The unit is designed to function either in the laboratory or in a mobile trailer. The system is on wheels and can be moved through standard double doors. The gamma system uses an HPGe detector with a Se-75 source for transmission corrections. The neutron detector uses 40 He-3 detectors connected to a JSR-12 neutron coincidence counter. The system's software is unique and is interactive with the user; it features a menu driven operator screen from which all functions regarding operations and calibrations can be selected. Single or combined assays with various setups, including containers smaller than 55 gallons, may be performed. The software and analysis is designed for unknown waste contents, but allows input of waste stream information prior to assay. The system was originally designed for safeguards' MC ampersand A requirements and has enough sensitivity to determine whether a drum is TRU or LLW in one assay pass. Typical counting times are approximately 1800 seconds for a dual pass. Preliminary testing of the system with the available Pu standards has shown the system will perform to the required levels stated in the Data Quality Objectives of the WIPP Performance Demonstration program. An overall study of the system is underway to determine the lower limit of detection (LLD) for different isotopes, to best utilize the combined assay results, and to apply the appropriate data corrections for more complete answers, such as corrections for the end effects. Results from these developments will be presented at the conference

  10. Two dimensional neutron transport calculation system for plate-reactors: experimental design and qualification with SILOE

    International Nuclear Information System (INIS)

    The main objective of this work is to create a neutronic calculations system for the SILOE-SILOETTE reactors, adaptable to other types of plate reactors. The author presents the methodology and the development of the APOLLO 1D (99 gr.) calculations for the creation of cross sections libraries. After a recall of the Discrete Ordinate Method (DOT), the method accuracy is studied in order to optimize the spatial discretization of the calculations; calculations of DOT 3.5 and of SILOETTE core are conducted and their convergence and costs are examined. DOT calculations of SILOETTE and experimental tests results are then compared

  11. Electrostatic design and beam transport for a folded tandem electrostatic quadrupole accelerator facility for accelerator-based boron neutron capture therapy

    Energy Technology Data Exchange (ETDEWEB)

    Thatar Vento, V., E-mail: Vladimir.ThatarVento@gmail.com [Gerencia de Investigacion y Aplicaciones, CNEA, Av. Gral. Paz 1499 (1650), San Martin, Buenos Aires (Argentina)] [CONICET, Av. Rivadavia 1917 (1033), Ciudad Autonoma de Buenos Aires (Argentina); Bergueiro, J.; Cartelli, D. [Gerencia de Investigacion y Aplicaciones, CNEA, Av. Gral. Paz 1499 (1650), San Martin, Buenos Aires (Argentina)] [CONICET, Av. Rivadavia 1917 (1033), Ciudad Autonoma de Buenos Aires (Argentina); Valda, A.A. [Gerencia de Investigacion y Aplicaciones, CNEA, Av. Gral. Paz 1499 (1650), San Martin, Buenos Aires (Argentina)] [Escuela de Ciencia y Tecnologia, UNSAM, M. Irigoyen 3100 (1650), San Martin, Buenos Aires (Argentina); Kreiner, A.J. [Gerencia de Investigacion y Aplicaciones, CNEA, Av. Gral. Paz 1499 (1650), San Martin, Buenos Aires (Argentina)] [CONICET, Av. Rivadavia 1917 (1033), Ciudad Autonoma de Buenos Aires (Argentina)] [Escuela de Ciencia y Tecnologia, UNSAM, M. Irigoyen 3100 (1650), San Martin, Buenos Aires (Argentina)

    2011-12-15

    Within the frame of an ongoing project to develop a folded Tandem-Electrostatic-Quadrupole (TESQ) accelerator facility for Accelerator-Based Boron Neutron Capture Therapy (AB-BNCT), we discuss here the electrostatic design of the machine, including the accelerator tubes with electrostatic quadrupoles and the simulations for the transport and acceleration of a high intensity beam.

  12. 3-D Deep Penetration Neutron Imaging of Thick Absorbing and Diffusive Objects Using Transport Theory. Final technical report

    International Nuclear Information System (INIS)

    here explores the inverse problem of optical tomography applied to heterogeneous domains. The neutral particle transport equation was used as the forward model for how neutral particles stream through and interact within these heterogeneous domains. A constrained optimization technique that uses Newtons method served as the basis of the inverse problem. Optical tomography aims at reconstructing the material properties using (a) illuminating sources and (b) detector readings. However, accurate simulations for radiation transport require that the particle (gamma and/or neutron) energy be appropriate discretize in the multigroup approximation. This, in turns, yields optical tomography problems where the number of unknowns grows (1) about quadratically with respect to the number of energy groups, G, (notably to reconstruct the scattering matrix) and (2) linearly with respect to the number of unknown material regions. As pointed out, a promising approach could rely on algorithms to appropriately select a material type per material zone rather than G2 values. This approach, though promising, still requires further investigation: (a) when switching from cross-section values unknowns to material type indices (discrete integer unknowns), integer programming techniques are needed since derivative information is no longer available; and (b) the issue of selecting the initial material zoning remains. The work reported here proposes an approach to solve the latter item, whereby a material zoning is proposed using one-group or few-groups transport approximations. The capabilities and limitations of the presented method were explored; they are briefly summarized next and later described in fuller details in the Appendices. The major factors that influenced the ability of the optimization method to reconstruct the cross sections of these domains included the locations of the sources used to illuminate the domains, the number of separate experiments used in the reconstruction, the

  13. Advanced sample environments for in situ neutron diffraction studies of nuclear materials

    Science.gov (United States)

    Reiche, Helmut Matthias

    Generation IV nuclear reactor concepts, such as the supercritical-water-cooled nuclear reactor (SCWR), are actively researched internationally. Operating conditions above the critical point of water (374°C, 22.1 MPa) and fuel core temperature that potentially exceed 1850°C put a high demand on the surrounding materials. For their safe application, it is essential to characterize and understand the material properties on an atomic scale such as crystal structure and grain orientation (texture) changes as a function of temperature and stress. This permits the refinement of models predicting the macroscopic behavior of the material. Neutron diffraction is a powerful tool in characterizing such crystallographic properties due to their deep penetration depth into condensed matter. This leads to the ability to study bulk material properties, as opposed to surface effects, and allows for complex sample environments to study e.g. the individual contributions of thermo-mechanical processing steps during manufacturing, operating or accident scenarios. I present three sample environments for in situ neutron diffraction studies that provide such crystallographic information and have been successfully commissioned and integrated into the user program of the High Pressure -- Preferred Orientation (HIPPO) diffractometer at the Los Alamos Neutron Science Center (LANSCE) user facility. I adapted a sample changer for reliable and fast automated texture measurements of multiple specimens. I built a creep furnace combining a 2700 N load frame with a resistive vanadium furnace, capable of temperatures up to 1000°C, and manipulated by a pair of synchronized rotation stages. This combination allows following deformation and temperature dependent texture and strain evolutions in situ. Utilizing the presented sample changer and creep furnace we studied pressure tubes made of Zr-2.5wt%Nb currently employed in CANDURTM nuclear reactors and proposed for future SCWRs, acting as the primary

  14. Some contributions towards the parallel simulation of time dependent neutron transport and the integration of observed data in real time

    International Nuclear Information System (INIS)

    In this thesis, we have first developed a time dependent 3D neutron transport solver on unstructured meshes with discontinuous Galerkin finite elements spatial discretization. The solver (called MINARET) represents in itself an important contribution in reactor physics thanks to the accuracy that it can provide in the knowledge of the state of the core during severe accidents. It will also play an important role on vessel fluence calculations. From a mathematical point of view, the most important contribution has consisted in the implementation of modern algorithms that are well adapted for modern parallel architectures and that significantly decrease the computing times. A special effort has been done in order to efficiently parallelize the time variable by the use of the parareal in time algorithm. For this, we have first analyzed the performances that the classical scheme of parareal can provide when applied to the resolution of the neutron transport equation in a reactor core. Then, with the purpose of improving these performances, a parareal scheme that takes more efficiently into account the presence of other iterative schemes in the resolution of each time step has been proposed. The main idea consists in limiting the number of internal iterations for each time step and to reach convergence across the parareal iterations. A second phase of our work has been motivated by the following question: given the high degree of accuracy that MINARET can provide in the modeling of the neutron population, could we somehow use it as a tool to monitor in real time the population of neutrons on the purpose of helping in the operation of the reactor? And, what is more, how to make such a tool be coherent in some sense with the measurements taken in situ? One of the main challenges of this problem is the real time aspect of the simulations. Indeed, despite all of our efforts to speed-up the calculations, the discretization methods used in MINARET do not provide simulations

  15. Transport methods: general. 3. An Additive Angular-Dependent Re-balance Acceleration Method for Neutron Transport Equations

    International Nuclear Information System (INIS)

    An additive angular-dependent re-balance (AADR) factor acceleration method is described to accelerate the source iteration of discrete ordinates transport calculation. The formulation of the AADR method follows that of the angular-dependent re-balance (ADR) method in that the re-balance factor is defined only on the cell interface and in that the low-order equation is derived by integrating the transport equation (high-order equation) over angular subspaces. But, the re-balance factor is applied additively. While the AADR method is similar to the boundary projection acceleration and the alpha-weighted linear acceleration, it is more general and does have distinct features. The method is easily extendible to DPN and low-order SN re-balancing, and it does not require consistent discretizations between the high- and low-order equations as in diffusion synthetic acceleration. We find by Fourier analysis and numerical results that the AADR method with a chosen form of weighting functions is unconditionally stable and very effective. There also exists an optimal weighting parameter that leads to the smallest spectral radius. The AADR acceleration method described in this paper is simple to implement, unconditionally stable, and very effective. It uses a physically based weighting function with an optimal parameter, leading to the best spectral radius of ρ<0.1865, compared to ρ<0.2247 of DSA. The application of the AADR acceleration method with the LMB scheme on a test problem shows encouraging results

  16. Analytical Tests for Ray Effect Errors in Discrete Ordinate Methods for Solving the Neutron Transport Equation

    Energy Technology Data Exchange (ETDEWEB)

    Chang, B

    2004-03-22

    This paper contains three analytical solutions of transport problems which can be used to test ray-effect errors in the numerical solutions of the Boltzmann Transport Equation (BTE). We derived the first two solutions and the third was shown to us by M. Prasad. Since this paper is intended to be an internal LLNL report, no attempt was made to find the original derivations of the solutions in the literature in order to cite the authors for their work.

