Benchmark field study of deep neutron penetration
Morgan, J. F.; Sale, K.; Gold, R.; Roberts, J. H.; Preston, C. C.
1991-06-01
A unique benchmark neutron field has been established at the Lawrence Livermore National Laboratory (LLNL) to study deep penetration neutron transport. At LLNL, a tandem accelerator is used to generate a monoenergetic neutron source that permits investigation of deep neutron penetration under conditions that are virtually ideal to model, namely the transport of mono-energetic neutrons through a single material in a simple geometry. General features of the Lawrence Tandem (LATAN) benchmark field are described with emphasis on neutron source characteristics and room return background. The single material chosen for the first benchmark, LATAN-1, is a steel representative of Light Water Reactor (LWR) Pressure Vessels (PV). Also included is a brief description of the Little Boy replica, a critical reactor assembly designed to mimic the radiation doses from the atomic bomb dropped on Hiroshima, and its us in neutron spectrometry.
Use of Neutron Benchmark Fields for the Validation of Dosimetry Cross Sections
Griffin, Patrick
2016-02-01
The evolution of validation metrics for dosimetry cross sections in neutron benchmark fields is explored. The strength of some of the metrics in providing validation evidence is examined by applying them to the 252Cf spontaneous fission standard neutron benchmark field, the 235U thermal neutron fission reference benchmark field, the ACRR pool-type reactor central cavity reference benchmark fields, and the SPR-III fast burst reactor central cavity. The IRDFF dosimetry cross section library is used in the validation study and observations are made on the amount of coverage provided to the library contents by validation data available in these benchmark fields.
Standard Guide for Benchmark Testing of Reactor Dosimetry in Standard and Reference Neutron Fields
American Society for Testing and Materials. Philadelphia
2010-01-01
Return to Contents page 1.1 This guide covers facilities and procedures for benchmarking neutron measurements and calculations. Particular sections of the guide discuss: the use of well-characterized benchmark neutron fields to calibrate integral neutron sensors; the use of certified-neutron-fluence standards to calibrate radiometric counting equipment or to determine interlaboratory measurement consistency; development of special benchmark fields to test neutron transport calculations; use of well-known fission spectra to benchmark spectrum-averaged cross sections; and the use of benchmarked data and calculations to determine the uncertainties in derived neutron dosimetry results. 1.2 The values stated in SI units are to be regarded as standard. No other units of measurement are included in this standard.
Neutron Reference Benchmark Field Specification: ACRR Free-Field Environment (ACRR-FF-CC-32-CL).
Vega, Richard Manuel [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Parma, Edward J. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Griffin, Patrick J. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Vehar, David W. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)
2015-07-01
This report was put together to support the International Atomic Energy Agency (IAEA) REAL- 2016 activity to validate the dosimetry community’s ability to use a consistent set of activation data and to derive consistent spectral characterizations. The report captures details of integral measurements taken in the Annular Core Research Reactor (ACRR) central cavity free-field reference neutron benchmark field. The field is described and an “a priori” calculated neutron spectrum is reported, based on MCNP6 calculations, and a subject matter expert (SME) based covariance matrix is given for this “a priori” spectrum. The results of 31 integral dosimetry measurements in the neutron field are reported.
Over 50 neutron benchmark calculations have recently been completed as part of an ongoing program to validate the MCNP Monte Carlo radiation transport code. The new and significant aspects of this work are as follows: These calculations are the first attempt at a validation program for MCNP and the first official benchmarking of version 4 of the code. We believe the chosen set of benchmarks is a comprehensive set that may be useful for benchmarking other radiation transport codes and data libraries. These calculations provide insight into how well neutron transport calculations can be expected to model a wide variety of problems
More than 50 neutron benchmark calculations have recently been completed as part of an ongoing program to validate the MCNP Monte Carlo radiation transport code. The benchmark calculations reported here are part of an ongoing multiyear, multiperson effort to benchmark version 4 of the MCNP code. The MCNP is a Monte Carlo three-dimensional general-purpose, continuous-energy neutron, photon, and electron transport code. It is used around the world for many applications including aerospace, oil-well logging, physics experiments, criticality safety, reactor analysis, medical imaging, defense applications, accelerator design, radiation hardening, radiation shielding, health physics, fusion research, and education. The first phase of the benchmark project consisted of analytic and photon problems. The second phase consists of the ENDF/B-V neutron problems reported in this paper and in more detail in the comprehensive report. A cooperative program being carried out a General Electric, San Jose, consists of light water reactor benchmark problems. A subsequent phase focusing on electron problems is planned
Vega, Richard Manuel [Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States); Parm, Edward J. [Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States); Griffin, Patrick J. [Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States); Vehar, David W. [Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States)
2015-07-01
This report was put together to support the International Atomic Energy Agency (IAEA) REAL- 2016 activity to validate the dosimetry community’s ability to use a consistent set of activation data and to derive consistent spectral characterizations. The report captures details of integral measurements taken in the Annular Core Research Reactor (ACRR) central cavity with the Polyethylene-Lead-Graphite (PLG) bucket, reference neutron benchmark field. The field is described and an “a priori” calculated neutron spectrum is reported, based on MCNP6 calculations, and a subject matter expert (SME) based covariance matrix is given for this “a priori” spectrum. The results of 37 integral dosimetry measurements in the neutron field are reported.
Measurements of integral cross section ratios in two dosimetry benchmark neutron fields
In the frame of a current interlaboratory effort devoted to the standardization of fuels and materials neutron dosimetry, the 103Rh(n,n')103mRh and 58Ni(n,p)58Co integral cross sections have been accurately measured relatively to the 115In(n,n')115m In cross section in the 235U thermal fission neutron spectrum and in the MOL-ΣΣ intermediate-energy standard neutron field. In this last neutron field, the data are related also to the 235U(n,f) cross section. The measurements are extensively documented and the results briefly compared to literature. Most noticeably, decisive support is provided for the selection of a specific 103Rh(n,n')103mRh differential-energy cross section among the existing, conflicting data. (author)
Benchmarking of activation reaction distribution in an intermediate energy neutron field.
Ogawa, Tatsuhiko; Morev, Mikhail N; Hirota, Masahiro; Abe, Takuya; Koike, Yuya; Iwai, Satoshi; Iimoto, Takeshi; Kosako, Toshiso
2011-07-01
Neutron-induced reaction rate depth profiles inside concrete shield irradiated by intermediate energy neutron were calculated using a Monte-Carlo code and compared with an experiment. An irradiation field of intermediate neutron produced in the forward direction from a thick (stopping length) target bombarded by 400 MeV nucleon(-1) carbon ions was arranged at the heavy ion medical accelerator in Chiba. Ordinary concrete shield of 90 cm thickness was installed 50 cm downstream the iron target. Activation detectors of aluminum, gold and gold covered with cadmium were inserted at various depths. Irradiated samples were extracted after exposure and gamma-ray spectrometry was performed for each sample. Comparison of experimental and calculated shows good agreement for both low- and high-energy neutron-induced reaction except for (27)Al(n,X)(24)Na reaction at the surface. PMID:21515619
3-D neutron transport benchmarks
A set of 3-D neutron transport benchmark problems proposed by the Osaka University to NEACRP in 1988 has been calculated by many participants and the corresponding results are summarized in this report. The results of Keff, control rod worth and region-averaged fluxes for the four proposed core models, calculated by using various 3-D transport codes are compared and discussed. The calculational methods used were: Monte Carlo, Discrete Ordinates (Sn), Spherical Harmonics (Pn), Nodal Transport and others. The solutions of the four core models are quite useful as benchmarks for checking the validity of 3-D neutron transport codes
Prismatic VHTR neutronic benchmark problems
Connolly, Kevin John, E-mail: connolly@gatech.edu [Nuclear and Radiological Engineering and Medical Physics Programs, George W. Woodruff School, Georgia Institute of Technology, Atlanta, GA (United States); Rahnema, Farzad, E-mail: farzad@gatech.edu [Nuclear and Radiological Engineering and Medical Physics Programs, George W. Woodruff School, Georgia Institute of Technology, Atlanta, GA (United States); Tsvetkov, Pavel V. [Department of Nuclear Engineering, Texas A& M University, College Station, TX (United States)
2015-04-15
Highlights: • High temperature gas-cooled reactor neutronics benchmark problems. • Description of a whole prismatic VHTR core in its full heterogeneity. • Modeled using continuous energy nuclear data at a representative hot operating temperature. • Benchmark results for core eigenvalue, block-averaged power, and some selected pin fission density results. - Abstract: This paper aims to fill an apparent scarcity of benchmarks based on high temperature gas-cooled reactors. Within is a description of a whole prismatic VHTR core in its full heterogeneity and modeling using continuous energy nuclear data at a representative hot operating temperature. Also included is a core which has been simplified for ease in modeling while attempting to preserve as faithfully as possible the neutron physics of the core. Fuel and absorber pins have been homogenized from the particle level, however, the blocks which construct the core remain strongly heterogeneous. A six group multigroup (discrete energy) cross section set has been developed via Monte Carlo using the original heterogeneous core as a basis. Several configurations of the core have been solved using these two cross section sets; eigenvalue results, block-averaged power results, and some selected pin fission density results are presented in this paper, along with the six-group cross section data, so that method developers may use these problems as a standard reference point.
Prismatic VHTR neutronic benchmark problems
Highlights: • High temperature gas-cooled reactor neutronics benchmark problems. • Description of a whole prismatic VHTR core in its full heterogeneity. • Modeled using continuous energy nuclear data at a representative hot operating temperature. • Benchmark results for core eigenvalue, block-averaged power, and some selected pin fission density results. - Abstract: This paper aims to fill an apparent scarcity of benchmarks based on high temperature gas-cooled reactors. Within is a description of a whole prismatic VHTR core in its full heterogeneity and modeling using continuous energy nuclear data at a representative hot operating temperature. Also included is a core which has been simplified for ease in modeling while attempting to preserve as faithfully as possible the neutron physics of the core. Fuel and absorber pins have been homogenized from the particle level, however, the blocks which construct the core remain strongly heterogeneous. A six group multigroup (discrete energy) cross section set has been developed via Monte Carlo using the original heterogeneous core as a basis. Several configurations of the core have been solved using these two cross section sets; eigenvalue results, block-averaged power results, and some selected pin fission density results are presented in this paper, along with the six-group cross section data, so that method developers may use these problems as a standard reference point
Vega, Richard Manuel [Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States); Parma, Edward J. [Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States); Griffin, Patrick J. [Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States); Vehar, David W. [Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States)
2015-07-01
This report was put together to support the International Atomic Energy Agency (IAEA) REAL- 2016 activity to validate the dosimetry community’s ability to use a consistent set of activation data and to derive consistent spectral characterizations. The report captures details of integral measurements taken in the Annular Core Research Reactor (ACRR) central cavity with the 44 inch Lead-Boron (LB44) bucket, reference neutron benchmark field. The field is described and an “a priori” calculated neutron spectrum is reported, based on MCNP6 calculations, and a subject matter expert (SME) based covariance matrix is given for this “a priori” spectrum. The results of 31 integral dosimetry measurements in the neutron field are reported.
AGENT code - neutron transport benchmark examples
The paper focuses on description of representative benchmark problems to demonstrate the versatility and accuracy of the AGENT (Arbitrary Geometry Neutron Transport) code. AGENT couples the method of characteristics and R-functions allowing true modeling of complex geometries. AGENT is optimized for robustness, accuracy, and computational efficiency for 2-D assembly configurations. The robustness of R-function based geometry generator is achieved through the hierarchical union of the simple primitives into more complex shapes. The accuracy is comparable to Monte Carlo codes and is obtained by following neutron propagation through true geometries. The computational efficiency is maintained through a set of acceleration techniques introduced in all important calculation levels. The selected assembly benchmark problems discussed in this paper are: the complex hexagonal modular high-temperature gas-cooled reactor, the Purdue University reactor and the well known C5G7 benchmark model. (author)
Fast neutron benchmark proposal at TRIGA-ACPR Reactor
The development of fast neutron benchmarks is a historical aim of reactor physics. The dry experimental tube situated in the central region of the core in TRIGA Annular-Core Pulsing Reactor (ACPR) offers a suitable neutron source for fast neutron benchmark development. Our proposal consists in mounting a high-enriched uranium annular converter into the dry channel of the core. Preliminary computations and measurements are presented in this paper. Neutron flux computations in the dry channel and the uranium converter were performed using MCNP and WIMS codes. Also neutron flux spectrum measurements and fast and thermal neutron flux distribution measurements were performed using foil activation techniques. (authors)
Benchmarking of neutron production of heavy-ion transport codes
Document available in abstract form only, full text of document follows: Accurate prediction of radiation fields generated by heavy ion interactions is important in medical applications, space missions, and in design and operation of rare isotope research facilities. In recent years, several well-established computer codes in widespread use for particle and radiation transport calculations have been equipped with the capability to simulate heavy ion transport and interactions. To assess and validate these capabilities, we performed simulations of a series of benchmark-quality heavy ion experiments with the computer codes FLUKA, MARS15, MCNPX, and PHITS. We focus on the comparisons of secondary neutron production. Results are encouraging; however, further improvements in models and codes and additional benchmarking are required. (authors)
Design of Pre-collimator System for Neutronics Benchmark Experiment
无
2011-01-01
In order to carry out evaluation of neutron nuclear data, in the last "Five-Year" period, China Institute of Atomic Energy has developed a set of neutron nuclear data benchmarking test system, and used the time-of-flight technique to measure the neutron
Measurement of neutron flux spectra in a tungsten benchmark by neutron foil activation method
The nuclear designs of fusion devices such as ITER (international thermonuclear experimental reactor), which is an experimental fusion reactor based on the ''tokamak'' concept, rely on the results of neutron physical calculations. These depend on the knowledge of the neutron and photon flux spectra which is particularly important because it permits to anticipate the possible answers of the whole structure to phenomena such as nuclear heating, tritium breeding, atomic displacements, radiation shielding, power generation and material activation. The flux spectra can be calculated with transport codes, but validating measurements are also required. An important constituent of structural materials and divertor areas of fusion reactors is tungsten. This thesis deals with the measurement of the neutron fluence and neutron energy spectrum in a tungsten assembly by means of multiple foil neutron activation technique. In order to check and qualify the experimental tools and the codes to be used in the tungsten benchmark experiment, test measurements in the D-T and D-D neutron fields of the neutron generator at Technische Universitaet Dresden were performed. The characteristics of the D-D and D-T reactions, used to produce monoenergetic neutrons, together with the selection of activation reactions suitable for fusion applications and details of the activation measurements are presented. Corrections related to the neutron irradiation process and those to the sample counting process are discussed, too. The neutron fluence and its energy distribution in a tungsten benchmark, irradiated at the frascati neutron generator with 14 MeV neutrons produced by the T(d,n)4He reaction, are then derived from the measurements of the neutron induced γ-ray activity in the foils using the STAYNL unfolding code, based on the linear least-squares-errors method, together with the IRDF-90.2 (international reactor dosimetry file) cross section library. The differences between the neutron flux
Collection of experimental data for fusion neutronics benchmark
During the recent ten years or more, many benchmark experiments for fusion neutronics have been carried out at two principal D-T neutron sources, FNS at JAERI and OKTAVIAN at Osaka University, and precious experimental data have been accumulated. Under an activity of Fusion Reactor Physics Subcommittee of Reactor Physics Committee, these experimental data are compiled in this report. (author)
POLCA-T Neutron Kinetics Model Benchmarking
Kotchoubey, Jurij
2015-01-01
The demand for computational tools that are capable to reliably predict the behavior of a nuclear reactor core in a variety of static and dynamic conditions does inevitably require a proper qualification of these tools for the intended purposes. One of the qualification methods is the verification of the code in question. Hereby, the correct implementation of the applied model as well as its flawless implementation in the code are scrutinized. The present work concerns with benchmarking as a ...
Computational benchmarking of fast neutron transport throughout large water thicknesses
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)
Bulk shielding benchmark experiment at Frascati neutron generator (FNG)
Batistoni, P.; Angelone, M.; Martone, M.; Pillon, M.; Rado, V. [Associazione EURATOM-ENEA sulla Fusione, Frascati (Italy); Santamarina, A.; Abidi, I.; Gastaldi, B.; Martini, M.; Marquette, J.P. [CEA Centre d`Etudes de Cadarache, 13 - Saint-Paul-lez-Durance (France)
1994-11-01
In the framework of the European Fusion Technology Program, ENEA (Italian Agency for New Technologies, Energy and the Environment) - Frascati and CEA (Commissariat a` l`Energie Atomique) - Cadarache, in collaboration performed a bulk shielding benchmark experiment, using the 14-MeV Frascati neutron generator (FNG), aimed at obtaining accurate experimental data for improving the nuclear data base and methods used in shielding designs. The experiment consisted of the irradiation of a stainless steel block by 14-MeV neutrons. The experimental results have been compared with numerical results calculated using both Sn and Monte Carlo transport codes and the cross section library EFF.1 (european fusion file).
Preliminary Neutronics Results for the OECD MHTGR-350 Benchmark
The benchmark problem is based on the MHTGR-350 reactor designed by General Atomics (GA). Phase I of the problem has three steady state exercises : Exercise 1 for neutronics stand alone with fixed cross-sections, Exercise 2 for thermal-fluids stand alone and Exercise 3 for coupled steady state. Phase II is defined for coupled transient cases. Phase III is defined to test the depletion capabilities of lattice physics codes. Phase III has two exercises : Exercise 1 for cold state and Exercise 2 for hot state. In this paper, a preliminary results for Exercise 1 of Phase I obtained by using CAPP code and the results for Phase III by McCARD code are presented. In this paper, some preliminary neutronics results for the OECD/NEA MHTGR-350 neutronics/thermal fluids coupled benchmark problem were presented and some of the global parameters for Phase I Exercise 1 were compared with those presented by INL research group. They showed a good agreement with each other. The results for Phase III were also reasonable. The benchmark is ongoing and more comparisons with the results of other research groups will be made as soon as they are available
Benchmark experiment on vanadium assembly with D-T neutrons. Leakage neutron spectrum measurement
Kokooo; Murata, I.; Nakano, D.; Takahashi, A. [Osaka Univ., Suita (Japan); Maekawa, F.; Ikeda, Y.
1998-03-01
The fusion neutronics benchmark experiments have been done for vanadium and vanadium alloy by using the slab assembly and time-of-flight (TOF) method. The leakage neutron spectra were measured from 50 keV to 15 MeV and comparison were done with MCNP-4A calculations which was made by using evaluated nuclear data of JENDL-3.2, JENDL-Fusion File and FENDL/E-1.0. (author)
Benchmarking of the FENDL-3 Neutron Cross-section Data Starter Library for Fusion Applications
Fischer, U., E-mail: ulrich.fischer@kit.edu [Association KIT-Euratom, Karlsruhe Institute of Technology, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Angelone, M. [Associazione ENEA-Euratom, ENEA Fusion Division, Via E. Fermi 27, I-00044 Frascati (Italy); Bohm, T. [University of Wisconsin-Madison, 1500 Engineering Dr, Madison, WI 53706 (United States); Kondo, K. [Association KIT-Euratom, Karlsruhe Institute of Technology, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Konno, C. [Japan Atomic Energy Agency, Tokai-mura, Naka-gun, Ibaraki-ken 319-1195 (Japan); Sawan, M. [University of Wisconsin-Madison, 1500 Engineering Dr, Madison, WI 53706 (United States); Villari, R. [Associazione ENEA-Euratom, ENEA Fusion Division, Via E. Fermi 27, I-00044 Frascati (Italy); Walker, B. [University of Wisconsin-Madison, 1500 Engineering Dr, Madison, WI 53706 (United States)
2014-06-15
This paper summarizes the benchmark analyses performed in a joint effort of ENEA (Italy), JAEA (Japan), KIT (Germany), and the University of Wisconsin (USA) on a computational ITER benchmark and a series of 14 MeV neutron benchmark experiments. The computational benchmark revealed a modest increase of the neutron flux levels in the deep penetration regions and a substantial increase of the gas production in steel components. The comparison to experimental results showed good agreement with no substantial differences between FENDL-3.0 and FENDL-2.1 for most of the responses. In general, FENDL-3 shows an improved performance for fusion neutronics applications.
Benchmarking of the FENDL-3 Neutron Cross-section Data Starter Library for Fusion Applications
This paper summarizes the benchmark analyses performed in a joint effort of ENEA (Italy), JAEA (Japan), KIT (Germany), and the University of Wisconsin (USA) on a computational ITER benchmark and a series of 14 MeV neutron benchmark experiments. The computational benchmark revealed a modest increase of the neutron flux levels in the deep penetration regions and a substantial increase of the gas production in steel components. The comparison to experimental results showed good agreement with no substantial differences between FENDL-3.0 and FENDL-2.1 for most of the responses. In general, FENDL-3 shows an improved performance for fusion neutronics applications
Benchmark experiment on vanadium assembly with D-T neutrons. In-situ measurement
Maekawa, Fujio; Kasugai, Yoshimi; Konno, Chikara; Wada, Masayuki; Oyama, Yukio; Ikeda, Yujiro [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Murata, Isao; Kokooo; Takahashi, Akito
1998-03-01
Fusion neutronics benchmark experimental data on vanadium were obtained for neutrons in almost entire energies as well as secondary gamma-rays. Benchmark calculations for the experiment were performed to investigate validity of recent nuclear data files, i.e., JENDL Fusion File, FENDL/E-1.0 and EFF-3. (author)
RADSAT Benchmarks for Prompt Gamma Neutron Activation Analysis Measurements
Burns, Kimberly A.; Gesh, Christopher J.
2011-07-01
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. High-resolution gamma-ray spectrometers are used in these applications to measure the spectrum of the emitted photon flux, which consists of both continuum and characteristic gamma rays with discrete energies. Monte Carlo transport is the most commonly used simulation tool for this type of problem, but computational times can be prohibitively long. This work explores the use of multi-group deterministic methods for the simulation of coupled neutron-photon problems. The main purpose of this work is to benchmark several problems modeled with RADSAT and MCNP to experimental data. Additionally, the cross section libraries for RADSAT are updated to include ENDF/B-VII cross sections. Preliminary findings show promising results when compared to MCNP and experimental data, but also areas where additional inquiry and testing are needed. The potential benefits and shortcomings of the multi-group-based approach are discussed in terms of accuracy and computational efficiency.
Highlights: • Performance estimation of nuclear-data benchmark was investigated. • Point detector contribution played a benchmark role not only to the neutron producing the detector contribution but also equally to all the upstream transport neutrons. • New functions were defined to give how well the contribution could be interpreted for benchmarking. • Benchmark performance could be evaluated only by a forward Monte Carlo calculation. -- Abstract: The author's group has been investigating how the performance estimation of nuclear-data benchmark using experiment and its analysis by Monte Carlo code should be carried out especially at 14 MeV. We have recently found that a detector contribution played a benchmark role not only to the neutron producing the detector contribution but also equally to all the upstream neutrons during the neutron history. This result would propose that the benchmark performance could be evaluated only by a forward Monte Carlo calculation. In this study, we thus defined new functions to give how well the contribution could be utilized for benchmarking using the point detector, and described that it was deeply related to the newly introduced “partial adjoint contribution”. By preparing these functions before benchmark experiments, one could know beforehand how well and for which nuclear data the experiment results could do benchmarking in forward Monte Carlo calculations
Benchmark-experiments for Pb and Bi neutron data testing
The expedience of accurate estimation of neutron data for Pb and Bi has increased recently in connection with the Accelerator-driven system (ADS) projects and the new generation fast reactors under development, which shall use lead or lead-bismuth coolant. Still the significant difference (10%) in the energy range of 100 keV - 500 keV, for the σtot from various data sets has been observed. The differences found are associated with the energy range, for which experimental information is lacking. The situation with Bi data is not better. In this connection, several benchmarks were assembled at BFS with uranium and plutonium fuel and lead or lead-bismuth coolant. The scope of the investigations included the measurements of the spectral indexes, distributions of the fission rates of the main isotopes, small samples worths and coolant voiding. The special program was connected with minor actinides. The influence of the plutonium isotope composition was investigated at the assemblies with reactor and weapon grade Pu. Calculations of the measured parameters were carried out using the most modern versions of nuclear data libraries. All the results of these experiments and their analysis have prepared for the construction of the benchmarks and planed as the candidates for the International data base IRPhEP. (authors)
Analytical benchmarks for nuclear engineering applications. Case studies in neutron transport theory
The developers of computer codes involving neutron transport theory for nuclear engineering applications seldom apply analytical benchmarking strategies to ensure the quality of their programs. A major reason for this is the lack of analytical benchmarks and their documentation in the literature. The few such benchmarks that do exist are difficult to locate, as they are scattered throughout the neutron transport and radiative transfer literature. The motivation for this benchmark compendium, therefore, is to gather several analytical benchmarks appropriate for nuclear engineering applications under one cover. We consider the following three subject areas: neutron slowing down and thermalization without spatial dependence, one-dimensional neutron transport in infinite and finite media, and multidimensional neutron transport in a half-space and an infinite medium. Each benchmark is briefly described, followed by a detailed derivation of the analytical solution representation. Finally, a demonstration of the evaluation of the solution representation includes qualified numerical benchmark results. All accompanying computer codes are suitable for the PC computational environment and can serve as educational tools for courses in nuclear engineering. While this benchmark compilation does not contain all possible benchmarks, by any means, it does include some of the most prominent ones and should serve as a valuable reference. (author)
Neutron scattering in magnetic fields
The use of magnetic fields in neutron scattering experimentation is reviewed briefly. Two general areas of application can be distinguished. In one the field acts to change the properties of the scattering sample; in the second the field acts on the neutron itself. Several examples are discussed. Precautions necessary for high precision polarized beam measurements are reviewed. 33 references
Neutron scattering in magnetic fields
Koehler, W.C.
1984-01-01
The use of magnetic fields in neutron scattering experimentation is reviewed briefly. Two general areas of application can be distinguished. In one the field acts to change the properties of the scattering sample ; in the second the field acts on the neutron itself. Several examples are discussed. Precautions necessary for high precision polarized beam measurements are reviewed.
The ITER IT has adopted the newly developed FEM, 3-D, and CAD-based Discrete Ordinates code, ATTILA for the neutronics studies contingent on its success in predicting key neutronics parameters and nuclear field according to the stringent QA requirements set forth by the Management and Quality Program (MQP). ATTILA has the advantage of providing a full flux and response functions mapping everywhere in one run where components subjected to excessive radiation level and strong streaming paths can be identified. The ITER neutronics community had agreed to use a standard CAD model of ITER (40 degree sector, denoted ''Benchmark CAD Model'') to compare results for several responses selected for calculation benchmarking purposes to test the efficiency and accuracy of the CAD-MCNP approach developed by each party. Since ATTILA seems to lend itself as a powerful design tool with minimal turnaround time, it was decided to benchmark this model with ATTILA as well and compare the results to those obtained with the CAD MCNP calculations. In this paper we report such comparison for five responses, namely: (1) Neutron wall load on the surface of the 18 shield blanket module (SBM), (2) Neutron flux and nuclear heating rate in the divertor cassette, (3) nuclear heating rate in the winding pack of the inner leg of the TF coil, (4) Radial flux profile across dummy port plug and shield plug placed in the equatorial port, and (5) Flux at seven point locations situated behind the equatorial port plug. (orig.)
