WorldWideScience

Sample records for calculated neutron air

  1. Calculated neutron air kerma strength conversion factors for a generically encapsulated Cf-252 brachytherapy source

    CERN Document Server

    Rivard, M J; D'Errico, F; Tsai, J S; Ulin, K; Engler, M J

    2002-01-01

    The sup 2 sup 5 sup 2 Cf neutron air kerma strength conversion factor (S sub K sub N /m sub C sub f) is a parameter needed to convert the radionuclide mass (mu g) provided by Oak Ridge National Laboratory into neutron air kerma strength required by modern clinical brachytherapy dosimetry formalisms indicated by Task Group No. 43 of the American Association of Physicists in Medicine (AAPM). The impact of currently used or proposed encapsulating materials for sup 2 sup 5 sup 2 Cf brachytherapy sources (Pt/Ir-10%, 316L stainless steel, nitinol, and Zircaloy-2) on S sub K sub N /m sub C sub f was calculated and results were fit to linear equations. Only for substantial encapsulation thicknesses, did S sub K sub N /m sub C sub f decrease, while the impact of source encapsulation composition is increasingly negligible as Z increases. These findings are explained on the basis of the non-relativistic kinematics governing the majority of sup 2 sup 5 sup 2 Cf neutron interactions. Neutron kerma and energy spectra resul...

  2. Calculation Package: Derivation of Facility-Specific Derived Air Concentration (DAC) Values in Support of Spallation Neutron Source Operations

    Energy Technology Data Exchange (ETDEWEB)

    McLaughlin, David A [ORNL

    2009-12-01

    Derived air concentration (DAC) values for 175 radionuclides* produced at the Oak Ridge National Laboratory (ORNL) Spallation Neutron Source (SNS), but not listed in Appendix A of 10 CFR 835 (01/01/2009 version), are presented. The proposed DAC values, ranging between 1 E-07 {micro}Ci/mL and 2 E-03 {micro}Ci/mL, were calculated in accordance with the recommendations of the International Commission on Radiological Protection (ICRP), and are intended to support an exemption request seeking regulatory relief from the 10 CFR 835, Appendix A, requirement to apply restrictive DACs of 2E-13 {micro}Ci/mL and 4E-11 {micro}Ci/mL and for non-listed alpha and non-alpha-emitting radionuclides, respectively.

  3. Calculated neutron intensities for SINQ

    Energy Technology Data Exchange (ETDEWEB)

    Atchison, F

    1998-03-01

    A fully detailed calculation of the performance of the SINQ neutron source, using the PSI version of the HETC code package, was made in 1996 to provide information useful for source commissioning. Relevant information about the formulation of the problem, cascade analysis and some of the results are presented. Aspects of the techniques used to verify the results are described and discussed together with a limited comparison with earlier results obtained from neutron source design calculations. A favourable comparison between the measured and calculated differential neutron flux in one of the guides gives further indirect evidence that such calculations can give answers close to reality in absolute terms. Due to the complex interaction between the many nuclear (and other) models involved, no quantitative evaluation of the accuracy of the calculational method in general terms can be given. (author) refs., 13 figs., 9 tabs.

  4. AIR COOLED NEUTRONIC REACTOR

    Science.gov (United States)

    Fermi, E.; Szilard, L.

    1958-05-27

    A nuclear reactor of the air-cooled, graphite moderated type is described. The active core consists of a cubicle mass of graphite, approximately 25 feet in each dimension, having horizontal channels of square cross section extending between two of the opposite faces, a plurality of cylindrical uranium slugs disposed in end to end abutting relationship within said channels providing a space in the channels through which air may be circulated, and a cadmium control rod extending within a channel provided in the moderator. Suitable shielding is provlded around the core, as are also provided a fuel element loading and discharge means, and a means to circulate air through the coolant channels through the fuel charels to cool the reactor.

  5. Relativistic calculations of coalescing binary neutron stars

    Indian Academy of Sciences (India)

    Relativistic calculations of coalescing binary neutron stars. JOSHUA FABER, PHILIPPE GRANDCLÉMENT and FREDERIC RASIO. Department of Physics and Astronomy, Northwestern University, Evanston,. IL 60208-0834, USA. E-mail: rasio@mac.com. Abstract. We have designed and tested a new relativistic Lagrangian ...

  6. Method for determining air humidity in wind tunnels using a neutron emitter

    Science.gov (United States)

    Chegorian, M. A.; Moskalev, V. M.; Ignatov, Iu. A.; Zhuravlev, V. N.

    The paper presents a theoretical analysis of a neutron-hygrometer technique for determining the humidity of an air jet in a wind tunnel. Formulas are presented for calculating the activity of the neutron emitter. Results are presented for a Pu-239 neutron source.

  7. Neutron Detector Signal Processing to Calculate the Effective Neutron Multiplication Factor of Subcritical Assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Talamo, Alberto [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division; Gohar, Yousry [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division

    2016-06-01

    This report describes different methodologies to calculate the effective neutron multiplication factor of subcritical assemblies by processing the neutron detector signals using MATLAB scripts. The subcritical assembly can be driven either by a spontaneous fission neutron source (e.g. californium) or by a neutron source generated from the interactions of accelerated particles with target materials. In the latter case, when the particle accelerator operates in a pulsed mode, the signals are typically stored into two files. One file contains the time when neutron reactions occur and the other contains the times when the neutron pulses start. In both files, the time is given by an integer representing the number of time bins since the start of the counting. These signal files are used to construct the neutron count distribution from a single neutron pulse. The built-in functions of MATLAB are used to calculate the effective neutron multiplication factor through the application of the prompt decay fitting or the area method to the neutron count distribution. If the subcritical assembly is driven by a spontaneous fission neutron source, then the effective multiplication factor can be evaluated either using the prompt neutron decay constant obtained from Rossi or Feynman distributions or the Modified Source Multiplication (MSM) method.

  8. Delayed neutron calculations using ENDF/B-V data

    Energy Technology Data Exchange (ETDEWEB)

    England, T.R.; Shenter, R.E.; Schmittroth, F.

    1979-01-01

    Data from 20 fission yield sets in ENDF/B-V are used with the emission probabilities (Pn) of 105 delayed neutron precursors to calculate 6-group and total equilibrium delayed neutron yields of (anti ..nu../sub d/). Results are compared with recent evaluations and selected measurements of anti ..nu../sub d/. Least-squares data adjustment methods are in progress to improve agreement; preliminary results indicate significant improvement when fission yields are adjusted. 2 tables.

  9. Neutronic investigation and activation calculation for CFETR HCCB blankets

    Science.gov (United States)

    Shuling, XU; Mingzhun, LEI; Sumei, LIU; Kun, LU; Kun, XU; Kun, PEI

    2017-12-01

    The neutronic calculations and activation behavior of the proposed helium cooled ceramic breeder (HCCB) blanket were predicted for the Chinese Fusion Engineering Testing Reactor (CFETR) design model using the MCNP multi-particle transport code and its associated data library. The tritium self-sufficiency behavior of the HCCB blanket was assessed, addressing several important breeding-related arrangements inside the blankets. Two candidate first wall armor materials were considered to obtain a proper tritium breeding ratio (TBR). Presentations of other neutronic characteristics, including neutron flux, neutron-induced damages in terms of the accumulated dpa and helium production were also conducted. Activation, decay heat levels and contact dose rates of the components were calculated to estimate the neutron-induced radioactivity and personnel safety. The results indicate that neutron radiation is efficiently attenuated and slowed down by components placed between the plasma and toroidal field coil. The dominant nuclides and corresponding isotopes in the structural steel were discussed. A radioactivity comparison between pure beryllium and beryllium with specific impurities was also performed. After a millennium cooling time, the decay heat of all the concerned components and materials is less than 1 × 10‑4 kW, and most associated in-vessel components qualify for recycling by remote handling. The results demonstrate that acceptable hands-on recycling and operation still require a further long waiting period to allow the activated products to decay.

  10. Relativistic calculations of coalescing binary neutron stars

    Indian Academy of Sciences (India)

    We have designed and tested a new relativistic Lagrangian hydrodynamics code, which treats gravity in the conformally flat approximation to general relativity. We have tested the resulting code extensively, finding that it performs well for calculations of equilibrium single-star models, collapsing relativistic dust clouds, and ...

  11. Calculation of prompt neutron spectra for curium isotopes

    Energy Technology Data Exchange (ETDEWEB)

    Ohsawa, Takaaki [Kinki Univ., Higashi-Osaka, Osaka (Japan). Atomic Energy Research Inst.