  17. Solution of the Neutron transport equation in hexagonal geometry using strongly discontinuous nodal schemes; Solucion de la Ecuacion de transporte de neutrones en geometria hexagonal usando esquemas nodales fuertemente discontinuos

    Energy Technology Data Exchange (ETDEWEB)

    Mugica R, C.A.; Valle G, E. del [IPN, ESFM, Departamento de Ingenieria Nuclear, 07738 Mexico D.F. (Mexico)]. e-mail: cmugica@ipn.mx

    2005-07-01

    In 2002, E. del Valle and Ernest H. Mund developed a technique to solve numerically the Neutron transport equations in discrete ordinates and hexagonal geometry using two nodal schemes type finite element weakly discontinuous denominated WD{sub 5,3} and WD{sub 12,8} (of their initials in english Weakly Discontinuous). The technique consists on representing each hexagon in the union of three rhombuses each one of which it is transformed in a square in the one that the methods WD{sub 5,3} and WD{sub 12,8} were applied. In this work they are solved the mentioned equations of transport using the same discretization technique by hexagon but using two nodal schemes type finite element strongly discontinuous denominated SD{sub 3} and SD{sub 8} (of their initials in english Strongly Discontinuous). The application in each case as well as a reference problem for those that results are provided for the effective multiplication factor is described. It is carried out a comparison with the obtained results by del Valle and Mund for different discretization meshes so much angular as spatial. (Author)

  18. OSMOSE an experimental program for improving neutronic predictions of advanced nuclear fuels.

    Energy Technology Data Exchange (ETDEWEB)

    Klann, R. T.; Aliberti, G.; Zhong, Z.; Graczyk, D.; Loussi, A.; Nuclear Engineering Division; Commissariat a l Energie Atomique

    2007-10-18

    This report describes the technical results of tasks and activities conducted in FY07 to support the DOE-CEA collaboration on the OSMOSE program. The activities are divided into five high-level tasks: reactor modeling and pre-experiment analysis, sample fabrication and analysis, reactor experiments, data treatment and analysis, and assessment for relevance to high priority advanced reactor programs (such as GNEP and Gen-IV).

  19. On an optimized neutron shielding for an advanced molten salt fast reactor design

    International Nuclear Information System (INIS)

    The molten salt reactor technology has gained renewed interest. In contrast to the historic molten salt reactors, the current projects are based on designing a molten salt fast reactor. Thus the shielding becomes significantly more challenging than in historic concepts. One very interesting and innovative result of the most recent EURATOM project on molten salt reactors – EVOL – is the fluid flow optimized design of the inner core vessel using curved blanket walls. The developed structure leads to a very uniform flow distribution. The design avoids all core internal structures. On the basis of this new geometry a model for neutron physics calculation is presented and applied for a shielding optimization. Based on these results an optimized shielding strategy is developed for the molten salt fast reactor to keep the fluence in the safety related outer vessel below expected limit values. A lifetime of 80 years can be assured, but the size of the core/blanket system has to be significantly increased and will finally be comparable to a sodium cooled fast reactor. The HELIOS results are verified against Monte-Carlo calculations with very satisfactory agreement for a deep penetration problem. (author)

  20. GOLEM: a versatile computer code for reactor neutronic calculation advances in qualification of the different modules

    International Nuclear Information System (INIS)

    The last 12 years studies about the CABRI, SCARABEE and PHEBUS projects are summarized. It describes the object and the genesis of the cores, the evolution of the core concept and the associated neutronic problems. The calculational scheme used is presented, together with its qualification. The formalism, and the qualification of the different modules of GOLEM are presented. COXYS: module of physical analysis in order to determine the best energetic and spatial mesh for the case of interest. GOLU.B: input data management module. VAREC: calculation module of perturbations due to materials enables to compute perturbed flux and reactivity variation. VARYX: calculation module of geometric perturbations. TRACASYN: module of 3D power shape calculation. Finally TRACASTORE: module of management and graphic exploitation of results. Then, one gives utilization directions for these different modules. Qualification results show that GOLEM is able to analyse the fine physics of many various cases, to calculate by perturbation effects greater than 5000 pcm, to rebuild perturbed flux with margins near 3% for difficult situations, like reactor voiding or spectral or spectral variation in a PWR. Furthermore, 3D hot spots are calculated within margins of a magnitude comparable to experimental ones

  1. Combining geoelectrical and advanced lysimeter methods to characterize heterogeneous flow and transport under unsaturated transient conditions

    Science.gov (United States)

    Wehrer, M.; Skowronski, J.; Binley, A. M.; Slater, L. D.

    2013-12-01

    Our ability to predict flow and transport processes in the unsaturated critical zone is considerably limited by two characteristics: heterogeneity of flow and transience of boundary conditions. The causes of heterogeneous - or preferential - flow and transport are fairly well understood, yet the characterization and quantification of such processes in natural profiles remains challenging. This is due to current methods of observation, such as staining and isotope tracers, being unable to observe multiple events on the same profile and offering limited spatial information. In our study we demonstrate an approach to characterize preferential flow and transport processes applying a combination of geoelectrical methods and advanced lysimeter techniques. On an agricultural soil profile, which was transferred undisturbed into a lysimeter container, we applied systematically varied input flow boundary conditions, resembling natural precipitation events. We simultaneously measured the breakthrough of a conservative tracer. Flow and transport in the soil column were observed using electrical resistivity tomography (ERT), tensiometers, water content probes and a multicompartment suction plate (MSP). These techniques allowed a direct ground-truthing of soil moisture and pore fluid resistivity changes estimated noninvasively using ERT. We were able to image both the advancing infiltration front and the advancing tracer front using time lapse ERT. Water content changes associated with the advancing infiltration front dominated over pore fluid conductivity changes during short term precipitation events. Conversely, long term displacement of the solute front was monitored during periods of constant water content in between infiltration events. We observed preferential flow phenomena through ERT and through the MSP, which agreed in general terms. The preferential flow fraction was observed to be independent of precipitation rate. This suggests the presence of a fingering process

  2. Algorithmic choices in WARP – A framework for continuous energy Monte Carlo neutron transport in general 3D geometries on GPUs

    International Nuclear Information System (INIS)

    Highlights: • WARP, a GPU-accelerated Monte Carlo neutron transport code, has been developed. • The NVIDIA OptiX high-performance ray tracing library is used to process geometric data. • The unionized cross section representation is modified for higher performance. • Reference remapping is used to keep the GPU busy as neutron batch population reduces. • Reference remapping is done using a key-value radix sort on neutron reaction type. - Abstract: In recent supercomputers, general purpose graphics processing units (GPGPUs) are a significant faction of the supercomputer’s total computational power. GPGPUs have different architectures compared to central processing units (CPUs), and for Monte Carlo neutron transport codes used in nuclear engineering to take advantage of these coprocessor cards, transport algorithms must be changed to execute efficiently on them. WARP is a continuous energy Monte Carlo neutron transport code that has been written to do this. The main thrust of WARP is to adapt previous event-based transport algorithms to the new GPU hardware; the algorithmic choices for all parts of which are presented in this paper. It is found that remapping history data references increases the GPU processing rate when histories start to complete. The main reason for this is that completed data are eliminated from the address space, threads are kept busy, and memory bandwidth is not wasted on checking completed data. Remapping also allows the interaction kernels to be launched concurrently, improving efficiency. The OptiX ray tracing framework and CUDPP library are used for geometry representation and parallel dataset-side operations, ensuring high performance and reliability

  3. Advanced Space Transportation Concepts and Propulsion Technologies for a New Delivery Paradigm

    Science.gov (United States)

    Robinson, John W.; McCleskey, Carey M.; Rhodes, Russel E.; Lepsch, Roger A.; Henderson, Edward M.; Joyner, Claude R., III; Levack, Daniel J. H.

    2013-01-01

    This paper describes Advanced Space Transportation Concepts and Propulsion Technologies for a New Delivery Paradigm. It builds on the work of the previous paper "Approach to an Affordable and Productive Space Transportation System". The scope includes both flight and ground system elements, and focuses on their compatibility and capability to achieve a technical solution that is operationally productive and also affordable. A clear and revolutionary approach, including advanced propulsion systems (advanced LOX rich booster engine concept having independent LOX and fuel cooling systems, thrust augmentation with LOX rich boost and fuel rich operation at altitude), improved vehicle concepts (autogeneous pressurization, turbo alternator for electric power during ascent, hot gases to purge system and keep moisture out), and ground delivery systems, was examined. Previous papers by the authors and other members of the Space Propulsion Synergy Team (SPST) focused on space flight system engineering methods, along with operationally efficient propulsion system concepts and technologies. This paper continues the previous work by exploring the propulsion technology aspects in more depth and how they may enable the vehicle designs from the previous paper. Subsequent papers will explore the vehicle design, the ground support system, and the operations aspects of the new delivery paradigm in greater detail.