The Spallation Neutron Source (SNS) will provide an intense source of low-energy neutrons for experimental use. The low-energy neutrons are produced by the interaction of a high-energy (1.0 GeV) proton beam on a mercury (Hg) target and slowed down in liquid hydrogen or light water moderators. Computer codes and computational techniques are being benchmarked against relevant experimental data to validate and verify the tools being used to predict the performance of the SNS. The LAHET Code System (LCS), which includes LAHET, HTAPE ad HMCNP (a modified version of MCNP version 3b), have been applied to the analysis of experiments that were conducted in the Alternating Gradient Synchrotron (AGS) facility at Brookhaven National Laboratory (BNL). In the AGS experiments, foils of various materials were placed around a mercury-filled stainless steel cylinder, which was bombarded with protons at 1.6 GeV. Neutrons created in the mercury target, activated the foils. Activities of the relevant isotopes were accurately measured and compared with calculated predictions. Measurements at BNL were provided in part by collaborating scientists from JAERI as part of the AGS Spallation Target Experiment (ASTE) collaboration. To date, calculations have shown good agreement with measurements
Measurement Methods in the field of benchmarking
István Szűts
2004-05-01
Full Text Available In benchmarking we often come across with parameters being difficultto measure while executing comparisons or analyzing performance, yet they haveto be compared and measured so as to be able to choose the best practices. Thesituation is similar in the case of complex, multidimensional evaluation as well,when the relative importance and order of different dimensions, parameters to beevaluated have to be determined or when the range of similar performanceindicators have to be decreased with regard to simpler comparisons. In suchcases we can use the ordinal or interval scales of measurement elaborated by S.S.Stevens.
The MUSE project, carried out within the European fifth Framework Program, focuses on the coupling of a sub-critical reactor core with an external neutron source. In the first stage of the project a benchmark has been defined in order to define a reference calculational route, which is able to accurately predict the neutronics behavior in an accelerator driven system. Benchmark calculations will be carried out by several members of the project and the results will be compared, also with experimental results. The contribution of NRG to the project consists of the benchmark calculations and additional work that focuses on the calculation of 3D distributions of reaction yields. This paper discusses the non-conventional methods used to perform the benchmark calculations, including the 3D reaction yield distributions. The 3D distributions calculated for the sub-critical core will be Shown and discussed. With the ORANGE-extension to MCNP it is possible to tally 3D distributions, without adding extra cells and surfaces to the geometry and without a significant slowing down of the calculation. These are major advantages when compared to the conventional way of tallying in the MCNP-code. The distributions show details that can be understood in terms of the expected neutron behavior in the different parts of the geometry. For instance, the results show that: 1) a large number of fast neutrons is found in the fuel regions, 2) the reflector region shows an increased number of slower neutrons and 3) the reaction yield in the shielding region declines steeply. The extension therefore seems a useful tool in generating a better understanding of the behavior of neutrons throughout large and complex geometries like accelerator driven systems, but we also expect to use the extension in a variety of different fields. (authors)
Benchmarking of the FENDL-3 Neutron Cross-Section Data Library for Fusion Applications
This report summarizes the benchmark analyses performed in a joint effort of ENEA (Italy), JAEA (Japan), KIT (Germany), and the University of Wisconsin (USA) with the objective to test and qualify the neutron induced general purpose FENDL-3.0 data library for fusion applications. The benchmark approach consisted of two major steps including the analysis of a simple ITER-like computational benchmark, and a series of analyses of benchmark experiments conducted previously at the 14 MeV neutron generator facilities at ENEA Frascati, Italy (FNG) and JAEA, Tokai-mura, Japan (FNS). The computational benchmark revealed a modest increase of the neutron flux levels in the deep penetration regions and a substantial increase of the gas production in steel components. The comparison to experimental results showed good agreement with no substantial differences between FENDL-3.0 and FENDL-2.1 for most of the responses analysed. There is a slight trend, however, for an increase of the fast neutron flux in the shielding experiment and a decrease in the breeder mock-up experiments. The photon flux spectra measured in the bulk shield and the tungsten experiments are significantly better reproduced with FENDL-3.0 data. In general, FENDL-3, as compared to FENDL-2.1, shows an improved performance for fusion neutronics applications. It is thus recommended to ITER to replace FENDL-2.1 as reference data library for neutronics calculation by FENDL-3.0. (author)
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.
Analysis of two different benchmark problems using one-dimensional neutron transport theory code
This paper focuses on the application of method of characteristics (MOC) for the solution of neutron transport equation in one-dimensional geometries. The paper discusses the results obtained for two different benchmark problems. The results compared well with the benchmark results. An interesting result is that, in case of MOC the unphysical flux dip at the centre of sphere (commonly found with SN - method) is absent. (author)
SCALE Modeling of Selected Neutronics Test Problems within the OECD UAM LWR’s Benchmark
Luigi Mercatali; Kostadin Ivanov; Victor Hugo Sanchez
2013-01-01
The OECD UAM Benchmark was launched in 2005 with the objective of determining the uncertainty in the simulation of Light Water Reactors (LWRs) system calculations at all the stages of the coupled reactor physics—thermal hydraulics modeling. Within the framework of the “Neutronics Phase” of the Benchmark the solutions of some selected test cases at the cell physics and lattice physics levels are presented. The SCALE 6.1 code package has been used for the neutronics modeling of the selected exe...
Benchmarks on neutron leakage from iron and Beryllium slavs and spheres
Five benchmarks, recommended by the IAEA for nuclear power engineering have been calculated for an assessment of the Iron and Beryllium neutron data from the recent FENDL-1 version. The FENDL/MG-1.0 multigroup data processed in the IAEA by NJOY code are in VITAMIN-J energy structure in MATXS format. These data have been converted to ANISN format by TRANSX code and collected to binary library by LIBFENDL code. The neutron transport calculations have been carried out by the codes ANISN, GRTUNCL and DORT. Two benchmarks corresponding to the 14 MeV neutron transmission through Iron sphere shell (Simakov S.P. at al, IPPE, Obninsk) and Iron slabs (Y. Oyama and H. Maekawa, FNS/JAERI) permit to test the FENDL-1 Iron data for fusion application. The benchmark on neutron leakage from 25 cm radius Iron sphere with 252Cf source allows to show the FENDL-1 Iron data applicability in LWRs tasks. The comparison of the calculated and measured results demonstrates discouraged inconsistency when material thickness exceeds 20 cm . Modelling of the 14 MeV neutrons' transmission through Beryllium slabs (H. Maekawa and Y. Oyama at FNS/JAERI), and through sphere shell (Simakov S.P. at al in IPPE, Obninsk) has been carried out to test the multiplication data for the Beryllium as a fusion blanket material . The calculated angular neutron leakage from the slabs and the scalar neutron leakage from the sphere are in relatively good consistency with the measured ones. (author)
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
Benchmark testing of a multiblade neutron velocity selector
Rosta, Laszlo [Research Institute for Solid State Physics and Optics, H-1525 Budapest (Hungary); Fuezi, Janos [Research Institute for Solid State Physics and Optics, H-1525 Budapest (Hungary); Transilvania University, R-2200 Brasov (Romania); Homanyi, Laszlo [MIRROTRON Ltd., H-1525 Budapest (Hungary)]. E-mail: lhomanyi@mirrotron.kfkipark.hu
2006-11-15
The fully operational prototype of a rotational velocity selector for neutron beam monochromation has been tested and the results compared to the theoretical characteristics (output wavelength, transmission and selectivity with respect to velocity and horizontal tilt angle of the multiblade rotor system). The minimum of the selected central wavelength is 2.5 A in the current construction. Measurements are performed on a cold neutron beam at the Budapest Neutron Centre by energy-resolved pinhole beam imaging, using a 2D position-sensitive detector in time-of-flight regime. Thus, the effects of the incoming beam divergence are also evaluated.
Neutron dosimetry in mixed fields with monoblock neutron spectrometer
Full text: The multi-sphere method of neutron spectrometry or namely Bonner spheres neutron spectrometry is currently playing an increasing role in the mixed radiation field measurements. The growing popularity of this methodology is caused by its relative availability, simplicity of measurement in a wide energy range, high sensitivity and satisfactory gamma-ray suppression. These qualities allow the usage of multi-sphere neutron spectrometers for adequate characterization of neutron field, particularly reliable measurements of neutron dose rate. However, the main difficulties in the application of this kind of neutron detector are the perturbation of the neutron field, caused by the detector itself, and the complex procedure required for unfolding the neutron spectrum. Furthermore, it is necessary to perform a relatively high number of measurements, one for each spherical moderator (as a rule, 5-7 pieces). This in turn may require a dedicated source monitoring system, otherwise significant errors may occur. These requirements hamper the application of the multi-sphere spectrometry method to pulsed neutron sources, for example. Other difficulties occur in the characterization of reactor neutron beams, in case the beam diameter is smaller than those of the spherical moderators. In this situation it is necessary to carry out a beam scanning and integrate the acquired data. To improve the methodology of neutron field parameter measurement the Monoblock Neutron Spectrometer (MNS) has been developed recently. The basic idea of the novel detector is to determine the neutron energy spectrum by unfolding a set of count rates from thermal neutron detectors located at different depths in the common polyethylene moderator. The unfolding algorithms for neutron spectrum and neutron dose rates have been specifically improved for operation with MNS. The testing results with well-know neutron reference fields and reactor neutron beam are presented. The application of MNS for
TRIPOLI-4® - MCNP5 ITER A-lite neutronic model benchmarking
Jaboulay, J.-C.; Cayla, P.-Y.; Fausser, C.; Lee, Y.-K.; Trama, J.-C.; Li-Puma, A.
2014-06-01
The aim of this paper is to present the capability of TRIPOLI-4®, the CEA Monte Carlo code, to model a large-scale fusion reactor with complex neutron source and geometry. In the past, numerous benchmarks were conducted for TRIPOLI-4® assessment on fusion applications. Experiments (KANT, OKTAVIAN, FNG) analysis and numerical benchmarks (between TRIPOLI-4® and MCNP5) on the HCLL DEMO2007 and ITER models were carried out successively. In this previous ITER benchmark, nevertheless, only the neutron wall loading was analyzed, its main purpose was to present MCAM (the FDS Team CAD import tool) extension for TRIPOLI-4®. Starting from this work a more extended benchmark has been performed about the estimation of neutron flux, nuclear heating in the shielding blankets and tritium production rate in the European TBMs (HCLL and HCPB) and it is presented in this paper. The methodology to build the TRIPOLI-4® A-lite model is based on MCAM and the MCNP A-lite model (version 4.1). Simplified TBMs (from KIT) have been integrated in the equatorial-port. Comparisons of neutron wall loading, flux, nuclear heating and tritium production rate show a good agreement between the two codes. Discrepancies are mainly included in the Monte Carlo codes statistical error.
Benchmarking of photon and coupled neutron and photon process of SuperMC 2.0
Super Monte Carlo Calculation Program for Nuclear and Radiation Process (SuperMC), developed by FDS Team in China, is a multi-functional simulation program mainly based on Monte Carlo (MC) method and advanced computer technology. This paper focuses on the benchmarking of physical process of photon and coupled neutron-photon of SuperMC2.0. Integral leakage rate of photon in the spherical and hemispherical shell experiment was tested to verify the physical process of photon and coupled neutron and photon transport. Vanadium assembly experiment and ADS benchmark were given as comprehensive benchmarks. The correctness was preliminarily verified by comparing calculation results of SuperMC with experimental results and MCNP calculation results. (author)
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
Neutron in Strong Magnetic Fields
Andreichikov, M A; Orlovsky, V D; Simonov, Yu A
2013-01-01
Relativistic world-line Hamiltonian for strongly interacting 3q systems in magnetic field is derived from the path integral for the corresponding Green's function. The neutral baryon Hamiltonian in magnetic field obeys the pseudomomentum conservation and allows a factorization of the c.m. and internal motion. The resulting expression for the baryon mass in magnetic field is written explicitly with the account of hyperfine, OPE and OGE (color Coulomb) interaction. The neutron mass is fast decreasing with magnetic field, losing 1/2 of its value at eB~0.25 GeV^2 and is nearly zero at eB~0.5 GeV^2. Possible physical consequences of the calculated mass trajectory of the neutron, M_n(B), are presented and discussed.
Data collection of fusion neutronics benchmark experiment conducted at FNS/JAERI
Maekawa, Fujio; Konno, Chikara; Kasugai, Yoshimi; Oyama, Yukio; Ikeda, Yujiro [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment
1998-08-01
Fusion neutronics benchmark experimental data have been continued at the Fusion Neutronics Source (FNS) facility in JAERI. This report compiles unpublished results of the in-situ measurement experiments conducted by the end of 1996. Experimental data of neutron spectra in entire energy range, dosimetry reaction rates, gamma-ray spectrum and gamma-ray heating rates are acquired for five materials of beryllium, vanadium, iron, copper and tungsten. These experimental data along with data previously reported are effective for validating cross section data stored in evaluated nuclear data files such as JENDL. (author)
Data collection of fusion neutronics benchmark experiment conducted at FNS/JAERI
Fusion neutronics benchmark experimental data have been continued at the Fusion Neutronics Source (FNS) facility in JAERI. This report compiles unpublished results of the in-situ measurement experiments conducted by the end of 1996. Experimental data of neutron spectra in entire energy range, dosimetry reaction rates, gamma-ray spectrum and gamma-ray heating rates are acquired for five materials of beryllium, vanadium, iron, copper and tungsten. These experimental data along with data previously reported are effective for validating cross section data stored in evaluated nuclear data files such as JENDL. (author)
Sustaining knowledge in the neutron generator community and benchmarking study.
Barrentine, Tameka C.; Kennedy, Bryan C.; Saba, Anthony W.; Turgeon, Jennifer L.; Schneider, Julia Teresa; Stubblefield, William Anthony; Baldonado, Esther
2008-03-01
In 2004, the Responsive Neutron Generator Product Deployment department embarked upon a partnership with the Systems Engineering and Analysis knowledge management (KM) team to develop knowledge management systems for the neutron generator (NG) community. This partnership continues today. The most recent challenge was to improve the current KM system (KMS) development approach by identifying a process that will allow staff members to capture knowledge as they learn it. This 'as-you-go' approach will lead to a sustainable KM process for the NG community. This paper presents a historical overview of NG KMSs, as well as research conducted to move toward sustainable KM.
Integral Data Benchmark of HENDL2.0/MG Compared with Neutronics Shielding Experiments
Jiang, Jieqiong; Xu, Dezheng; Zheng, Shanliang; He, Zhaozhong; Hu, Yanglin; Li, Jingjing; Zou, Jun; Zeng, Qin; Chen, Mingliang; Wang, Minghuang
2009-10-01
HENDL2.0, the latest version of the hybrid evaluated nuclear data library, was developed based upon some evaluated data from FENDL2.1 and ENDF/B-VII. To qualify and validate the working library, an integral test for the neutron production data of HENDL2.0 was performed with a series of existing spherical shell benchmark experiments (such as V, Be, Fe, Pb, Cr, Mn, Cu, Al, Si, Co, Zr, Nb, Mo, W and Ti). These experiments were simulated numerically using HENDL2.0/MG and a home-developed code VisualBUS. Calculations were conducted with both FENDL2.1/MG and FENDL2.1/MC, which are based on a continuous-energy Monte Carlo Code MCNP/4C. By comparison and analysis of the neutron leakage spectra and the integral test, benchmark results of neutron production data are presented in this paper.
Integral Data Benchmark of HENDL2.0/MG Compared with Neutronics Shielding Experiments
JIANG Jieqiong; XU Dezheng; ZHENG Shanliang; HE Zhaozhong; HU Yanglin; LI Jingjing; ZOU Jun; ZENG Qin; CHEN Mingliang; WANG Minghuang
2009-01-01
HENDL2.0,the latest version of the hybrid evaluated nuclear data library,was developed based upon some evaluated data from FENDL2.1 and ENDF/B-VII.To qualify and validate the working library,an integral test for the neutron production data of HENDL2.0 was performed with a series of existing spherical shell benchmark experiments (such as V,Be,Fe,Pb,Cr,Mn,Cu,Al,Si,Co,Zr,Nb,Mo,W and Ti).These experiments were simulated numerically using HENDL2.0/MG and a home-developed code VisualBUS.Calculations were conducted with both FENDL2.1/MG and FENDL2.1/MC,which are based on a continuous-energy Monte Carlo Code MCNP/4C.By comparison and analysis of the neutron leakage spectra and the integral test,benchmark results of neutron production data are presented in this paper.
Benchmarking the inelastic neutron scattering soil carbon method
The herein described inelastic neutron scattering (INS) method of measuring soil carbon was based on a new procedure for extracting the net carbon signal (NCS) from the measured gamma spectra and determination of the average carbon weight percent (AvgCw%) in the upper soil layer (~8 cm). The NCS ext...
Neutron transport benchmark examples with web-based AGENT
The AGENT (Arbitrary GEometry Neutron Transport) an open-architecture reactor modeling tool is deterministic neutron transport code for two- or three-dimensional heterogeneous neutronic design and analysis of the whole reactor cores regardless of geometry types and material configurations. The AGENT neutron transport methodology is applicable to all generations of nuclear power and research reactors. It combines three theories: (1) mathematical theory of R-functions that is used to generate real three-dimensional geometries of square or hexagonal heterogeneous geometries, (2) the x-y method of characteristics (MOC) used to solve isotropic neutron transport in non-homogenized 2D reactor slices, and (3) the one-dimensional diffusion theory or MOC theory used to couple the x-y and z neutron tracks through the transverse leakage and angular mesh-wise flux values. The R-function geometrical module allows a sequential building of the layers of geometry and automatic submeshing based on the network of geometric domain functions. The simplicity of geometry description and selection of parameters for accurate treatment of neutron propagation is achieved through the Boolean algebraic hierarchically organized simple primitives into complex domains (both being represented with corresponding domain functions). AGENT methodologies and numerical solutions are applicable in validating neutronic analysis for GenIV reactor designs while the effect of double heterogeneity in very high temperature reactors (VHTRs) is under development. The accuracy is comparable to Monte Carlo codes and is obtained by following neutron propagation through real geometrical domains that does not require homogenization or simplifications. The efficiency is maintained through set of acceleration techniques introduced at all important calculation levels. The flux solution incorporates power iteration with two different acceleration techniques: coarse mesh rebalancing (CMR) and coarse mesh finite difference
Highlights: • A Neutronics and Thermal-hydraulics Coupled code is developed for transient analysis. • The spatial kinetics model was employed in the benchmark. • The simulation correctness of NTC accuracy demonstrated by benchmark. - Abstract: The Neutronics and Thermal-hydraulics Coupled Simulation program (NTC) is developed by FDS Team, which is a code used for transient analysis of advanced reactors. To investigate the capacity and calculation correctness of NTC for transient simulation, a benchmark on beam interruptions in an 80 MWth LBE-cooled and MOX-fuelled experimental accelerator-driven sub-critical system XADS was carried out by NTC. The benchmark on beam interruptions used in this paper was developed by the OECD/NEA Working Party on Scientific Issues in Partitioning and Transmutation (WPPT). The calculation model had the minimum phenomenological and computational complexity which concerned a simple model (single fuel channel thermal-hydraulics) of the average fuel pin corresponding to the BOL fuel condition. This benchmark was designed to investigate the temperature and power responses caused by beam interruption of different durations, which aimed at comparative assessment of NTC and other computation methods. A comparison of NTC and other ten sets of temperature and power was provided, which showed that the results had good agreement
V Wagner; A Krása; M Majerla; F Křížek; O Svoboda; A Kugler; J Adam; V M Tsoupko-Sitnikov; M I Krivopustov; I V Zhuk; W Westmeier
2007-02-01
The set-up `energy plus transmutation', consisting of a thick lead target and a natural uranium blanket, was irradiated by relativistic proton beams with the energy from 0.7 GeV up to 2 GeV. Neutron field was measured in different places of this set-up using different activation detectors. The possibilities of using the obtained data for benchmark studies are analyzed in this paper. Uncertainties of experimental data are shown and discussed. The experimental data are compared with results of simulation with MCNPX code.
Ueki, Kohtaro; Ohashi, Atsuto (Ship Research Inst., Mitaka, Tokyo (Japan)); Kawai, Masayoshi
1993-04-01
The iron, carbon and beryllium cross sections in JENDL-3 have been tested by the continuous energy Monte Carlo analysis of the neutron shielding benchmark experiments. The iron cross sections have been tested with analysis of the ORNL and the Winfrith experiments using the fission neutron sources, and also the LLNL iron experiment using the D-T neutron source. The carbon and beryllium cross sections have been tested with the JAERI-FNS TOF experiments using the D-T neutron source. Revision of the subroutine TALLYD and an appropriate weight-window-parameter assignment have been accomplished in the MCNP code. In consequence, the FSD for each energy bin is reduced so small that the Monte Carlo results for neutron energy spectra could be recognized to be reliable. The Monte Carlo calculations with JENDL-3 indicate a good agreement with the benchmark experiments in a wide energy range, as a whole. Particularly, for the Winfrith iron experiment, the results with JENDL-3 give better agreement, just below the iron 24keV window, than that with ENDF/B-IV. For the JAERI-FNS TOF graphite experiment, the calculated angular fluxes with JENDL-3 give closer agreement than that with ENDF/B-IV at several peaks and dips caused by the inelastic scattering. However, distinct underestimation is observed in the calculated energy spectrum with JENDL-3 between 0.8 and 3.0 MeV for the two iron experiments using fission neutron sources. (author).
Sustaining knowledge in the neutron generator community and benchmarking study. Phase II.
Huff, Tameka B.; Stubblefield, William Anthony; Cole, Benjamin Holland, II; Baldonado, Esther
2010-08-01
This report documents the second phase of work under the Sustainable Knowledge Management (SKM) project for the Neutron Generator organization at Sandia National Laboratories. Previous work under this project is documented in SAND2008-1777, Sustaining Knowledge in the Neutron Generator Community and Benchmarking Study. Knowledge management (KM) systems are necessary to preserve critical knowledge within organizations. A successful KM program should focus on people and the process for sharing, capturing, and applying knowledge. The Neutron Generator organization is developing KM systems to ensure knowledge is not lost. A benchmarking study involving site visits to outside industry plus additional resource research was conducted during this phase of the SKM project. The findings presented in this report are recommendations for making an SKM program successful. The recommendations are activities that promote sharing, capturing, and applying knowledge. The benchmarking effort, including the site visits to Toyota and Halliburton, provided valuable information on how the SEA KM team could incorporate a KM solution for not just the neutron generators (NG) community but the entire laboratory. The laboratory needs a KM program that allows members of the workforce to access, share, analyze, manage, and apply knowledge. KM activities, such as communities of practice (COP) and sharing best practices, provide a solution towards creating an enabling environment for KM. As more and more people leave organizations through retirement and job transfer, the need to preserve knowledge is essential. Creating an environment for the effective use of knowledge is vital to achieving the laboratory's mission.
Sustaining knowledge in the neutron generator community and benchmarking study. Phase II
This report documents the second phase of work under the Sustainable Knowledge Management (SKM) project for the Neutron Generator organization at Sandia National Laboratories. Previous work under this project is documented in SAND2008-1777, Sustaining Knowledge in the Neutron Generator Community and Benchmarking Study. Knowledge management (KM) systems are necessary to preserve critical knowledge within organizations. A successful KM program should focus on people and the process for sharing, capturing, and applying knowledge. The Neutron Generator organization is developing KM systems to ensure knowledge is not lost. A benchmarking study involving site visits to outside industry plus additional resource research was conducted during this phase of the SKM project. The findings presented in this report are recommendations for making an SKM program successful. The recommendations are activities that promote sharing, capturing, and applying knowledge. The benchmarking effort, including the site visits to Toyota and Halliburton, provided valuable information on how the SEA KM team could incorporate a KM solution for not just the neutron generators (NG) community but the entire laboratory. The laboratory needs a KM program that allows members of the workforce to access, share, analyze, manage, and apply knowledge. KM activities, such as communities of practice (COP) and sharing best practices, provide a solution towards creating an enabling environment for KM. As more and more people leave organizations through retirement and job transfer, the need to preserve knowledge is essential. Creating an environment for the effective use of knowledge is vital to achieving the laboratory's mission.
Benchmark experiment on a copper slab assembly bombarded by D-T neutrons
Copper is a very important material for fusion reactor because it is used in superconducting magnets or first walls and so on. To verify nuclear data of copper, a benchmark experiment was performed using the D-T neutron source of the FNS facility in Japan Atomic Energy Research Institute. An cylindrical experimental assembly of 629 mm in diameter and 608 mm in thickness made of pure copper was located at 200 mm from the D-T neutron source. In the assembly, the following quantities were measured; i) neutron spectra in energy regions of MeV and keV, ii) neutron reaction rates, iii) prompt and decay gamma-ray spectra and iv) gamma-ray heating rates. The obtained experimental data were compiled in this report. (author)
D. W. Nigg; J. K. Hartwell; J. R. Venhuizen; C. A. Wemple; R. Risler; G. E. Laramore; W. Sauerwein; G. Hudepohl; A. Lennox
2006-06-01
The Idaho National Laboratory (INL), the University of Washington (UW) Neutron Therapy Center, the University of Essen (Germany) Neutron Therapy Clinic, and the Northern Illinois University(NIU) Institute for Neutron Therapy at Fermilab have been collaborating in the development of fast-neutron therapy (FNT) with concurrent neutron capture (NCT) augmentation [1,2]. As part of this effort, we have conducted measurements to produce suitable benchmark data as an aid in validation of advanced three-dimensional treatment planning methodologies required for successful administration of FNT/NCT. Free-beam spectral measurements as well as phantom measurements with Lucite{trademark} cylinders using thermal, resonance, and threshold activation foil techniques have now been completed at all three clinical accelerator facilities. The same protocol was used for all measurements to facilitate intercomparison of data. The results will be useful for further detailed characterization of the neutron beams of interest as well as for validation of various charged particle and neutron transport codes and methodologies for FNT/NCT computational dosimetry, such as MCNP [3], LAHET [4], and MINERVA [5].
Calculation of the IAEA ADS neutronics benchmark (stage-1) (2D discrete coordinate method)
To study the neutronics for the ADS system, a set of computation software based on discrete ordinate method is selected and established. The set is tested through an IAEA benchmark. In the test process, the understanding and using of this software set are improved. The benchmark is analyzed. The calculations include the effective multiplication factor keff , the required strength of the spallation neutron source for 1.5 GW thermal power, the distribution of power density and the spectrum index, and the void effect at the beginning of life, BOL; the spatial and time-dependent density distribution of various nuclides (actinides and fission products) for burn-up process. The results are given in figures and tables and are consistent with calculations made abroad. The conclusion is that this software set can be applied to the optimization of design study for the ADS system
Development and benchmarking of higher energy neutron transport data libraries
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
Full text: New releases of nuclear data files made available during the few recent years. The reference MCNP5 code (1) for Monte Carlo calculations is usually distributed with only one standard nuclear data library for neutron interactions based on ENDF/B-VI. The main goal of this work is to process new neutron cross sections libraries in ACE continuous format for MCNP code based on the most recent data files recently made available for the scientific community : ENDF/B-VII.b2, ENDF/B-VI (release 8), JEFF3.0, JEFF-3.1, JENDL-3.3 and JEF2.2. In our data treatment, we used the modular NJOY system (release 99.9) (2) in conjunction with its most recent upadates. Assessment of the processed point wise cross sections libraries performances was made by means of some criticality prediction and analysis of other integral parameters for a set of reactor benchmarks. Almost all the analyzed benchmarks were taken from the international handbook of Evaluated criticality safety benchmarks experiments from OECD (3). Some revised benchmarks were taken from references (4,5). These benchmarks use Pu-239 or U-235 as the main fissionable materiel in different forms, different enrichments and cover various geometries. Monte Carlo calculations were performed in 3D with maximum details of benchmark description and the S(α,β) cross section treatment was adopted in all thermal cases. The resulting one standard deviation confidence interval for the eigenvalue is typically +/-13% to +/-20 pcm
Parareal in time 3D numerical solver for the LWR Benchmark neutron diffusion transient model
Baudron, Anne-Marie A -M; Maday, Yvon; Riahi, Mohamed Kamel; Salomon, Julien
2014-01-01
We present a parareal in time algorithm for the simulation of neutron diffusion transient model. The method is made efficient by means of a coarse solver defined with large time steps and steady control rods model. Using finite element for the space discretization, our implementation provides a good scalability of the algorithm. Numerical results show the efficiency of the parareal method on large light water reactor transient model corresponding to the Langenbuch-Maurer-Werner (LMW) benchmark [1].