    1997-03-01

    With the aim of checking the existing evaluations contained in JENDL-3.2 and providing new evaluations based on a methodology proposed by the author, a series of calculations of prompt neutron spectra have been undertaken for curium isotopes. Some of the evaluations in JENDL-3.2 was found to be unphysically hard and should be revised. (author)

  12. Software for neutron physical calculations of fast neutron reactors. [PRIDAN, MED, ANALIT

    Energy Technology Data Exchange (ETDEWEB)

    Jordanov, J.; Antonov, N. (Bylgarska Akademiya na Naukite, Sofia. Inst. za Yadrena Izsledvaniya i Yadrena Energetika)

    1983-01-01

    A set of programs has been developed for neutron physical calculations of fast neutron reactors. The set includes the PRIDAN program for calculating the effective cross-sections of the media, characteristics of the fast reactor systems, the MED program - a single dimension multigroup program for calculating fast reactors in multigroup diffusion approximation and the ANALIT program for calculating the criticality in plane geometry. PRIDAN uses the formalism of the self-shielding factors and prepares the effective cross-sections using an iterative procedure. The values of the multigroup cross-sections obtained for different temperatures are used directly as input data for the other programs. MED calculates the critical dimensions, the coefficient of effective multiplication, the real and conjugated neutron fluxes for each enegry group and each spatial zone, the distribution of the neutron fission sources. The maximum number of enegry groups - 26 and the maximum number of spatial points - 170, are distributed in seven spatial zones at most. The ANALIT program solves under given conditions the diffusion multigroup equation analytically using a method proposed by the authors. The body of mathematic pertaining to the case is presented. The results obtained, which demonstrate how functional the programs are and how applicable they are for neutron-physical analysis, are presented in tabular form.

  13. Influence of scission neutrons on the prompt fission neutron spectrum calculations

    Science.gov (United States)

    Serot, Olivier; Litaize, Olivier; Chebboubi, Abdelaziz

    2017-09-01

    The calculation of the Prompt Fission Neutron Spectrum (PFNS) was performed using the FIFRELIN Monte Carlo code simulating the de-excitation of the whole fission fragments. This de-excitation is governed by the Hauser-Feshbach statistical model, which has the advantage to take into account the conservation laws for the energy, spin and parity of the initial and final states. In this way, the competition between prompt neutron and prompt gamma emission can be properly accounted for. Assuming that the prompt neutron emission comes only from an evaporation process of the fully accelerated fission fragments, our calculations are not able to reproduce satisfactorily the experimental data. In this context, we have added an additional source of neutrons that may arise during the sudden rupture of the neck (the so-called scission neutrons). Applied in the case of the spontaneous fission of 252Cf, our PFNS calculations show a very good agreement with the Mannhart evaluation by accounting for a 2% scission neutron contribution. Note to the reader: the pdf file has been changed on October 03, 2017.

  14. Exact nuclear data uncertainty propagation for fusion neutronics calculations

    Energy Technology Data Exchange (ETDEWEB)

    Rochman, D., E-mail: rochman@nrg.e [Nuclear Research and Consultancy Group NRG, P.O. Box 25, 1755 ZG Petten (Netherlands); Koning, A.J.; Marck, S.C. van der [Nuclear Research and Consultancy Group NRG, P.O. Box 25, 1755 ZG Petten (Netherlands)

    2010-08-15

    Recently, we have presented an exact method (now called 'Total Monte Carlo') to propagate uncertainties of fundamental nuclear physics experiments, models and parameters to different types of criticality-safety benchmarks. We now show that such exact uncertainty calculations are directly relevant to the optimal and safe design of fusion reactors by applying this methodology to a series of fusion shielding benchmarks, namely those connected to the Oktavian, Fusion Neutronics Source and LLNL Pulsed Sphere experiments. Uncertainties on neutron and gamma leakage fluxes for 13 shielding benchmarks are obtained, in the mass range from {sup nat}Mg to {sup nat}W. Uncertainties for cross-sections, angular distributions, single- and double-differential emission spectra, and gamma-ray production cross-sections are considered in this uncertainty propagation scheme.

  15. Calculation of neutron flux and spectrum in the irradiation test capsule at HANARO

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Seong Woo; Cho, Man Soon; Choo, Kee Nam; Park, Sang Jun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-05-15

    The irradiation test capsules were mostly used for the irradiation test in CT and OR5 irradiation hole. Since the neutron fluence is an important factor, fluence monitor(F/M)s were inserted in the irradiation test capsule in order to measure the neutron fluence of test specimen. Not only the good measurement technique but also the calculation data is necessary to accurately evaluate the neutron fluence of irradiated material. Therefore, following factors should be calculated for detailed evaluation of the neutron fluence; Neutron flux and spectrum with the position of control absorber rod(CAR), Neutron flux and spectrum at the candidate F/M irradiated position, Neutron fluence difference between F/M and specimen From this calculation data, the neutron fluence of irradiated specimen and F/M can be predicted. In this paper, the neutron flux and spectrum were calculated for the irradiation capsule. This data can be a basic data of neutron dosimetry for the irradiation test and applied to select the optimum F/M installation position and verify the neutron fluence of the specimen. The neutron flux and spectrum was calculated for irradiation test capsule. The difference of neutron flux and spectrum of the irradiation test capsule in CT and OR5 irradiation hole was observed. Also the spectral averaged cross section was calculated and applied to the fast neutron fluence evaluation. As a result of this evaluation, the good agreement between calculated and measured data was shown.

  16. Two level calculation of assembly neutronic data libraries; Schema de calcul de bibliotheques a deux niveaux

    Energy Technology Data Exchange (ETDEWEB)

    Benomar, M

    1998-09-01

    The neutronic modeling of a nuclear reactor core requires 2 steps. The first step that is called transport calculation, is an accurate modeling of each type of assemblies put in a simple configuration. APOLLO2, a French neutronic code is used. This step allows the constitution of assembly data libraries. The second step represents the computing of the whole core by the diffusion theory and by using the data libraries defined in the first step. This work is dedicated to the improvement of the first step by allowing both a 172 group energy meshing and a two-dimension spatial processing. (A.C.) 7 refs.

  17. Neutron and photon transport calculations in fusion system. 2

    Energy Technology Data Exchange (ETDEWEB)

    Sato, Satoshi [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment

    1998-03-01

    On the application of MCNP to the neutron and {gamma}-ray transport calculations for fusion reactor system, the wide range design calculation has been carried out in the engineering design activities for the international thermonuclear fusion experimental reactor (ITER) being developed jointly by Japan, USA, EU and Russia. As the objects of shielding calculation for fusion reactors, there are the assessment of dose equivalent rate for living body shielding and the assessment of the nuclear response for the soundness of in-core structures. In the case that the detailed analysis of complicated three-dimensional shapes is required, the assessment using MCNP has been carried out. Also when the nuclear response of peripheral equipment due to the gap streaming between blanket modules is evaluated with good accuracy, the calculation with MCNP has been carried out. The analyses of the shieldings for blanket modules and NBI port are explained, and the examples of the results of analyses are shown. In the blanket modules, there are penetrating holes and continuous gap. In the case of the NBI port, shielding plug cannot be installed. These facts necessitate the MCNP analysis with high accuracy. (K.I.)

  18. Calculation of isodose curves from initial neutron radiation of a hypothetical nuclear explosion using Monte Carlo Method

    Energy Technology Data Exchange (ETDEWEB)

    Medeiros, Marcos P.C.; Rebello, Wilson F.; Andrade, Edson R., E-mail: rebello@ime.eb.br, E-mail: daltongirao@yahoo.com.br [Instituto Militar de Engenharia (IME), Rio de Janeiro, RJ (Brazil). Secao de Engenharia Nuclear; Silva, Ademir X., E-mail: ademir@nuclear.ufrj.br [Corrdenacao dos Programas de Pos-Graduacao em Egenharia (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Programa de Engenharia Nuclear

    2015-07-01

    Nuclear explosions are usually described in terms of its total yield and associated shock wave, thermal radiation and nuclear radiation effects. The nuclear radiation produced in such events has several components, consisting mainly of alpha and beta particles, neutrinos, X-rays, neutrons and gamma rays. For practical purposes, the radiation from a nuclear explosion is divided into {sup i}nitial nuclear radiation{sup ,} referring to what is issued within one minute after the detonation, and 'residual nuclear radiation' covering everything else. The initial nuclear radiation can also be split between 'instantaneous or 'prompt' radiation, which involves neutrons and gamma rays from fission and from interactions between neutrons and nuclei of surrounding materials, and 'delayed' radiation, comprising emissions from the decay of fission products and from interactions of neutrons with nuclei of the air. This work aims at presenting isodose curves calculations at ground level by Monte Carlo simulation, allowing risk assessment and consequences modeling in radiation protection context. The isodose curves are related to neutrons produced by the prompt nuclear radiation from a hypothetical nuclear explosion with a total yield of 20 KT. Neutron fluency and emission spectrum were based on data available in the literature. Doses were calculated in the form of ambient dose equivalent due to neutrons H*(10){sub n}{sup -}. (author)

  19. A method for transient, three-dimensional neutron transport calculations

    Energy Technology Data Exchange (ETDEWEB)

    Waddell, M.W. Jr. (Oak Ridge Y-12 Plant, TN (United States)); Dodds, H.L. (Tennessee Univ., Knoxville, TN (United States))

    1992-12-28

    This paper describes the development and evaluation of a method for solving the time-dependent, three-dimensional Boltzmann transport model with explicit representation of delayed neutrons. A hybrid stochastic/deterministic technique is utilized with a Monte Carlo code embedded inside of a quasi-static kinetics framework. The time-dependent flux amplitude, which is usually fast varying, is computed deterministically by a conventional point kinetics algorithm. The point kinetics parameters, reactivity and generation time as well as the flux shape, which is usually slowly varying in time, are computed stochastically during the random walk of the Monte Carlo calculation. To verify the accuracy of this new method, several computational benchmark problems from the Argonne National Laboratory benchmark book, ANL-7416, were calculated. The results are shown to be in reasonably good agreement with other independently obtained solutions. The results obtained in this work indicate that the method/code is working properly and that it is economically feasible for many practical applications provided a dedicated high performance workstation is available.