  4. Validation of neutron-transport calculations in benchmark facilities for improved damage-fluence predictions

    International Nuclear Information System (INIS)

    An accurate determination of damage fluence accumulated by reactor pressure vessels (RPV) as a function of time is essential in order to evaluate the vessel integrity for both pressurized thermal shock (PTS) transients and end-of-life considerations. The desired accuracy for neutron exposure parameters such as displacements per atom or fluence (E > 1 MeV) is of the order of 20 to 30%. However, these types of accuracies can only be obtained realistically by validation of nuclear data and calculational methods in benchmark facilities. The purposes of this paper are to review the needs and requirements for benchmark experiments, to discuss the status of current benchmark experiments, to summarize results and conclusions obtained so far, and to suggest areas where further benchmarking is needed

  5. Study of the long-range transport of atmospheric pollutants by instrumental neutron activation analysis

    International Nuclear Information System (INIS)

    Aerosol samples were collected to study the characteristics of marine aerosols in the different western Pacific ocean areas. During the first cruise from 15 October to 25 November 1989, aerosol samples were collected with a kA-200 Andersen cascade impactor and a kB-120 sampler. Instrumental neutron activation analysis was used to determine the elemental composition of the aerosols. The concentrations of crustal and pollution elements in aerosols were higher over the ocean area close to the China coast and decreased very rapidly with increasing distance from land. The morphology and elemental composition of aerosol particles showed that the seasalt particles may conglomerate with small crustal and pollution particles from land to form large particles. (author). 4 refs, 1 fig., 1 tab

  6. Neutron transport-burnup code MCORGS and its application in fusion fission hybrid blanket conceptual research

    Science.gov (United States)

    Shi, Xue-Ming; Peng, Xian-Jue

    2016-09-01

    Fusion science and technology has made progress in the last decades. However, commercialization of fusion reactors still faces challenges relating to higher fusion energy gain, irradiation-resistant material, and tritium self-sufficiency. Fusion Fission Hybrid Reactors (FFHR) can be introduced to accelerate the early application of fusion energy. Traditionally, FFHRs have been classified as either breeders or transmuters. Both need partition of plutonium from spent fuel, which will pose nuclear proliferation risks. A conceptual design of a Fusion Fission Hybrid Reactor for Energy (FFHR-E), which can make full use of natural uranium with lower nuclear proliferation risk, is presented. The fusion core parameters are similar to those of the International Thermonuclear Experimental Reactor. An alloy of natural uranium and zirconium is adopted in the fission blanket, which is cooled by light water. In order to model blanket burnup problems, a linkage code MCORGS, which couples MCNP4B and ORIGEN-S, is developed and validated through several typical benchmarks. The average blanket energy Multiplication and Tritium Breeding Ratio can be maintained at 10 and 1.15 respectively over tens of years of continuous irradiation. If simple reprocessing without separation of plutonium from uranium is adopted every few years, FFHR-E can achieve better neutronic performance. MCORGS has also been used to analyze the ultra-deep burnup model of Laser Inertial Confinement Fusion Fission Energy (LIFE) from LLNL, and a new blanket design that uses Pb instead of Be as the neutron multiplier is proposed. In addition, MCORGS has been used to simulate the fluid transmuter model of the In-Zinerater from Sandia. A brief comparison of LIFE, In-Zinerater, and FFHR-E will be given.

  7. The bidimensional neutron transport code TWOTRAN-GG. Users manual and input data TWOTRAN-TRACA version; El codigo de transporte bidimensional TWOTRAN-GG. Manual de usuario y datos de entrada version TWOTRAN-TRACA

    Energy Technology Data Exchange (ETDEWEB)

    Ahnert, C.; Aragones, J. M.

    1981-07-01

    This Is a users manual of the neutron transport code TWOTRAN-TRACA, which is a version of the original TWOTRAN-GG from the Los Alamos Laboratory, with some modifications made at JEN. A detailed input data description is given as well as the new modifications developed at JEN. (Author) 8 refs.

  8. Analysis of a HP-refinement method for solving the neutron transport equation using two error estimators

    International Nuclear Information System (INIS)

    The solution of the time-independent neutron transport equation in a deterministic way invariably consists in the successive discretization of the three variables: energy, angle and space. In the SNATCH solver used in this study, the energy and the angle are respectively discretized with a multigroup approach and the discrete ordinate method. A set of spatial coupled transport equations is obtained and solved using the Discontinuous Galerkin Finite Element Method (DGFEM). Within this method, the spatial domain is decomposed into elements and the solution is approximated by a hierarchical polynomial basis in each one. This approach is time and memory consuming when the mesh becomes fine or the basis order high. To improve the computational time and the memory footprint, adaptive algorithms are proposed. These algorithms are based on an error estimation in each cell. If the error is important in a given region, the mesh has to be refined (h−refinement) or the polynomial basis order increased (p−refinement). This paper is related to the choice between the two types of refinement. Two ways to estimate the error are compared on different benchmarks. Analyzing the differences, a hp−refinement method is proposed and tested. (author)

  9. Neutron Transport and Nuclear Burnup Analysis for the Laser Inertial Confinement Fusion-Fission Energy (LIFE) Engine

    Energy Technology Data Exchange (ETDEWEB)

    Kramer, K J; Latkowski, J F; Abbott, R P; Boyd, J K; Powers, J J; Seifried, J E

    2008-10-24

    Lawrence Livermore National Laboratory is currently developing a hybrid fusion-fission nuclear energy system, called LIFE, to generate power and burn nuclear waste. We utilize inertial confinement fusion to drive a subcritical fission blanket surrounding the fusion chamber. It is composed of TRISO-based fuel cooled by the molten salt flibe. Low-yield (37.5 MJ) targets and a repetition rate of 13.3 Hz produce a 500 MW fusion source that is coupled to the subcritical blanket, which provides an additional gain of 4-8, depending on the fuel. In the present work, we describe the neutron transport and nuclear burnup analysis. We utilize standard analysis tools including, the Monte Carlo N-Particle (MCNP) transport code, ORIGEN2 and Monteburns to perform the nuclear design. These analyses focus primarily on a fuel composed of depleted uranium not requiring chemical reprocessing or enrichment. However, other fuels such as weapons grade plutonium and highly-enriched uranium are also under consideration. In addition, we have developed a methodology using {sup 6}Li as a burnable poison to replace the tritium burned in the fusion targets and to maintain constant power over the lifetime of the engine. The results from depleted uranium analyses suggest up to 99% burnup of actinides is attainable while maintaining full power at 2GW for more than five decades.

  10. Coupling of discrete ordinates methods by transmission of boundary conditions in solving the neutron transport equation in slab geometry

    International Nuclear Information System (INIS)

    Neutron transport in nuclear reactors is quite well modelled by the linear Boltzmann transport equation. Its solution is relatively easy, but unfortunately too expensive to achieve whole core computations. Thus, we have to simplify it, for example by homogenizing some physical characteristics. However, the solution may then be inaccurate. Moreover, in strongly homogeneous areas, the error may be too big. Then we would like to deal with such an inconvenient by solving the equation accurately on this area, but more coarsely away from it, so that the computation is not too expensive. This problem is the subject of a thesis. We present here some results obtained for slab geometry. The couplings between the fine and coarse discretization regions could be conceived in a number of approaches. Here, we only deal with the coupling at crossing the interface between two sub-domains. In the first section, we present the coupling of discrete ordinate methods for solving the homogeneous, isotropic and mono-kinetic equation. Coupling operators are defined and shown to be optimal. The second and the third sections are devoted to an extension of the previous results when the equation is non-homogeneous, anisotropic and multigroup (under some restrictive assumptions). Some numerical results are given in the case of isotropic and mono-kinetic equations. (author)

  11. Design of a high-current low-energy beam transport line for an intense D-T/D-D neutron generator

    Science.gov (United States)

    Lu, Xiaolong; Wang, Junrun; Zhang, Yu; Li, Jianyi; Xia, Li; Zhang, Jie; Ding, Yanyan; Jiang, Bing; Huang, Zhiwu; Ma, Zhanwen; Wei, Zheng; Qian, Xiangping; Xu, Dapeng; Lan, Changlin; Yao, Zeen

    2016-03-01

    An intense D-T/D-D neutron generator is currently being developed at the Lanzhou University. The Cockcroft-Walton accelerator, as a part of the neutron generator, will be used to accelerate and transport the high-current low-energy beam from the duoplasmatron ion source to the rotating target. The design of a high-current low-energy beam transport (LEBT) line and the dynamics simulations of the mixed beam were carried out using the TRACK code. The results illustrate that the designed beam line facilitates smooth transportation of a deuteron beam of 40 mA, and the number of undesired ions can be reduced effectively using two apertures.

  12. NUMERICAL SIMULATION OF SINGLE-GROUP, STEADY STATE AND ISOTROPIC NEUTRON TRANSPORT EQUATION IN DIFFUSIVE REGIMES

    Institute of Scientific and Technical Information of China (English)

    Ying Genjun; Fu Ying; Ma Yichen; Zhang Zhipeng

    2006-01-01

    We present an algorithm for numerical solution of transport equation in diffusive regimes, in which the transport equation is nearly singular and its solution becomes a solution of a diffusion equation. This algorithm, which is based on the Least-squares FEM in combination with a scaling transformation, presents a good approximation of a diffusion operator in diffusive regimes and guarantees an accurate discrete solution. The numerical experiments in 2D and 3D case are given, and the numerical results show that this algorithm is correct and efficient.

  13. Ultra low emission vehicle - transport using advance propulsion (ULEV-TAP)

    Energy Technology Data Exchange (ETDEWEB)

    Etemad, S. [Imperial College of Science, Technology and Medicine, London (United Kingdom)

    1999-07-01

    The increasing concern over emission pollutants that is related to the transportation sector has prompted rigorous environmental legislation in the USA and Europe. The zero emission vehicle solution based on battery technology, although attractive, appears to require step improvements in technology not realisable in the short term. The alternative, hybrid electric vehicle, is presently gaining support, with many attractive examples being made available in the market for the passenger vehicle application. The present contribution describes a newly formed project sponsored by the European Community (Brite-EuRam DG12 programme). The Ultra Low Emission Vehicle - Transport using Advanced Propulsion (ULEV-TAP) commenced in August 1997, and aims to demonstrate a hybrid powertrain based on a turbogenerator (high speed generator directly coupled to a gas turbine engine) prime mover and a flywheel energy storage unit. The demonstration vehicle is to be a regional rail vehicle operating in the city of Karlsruhe. (author)

  14. Advanced Transport Operating System (ATOPS) color displays software description microprocessor system

    Science.gov (United States)

    Slominski, Christopher J.; Plyler, Valerie E.; Dickson, Richard W.

    1992-01-01

    This document describes the software created for the Sperry Microprocessor Color Display System used for the Advanced Transport Operating Systems (ATOPS) project on the Transport Systems Research Vehicle (TSRV). The software delivery known as the 'baseline display system', is the one described in this document. Throughout this publication, module descriptions are presented in a standardized format which contains module purpose, calling sequence, detailed description, and global references. The global reference section includes procedures and common variables referenced by a particular module. The system described supports the Research Flight Deck (RFD) of the TSRV. The RFD contains eight cathode ray tubes (CRTs) which depict a Primary Flight Display, Navigation Display, System Warning Display, Takeoff Performance Monitoring System Display, and Engine Display.

  15. Advanced Transport Operating System (ATOPS) color displays software description: MicroVAX system

    Science.gov (United States)

    Slominski, Christopher J.; Plyler, Valerie E.; Dickson, Richard W.