A neutron collimator with adjustable radiation field
An adjustable neutron collimator for neutron therapy purposes is described. The collimator is designed to give a very sharp radiation field and a high freedom of choice for the radiation geometrics. (L.E.)
EA-MC Neutronic Calculations on IAEA ADS Benchmark 3.2
The neutronics and the transmutation properties of the IAEA ADS benchmark 3.2 setup, the 'Yalina' experiment or ISTC project B-70, have been studied through an extensive amount of 3-D Monte Carlo calculations at CERN. The simulations were performed with the state-of-the-art computer code package EA-MC, developed at CERN. The calculational approach is outlined and the results are presented in accordance with the guidelines given in the benchmark description. A variety of experimental conditions and parameters are examined; three different fuel rod configurations and three types of neutron sources are applied to the system. Reactivity change effects introduced by removal of fuel rods in both central and peripheral positions are also computed. Irradiation samples located in a total of 8 geometrical positions are examined. Calculations of capture reaction rates in 129I, 237Np and 243Am samples and of fission reaction rates in 235U, 237Np and 243Am samples are presented. Simulated neutron flux densities and energy spectra as well as spectral indices inside experimental channels are also given according to benchmark specifications. Two different nuclear data libraries, JAR-95 and JENDL-3.2, are applied for the calculations
Benchmark calculations on neutrons streaming through mazes at proton accelerator facilities
In accelerator shielding designs one of the important issues is to estimate radiation streaming through mazes and ducts. In order to validate the accuracy of the calculation methods concerning such neutron streaming, benchmark analyses were carried out using two kinds of benchmark problems based on past experiments. The analyses showed that the design methods were applicable to neutron streaming calculations of proton accelerator facilities with an uncertainty within a factor of two. In the analyses, relative comparisons were conducted using a radiation source generated by GeV energy protons, and absolute comparisons were conducted using a low-energy neutron source of a few tens of MeV. A radiation streaming experiment was planned and carried out at KEK using a radiation source produced by a thin copper target irradiated by 12 GeV protons. The preliminary experimental analysis is presented below. In addition, the authors propose to compile benchmark problems on radiation streaming for accelerator facilities and to search for possible new streaming experiments at other facilities. (authors)
Benchmark evaluation of the neutron radiography (NRAD) reactor upgraded LEU-fuel core
Benchmark models were developed to evaluate the cold-critical start-up measurements performed during the fresh core reload of the Neutron Radiography (NRAD) reactor with Low Enriched Uranium (LEU) fuel. The final upgraded core configuration with 64 fuel elements has been completed. Evaluated benchmark measurement data include criticality, control-rod worth measurements, shutdown margin, and excess reactivity. Dominant uncertainties in keff include the manganese content and impurities contained within the stainless steel cladding of the fuel and the 236U and erbium poison content in the fuel matrix. Calculations with MCNP5 and ENDF/B-VII.0 nuclear data are approximately 1.4% greater than the benchmark model eigenvalue, supporting contemporary research regarding errors in the cross section data necessary to simulate TRIGA-type reactors. Uncertainties in reactivity effects measurements are estimated to be ∼10% with calculations in agreement with benchmark experiment values within 2σ. The completed benchmark evaluation details are available in the 2014 edition of the International Handbook of Evaluated Reactor Physics Experiments (IRPhEP Handbook). Evaluation of the NRAD LEU cores containing 56, 60, and 62 fuel elements have also been completed, including analysis of their respective reactivity effects measurements; they are also available in the IRPhEP Handbook but will not be included in this summary paper. (author)
D.C. Blitz (David)
2011-01-01
textabstractBenchmarking benchmarks is a bundle of six studies that are inspired by the prevalence of benchmarking in academic finance research as well as in investment practice. Three studies examine if current benchmark asset pricing models adequately describe the cross-section of stock returns. W
Polarized neutron reflectometry in high magnetic fields
Fritzsche, H.
2005-11-01
A simple method is described to maintain the polarization of a neutron beam on its way through the large magnetic stray fields produced by a vertical field of a cryomagnet with a split-coil geometry. The two key issues are the proper shielding of the neutron spin flippers and an additional radial field component in order to guide the neutron spin through the region of the null point (i.e., point of reversal for the vertical field component). Calculations of the neutron's spin rotation as well as polarized neutron reflectometry experiments on an ErFe2/DyFe2 multilayer show the perfect performance of the used setup. The recently commissioned cryomagnet M5 with a maximum vertical field of up to 7.2T in asymmetric mode for polarized neutrons and 9T in symmetric mode for unpolarized neutrons was used on the C5 spectrometer in reflectometry mode, at the NRU reactor in Chalk River, Canada.
Development of a 3D neutron transport code and benchmark tests
Results are reported of NEACRP '3D Neutron Transport Benchmarks' proposed from Osaka UNiversity, and of recent progress in the development of a 3D neutron transport code. Takeda et al. proposed four problems to NEACRP as 3D neutron transport benchmarks, and 22 results from 20 organizations were submitted. A variety of methods have been used, such as the Monte Carlo, Sn, Pn, synthetic, and nodal method. The results for k-eff, control-rod worths, and region-averaged fluxes are summarized with the conclusions that (1) in XYZ geometry the Sn method with n=8 shows a good agreement with the Monte-Carlo method, and gives even better results in some cases, (2) the Pn method has significant spatial mesh effects, and (3) the Sn method is not satisfactory in hexagonal-Z geometry, and improvements in accuracy are desirable. Improvement of a 3D neutron transport code is in progress to resolve the problem in the hexagonal-Z geometry by considering new diamond difference schemes and an improved coarse-mesh method, and also by applying the nodal method. (author)
Benchmark experiment on the model of fusion reactor blanket with uranium neutron multiplier
Benchmark experiment on the model of thermonuclear reactor blanket with 14 MeV neutron source is described. The model design corresponds to the known concept of the fast hybrid blanket with 238U neutron multiplier and main tritium production on 6Li. Detailed measurements of the following process velocities were carried out: tritium production on lithium isotopes; reactions modelling tritium production; (n, γ) and (n, 2n) processes for 238U; fission reactions for 235,238U, 239Pu, 237Np. Neutron flux integral measurements were performed by a set of threshold detectors on the basic of the 115In(n, n'), 204Pb(n, n'), 64Zn(n, p), 27Al(n, p), 56Fe(n, p), 107Ag(n, 2n), 63Cu(n, 2n) and 64(n, 2n) reactions
Intercomparison of Monte Carlo and SN sensitivity calculations for a 14 MeV neutron benchmark
An inter-comparison has been performed of probabilistic and deterministic sensitivity calculations with the objective to check and validate the Monte Carlo technique for calculating point detector sensitivities as being implemented in MCSEN, a local version of the MCNP4A code. A suitable 14 MeV neutron benchmark problem on an iron assembly has been considered to this end. Good agreement has been achieved for the calculated individual sensitivity profiles, the uncertainties and the neutron flux spectra as well. It is concluded that the Monte Carlo technique for calculating point detector sensitivities and related uncertainties as being implemented in MCSEN is well qualified for sensitivity and uncertainty analyses of fusion neutronics integral experiments. (orig.)
Benchmark calculations of neutron dose rates at transport and storage casks
The application of numerical calculations methods for demonstration of sufficient radiation shielding of radioactive waste transport and storage casks requires a validation based on appropriate measurements of gamma and neutron sources. The results of the comparison of measured data and calculations using the Monte Carlo program MCNP show deviations dependent on the loading of the cask within the standard deviation which is dominated by the measuring method. Considering the neutrons scattered at the salt MCNP (in case of disposal in the salt) tends to underestimate the nominal values, but still within the double standard deviation. This accuracy is not reached with MAVRIC. Based on AHE (active handling experiments) data benchmark calculations were performed that can be used as reference value. The total accuracy results from the accuracy of the source term and the measurement of the neutron dose rate with a deviation of 15%.
The bulk shielding benchmark experiment at the Frascati Neutron Generator (FNG)
Batistoni, P. [Associazione EURATOM-ENEA sulla Fusione, Frascati (Italy); Angelone, M. [Associazione EURATOM-ENEA sulla Fusione, Frascati (Italy); Martone, M. [Associazione EURATOM-ENEA sulla Fusione, Frascati (Italy); Pillon, M. [Associazione EURATOM-ENEA sulla Fusione, Frascati (Italy); Rado, V. [Associazione EURATOM-ENEA sulla Fusione, Frascati (Italy); Santamarina, A. [Commissariat al`Energie Atomique, Centre d`Etudes de Cadarache, F-13108 St. Paul-lez-Durance Cedex (France); Abidi, I. [Commissariat al`Energie Atomique, Centre d`Etudes de Cadarache, F-13108 St. Paul-lez-Durance Cedex (France); Gastaldi, B. [Commissariat al`Energie Atomique, Centre d`Etudes de Cadarache, F-13108 St. Paul-lez-Durance Cedex (France); Martini, M. [Commissariat al`Energie Atomique, Centre d`Etudes de Cadarache, F-13108 St. Paul-lez-Durance Cedex (France); Marquette, J.P. [Commissariat al`Energie Atomique, Centre d`Etudes de Cadarache, F-13108 St. Paul-lez-Durance Cedex (France)
1995-03-01
In the design of next-step fusion devices such as NET/ITER the nuclear performance of shielding blankets is of key importance in terms of nuclear heating of superconducting magnets and radiation damage. In the framework of the European Fusion Technology Program, ENEA Frascati and CEA Cadarache in collaboration performed a bulk shielding benchmark experiment using the 14MeV Frascati Neutron Generator (FNG), aimed at obtaining accurate experimental data for improving the nuclear database and methods used in shielding designs. The experiment consisted of the irradiation of a stainless steel block by 14MeV neutrons. The neutron reaction rates at various depths inside the block have been measured using fission chambers and activation foils characterized by different energy response ranges. The experimental results have been compared with numerical results calculated using both S{sub n} and Monte Carlo transport codes and the cross-section library EFF.1 (European Fusion File). (orig.).
Effective Field Theory of Neutron Star Superfluidity
Hormuzdiar, James; Hsu, Stephen
1998-01-01
We apply effective field theory and renormalization group techniques to the problem of Cooper pair formation in neutron stars. Simple analytical expressions for the $^1 S_0$ condensate are derived which are free of nuclear potential model dependencies. The condensate is evaluated using phase shift data from neutron-neutron scattering.
Utilizing the slowing-down-time technique for benchmarking neutron thermalization in graphite
Graphite is the moderator/reflector in the Very High Temperature Reactor (VHTR) concept of Generation IV reactors. As a thermal reactor, the prediction of the thermal neutron spectrum in the VHTR is directly dependent on the accuracy of the thermal neutron scattering libraries of graphite. In recent years, work has been on-going to benchmark and validate neutron thermalization in 'reactor grade' graphite. Monte Carlo simulations using the MCNP5 code were used to design a pulsed neutron slowing-down-time experiment and to investigate neutron slowing down and thermalization in graphite at temperatures relevant to VHTR operation. The unique aspect of this experiment is its ability to observe the behavior of neutrons throughout an energy range extending from the source energy to energies below 0.1 eV. In its current form, the experiment is designed and implemented at the Oak Ridge Electron Linear Accelerator (ORELA). Consequently, ORELA neutron pulses are injected into a 70 cm x 70 cm x 70 cm graphite pile. A furnace system that surrounds the pile and is capable of heating the graphite to a centerline temperature of 1200 K has been designed and built. A system based on U-235 fission chambers and Li-6 scintillation detectors surrounds the pile. This system is coupled to multichannel scaling instrumentation and is designed for the detection of leakage neutrons as a function of the slowing-down-time (i.e., time after the pulse). To ensure the accuracy of the experiment, careful assessment was performed of the impact of background noise (due to room return neutrons) and pulse-to-pulse overlap on the measurement. Therefore, the entire setup is surrounded by borated polyethylene shields and the experiment is performed using a source pulse frequency of nearly 130 Hz. As the basis for the benchmark, the calculated time dependent reaction rates in the detectors (using the MCNP code and its associated ENDF-B/VI thermal neutron scattering libraries) are compared to measured
Magnetic field visualization technique using neutrons
Neutron radiography is utilized in the internal inspection of various materials due to the high sensitivity against light elements and excellent material transmission capability of neutrons. On the other hand, neutrons can interact directly with magnetic field because they have magnetic moment. As a result, neutron beams cause changes in spin state and orbit while passing through the magnetic field. If these changes can be detected for each position, the information about the magnetic field can be expressed as an image. This paper explains the characteristics of the magnetic field imaging using neutrons, in comparison with those of other techniques. Regarding the experimental examples of the visualization techniques using pulsed neutrons that have been performed in Japan, it introduces several examples in the stage of development at the Materials and Life Science Facility of J-PARC. In addition, it looks forward to the application and future of magnetic field imaging. (A.O.)
Magnetic fields in Neutron Stars
Viganò, Daniele; Miralles, Juan A; Rea, Nanda
2015-01-01
Isolated neutron stars show a diversity in timing and spectral properties, which has historically led to a classification in different sub-classes. The magnetic field plays a key role in many aspects of the neutron star phenomenology: it regulates the braking torque responsible for their timing properties and, for magnetars, it provides the energy budget for the outburst activity and high quiescent luminosities (usually well above the rotational energy budget). We aim at unifying this observational variety by linking the results of the state-of-the-art 2D magneto-thermal simulations with observational data. The comparison between theory and observations allows to place two strong constraints on the physical properties of the inner crust. First, strong electrical currents must circulate in the crust, rather than in the star core. Second, the innermost part of the crust must be highly resistive, which is in principle in agreement with the presence of a novel phase of matter so-called nuclear pasta phase.
Benchmark experiment on titanium with DT neutron at JAEA/FNS
Titanium is contained in lithium titanate which is a tritium breeding material candidate. In the nuclear design, accurate nuclear data are needed. However, few benchmark experiments had been performed for titanium. We performed a benchmark experiment with a titanium assembly and a DT neutron source at JAEA/FNS. The titanium assembly was covered with Li2O blocks in order to reduce background neutrons. Dosimetry reaction rates were measured with niobium, indium and gold foils inside the assembly. And fission rates of 235U were measured by using micro fission chambers. This experiment was analyzed by using the Monte Carlo neutron transport code MCNP5-1.40 with recent nuclear data libraries of ENDF/B-VII.0, ENDF/B-VII.1, JEFF-3.1.2, JENDL-4.0 and JENDL-4.0u1. The calculation results were compared with the measured one in order to validate the nuclear data libraries of titanium. The calculated results with ENDF/B-VII.1 agreed with the measured one the best because the (n,2n) and (n,n′cont) reaction cross section data and resonance parameters were improved
2014-01-01
A coupled neutronics thermal-hydraulics code NODAL3 has been developed based on the few-group neutron diffusion equation in 3-dimensional geometry for typical PWR static and transient analyses. The spatial variables are treated by using a polynomial nodal method while for the neutron dynamic solver the adiabatic and improved quasistatic methods are adopted. In this paper we report the benchmark calculation results of the code against the OECD/NEA CRP PWR rod ejection cases. The objective of t...
Testing of the IRDF-90 cross-section library in benchmark neutron spectra
The new version of the International Reactor Dosimetry File IRDF-90 (called ''Version April 1993'') has been tested by calculation of average cross-sections and their uncertainties in a coarse three energy group structure and by neutron spectrum adjustments in reference neutron spectra. This paper presents the results obtained and compares them with the corresponding ones of the old IRDF-85 and with the data of the Nuclear Data Guide for Reactor Neutron Metrology. The applicability of the new library in the field of neutron metrology is discussed. (orig.)
Benchmark results for the critical slab and sphere problem in one-speed neutron transport theory
Research highlights: → The critical slab and sphere problem in neutron transport under Case eigenfunction formalism is considered. → These equations reduce to integral expressions involving X functions. → Gauss quadrature is not ideal but DE quadrature is well-suited. → Several fold decrease in computational effort with improved accuracy is realisable. - Abstract: In this paper benchmark numerical results for the one-speed criticality problem with isotropic scattering for the slab and sphere are reported. The Fredholm integral equations of the second kind based on the Case eigenfunction formalism are numerically solved by Neumann iterations with the Double Exponential quadrature.
Prediction of neutron embrittlement in the reactor pressure vessel. Venus-1 and Venus-3 benchmarks
The OECD/NEA Task Force on Computing Radiation Dose and Modelling of Radiation-Induced Degradation of Reactor Components (TFRDD) launched two international blind intercomparison exercises to examine the current computation techniques used in NEA Member countries for calculating neutron and gamma doses to reactor components. Various methodologies and different nuclear data were applied to predict dose rates in the Belgian VENUS-1 and three-dimensional VENUS-3 configurations for comparison with measured data. This report provides the detailed results from the two benchmarks.The exercise revealed that three-dimensional neutron fluence calculations provide results that are significantly more accurate than those obtained from two-dimensional calculations. Performing three-dimensional calculations is technically feasible given the power of today's computers. (author)
Evolution of Neutron Star Magnetic Fields
Dipankar Bhattacharya
2002-03-01
This paper reviews the current status of the theoretical models of the evolution of the magnetic fields of neutron stars other than magnetars. It appears that the magnetic fields of neutron stars decay significantly only if they are in binary systems. Three major physical models for this, namely spindown-induced flux expulsion, ohmic evolution of crustal field and diamagnetic screening of the field by accreted plasma, are reviewed.
Monoenergetic fast neutron reference fields: II. Field characterization
Nolte, Ralf; Thomas, David J.
2011-12-01
Monoenergetic neutron reference fields are required for the calibration of neutron detectors and dosemeters for various applications ranging from nuclear physics and nuclear data measurements to radiation protection. In a series of two separate publications the metrological aspects of the production and measurement of fast neutrons are reviewed. In the first part, requirements for the nuclear reactions used to produce neutron fields as well as methods for target characterization and the general layout of reference facilities were discussed. This second part focuses on the most important techniques for field characterization and includes the determination of the neutron fluence as well as the spectral neutron distribution and the determination of the fluence of contaminating photons. The measurements are usually carried out relative to reference cross sections which are reviewed in a separate contribution, but for certain conditions 'absolute' methods for neutron measurements can be used which are directly traceable to the international system of units (SI).
The Benchmark experiment on stainless steel bulk shielding at the Frascati neutron generator
In the framework of the European Technology Program for NET/ITER, ENEA (Italian Agency for New Technologies, Energy and Environment) - Frascati and CEA (Commissariat a L'Energie Atomique) - Cadarache collaborated on a Bulk Shield Benchmark Experiment using the 14-MeV Frascati Neutron Generator (FNG). The aim of the experiment was to obtain accurate experimental data for improving the nuclear database and methods used in shielding designs, through a rigorous analysis of the results. The experiment consisted of the irradiation of a stainless steel block by 14-MeV neutrons. The neutron reaction rates at different depths inside the block were measured by fission chambers and activation foils characterized by different energy response ranges. The experimental results have been compared with numerical results calculated using both SN and Monte Carlo transport codes and as transport cross section library the European Fusion File (EFF). In particular, the present report describes the experimental and numerical activity, including neutron measurements and Monte Carlo calculations, carried out by the ENEA Italian Agency for New Technologies, Energy and Environment) team
The fusion reactor inventory code FISPACT, together with the European Activation File EAF, is the European reference software for calculating the neutron-induced activation of fusion reactor relevant materials. Experimental verifications (benchmarks) of the code predictions have been performed at ENEA Frascati by means of an irradiation facility consisting of a D-T neutron generator and a moderator/reflector structure which is employed to mimic the neutron spectrum at the a fusion device first wall. Various materials (vanadium alloy, SiC, AlSI 316, martensitic steel F82H, copper, tungsten, iron, niobium), candidates to e used in a fusion reactor, have been exposed to neutrons produced in the facility (about 109 n x cm-2 x s-1) and the short and medium-lived induced radioactivity has been measured by gamma-ray spectroscopy. The experimental results have been used to validate the inventory code FISPACT, the physical database EAF, including its uncertainty predictions, and the composition of the material irradiated in particular for its minor elements and impurities. The comparison between calculated (C) and experimental results (E) is reported as C/E values and shows a satisfactory agreement for almost all radionuclides. Radionuclides for which there is not agreement between calculations and experiments are also discussed and an analysis of the causes of the lack of agreement is carried out. (author)
Benchmark experiment on stainless steel bulk shielding at Frascati neutron generator
Batistoni, P.; Angelone, M.; Martone, M.; Pillon, M.; Rado, V. [ENEA, Frascati (Italy). Centro Ricerche Energia - Area Energia e Innovazione
1994-11-01
In the framework of the European Technology Program for NET/ITER, ENEA (Italian Agency for New Technologies, Energy and Environment) - Frascati and CEA (Commissariat a L`Energie Atomique) - Cadarache collaborated on a Bulk Shield Benchmark Experiment using the 14-MeV Frascati Neutron Generator (FNG). The aim of the experiment was to obtain accurate experimental data for improving the nuclear database and methods used in shielding designs, through a rigorous analysis of the results. The experiment consisted of the irradiation of a stainless steel block by 14-MeV neutrons. The neutron reaction rates at different depths inside the block were measured by fission chambers and activation foils characterized by different energy response ranges. The experimental results have been compared with numerical results calculated using both S{sub N} and Monte Carlo transport codes and as transport cross section library the European Fusion File (EFF). In particular, the present report describes the experimental and numerical activity, including neutron measurements and Monte Carlo calculations, carried out by the (ENEA Italian Agency for New Technologies, Energy and Environment) team.
Evaluation of large 3600 MWth sodium-cooled fast reactor OECD neutronic benchmarks
Within the activities of the Working Party on Scientific Issues of Reactor Systems (WPRS) of the OECD, an international collaboration is ongoing on the neutronic analyses of several Generation-IV Sodium-cooled Fast Reactor (SFR) concepts. This paper summarizes the results obtained by participants from institutions of different countries (ANL, CEA, ENEA, HZDR, JAEA, CER, KIT, UIUC) for the large core numerical benchmarks. These results have been obtained using different calculation methods and analysis tools to estimate the core reactivity and isotopic composition evolution, neutronic feedbacks and power distribution. For the different core concepts analyzed, a satisfactory agreement was obtained between participants despite the different calculation schemes used. A good agreement was generally obtained when comparing compositions after burnup, the delayed neutron fraction, the Doppler coefficient, and the sodium void worth. However, some noticeable discrepancies between the k-effective values were observed and are explained in this paper. These are mostly due to the different neutronic libraries employed (JEFF3.1, ENDFB7.0 or JENDL-4.0) and to a lesser extent the calculations methods. (author)
SCALE Modeling of Selected Neutronics Test Problems within the OECD UAM LWR’s Benchmark
Luigi Mercatali
2013-01-01
Full Text Available The OECD UAM Benchmark was launched in 2005 with the objective of determining the uncertainty in the simulation of Light Water Reactors (LWRs system calculations at all the stages of the coupled reactor physics—thermal hydraulics modeling. Within the framework of the “Neutronics Phase” of the Benchmark the solutions of some selected test cases at the cell physics and lattice physics levels are presented. The SCALE 6.1 code package has been used for the neutronics modeling of the selected exercises. Sensitivity and Uncertainty analysis (S/U based on the generalized perturbation theory has been performed in order to assess the uncertainty of the computation of some selected reactor integral parameters due to the uncertainty in the basic nuclear data. As a general trend, it has been found that the main sources of uncertainty are the 238U (n, and the 239Pu nubar for the UOX- and the MOX-fuelled test cases, respectively. Moreover, the reference solutions for the test cases obtained using Monte Carlo methodologies together with a comparison between deterministic and stochastic solutions are presented.
Reference neutron fields for radiation monitoring
A set of reference neutron fields, generated at the Joint Institute of Nuclear Research (JINR) for radiation monitoring, is described. A calibration algorithm is proposed for detectors used in express and individual monitoring. The reference fields cover practically the whole energy range of neutron radiation produced at the nuclear installations of the JINR (from 10-8 up to hundreds of megaelectronvolts). The set includes the fields obtained from a 252Cf source in polyethylene moderators 12.7 and 29.2 cm in diameter; a soft-radiation field produced by multiply scattered neutrons in a synchrocyclotron labyrinth; and the hard radiation field formed by neutrons leaking from a whole concrete shield surrounding a synchrocyclotron when the shield is irradiated by the secondary radiation from the accelerator chamber and target. The latter two fields appear when the synchrocyclotron of the Laboratory of Nuclear Problems operates at an energy of 660 MeV
Rose, P. F.; Alter, H.; Paschall, R. K.; Thiele, A. W.
1973-01-15
The experimental details and the calculational specifications for a CSEWG integral data test shielding experiment are presented. The shielding experiment described in the benchmark model is a combination of sodium and stainless steel that simulates the FFTF radial shield. The measurements in general include use of foil activation techniques using resonance and threshold detectors and proton recoil neutron spectrometer measurements in the range 5 kev to 2 MeV. The benchmark model is a test of the neutron cross-section data for sodium and the material components of stainless steel.
Neutron interferometry constrains dark energy chameleon fields
H. Lemmel
2015-04-01
Full Text Available We present phase shift measurements for neutron matter waves in vacuum and in low pressure Helium using a method originally developed for neutron scattering length measurements in neutron interferometry. We search for phase shifts associated with a coupling to scalar fields. We set stringent limits for a scalar chameleon field, a prominent quintessence dark energy candidate. We find that the coupling constant β is less than 1.9×107 for n=1 at 95% confidence level, where n is an input parameter of the self-interaction of the chameleon field φ inversely proportional to φn.
Benchmarking mean-field approximations to level densities
Alhassid, Y.; Bertsch, G. F.; Gilbreth, C. N.; Nakada, H.