  20. A method for transient, three-dimensional neutron transport calculations

    Energy Technology Data Exchange (ETDEWEB)

    Waddell, M.W. Jr. (Martin Marietta Energy Systems, Inc. (United States)); Dodds, H.L. (Univ. of Tennessee (United States))

    1993-04-01

    This paper describes the development and evaluation of a method for solving the time-dependent, three-dimensional Boltzmann transport model with explicit representation of delayed neutrons. A hybrid stochastic/deterministic technique is utilized with a Monte Carlo code embedded inside of a quasi-static kinetics framework. The time-dependent flux amplitude, which is usually fast varying, is computed deterministically by a conventional point kinetics algorithm. The point kinetics parameters, reactivity and generation time as well as the flux shape, which is usually slowly varying in time, are computed stochastically during the random walk of the Monte Carlo calculation. To verify the accuracy of this new method, several computational benchmark problems from the Argonne National Laboratory benchmark book, ANL-7416, were calculated. The results are shown to be in reasonably good agreement with other independently obtained solutions. The results obtained in this work indicate that the method/code is working properly and that it is economically feasible for many practical applications provided a dedicated high performance workstation is available. (orig.)

  1. Thermal neutron scattering law calculations using ab initio molecular dynamics

    Science.gov (United States)

    Wormald, Jonathan; Hawari, Ayman I.

    2017-09-01

    In recent years, methods for the calculation of the thermal scattering law (i.e. S(α,β), where α and β are dimensionless momentum and energy transfer variables, respectively) were developed based on ab initio lattice dynamics (AILD) and/or classical molecular dynamics (CMD). While these methods are now mature and efficient, further advancement in the application of such atomistic techniques is possible using ab initio molecular dynamics (AIMD) methods. In this case, temperature effects are inherently included in the calculation, e.g. phonon density of states (DOS), while using ab initio force fields that eliminate the need for parameterized semi-empirical force fields. In this work, AIMD simulations were performed to predict the phonon spectra as a function of temperature for beryllium and graphite, which are representative nuclear reactor moderator and reflector materials. Subsequently, the calculated phonon spectra were utilized to predict S(α,β) using the LEAPR module of the NJOY code. The AIMD models of beryllium and graphite were 5 × 5 × 5 crystal unit cells (250 atoms and 500 atoms respectively). Electronic structure calculations for the prediction of Hellman-Feynman forces were performed using density functional theory with a GGA exchange correlation functional and corresponding core electron pseudopotentials. AIMD simulations of 1000-10,000 time-steps were performed with the canonical ensemble (NVT thermostat) for several temperatures between 300 K and 900 K. The phonon DOS were calculated as the power spectrum of the AIMD predicted velocity autocorrelation functions. The resulting AIMD phonon DOS and corresponding inelastic thermal neutron scattering cross sections at 300 K, where anharmonic effects are expected to be small, were found to be in reasonable agreement with the results generated using traditional AILD. This illustrated the validity of the AIMD approach. However, since the impact of the temperature on the phonon DOS (e.g. broadening of

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-01-01

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

  3. Tables for simplifying calculations of activities produced by thermal neutrons

    Science.gov (United States)

    Senftle, F.E.; Champion, W.R.

    1954-01-01

    The method of calculation described is useful for the types of work of which examples are given. It is also useful in making rapid comparison of the activities that might be expected from several different elements. For instance, suppose it is desired to know which of the three elements, cobalt, nickel, or vanadium is, under similar conditions, activated to the greatest extent by thermal neutrons. If reference is made to a cross-section table only, the values may be misleading unless properly interpreted by a suitable comparison of half-lives and abundances. In this table all the variables have been combined and the desired information can be obtained directly from the values of A 3??, the activity produced per gram per second of irradiation, under the stated conditions. Hence, it is easily seen that, under similar circumstances of irradiation, vanadium is most easily activated even though the cross section of one of the cobalt isotopes is nearly five times that of vanadium and the cross section of one of the nickel isotopes is three times that of vanadium. ?? 1954 Societa?? Italiana di Fisica.

  4. Neutron and gamma ray transport calculations in shielding system

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-03-01

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

  5. Determination of air/water ratio in pipes by fast neutrons: experiment and Monte Carlo simulation.

    Science.gov (United States)

    AboAlfaraj, Tareq; Abdul-Majid, Samir

    2012-04-01

    Fast neutron dose attenuation from a (252)Cf neutron source is used for the determination of air to water ratio in pipes. Such measurement of the two-phase flow volume fraction is important for many industrial plants such as desalination plants and oil refineries. Fast neutrons penetrate liquid more than slow neutrons or gamma rays. Using diameters from 11.5 cm to 20.76 cm and with wall thicknesses from 0.45 to 1.02 cm, attenuation was independent of pipe wall thicknesses and diameters. Experimental data was in good agreement with values calculated using MCNP codes. The measured neutron flux values decreased with increasing water levels in pipes up to about 14 cm, indicating that our system can be used successfully in desalination plants in pipes of different sizes. The experimental sensitivity was found to be about 0.015 mSv/hcm and the system can be used to measure water level changes down to few millimeters. Use of such a system in fixed positions in the plant can provide information on plant's overall performance and can detect loss of flow immediately before any consequences. A portable system could be designed to measure the air to water ratio in different locations in the plant in a relatively short time. Copyright © 2011 Elsevier Ltd. All rights reserved.

  6. An update on the discrepancy between calculated and measured neutron-induced radioactivity levels in Hiroshima.

    Science.gov (United States)

    Hunter, Nezahat; Charles, Monty W

    2002-12-01

    The thermal neutron activation measurements carried out over many years in Hiroshima and Nagasaki have been the subject of ongoing debate in recent years because they indicate that current DS86 neutron doses may have been significantly underestimated in Hiroshima. Long-lived neutron activation products, 60Co, 152Eu, 154Eu and 36Cl, which are still detectable today using modern analytical techniques, appear to indicate that DS86 calculated thermal neutron activation products decrease with distance more rapidly than the measured values. The latest thermal neutron activation measurements have been collated and a new relationship for the measured to calculated (M/C) ratio of induced activity has been derived as a function of slant range. This indicates a stronger dependence of M/C on slant range than previously derived by Straume et al (1992 Health Phys. 63 421-6) and emphasises even more the discrepancy between measured and calculated (DS86) neutron doses at distances beyond 1 km. While the main body of thermal neutron activation data appears to support a significant increase in the DS86 neutron dose component in Hiroshima, there are some thermal neutron activation measurements and some very recent fast neutron activation measurements which suggest that the discrepancy may not be so great. The extent of the required revision to the neutron component of the DS86 dosimetry remains the subject of ongoing new neutron activation measurements and re-analysis of existing published measurements. A companion paper considers the impact on radiation risk estimates of possible modifications to the DS86 dosimetry system on the basis of a broad range of interpretations of the neutron activation data.

  7. Neutron transport calculation for Activation Evaluation for Decommissioning of PET cyclotron Facility

    Science.gov (United States)

    Nobuhara, Fumiyoshi; Kuroyanagi, Makoto; Masumoto, Kazuyoshi; Nakamura, Hajime; Toyoda, Akihiro; Takahashi, Katsuhiko

    2017-09-01

    In order to evaluate the state of activation in a cyclotron facility used for the radioisotope production of PET diagnostics, we measured the neutron flux by using gold foils and TLDs. Then, the spatial distribution of neutrons and induced activity inside the cyclotron vault were simulated with the Monte Calro calculation code for neutron transport and DCHAIN-SP for activation calculation. The calculated results are in good agreement with measured values within factor 3. Therefore, the adaption of the advanced evaluation procedure for activation level is proved to be important for the planning of decommissioning of these facilities.

  8. Neutron transport calculation for Activation Evaluation for Decommissioning of PET cyclotron Facility

    Directory of Open Access Journals (Sweden)

    Nobuhara Fumiyoshi

    2017-01-01

    Full Text Available In order to evaluate the state of activation in a cyclotron facility used for the radioisotope production of PET diagnostics, we measured the neutron flux by using gold foils and TLDs. Then, the spatial distribution of neutrons and induced activity inside the cyclotron vault were simulated with the Monte Calro calculation code for neutron transport and DCHAIN-SP for activation calculation. The calculated results are in good agreement with measured values within factor 3. Therefore, the adaption of the advanced evaluation procedure for activation level is proved to be important for the planning of decommissioning of these facilities.