    1992-01-01

    This document describes the software created for the Display MicroVAX computer used for the Advanced Transport Operating Systems (ATOPS) project on the Transport Systems Research Vehicle (TSRV). The software delivery of February 27, 1991, known as the 'baseline display system', is the one described in this document. Throughout this publication, module descriptions are presented in a standardized format which contains module purpose, calling sequence, detailed description, and global references. The global references section includes subroutines, functions, and common variables referenced by a particular module. The system described supports the Research Flight Deck (RFD) of the TSRV. The RFD contains eight Cathode Ray Tubes (CRTs) which depict a Primary Flight Display, Navigation Display, System Warning Display, Takeoff Performance Monitoring System Display, and Engine Display.

  16. Advanced Transport Operating System (ATOPS) Flight Management/Flight Controls (FM/FC) software description

    Science.gov (United States)

    Wolverton, David A.; Dickson, Richard W.; Clinedinst, Winston C.; Slominski, Christopher J.

    1993-01-01

    The flight software developed for the Flight Management/Flight Controls (FM/FC) MicroVAX computer used on the Transport Systems Research Vehicle for Advanced Transport Operating Systems (ATOPS) research is described. The FM/FC software computes navigation position estimates, guidance commands, and those commands issued to the control surfaces to direct the aircraft in flight. Various modes of flight are provided for, ranging from computer assisted manual modes to fully automatic modes including automatic landing. A high-level system overview as well as a description of each software module comprising the system is provided. Digital systems diagrams are included for each major flight control component and selected flight management functions.

  17. Advanced bulk processing of lightweight materials for utilization in the transportation sector

    Science.gov (United States)

    Milner, Justin L.

    The overall objective of this research is to develop the microstructure of metallic lightweight materials via multiple advanced processing techniques with potentials for industrial utilization on a large scale to meet the demands of the aerospace and automotive sectors. This work focused on (i) refining the grain structure to increase the strength, (ii) controlling the texture to increase formability and (iii) directly reducing processing/production cost of lightweight material components. Advanced processing is conducted on a bulk scale by several severe plastic deformation techniques including: accumulative roll bonding, isolated shear rolling and friction stir processing to achieve the multiple targets of this research. Development and validation of the processing techniques is achieved through wide-ranging experiments along with detailed mechanical and microstructural examination of the processed material. On a broad level, this research will make advancements in processing of bulk lightweight materials facilitating industrial-scale implementation. Where accumulative roll bonding and isolated shear rolling, currently feasible on an industrial scale, processes bulk sheet materials capable of replacing more expensive grades of alloys and enabling low-temperature and high-strain-rate formability. Furthermore, friction stir processing to manufacture lightweight tubes, made from magnesium alloys, has the potential to increase the utilization of these materials in the automotive and aerospace sectors for high strength - high formability applications. With the increased utilization of these advanced processing techniques will significantly reduce the cost associated with lightweight materials for many applications in the transportation sectors.

  18. Status of advanced light-duty transportation technologies in the US

    International Nuclear Information System (INIS)

    The need to reduce oil consumption and greenhouse gases is driving a fundamental change toward more efficient, advanced vehicles, and fuels in the transportation sector. The paper reviews the current status of light duty vehicles in the US and discusses policies to improve fuel efficiency, advanced electric drives, and sustainable cellulosic biofuels. The paper describes the cost, technical, infrastructure, and market barriers for alternative technologies, i.e., advanced biofuels and light-duty vehicles, including diesel vehicles, natural-gas vehicles, hybrid electric vehicles, plug-in hybrid electric vehicles, and fuel-cell electric vehicles. The paper also presents R and D targets and technology validation programs of the US government. - Highlights: ► Summary of the current status of LDVs and fuels. ► Overview of government policies and incentives for advanced vehicles and fuels. ► Technical and infrastructure barriers for biofuels, PHEVs, and FCEVs. ► Cost targets and research challenges for batteries and fuel cells. ► Summary of near- to mid-term market considerations for vehicles and fuels.

  19. A coupled diffusion-transport computational method and its application for the determination of space dependent angular flux distributions at a cold neutron source

    International Nuclear Information System (INIS)

    A fast calculation program ''BRIDGE'' was developed for the calculation of a Cold Neutron Source (CNS) at a radial beam tube of the FRG-I reactor, which couples a total assembly diffusion calculation to a transport calculation for a certain subregion. For the coupling flux and current boundary values at the common surfaces are taken from the diffusion calculation and are used as driving conditions in the transport calculation. 'Equivalence Theorie' is used for the transport feedback effect on the diffusion calculation to improve the consistency of the boundary values. The optimization of a CNS for maximizing the subthermal flux in the wavelength range 4 - 6 A is discussed. (orig.)

  20. ULTRACLEAN FUELS PRODUCTION AND UTILIZATION FOR THE TWENTY-FIRST CENTURY: ADVANCES TOWARDS SUSTAINABLE TRANSPORTATION FUELS

    Energy Technology Data Exchange (ETDEWEB)

    Fox, E.

    2013-06-17

    Ultraclean fuels production has become increasingly important as a method to help decrease emissions and allow the introduction of alternative feed stocks for transportation fuels. Established methods, such as Fischer-Tropsch, have seen a resurgence of interest as natural gas prices drop and existing petroleum resources require more intensive clean-up and purification to meet stringent environmental standards. This review covers some of the advances in deep desulfurization, synthesis gas conversion into fuels and feed stocks that were presented at the 245th American Chemical Society Spring Annual Meeting in New Orleans, LA in the Division of Energy and Fuels symposium on "Ultraclean Fuels Production and Utilization".

  1. Ultraclean Fuels Production and Utilization for the Twenty-First Century: Advances toward Sustainable Transportation Fuels

    Energy Technology Data Exchange (ETDEWEB)

    Fox, Elise B.; Liu, Zhong-Wen; Liu, Zhao-Tie

    2013-11-21

    Ultraclean fuels production has become increasingly important as a method to help decrease emissions and allow the introduction of alternative feed stocks for transportation fuels. Established methods, such as Fischer-Tropsch, have seen a resurgence of interest as natural gas prices drop and existing petroleum resources require more intensive clean-up and purification to meet stringent environmental standards. This review covers some of the advances in deep desulfurization, synthesis gas conversion into fuels and feed stocks that were presented at the 245th American Chemical Society Spring Annual Meeting in New Orleans, LA in the Division of Energy and Fuels symposium on "Ultraclean Fuels Production and Utilization".

  2. The contribution to the energy balance and transport in an advanced-fuel tokamak reactor

    International Nuclear Information System (INIS)

    The influence of synchrotron radiation emission on the energy balance of an advanced-fuel (such as D-3He, or catalyzed-D) tokamak plasma is considered. It is shown that a region in the β-T space exists, where the fusion energy delivered to the plasma overcomes synchrotron and bremsstrahlung energy losses, and which could then allow for ignited operation. 1-Dimensional codes results are also presented, which illustrate the main features of radial transport in a ignited, D-3He tokamak plasma

  3. Materials research with neutron beams from a research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Root, J.; Banks, D. [Canadian Neutron Beam Centre, Chalk River Laboratories, Chalk River, Ontario (Canada)

    2015-03-15

    Because of the unique ways that neutrons interact with matter, neutron beams from a research reactor can reveal knowledge about materials that cannot be obtained as easily with other scientific methods. Neutron beams are suitable for imaging methods (radiography or tomography), for scattering methods (diffraction, spectroscopy, and reflectometry) and for other possibilities. Neutron-beam methods are applied by students and researchers from academia, industry and government to support their materials research programs in several disciplines: physics, chemistry, materials science and life science. The arising knowledge about materials has been applied to advance technologies that appear in everyday life: transportation, communication, energy, environment and health. This paper illustrates the broad spectrum of materials research with neutron beams, by presenting examples from the Canadian Neutron Beam Centre at the NRU research reactor in Chalk River. (author)

  4. The use of symbolic computation in radiative, energy, and neutron transport calculations. Final report

    International Nuclear Information System (INIS)

    This investigation used sysmbolic manipulation in developing analytical methods and general computational strategies for solving both linear and nonlinear, regular and singular integral and integro-differential equations which appear in radiative and mixed-mode energy transport. Contained in this report are seven papers which present the technical results as individual modules

  5. The use of symbolic computation in radiative, energy, and neutron transport calculations. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Frankel, J.I.

    1997-09-01

    This investigation used sysmbolic manipulation in developing analytical methods and general computational strategies for solving both linear and nonlinear, regular and singular integral and integro-differential equations which appear in radiative and mixed-mode energy transport. Contained in this report are seven papers which present the technical results as individual modules.

  6. Study on void fraction distribution in the moderator cell of Cold Neutron Source systems in China Advanced Research Reactor

    Science.gov (United States)

    Li, Liangxing; Li, Huixiong; Hu, Jinfeng; Bi, Qincheng; Chen, Tingkuan

    2007-04-01

    A physical model is developed for analyzing and evaluating the void fraction profiles in the moderator cell of the Cold Neutron Source (CNS) of the China Advanced Research Reactor (CARR), which is now constructing in the China Institute of Atomic Energy (CIAE). The results derived from the model are compared with the related experimental data and its propriety is verified. The model is then used to explore the influence of various factors, including the diameter of boiling vapor bubbles, liquid density, liquid viscosity and the total heating power acted on the moderator cell, on the void fraction profiles in the cell. The results calculated with the present model indicate that the void fraction in the moderator cell increases linearly with heating power, and increases with the liquid viscosity, but decreases as the size of bubbles increases, and increases linearly with heating power. For the case where hydrogen is being used as a moderator, calculation results show that the void fraction in the moderator cell may be less than 30%, which is the maximum void fraction permitted from the nuclear physics point of view. The model and the calculation results will help to obtain insight of the mechanism that controls the void fraction distribution in the moderator cell, and provide theoretical supports for the moderator cell design.