2016-04-01
We assess the accuracy of finite-temperature mean-field theory using as a standard the Hamiltonian and model space of the shell model Monte Carlo calculations. Two examples are considered: the nucleus 162Dy, representing a heavy deformed nucleus, and 148Sm, representing a nearby heavy spherical nucleus with strong pairing correlations. The errors inherent in the finite-temperature Hartree-Fock and Hartree-Fock-Bogoliubov approximations are analyzed by comparing the entropies of the grand canonical and canonical ensembles, as well as the level density at the neutron resonance threshold, with shell model Monte Carlo calculations, which are accurate up to well-controlled statistical errors. The main weak points in the mean-field treatments are found to be: (i) the extraction of number-projected densities from the grand canonical ensembles, and (ii) the symmetry breaking by deformation or by the pairing condensate. In the absence of a pairing condensate, we confirm that the usual saddle-point approximation to extract the number-projected densities is not a significant source of error compared to other errors inherent to the mean-field theory. We also present an alternative formulation of the saddle-point approximation that makes direct use of an approximate particle-number projection and avoids computing the usual three-dimensional Jacobian of the saddle-point integration. We find that the pairing condensate is less amenable to approximate particle-number projection methods because of the explicit violation of particle-number conservation in the pairing condensate. Nevertheless, the Hartree-Fock-Bogoliubov theory is accurate to less than one unit of entropy for 148Sm at the neutron threshold energy, which is above the pairing phase transition. This result provides support for the commonly used "back-shift" approximation, treating pairing as only affecting the excitation energy scale. When the ground state is strongly deformed, the Hartree-Fock entropy is significantly
The stainless steel bulk shielding benchmark experiment at the Frascati Neutron Generator (FNG)
Batistoni, P. (Associazione Euratom-ENEA sulla Fusione, CRE Frascati, I-00044 Frascati, Rome (Italy)); Angelone, M. (Associazione Euratom-ENEA sulla Fusione, CRE Frascati, I-00044 Frascati, Rome (Italy)); Martone, M. (Associazione Euratom-ENEA sulla Fusione, CRE Frascati, I-00044 Frascati, Rome (Italy)); Petrizzi, L. (Associazione Euratom-ENEA sulla Fusione, CRE Frascati, I-00044 Frascati, Rome (Italy)); Pillon, M. (Associazione Euratom-ENEA sulla Fusione, CRE Frascati, I-00044 Frascati, Rome (Italy)); Rado, V. (Associazione Euratom-ENEA sulla Fusione, CRE Frascati, I-00044 Frascati, Rome (Italy)); Santamarina, A. (Commissariat a l' Energie Atomique, Centre d' Etudes Nucleaires Cadarache, 13108, St.-Paul-lez-Durance Cedex (France)); Abidi, I. (Commissariat a l' Energie Atomique, Centre d' Etudes Nucleaires Cadarache, 13108, St.-Paul-lez-Durance Cedex (France)); Gastaldi, G. (Commissariat a l' Energie Atomique, Centre d' Etudes Nucleaires Cadarache, 13108, St.-Paul-lez-Durance Cedex
1994-09-01
In the framework of the European Technology Program for NET/ITER, ENEA (Ente Nazionale per le Nuove Tecnologie, l'Energia e l'Ambiente), Frascati and CEA (Commissariat a l'Energie Atomique), Cadarache, are collaborating on a bulk shielding benchmark experiment using the 14 MeV Frascati Neutron Generator (FNG). The aim of the experiment is to obtain accurate experimental data for improving the nuclear database and methods used in the shielding designs, through a rigorous analysis of the results. The experiment consists of the irradiation of a stainless steel block by 14 MeV neutrons. The neutron flux and spectra at different depths, up to 65 cm inside the block, are measured by fission chambers and activation foils characterized by different energy response ranges. The [gamma]-ray dose measurements are performed with ionization chambers and thermo-luminescent dosimeters (TLD). The first results are presented, as well as the comparison with calculations using the cross section library EFF (European Fusion File). ((orig.))
The stainless steel bulk shielding benchmark experiment at the Frascati Neutron Generator (FNG)
Batistoni, P.; Angelone, M.; Martone, M.; Petrizzi, L.; Pillon, M.; Rado, V.; Santamarina, A.; Abidi, I.; Gastaldi, G.; Joyer, P.; Marquette, J. P.; Martini, M.
1994-09-01
In the framework of the European Technology Program for NET/ITER, ENEA (Ente Nazionale per le Nuove Tecnologie, l'Energia e l'Ambiente), Frascati and CEA (Commissariat à l'Energie Atomique), Cadarache, are collaborating on a bulk shielding benchmark experiment using the 14 MeV Frascati Neutron Generator (FNG). The aim of the experiment is to obtain accurate experimental data for improving the nuclear database and methods used in the shielding designs, through a rigorous analysis of the results. The experiment consists of the irradiation of a stainless steel block by 14 MeV neutrons. The neutron flux and spectra at different depths, up to 65 cm inside the block, are measured by fission chambers and activation foils characterized by different energy response ranges. The γ-ray dose measurements are performed with ionization chambers and thermo-luminescent dosimeters (TLD). The first results are presented, as well as the comparison with calculations using the cross section library EFF (European Fusion File).
Peneliau, Y.; Litaize, O.; Archier, P.; De Saint Jean, C.
2014-04-01
A large set of nuclear data are investigated to improve the calculation predictions of the new neutron transport simulation codes. With the next generation of nuclear power plants (GEN IV projects), one expects to reduce the calculated uncertainties which are mainly coming from nuclear data and are still very important, before taking into account integral information in the adjustment process. In France, future nuclear power plant concepts will probably use MOX fuel, either in Sodium Fast Reactors or in Gas Cooled Fast Reactors. Consequently, the knowledge of 239Pu cross sections and other nuclear data is crucial issue in order to reduce these sources of uncertainty. The Prompt Fission Neutron Spectra (PFNS) for 239Pu are part of these relevant data (an IAEA working group is even dedicated to PFNS) and the work presented here deals with this particular topic. The main international data files (i.e. JEFF-3.1.1, ENDF/B-VII.0, JENDL-4.0, BRC-2009) have been considered and compared with two different spectra, coming from the works of Maslov and Kornilov respectively. The spectra are first compared by calculating their mathematical moments in order to characterize them. Then, a reference calculation using the whole JEFF-3.1.1 evaluation file is performed and compared with another calculation performed with a new evaluation file, in which the data block containing the fission spectra (MF=5, MT=18) is replaced by the investigated spectra (one for each evaluation). A set of benchmarks is used to analyze the effects of PFNS, covering criticality cases and mock-up cases in various neutron flux spectra (thermal, intermediate, and fast flux spectra). Data coming from many ICSBEP experiments are used (PU-SOL-THERM, PU-MET-FAST, PU-MET-INTER and PU-MET-MIXED) and French mock-up experiments are also investigated (EOLE for thermal neutron flux spectrum and MASURCA for fast neutron flux spectrum). This study shows that many experiments and neutron parameters are very sensitive to
Benchmarking mean-field approximations to level densities
Alhassid, Y; Gilbreth, C N; Nakada, H
2015-01-01
We assess the accuracy of finite-temperature mean-field theory using as a standard the Hamiltonian and model space of the shell model Monte Carlo calculations. Two examples are considered: the nucleus $^{162}$Dy, representing a heavy deformed nucleus, and $^{148}$Sm, representing a nearby heavy spherical nucleus with strong pairing correlations. The errors inherent in the finite-temperature Hartree-Fock and Hartree-Fock-Bogoliubov approximations are analyzed by comparing the entropies of the grand canonical and canonical ensembles, as well as the level density at the neutron resonance threshold, with shell model Monte Carlo (SMMC) calculations, which are accurate up to well-controlled statistical errors. The main weak points in the mean-field treatments are seen to be: (i) the extraction of number-projected densities from the grand canonical ensembles, and (ii) the symmetry breaking by deformation or by the pairing condensate. In the absence of a pairing condensate, we confirm that the usual saddle-point appr...
DOE–CEA Benchmark on SFR ASTRID Innovative Core: Neutronic and Safety Transients Simulation
ASTRID is a fast reactor being designed by the CEA to achieve a level of safety that exceeds that of conventional fast reactors. In particular, an axially heterogeneous core with an upper sodium plenum is employed to achieve a non-positive sodium void reactivity worth. In order to address the simulation challenges for this innovative concept, the US Department of Energy’s (DOE) Laboratories (Argonne National Laboratory and Idaho National Laboratory) and the CEA are performing neutronic and transient benchmark calculations for an ASTRID model based on design specifications provided by the CEA. The blind comparison of the initial DOE and CEA results are found to be in good agreement, enhancing confidence in CEA predictions of key ASTRID safety relevant parameters and transient behaviour. For several parameters, compared uncertainties in computed values are significant and further studies are needed to reduce them. (author)
Benchmark experiments of effective delayed neutron fraction βeff at FCA
Benchmark experiments of effective delayed neutron fraction βeff were performed at Fast Critical Assembly (FCA) in the Japan Atomic Energy Research Institute. The experiments were made in three cores providing systematic change of nuclide contribution to the βeff: XIX-1 core fueled with 93% enriched uranium, XIX-2 core fueled with plutonium and uranium (23% enrichment) and XIX-3 core fueled with plutonium (92% fissile Pu). Six organizations from five countries participated in these experiments and measured the βeff by using their own methods and instruments. Target accuracy in the βeff was achieved to be better than ±3% by averaging the βeff values measured using a wide variety of experimental methods. (author)
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)
Magnetic Field Evolution During Neutron Star Recycling
Cumming, A
2004-01-01
I describe work on two aspects of magnetic field evolution relevant for the "recycling" scenario for making millisecond radio pulsars. First, many of the theoretical ideas for bringing about accretion-induced field decay rely on dissipation of currents in the neutron star crust. I discuss field evolution in the crust due to the Hall effect, and outline when it dominates Ohmic decay. This emphasises the importance of understanding the impurity level in the crust. Second, I briefly discuss the progress that has been made in understanding the magnetic fields of neutron stars currently accreting matter in low mass X-ray binaries. In particular, thermonuclear X-ray bursts offer a promising probe of the magnetic field of these neutron stars.
Neutron field characteristics of Ciemat's Neutron Standards Laboratory Hector Rene Vega-Carrillo
Guzmán-García, Karen Arlete; Méndez Villafañe, Roberto; Vega-Carrillo, Héctor René
2015-01-01
Monte Carlo calculations were carried out to characterize the neutron field produced by the calibration neutron sources of the Neutron Standards Laboratory at the Research Center for Energy, Environment and Technology (CIEMAT) in Spain. For 241AmBe and 252Cf neutron sources, the neutron spectra, the ambient dose equivalent rates and the total neutron fluence rates were estimated. In the calibration hall, there are several items that modify the neutron field. To evaluate their effects differen...
Parareal in time 3D numerical solver for the LWR Benchmark neutron diffusion transient model
In this paper we present a time-parallel algorithm for the 3D neutrons calculation of a transient model in a nuclear reactor core. The neutrons calculation consists in numerically solving the time dependent diffusion approximation equation, which is a simplified transport equation. The numerical resolution is done with finite elements method based on a tetrahedral meshing of the computational domain, representing the reactor core, and time discretization is achieved using a θ-scheme. The transient model presents moving control rods during the time of the reaction. Therefore, cross-sections (piecewise constants) are taken into account by interpolations with respect to the velocity of the control rods. The parallelism across the time is achieved by an adequate use of the parareal in time algorithm to the handled problem. This parallel method is a predictor corrector scheme that iteratively combines the use of two kinds of numerical propagators, one coarse and one fine. Our method is made efficient by means of a coarse solver defined with large time step and fixed position control rods model, while the fine propagator is assumed to be a high order numerical approximation of the full model. The parallel implementation of our method provides a good scalability of the algorithm. Numerical results show the efficiency of the parareal method on large light water reactor transient model corresponding to the Langenbuch–Maurer–Werner benchmark
Parareal in time 3D numerical solver for the LWR Benchmark neutron diffusion transient model
Baudron, Anne-Marie, E-mail: anne-marie.baudron@cea.fr [Laboratoire de Recherche Conventionné MANON, CEA/DEN/DANS/DM2S and UPMC-CNRS/LJLL (France); CEA-DRN/DMT/SERMA, CEN-Saclay, 91191 Gif sur Yvette Cedex (France); Lautard, Jean-Jacques, E-mail: jean-jacques.lautard@cea.fr [Laboratoire de Recherche Conventionné MANON, CEA/DEN/DANS/DM2S and UPMC-CNRS/LJLL (France); CEA-DRN/DMT/SERMA, CEN-Saclay, 91191 Gif sur Yvette Cedex (France); Maday, Yvon, E-mail: maday@ann.jussieu.fr [Sorbonne Universités, UPMC Univ Paris 06, UMR 7598, Laboratoire Jacques-Louis Lions and Institut Universitaire de France, F-75005, Paris (France); Laboratoire de Recherche Conventionné MANON, CEA/DEN/DANS/DM2S and UPMC-CNRS/LJLL (France); Brown Univ, Division of Applied Maths, Providence, RI (United States); Riahi, Mohamed Kamel, E-mail: riahi@cmap.polytechnique.fr [Laboratoire de Recherche Conventionné MANON, CEA/DEN/DANS/DM2S and UPMC-CNRS/LJLL (France); CMAP, Inria-Saclay and X-Ecole Polytechnique, Route de Saclay, 91128 Palaiseau Cedex (France); Salomon, Julien, E-mail: salomon@ceremade.dauphine.fr [CEREMADE, Univ Paris-Dauphine, Pl. du Mal. de Lattre de Tassigny, F-75016, Paris (France)
2014-12-15
In this paper we present a time-parallel algorithm for the 3D neutrons calculation of a transient model in a nuclear reactor core. The neutrons calculation consists in numerically solving the time dependent diffusion approximation equation, which is a simplified transport equation. The numerical resolution is done with finite elements method based on a tetrahedral meshing of the computational domain, representing the reactor core, and time discretization is achieved using a θ-scheme. The transient model presents moving control rods during the time of the reaction. Therefore, cross-sections (piecewise constants) are taken into account by interpolations with respect to the velocity of the control rods. The parallelism across the time is achieved by an adequate use of the parareal in time algorithm to the handled problem. This parallel method is a predictor corrector scheme that iteratively combines the use of two kinds of numerical propagators, one coarse and one fine. Our method is made efficient by means of a coarse solver defined with large time step and fixed position control rods model, while the fine propagator is assumed to be a high order numerical approximation of the full model. The parallel implementation of our method provides a good scalability of the algorithm. Numerical results show the efficiency of the parareal method on large light water reactor transient model corresponding to the Langenbuch–Maurer–Werner benchmark.
Analytical benchmarks for nuclear engineering applications. Case studies in neutron transport theory
Preservation of know-how in the nuclear field is promoted through the activities of the OECD Nuclear Energy Agency Data Bank. One area of importance concerns methods for solving radiation transport problems, especially with regard to neutrons. This handbook (in the form of a case study), prepared by Barry D Ganapol, is the result of such an initiative. It is a compilation of solutions to the transport equation for which analytical representations can be found. It is designed for educational use in courses on analytical transport methods and numerical methods with application to reactor physics. In addition, it contains elements for the continuous improvement of transport methods and for computer code verification. The areas of neutron slowing down, thermalization and one-, two- and three-dimensional neutron transport theory are covered. A series of training courses, based on this compilation of solutions has recently begun. (author)
Benchmarking shielding simulations for an accelerator-driven spallation neutron source
Cherkashyna, Nataliia; DiJulio, Douglas D.; Panzner, Tobias; Rantsiou, Emmanouela; Filges, Uwe; Ehlers, Georg; Bentley, Phillip M.
2015-08-01
The shielding at an accelerator-driven spallation neutron facility plays a critical role in the performance of the neutron scattering instruments, the overall safety, and the total cost of the facility. Accurate simulation of shielding components is thus key for the design of upcoming facilities, such as the European Spallation Source (ESS), currently in construction in Lund, Sweden. In this paper, we present a comparative study between the measured and the simulated neutron background at the Swiss Spallation Neutron Source (SINQ), at the Paul Scherrer Institute (PSI), Villigen, Switzerland. The measurements were carried out at several positions along the SINQ monolith wall with the neutron dosimeter WENDI-2, which has a well-characterized response up to 5 GeV. The simulations were performed using the Monte-Carlo radiation transport code geant4, and include a complete transport from the proton beam to the measurement locations in a single calculation. An agreement between measurements and simulations is about a factor of 2 for the points where the measured radiation dose is above the background level, which is a satisfactory result for such simulations spanning many energy regimes, different physics processes and transport through several meters of shielding materials. The neutrons contributing to the radiation field emanating from the monolith were confirmed to originate from neutrons with energies above 1 MeV in the target region. The current work validates geant4 as being well suited for deep-shielding calculations at accelerator-based spallation sources. We also extrapolate what the simulated flux levels might imply for short (several tens of meters) instruments at ESS.
Ford, Donald J.
1993-01-01
Discusses benchmarking, the continuous process of measuring one's products, services, and practices against those recognized as leaders in that field to identify areas for improvement. Examines ways in which benchmarking can benefit human resources functions. (JOW)
Hebeler, K.; Schwenk, A.
2014-01-01
We discuss neutron matter calculations based on chiral effective field theory interactions and their predictions for the symmetry energy, the neutron skin of 208 Pb, and for the radius of neutron stars.
Applications of Neutron Bubble Dosimeters for Neutron Dose Monitoring in Mixed n-γ Fields
2008-01-01
<正>Bubble dosimeter is a promising technology in the field of neutron dosimetry. It provides real-time monitoring of neutron dose, stable energy response over wide range of neutron energy, and a very low
Ellis, R.J.
2001-01-11
A series of unit pin-cell benchmark problems have been analyzed related to irradiation of mixed oxide fuel in VVER-1000s (water-water energetic reactors). One-dimensional, discrete-ordinates eigenvalue calculations of these benchmarks were performed at ORNL using the SAS2H control sequence module of the SCALE-4.3 computational code system, as part of the Fissile Materials Disposition Program (FMDP) of the US DOE. Calculations were also performed using the SCALE module CSAS to confirm the results. The 238 neutron energy group SCALE nuclear data library 238GROUPNDF5 (based on ENDF/B-V) was used for all calculations. The VVER-1000 pin-cell benchmark cases modeled with SAS2H included zero-burnup calculations for eight fuel material variants (from LEU UO{sub 2} to weapons-grade MOX) at five different reactor states, and three fuel depletion cases up to high burnup. Results of the SAS2H analyses of the VVER-1000 neutronics benchmarks are presented in this report. Good general agreement was obtained between the SAS2H results, the ORNL results using HELIOS-1.4 with ENDF/B-VI nuclear data, and the results from several Russian benchmark studies using the codes TVS-M, MCU-RFFI/A, and WIMS-ABBN. This SAS2H benchmark study is useful for the verification of HELIOS calculations, the HELIOS code being the principal computational tool at ORNL for physics studies of assembly design for weapons-grade plutonium disposition in Russian reactors.
Real fields of neutron reference in Argentina
In order to improve the personal and area dose determination of mixed fields working areas the characterization of the radiation field inside the zone of experimental reactor, RA-1, have been made. The installation is a representative working place. Personal dosemeters belonging to the ARN and those from external personal neutron dosimetry laboratories have been calibrated in Hp(10) quantities as a first aim. The calibration points were determined using the multisphere neutron spectrometric system (MNSS) coupled with TLD for gamma measurements. The MNSS has a set of 12 high density polyethylene spheres (diameters from 2' till 15') a 3 He detector, 4 atm pressure located in the centre of the spheres and the associated electronics. The neutron response matrix for our MNSS was calculated using the MCNP Monte Carlo 4B code version 4B with the cross sections library ENDF/B-VI in the energy range between thermal neutron and 100 MeV. The neutron spectrum was obtained using the LOUHI82 deconvolution code. The calibration of the system was validated using a Am Be source with an fluence error less than 10%. In this work the spectrum data obtained with the MNSS is shown. (author)
Benchmark study on neutron cross sections based on pulsed sphere experiment
Benchmark validation of neutron cross sections was performed by comparing theoretical calculated leakage spectra with measured ones by means of pulsed sphere experiment conducted at OKTAVIAN of Osaka University, FZK, German and IPPE Obninsk, Russia. It was found out that the nuclear data of Be in both JENDL Fusion File and ENDF/B-VI had similar trend in each experiment. However, there exists some discrepancy among the three different experiments, which suggests further study is needed to validate Be nuclear data. Calculated spectra for Li, Cr, Mn, Cu, Zr, Nb and Mo using JENDL Fusion File predict the experiment fairly well. However, for LiF, (CF2)n, Si, Ti, Co and W, the calculated spectra are not in good agreement with the measurement. The prediction using FENDL-1.0 data gives NN agreement in case of Li, Cr, Mn, Cu and Mo, whereas in other case, the prediction gives insufficient result. The analysis of Fe and Pb experiment conducted at IPPE Obninsk showed that Fe data in JENDL Fusion File and JENDL-3.2 were much better than FENDL/E-1.0. Pb data in JENDL Fusion File appeared to have been much improved as compared with JENDL-3.2 evaluation. (author)
Description of benchmark: This international benchmark, concerns Pebble-Bed Modular Reactor (PBMR) coupled neutronics/thermal hydraulics transients based on the PBMR-400 MW design. The deterministic neutronics, thermal-hydraulics and transient analysis tools and methods available to design and analyse PBMRs lag, in many cases, behind the state of the art compared to other reactor technologies. This has motivated the testing of existing methods for HTGRs but also the development of more accurate and efficient tools to analyse the neutronics and thermal-hydraulic behaviour for the design and safety evaluations of the PBMR. In addition to the development of new methods, this includes defining appropriate benchmarks to verify and validate the new methods in computer codes. The scope of the benchmark is to establish well-defined problems, based on a common given set of cross sections, to compare methods and tools in core simulation and thermal hydraulics analysis with a specific focus on transient events through a set of multi-dimensional computational test problems. The benchmark exercise has the following objectives: - Establish a standard benchmark for coupled codes (neutronics/thermal-hydraulics) for PBMR design; - Code-to-code comparison using a common cross section library ; - Obtain a detailed understanding of the events and the processes; - Benefit from different approaches, understanding limitations and approximations. Major Design and Operating Characteristics of the PBMR (PBMR Characteristic and Value): Installed thermal capacity: 400 MW(t); Installed electric capacity: 165 MW(e); Load following capability: 100-40-100%; Availability: ≥ 95%; Core configuration: Vertical with fixed centre graphite reflector; Fuel: TRISO ceramic coated U-235 in graphite spheres; Primary coolant: Helium; Primary coolant pressure: 9 MPa; Moderator: Graphite; Core outlet temperature: 900 C.; Core inlet temperature: 500 C.; Cycle type: Direct; Number of circuits: 1; Cycle
Neutron stars and their magnetic fields
Reisenegger, Andreas
2008-01-01
Neutron stars have the strongest magnetic fields known anywhere in the Universe. In this review, I intend to give a pedagogical discussion of some of the related physics. Neutron stars exist because of Pauli's exclusion principle, in two senses: 1) It makes it difficult to squeeze particles too close together, in this way allowing a mechanical equilibrium state in the presence of extremely strong gravity. 2) The occupation of low-energy proton and electron states makes it impossible for low-energy neutrons to beta decay. A corollary of the second statement is that charged particles are necessarily present inside a neutron star, allowing currents to flow. Since these particles are degenerate, they collide very little, and therefore make it possible for the star to support strong, organized magnetic fields over long times. These show themselves in pulsars and are the most likely energy source for the high X-ray and gamma-ray luminosity ``magnetars''. I briefly discuss the possible origin of this field and some ...
Flux Expulsion Field Evolution in Neutron Stars
Jahan-Miri, M
1999-01-01
Models for the evolution of magnetic fields of neutron stars are constructed, assuming the field is embedded in the proton superconducting core of the star. The rate of expulsion of the magnetic flux out of the core, or equivalently the velocity of outward motion of flux-carrying proton-vortices is determined from a solution of the Magnus equation of motion for these vortices. A force due to the pinning interaction between the proton-vortices and the neutron-superfluid vortices is also taken into account in addition to the other more conventional forces acting on the proton-vortices. Alternative models for the field evolution are considered based on the different possibilities discussed for the effective values of the various forces. The coupled spin and magnetic evolution of single pulsars as well as those processed in low-mass binary systems are computed, for each of the models. The predicted lifetimes of active pulsars, field strengths of the very old neutron stars, and distribution of the magnetic fields ...
Gravitational field energy contribution to the neutron star mass
Dyrda, M.; Kinasiewicz, B.; Kutschera, M.; Szmaglinski, A.
2006-01-01
Neutron stars are discussed as laboratories of physics of strong gravitational fields. The mass of a neutron star is split into matter energy and gravitational field energy contributions. The energy of the gravitational field of neutron stars is calculated with three different approaches which give the same result. It is found that up to one half of the gravitational mass of maximum mass neutron stars is comprised by the gravitational field energy. Results are shown for a number of realistic ...
Magnetic field evolution of accreting neutron stars
Istomin, Ya N
2016-01-01
The flow of a matter, accreting onto a magnetized neutron star, is accompanied by an electric current. The closing of the electric current occurs in the crust of a neutron stars in the polar region across the magnetic field. But the conductivity of the crust along the magnetic field greatly exceeds the conductivity across the field, so the current penetrates deep into the crust down up to the super conducting core. The magnetic field, generated by the accretion current, increases greatly with the depth of penetration due to the Hall conductivity of the crust is also much larger than the transverse conductivity. As a result, the current begins to flow mainly in the toroidal direction, creating a strong longitudinal magnetic field, far exceeding an initial dipole field. This field exists only in the narrow polar tube of $r$ width, narrowing with the depth, i.e. with increasing of the crust density $\\rho$, $r\\propto \\rho^{-1/4}$. Accordingly, the magnetic field $B$ in the tube increases with the depth, $B\\propto...
The three-dimensional continuous-energy Monte Carlo code MCNP4C was used to develop a versatile and accurate full-core model of the TRIGA MARK II research reactor at AERE, Savar. Thr consistency and accuracy of both the Monte Carlo simulation and neutron transport physics was established by benchmarking the TRIGA experiments. Analysis of neutron flux and reactivity experiments comprising control rod worths, critical rod height, excess reactivity and shutdown margin were used in the validation process. Calculations of fast neutron flux, and fuel and graphite element worths distribution are also presented. Good agreement between the experiments and MCNP calculations indicate that the simulation of TRIGA reactor is treated adequately. (author)
Walsh, Jonathan A.; Forget, Benoit; Smith, Kord S.; Brown, Forrest B.
2016-03-01
In this work we describe the development and application of computational methods for processing neutron cross section data in the unresolved resonance region (URR). These methods are integrated with a continuous-energy Monte Carlo neutron transport code, thereby enabling their use in high-fidelity analyses. Enhanced understanding of the effects of URR evaluation representations on calculated results is then obtained through utilization of the methods in Monte Carlo integral benchmark simulations of fast spectrum critical assemblies. First, we present a so-called on-the-fly (OTF) method for calculating and Doppler broadening URR cross sections. This method proceeds directly from ENDF-6 average unresolved resonance parameters and, thus, eliminates any need for a probability table generation pre-processing step in which tables are constructed at several energies for all desired temperatures. Significant memory reduction may be realized with the OTF method relative to a probability table treatment if many temperatures are needed. Next, we examine the effects of using a multi-level resonance formalism for resonance reconstruction in the URR. A comparison of results obtained by using the same stochastically-generated realization of resonance parameters in both the single-level Breit-Wigner (SLBW) and multi-level Breit-Wigner (MLBW) formalisms allows for the quantification of level-level interference effects on integrated tallies such as keff and energy group reaction rates. Though, as is well-known, cross section values at any given incident energy may differ significantly between single-level and multi-level formulations, the observed effects on integral results are minimal in this investigation. Finally, we demonstrate the calculation of true expected values, and the statistical spread of those values, through independent Monte Carlo simulations, each using an independent realization of URR cross section structure throughout. It is observed that both probability table
Walsh Jonathan A.