  9. Neutronic calculation of fast reactors by the EUCLID/V1 integrated code

    Science.gov (United States)

    Koltashev, D. A.; Stakhanova, A. A.

    2017-01-01

    This article considers neutronic calculation of a fast-neutron lead-cooled reactor BREST-OD-300 by the EUCLID/V1 integrated code. The main goal of development and application of integrated codes is a nuclear power plant safety justification. EUCLID/V1 is integrated code designed for coupled neutronics, thermomechanical and thermohydraulic fast reactor calculations under normal and abnormal operating conditions. EUCLID/V1 code is being developed in the Nuclear Safety Institute of the Russian Academy of Sciences. The integrated code has a modular structure and consists of three main modules: thermohydraulic module HYDRA-IBRAE/LM/V1, thermomechanical module BERKUT and neutronic module DN3D. In addition, the integrated code includes databases with fuel, coolant and structural materials properties. Neutronic module DN3D provides full-scale simulation of neutronic processes in fast reactors. Heat sources distribution, control rods movement, reactivity level changes and other processes can be simulated. Neutron transport equation in multigroup diffusion approximation is solved. This paper contains some calculations implemented as a part of EUCLID/V1 code validation. A fast-neutron lead-cooled reactor BREST-OD-300 transient simulation (fuel assembly floating, decompression of passive feedback system channel) and cross-validation with MCU-FR code results are presented in this paper. The calculations demonstrate EUCLID/V1 code application for BREST-OD-300 simulating and safety justification.

  10. Development of Library Processing System for Neutron Transport Calculation

    Energy Technology Data Exchange (ETDEWEB)

    Song, J. S.; Park, S. Y.; Kim, H. Y. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)] (and others)

    2008-12-15

    A system for library generation was developed for the lattice neutron transport program for pressurized water reactor core analysis. The system extracts multi energy group nuclear data for requested nuclides from ENDF/B whose data are based on continuous energy, generates hydrogen equivalent factor and resonance integral table as functions of temperature and background cross section for resonance nuclides, generates subgroup data for the lattice program to treat resonance exactly as possible, and generates multi-group neutron library file including nuclide depletion data for use of the lattice program.

  11. Neutron Deep Penetration Calculations in Light Water with Monte Carlo TRIPOLI-4® Variance Reduction Techniques

    Directory of Open Access Journals (Sweden)

    Lee Yi-Kang

    2017-01-01

    Full Text Available Nuclear decommissioning takes place in several stages due to the radioactivity in the reactor structure materials. A good estimation of the neutron activation products distributed in the reactor structure materials impacts obviously on the decommissioning planning and the low-level radioactive waste management. Continuous energy Monte-Carlo radiation transport code TRIPOLI-4 has been applied on radiation protection and shielding analyses. To enhance the TRIPOLI-4 application in nuclear decommissioning activities, both experimental and computational benchmarks are being performed. To calculate the neutron activation of the shielding and structure materials of nuclear facilities, the knowledge of 3D neutron flux map and energy spectra must be first investigated. To perform this type of neutron deep penetration calculations with the Monte Carlo transport code, variance reduction techniques are necessary in order to reduce the uncertainty of the neutron activation estimation. In this study, variance reduction options of the TRIPOLI-4 code were used on the NAIADE 1 light water shielding benchmark. This benchmark document is available from the OECD/NEA SINBAD shielding benchmark database. From this benchmark database, a simplified NAIADE 1 water shielding model was first proposed in this work in order to make the code validation easier. Determination of the fission neutron transport was performed in light water for penetration up to 50 cm for fast neutrons and up to about 180 cm for thermal neutrons. Measurement and calculation results were benchmarked. Variance reduction options and their performance were discussed and compared.

  12. Neutron Deep Penetration Calculations in Light Water with Monte Carlo TRIPOLI-4® Variance Reduction Techniques

    Science.gov (United States)

    Lee, Yi-Kang

    2017-09-01

    Nuclear decommissioning takes place in several stages due to the radioactivity in the reactor structure materials. A good estimation of the neutron activation products distributed in the reactor structure materials impacts obviously on the decommissioning planning and the low-level radioactive waste management. Continuous energy Monte-Carlo radiation transport code TRIPOLI-4 has been applied on radiation protection and shielding analyses. To enhance the TRIPOLI-4 application in nuclear decommissioning activities, both experimental and computational benchmarks are being performed. To calculate the neutron activation of the shielding and structure materials of nuclear facilities, the knowledge of 3D neutron flux map and energy spectra must be first investigated. To perform this type of neutron deep penetration calculations with the Monte Carlo transport code, variance reduction techniques are necessary in order to reduce the uncertainty of the neutron activation estimation. In this study, variance reduction options of the TRIPOLI-4 code were used on the NAIADE 1 light water shielding benchmark. This benchmark document is available from the OECD/NEA SINBAD shielding benchmark database. From this benchmark database, a simplified NAIADE 1 water shielding model was first proposed in this work in order to make the code validation easier. Determination of the fission neutron transport was performed in light water for penetration up to 50 cm for fast neutrons and up to about 180 cm for thermal neutrons. Measurement and calculation results were benchmarked. Variance reduction options and their performance were discussed and compared.

  13. Dose measurements and calculations in the epithermal neutron beam at the Brookhaven Medical Research Reactor (BMRR)

    Energy Technology Data Exchange (ETDEWEB)

    Fairchild, R.G.; Greenberg, D.; Kamen, Y.; Fiarman, S. (Brookhaven National Lab., Upton, NY (USA). Medical Dept.); Benary, V. (Brookhaven National Lab., Upton, NY (USA). Medical Dept. Tel Aviv Univ. (Israel)); Kalef-Ezra, J. (Brookhaven National Lab., Upton, NY (USA). Medical Dept. Ioannina Univ. (Greece)); Wielopolski, L. (Brookhaven National Lab., Upton, NY (USA). Medical Dept. State Univ. of New

    1990-01-01

    The characteristics of the epithermal neutron beam at BMRR were measured, calculated, and reported. This beam has already been used for animal irradiations. We anticipate that it will be used for clinical trials. Thermal and epithermal neutron flux densities distributions, and dose rate distributions, as a function of depth were measured in a lucite dog-head phantom. Monte Carlo calculations were performed and compared with the measured values. 2 refs., 4 figs., 1 tab.

  14. Shielding calculation constants for use in effective dose evaluation for photons, neutrons and Bremsstrahlung from beta-ray

    Energy Technology Data Exchange (ETDEWEB)

    Sakamoto, Yukio; Endo, Akira; Tsuda, Shuichi; Takahashi, Fumiaki; Yamaguchi, Yasuhiro [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2001-01-01

    Dose quantity in the shielding design calculation will be changed from the 1 cm depth dose equivalent to effective dose on the occasion of the introduction of International Commission on Radiological Protection (ICRP) 1990 Recommendations (ICRP Publication 60) into domestic laws. In the shielding calculation for the radiation facilities, simple dose estimation methods by using the shielding calculation constants instead of calculation of radiation energy spectra behind the shielding materials are effective and widely used. These shielding calculation constants depend on the dose quantity to be estimated and those for the evaluation of 1 cm depth dose equivalents should be replaced by those for the evaluation of effective dose. In the present report, the shielding calculation constants are summarized for photons, neutrons and Bremsstrahlung from beta-ray. For mono-energetic photons with energies from 0.015 MeV to 10 MeV, effective dose buildup factors, effective conversion coefficients from air kerma to effective dose and transmission data of effective dose were calculated. Effective dose rate constants, which represent an effective dose value at 1 m apart from a source without shielding, and transmission data of effective dose were also calculated for gamma-ray and X-ray from 33 radioisotopes, Bremsstrahlung from 13 radioisotopes beta-ray and 4 neutron sources. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Byung Cheol; Chang, Ki Oak

    1997-05-01

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

  16. Comparison of the neutron ambient dose equivalent and ambient absorbed dose calculations with different GEANT4 physics lists

    Science.gov (United States)

    Ribeiro, Rosane Moreira; Souza-Santos, Denison

    2017-10-01

    A comparison between neutron physics lists given by GEANT4, is made in the calculation of the ambient dose equivalent, and ambient absorbed dose, per fluence conversion coefficients (H* (10) / ϕ and D* (10) / ϕ) for neutrons in the range of 10-9 MeV to 15 MeV. Physics processes are included for neutrons, photons and charged particles, and calculations are made for neutrons and secondary particles. Results obtained for QBBC, QGSP_BERT, QGSP_BIC and Neutron High Precision physics lists are compared with values published in ICRP 74 and previously published articles. Neutron high precision physics lists showed the best results in the studied energy range.