  7. Analyses of the reflector tank, cold source, and beam tube cooling for ANS reactor. [Advanced Neutron Source (ANS)

    Energy Technology Data Exchange (ETDEWEB)

    Marland, S. (Tennessee Univ., Knoxville, TN (United States))

    1992-07-01

    This report describes my work as an intern with Martin Marietta Energy Systems, Inc., in the summer of 1991. I was assigned to the Reactor Technology Engineering Department, working on the Advanced Neutron Source (ANS). My first project was to select and analyze sealing systems for the top of the diverter/reflector tank. This involved investigating various metal seals and calculating the forces necessary to maintain an adequate seal. The force calculations led to an analysis of several bolt patterns and lockring concepts that could be used to maintain a seal on the vessel. Another project involved some pressure vessel stress calculations and the calculation of the center of gravity for the cold source assembly. I also completed some sketches of possible cooling channel patterns for the inner vessel of the cold source. In addition, I worked on some thermal design analyses for the reflector tank and beam tubes, including heat transfer calculations and assisting in Patran and Pthermal analyses. To supplement the ANS work, I worked on other projects. I completed some stress/deflection analyses on several different beams. These analyses were done with the aid of CAASE, a beam-analysis software package. An additional project involved bending analysis on a carbon removal system. This study was done to find the deflection of a complex-shaped beam when loaded with a full waste can.

  8. Updated version of the DOT 4 one- and two-dimensional neutron/photon transport code

    Energy Technology Data Exchange (ETDEWEB)

    Rhoades, W.A.; Childs, R.L.

    1982-07-01

    DOT 4 is designed to allow very large transport problems to be solved on a wide range of computers and memory arrangements. Unusual flexibilty in both space-mesh and directional-quadrature specification is allowed. For example, the radial mesh in an R-Z problem can vary with axial position. The directional quadrature can vary with both space and energy group. Several features improve performance on both deep penetration and criticality problems. The program has been checked and used extensively.

  9. Updated version of the DOT 4 one- and two-dimensional neutron/photon transport code

    International Nuclear Information System (INIS)

    DOT 4 is designed to allow very large transport problems to be solved on a wide range of computers and memory arrangements. Unusual flexibilty in both space-mesh and directional-quadrature specification is allowed. For example, the radial mesh in an R-Z problem can vary with axial position. The directional quadrature can vary with both space and energy group. Several features improve performance on both deep penetration and criticality problems. The program has been checked and used extensively

  10. Advances in Dynamic Transport of Organic Contaminants in Karst Groundwater Systems

    Science.gov (United States)

    Padilla, I. Y.; Vesper, D.; Alshawabkeh, A.; Hellweger, F.

    2011-12-01

    Karst groundwater systems develop in soluble rocks such as limestone, and are characterized by high permeability and well-developed conduit porosity. These systems provide important freshwater resources for human consumption and ecological integrity of streams, wetlands, and coastal zones. The same characteristics that make karst aquifers highly productive make them highly vulnerable to contamination. As a result, karst aquifers serve as an important route for contaminants exposure to humans and wildlife. Transport of organic contaminants in karst ground-water occurs in complex pathways influenced by the flow mechanism predominating in the aquifer: conduit-flow dominated systems tend to convey solutes rapidly through the system to a discharge point without much attenuation; diffuse-flow systems, on the other hand, can cause significant solute retardation and slow movement. These two mechanisms represent end members of a wide spectrum of conditions found in karst areas, and often a combination of conduit- and diffuse-flow mechanisms is encountered, where both flow mechanisms can control the fate and transport of contaminants. This is the case in the carbonate aquifers of northern Puerto Rico. This work addresses advances made on the characterization of fate and transport processes in karst ground-water systems characterized by variable conduit and/or diffusion dominated flow under high- and low-flow conditions. It involves laboratory-scale physical modeling and field-scale sampling and historical analysis of contaminant distribution. Statistical analysis of solute transport in Geo-Hydrobed physical models shows the heterogeneous character of transport dynamics in karstic units, and its variability under different flow regimes. Field-work analysis of chlorinated volatile organic compounds and phthalates indicates a large capacity of the karst systems to store and transmit contaminants. This work is part of the program "Puerto Rico Testsite for Exploring Contamination

  11. New advances in the pathophysiology of intestinal ion transport and barrier function in diarrhea and the impact on therapy.

    Science.gov (United States)

    Hoque, Kazi Mirajul; Chakraborty, Subhra; Sheikh, Irshad Ali; Woodward, Owen M

    2012-06-01

    Diarrhea remains a continuous threat to human health worldwide. Scaling up the best practices for diarrhea prevention requires improved therapies. Diarrhea results from dysregulation of normal intestinal ion transport functions. Host-microbe contact is a key determinant of this response. Underlying mechanisms in the disease state are regulated by intracellular signals that modulate the activity of individual transport proteins responsible for ion transport and barrier function. Similarly, virulence factors of pathogens and their complex interaction with the host has shed light on the mechanism of enteric infection. Great advances in our understanding of the pathophysiologic mechanisms of epithelial transport, and host-microbe interaction have been made in recent years. Application of these new advances may represent strategies to decrease pathogen attachment, enhance intestinal cation absorption, decrease anion secretion and repair barrier function. This review highlights the new advances and better understanding in the pathophysiology of diarrheal diseases and their impact on therapy.

  12. Ripple transport in helical-axis advanced stellarators - a comparison with classical stellarator/torsatrons

    International Nuclear Information System (INIS)

    Calculations of the neoclassical transport rates due to particles trapped in the helical ripples of a stellarator's magnetic field are carried out, based on solutions of the bounce-averaged kinetic equation. These calculations employ a model for the magnetic field strength, B, which is an accurate approximation to the actual B for a wide variety of stellarator-type devices, among which are Helical-Axis Advanced Stellarators (Helias) as well as conventional stellarators and torsatrons. Comparisons are carried out in which it is shown that the Helias concept leads to significant reductions in neoclassical transport rates throughout the entire long-mean-free-path regime, with the reduction being particularly dramatic in the ν-1 regime. These findings are confirmed by numerical simulations. Further, it is shown that the behavior of deeply trapped particles in Helias can be fundamentally different from that in classical stellarator/torsatrons; as a consequence, the beneficial effects of a radial electric field on the transport make themselves felt at lower collision frequency than is usual. (orig.)

  13. 3D Monte-Carlo transport calculations of whole slab reactor cores: validation of deterministic neutronic calculation routes

    Energy Technology Data Exchange (ETDEWEB)

    Palau, J.M. [CEA Cadarache, Service de Physique des Reacteurs et du Cycle, Lab. de Projets Nucleaires, 13 - Saint-Paul-lez-Durance (France)

    2005-07-01

    This paper presents how Monte-Carlo calculations (French TRIPOLI4 poly-kinetic code with an appropriate pre-processing and post-processing software called OVNI) are used in the case of 3-dimensional heterogeneous benchmarks (slab reactor cores) to reduce model biases and enable a thorough and detailed analysis of the performances of deterministic methods and their associated data libraries with respect to key neutron parameters (reactivity, local power). Outstanding examples of application of these tools are presented regarding the new numerical methods implemented in the French lattice code APOLLO2 (advanced self-shielding models, new IDT characteristics method implemented within the discrete-ordinates flux solver model) and the JEFF3.1 nuclear data library (checked against JEF2.2 previous file). In particular we have pointed out, by performing multigroup/point-wise TRIPOLI4 (assembly and core) calculations, the efficiency (in terms of accuracy and computation time) of the new IDT method developed in APOLLO2. In addition, by performing 3-dimensional TRIPOLI4 calculations of the whole slab core (few millions of elementary volumes), the high quality of the new JEFF3.1 nuclear data files and revised evaluations (U{sup 235}, U{sup 238}, Hf) for reactivity prediction of slab cores critical experiments has been stressed. As a feedback of the whole validation process, improvements in terms of nuclear data (mainly Hf capture cross-sections) and numerical methods (advanced quadrature formulas accounting validation results, validation of new self-shielding models, parallelization) are suggested to improve even more the APOLLO2-CRONOS2 standard calculation route. (author)

  14. 3D Monte-Carlo transport calculations of whole slab reactor cores: validation of deterministic neutronic calculation routes

    International Nuclear Information System (INIS)

    This paper presents how Monte-Carlo calculations (French TRIPOLI4 poly-kinetic code with an appropriate pre-processing and post-processing software called OVNI) are used in the case of 3-dimensional heterogeneous benchmarks (slab reactor cores) to reduce model biases and enable a thorough and detailed analysis of the performances of deterministic methods and their associated data libraries with respect to key neutron parameters (reactivity, local power). Outstanding examples of application of these tools are presented regarding the new numerical methods implemented in the French lattice code APOLLO2 (advanced self-shielding models, new IDT characteristics method implemented within the discrete-ordinates flux solver model) and the JEFF3.1 nuclear data library (checked against JEF2.2 previous file). In particular we have pointed out, by performing multigroup/point-wise TRIPOLI4 (assembly and core) calculations, the efficiency (in terms of accuracy and computation time) of the new IDT method developed in APOLLO2. In addition, by performing 3-dimensional TRIPOLI4 calculations of the whole slab core (few millions of elementary volumes), the high quality of the new JEFF3.1 nuclear data files and revised evaluations (U235, U238, Hf) for reactivity prediction of slab cores critical experiments has been stressed. As a feedback of the whole validation process, improvements in terms of nuclear data (mainly Hf capture cross-sections) and numerical methods (advanced quadrature formulas accounting validation results, validation of new self-shielding models, parallelization) are suggested to improve even more the APOLLO2-CRONOS2 standard calculation route. (author)

  15. Report by the AERES on the unit: Unit of researches on neutron transport and radioactivity confinement in nuclear installations under the supervision of the establishments and bodies: IRSN

    International Nuclear Information System (INIS)

    This report is a kind of audit report on a research laboratory whose activity is organized in three departments: neutron transport and criticality (themes: numerical methods, maths and statistics related to the simulation of neutral particle propagation, nuclear data, uncertainty propagation and bias estimation, code qualification and associated experimental programs, neutron transport in reactors and fuel cycle, criticality accidents), radionuclide transfer in radioactive waste disposals (site identification strategy, hydro-mechanical phenomena affecting storage performance, physical-chemical evolution factors, storage modelling), and metrology and confinement of radioactive gases and aerosols. The authors discuss an assessment of the unit activities in terms of strengths and opportunities, aspects to be improved and recommendations, productions and publications. A more detailed assessment is presented for each department in terms of scientific quality, influence and attractiveness (awards, recruitment capacity, capacity to obtain financing and to tender, participation to international programs), strategy and governance, and project

  16. TN approximation on the critical size of time-dependent, one-speed and one-dimensional neutron transport problem with anisotropic scattering

    International Nuclear Information System (INIS)

    The criticality problem is studied based on one-speed time-dependent neutron transport theory, for a uniform and finite slab, using the Marshak boundary condition. The time-dependent neutron transport equation is reduced to a stationary equation. The variation of the critical thickness of the time-dependent system is investigated by using the linear anisotropic scattering kernel together with the combination of forward and backward scattering. Numerical calculations for various combinations of the scattering parameters and selected values of the time decay constant and the reflection coefficient are performed by using the Chebyshev polynomials approximation method. The results are compared with those previously obtained by other methods which are available in the literature.