2016-01-01
Full Text Available In this work we describe the development and application of computational methods for processing neutron cross section data in the unresolved resonance region (URR. These methods are integrated with a continuous-energy Monte Carlo neutron transport code, thereby enabling their use in high-fidelity analyses. Enhanced understanding of the effects of URR evaluation representations on calculated results is then obtained through utilization of the methods in Monte Carlo integral benchmark simulations of fast spectrum critical assemblies. First, we present a so-called on-the-fly (OTF method for calculating and Doppler broadening URR cross sections. This method proceeds directly from ENDF-6 average unresolved resonance parameters and, thus, eliminates any need for a probability table generation pre-processing step in which tables are constructed at several energies for all desired temperatures. Significant memory reduction may be realized with the OTF method relative to a probability table treatment if many temperatures are needed. Next, we examine the effects of using a multi-level resonance formalism for resonance reconstruction in the URR. A comparison of results obtained by using the same stochastically-generated realization of resonance parameters in both the single-level Breit-Wigner (SLBW and multi-level Breit-Wigner (MLBW formalisms allows for the quantification of level-level interference effects on integrated tallies such as keff and energy group reaction rates. Though, as is well-known, cross section values at any given incident energy may differ significantly between single-level and multi-level formulations, the observed effects on integral results are minimal in this investigation. Finally, we demonstrate the calculation of true expected values, and the statistical spread of those values, through independent Monte Carlo simulations, each using an independent realization of URR cross section structure throughout. It is observed that both
Photon spectrometry in thermal neutron standard field
Kudo, K; Koshikawa, S; Toyokawa, H; Ohgaki, H; Matzke, M
2002-01-01
An NE213 liquid scintillation counter (5.08 cm in diameter and 5.08 cm long) with an LiF filter was used to measure the energy distribution of photons mixed in a thermal neutron field. The response function matrix of photons in an energy range up to 10 MeV was calculated by the EGS4/PRESTA code and properly folded with a resolution function. Pulse height spectra measured with a set of reference gamma-ray sources were compared to the calculated response function and agreed very well for all reference gamma-ray sources. The GRAVEL and MIEKE codes from the HEPRO program were used to unfold measured pulse height spectra. Energy distributions obtained by the unfolding were applied to evaluate the effective dose equivalent of photons mixed in a thermal neutron field.
Calibration of the IRD two-component TLD albedo neutron dosemeter in some moderated neutron fields
Freitas, Bruno M.; Silva, Ademir X. da, E-mail: bfreitas@nuclear.ufrj.br, E-mail: ademir@nuclear.ufrj.br [Coordenacao dos Programas de Pos-Graduacao em Engenharia (PEN/COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Programa de Engenharia Nuclear; Martins, Marcelo M.; Pereira, Walsan W.; Mauricio, Claudia L.P., E-mail: marcelo@ird.gov.br, E-mail: walsan@ird.gov.br, E-mail: claudia@ird.gov.br [Instituto de Radioprotecao e Dosimetria (IRD/CNEN-RJ), Rio de Janeiro, RJ (Brazil)
2015-07-01
In some stray neutron fields, like those found in practices involving the handling of radionuclide sources, the neutron calibration factor for albedo neutron dosemeter can vary widely compared to the factor for bare sources. This is the case for well logging, which is the area with the largest number of workers exposed to neutrons in Brazil. The companies employ routinely {sup 241}Am-Be neutron sources. The albedo response variation is mainly due to the presence of scattered and moderated neutrons. This paper studies the response variation of the two-component TLD albedo neutron dosemeter used in the neutron individual monitoring service of Instituto de Radioprotecao e Dosimetria, in different radionuclide neutron source beams. The neutron spectra were evaluated applying a Bonner sphere spectrometer with a {sup 6}LiI(Eu) detector in the Brazilian National Metrology Neutron Laboratory. Standard neutron sources of {sup 241}Am-Be and {sup 252}Cf were employed, besides {sup 238}Pu-Be. Measurements were also made with scattered and moderated neutron beams, including {sup 252}Cf(D{sub 2}O) reference spectrum, {sup 241}Am-Be moderated with paraffin and silicone and a thermal neutron flux facility. New neutron calibration factors, as a function of the incident to albedo neutron ratio, were proposed for use in the albedo algorithm for occupational fields where the primary neutron beam is one of those studied sources. (author)
Calibration of the IRD two-component TLD albedo neutron dosemeter in some moderated neutron fields
In some stray neutron fields, like those found in practices involving the handling of radionuclide sources, the neutron calibration factor for albedo neutron dosemeter can vary widely compared to the factor for bare sources. This is the case for well logging, which is the area with the largest number of workers exposed to neutrons in Brazil. The companies employ routinely 241Am-Be neutron sources. The albedo response variation is mainly due to the presence of scattered and moderated neutrons. This paper studies the response variation of the two-component TLD albedo neutron dosemeter used in the neutron individual monitoring service of Instituto de Radioprotecao e Dosimetria, in different radionuclide neutron source beams. The neutron spectra were evaluated applying a Bonner sphere spectrometer with a 6LiI(Eu) detector in the Brazilian National Metrology Neutron Laboratory. Standard neutron sources of 241Am-Be and 252Cf were employed, besides 238Pu-Be. Measurements were also made with scattered and moderated neutron beams, including 252Cf(D2O) reference spectrum, 241Am-Be moderated with paraffin and silicone and a thermal neutron flux facility. New neutron calibration factors, as a function of the incident to albedo neutron ratio, were proposed for use in the albedo algorithm for occupational fields where the primary neutron beam is one of those studied sources. (author)
Benchmarking of Force Fields for Molecule-Membrane Interactions.
Paloncýová, Markéta; Fabre, Gabin; DeVane, Russell H; Trouillas, Patrick; Berka, Karel; Otyepka, Michal
2014-09-01
Studies of drug-membrane interactions witness an ever-growing interest, as penetration, accumulation, and positioning of drugs play a crucial role in drug delivery and metabolism in human body. Molecular dynamics simulations complement nicely experimental measurements and provide us with new insight into drug-membrane interactions; however, the quality of the theoretical data dramatically depends on the quality of the force field used. We calculated the free energy profiles of 11 molecules through a model dimyristoylphosphatidylcholine (DMPC) membrane bilayer using five force fields, namely Berger, Slipids, CHARMM36, GAFFlipids, and GROMOS 43A1-S3. For the sake of comparison, we also employed the semicontinuous tool COSMOmic. High correlation was observed between theoretical and experimental partition coefficients (log K). Partition coefficients calculated by all-atomic force fields (Slipids, CHARMM36, and GAFFlipids) and COSMOmic differed by less than 0.75 log units from the experiment and Slipids emerged as the best performing force field. This work provides the following recommendations (i) for a global, systematic and high throughput thermodynamic evaluations (e.g., log K) of drugs COSMOmic is a tool of choice due to low computational costs; (ii) for studies of the hydrophilic molecules CHARMM36 should be considered; and (iii) for studies of more complex systems, taking into account all pros and cons, Slipids is the force field of choice. PMID:26588554
Reference mixed neutron-gamma fields are used for test and calibration of dosimetric and spectrometric systems, intercomparison measurements, and benchmark tests and represent experimental base for reactor studies. Set of the spherical experimental assemblies for the mixed neutron-gamma reference fields implementation was build in the NRI Rez. Extended sets of measurements and simulation calculations were done to describe the reference mixed field dosimetry and spectral characteristics with best achievable precision. The Monte Carlo technique was used for different experimental setups models description, comparison and verification and field characteristics simulation. Effects (hardly distinguishable experimentally) were also studied ( contributions from individual parts of experimental setup, field individual components and next effects as shadow shield cones transparency, etc.). Some results and main conclusions of these studies and calculations are presented and discussed. (authors)
Coarse-graining polymers with the MARTINI force-field: polystyrene as a benchmark case
Rossi, G.; Monticelli, L.; Puisto, S. R.;
2011-01-01
parameterization. We refine the MARTINI procedure by including one additional target property related to the structure of the polymer, namely the radius of gyration. The force-field optimization is mainly based on experimental data. We test our procedure on polystyrene, a standard benchmark for coarse-grained (CG...
Multiphysics field analysis and multiobjective design optimization: a benchmark problem
di Barba, P.; Doležel, Ivo; Karban, P.; Kůs, P.; Mach, F.; Mognaschi, M. E.; Savini, A.
2014-01-01
Roč. 22, č. 7 (2014), s. 1214-1225. ISSN 1741-5977 R&D Projects: GA ČR(CZ) GAP102/11/0498 Institutional support: RVO:61388998 Keywords : coupled-field problems * finite-element analysis * hp-FEM adaptation Subject RIV: JA - Electronics ; Optoelectronics, Electrical Engineering Impact factor: 0.868, year: 2014
Highlights: • Benchmark study performed for the neutronic calculations of TRIGA research reactors. • WIMSD-5B/CITATION is the utilized code system along with the WIMSD-IAEA-69 library. • The studied condensed spectra are five and seven energy groups spectra. • Analyzed: lattice parameters, reactivities, CR worth, flux and power distribution. • The lattice and neutronic parameters showed the accuracy of both condensed spectra. - Abstract: The objective of this paper is to assess the suitability and accuracy of the deterministic diffusion method for the neutronic calculations of the TRIGA Mark-III research reactors using the WIMSD/CITATION code system in proposed condensed energy spectra of five and seven energy groups with one and three thermal groups respectively. The utilized cell transport calculations code and core diffusion calculations code are the WIMSD-5B and the CITVAP v3.1 codes respectively, along with the WIMSD-IAEA-69 nuclear data library. Firstly, the assessment goes through analyzing the integral parameters – keff, ρ238, δ235, δ238, and C* – of the TRX and BAPL benchmark lattices and comparison with experimental and previous reference results using other ENDLs at the full energy spectra which show good agreement with the references at both spectra. Secondly, evaluation of the 3D nuclear characteristics of three different cores of the TRR-1/M1 TRIGA Mark-III Thai research reactor at the condensed energy spectra. The results include the excess reactivities of the cores and the worth of selected control rods which were compared with reference Monte Carlo results and experimental values. The results show good agreement with the references at both energy spectra and the better accuracy are attainable in the five energy groups spectrum. The results also include neutron flux distributions which are evaluated for future comparisons with other calculational techniques even they are comparable to reactors and fuels of the same type. The
Absorbed neutron doses in air holes of fast neutron fields at the RB reactor
Different experimental fast neutron fields are created at the RB reactor. The absorbed neutron doses in their air holes are determined on the basis of intermediate and fast neutron spectra measurements. The obtained results are analyzed in connection with application of these fields. (author)
Compilation of neutron flux density spectra and reaction rates in different neutron fields
Upon the recommendation of International Working Group of Reactor Radiation Measurements (IWGRRM), the compilation of neutron flux density spectra and the reaction rates obtained by activation and fission foils in different neutron fields is presented. The neutron fields considered are as follows: 1/E; iron block; LWR core and pressure vessel; LMFBR core and blanket; CTR first wall and blanket; fission spectrum
Since eight years at our laboratory, MCNP code has been used as a fundamental tool in many fusion directed activities in which we have been or we still are involved. Mainly they are: neutronics analysis of the performances of blanket components, supporting and optimizing their design; the estimation of the nuclear heat and radiation loads on the toroidal superconducting coils to assess the system shielding performances; then, a 14 MeV neutron generator is recently operating in Frascati and an experimental programme started with a benchmark neutron transport in a stainless steel block, MCNP is used to perform calculations. Present status of these experiments are reviewed. (K.A.)
Standard Guide for Benchmark Testing of Light Water Reactor Calculations
American Society for Testing and Materials. Philadelphia
2010-01-01
1.1 This guide covers general approaches for benchmarking neutron transport calculations in light water reactor systems. A companion guide (Guide E2005) covers use of benchmark fields for testing neutron transport calculations and cross sections in well controlled environments. This guide covers experimental benchmarking of neutron fluence calculations (or calculations of other exposure parameters such as dpa) in more complex geometries relevant to reactor surveillance. Particular sections of the guide discuss: the use of well-characterized benchmark neutron fields to provide an indication of the accuracy of the calculational methods and nuclear data when applied to typical cases; and the use of plant specific measurements to indicate bias in individual plant calculations. Use of these two benchmark techniques will serve to limit plant-specific calculational uncertainty, and, when combined with analytical uncertainty estimates for the calculations, will provide uncertainty estimates for reactor fluences with ...
A new multidimensional semi-analytical benchmark capability is developed. The key feature in the solution is the point kernel formulation. The 3D nature of the source is inherited in the flux making this a true multidimensional test. In addition, an efficient numerical scheme, called iterative interpolation, is used to evaluate the required point kernel solution and maintain benchmark accuracy. The EVENT finite element transport algorithm is compared to the point source solution as the first step of embedding the benchmark directly with the EVENT code. Additional code comparisons will be presented. (authors)
A new technique for neutron monitoring in stray radiation fields
At reactors, accelerators and therapy facilities including linear accelerators there is the need to monitor low level stray radiation fields, Neutron monitoring today is based mainly on the application of rem counters. A new approach to neutron monitoring is described which allows one to measure the dose equivalent of neutrons and gamma rays and to interpret neutron spectra in stray radiation fields in terms of Esub(e)sub(f)sub(f) and the dose fractnon of thermat neutrons. Compared to the muttisphere technique only a single moderator sphere of 30 cm diam and various pairs of TLD600/TLD7O0 detectors as passive neutron/gamma detectors are applied to measure moderated neutrons in the center of the sphere, backscattered albedo neutrons on the surface of the sphere and thermal as well as epithermal neutrons from the stray radiation field directly. The passive dosimeter system is sensitive to neutrons in the dose range 10 mrem - 500 rem and permits long-term exposures of several months. Field exposures performed at different facilities are described showing representative results for a neutron stray radiation field with Esub(e)sub(f)sub(f) between 2 MeV down to 100 keV and dose contributions from thermal neutrons between 1% and 30% depending on the kind of neutron stray source, the distance from the source, and the concrete shieldings in the room. (H.K.)
Compilation of neutron flux density spectra and reaction rates in different neutron fields. V.3
Upon the recommendation of the International Working Group of Reactor Radiation Measurements (IWGRRM) a compilation of documents containing neutron flux density spectra and the reaction rates obtained by activiation and fission foils in different neutron fields is presented
Various benchmark cores were analyzed with the 4S nuclear design methodology (in combination of transport calculation methods and JENDL-3.3 base libraries), in order to enhance the data base for evaluation of uncertainties of nuclear characteristics. Delayed neutron fraction analyses were made for support of the safety analyses of the 4S core. Using critical benchmark data obtained in leakage dominant small ZPR and ZPPR cores, low enriched uranium fast spectra cores have been analyzed using JENDL-3.3 and ENDF/B-VII.0 libraries to understand the results obtained from the recent reflector control physics benchmark FCA XXIII cores. The evaluation showed delayed neutron fractions and criticality were reproduced in good agreement with the 4S nuclear design methodology. (author)
Boldyreva, Anna
2014-01-01
This bachelor's thesis is focused on financial benchmarking of TULIPA PRAHA s.r.o. The aim of this work is to evaluate financial situation of the company, identify its strengths and weaknesses and to find out how efficient is the performance of this company in comparison with top companies within the same field by using INFA benchmarking diagnostic system of financial indicators. The theoretical part includes the characteristic of financial analysis, which financial benchmarking is based on a...
The ENDF/B-VII.1 library is the latest revision to the United States' Evaluated Nuclear Data File (ENDF). The ENDF library is currently in its seventh generation, with ENDF/B-VII.0 being released in 2006. This revision expands upon that library, including the addition of new evaluated files (was 393 neutron files previously, now 418 including replacement of elemental vanadium and zinc evaluations with isotopic evaluations) and extension or updating of many existing neutron data files. Complete details are provided in the companion paper [1]. This paper focuses on how accurately application libraries may be expected to perform in criticality calculations with these data. Continuous energy cross section libraries, suitable for use with the MCNP Monte Carlo transport code, have been generated and applied to a suite of nearly one thousand critical benchmark assemblies defined in the International Criticality Safety Benchmark Evaluation Project's International Handbook of Evaluated Criticality Safety Benchmark Experiments. This suite covers uranium and plutonium fuel systems in a variety of forms such as metallic, oxide or solution, and under a variety of spectral conditions, including unmoderated (i.e., bare), metal reflected and water or other light element reflected. Assembly eigenvalues that were accurately predicted with ENDF/B-VII.0 cross sections such as unmoderated and uranium reflected 235U and 239Pu assemblies, HEU solution systems and LEU oxide lattice systems that mimic commercial PWR configurations continue to be accurately calculated with ENDF/B-VII.1 cross sections, and deficiencies in predicted eigenvalues for assemblies containing selected materials, including titanium, manganese, cadmium and tungsten are greatly reduced. Improvements are also confirmed for selected actinide reaction rates such as 236U capture. Other deficiencies, such as the overprediction of Pu solution system critical eigenvalues and a decreasing trend in calculated eigenvalue for
Research of Multi Detectors of Neutron Spectrum in Mix Fields
LI; Wei; CHEN; Jun; WANG; Zhi-qiang; LI; Chun-juan; LIU; Yi-na; LUO; Hai-long; ZHANG; Wei-hua
2013-01-01
This neutron spectrometer can be used to measure neutron spectrum and neutron equivalent dosimetry.The range of neutron spectrum is thermal-20 MeV,and the range of neutron equivalent dosimetry is 1μSv·h-1-4 mSv·h-1.The sensor head of the neutron spectrum of multi detectors in mix fields houses five gas-filled sensors and a photo-scintillator column.There are two boron tri-fluoride(BF3)and three hydrogen
Benchmarking shielding simulations for an accelerator-driven spallation neutron source
Cherkashyna, Nataliia; DiJulio, Douglas D.; Panzner, Tobias; Rantsiou, Emmanouela; Filges, Uwe; Ehlers, Georg; Bentley, Phillip M.
2015-01-01
The shielding at an accelerator-driven spallation neutron facility plays a critical role in the performance of the neutron scattering instruments, the overall safety, and the total cost of the facility. Accurate simulation of shielding components is thus key for the design of upcoming facilities, such as the European Spallation Source (ESS), currently in construction in Lund, Sweden. In this paper, we present a comparative study between the measured and the simulated neutron background at the...
NONE
1998-06-01
This volume of the progress report provides documentation of reactor physics and criticality safety studies conducted in the Russian Federation during fiscal year 1997 and sponsored by the Fissile Materials Disposition Program of the US Department of Energy. Descriptions of computational and experimental benchmarks for the verification and validation of computer programs for neutron physics analyses are included. All benchmarks include either plutonium, uranium, or mixed uranium and plutonium fuels. Calculated physics parameters are reported for all of the contaminated benchmarks that the United States and Russia mutually agreed in November 1996 were applicable to mixed-oxide fuel cycles for light-water reactors.
Parallel PENTRAN solutions of the 'TIEL' steady state neutron transport benchmark problems
The TIEL benchmarks include a set of analytical benchmark solutions to the transport equation in infinite media. In this paper, we present solutions for the infinite planar, point, spherical shell, and solid spherical sources solved using the PENTRAN 3-D Parallel SN code compared to reference solutions of the TIEL benchmark problems. Excellent agreement (less than 0.7% difference) was achieved when comparing reference solutions with PENTRAN scalar fluxes in the case of planar, shell and solid sphere sources. Depending on the scattering properties and the geometry of the system, Legendre-Chebychev quadratures as high as S34 for the planar or S54 for the spherical symmetry sources, with appropriate spatial discretization, were needed to properly represent the flux. Inherent numerical difficulties were encountered when finding the solution in the very close proximity (< 0.1 mfp) of the sources. (authors)
Background: Shields are commonly constructed for a radionuclide neutron source m an actual calibration room in practice. Purpose: Monte Carlo (MC) calculation and experiments were applied to evaluate the effects of scattered neutrons on the neutron radiation field generated by Cf-252 neutron source with a shield. Methods: The effects of scattered neutrons caused by the shield of Cf-252 neutron source were evaluated by calculating the neutron spectra, neutron flux rate and neutron ambient dose equivalent with MC simulation. Similarly, the effects of scattered neutrons caused by the walls, ground and roof of source room were analyzed. Results: The calculation results show that the neutron flux-ambient dose equivalent conversion factor changes from 385 pSv·cm2 of a bare Cf-252 radionuclide from an idealized situation to 280 pSv·cm2 with the shield. The contribution of scattered neutrons from the walls, ground and roof is proportional to the square of distance between wall and source. The experimental data on dose rate are consistent with the calculated results and indicate the reliability of this method. Conclusion: This study provides a practical and feasible way to calibrate the radiation protection instruments using a non-standard radionuclide neutron radiation field. (authors)
Characterization of neutron field in a NPP workplace
At the Krsko Nuclear Power Plant (NPP), albedo dosimeters are used for personal neutron dosimetry. Spectrometric measurements allow determination of reference dosimetric values of realistic neutron fields to be used for calibration of albedo dosimeters. The Laboratory for Neutron Metrology and Dosimetry from the Inst. for Radiological Protection and Nuclear Safety (IRSN) was in charge of characterising neutron fields in the plant at two representative points with high neutron and gamma dose rate. Calibration of the dosimeters in the workplace used to be performed only by a spherical survey meter. Based on the reference dosimetric values, the Plant Dosimetry Laboratory has verified the response of albedo dosimeters. (authors)
Design of a pre-collimator system for neutronics benchmark experiment
Benchmark experiment is an important means to inspect the reliability and accuracy of the evaluated nuclear data, the effect/background ratios are the important parameters to weight the quality of experimental data. In order to obtain higher effect/background ratios, a pre-collimator system was designed for benchmark experiment. This system mainly consists of a pre-collimator and a shadow cone, The MCNP-4C code was used to simulate the background spectra under various conditions, from the results we found that with the pre-collimator system have a very marked improvement in the effect/background ratios. (authors)
Status of international benchmark experiment for effective delayed neutron fraction ({beta}eff)
Okajima, S.; Sakurai, T.; Mukaiyama, T. [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment
1997-03-01
To improve the prediction accuracy of the {beta}eff, the program of the international benchmark experiment (Beta Effect Reactor Experiment for a New International Collaborative Evaluation: BERNICE) was planned. This program composed of two parts; BERNICE-MASURCA and BERNICE-FCA. The former one was carried out in the fast critical facility MASURCA of CEA, FRANCE between 1993 and 1994. The latter one started in the FCA, JAERI in 1995 and still is going. In these benchmark experiments, various experimental techniques have been applied for in-pile measurements of the {beta}eff. The accuracy of the measurements was better than 3%. (author)
Development of magnetic field imaging using polarized pulsed neutrons
Magnetic field imaging using polarized pulsed neutrons is one of the most attractive applications of the energy-resolved neutron imaging technique. As the interaction between magnetic fields and a neutron spin depends on the neutron wavelength, analysis of the wavelength dependent polarization makes it possible to quantify the both magnetic field strength and direction. Hence, the combination of Time-of-flight method of pulsed neutron beam and three-dimensional polarization analysis is very suitable to the neutron magnetic field imaging technique. In this paper, we will explain about the results on magnetic steel sample and on spatial magnetic field from the electromagnet performed at BL10 of MLF in J-PARC. (author)
An in-phantom comparison of neutron fields for BNCT
Previously, the authors have developed the in-phantom neutron field assessment parameters T and D (Tumor) for the evaluation of epithermal neutron fields for use in BNCT. These parameters are based on an energy-spectrum-dependent neutron normal-tissue RBE and the treatment planning methodology of Gahbauer and his co-workers, which includes the effects of dose fractionation. In this paper, these neutron field assessment parameters were applied to The Ohio State University (OSU) design of an Accelerator Based Neutron Source (ABNS) (hereafter called the OSU-ABNS) and the Brookhaven Medical Research Reactor (BMRR) epithermal neutron beam (hereafter called the BMRR-ENB), in order to judge the suitability of the OSU-ABNS for BNCT. The BMRR-ENB was chosen as the basis for comparison because it is presently being used in human clinical trials of BNCT and because it is the standard to which other neutron beams are most often compared
Large-Signal Model of Graphene Field-Effect Transistors -- Part II: Circuit Performance Benchmarking
Pasadas, Francisco; Jiménez, David
2016-01-01
This paper presents a circuit performance benchmarking using the large-signal model of graphene field effect transistor reported in Part I of this two-part paper. To test the model, it has been implemented in a circuit simulator. Specifically we have simulated a high-frequency performance amplifier, together with other circuits that take advantage of the ambipolarity of graphene, such as a frequency doubler, a radio-frequency subharmonic mixer and a multiplier phase detector. A variety of sim...
Lin, Yi-Chun; Huang, Tseng-Te; Liu, Yuan-Hao; Chen, Wei-Lin; Chen, Yen-Fu; Wu, Shu-Wei; Nievaart, Sander; Jiang, Shiang-Huei
2015-06-01
The paired ionization chambers (ICs) technique is commonly employed to determine neutron and photon doses in radiology or radiotherapy neutron beams, where neutron dose shows very strong dependence on the accuracy of accompanying high energy photon dose. During the dose derivation, it is an important issue to evaluate the photon and electron response functions of two commercially available ionization chambers, denoted as TE(TE) and Mg(Ar), used in our reactor based epithermal neutron beam. Nowadays, most perturbation corrections for accurate dose determination and many treatment planning systems are based on the Monte Carlo technique. We used general purposed Monte Carlo codes, MCNP5, EGSnrc, FLUKA or GEANT4 for benchmark verifications among them and carefully measured values for a precise estimation of chamber current from absorbed dose rate of cavity gas. Also, energy dependent response functions of two chambers were calculated in a parallel beam with mono-energies from 20 keV to 20 MeV photons and electrons by using the optimal simple spherical and detailed IC models. The measurements were performed in the well-defined (a) four primary M-80, M-100, M120 and M150 X-ray calibration fields, (b) primary 60Co calibration beam, (c) 6 MV and 10 MV photon, (d) 6 MeV and 18 MeV electron LINACs in hospital and (e) BNCT clinical trials neutron beam. For the TE(TE) chamber, all codes were almost identical over the whole photon energy range. In the Mg(Ar) chamber, MCNP5 showed lower response than other codes for photon energy region below 0.1 MeV and presented similar response above 0.2 MeV (agreed within 5% in the simple spherical model). With the increase of electron energy, the response difference between MCNP5 and other codes became larger in both chambers. Compared with the measured currents, MCNP5 had the difference from the measurement data within 5% for the 60Co, 6 MV, 10 MV, 6 MeV and 18 MeV LINACs beams. But for the Mg(Ar) chamber, the derivations reached 7
Sensitivity analysis in the fast neutron field
Report presents first part of activities which were completed within task 7: 'Sensitivity analysis in the fast neutron field'. It includes general mathematical formulation of linear and bilinear relevant functionals, as well as special forms of characteristic values. In addition, explicit form of transport equation is derived. It should be solved for the need of sensitivity analysis. Based on presented mathematical method and review of existing methods, a computation procedure is conceived. It is made of 3 parts: preparation of multigroup constants, solving the transport equations and calculation of functionals. ENDF/B-IV data, service code NJOY, RFPN code for solving transport equations and ANOS code for calculating the functionals and sensitivity analysis will be used. RFPN code ws adapted for this purpose and the ANOS code needs additional work in the forthcoming phase
Surian Pinem
2014-01-01
Full Text Available A coupled neutronics thermal-hydraulics code NODAL3 has been developed based on the few-group neutron diffusion equation in 3-dimensional geometry for typical PWR static and transient analyses. The spatial variables are treated by using a polynomial nodal method while for the neutron dynamic solver the adiabatic and improved quasistatic methods are adopted. In this paper we report the benchmark calculation results of the code against the OECD/NEA CRP PWR rod ejection cases. The objective of this work is to determine the accuracy of NODAL3 code in analysing the reactivity initiated accident due to the control rod ejection. The NEACRP PWR rod ejection cases are chosen since many organizations participated in the NEA project using various methods as well as approximations, so that, in addition to the reference solutions, the calculation results of NODAL3 code can also be compared to other codes’ results. The transient parameters to be verified are time of power peak, power peak, final power, final average Doppler temperature, maximum fuel temperature, and final coolant temperature. The results of NODAL3 code agree well with the PHANTHER reference solutions in 1993 and 1997 (revised. Comparison with other validated codes, DYN3D/R and ANCK, shows also a satisfactory agreement.