  17. Delayed neutron spectra and their uncertainties in fission product summation calculations

    Energy Technology Data Exchange (ETDEWEB)

    Miyazono, T.; Sagisaka, M.; Ohta, H.; Oyamatsu, K.; Tamaki, M. [Nagoya Univ. (Japan)

    1997-03-01

    Uncertainties in delayed neutron summation calculations are evaluated with ENDF/B-VI for 50 fissioning systems. As the first step, uncertainty calculations are performed for the aggregate delayed neutron activity with the same approximate method as proposed previously for the decay heat uncertainty analyses. Typical uncertainty values are about 6-14% for {sup 238}U(F) and about 13-23% for {sup 243}Am(F) at cooling times 0.1-100 (s). These values are typically 2-3 times larger than those in decay heat at the same cooling times. For aggregate delayed neutron spectra, the uncertainties would be larger than those for the delayed neutron activity because much more information about the nuclear structure is still necessary. (author)

  18. Neutron dosimetry and damage calculations for the HFIR-JP-23 irradiations

    Energy Technology Data Exchange (ETDEWEB)

    Greenwood, L.R.; Ratner, R.T. [Pacific Northwest National Lab., Richland, WA (United States)

    1996-10-01

    Neutron fluence measurements and radiation damage calculations are reported for the joint US-Japanese experiment JP-23, which was conducted in target position G6 of the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL). The maximum neutron fluence at midplane was 4.4E+22 n/cm{sup 2} resulting in about 9.0 dpa in type 316 stainless steel.

  19. Neutron dosimetry and damage calculations for the ATR-A1 irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Greenwood, L.R.; Ratner, R.T. [Pacific Northwest National Lab., Richland, WA (United States)

    1998-09-01

    Neutron fluence measurements and radiation damage calculations are reported for the collaborative US/Japan ATR-A1 irradiation in the Advanced Test Reactor (ATR) at Idaho National Engineering Laboratory (INEL). The maximum total neutron fluence at midplane was 9.4 {times} 10{sup 21} n/cm{sup 2} (5.5 {times} 10{sup 21} n/cm{sup 2} above 0.1 MeV), resulting in about 4.6 dpa in vanadium.

  20. The shell structure effects in neutron cross section calculation by a ...

    African Journals Online (AJOL)

    The role of the shell structure properties of the nucleus in the calculation of neutron-induced reaction cross-section data based on nuclear reaction theory has been investigated. In this investigation, measured, evaluated and calculated (n.p) reaction cross-section data on la spherical nucleus (i.e. 112Sn) and a deformed ...

  1. Ab initio calculation of the neutron-proton mass difference

    CERN Multimedia

    CERN. Geneva

    2015-01-01

    The existence and stability of atoms relies on the fact that neutrons are more massive than protons. The mass difference is only 0.14% of the average and has significant astrophysical and cosmological implications. A slightly smaller or larger value would have led to a dramatically different universe. After an introduction to the problem and to lattice quantum chromodynamics (QCD), I will show how this difference can be computed precisely by carefully accounting for electromagnetic and mass isospin breaking effects in lattice computations. I will also report on results for splittings in the \\Sigma, \\Xi, D and \\Xi_{cc} isospin multiplets, some of which are predictions. The computations are performed in lattice QCD plus QED with four, non-degenerate quark flavors.

  2. Uncertainty in the delayed neutron fraction in fuel assembly depletion calculations

    Directory of Open Access Journals (Sweden)

    Aures Alexander

    2017-01-01

    Full Text Available This study presents uncertainty and sensitivity analyses of the delayed neutron fraction of light water reactor and sodium-cooled fast reactor fuel assemblies. For these analyses, the sampling-based XSUSA methodology is used to propagate cross section uncertainties in neutron transport and depletion calculations. Cross section data is varied according to the SCALE 6.1 covariance library. Since this library includes nu-bar uncertainties only for the total values, it has been supplemented by delayed nu-bar uncertainties from the covariance data of the JENDL-4.0 nuclear data library. The neutron transport and depletion calculations are performed with the TRITON/NEWT sequence of the SCALE 6.1 package. The evolution of the delayed neutron fraction uncertainty over burn-up is analysed without and with the consideration of delayed nu-bar uncertainties. Moreover, the main contributors to the result uncertainty are determined. In all cases, the delayed nu-bar uncertainties increase the delayed neutron fraction uncertainty. Depending on the fuel composition, the delayed nu-bar values of uranium and plutonium in fact give the main contributions to the delayed neutron fraction uncertainty for the LWR fuel assemblies. For the SFR case, the uncertainty of the scattering cross section of U-238 is the main contributor.

  3. Air ingression calculations for selected plant transients using MELCOR

    Energy Technology Data Exchange (ETDEWEB)

    Kmetyk, L.N.

    1994-01-01

    Two sets of MELCOR calculations have been completed studying the effects of air ingression on the consequences of various severe accident scenarios. One set of calculations analyzed a station blackout with surge line failure prior to vessel breach, starting from nominal operating conditions; the other set of calculations analyzed a station blackout occurring during shutdown (refueling) conditions. Both sets of analyses were for the Surry plant, a three-loop Westinghouse PWR. For both accident scenarios, a basecase calculation was done, and then repeated with air ingression from containment into the core region following core degradation and vessel failure. In addition to the two sets of analyses done for this program, a similar air-ingression sensitivity study was done as part of a low-power/shutdown PRA, with results summarized here; that PRA study also analyzed a station blackout occurring during shutdown (refueling) conditions, but for the Grand Gulf plant, a BWR/6 with Mark III containment. These studies help quantify the amount of air that would have to enter the core region to have a significant impact on the severe accident scenario, and demonstrate that one effect, of air ingression is substantial enhancement of ruthenium release. These calculations also show that, while the core clad temperatures rise more quickly due to oxidation with air rather than steam, the core also degrades and relocates more quickly, so that no sustained, enhanced core heatup is predicted to occur with air ingression.

  4. Calculated neutron-source spectra from selected irradiated PWR fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Rinard, P.M.; Bosler, G.E.; Phillips, J.R.

    1981-12-01

    The energy spectra of neutrons emitted from a pressurized-water-reactor fuel assembly have been calculated for a variety of exposures and cooling times. They are presented in graphical form. Some effects of initial enrichment are also included. Neutrons from spontaneous fissions were given either a Maxwellian temperature of 1.2 or 1.5 MeV, depending on whether they were due to plutonium and uranium nuclides or curium nuclides. A single (..cap alpha..,n) spectrum was deemed sufficient to represent the neutrons from all the alpha-emitting nuclides. The proportions of the nuclides undergoing spontaneous fission and those emitting alpha particles were determined from calculated atom densities. The particular pressurized-water-reactor fuel assembly assumed for this purpose was of the type used in the H.B. Robinson Unit-2 power plant (740 MWe).

  5. Neutron and gamma ray calculation for Hiroshima-type atomic bomb

    Energy Technology Data Exchange (ETDEWEB)

    Hoshi, Masaharu; Endo, Satoru; Takada, Jun [Hiroshima Univ. (Japan). Research Inst. for Radiation Biology and Medicine; Iwatani, Kazuo; Oka, Takamitsu; Shizuma, Kiyoshi; Fujita, Shoichiro; Hasai, Hiromi

    1998-03-01

    We looked at the radiation dose of Hiroshima and Nagasaki atomic bomb again in 1986. We gave it the name of ``Dosimetry System 1986`` (DS86). We and other groups have measured the expose dose since 1986. Now, the difference between data of {sup 152}Eu and the calculation result on the basis of DS86 was found. To investigate the reason, we carried out the calculations of neutron transport and neutron absorption gamma ray for Hiroshima atomic bomb by MCNP3A and MCNP4A code. The problems caused by fast neutron {sup 32}P from sulfur in insulator of pole. To correct the difference, we investigated many models and found agreement of all data within 1 km. (S.Y.)

  6. Talys calculations for evaluation of neutron-induced single-event upset cross sections

    Energy Technology Data Exchange (ETDEWEB)

    Bourselier, Jean-Christophe

    2005-08-15

    The computer code TALYS has been used to calculate interactions between cosmic-ray neutrons and silicon nuclei with the goal to describe single-event upset (SEU) cross sections in microelectronics devices. Calculations for the Si(n,X) reaction extend over an energy range of 2 to 200 MeV. The obtained energy spectra of the resulting residuals and light-ions have been integrated using several different critical charges as SEU threshold. It is found that the SEU cross section seems largely to be dominated by {sup 28}Si recoils from elastic scattering. Furthermore, the shape of the SEU cross section as a function of the energy of the incoming neutron changes drastically with decreasing critical charge. The results presented in this report stress the importance of performing studies at mono-energetic neutron beams to advance the understanding of the underlying mechanisms causing SEUs.

  7. A new method for calculation of an air quality index

    Energy Technology Data Exchange (ETDEWEB)

    Ilvessalo, P. [Finnish Meteorological Inst., Helsinki (Finland). Air Quality Dept.