  17. Accurate Least-Squares P$_N$ Scaling based on Problem Optical Thickness for solving Neutron Transport Problems

    CERN Document Server

    Zheng, Weixiong

    2016-01-01

    In this paper, we present an accurate and robust scaling operator based on material optical thickness (OT) for the least-squares spherical harmonics (LSP$_N$) method for solving neutron transport problems. LSP$_N$ without proper scaling is known to be erroneous in highly scattering medium, if the optical thickness of the material is large. A previously presented scaling developed by Manteuffel, et al.\\ does improve the accuracy of LSP$_N$, in problems where the material is optically thick. With the method, however, essentially no scaling is applied in optically thin materials, which can lead to an erroneous solution with presence of highly scattering medium. Another scaling approach, called the reciprocal-removal (RR) scaled LSP$_N$, which is equivalent to the self-adjoint angular flux (SAAF) equation, has numerical issues in highly-scattering materials due to a singular weighting. We propose a scaling based on optical thickness that improves the solution in optically thick media while avoiding the singularit...

  18. Transmission probability method based on triangle meshes for solving the unstructured geometry neutron transport problem

    Institute of Scientific and Technical Information of China (English)

    WU Hongchun; LIU Pingping; ZHOU Yongqiang; CAO Liangzhi

    2007-01-01

    The fuel assembly or core with unstructured geometry is frequently used in the advanced reactor. To calculate the fuel assembly, the transmission probability method (TPM) is widely used. However, the rectangular or hexagonal meshes are mainly used in the TPM codes for the normal core structure. The triangle meshes are most useful for expressing the complicated unstructured geometry. Even though the finite element method and Monte-Carlo methodare well suited for solving the unstructured geometry problem, they are very time-consuming. Therefore, a TPM code based on the triangle meshes is developed here. This code was applied to the hybrid fuel geometry, and compared with the results of the MCNP code and other codes. The results of the comparison were consistent with each other. The TPM with triangle meshes can thus be applied to the two-dimensional arbitrary fuel assembly.

  19. A High Temperature-Tolerant and Radiation-Resistant In-Core Neutron Sensor for Advanced Reactors. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Cao, Lei [The Ohio State Univ., Columbus, OH (United States); Miller, Don [The Ohio State Univ., Columbus, OH (United States)

    2015-01-23

    The objectives of this project are to develop a small and reliable gallium nitride (GaN) neutron sensor that is capable of withstanding high neutron fluence and high temperature, isolating gamma background, and operating in a wide dynamic range. The first objective will be the understanding of the fundamental materials properties and electronic response of a GaN semiconductor materials and device in an environment of high temperature and intense neutron field. To achieve such goal, an in-situ study of electronic properties of GaN device such as I-V, leakage current, and charge collection efficiency (CCE) in high temperature using an external neutron beam will be designed and implemented. We will also perform in-core irradiation of GaN up to the highest yet fast neutron fluence and an off-line performance evaluation.

  20. A new modelling of the multigroup scattering cross section in deterministic codes for neutron transport

    International Nuclear Information System (INIS)

    In reactor physics, calculation schemes with deterministic codes are validated with respect to a reference Monte Carlo code. The remaining biases are attributed to the approximations and models induced by the multigroup theory (self-shielding models and expansion of the scattering law using Legendre polynomials) to represent physical phenomena (resonant absorption and scattering anisotropy respectively). This work focuses on the relevance of a polynomial expansion to model the scattering law. Since the outset of reactor physics, the latter has been expanded on a truncated Legendre polynomial basis. However, the transfer cross sections are highly anisotropic, with non-zero values for a very small range of the cosine of the scattering angle. Besides, the finer the energy mesh and the lighter the scattering nucleus, the more exacerbated is the peaked shape of this cross section. As such, the Legendre expansion is less suited to represent the scattering law. Furthermore, this model induces negative values which are non-physical. In this work, various scattering laws are briefly described and the limitations of the existing model are pointed out. Hence, piecewise-constant functions have been used to represent the multigroup scattering cross section. This representation requires a different model for the diffusion source. The discrete ordinates method which is widely employed to solve the transport equation has been adapted. Thus, the finite volume method for angular discretization has been developed and implemented in Paris environment which hosts the Sn solver, Snatch. The angular finite volume method has been compared to the collocation method with Legendre moments to ensure its proper performance. Moreover, unlike the latter, this method is adapted for both the Legendre moments and the piecewise-constant functions representations of the scattering cross section. This hybrid-source method has been validated for different cases: fuel cell in infinite lattice

  1. Analysis and development of spatial hp-refinement methods for solving the neutron transport equation; Analyse et developpement de methodes de raffinement hp en espace pour l'equation de transport des neutrons

    Energy Technology Data Exchange (ETDEWEB)

    Fournier, D.

    2011-10-10

    The different neutronic parameters have to be calculated with a higher accuracy in order to design the 4. generation reactor cores. As memory storage and computation time are limited, adaptive methods are a solution to solve the neutron transport equation. The neutronic flux, solution of this equation, depends on the energy, angle and space. The different variables are successively discretized. The energy with a multigroup approach, considering the different quantities to be constant on each group, the angle by a collocation method called SN approximation. Once the energy and angle variable are discretized, a system of spatially-dependent hyperbolic equations has to be solved. Discontinuous finite elements are used to make possible the development of hp-refinement methods. Thus, the accuracy of the solution can be improved by spatial refinement (h-refinement), consisting into subdividing a cell into sub-cells, or by order refinement (p-refinement), by increasing the order of the polynomial basis. In this thesis, the properties of this methods are analyzed showing the importance of the regularity of the solution to choose the type of refinement. Thus, two error estimators are used to lead the refinement process. Whereas the first one requires high regularity hypothesis (analytical solution), the second one supposes only the minimal hypothesis required for the solution to exist. The comparison of both estimators is done on benchmarks where the analytic solution is known by the method of manufactured solutions. Thus, the behaviour of the solution as a regard of the regularity can be studied. It leads to a hp-refinement method using the two estimators. Then, a comparison is done with other existing methods on simplified but also realistic benchmarks coming from nuclear cores. These adaptive methods considerably reduces the computational cost and memory footprint. To further improve these two points, an approach with energy-dependent meshes is proposed. Actually, as the

  2. Transmutation Scenarios Impacts on Advanced Nuclear Cycles (fabrication/reprocessing/transportation)

    International Nuclear Information System (INIS)

    In the frame of the French Law for waste management, minor actinides transmutation scenarios have been studied for a sodium-cooled fast reactors fleet using homogeneous or heterogeneous recycling modes. Americium, neptunium and curium can be transmuted once included together in the standard MOX fuel, or the sole Americium can be incorporated in Am-bearing radial blanket. MAs transmutation in Accelerator Driven System has also been studied while Plutonium is recycling in SFR. Assessments and comparisons of these advanced cycles have been performed in light of technical and economic aspects criteria. The purpose of this study is to present the results in terms of impacts of the transmutation scenarios on fuel cycle plants (fabrication, reprocessing) and transportations taking into account thermal, radiation and criticality parameters. Comparison with no transmutation option is also presented. (author)

  3. Motion-base simulator results of advanced supersonic transport handling qualities with active controls

    Science.gov (United States)

    Feather, J. B.; Joshi, D. S.

    1981-01-01

    Handling qualities of the unaugmented advanced supersonic transport (AST) are deficient in the low-speed, landing approach regime. Consequently, improvement in handling with active control augmentation systems has been achieved using implicit model-following techniques. Extensive fixed-based simulator evaluations were used to validate these systems prior to tests with full motion and visual capabilities on a six-axis motion-base simulator (MBS). These tests compared the handling qualities of the unaugmented AST with several augmented configurations to ascertain the effectiveness of these systems. Cooper-Harper ratings, tracking errors, and control activity data from the MBS tests have been analyzed statistically. The results show the fully augmented AST handling qualities have been improved to an acceptable level.

  4. A Psychoacoustic Evaluation of Noise Signatures from Advanced Civil Transport Aircraft

    Science.gov (United States)

    Rizzi, Stephen A.; Christian, Andrew

    2016-01-01

    The NASA Environmentally Responsible Aviation project has been successful in developing and demonstrating technologies for integrated aircraft systems that can simultaneously meet aggressive goals for fuel burn, noise and emissions. Some of the resulting systems substantially differ from the familiar tube and wing designs constituting the current civil transport fleet. This study attempts to explore whether or not the effective perceived noise level metric used in the NASA noise goal accurately reflects human subject response across the range of vehicles considered. Further, it seeks to determine, in a quantitative manner, if the sounds associated with the advanced aircraft are more or less preferable to the reference vehicles beyond any differences revealed by the metric. These explorations are made through psychoacoustic tests in a controlled laboratory environment using simulated stimuli developed from auralizations of selected vehicles based on systems noise assessments.

  5. New developments in APSTNG neutron probe diagnostics

    Energy Technology Data Exchange (ETDEWEB)

    Rhodes, E.; Dickerman, C.E.

    1995-12-31

    The development and investigation of a small associated-particle sealed-tube neutron generator (APSTNG) show 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 separate the prompt and delayed gamma-ray spectra and 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. The gamma-ray dector and neutron generator can be located on the same side of the interrogated object, so spaces behind walls and other confirmed areas can be inspected. No collimators or radiation shielding are needed, the neutron generator is relatively simple and small, and commercial-grade electronics are employed. A complete system could be transported in an automotive van. Proof-of-concept laboratory experiments have been successfully performed for simulated nuclear, chemical warfare, and conventional munitions. Inspection applications have been investigated for presence of cocaine in propane tanks, uranium and plutonium smuggling, and radioactive and toxic waste characterization. An advanced APSTNG tube is being designed and constructed 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.