Neutron induced degradation in nitrided pyrogenic field oxide MOS capacitors
Vaidya, S J; Shaikh, A M; Chandorkar, A N
2002-01-01
Neutron induced oxide charge trapping and generation of interface states in MOS capacitors with pyrogenic and nitrided pyrogenic field oxides have been studied. In order to assess the damage due to neutrons alone, it is necessary to account for the damage produced by the accompanying gamma rays from neutron radiation. This is done by measuring the intensity of gamma radiation accompanying neutrons at different neutron fluences at the irradiation position. MOS capacitor structures were subjected to neutron radiation in a swimming pool type of reactor. Other samples from the same batch were then subjected to an equivalent dose of gamma radiation from a Co sup 6 sup 0 source. The difference in the damage observed was used to characterize the damage caused by neutrons. It is observed that neutrons, though uncharged, are capable of causing ionization damage. This damage is found to be significant when the radiation is performed under biased conditions. Nitridation in different ambients is found to improve the radi...
Weakly bound states of neutrons in gravitational fields
Khugaev, Avas V.; Sultanov, Renat A.; Guster, Dennis
2010-01-01
In this paper a quantum-mechanical behaviour of neutrons in gravitational fields is considered. A first estimation is made using the semiclassical approximation, neglecting General Relativity, magnetic and rotation effects, for neutrons in weakly bound states in the weak gravitational field of the Earth. This result was generalized for a case, in which the Randall - Sundrum correction to Newton's gravitational law on the small scales was applied. Application of the results to Neutron Star phy...
The Response of Alanine Dosimeters in Thermal Neutron Fields
Schmitz, T; Bassler, Niels; Sharpe, P; Palmans, H.; KRATZ J.v.; Langgruth, P.; HAMPEL G.
2012-01-01
Purpose:Boron Neutron Capture Therapy (BNCT) is a special kind of particle therapy, based on the neutron induced fission of the boron isotope 10B [1]. We have performed dosimetry experiments on the mixed neutron and gamma fields at the TRIGA Mark II research reactor in Mainz. Commonly, dosimetry in such fields is realized by foil activation and ion chambers [2]. Here we investigate alanine as an easier and more robust alternative dosimeter.Methods:We have performed four phantom experiments at...
Core management and fast neutron field characterization of JOYO
Twenty eight years of operations at the experimental fast reactor JOYO provide a wealth of experience with core management and characterization of fast neutron field. This experience has been applied to several core modifications to upgrade JOYO's irradiation capability. Reactor physics test and neutron n field.flux measurement results have been used to confirm the accuracy of neutron diffusion theory calculations. These experiences and accumulated data will be useful for the core design in future fast reactors. (author)
Lin, Yi-Chun [Health Physics Division, Institute of Nuclear Energy Research, Taoyuan County, Taiwan (China); Huang, Tseng-Te, E-mail: huangtt@iner.gov.tw [Health Physics Division, Institute of Nuclear Energy Research, Taoyuan County, Taiwan (China); Liu, Yuan-Hao [Nuclear Science and Technology Development Center, National Tsing Hua University, Hsinchu City, Taiwan (China); Chen, Wei-Lin [Institute of Nuclear Engineering and Science, National Tsing Hua University, Hsinchu City, Taiwan (China); Chen, Yen-Fu [Atomic Energy Council, New Taipei City, Taiwan (China); Wu, Shu-Wei [Dept. of Biomedical Engineering and Environmental Sciences, National Tsing Hua University, Hsinchu, Taiwan (China); Nievaart, Sander [Institute for Energy, Joint Research Centre, European Commission, Petten (Netherlands); Jiang, Shiang-Huei [Dept. of Engineering and System Science, National Tsing Hua University, Hsinchu, Taiwan (China)
2015-06-01
The paired ionization chambers (ICs) technique is commonly employed to determine neutron and photon doses in radiology or radiotherapy neutron beams, where neutron dose shows very strong dependence on the accuracy of accompanying high energy photon dose. During the dose derivation, it is an important issue to evaluate the photon and electron response functions of two commercially available ionization chambers, denoted as TE(TE) and Mg(Ar), used in our reactor based epithermal neutron beam. Nowadays, most perturbation corrections for accurate dose determination and many treatment planning systems are based on the Monte Carlo technique. We used general purposed Monte Carlo codes, MCNP5, EGSnrc, FLUKA or GEANT4 for benchmark verifications among them and carefully measured values for a precise estimation of chamber current from absorbed dose rate of cavity gas. Also, energy dependent response functions of two chambers were calculated in a parallel beam with mono-energies from 20 keV to 20 MeV photons and electrons by using the optimal simple spherical and detailed IC models. The measurements were performed in the well-defined (a) four primary M-80, M-100, M120 and M150 X-ray calibration fields, (b) primary {sup 60}Co calibration beam, (c) 6 MV and 10 MV photon, (d) 6 MeV and 18 MeV electron LINACs in hospital and (e) BNCT clinical trials neutron beam. For the TE(TE) chamber, all codes were almost identical over the whole photon energy range. In the Mg(Ar) chamber, MCNP5 showed lower response than other codes for photon energy region below 0.1 MeV and presented similar response above 0.2 MeV (agreed within 5% in the simple spherical model). With the increase of electron energy, the response difference between MCNP5 and other codes became larger in both chambers. Compared with the measured currents, MCNP5 had the difference from the measurement data within 5% for the {sup 60}Co, 6 MV, 10 MV, 6 MeV and 18 MeV LINACs beams. But for the Mg(Ar) chamber, the derivations
The neutron imaging system fielded at the National Ignition Facility
We have fielded a neutron imaging system at the National Ignition Facility to collect images of fusion neutrons produced in the implosion of inertial confinement fusion experiments and scattered neutrons from (n, n') reactions of the source neutrons in the surrounding dense material. A description of the neutron imaging system is presented, including the pinhole array aperture, the line-of-sight collimation, the scintillator-based detection system and the alignment systems and methods. Discussion of the alignment and resolution of the system is presented. We also discuss future improvements to the system hardware. (authors)
The Neutron Imaging System Fielded at the National Ignition Facility
We have fielded a neutron imaging system at the National Ignition Facility to collect images of fusion neutrons produced in the implosion of inertial confinement fusion experiments and scattered neutrons from (n, n') reactions of the source neutrons in the surrounding dense material. A description of the neutron imaging system will be presented, including the pinhole array aperture, the line-of-sight collimation, the scintillator-based detection system and the alignment systems and methods. Discussion of the alignment and resolution of the system will be presented. We will also discuss future improvements to the system hardware.
The Neutron Imaging System Fielded at the National Ignition Facility
Fittinghoff, D N; Atkinson, D P; Bower, D E; Drury, O B; Dzenitis, J M; Felker, B; Frank, M; Liddick, S N; Moran, M J; Roberson, G P; Weiss, P B; Grim, G P; Aragonez, R J; Archuleta, T N; Batha, S H; Clark, D D; Clark, D J; Danly, C R; Day, R D; Fatherley, V E; Finch, J P; Garcia, F P; Gallegos, R A; Guler, N; Hsu, A H; Jaramillo, S A; Loomis, E N; Mares, D; Martinson, D D; Merrill, F E; Morgan, G L; Munson, C; Murphy, T J; Oertel, J A; Polk, P J; Schmidt, D W; Tregillis, I L; Valdez, A C; Volegov, P L; Wang, T F; Wilde, C H; Wilke, M D; Wilson, D C; Buckles, R A; Cradick, J R; Kaufman, M I; Lutz, S S; Malone, R M; Traille, A
2011-10-24
We have fielded a neutron imaging system at the National Ignition Facility to collect images of fusion neutrons produced in the implosion of inertial confinement fusion experiments and scattered neutrons from (n, n') reactions of the source neutrons in the surrounding dense material. A description of the neutron imaging system will be presented, including the pinhole array aperture, the line-of-sight collimation, the scintillator-based detection system and the alignment systems and methods. Discussion of the alignment and resolution of the system will be presented. We will also discuss future improvements to the system hardware.
Interaction of neutrons with the matter in the laser field
The interactions of neutrons with the molecules, atoms and nuclei in the presence of the coherent electromagnetic radiation are considered. There are two effects which are discussed in detail: 1) the ''acceleration'' of thermal neutrons passed through the excited by the resonance laser wave molecular gas; 2) the induced by the laser field the slow neutron capture accompanied by the compound nucleus level excitation. The given effects, if they are experimentally detected, give the possibility to control the neutron flux (spectrum change, polarization, spatial modulation and etc.) and change the interaction cross sections of thermal and resonance neutrons with nuclei due to excitation of p levels of the compound nucleus
The neutron imaging system fielded at the National Ignition Facility
Fittinghoff D.N.
2013-11-01
Full Text Available We have fielded a neutron imaging system at the National Ignition Facility to collect images of fusion neutrons produced in the implosion of inertial confinement fusion experiments and scattered neutrons from (n, n′ reactions of the source neutrons in the surrounding dense material. A description of the neutron imaging system is presented, including the pinhole array aperture, the line-of-sight collimation, the scintillator-based detection system and the alignment systems and methods. Discussion of the alignment and resolution of the system is presented. We also discuss future improvements to the system hardware.
Neutron-kinetic and thermo-hydraulic uncertainties in the study of Kalinin-3 benchmark
The effects of nuclear data covariance on important reactor parameters are investigated. The analyses are performed on the base of the OECD/NEA coolant transient Benchmark (K-3) on measured data at Kalinin-3 Nuclear Power Plant (NPP). For this purpose the GRS uncertainty and sensitivity software package XSUSA is applied to propagate uncertainties in nuclear data libraries to the full core coupled transient calculations. Moreover, based on the previous thermo-hydraulic studies a set of most important thermo-hydraulic parameters is chosen and added to the uncertain input vector. A statistically representative set of coupled ATHLET PARCS code steady state calculations is analyzed and both integral and local output quantities are compared with the measurements available in the benchmark. The work is a step forward in establishing a ''best-estimate calculations in combination with performing uncertainty analysis'' methodology for coupled full core calculations.
Neutron-kinetic and thermo-hydraulic uncertainties in the study of Kalinin-3 benchmark
Pasichnyk, Ihor; Zwermann, Winfried; Velkov, Kiril [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) gGmbH, Garching (Germany); Nikonov, Sergey [All-Russain Research Institute for NPP Operation (VNIIAES), Moscow (Russian Federation)
2015-09-15
The effects of nuclear data covariance on important reactor parameters are investigated. The analyses are performed on the base of the OECD/NEA coolant transient Benchmark (K-3) on measured data at Kalinin-3 Nuclear Power Plant (NPP). For this purpose the GRS uncertainty and sensitivity software package XSUSA is applied to propagate uncertainties in nuclear data libraries to the full core coupled transient calculations. Moreover, based on the previous thermo-hydraulic studies a set of most important thermo-hydraulic parameters is chosen and added to the uncertain input vector. A statistically representative set of coupled ATHLET PARCS code steady state calculations is analyzed and both integral and local output quantities are compared with the measurements available in the benchmark. The work is a step forward in establishing a ''best-estimate calculations in combination with performing uncertainty analysis'' methodology for coupled full core calculations.
Cross-section evaluation utilizing integral reaction-rate measurements in fast neutron fields
The role of integral reaction-rate data for cross-section evaluation is reviewed. The subset of integral data considered comprises integral reaction rates measured for dosimeter, fission-product, and actinide-type materials irradiated in reactor dosimetry fast neutron benchmark fields and in the EBR-II. Utilization of these integral data for integral testing, multigroup cross-section adjustment and pointwise cross section adjustment is treated in some detail. Examples are given that illustrate the importance of considering a priori uncertainty and correlation information for these analyses. 3 figures, 3 tables
The ADS (accelerator driven sub-critical system) has great energy spans, complex energy spectrum structures and strong physical effects. Hence, the existing nuclear data libraries can't fully meet the needs of nuclear analysis in ADS. In order to do nuclear analysis for ADS system, a point-wise data library HENDL-ADS/MC (hybrid evaluated nuclear data library) was produced by FDS team. Meanwhile, to test the availability and reliability of the HENDL-ADS/MC data library, a series of shielding and critical safety benchmarks were performed. To validate and qualify the reliability of the high-energy cross section for HENDL-ADS/MC library further, a series of high neutronics integral experiments have been performed. The testing results confirm the accuracy and reliability of HENDL-ADS/MC. (authors)
Characteristics and contents of the CSRL-V (Criticality Safety Reference Library based on ENDF/B-V data) 227-neutron-group AMPX master and pointwise cross-section libraries are described. Results obtained in using CSRL-V to calculate performance parameters of selected thermal reactor and criticality safety benchmarks are discussed
The facility for incineration of long-lived minor actinides and some dangerous fission products should be an important feature of the future nuclear power (NP). For many reasons the liquid-fuel reactor driven by accelerator can be considered as the perspective reactor- burner for radioactive waste. The fuel of such reactor is the fluoride molten salt composition with minor actinides (Np, Cm, Am) and some fission products (99Tc, 129I, etc.). Preliminary analysis shows that the values of keff, calculated with different codes and nuclear data differ up to several percents for such fuel compositions. Reliable critical and subcritical benchmark experiments with molten salt fuel compositions with significant quantities of minor actinides are absent. One of the main tasks for the numerical study of this problem is the estimation of nuclear data for such fuel compositions and verification of the different numerical codes used for the calculation of keff, neutron spectra and reaction rates. It is especially important for the resonance region where experimental data are poor or absent. The calculation benchmark of the cascade subcritical molten salt reactor is developed. For the chosen nuclear fuel composition the comparison of the results obtained by three different Monte-Carlo codes (MCNP4A, MCU, and C95) using three different nuclear data libraries are presented. This report concerns the investigation of subcritical molten salt reactor unit main peculiarities carried out at the beginning of ISTC project 1486. (author)
Recent findings indicate that gamma radiation can contribute to the embrittlement of reactor materials. On this background an experimental benchmark programme at two low power reactors was started to measure both, neutron and gamma spectral fluences behind and inside of transmission modules consisting of variable iron and water slabs using a NE213 scintillation spectrometer and partly a HPGe detector. The experimental results are used to validate Monte Carlo calculation methods for coupled neutron/gamma problems. The experiment and results of a first series of measurements and comparisons to MCNP calculations for neutron and gamma energy spectra are presented. (author)
US/JAERI fusion neutronics calculational benchmarks for nuclear data and codes intercomparison
The US Department of Energy (DOE) and the Japan Atomic Energy Research Institute (JAERI) have been involved in a collaborative research program on fusion neutronics. The program focuses on performing post- and pre-analyses for the integral experiments conducted at the fusion neutronics source (FNS) facility at JAERI. The main general objectives of the program are: (a) to provide experimental data needed to determine the accuracy, guide the development and establish the validity of computational methods and the nuclear data base; (b) to provide the data base required to evaluate the overall uncertainty (both analytical and experimental) in estimating key parameters of importance in fusion blanket design (e.g., tritium production rate, nuclear heating rate, dose rate, etc.); (c) to intercompare various measuring techniques and to increase the reliability of measurements by developing more advanced detectors; (d) to provide experimental data to assist in the selection of materials and configuration of candidate blanket concepts from the neutronics viewpoint
Energy of Gravitational Field of Static Spherically Symmetric Neutron Stars
WENDe-Hua; CHENWei; WANGXian-Ju; AIBao-Quan; LIUGuo-Tao; LIULiang-Gang
2003-01-01
By using the Einstein-Tolman expression of the energy-momentum pseudo-tensor, the energy density of the gravitational field of the static spherically symmetric neutron stars is calculated in the Cartesian coordinate system.It is exciting that the energy density of gravitational field is positive and rational The xmmerical results of the energy density of gravitational field of neutron stars are calculated. For neutron stars with M=2M, the ratio of the energy density of gravitational field to the energy density of pure matters would be up to 0.54 at the surface.
Energy of Gravitational Field of Static Spherically Symmetric Neutron Stars
WEN De-Hua; CHEN Wei; WANG Xian-Ju; AI Bao-Quan; LIU Guo-Tao; LIU Liang-Gang
2003-01-01
By using the Einstein-Tolman expression of the energy-momentum pseudo-tensor, the energy density ofthe gravitational field of the static spherically symmetric neutron stars is calculated in the Cartesian coordinate system.It is exciting that the energy density of gravitational field is positive and rational. The numerical results ot the energydensity of gravitational field of neutron stars are calculated. For neutron stars with M = 2M , the ratio of the energydensity of gravitational field to the energy density of pure matters would be up to 0.54 at the surface.
Rotating proto-neutron stars under strong magnetic fields
Franzon, B; Schramm, S
2016-01-01
In this work, we study the effects of magnetic fields and rotation on the structure and composition of proto-neutron stars (PNSs). A hadronic chiral SU(3) model is applied to cold neutron stars (NS) and proto-neutron stars with trapped neutrinos and at fixed entropy per baryon. We obtain general relativistic solutions for neutron and proto-neutron stars endowed with a poloidal magnetic field by solving Einstein-Maxwell field equations in a self-consistent way. As the neutrino chemical potential decreases in value over time, this alters the chemical equilibrium and the composition inside the star, leading to a change in the structure and in the particle population of these objects. We find that the magnetic field deforms the star and significantly alters the number of trapped neutrinos in the stellar interior, together with strangeness content and temperature in each evolution stage.
Benchmark analysis of neutronics performances of a SiC block irradiated with 14 MeV neutrons
Silicon carbide (SiC) in the form of ceramic matrix is a low activation structural material proposed for fusion reactors. Its development is pursued in the European Fusion Technology Program. A SiC block (457x457x711 mm3), borrowed from JAERI, was irradiated with 14 MeV neutrons at the FNG facility of ENEA Frascati. Activation reaction rates, neutron fluxes and spectra, as well as nuclear heating were measured in four selected experimental positions inside the block. The experimental analysis was performed using the Monte Carlo transport code MCNP-4C and point-wise cross sections derived from FENDL-2.0, EFF-2.4 and EFF-3.0 evaluated nuclear data files. Deterministic transport calculations were also performed using the discrete ordinates code DORT. The sensitivity and uncertainty analysis were performed as well using the SUSD3D code. Results indicate that calculation based on EFF-3.0 nuclear data file estimates the neutron flux and spectra with a reasonable uncertainty which is still lower than ±30% for all measured quantities
Both the (ICRP) and the (NCPR) have recommended an increase in neutron quality factors and the adoption of effective dose equivalent methods. The series of reports entitled Personnel Neutron Dose Assessment Upgrade (PNL-6620) addresses these changes. Volume 1 in this series of reports (Personnel Neutron Dosimetry Assessment) provided guidance on the characteristics, use, and calibration of personnel neutron dosimeters in order to meet the new recommendations. This report, Volume 2: Field Neutron Spectrometer for Health Physics Applications describes the development of a portable field spectrometer which can be set up for use in a few minutes by a single person. The field spectrometer described herein represents a significant advance in improving the accuracy of neutron dose assessment. It permits an immediate analysis of the energy spectral distribution associated with the radiation from which neutron quality factor can be determined. It is now possible to depart from the use of maximum Q by determining and realistically applying a lower Q based on spectral data. The field spectrometer is made up of two modules: a detector module with built-in electronics and an analysis module with a IBM PC/reg sign/-compatible computer to control the data acquisition and analysis of data in the field. The unit is simple enough to allow the operator to perform spectral measurements with minimal training. The instrument is intended for use in steady-state radiation fields with neutrons energies covering the fission spectrum range. The prototype field spectrometer has been field tested in plutonium processing facilities, and has been proven to operate satisfactorily. The prototype field spectrometer uses a 3He proportional counter to measure the neutron energy spectrum between 50 keV and 5 MeV and a tissue equivalent proportional counter (TEPC) to measure absorbed neutron dose
Four calculational benchmarks have been selected to compare various nuclear data libraries based on both ENDF/B-IV and V, and to compare results from various transport codes. Discrepancies up to 20% in tritium production from 7Li were found and have been attributed mainly to differences in current ENDF/B-IV and V evaluations, while approx.4% is attributed to differences in the group structure of the libraries used. Results from MCNP and VIP Monte Carlo codes are in good agreement, but MORSE calculations show good agreement only for high threshold reactions
The Monte Carlo (MC)-discrete ordinates (SN) coupled method is an efficient approach to solve shielding calculations of nuclear device with complex geometries and deep penetration. The 3D MC-SN coupled method has been used for PWR shielding calculation for the first time. According to characteristics of NUREG/CR-6115 PWR model, the thermal shield is specified as the common surface to link the Monte Carlo complex geometrical model and the deep penetration SN model. 3D Monte Carlo code is employed to accurately simulate the structure from core to thermal shield. The neutron tracks crossing the thermal shield inner surface are recorded by MC code. The SN boundary source is generated by the interface program and used by the 3D SN code to treat the calculation from thermal shield to pressure vessel. The calculation results include the circular distributions of fast neutron flux at pressure vessel inner wall, pressure vessel T/4 and lower weld locations. The calculation results are performed with comparison to MCNP and DORT solutions of benchmark report and satisfactory agreements are obtained. The validity of the method and the correctness of the programs are proved. (authors)
Simulation analysis of radiation fields inside phantoms for neutron irradiation
Radiation fields inside phantoms have been calculated for neutron irradiation. Particle and heavy-ion transport code system PHITS was employed for the calculation. Energy and size dependences of neutron dose were analyzed using tissue equivalent spheres of different size. A voxel phantom of mouse was developed based on CT images of an 8-week-old male C3H/HeNs mouse. Deposition energy inside the mouse was calculated for 2- and 10-MeV neutron irradiation. (author)
TRIPOLI-2: neutron gamma coupling - applications to shielding benchmarks and designs
Recent additions in the on-going development of the TRIPOLI Monte Carlo code system include conversion to the ENDF/B data format and an automated coupling scheme for neutron secondary gamma-ray calculations. Two shielding calculations are presented here which feature these two new developments
Neutronics benchmarking study of breeding shield for the fusion experimental reactor (FER)
Neutron flux distribution and tritium breeding ratio (TBR) were calculated with the one-dimensional discrete ordinates code using the various cross section libraries and with the continuous energy Monte Carlo code. Two types of the geometrical representation were considered; one is the homogeneous mixture of structural material, aqueous lithium salt coolant and, if any, neutron multiplier; the other is the alternating layers of the structure and the coolant. Results were compared in terms of neutron flux attenuation, integrated TBR, and energy profile and spatial distribution of tritium production rate. SN/MC ratios of TBRs for the homogeneous model are smaller than unity by a few percents and several percents for the configurations with and without multiplier, respectively. For the heterogeneous model without multiplier, the total TBRs agree within several percents and are larger than those of the homogeneous model by 20-40%. While, total neutron fluxes are underestimated with SN calculatuins by 30-40% compared to the MC results for both the homogeous and heterogeneous models. (author). 14 refs.; 5 figs.; 5 tabs
Fast neutron fluence calculation benchmark analysis based on 3D MC-SN bidirectional coupling method
The Monte Carlo (MC)-discrete ordinates (SN) bidirectional coupling method is an efficient approach to solve shielding calculation of the large complex nuclear facility. The test calculation was taken by the application of the MC-SN bidirectional coupling method on the shielding calculation of the large PWR nuclear facility. Based on the characteristics of NUREG/CR-6115 PWR benchmark model issued by the NRC, 3D Monte Carlo code was employed to accurately simulate the structure from the core to the thermal shield and the dedicated model of the calculation parts locating in the pressure vessel, while the TORT was used for the calculation from the thermal shield to the second down-comer region. The transform between particle probability distribution of MC and angular flux density of SN was realized by the interface program to achieve the coupling calculation. The calculation results were compared with MCNP and DORT solutions of benchmark report and satisfactory agreements were obtained. The preliminary validity of feasibility by using the method to solve shielding problem of a large complex nuclear device was proved. (authors)
Neutron transport benchmark on iron using a white highenergy neutron field
Bém, Pavel; Fischer, U.; Šimakov, S.; von Mollendorff, U.
2003-01-01
Roč. 69, 1, 2, 3, 4 (2003), s. 479-483. ISSN 0920-3796 Grant ostatní: GA-(XX) EFDA-TWO-TTMI-003/D14 Institutional research plan: CEZ:AV0Z1048901 Keywords : accelerator-driven system Subject RIV: BG - Nuclear, Atomic and Molecular Physics, Colliders Impact factor: 0.753, year: 2003
Electromagnetic multipole fields of neutron stars
There is now indisputable evidence that some pulsars possess space velocities so high that internal asymmetries in the dynamics of their formation are strongly implied. We develop in this paper a complete formalism for the calculation of the only such mechanism that has yet been subjected to quantitative analysis: electromagnetic recoil radiation. To make the general problem tractable without doing violence to the physics, we have made the following simplifying assumptions: (1) the magnetic induction B in athin shell enclosing the surface can be satisfactorily approximated by a sum of vacuum multipole fields; (2) the star is spherical, and all parts are in good electrical contact; (3) vertical-bar Ω X r vertical-barvery-much-less-thanc everywhere within the star; and (4) the star is surrounded by a vacuum. Our qualitative conclusions hold even if these assumptions are violated, but corrections to our quantitative results required by a relaxation of our assumptions are not easily computed.Given this simple electrodynamic model of a neutron star, we solve the following problems: (1) What electric multipoles are induced by each magnetic multipole. (2) What is the general formula for the recoil produced by the projection on the rotational axis of a net linear momentum flux produced by the rotation of any two magnetic multipoles. (3) What is the set of centered multipoles that represents the field of an arbitrary off-centered multipole. We use these general results go perform a detailed analysis of the linear momentum radiated by an off-centered dipole. We find a force larger by a factor 6 than that obtained for the special case treated in the best previous calculation. In spite of this considerable increase in the computed strengrh of the effect, we still believe it to be too weak to produce the large space velocities observed for pulsars. For the mechanism to be effective, the pulsar must be born rotating near the breakup velocity
Magnetic structure determination using zero-field neutron polarimetry
A simple interpretation of the formulae which predict the polarisation of elastically scattered neutrons and a pictorial representation of the polarisation directions before and after the scattering process are presented. Some results from recent zero-field neutron polarimetry experiments are used to demonstrate the theory. (orig.)