    1995-12-31

    Air quality measurement programs in Finnish towns have expanded during the last few years. As a result of this it is more and more difficult to make use of all the measured concentration data. Citizens of Finnish towns are nowadays taking more of an interest in the air quality of their surroundings. The need to describe air quality in a simplified form has increased. Air quality indices permit the presentation of air quality data in such a way that prevailing conditions are more easily understandable than when using concentration data as such. Using an air quality index always means that some of the information about concentrations of contaminants in the air will be lost. How much information is possible to extract from a single index number depends on the calculation method. A new method for the calculation of an air quality index has been developed. This index always indicates the overstepping of an air quality guideline level. The calculation of this air quality index is performed using the concentrations of all the contaminants measured. The index gives information both about the prevailing air quality and also the short-term trend. It can also warn about the expected exceeding of guidelines due to one or several contaminants. The new index is especially suitable for the real-time monitoring and notification of air quality values. The behaviour of the index was studied using material from a measurement period in the spring of 1994 in Kaepylae, Helsinki. Material from a pre-operational period in the town of Oulu was also available. (author)

  8. Calculation of neutron fluence to dose equivalent conversion coefficients using GEANT4; Calculo de coeficientes de fluencia de neutrons para equivalente de dose individual utilizando o GEANT4

    Energy Technology Data Exchange (ETDEWEB)

    Ribeiro, Rosane M.; Santos, Denison de S.; Queiroz Filho, Pedro P. de; Mauricio, CLaudia L.P.; Silva, Livia K. da; Pessanha, Paula R., E-mail: rosanemribeiro@oi.com.br [Instituto de Radioprotecao e Dosimetria (IRD/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2014-07-01

    Fluence to dose equivalent conversion coefficients provide the basis for the calculation of area and personal monitors. Recently, the ICRP has started a revision of these coefficients, including new Monte Carlo codes for benchmarking. So far, little information is available about neutron transport below 10 MeV in tissue-equivalent (TE) material performed with Monte Carlo GEANT4 code. The objective of this work is to calculate neutron fluence to personal dose equivalent conversion coefficients, H{sub p} (10)/Φ, with GEANT4 code. The incidence of monoenergetic neutrons was simulated as an expanded and aligned field, with energies ranging between thermal neutrons to 10 MeV on the ICRU slab of dimension 30 x 30 x 15 cm{sup 3}, composed of 76.2% of oxygen, 10.1% of hydrogen, 11.1% of carbon and 2.6% of nitrogen. For all incident energy, a cylindrical sensitive volume is placed at a depth of 10 mm, in the largest surface of the slab (30 x 30 cm{sup 2}). Physic process are included for neutrons, photons and charged particles, and calculations are made for neutrons and secondary particles which reach the sensitive volume. Results obtained are thus compared with values published in ICRP 74. Neutron fluence in the sensitive volume was calculated for benchmarking. The Monte Carlo GEANT4 code was found to be appropriate to calculate neutron doses at energies below 10 MeV correctly. (author)

  9. Calculation of beams of positrons, neutrons and protons associated with terrestrial gamma-ray flashes

    NARCIS (Netherlands)

    C. Köhn (Christoph); U. Ebert (Ute)

    2015-01-01

    htmlabstractPositron beams have been observed by the Fermi satellite to be correlated with lightning leaders, and neutron emissions have been attributed to lightning and to laboratory sparks as well. Here we discuss the cross sections to be used for modeling these emissions, and we calculate the

  10. Easy-to-use application programs for decay heat and delayed neutron calculations on personal computers

    Energy Technology Data Exchange (ETDEWEB)

    Oyamatsu, Kazuhiro [Nagoya Univ. (Japan)

    1998-03-01

    Application programs for personal computers are developed to calculate the decay heat power and delayed neutron activity from fission products. The main programs can be used in any computers from personal computers to main frames because their sources are written in Fortran. These programs have user friendly interfaces to be used easily not only for research activities but also for educational purposes. (author)

  11. Model-Independent Calculation of Radiative Neutron Capture on Lithium-7

    NARCIS (Netherlands)

    Rupak, Gautam; Higa, Renato

    2011-01-01

    The radiative neutron capture on lithium-7 is calculated model independently using a low-energy halo effective field theory. The cross section is expressed in terms of scattering parameters directly related to the S-matrix elements. It depends on the poorly known p-wave effective range parameter

  12. Implementation of variance-reduction techniques for Monte Carlo nuclear logging calculations with neutron sources

    NARCIS (Netherlands)

    Maucec, M

    2005-01-01

    Monte Carlo simulations for nuclear logging applications are considered to be highly demanding transport problems. In this paper, the implementation of weight-window variance reduction schemes in a 'manual' fashion to improve the efficiency of calculations for a neutron logging tool is presented.

  13. Neutron total cross section calculation within the framework of quasi-harmonic approximation

    Science.gov (United States)

    Cai, Xiao-Xiao; Klinkby, Esben

    2017-10-01

    The accuracy of neutron scattering cross sections is the gauge for the realistic outcome of a neutron transport simulation. To improve the traditional harmonic physics model used in such simulations, we revisit the slow neutron transport theory in crystalline materials and aim to develop a unified model that has good performance for neutron transport problems in crystals in a wide range of temperatures and pressures. The quasi-harmonic approximation (QHA) correlates phonon evolution explicitly with unit cell volume. Therefore, it is capable of evaluating a variety of material properties at finite temperatures. In this work, we show numerically that it is a very effective tool for our application as well. Within the framework of QHA, we calculate the temperature dependent characteristics of phonons in three elemental crystals, namely Be, Mg and Al. Based on the obtained results, our calculated neutron total cross sections agree closely with experimental transmission cross sections in a large temperature range below the melting point. We show that as the harmonic cross section model ignores the effects of phonon softening in these crystals, it underestimates the total inelastic cross sections at high temperatures. In the case of Al, we observe that such underestimation is up to 7% at room temperature. In addition, we study the phonon–phonon scatterings in Al. We observe that the cross section is insensitive to the finite phonon lifetimes even at 800 K.

  14. Verification of the DUCT-III for calculation of high energy neutron streaming

    CERN Document Server

    Masukawa, F; Hayashi, K; Hirayama, H; Nakano, H; Nakashima, H; Sasamoto, N; Shin, K; Tayama, R I

    2003-01-01

    A large number of radiation streaming calculations under a variety of conditions are required as a part of shielding design for a high energy proton accelerator facility. Since sophisticated methods are very time consuming, simplified methods are employed in many cases. For accuracy evaluation of a simplified code DUCT-III for high energy neutron streaming calculations, two kinds of benchmark problems based on the experiments were analyzed. Through comparison of the DUCT-III calculations with both the measurements and the sophisticated Monte Carlo calculations, DUCT-III was seen reliable enough for applying to the shielding design for the Intense Proton Accelerator Facility.

  15. Monte Carlo calculation of the neutron dose to a fetus at commercial flight altitudes

    Science.gov (United States)

    Alves, M. C.; Galeano, D. C.; Santos, W. S.; Hunt, John G.; d'Errico, Francesco; Souza, S. O.; de Carvalho Júnior, A. B.

    2017-11-01

    Aircrew members are exposed to primary cosmic rays as well as to secondary radiations from the interaction of cosmic rays with the atmosphere and with the aircraft. The radiation field at flight altitudes comprises neutrons, protons, electrons, positrons, photons, muons and pions. Generally, 50% of the effective dose to airplane passengers is due to neutrons. Care must be taken especially with pregnant aircrew members and frequent fliers so that the equivalent dose to the fetus will not exceed prescribed limits during pregnancy (1 mSv according to ICRP, and 5 mSv according to NCRP). Therefore, it is necessary to evaluate the equivalent dose to a fetus in the maternal womb. Up to now, the equivalent dose rate to a fetus at commercial flight altitudes was obtained using stylized pregnant-female phantom models. The aim of this study was calculating neutron fluence to dose conversion coefficients for a fetus of six months of gestation age using a new, realistic pregnant-female mesh-phantom. The equivalent dose rate to a fetus during an intercontinental flight was also calculated by folding our conversion coefficients with published spectral neutron flux data. The calculated equivalent dose rate to the fetus was 2.35 μSv.h-1, that is 1.5 times higher than equivalent dose rates reported in the literature. The neutron fluence to dose conversion coefficients for the fetus calculated in this study were 2.7, 3.1 and 3.9 times higher than those from previous studies using fetus models of 3, 6 and 9 months of gestation age, respectively. The differences between our study and data from the literature highlight the importance of using more realistic anthropomorphic phantoms to estimate doses to a fetus in pregnant aircrew members.

  16. FLUKA Calculation of the Neutron Albedo Encountered at Low Earth Orbits

    CERN Document Server

    Claret, Arnaud; Combier, Natacha; Ferrari, Alfredo; Laurent, Philippe

    2014-01-01

    This paper presents Monte-Carlo simulations based on the Fluka code aiming to calculate the contribution of the neutron albedo at a given date and altitude above the Earth chosen by the user. The main input parameters of our model are the solar modulation affecting the spectra of cosmic rays, and the date of the Earth’s geomagnetic fi eld. The results consist in a two-parameter distribution, the neutron energy and the angle to the tangent plane of the sphere containing the orbi t of interest, and are provided by geographical position above the E arth at the chosen altitude. This model can be used to predict the te mporal variation of the neutron fl ux encountered along the orbit, and thus constrain the determination of the instrumental backg round noise of space experiments in low earth orbit.