  6. Spatial homogenization methods for pin-by-pin neutron transport calculations

    Science.gov (United States)

    Kozlowski, Tomasz

    For practical reactor core applications low-order transport approximations such as SP3 have been shown to provide sufficient accuracy for both static and transient calculations with considerably less computational expense than the discrete ordinate or the full spherical harmonics methods. These methods have been applied in several core simulators where homogenization was performed at the level of the pin cell. One of the principal problems has been to recover the error introduced by pin-cell homogenization. Two basic approaches to treat pin-cell homogenization error have been proposed: Superhomogenization (SPH) factors and Pin-Cell Discontinuity Factors (PDF). These methods are based on well established Equivalence Theory and Generalized Equivalence Theory to generate appropriate group constants. These methods are able to treat all sources of error together, allowing even few-group diffusion with one mesh per cell to reproduce the reference solution. A detailed investigation and consistent comparison of both homogenization techniques showed potential of PDF approach to improve accuracy of core calculation, but also reveal its limitation. In principle, the method is applicable only for the boundary conditions at which it was created, i.e. for boundary conditions considered during the homogenization process---normally zero current. Therefore, there exists a need to improve this method, making it more general and environment independent. The goal of proposed general homogenization technique is to create a function that is able to correctly predict the appropriate correction factor with only homogeneous information available, i.e. a function based on heterogeneous solution that could approximate PDFs using homogeneous solution. It has been shown that the PDF can be well approximated by least-square polynomial fit of non-dimensional heterogeneous solution and later used for PDF prediction using homogeneous solution. This shows a promise for PDF prediction for off

  7. DANCE (Detector for Advanced Neutron Capture Experiments) is a 4π array of BaF2 crystals installed at LANSCE, Lujan Center. Neutron capture measurements on ^157Gd and ^89Y nuclei were conducted using this facility.

    Science.gov (United States)

    Chyzh, A.; Mitchell, G.; Vieira, D.; Bredeweg, T.; Ullmann, J.; Jandel, M.; Couture, A.; Keksis, A.; Rundberg, R.; Wilhelmy, J.; O'Donnell, J.; Baramsai, B.; Haight, R.; Wouters, J.; Krticka, M.; Parker, W.; Becker, J.; Agvaanlusan, U.

    2009-10-01

    DANCE (Detector for Advanced Neutron Capture Experiments) is a 4π array of BaF2 crystals installed at LANSCE, Lujan Center. Neutron capture measurements on ^157Gd and ^89Y nuclei were conducted using this facility. The absolute cross sections of the ^89Y(n,γ) reaction was measured for the first time ever in the neutron energy range of 10 eV -- 10 keV and improvements were made in the 10 -- 300 keV range. The error bars were significantly reduced and number of cross section points was increased since the past ^89Y(n,γ) experiments. The ^157Gd(n,γ) cross section was determined at En = 20 eV -- 300 keV by normalizing the experimental DANCE data to a well known resonance taken from the ENDF/B-VII library. Computer simulations of the ^157Gd(n,γ) cascades and DANCE pulse height function were made using DICEBOX and GEANT4 codes and simulated Esum and Eγ spectra are compared to the experimental DANCE data. Values of spin and photon strength function (PSF) of the ^157Gd(n,γ) resonances are provided in the range of En = 2 -- 300 eV using spin dependence upon a γ-ray multiplicity.

  8. Systematic and statistical errors in a bayesian approach to the estimation of the neutron-star equation of state using advanced gravitational wave detectors

    CERN Document Server

    Wade, Leslie; Ochsner, Evan; Lackey, Benjamin D; Farr, Benjamin F; Littenberg, Tyson B; Raymond, Vivien

    2014-01-01

    Advanced ground-based gravitational-wave detectors are capable of measuring tidal influences in binary neutron-star systems. In this work, we report on the statistical uncertainties in measuring tidal deformability with a full Bayesian parameter estimation implementation. We show how simultaneous measurements of chirp mass and tidal deformability can be used to constrain the neutron-star equation of state. We also study the effects of waveform modeling bias and individual instances of detector noise on these measurements. We notably find that systematic error between post-Newtonian waveform families can significantly bias the estimation of tidal parameters, thus motivating the continued development of waveform models that are more reliable at high frequencies.

  9. Application of particle transport code PHITS for design of J-PARC 1MW spallation neutron source and its validation

    International Nuclear Information System (INIS)

    Japan Spallation Neutron Source (JSNS) is one of major experimental facilities in Japan Proton Accelerator Research Complex (J-PARC). JSNS operated by 3-GeV and 1-MW pulsed proton beams has the highest class neutron intensity in the world. In the design stage, aiming to the best neutronic performance, the PHITS code was fully applied to JSNS neutronics designs and several thousands calculation cases were done with complicated models. Not only optimization of neutronic performance, but also shielding calculation, nuclear heat estimation for the engineering design, residual radioactivity estimation for the cask design and radiation damage estimation for the life and maintenance design were done with the PHITS code. JSNS is one of the first facilities in the world fully adapted such a simulation code to the neutronics design. In these calculations, note that the particle energy change from GeV to meV (12 decades) and neutron fluxes reduce by 10 decades or more. To confirm the reliability of these calculations, neutron spectral intensities were measured. As a result, the measured values were good agreement with the calculated values. We could confirm that the PHITS were reliable for such a design calculation. (author)

  10. Coupled multi-group neutron photon transport for the simulation of high-resolution gamma-ray spectroscopy applications

    Energy Technology Data Exchange (ETDEWEB)

    Burns, Kimberly A. [Georgia Inst. of Technology, Atlanta, GA (United States)

    2009-08-01

    The accurate and efficient simulation of coupled neutron-photon problems is necessary for several important radiation detection applications. Examples include the detection of nuclear threats concealed in cargo containers and prompt gamma neutron activation analysis for nondestructive determination of elemental composition of unknown samples.

  11. Advanced Penning-type ion source development and passive beam focusing techniques for an associated particle imaging neutron generator

    OpenAIRE

    Sy, Amy

    2013-01-01

    The use of accelerator-based neutron generators for non-destructive imaging and analysis in commercial and security applications is continuously under development, with improvements to available systems and combinations of available techniques revealing new capabilities for real-time elemental and isotopic analysis. The recent application of associated particle imaging (API) techniques for time- and directionally-tagged neutrons to induced fission and transmission imaging methods demonstrate...

  12. Fusion Neutron Flux Monitor for ITER

    Institute of Scientific and Technical Information of China (English)

    YANG Jinwei; YANG Qingwei; XIAO Gongshan; ZHANG Wei; SONG Xianying; LI Xu

    2008-01-01

    Neutron flux monitor (NFM) as an important diagnostic sub-system in ITER (international thermonuclear experimental reactor) provides a global neutron source intensity, fusion power and neutron flux in real time. Three types of neutron flux monitor assemblies with different sensitivities and shielding materials have been designed. Through MCNP (Mante-Carlo neutral particle transport code) calculations, this extended system of NFM can detect the neutron flux in a range of 104 n/(cm2·s) to 1014 n/(cm2·s). It is capable of providing accurate neutron yield measurements for all operational modes encountered in the ITER experiments including the in-situ calibration. Combining both the counting mode and Campbelling (MSV; Mean Square Voltage) mode in the signal processing units, the requirement of the dynamic range (107) for these NFMs and time resolution (1 ms) can be met. Based on a uncertainty analysis, the estimated absolute measurement accuracies of the total fusion neutron yield can reach the required 10% level in both the early stage of the DD-phase and the full power DT operation mode. In the advanced DD-phase, the absolute measurement accuracy would be better than 20%.

  13. An Advanced Integrated Diffusion/Transport Method for the Design, Analysis and Optimization of the Very-High-Temperature Reactors

    International Nuclear Information System (INIS)

    The main objective of this research is to develop an integrated diffusion/transport (IDT) method to substantially improve the accuracy of nodal diffusion methods for the design and analysis of Very High Temperature Reactors (VHTR). Because of the presence of control rods in the reflector regions in the Pebble Bed Reactor (PBR-VHTR), traditional nodal diffusion methods do not accurately model these regions, within which diffusion theory breaks down in the vicinity of high neutron absorption and steep flux gradients. The IDT method uses a local transport solver based on a new incident flux response expansion method in the controlled nodes. Diffusion theory is used in the rest of the core. This approach improves the accuracy of the core solution by generating transport solutions of controlled nodes while maintaining computational efficiency by using diffusion solutions in nodes where such a treatment is sufficient. The transport method is initially developed and coupled to the reformulated 3-D nodal diffusion model in the CYNOD code for PBR core design and fuel cycle analysis. This method is also extended to the prismatic VHTR. The new method accurately captures transport effects in highly heterogeneous regions with steep flux gradients. The calculations of these nodes with transport theory avoid errors associated with spatial homogenization commonly used in diffusion methods in reactor core simulators

  14. An Advanced Integrated Diffusion/Transport Method for the Design, Analysis and Optimization of the Very-High-Temperature Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Farzad Rahnema; Dingkang Zhang; Abderrafi Ougouag; Frederick Gleicher

    2011-04-04

    The main objective of this research is to develop an integrated diffusion/transport (IDT) method to substantially improve the accuracy of nodal diffusion methods for the design and analysis of Very High Temperature Reactors (VHTR). Because of the presence of control rods in the reflector regions in the Pebble Bed Reactor (PBR-VHTR), traditional nodal diffusion methods do not accurately model these regions, within which diffusion theory breaks down in the vicinity of high neutron absorption and steep flux gradients. The IDT method uses a local transport solver based on a new incident flux response expansion method in the controlled nodes. Diffusion theory is used in the rest of the core. This approach improves the accuracy of the core solution by generating transport solutions of controlled nodes while maintaining computational efficiency by using diffusion solutions in nodes where such a treatment is sufficient. The transport method is initially developed and coupled to the reformulated 3-D nodal diffusion model in the CYNOD code for PBR core design and fuel cycle analysis. This method is also extended to the prismatic VHTR. The new method accurately captures transport effects in highly heterogeneous regions with steep flux gradients. The calculations of these nodes with transport theory avoid errors associated with spatial homogenization commonly used in diffusion methods in reactor core simulators

  15. TIMEX: a time-dependent explicit discrete ordinates program for the solution of multigroup transport equations with delayed neutrons

    Energy Technology Data Exchange (ETDEWEB)

    Hill, T.R.; Reed, W.H.