Reference neutron fields for metrology of radiation monitoring
Aleinikov, V.E.; Bamblevskij, V.P.; Komochkov, M.M.; Krylov, A.R.; Mokrov, Yu.V.; Timoshenko, G.N. (Joint Inst. for Nuclear Research, Dubna (Russian Federation))
1994-01-01
A set of reference neutron fields has been created in the Joint Institute for Nuclear Research for metrology of radiation monitoring. The set covers the common energy range of neutrons generated by nuclear installations of the Institute (from [approx] 10[sup -8] to hundreds of MeV). The set comprises reference fields based on [sup 252]Cf in polyethylene moderators with diameters 12.7 and 29.2 cm and two reference fields, based on a 600 MeV phasotron. A soft field is created in the labyrinth of the phasotron by scattered neutrons. A hard field is formed by leakage neutrons from a concrete shield without holes. This shield is irradiated by secondary radiation produced in the accelerator chamber and the target station. (author).
Reference neutron fields for metrology of radiation monitoring
A set of reference neutron fields has been created in the Joint Institute for Nuclear Research for metrology of radiation monitoring. The set covers the common energy range of neutrons generated by nuclear installations of the Institute (from ∼ 10-8 to hundreds of MeV). The set comprises reference fields based on 252Cf in polyethylene moderators with diameters 12.7 and 29.2 cm and two reference fields, based on a 600 MeV phasotron. A soft field is created in the labyrinth of the phasotron by scattered neutrons. A hard field is formed by leakage neutrons from a concrete shield without holes. This shield is irradiated by secondary radiation produced in the accelerator chamber and the target station. (author)
Neutron and photon shielding benchmark calculations by MCNP on the LR-0 experimental facility.
Hordósy, G
2005-01-01
In the framework of the REDOS project, the space-energy distribution of the neutron and photon flux has been calculated over the pressure vessel simulator thickness of the LR-0 experimental reactor, Rez, Czech Republic. The results calculated by the Monte Carlo code MCNP4C are compared with the measurements performed in the Nuclear Research Institute, Rez. The spectra have been measured at the barrel, in front of, inside and behind the pressure vessel in different configurations. The neutron measurements were performed in the energy range 0.1-10 MeV. This work has been done in the frame of the 5th Frame Work Programme of the European Community 1998-2002. PMID:16604591
Computational evaluation oa a neutron field facility
This paper describes the results of a study based on computer simulation for a realistic 3D model of Ionizing Radiation Laboratory of the Institute for Advanced Studies (IEAv) using the MCNP5 (Monte Carlo N-Particle) code, in order to guide the installing a neutron generator, produced by reaction 3H(d,n)4He. The equipment produces neutrons with energy of 14.1 MeV and 2 x 108 n/s production rate in 4 πgeometry, which can also be used for neutron dosimetry studies. This work evaluated the spectra and neutron fluence provided on previously selected positions inside the facility, chosen due to the interest to evaluate the assessment of ambient dose equivalent so that they can be made the necessary adjustments to the installation to be consistent with the guidelines of radiation protection and radiation safety, determined by the standards of National Nuclear Energy Commission (CNEN). (author)
Computational evaluation oa a neutron field facility
Pinto, Jose Julio de O.; Pazianotto, Mauricio T., E-mail: jjfilos@hotmail.com, E-mail: mpazianotto@gmail.com [Instituto Tecnologico de Aeronautica (ITA/DCTA), Sao Jose dos Campos, SP (Brazil); Federico, Claudio A.; Passaro, Angelo, E-mail: claudiofederico@ieav.cta.br, E-mail: angelo@ieav.cta.br [Instituto de Estudos Avancados (IEAv/DCTA), Sao Jose dos Campos, SP (Brazil)
2015-07-01
This paper describes the results of a study based on computer simulation for a realistic 3D model of Ionizing Radiation Laboratory of the Institute for Advanced Studies (IEAv) using the MCNP5 (Monte Carlo N-Particle) code, in order to guide the installing a neutron generator, produced by reaction {sup 3}H(d,n){sup 4}He. The equipment produces neutrons with energy of 14.1 MeV and 2 x 10{sup 8} n/s production rate in 4 πgeometry, which can also be used for neutron dosimetry studies. This work evaluated the spectra and neutron fluence provided on previously selected positions inside the facility, chosen due to the interest to evaluate the assessment of ambient dose equivalent so that they can be made the necessary adjustments to the installation to be consistent with the guidelines of radiation protection and radiation safety, determined by the standards of National Nuclear Energy Commission (CNEN). (author)
Study on neutron radiation field of carbon ions therapy
Xu, Jun-Kui; Li, Wu-Yuan; Yan, Wei-Wei; Chen, Xi-Meng; Mao, Wang; Pang, Cheng-Guo
2015-01-01
Carbon ions offer significant advantages for deep-seated local tumors therapy due to their physical and biological properties. Secondary particles, especially neutrons caused by heavy ion reactions should be carefully considered in treatment process and radiation protection. For radiation protection purposes, the FLUKA Code was used in order to evaluate the radiation field at deep tumor therapy room of HIRFL in this paper. The neutron energy spectra, neutron dose and energy deposition of carbon ion and neutron in tissue-like media was studied for bombardment of solid water target by 430MeV/u C ions. It is found that the calculated neutron dose have a good agreement with the experimental date, and the secondary neutron dose may not exceed one in a thousand of the carbon ions dose at Bragg peak area in tissue-like media.
Evaluation of neutron radiation field in carbon ion therapy
Xu, Jun-Kui; Su, You-Wu; Li, Wu-Yuan; Yan, Wei-Wei; Chen, Xi-Meng; Mao, Wang; Pang, Cheng-Guo
2016-01-01
Carbon ions have significant advantages in tumor therapy because of their physical and biological properties. In view of the radiation protection, the safety of patients is the most important issue in therapy processes. Therefore, the effects of the secondary particles produced by the carbon ions in the tumor therapy should be carefully considered, especially for the neutrons. In the present work, the neutron radiation field induced by carbon ions was evaluated by using the FLUKA code. The simulated results of neutron energy spectra and neutron dose was found to be in good agreement with the experiment data. In addition, energy deposition of carbon ions and neutrons in tissue-like media was studied, it is found that the secondary neutron energy deposition is not expected to exceed 1% of the carbon ion energy deposition in a typical treatment.
Characterization of neutron field in a NPP workplace
Full text: At Krsko Nuclear Power Plant (NPP) with pressurized water reactor, albedo dosemeters are used for personal neutron dosimetry while survey meters, based on a thermal-neutron detector inside a spherical moderator, are used for dose rate assessment in routine monitoring. The response of both systems is dependent on the energy of the existing neutron fields. Sphere dose rate detector was considered as reference for the calibration of the dosimeters in the workplace. Spectrometric measurements allow determination of the reference dosimetric values and verification of calibration of albedo dosimeters and response of the dose rate detector. The Laboratory for Neutron Metrology and Neutron Dosimetry from the Institute for Radiological Protection and Nuclear Safety (IRSN) was engaged in characterization of neutron fields at several plant locations having high neutron and gamma dose rate. The neutron fields were determined at three typical locations using newly characterized Bonner Sphere System (BSS), based on a cylindrical 3He counter. Measurement results of BSS and of albedo dosimeters are presented in this article. Based on the BSS results, in-situ calibration of NPP dosemeters is discussed. (author)
A Nuclear Energy Agency (NEA), Organization for Economic Co-operation and Development (OECD) benchmark for Uncertainty Analysis in Modeling (UAM) is defined in order to facilitate the development and validation of available uncertainty analysis and sensitivity analysis methods for best-estimate Light water Reactor (LWR) design and safety calculations. The benchmark has been named the OECD/NEA UAM-LWR benchmark, and has been divided into three phases each of which focuses on a different portion of the uncertainty propagation in LWR multi-physics and multi-scale analysis. Several different reactor cases are modeled at various phases of a reactor calculation. This paper discusses Phase I, known as the 'Neutronics Phase', which is devoted mostly to the propagation of nuclear data (cross-section) uncertainty throughout steady-state stand-alone neutronics core calculations. Three reactor systems (for which design, operation and measured data are available) are rigorously studied in this benchmark: Peach Bottom Unit 2 BWR, Three Mile Island Unit 1 PWR, and VVER-1000 Kozloduy-6/Kalinin-3. Additional measured data is analyzed such as the KRITZ LEU criticality experiments and the SNEAK-7A and 7B experiments of the Karlsruhe Fast Critical Facility. Analyzed results include the top five neutron-nuclide reactions, which contribute the most to the prediction uncertainty in keff, as well as the uncertainty in key parameters of neutronics analysis such as microscopic and macroscopic cross-sections, six-group decay constants, assembly discontinuity factors, and axial and radial core power distributions. Conclusions are drawn regarding where further studies should be done to reduce uncertainties in key nuclide reaction uncertainties (i.e.: 238U radiative capture and inelastic scattering (n, n') as well as the average number of neutrons released per fission event of 239Pu).
Automated full-field range OPD and MTF measurement bench for automotive objective benchmark
Boucher, W.; Yonnet, M.; Brahmi, D.; Gascon, A.; Deprez, M.; Wattellier, B.; Lavergne, O.
2015-10-01
We present a metrology bench used to benchmark the optical quality of lenses used in automotive applications. These lenses have a wide field of view, typically larger than 180° and fast numerical apertures (F/2). Lenses are compared in terms of OPD and MTF. For this aim, interferometric solutions would require double-passing the lens. Between two field positions, it would be necessary to realign the reference sphere. This is not compatible with the requested high measurement throughput. Therefore wave front sensing is the best solution to this problem because it is able to characterize lenses in single pass configurations. We chose to measure the transmitted wave front with a QuadriWave Lateral Shearing Inferferometer wave front sensor placed directly after the focal plane. This technology is best suited in this case because it has the unique ability to measure fast diverging beams with high accuracy. It can also accept the large tilt angles due to the lens non telecentricity. Because the information needed for lens comparison is the OPD in the exit pupil, the wave front is then back-propagated to it thanks to an innovative ray-trace algorithm. This algorithm is also able to deduce the aperture transmission to take into account vignetting effects that appear at the field edges. Finally the MTF is simulated from the OPD and the pupil function. This bench is fully automated to rapidly benchmark a large number of lenses in their complete field of view. We characterized the bench according to the ISO 5725 standard. Its precision was tested with commercial lenses whereas trueness was assessed with calibrated lenses. The accuracy on MTF was found to be below 1% over the whole field of view.
Neutron Limit on the Strongly-Coupled Chameleon Field
Li, K; Cory, D G; Haun, R; Heacock, B; Huber, M G; Nsofini, J; Pushin, D A; Saggu, P; Sarenac, D; Shahi, C B; Skavysh, V; Snow, W M; Young, A R
2016-01-01
The physical origin of the dark energy that causes the accelerated expansion rate of the universe is one of the major open questions of cosmology. One set of theories postulates the existence of a self-interacting scalar field for dark energy coupling to matter. In the chameleon dark energy theory, this coupling induces a screening mechanism such that the field amplitude is nonzero in empty space but is greatly suppressed in regions of terrestrial matter density. However measurements performed under appropriate vacuum conditions can enable the chameleon field to appear in the apparatus, where it can be subjected to laboratory experiments. Here we report the most stringent upper bound on the free neutron-chameleon coupling in the strongly-coupled limit of the chameleon theory using neutron interferometric techniques. Our experiment sought the chameleon field through the relative phase shift it would induce along one of the neutron paths inside a perfect crystal neutron interferometer. The amplitude of the cham...
Neutron in a Strong Magnetic Field: Finite Volume Effects
Tiburzi, Brian C
2014-01-01
We investigate the neutron's response to magnetic fields on a torus with the aid of chiral perturbation theory, and expose effects from non-vanishing holonomies. The determination of such effects necessitates non-perturbative treatment of the magnetic field; and, to this end, a strong-field power counting is employed. Using a novel coordinate-space method, we find the neutron propagates in a coordinate-dependent effective potential that we obtain by integrating out charged pions winding around the torus. Knowledge of these finite volume effects will aid in the extraction of neutron properties from lattice QCD computations in external magnetic fields. In particular, we obtain finite volume corrections to the neutron magnetic moment and magnetic polarizability. These quantities have not been computed correctly in the literature. In addition to effects from non-vanishing holonomies, finite volume corrections depend on the magnetic flux quantum through an Aharonov-Bohm effect. We make a number of observations tha...
The neutron field perturbation effect in the Dalat reactor
The perturbation of the thermal neutron field of the Dalat Nuclear Research Reactor is investigated when replacing the fuel element by another material. The similarity between the thermal neutron field distribution curves of the water column and the methylmetha copulation rod is obtained for 5 cells at the different positions in the core. The perturbation of the thermal neutron field when replacing the fuel element or water column by the methylmetha copulation rod is verified. In consequence, it is possible to apply the method of replacing the measurement of the relative distribution of the thermal neutron field on the surface of fuel element by that in the water column or in the methylmetha copulation rod. The measurement may be carried out at power levels of 30 - 50 watts. (author). 5 refs., 4 figs., 4 tabs
Benchmark Modeling of the Near-Field and Far-Field Wave Effects of Wave Energy Arrays
Rhinefrank, Kenneth E; Haller, Merrick C; Ozkan-Haller, H Tuba
2013-01-26
This project is an industry-led partnership between Columbia Power Technologies and Oregon State University that will perform benchmark laboratory experiments and numerical modeling of the near-field and far-field impacts of wave scattering from an array of wave energy devices. These benchmark experimental observations will help to fill a gaping hole in our present knowledge of the near-field effects of multiple, floating wave energy converters and are a critical requirement for estimating the potential far-field environmental effects of wave energy arrays. The experiments will be performed at the Hinsdale Wave Research Laboratory (Oregon State University) and will utilize an array of newly developed Buoys' that are realistic, lab-scale floating power converters. The array of Buoys will be subjected to realistic, directional wave forcing (1:33 scale) that will approximate the expected conditions (waves and water depths) to be found off the Central Oregon Coast. Experimental observations will include comprehensive in-situ wave and current measurements as well as a suite of novel optical measurements. These new optical capabilities will include imaging of the 3D wave scattering using a binocular stereo camera system, as well as 3D device motion tracking using a newly acquired LED system. These observing systems will capture the 3D motion history of individual Buoys as well as resolve the 3D scattered wave field; thus resolving the constructive and destructive wave interference patterns produced by the array at high resolution. These data combined with the device motion tracking will provide necessary information for array design in order to balance array performance with the mitigation of far-field impacts. As a benchmark data set, these data will be an important resource for testing of models for wave/buoy interactions, buoy performance, and far-field effects on wave and current patterns due to the presence of arrays. Under the proposed project we will initiate
Neutron spectrum determination by activation method in fast neutron fields at the RB reactors
The fast neutron fields of the RB reactor are presented in this paper. The activation method for spectrum determination is described and explained. The obtained results for intermediate and fast spectrum are given and discussed. (authors). 7 refs., 3 tabs
It is important to measure the microdistribution of 10B in a cell to predict the cell-killing effect of new boron compounds in the field of boron neutron capture therapy. Alpha autoradiography has generally been used to detect the microdistribution of 10B in a cell. Although it has been performed using a reactor-based neutron source, the realization of an accelerator-based thermal neutron irradiation field is anticipated because of its easy installation at any location and stable operation. Therefore, we propose a method using a cyclotron-based epithermal neutron source in combination with a water phantom to produce a thermal neutron irradiation field for alpha autoradiography. This system can supply a uniform thermal neutron field with an intensity of 1.7×109 (cm−2 s−1) and an area of 40 mm in diameter. In this paper, we give an overview of our proposed system and describe a demonstration test using a mouse liver sample injected with 500 mg/kg of boronophenyl-alanine. - Highlights: • We developed a thermal neutron irradiation field using cyclotron based epithermal neutron source combination with a water phantom for alpha autoradiography. • The uniform thermal neutron irradiation field with an intensity of 1.7×109 (cm−2 s−1) with a size of 40 mm in diameter was obtained. • Demonstration test of alpha autoradiography using a liver sample with the injection of BPA was performed. • Boron image discriminated with the background event of protons was clearly shown by means of the particle identification
Agent code: Neutron transport benchmark example and extension to 3D lattice geometry
Hursin Mathieu
2005-01-01
Full Text Available The general methodology be hind 2D arbitrary geometry neutron transport AGENT code is the theory of R-functions, which al lows for simple modeling of complex geometries, and the method of characteristics, which solves the integral transport equation along characteristic neutron trajectories. This paper focuses on the extension of the methodology to ac count for 3D lattice geometries. Since the direct application of method of characteristics to 3D non-homogenized core con figuration may re quire a tremendous amount of memory and computing time, an alternative approximate solution based on coupling 2D method of characteristics and 1D diffusion solution is developed. The planar 2D method of characteristics and axial 1D diffusion solutions are coupled through the trans verse leak age. The use of a lower order 1D solution in the axial direction is justified by the fact that more heterogeneity in current PWR and BWR reactor cores occurs in the radial direction than in the axial one. In order to demonstrate the versatility and accuracy of the AGENT code, a 2D heterogeneous lattice problem, C5G7 is described in details. A theoretical description of the coupling methodology for 3D method of characteristics solution is followed by preliminary validation in comparison to the DeCART code.
The Binary History and the Magnetic Field of Neutron Star
Konar, Sushan
2009-01-01
There has been strong observational evidence suggesting a causal connection between the binary history of neutron stars and the evolution of their magnetic field. In this article we discuss one of the plausible mechanisms proposed for the evolution of the surface magnetic field, that of the diamagnetic screening of the field by accreted material.
Reference neutron fields for metrology of radiation monitoring
The set of reference neutron fields created in the JINR for metrology of radiation monitoring is described. The set comprises reference fields based on Cf-252 in polyethylene moderators with diameters 12.7 cm and 29.2 cm and two reference fields based on 660 MeV Laboratory of Nuclear Problems phasotron. 11 refs.; 7 figs.; 1 tab
Neutron Star Magnetic Field as for Nonzero Photon Mass
WANG Qing-Wu; L(U) Xiao-Fu
2005-01-01
We investigate the neutron star magnetic field by the relative mean-field theory, where the photon effective mass depending on baryon density of charged particles is nonzero. This field is produced by star itself, which is the function of baryon density. The result fits the observations.
Calculations of radiation flux and dose distributions for Boron Neutron Capture Therapy (BNCT) of brain tumors are typically performed using sophisticated three-dimensional analytical models based on either a homogeneous approximation or a simplified few-region approximation to the actual highly-heterogeneous geometry of the irradiation volume. Such models should be validated by comparison with calculations using detailed models in which all significant macroscopic tissue heterogeneities and geometric structures are explicitly represented as faithfully as possible. This work describes a validation exercise for BNCT of canine brain tumors. Geometric measurements of the canine anatomical structures of interest for this work were performed by dissecting and examining two essentially identical Labrador Retriever heads. Chemical analyses of various tissue samples taken during the dissections were conducted to obtain measurements of elemental compositions for tissues of interest. The resulting geometry and tissue composition data were then used to construct a detailed heterogeneous calculational model of the Labrador Retriever head. Calculations of three-dimensional radiation flux distributions pertinent to BNCT were performed for the model using the TORT discrete-ordinates radiation transport code. The calculations were repeated for a corresponding volume-weighted homogeneous tissue model. Comparison of the results showed that the peak neutron and photon flux magnitudes were quite similar for the two models (within 5%), but that the spatial flux profiles were shifted in the heterogeneous model such that the fluxes in some locations away from the peak differed from the corresponding fluxes in the homogeneous model by as much as 10-20%. Differences of this magnitude can be therapeutically significant, emphasizing the need for proper validation of simplified treatment planning models
Moran, J.M.
1992-02-01
Calculations of radiation flux and dose distributions for Boron Neutron Capture Therapy (BNCT) of brain tumors are typically performed using sophisticated three-dimensional analytical models based on either a homogeneous approximation or a simplified few-region approximation to the actual highly-heterogeneous geometry of the irradiation volume. Such models should be validated by comparison with calculations using detailed models in which all significant macroscopic tissue heterogeneities and geometric structures are explicitly represented as faithfully as possible. This work describes a validation exercise for BNCT of canine brain tumors. Geometric measurements of the canine anatomical structures of interest for this work were performed by dissecting and examining two essentially identical Labrador Retriever heads. Chemical analyses of various tissue samples taken during the dissections were conducted to obtain measurements of elemental compositions for tissues of interest. The resulting geometry and tissue composition data were then used to construct a detailed heterogeneous calculational model of the Labrador Retriever head. Calculations of three-dimensional radiation flux distributions pertinent to BNCT were performed for the model using the TORT discrete-ordinates radiation transport code. The calculations were repeated for a corresponding volume-weighted homogeneous tissue model. Comparison of the results showed that the peak neutron and photon flux magnitudes were quite similar for the two models (within 5%), but that the spatial flux profiles were shifted in the heterogeneous model such that the fluxes in some locations away from the peak differed from the corresponding fluxes in the homogeneous model by as much as 10-20%. Differences of this magnitude can be therapeutically significant, emphasizing the need for proper validation of simplified treatment planning models.
In current radiotherapy, neutrons are produced in a photonuclear reaction when incident photon energy is higher than the threshold. In the present study, a method of discriminating the neutron component was investigated using an imaging plate (IP) in the neutron-gamma-ray mixed field. Two types of IP were used: a conventional IP for beta- and gamma rays, and an IP doped with Gd for detecting neutrons. IPs were irradiated in the mixed field, and the photo-stimulated luminescence (PSL) intensity of the thermal neutron component was discriminated using an expression proposed herein. The PSL intensity of the thermal neutron component was proportional to thermal neutron fluence. When additional irradiation of photons was added to constant neutron irradiation, the PSL intensity of the thermal neutron component was not affected. The uncertainty of PSL intensities was approximately 11.4 %. This method provides a simple and effective means of discriminating the neutron component in a mixed field. (authors)
Thermal evolution of neutron stars with decaying magnetic fields
Rotochemical heating originates in the deviation from beta equilibrium due to spin-down compression, which is closely related to the dipole magnetic field. We numerically calculate the deviation from chemical equilibrium and thermal evolution of neutron stars with decaying magnetic fields. We find that the power-law long term decay of the magnetic field slightly affects the deviation from chemical equilibrium and surface temperature. However, the magnetic decay leads to older neutron stars that could have a different surface temperature with the same magnetic field strength. That is, older neutron stars with a low magnetic field (108 G) could have a lower temperature even with rotochemical heating in operation, which probably explains the lack of other observations on older millisecond pulsars with higher surface temperature, except millisecond pulsar J0437–4715. (paper)
A reference neutron field (RFN) is used as a standard neutron source (SNS) that is influenced by the changes in the reactor core due to recharging or other causes. A whole range of measurements is carried out in a full scope, to specify its characteristics precisely. The SNS comprises: 1) the RNF certificated to the neutron energy spectrum, its location in the reactor field, being a reference measure of the differential energy distribution in the neutron flux; 2) exposure monitoring tools (detectors revealing the certified physical characteristics); 3) functional measurement apparatus (revealing the spectral characteristics). The following basic metrological characteristics are given: differential neutron energy spectrum, described by F(E) [1/cm2.s.MeV], normalized by 1 in the range 3-19 MeV and the measurement error; the conventional neutron flux density and its error. The methodology of measuring the neutron flux integral density comprises the following six steps: 1) assessment of the influence of the changes in the core configuration on the stability of the RNF (estimated in six energy ranges); 2) demonstration of RNF application in reactor physics studies; 3) irradiation of two sets of activation detectors (Au, Sc and Au, Sc, S in Al and Cd shields); 4) measurement of the detector activities by calibrated gamma- and beta- spectrometric apparatus; 5) determination of the neutron field characteristics at a certain point of the RNF by the method of activating ratios; 6) the result accuracy assessment and probabilistic error limits determination with 95% upper bound frequency. The RNF neutron energy range have been measured 6 times for a period of two years. 6 refs., 8 figs. (M.A.)
Validation of IRDFF in 252Cf Standard and IRDF-2002 Reference Neutron Fields
Simakov, Stanislav; Capote, Roberto; Greenwood, Lawrence; Griffin, Patrick; Kahler, Albert; Pronyaev, Vladimir; Trkov, Andrej; Zolotarev, Konstantin
2016-02-01
The results of validation of the latest release of International Reactor Dosimetry and Fusion File, IRDFF-1.03, in the standard 252Cf(s.f.) and reference 235U(nth,f) neutron benchmark fields are presented. The spectrum-averaged cross sections were shown to confirm IRDFF-1.03 in the 252Cf standard spontaneous fission spectrum; that was not the case for the current recommended spectra for 235U(nth,f). IRDFF was also validated in the spectra of the research reactor facilities ISNF, Sigma-Sigma and YAYOI, which are available in the IRDF-2002 collection. The ISNF facility was re-simulated to remove unphysical oscillations in the spectrum. IRDFF-1.03 was shown to reproduce reasonably well the spectrum-averaged data measured in these fields except for the case of YAYOI.
Validation of IRDFF in 252Cf Standard and IRDF-2002 Reference Neutron Fields
Simakov Stanislav
2016-01-01
Full Text Available The results of validation of the latest release of International Reactor Dosimetry and Fusion File, IRDFF-1.03, in the standard 252Cf(s.f. and reference 235U(nth,f neutron benchmark fields are presented. The spectrum-averaged cross sections were shown to confirm IRDFF-1.03 in the 252Cf standard spontaneous fission spectrum; that was not the case for the current recommended spectra for 235U(nth,f. IRDFF was also validated in the spectra of the research reactor facilities ISNF, Sigma-Sigma and YAYOI, which are available in the IRDF-2002 collection. The ISNF facility was re-simulated to remove unphysical oscillations in the spectrum. IRDFF-1.03 was shown to reproduce reasonably well the spectrum-averaged data measured in these fields except for the case of YAYOI.
The freedom to choose neutron star magnetic field equilibria
Glampedakis, Kostas; Lasky, Paul D.
2016-01-01
Our ability to interpret and glean useful information from the large body of observations of strongly magnetised neutron stars rests largely on our theoretical understanding of magnetic field equilibria. We answer the following question: is one free to arbitrarily prescribe magnetic equilibria such that fluid degrees of freedom can balance the equilibrium equations? We examine this question for various models for neutron star matter; from the simplest single-fluid barotrope to more realistic ...