  17. Neutron dosimetry and damage calculations for the TRIGA MARK-II reactor in Vienna

    Science.gov (United States)

    Weber, H. W.; Böck, H.; Unfried, E.; Greenwood, L. R.

    1986-02-01

    In order to improve the source characterization of the reactor, especially for recent irradiation experiments in the central irradiation thimble, neutron activation experiments were made on 16 nuclides and the neutron flux spectrum was adjusted using the computer code STAY'SL. The results for the total, thermal and fast neutron flux density at a reactor power of 250 kW are as follows: 2.1 × 10 17, 6.1 × 10 16 ( E 0.1 MeV) and 4.0 × 10 16 ( E > 1 MeV) m -2 s -1. respectively. Calculated damage energy cross sections and gas production rates are presented for selected elements.

  18. Nuclear structure of 76Ge from proton-neutron interacting boson model calculations

    Science.gov (United States)

    Zhang, DaLi; Mu, ChengFu

    2018-01-01

    The properties of low-lying states in 76Ge, especially the characteristics of the mixed-symmetry states, have been investigated within the neutron-proton interacting boson model (IBM-2). By considering the relative energy of d proton boson to be different from that of neutron boson, the low-lying positive parity levels and M1, E2 transition strengths have been calculated. The IBM-2 calculated results are in good agreement with the experimental data. Particularly, the mixed-symmetry states have been reproduced quite well. The calculation and systematic analysis demonstrated that the collective character of 76Ge lies closest to the SU*πv(3), with some possible Oπv(6) dynamic symmetry in IBM-2 viewpoint.

  19. CALCULATION OF AIR ION REGIME IN THE CASE OF ARTIFICIAL AIR IONIZATION

    Directory of Open Access Journals (Sweden)

    BILIAIEV M. M.

    2015-10-01

    Full Text Available Purpose. One of the major tasks in the field of labor protection is providing of the necessary qualitative composition of air in the working areas of office and industrial spaces. In order to maintain the necessary air ion level in the air space premises, the artificial ionization of air is used often in the premises. At present in Ukraine analytical model are used for the calculation of air ion regime in premises, influencing on the formation process of air ions concentration field. An alternative solution is the use of CFD models, developing including the air jets aerodynamics in the premise, the presence of furniture, equipment, transfer of ions under an electric field, and other physical factors, determining intensity and shape of air ions concentration field in the premise. Methodology. Influence of air flow was taken into account in the development of CFD models for calculation of air ion regime in the apartment, caused by operation of ventilation, diffusion, electric field impact, as well as the interaction of different polarity ions with each other, and their interaction with dust particles. The proposed model of calculation of air ion regime in premises based on the use of aerodynamics, electrostatics and mass transfer levels. This model allows operatively to calculate air ions concentration field with the influence of the walls, floor, ceiling and obstacles on the process of air ions dispersion, the specific location of different polarity ions emission and their interaction in the premise and work areas in conditions of artificial air ionization. Results. The calculated data were obtained and on their base could be estimated the concentration of air ion anywhere in the premise with artificial air ionization. Ions concentration field, being calculated using this CFD model, as concentration field isolines is presented. Originality. The results of the air ion regime calculation in the premise are presented, based on numerical 2D CFD model

  20. Model-independent calculation of radiative neutron capture on lithium-7.

    Science.gov (United States)

    Rupak, Gautam; Higa, Renato

    2011-06-03

    The radiative neutron capture on lithium-7 is calculated model independently using a low-energy halo effective field theory. The cross section is expressed in terms of scattering parameters directly related to the S-matrix elements. It depends on the poorly known p-wave effective range parameter r(1). This constitutes the largest uncertainty in traditional model calculations. It is explicitly demonstrated by comparing with potential model calculations. A single parameter fit describes the low-energy data extremely well and yields r(1)≈-1.47  fm(-1).

  1. Fuel management and core design code systems for pressurized water reactor neutronic calculations

    Energy Technology Data Exchange (ETDEWEB)

    Ahnert, C.; Arayones, J.M.

    1985-06-01

    A package of connected code systems for the neutronic calculations relevant in fuel management and core design has been developed and applied for validation to the startup tests and first operating cycle of a 900MW (electric) PWR. The package includes the MARIA code system for the modeling of the different types of PWR fuel assemblies, the CARMEN code system for detailed few group diffusion calculations for PWR cores at operating and burnup conditions, and the LOLA code system for core simulation using onegroup nodal theory parameters explicitly calculated from the detailed solutions.

  2. Calculation of fast neutron flux in reactor pressure tubes and experimental facilities

    Energy Technology Data Exchange (ETDEWEB)

    Barnett, P.C. [Canadian General Electric (Canada)

    1968-07-15

    The computer program EPITHET was used to calculate the fast neutron flux (>1 MeV) in several reactor pressure tubes and experimental facilities in order to compare the fast neutron flux in the different cases and to provide a self-consistent set of flux values which may be used to relate creep strain to fast neutron flux . The facilities considered are shown below together with the calculated fast neutron flux (>1 MeV). Fast flux 10{sup 13} n/cm{sup 2}s: NPD 1.14, Douglas Point 2.66, Pickering 2.89, Gentilly 2.35, SGHWR 3.65, NRU U-1 and U-2 3.25'' pressure tube - 19 element fuel 3.05, NRU U-1 and U-2 4.07'' pressure tube - 28 element fuel 3.18, NRU U-1 and U-2 4.07'' pressure tube - 18 element fuel 2.90, NRX X-5 0.88, PRTR Mk I fuel 2.81, PRTR HPD fuel 3.52, WR-1 2.73, Mk IV creep machine (NRX) 0.85, Mk VI creep machine (NRU) 2.04, Biaxial creep insert (NRU U-49) 2.61.

  3. Neutron polarization evolution calculations along the SNS magnetism reflectometer beam line

    Energy Technology Data Exchange (ETDEWEB)

    Parizzi, Andre de [Oak Ridge National Laboratory, Spallation Neutron Source (SNS), Building 8600, Bethel Valley Road, Oak Ridge, TN 37831 (United States)]. E-mail: parizziad@ornl.gov; Klose, Frank [Oak Ridge National Laboratory, Spallation Neutron Source (SNS), Building 8600, Bethel Valley Road, Oak Ridge, TN 37831 (United States); Christoph, Volker [University of Applied Sciences (HTW), Dresden (Germany)

    2005-02-15

    In polarized neutron scattering instruments, most polarization devices apply magnetic fields of different space and time profiles for achieving the desired conditioning of the beam. Magnetic fields created at each device impose fringe/stray fields onto other devices in the beam line, which may affect their functionalities as well as the evolution of the neutron polarization. For the SNS magnetism reflectometer, it is desirable that different sample environment magnets and beam conditioning devices can be used in variable experimental conditions. Spin polarizers and analyzers, broad-band spin flippers and other polarized neutron devices must be capable of working reliably in the vicinity of small magnetic fields generated by an iron-yoke electromagnet and of much larger magnetic fields created, for example, by a high-field superconducting magnet. The latter may not only impose relatively large stray fields along the beam path, but also produce relatively large field gradients. In this paper, we present calculations treating the magnetic field interference between devices, the effect of sample environment magnets and the resultant neutron polarization evolution along the beam line. Calculations are presented for polarized instrumentation configurations that will typically be applied in standard experimental conditions at the SNS magnetism reflectometer.

  4. Boron Neutron Capture Therapy (BNCT) Dose Calculation using Geometrical Factors Spherical Interface for Glioblastoma Multiforme

    Science.gov (United States)

    Zasneda, Sabriani; Widita, Rena

    2010-06-01

    Boron Neutron Capture Therapy (BNCT) is a cancer therapy by utilizing thermal neutron to produce alpha particles and lithium nuclei. The superiority of BNCT is that the radiation effects could be limited only for the tumor cells. BNCT radiation dose depends on the distribution of boron in the tumor. Absorbed dose to the cells from the reaction 10B (n, α) 7Li was calculated near interface medium containing boron and boron-free region. The method considers the contribution of the alpha particle and recoiled lithium particle to the absorbed dose and the variation of Linear Energy Transfer (LET) charged particles energy. Geometrical factor data of boron distribution for the spherical surface is used to calculate the energy absorbed in the tumor cells, brain and scalp for case Glioblastoma Multiforme. The result shows that the optimal dose in tumor is obtained for boron concentrations of 22.1 mg 10B/g blood.