    1976-01-01

    TIMEX solves the time-dependent, one-dimensional multigroup transport equation with delayed neutrons in plane, cylindrical, spherical, and two-angle plane geometries. Both regular and adjoint, inhomogeneous and homogeneous problems subject to vacuum, reflective, periodic, white, albedo or inhomogeneous boundary flux conditions are solved. General anisotropic scattering is allowed and anisotropic inhomogeneous sources are permitted. The discrete ordinates approximation for the angular variable is used with the diamond (central) difference approximation for the angular extrapolation in curved geometries. A linear discontinuous finite element representation for the angular flux in each spatial mesh cell is used. The time variable is differenced by an explicit technique that is unconditionally stable so that arbitrarily large time steps can be taken. Because no iteration is performed the method is exceptionally fast in terms of computing time per time step. Two acceleration methods, exponential extrapolation and rebalance, are utilized to improve the accuracy of the time differencing scheme. Variable dimensioning is used so that any combination of problem parameters leading to a container array less than MAXCOR can be accommodated. The running time for TIMEX is highly problem-dependent, but varies almost linearly with the total number of unknowns and time steps. Provision is made for creation of standard interface output files for angular fluxes and angle-integrated fluxes. Five interface units (use of interface units is optional), five output units, and two system input/output units are required. A large bulk memory is desirable, but may be replaced by disk, drum, or tape storage. 13 tables, 9 figures. (auth)

  16. The Advanced Re-Entry Vehicle (ARV) a Development Step from ATV Toward Manned Transportation Systems

    Science.gov (United States)

    Bottacini, M.; Berthe, P.; Vo, X.; Pietsch, K.

    2011-08-01

    The Advanced Re-entry Vehicle (ARV) programme has been undertaken by Europe with the objective to contribute to the preparation of a future European crew transportation system, while providing a valuable logistic support to the ISS through an operational cargo return system. This development would allow: - the early acquisition of critical technologies; - the design, development and testing of elements suitable for the follow up human rated transportation system. These vehicles should also serve future LEO infrastructures and exploration missions. With the aim to satisfy the above objectives a team composed by major European industries and led by EADS Astrium Space Transportation is currently conducting the phase A of the programme under contract with the European Space Agency (ESA). Two vehicle versions are being investigated: a Cargo version, transporting cargo only to/from the ISS, and a Crew version, which will allow the transfer of both crew and cargo to/from the ISS. The ARV Cargo version, in its present configuration, is composed of three modules. The Versatile Service Module (VSM) provides to the system the propulsion/GNC for orbital manoeuvres and attitude control and the orbital power generation. Its propulsion system and GNC shall be robust enough to allow its use for different launch stacks and different LEO missions in the future. The Un-pressurised Cargo Module (UCM) provides the accommodation for about 3000 kg of un-pressurised cargo and is to be sufficiently flexible to ensure the transportation of: - orbital infrastructure components (ORU's); - scientific / technological experiments; - propellant for re-fuelling, re-boost (and deorbiting) of the ISS. The Re-entry Module (RM) provides a pressurized volume to accommodate active/passive cargo (2000 kg upload/1500 kg download). It is conceived as an expendable conical capsule with spherical heat- hield, interfacing with the new docking standard of the ISS, i.e. it carries the IBDM docking system, on a

  17. Development and application of neutron transport methods and uncertainty analyses for reactor core calculations. Technical report; Entwicklung und Einsatz von Neutronentransportmethoden und Unsicherheitsanalysen fuer Reaktorkernberechnungen. Technischer Bericht

    Energy Technology Data Exchange (ETDEWEB)

    Zwermann, W.; Aures, A.; Bernnat, W.; and others

    2013-06-15

    This report documents the status of the research and development goals reached within the reactor safety research project RS1503 ''Development and Application of Neutron Transport Methods and Uncertainty Analyses for Reactor Core Calculations'' as of the 1{sup st} quarter of 2013. The superordinate goal of the project is the development, validation, and application of neutron transport methods and uncertainty analyses for reactor core calculations. These calculation methods will mainly be applied to problems related to the core behaviour of light water reactors and innovative reactor concepts. The contributions of this project towards achieving this goal are the further development, validation, and application of deterministic and stochastic calculation programmes and of methods for uncertainty and sensitivity analyses, as well as the assessment of artificial neutral networks, for providing a complete nuclear calculation chain. This comprises processing nuclear basis data, creating multi-group data for diffusion and transport codes, obtaining reference solutions for stationary states with Monte Carlo codes, performing coupled 3D full core analyses in diffusion approximation and with other deterministic and also Monte Carlo transport codes, and implementing uncertainty and sensitivity analyses with the aim of propagating uncertainties through the whole calculation chain from fuel assembly, spectral and depletion calculations to coupled transient analyses. This calculation chain shall be applicable to light water reactors and also to innovative reactor concepts, and therefore has to be extensively validated with the help of benchmarks and critical experiments.

  18. Development and characterization of a high yield transportable pulsed neutron source with efficient and compact pulsed power system

    Science.gov (United States)

    Verma, Rishi; Mishra, Ekansh; Dhang, Prosenjit; Sagar, Karuna; Meena, Manraj; Shyam, Anurag

    2016-09-01

    The results of characterization experiments carried out on a newly developed dense plasma focus device based intense pulsed neutron source with efficient and compact pulsed power system are reported. Its high current sealed pseudospark switch based low inductance capacitor bank with maximum stored energy of ˜10 kJ is segregated into four modules of ˜2.5 kJ each and it cumulatively delivers peak current in the range of 400 kA-600 kA (corresponding to charging voltage range of 14 kV-18 kV) in a quarter time period of ˜2 μs. The neutron yield performance of this device has been optimized by discretely varying deuterium filling gas pressure in the range of 6 mbar-11 mbar at ˜17 kV/550 kA discharge. At ˜7 kJ/8.5 mbar operation, the average neutron yield has been measured to be in the order of ˜4 × 109 neutrons/pulse which is the highest ever reported neutron yield from a plasma focus device with the same stored energy. The average forward to radial anisotropy in neutron yield is found to be ˜2. The entire system is contained on a moveable trolley having dimensions 1.5 m × 1 m × 0.7 m and its operation and control (up to the distance of 25 m) are facilitated through optically isolated handheld remote console. The overall compactness of this system provides minimum proximity to small as well as large samples for irradiation. The major intended application objective of this high neutron yield dense plasma focus device development is to explore the feasibility of active neutron interrogation experiments by utilization of intense pulsed neutron sources.

  19. Advanced Nodal P3/SP3 Axial Transport Solvers for the MPACT 2D/1D Scheme

    Energy Technology Data Exchange (ETDEWEB)

    Stimpson, Shane G [ORNL; Collins, Benjamin S [ORNL

    2015-01-01

    As part of its initiative to provide multiphysics simulations of nuclear reactor cores, the Consortium for Advanced Simulation of Light Water Reactors (CASL) is developing the Virtual Environment for Reactor Applications Core Simulator (VERA-CS). The MPACT code, which is the primary neutron transport solver of VERA-CS, employs the two-dimensional/one-dimensional (2D/1D) method to solve 3-dimensional neutron transport problems and provide sub-pin-level resolution of the power distribution. While 2D method of characteristics is used to solve for the transport effects within each plane, 1D-nodal methods are used axially. There have been extensive studies of the 2D/1D method with a variety nodal methods, and the P3/SP3 solver has proved to be an effective method of providing higher-fidelity solutions while maintaining a low computational burden.The current implementation in MPACT wraps a one-node nodal expansion method (NEM) kernel for each moment, iterating between them and performing multiple sweeps to resolve flux distributions. However, it has been observed that this approach is more sensitive to convergence problems. This paper documents the theory and application two new nodal P3/SP3 approaches to be used within the 2D/1D method in MPACT. These two approaches aim to provide enhanced stability compared with the pre-existing one-node approach. Results from the HY-NEM-SP3 solver show that the accuracy is consistent with the one-node formulations and provides improved convergence for some problems; but the solver has issues with cases in thin planes. Although the 2N-SENM-SP3 solver is still under development, it is intended to resolve the issues with HY-NEM-SP3 but it will incur some additional computational burden by necessitating an additional 1D-CMFD-P3 solver to generate the second moment cell-averaged scalar flux.

  20. Neutron analysis of spent fuel storage installation using parallel computing and advance discrete ordinates and Monte Carlo techniques.

    Science.gov (United States)

    Shedlock, Daniel; Haghighat, Alireza

    2005-01-01

    In the United States, the Nuclear Waste Policy Act of 1982 mandated centralised storage of spent nuclear fuel by 1988. However, the Yucca Mountain project is currently scheduled to start accepting spent nuclear fuel in 2010. Since many nuclear power plants were only designed for -10 y of spent fuel pool storage, > 35 plants have been forced into alternate means of spent fuel storage. In order to continue operation and make room in spent fuel pools, nuclear generators are turning towards independent spent fuel storage installations (ISFSIs). Typical vertical concrete ISFSIs are -6.1 m high and 3.3 m in diameter. The inherently large system, and the presence of thick concrete shields result in difficulties for both Monte Carlo (MC) and discrete ordinates (SN) calculations. MC calculations require significant variance reduction and multiple runs to obtain a detailed dose distribution. SN models need a large number of spatial meshes to accurately model the geometry and high quadrature orders to reduce ray effects, therefore, requiring significant amounts of computer memory and time. The use of various differencing schemes is needed to account for radial heterogeneity in material cross sections and densities. Two P3, S12, discrete ordinate, PENTRAN (parallel environment neutral-particle TRANsport) models were analysed and different MC models compared. A multigroup MCNP model was developed for direct comparison to the SN models. The biased A3MCNP (automated adjoint accelerated MCNP) and unbiased (MCNP) continuous energy MC models were developed to assess the adequacy of the CASK multigroup (22 neutron, 18 gamma) cross sections. The PENTRAN SN results are in close agreement (5%) with the multigroup MC results; however, they differ by -20-30% from the continuous-energy MC predictions. This large difference can be attributed to the expected difference between multigroup and continuous energy cross sections, and the fact that the CASK library is based on the old ENDF