Intercomparison of radiation protection instrumentation in a pulsed neutron field
In the framework of the EURADOS working group 11, an intercomparison of active neutron survey meters was performed in a pulsed neutron field (PNF). The aim of the exercise was to evaluate the performances of various neutron instruments, including commercially available rem-counters, personal dosemeters and instrument prototypes. The measurements took place at the cyclotron of the Helmholtz-Zentrum Berlin für Materialien und Energie GmbH. The cyclotron is routinely used for proton therapy of ocular tumours, but an experimental area is also available. For the therapy the machine accelerates protons to 68 MeV. The interaction of the proton beam with a thick tungsten target produces a neutron field with energy up to about 60 MeV. One interesting feature of the cyclotron is that the beam can be delivered in bursts, with the possibility to modify in a simple and flexible way the burst length and the ion current. Through this possibility one can obtain radiation bursts of variable duration and intensity. All instruments were placed in a reference position and irradiated with neutrons delivered in bursts of different intensity. The analysis of the instrument response as a function of the burst charge (the total electric charge of the protons in the burst shot onto the tungsten target) permitted to assess for each device the dose underestimation due to the time structure of the radiation field. The personal neutron dosemeters were exposed on a standard PMMA slab phantom and the response linearity was evaluated
Intercomparison of radiation protection instrumentation in a pulsed neutron field
Caresana, M., E-mail: marco.caresana@polimi.it [Politecnico di Milano, CESNEF, Dipartimento di Energia, via Ponzio 34/3, 20133 Milano (Italy); Denker, A. [Helmholtz-Zentrum Berlin für Materialien und Energie, Hahn-Meitner-Platz 1, D-14109 Berlin (Germany); Esposito, A. [IFNF-LNF, FISMEL, via E. Fermi 40, 00044 Frascati (Italy); Ferrarini, M. [CNAO, Via Privata Campeggi, 27100 Pavia (Italy); Golnik, N. [Institute of Metrology and Biomedical Engineering, Warsaw University of Technology, Sw. A. Boboli 8, 02-525 Warsaw (Poland); Hohmann, E. [Paul Scherrer Institut (PSI), Radiation Metrology Section, CH-5232 Villigen PSI (Switzerland); Leuschner, A. [Deutsches Elektronen-Synchrotron DESY, Notkestr. 85, 22603 Hamburg (Germany); Luszik-Bhadra, M. [Physikalisch-Technische Bundesanstalt (PTB), Bundesallee 100, 38116 Braunschweig (Germany); Manessi, G. [CERN, 1211 Geneva 23 (Switzerland); University of Liverpool, Department of Physics, L69 7ZE Liverpool (United Kingdom); Mayer, S. [Paul Scherrer Institut (PSI), Radiation Metrology Section, CH-5232 Villigen PSI (Switzerland); Ott, K. [Helmholtz-Zentrum Berlin, BESSYII, Albert-Einstein-Str.15, 12489 Berlin (Germany); Röhrich, J. [Helmholtz-Zentrum Berlin für Materialien und Energie, Hahn-Meitner-Platz 1, D-14109 Berlin (Germany); Silari, M. [CERN, 1211 Geneva 23 (Switzerland); Trompier, F. [Institute for Radiological Protection and Nuclear Safety, F-92262 Fontenay aux Roses (France); Volnhals, M.; Wielunski, M. [Helmholtz Zentrum München, Ingolstädter Landstr. 1, D-85764 Neuherberg (Germany)
2014-02-11
In the framework of the EURADOS working group 11, an intercomparison of active neutron survey meters was performed in a pulsed neutron field (PNF). The aim of the exercise was to evaluate the performances of various neutron instruments, including commercially available rem-counters, personal dosemeters and instrument prototypes. The measurements took place at the cyclotron of the Helmholtz-Zentrum Berlin für Materialien und Energie GmbH. The cyclotron is routinely used for proton therapy of ocular tumours, but an experimental area is also available. For the therapy the machine accelerates protons to 68 MeV. The interaction of the proton beam with a thick tungsten target produces a neutron field with energy up to about 60 MeV. One interesting feature of the cyclotron is that the beam can be delivered in bursts, with the possibility to modify in a simple and flexible way the burst length and the ion current. Through this possibility one can obtain radiation bursts of variable duration and intensity. All instruments were placed in a reference position and irradiated with neutrons delivered in bursts of different intensity. The analysis of the instrument response as a function of the burst charge (the total electric charge of the protons in the burst shot onto the tungsten target) permitted to assess for each device the dose underestimation due to the time structure of the radiation field. The personal neutron dosemeters were exposed on a standard PMMA slab phantom and the response linearity was evaluated.
Stability of magnetic fields of isolated and binary neutron stars
It is suggested that convective instabilities in cooling neutron stars may lead to magnetic field decay. Since rotation may have a stabilizing influence, the rotational history of the star is more important, than the age of the star, in determining whether its magnetic field decays or not. 27 references
The calibration of a personnel neutron dosemeter in different neutron fields is described. The badge-like dosemeter contains 5 detectors: polycarbonate foil (10 μm, Makrofol KG), 232Th, natural uranium, natural uranium with boron, and natural uranium with cadmium. Detector sensitivity and calibration factors have been calculated and measured in radiation fields of 252Cf fission neutrons, WWR-S reactor neutrons with and without Cd and Fe shielding, 3-MeV (d,t) generator neutrons, and 238PuBe neutrons. Measurement range and achievable accuracy are discussed from the point of view of applying the dosemeter in routine and emergency uses
An active pixels spectrometers for neutronic fields metrology
The fundamental metrology is responsible for the sustainability of the measurement systems and handles to supply the reference standards. Concerning the metrology of ionizing radiations and, in particular the neutron metrology, detectors standards are used to characterize reference fields, in terms of energy and fluence. The dosimeters or particle detectors are calibrated on these reference fields. This thesis presents the development of a neutron spectrometer neutron candidate to the status of primary standard for the characterization of neutron fields in the range from 5 to 20 MeV. The spectrometer uses the recoil proton telescope as detection principle; the CMOS technology, through three sensor positions, is taking advantage to realize the tracking of protons. A Si(Li) detector handles the measure of the residual proton energy. The device simulations, realized under MCNPX, allow to estimate its performances and to validate the neutron energy reconstruction. An essential step of characterization of the telescope elements and in particular of CMOS sensors is also proposed to guarantee the validity of posterior experimental measurements. The tests realized as well in mono-energy fields as in radionuclide source show the very good performances of the system. The quantification of uncertainties indicates an energy estimation with 1.5 % accuracy and a resolution of less than 6 %. The fluence measurement is performed with an uncertainty about 4 to 6%. (author)
Neutron Limit on the Strongly-Coupled Chameleon Field
Pushin, Dmitry
2016-03-01
One of the major open questions of cosmology is the physical origin of the dark energy. There are a few sets of theories which might explain this origin that could be tested experimentally. The chameleon dark energy theory postulates self-interacting scalar field that couples to matter. This coupling induces a screening mechanism chosen so that the field amplitude is nonzero in empty space but is greatly suppressed in regions of terrestrial matter density. On behalf of the INDEX collaboration, I will report the most stringent upper bound on the free neutron-chameleon coupling in the strongly-coupled limit of the chameleon theory using neutron interferometric techniques. In our experiment we measure neutron phase induced by chameleon field. We report a 95 % confidence level upper bound on the neutron-chameleon coupling ranging from β < 4 . 7 ×106 for a Ratra-Peebles index of n = 1 in the nonlinear scalar field potential to β < 2 . 4 ×107 for n = 6 , one order of magnitude more sensitive than the most recent free neutron limit for intermediate n. This work was supported by NIST; NSF Grants: PHY-1205342, PHY-1068712, PHY-1307426; DOE award DE-FG02-97ER41042; NSERC CREATE and DISCOVERY programs; CERC; IUCSS and IU FRS program.
Cooling of Neutron Stars with Strong Toroidal Magnetic Fields
Page, D; Küker, M; Page, Dany; Geppert, Ulrich; Kueker, Manfred
2007-01-01
We present models of temperature distribution in the crust of a neutron star in the presence of a strong toroidal component superposed to the poloidal component of the magnetic field. The presence of such a toroidal field hinders heat flow toward the surface in a large part of the crust. As a result, the neutron star surface presents two warm regions surrounded by extended cold regions and has a thermal luminosity much lower than in the case the magnetic field is purely poloidal. We apply these models to calculate the thermal evolution of such neutron stars and show that the lowered photon luminosity naturally extends their life-time as detectable thermal X-ray sources.
Neutron matter under strong magnetic fields: a comparison of models
Aguirre, R; Vidaña, I
2013-01-01
The equation of state of neutron matter is affected by the presence of a magnetic field due to the intrinsic magnetic moment of the neutron. Here we study the equilibrium configuration of this system for a wide range of densities, temperatures and magnetic fields. Special attention is paid to the behavior of the isothermal compressibility and the magnetic susceptibility. Our calculation is performed using both microscopic and phenomenological approaches of the neutron matter equation of state, namely the Brueckner--Hartree--Fock (BHF) approach using the Argonne V18 nucleon-nucleon potential supplemented with the Urbana IX three-nucleon force, the effective Skyrme model in a Hartree--Fock description, and the Quantum Hadrodynamic formulation with a mean field approximation. All these approaches predict a change from completely spin polarized to partially polarized matter that leads to a continuous equation of state. The compressibility and the magnetic susceptibility show characteristic behaviors, which reflec...
Neutron limit on the strongly-coupled chameleon field
Li, K.; Arif, M.; Cory, D. G.; Haun, R.; Heacock, B.; Huber, M. G.; Nsofini, J.; Pushin, D. A.; Saggu, P.; Sarenac, D.; Shahi, C. B.; Skavysh, V.; Snow, W. M.; Young, A. R.; Index Collaboration
2016-03-01
The physical origin of the dark energy that causes the accelerated expansion rate of the Universe is one of the major open questions of cosmology. One set of theories postulates the existence of a self-interacting scalar field for dark energy coupling to matter. In the chameleon dark energy theory, this coupling induces a screening mechanism such that the field amplitude is nonzero in empty space but is greatly suppressed in regions of terrestrial matter density. However measurements performed under appropriate vacuum conditions can enable the chameleon field to appear in the apparatus, where it can be subjected to laboratory experiments. Here we report the most stringent upper bound on the free neutron-chameleon coupling in the strongly coupled limit of the chameleon theory using neutron interferometric techniques. Our experiment sought the chameleon field through the relative phase shift it would induce along one of the neutron paths inside a perfect crystal neutron interferometer. The amplitude of the chameleon field was actively modulated by varying the millibar pressures inside a dual-chamber aluminum cell. We report a 95% confidence level upper bound on the neutron-chameleon coupling β ranging from β <4.7 ×106 for a Ratra-Peebles index of n =1 in the nonlinear scalar field potential to β <2.4 ×107 for n =6 , one order of magnitude more sensitive than the most recent free neutron limit for intermediate n . Similar experiments can explore the full parameter range for chameleon dark energy in the foreseeable future.
A compact neutron scatter camera for field deployment.
Goldsmith, John E M; Gerling, Mark D; Brennan, James S
2016-08-01
We describe a very compact (0.9 m high, 0.4 m diameter, 40 kg) battery operable neutron scatter camera designed for field deployment. Unlike most other systems, the configuration of the sixteen liquid-scintillator detection cells are arranged to provide omnidirectional (4π) imaging with sensitivity comparable to a conventional two-plane system. Although designed primarily to operate as a neutron scatter camera for localizing energetic neutron sources, it also functions as a Compton camera for localizing gamma sources. In addition to describing the radionuclide source localization capabilities of this system, we demonstrate how it provides neutron spectra that can distinguish plutonium metal from plutonium oxide sources, in addition to the easier task of distinguishing AmBe from fission sources. PMID:27587113
Primm III, RT
2002-05-29
This volume of the progress report provides documentation of reactor physics and criticality safety studies conducted in the US during fiscal year 1997 and sponsored by the Fissile Materials Disposition Program of the US Department of Energy. Descriptions of computational and experimental benchmarks for the verification and validation of computer programs for neutron physics analyses are included. All benchmarks include either plutonium, uranium, or mixed uranium and plutonium fuels. Calculated physics parameters are reported for all of the computational benchmarks and for those experimental benchmarks that the US and Russia mutually agreed in November 1996 were applicable to mixed-oxide fuel cycles for light-water reactors.
Quantification of neutron field at the neutron therapy room of KCCH using a Bonner sphere
In order to quantify the neutron fields at the neutron therapy room of KCCH the Bonner Sphere spectrometry system (BS) was used for the measurement of neutron spectra produced from two kinds of Be targets (1.0 and 10.5 mm bombarded by protons of 35 and 45 MeV. It was found that additional neutrons produced from the beam line tube and the beam stopper, which are made of Aluminum, were included considerably as a part of neutron spectrum in the neutron field made from the thin (1.0 mm) Be target. Neutrons from the thick (10.5 mm) Be were hardened by a iron filter of 2.6 cm and collimated by the gantry, and the beam size was fitted 26 x 16 cm2) to cover the cross sectional area of a BS used in this measurement. Six kinds of neutron spectra were measured and the dosimetric quantities such as the fluence averaged energy (Eave.), the spectrum weighted dose conversion coefficient (h*) and the dose equivalent rate (H) per nano ampere were determined. These were ranged as follows, Eave. was from 4.3 to 15.1 MeV, and h* was from 326 to 447 pSv.cm2, and H was from 0.17 to 5.66 mSv.h-1.nA-1. The MXDFC31 code was used to unfold the measured data of BS and the MCNPX code (Ver. 2.4) implemented to calculate the default spectra which are necessary for unfolding as a prior information
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
The Covariance and Bicovariance of the Stochastic Neutron Field
On the basis of the general stochastic neutron field theory developed by Munoz-Cobo et al, results on the covariance and bicovariance of the neutron field have been presented. These two statistical quantities are obtained from the counts observed in detectors operating during a period of time (gate length), Δqc. A classical example is the so called Feynmann Y-function that is defined as the variance to mean ratio of the neutron field. Upon taking the limit of the covariance and bicovariance function for Δqc rarrow O , one obtains the two and three detector cross correlation functions respectively. The mathematical structure of the results so obtained have a transparent physical interpretation in terms of the space and delay time overlap between the field-of-view of the detectors. For the first time, an expression has been obtained for the bispectrum function of the stochastic neutron field and for the appropriate weight functions to be used as space-energy-angle correction factors for the one-point kinetics approximation
Evolutions of Neutron Stars and their Magnetic Fields
Bisnovatyi-Kogan, G S
2016-01-01
Estimations of magnetic fields of neutron stars, observed as radio and X-ray pulsars, are discussed. It is shown, that theoretical and observational values for different types of radiopulsars are in good correspondence. Radiopulsars in close binaries and millisecond pulsars, which have passed the stage of disk accretion (recycled radiopulsars), have magnetic fields 2-4 orders of magnitude smaller than ordinary single pulsars. Most probably, the magnetic field of the neutron star was screened by the infalling material. Several screening models are considered. Formation of single recycled pulsars loosing its companion is discussed. Magnetic fields of some X-ray pulsars are estimated from the cyclotron line energy. In the case of Her X-1 this estimation exceeds considerably the value of its magnetic field obtained from long term observational data related to the beam structure evolution. Another interpretation of the cyclotron feature, based on the relativistic dipole radiation mechanism, could remove this discr...
Due to a need for security screening instruments capable of detecting explosives and nuclear materials there is growing interest in neutron generator systems suitable for field use for applications broadly referred to as active neutron interrogation (ANI). Over the past two years Thermo Electron Corporation has developed a suite of different compact accelerator neutron generator products specifically designed for ANI field work to meet this demand. These systems incorporate hermetically-sealed particle accelerator tubes designed to produce fast neutrons using either the deuterium-deuterium (En = 2.5 MeV) or deuterium-tritium (En = 14.1 MeV) fusion reactions. Employing next-generation features including advanced sealed-tube accelerator designs, all-digital control electronics and innovative housing configurations these systems are suitable for many different uses. A compact system weighing less than 14 kg (MP 320) with a lifetime exceeding 1000 hours has been developed for portable applications. A system for fixed installations (P 325) has been developed with an operating life exceeding 4500 hours that incorporates specific serviceability features for permanent facilities with difficult-to-access shield blocks. For associated particle imaging (API) investigations a second-generation system (API 120) with an operating life of greater than 1000 hours has been developed for field use in which a high resolution fiberoptic imaging plate is specially configured to take advantage of a neutron point-source spot size of ∼2 mm. (author)
The effect of angular anisotropy in inelastic secondary neutron scattering on neutron leakage spectra from the surface of spherical specimens is investigated. It is shown how inadequate representation of the cross-section structure in the neutron energy resonance region can affect the neutron leakage spectrum. (author). 19 refs, 5 figs, 6 tabs
Field neutron spectrometer using 3He, TEPC, and multisphere detectors
Since the last DOE Neutron Dosimetry Workshop, there have been a number of changes in radiation protection standards proposed by national and international advisory bodies. These changes include: increasing quality factors for neutrons by a factor of two, defining quality factors as a function of lineal energy rather than linear energy transfer (see ACCRUE-40; Joint Task Group 1986), and adoption of effective dose equivalent methodologies. In order to determine the effects of these proposed changes, it is necessary to know the neutron energy spectrum in the work place. In response to the possible adoption of these proposals, the Department of Energy (DOE) initiated a program to develop practical neutron spectrometry systems for use by health physicists. One part of this program was the development of a truly portable, battery operated liquid scintillator spectrometer using proprietary electronics developed at Lawrence Livermore National Laboratory (LLNL); this instrument will be described in the following paper. The second part was the development at PNL of a simple transportable spectrometer based on commercially available electronics. This open-quotes field neutron spectrometerclose quotes described in this paper is intended to be used over a range of neutron energies extending from thermal to 20 MeV
A compact neutron generator using a field ionization source
Persaud, Arun; Waldmann, Ole; Schenkel, Thomas [E.O. Lawrence Berkeley National Laboratory, Berkeley, California 94720 (United States); Kapadia, Rehan; Takei, Kuniharu; Javey, Ali [Department of Electrical Engineering and Computer Sciences, University of California at Berkeley, Berkeley, California 94720 (United States)
2012-02-15
Field ionization as a means to create ions for compact and rugged neutron sources is pursued. Arrays of carbon nano-fibers promise the high field-enhancement factors required for efficient field ionization. We report on the fabrication of arrays of field emitters with a density up to 10{sup 6} tips/cm{sup 2} and measure their performance characteristics using electron field emission. The critical issue of uniformity is discussed, as are efforts towards coating the nano-fibers to enhance their lifetime and surface properties.
A compact neutron generator using a field ionization source
Field ionization as a means to create ions for compact and rugged neutron sources is pursued. Arrays of carbon nano-fibers promise the high field-enhancement factors required for efficient field ionization. We report on the fabrication of arrays of field emitters with a density up to 106 tips/cm2 and measure their performance characteristics using electron field emission. The critical issue of uniformity is discussed, as are efforts towards coating the nano-fibers to enhance their lifetime and surface properties.
The meeting of the Radiation Energy Spectra Unfolding Workshop organized by the Radiation Shielding Information Center is discussed. The plans of the unfolding code benchmarking effort to establish methods of standardization for both the few channel neutron and many channel gamma-ray and neutron spectroscopy problems are presented
Benchmarking of CAD-based SuperMC with ITER benchmark model
Neutronics design and analysis of fusion reactors is significantly complex mainly on geometry and physical process of neutron. The great challenges brought by advanced nuclear energy system promote the development of Super Monte Carlo Calculation Program for Nuclear and Radiation Process (SuperMC). The ITER benchmark model, a verification model created by ITER International Organization, was used for benchmarking the latest SuperMC which can perform CAD-based neutron and photon transport calculation. The calculation results of SuperMC for the first wall, divertor cassettes, inboard toroidal field coils and equatorial port were compared with the results of MCNP and the results were coincident. The intelligence and advantage of SuperMC on automatic conversion from complicated CAD model to full format calculation model, complex source construction and geometry description method was demonstrated. The correctness of neutron and photon transport in energy range corresponding to fusion reactors was also demonstrated
Fast neutron fields imaging with a CCD-based luminescent detector
The paper considers some questions concerned with the development of an imaging system based on a CCD-detector for visualising fast neutron fields. From those the most important are: development of fast neutron screens, detector resistance to irradiation fields, and feasibility of fast neutron radiography and tomography at various neutron sources
Fast neutron fields imaging with a CCD-based luminescent detector
Mikerov, V
1999-01-01
The paper considers some questions concerned with the development of an imaging system based on a CCD-detector for visualising fast neutron fields. From those the most important are: development of fast neutron screens, detector resistance to irradiation fields, and feasibility of fast neutron radiography and tomography at various neutron sources.
Semiclassical description of neutron polarization in an inhomogeneous magnetic field
Naida, O.N.; Prudkovskii, A.G.
1978-12-01
A method is obtained for constructing the semiclassical solutions to the Pauli equation for the neutron in a magnetic field which describe both the Stern-Gerlach beam splitting and spin-flip processes. The accuracy of these solutions makes it possible (in contrast to the known semiclassical methods) to describe interference phenomena, which can be interpreted as corrections to the Larmor precession.
Phantom experiments to evaluate thermal neutron flux distribution were performed using the Scintillator with Optical Fiber (SOF) detector, which was developed as a thermal neutron monitor during boron neutron capture therapy (BNCT) irradiation. Compared with the gold wire activation method and Monte Carlo N-particle (MCNP) calculations, it was confirmed that the SOF detector is capable of measuring thermal neutron flux as low as 105 n/cm2/s with sufficient accuracy. The SOF detector will be useful for phantom experiments with BNCT neutron fields from low-current accelerator-based neutron sources. (author)
The Gamma-3 assembly is located at the Joint Institute for Nuclear Research, Dubna, Russia. It consists of a cylindrical lead target (ø = 8 cm, L = 58.8 cm) surrounded by reactor grade graphite (110 × 110 × 60 cm). The target was irradiated with a beam of 1.6 GeV deuterons from the Nuclotron accelerator and CR-39 track detectors coupled to LR-115 2B film were used to measure the slow neutron distribution on the surface of the graphite. The detection efficiency of the CR-39 in the CR-39/LR-115 2B system was measured using a custom made calibration setup and found to be (1.12 ± 0.05) × 10−3 and (6.1 ± 1.2) × 10−4 tracks per neutron, for thermal and epithermal neutrons respectively, under the etching and counting procedures described in this work. The irradiation of the Gamma-3 was also simulated using MCNPX 2.7 Monte Carlo code and good agreement between the experimental and calculated track densities was found. This serves as a good validation for the computational models used to simulate spallation neutron production, transport and moderation. - Highlights: • Distribution of graphite moderated spallation neutrons measured with CR39/LR115 2B. • The spallation neutrons were generated by interaction of 1.6 GeV d with Pb-target. • CR-39 detector was calibrated using a standard neutron field. • The thermal and epithermal neutron fluences were determined. • Experimental findings are in good agreement with MCNPX code predictions
Tissue-equivalent proportional counters (TEPC) can potentially be used as a portable and personal dosemeter in mixed neutron and gamma-ray fields, but what hinders this use is their typically large physical size. To formulate compact TEPC designs, the use of a Monte Carlo transport code is necessary to predict the performance of compact designs in these fields. To perform this modelling, three candidate codes were assessed: MCNPX 2.7.E, FLUKA 2011.2 and PHITS 2.24. In each code, benchmark simulations were performed involving the irradiation of a 5-in. TEPC with monoenergetic neutron fields and a 4-in. wall-less TEPC with monoenergetic gamma-ray fields. The frequency and dose mean lineal energies and dose distributions calculated from each code were compared with experimentally determined data. For the neutron benchmark simulations, PHITS produces data closest to the experimental values and for the gamma-ray benchmark simulations, FLUKA yields data closest to the experimentally determined quantities. (authors)
In order to reassesses the available design results of Test Breeder Modules (TBMs) a framework contract agreement between F4E and IDOM-Spain has been signed. SEA SL-Spain and UNED-Spain participate as sub-contractors of IDOM. In this study, a qualification of MCNPX code and nuclear data libraries are performed with benchmarking of measured tritium production and neutron flux at the mock-up of the EU TBM, HCPB concept. The irradiation and measurements had been performed in the frame of European Fusion Technology Program by ENEA-Italy, TUD-Germany and JAERI -Japan.
A comparison of neutron beams for BNCT based on in-phantom neutron field assessment parameters
In this paper our in-phantom neutron field assessment parameters, T and DTumor, were used to evaluate several neutron sources for use in BNCT. Specifically, neutron fields from The Ohio State University (OSU) Accelerator-Based Neutron Source (ABNS) design, two alternative ABNS designs from the literature (the Al/AlF3-Al2O3 ABNS and the 7LiF-Al2O3 ABNS), a fission-convertor plate concept based on the 500-kW OSU Research Reactor (OSURR), and the Brookhaven Medical Research Reactor (BMRR) facility were evaluated. In order to facilitate a comparison of the various neutron fields, values of T and DTumor were calculated in a 14 cmx14 cmx14 cm lucite cube phantom located in the treatment port of each neutron source. All of the other relevant factors, such as phantom materials, kerma factors, and treatment parameters, were kept the same. The treatment times for the OSURR, the 7LiF-Al2O3 ABNS operating at a beam current of 10 mA, and the BMRR were calculated to be comparable and acceptable, with a treatment time per fraction of approximately 25 min for a four fraction treatment scheme. The treatment time per fraction for the OSU ABNS and the Al/AlF3-Al2O3 ABNS can be reduced to below 30 min per fraction for four fractions, if the proton beam current is made greater than approximately 20 mA. DTumor was calculated along the beam centerline for tumor depths in the phantom ranging from 0 to 14 cm. For tumor depths ranging from 0 to approximately 1.5 cm, the value of DTumor for the OSURR is largest, while for tumor depths ranging from 1.5 to approximately 14 cm, the value of DTumor for the OSU-ABNS is the largest
A compact neutron generator using a field ionization source
Persaud, Arun; Waldmann, Ole; Kapadia, Rehan; Takei, Kuniharu; Javey, Ali; Schenkel, Thomas
2011-10-31
Field ionization as a means to create ions for compact and rugged neutron sources is pursued. Arrays of carbon nano-bers promise the high eld-enhancement factors required for efficient field ionization. We report on the fabrication of arrays of field emitters with a density up to 10{sup 6} tips/cm{sup 2} and measure their performance characteristics using electron field emission. The critical issue of uniformity is discussed, as are efforts towards coating the nano-fibers to enhance their lifetime and surface properties.
An investigation of neutron and gamma fields originating from the operation of a nuclear reactor
The testing and usage of a suite of computer programs applicable to 1-D shielding analyses is described. The motion of radiation through matter and its interaction with the medium through which it propagates is described by the Boltzmann transport equation. This complex equation usually cannot be solved analytically. It can however be re-written in the so-called Discrete Ordinates form which is successfully solved by digital computer. Computer software is an invaluable tool in finding cost effective solutions to the problem of ensuring that people and materials are safely shielded from the harmful effects of radiation. The package of computer program SCALE-3 contains several modules which can be used to devise solutions to 1-D shielding problems. However, before they can be applied, the user's ability to obtain and interpret meaningful results must be validated. This is achieved by means of comparative studies called benchmarks. Two benchmark studies were performed. The first assesses the ability to apply the software in question to solve a problem associated with the safe transportation of spent nuclear fuel. This was achieved by repeating the shielding analysis of a simplified cast iron cask found in the literature and comparing results. For the second benchmark, the properties of neutron and gamma-ray fields found around an operating nuclear reactor were evaluated and compared to those obtained using alternative technology. Using the computer technology thus tested, a complete shielding analysis of a cask being considered for the transportation of spent nuclear fuel was performed. 27 figs., 9 tabs., 34 refs
Field integral correction in neutron resonance spin echo
Neutron resonance spin echo (NRSE) as a variant of neutron spin echo (NSE) has the advantage that it needs only relatively small magnetic coils. Field inhomogeneities are therefore less important than in NSE. We have built a new type of NRSE spectrometer that overcomes the main limitation of NRSE towards high-energy resolution. Our setup profits from a new longitudinal NRSE field geometry which allows to use Fresnel coils correcting for the beam divergence effect, while former NRSE setups with transversal static magnetic fields could not use Fresnel coils. We demonstrate the function of the longitudinal resonance flip coils, and show first results of spin echo test measurements performed by means of the new setup