  5. Calculation principles of humid air in a reversed Brayton cycle

    Energy Technology Data Exchange (ETDEWEB)

    Backman, J. [Lappeenranta Univ. of Technology (Finland). Dept. of Energy Technology

    1997-12-31

    The article presents a calculation method for reversed Brayton cycle that uses humid air as working medium. The reversed Brayton cycle can be employed as an air dryer, a heat pump or a refrigerating machine. In this research the use of humid air as a working fluid has an environmental advantage, as well. In this method especially the expansion process in the turbine is important because of the condensation of the water vapour in the humid air. This physical phenomena can have significant effects on the level of performance of the application. The expansion process differs physically from the compression process, when the water vapour in the humid air begins to condensate. In the thermodynamic equilibrium of the flow, the water vapour pressure in humid air cannot exceed the pressure of saturated water vapour in corresponding temperature. Expansion calculation during operation around the saturation zone is based on a quasistatic expansion, in which the system after the turbine is in thermodynamical equilibrium. The state parameters are at every moment defined by the equation of state, and there is no supercooling in the vapour. Following simplifications are used in the calculations: The system is assumed to be adiabatic. This means that there is no heat transfer to the surroundings. This is a common practice, when the temperature differences are moderate as here; The power of the cooling is omitted. The cooling construction is very dependent on the machine and the distribution of the losses; The flow is assumed to be one-dimensional, steady-state and homogenous. The water vapour condensing in the turbine can cause errors, but the errors are mainly included in the efficiency calculation. (author) 11 refs.

  6. Development of SCINFUL-CG code to calculate response functions of scintillators in various shapes used for neutron measurement

    Energy Technology Data Exchange (ETDEWEB)

    Endo, Akira; Kim, Eunjoo; Yamaguchi, Yasuhiro [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2001-10-01

    A Monte Carlo code SCINFUL has been utilized for calculating response functions of organic scintillators for high-energy neutron spectroscopy. However, the applicability of SCINFUL is limited to the calculations for cylindrical NE213 and NE110 scintillators. In the present study, SCINFUL-CG was developed by introducing a geometry specifying function and high-energy neutron cross section data into SCINFUL. The geometry package MARS-CG, the extended version of the CG (Combinatorial Geometry), was programmed into SCINFUL-CG to express various geometries of detectors. Neutron spectra in the regions specified by the CG can be evaluated by the track length estimator. The cross section data of silicon, oxygen and aluminum for neutron transport calculation were incorporated up to 100 MeV using the data of LA150 library. Validity of SCINFUL-CG was examined by comparing calculated results with those by SCINFUL and MCNP and experimental data measured using high-energy neutron fields. SCINFUL-CG can be used for the calculations of the response functions and neutron spectra in the organic scintillators in various shapes. The computer code will be applicable to the designs of high-energy neutron spectrometers and neutron monitors using the organic scintillators. The present report describes the new features of SCINFUL-CG and explains how to use the code. (author)

  7. Calculations of air cooler for new subsonic wind tunnel

    Science.gov (United States)

    Rtishcheva, A. S.

    2017-10-01

    As part of the component development of TsAGI’s new subsonic wind tunnel where the air flow velocity in the closed test section is up to 160 m/sec hydraulic and thermal characteristics of air cooler are calculated. The air cooler is one of the most important components due to its highest hydraulic resistance in the whole wind tunnel design. It is important to minimize its hydraulic resistance to ensure the energy efficiency of wind tunnel fans and the cost-cutting of tests. On the other hand the air cooler is to assure the efficient cooling of air flow in such a manner as to maintain the temperature below 40 °C for seamless operation of measuring equipment. Therefore the relevance of this project is driven by the need to develop the air cooler that would demonstrate low hydraulic resistance of air and high thermal effectiveness of heat exchanging surfaces; insofar as the cooling section must be given up per unit time with the amount of heat Q=30 MW according to preliminary evaluations. On basis of calculation research some variants of air cooler designs are proposed including elliptical tubes, round tubes, and lateral plate-like fins. These designs differ by the number of tubes and plates, geometrical characteristics and the material of finned surfaces (aluminium or cooper). Due to the choice of component configurations a high thermal effectiveness is achieved for finned surfaces. The obtained results form the basis of R&D support in designing the new subsonic wind tunnel.

  8. OECD/NEA benchmark for time-dependent neutron transport calculations without spatial homogenization

    Energy Technology Data Exchange (ETDEWEB)

    Hou, Jason, E-mail: jason.hou@ncsu.edu [Department of Nuclear Engineering, North Carolina State University, Raleigh, NC 27695 (United States); Ivanov, Kostadin N. [Department of Nuclear Engineering, North Carolina State University, Raleigh, NC 27695 (United States); Boyarinov, Victor F.; Fomichenko, Peter A. [National Research Centre “Kurchatov Institute”, Kurchatov Sq. 1, Moscow (Russian Federation)

    2017-06-15

    Highlights: • A time-dependent homogenization-free neutron transport benchmark was created. • The first phase, known as the kinetics phase, was described in this work. • Preliminary results for selected 2-D transient exercises were presented. - Abstract: A Nuclear Energy Agency (NEA), Organization for Economic Co-operation and Development (OECD) benchmark for the time-dependent neutron transport calculations without spatial homogenization has been established in order to facilitate the development and assessment of numerical methods for solving the space-time neutron kinetics equations. The benchmark has been named the OECD/NEA C5G7-TD benchmark, and later extended with three consecutive phases each corresponding to one modelling stage of the multi-physics transient analysis of the nuclear reactor core. This paper provides a detailed introduction of the benchmark specification of Phase I, known as the “kinetics phase”, including the geometry description, supporting neutron transport data, transient scenarios in both two-dimensional (2-D) and three-dimensional (3-D) configurations, as well as the expected output parameters from the participants. Also presented are the preliminary results for the initial state 2-D core and selected transient exercises that have been obtained using the Monte Carlo method and the Surface Harmonic Method (SHM), respectively.

  9. Calculation of the Effective Delayed Neutron Fraction by Deterministic and Monte Carlo Methods

    Directory of Open Access Journals (Sweden)

    M. Carta

    2011-01-01

    Full Text Available The studies on Accelerator-Driven Systems (ADSs have renewed the interest in the theoretical and computational evaluation of the main integral parameters characterizing subcritical systems (e.g., reactivity, effective delayed neutron fraction βeff, and mean prompt neutron generation time. In particular, some kinetic parameters, as the effective delayed neutron fraction, are evaluated in Monte Carlo codes by formulations which do not require the calculation of the adjoint flux. This paper is focused on a theoretical and computational analysis about how the different βeff definitions are connected and which are the approximations inherent to the Monte Carlo definition with respect to the standard definition involving weighted integrals. By means of a refined transport computational analysis carried out in a coherent and consistent way, that is, using the same deterministic code and neutron data library for the βeff evaluation in different ways, the theoretical analysis is numerically confirmed. Both theoretical and numerical results confirm the effectiveness of the Monte Carlo βeff evaluation, at least in cases where spectral differences between total and prompt fluxes are negligible with respect to the value of the functionals entering the classical βeff formulation.

  10. RAMA Methodology for the Calculation of Neutron Fluence; Metodologia RAMA para el Calculo de la Fluencia Neutronica

    Energy Technology Data Exchange (ETDEWEB)

    Villescas, G.; Corchon, F.

    2013-07-01

    he neutron fluence plays an important role in the study of the structural integrity of the reactor vessel after a certain time of neutron irradiation. The NRC defined in the Regulatory Guide 1.190, the way must be estimated neutron fluence, including uncertainty analysis of the validation process (creep uncertainty is ? 20%). TRANSWARE Enterprises Inc. developed a methodology for calculating the neutron flux, 1,190 based guide, known as RAMA. Uncertainty values obtained with this methodology, for about 18 vessels, are less than 10%.

  11. Theoretical calculation of solid particles deposition from the air

    Directory of Open Access Journals (Sweden)

    Bobro Milan

    2002-03-01

    Full Text Available This paper presents the calculation of harmful substance deposition (air pollution from the point source (Slanèo, et al., 2001 using equation (1. The point source shall be understood as e.g. chimneys of factory, heat plant, incinerator, boiler plant, local heating plant, etc.The theoretical calculation of concentration (1, or deposition (8 is based on the study of transfer and dispersion of pollution in air (Slanèo, et al., 2000a. The movement of pollution in air consists of a movement of the air itself and a relative movement of pollution particles and air, while the movement of harmful substance in the smoke trail is under the influence of turbulent diffusion, convection and gravitation. Molecular diffusion is not important in this process. When calculating concentrations (1 and deposition (8 of air pollution on a particular place near the source, it is assumed that the air speed is constant, the direction of wind does not change with the height and the source of air pollution is time-constant. The change in the wind speed with the height depends on the stability class of atmosphere (temperature gradient (Slanèo, et al., 2000a and it is calculated using equation (10.The theoretical calculation of concentration and or deposition of harmful substance from the point source (1 and (8 shall be applied if the harmful substance particles, which leave the source, have the same density (composition, shape (spherical and size.The experimental observations of dust deposition showed the significance of 0.1-20 µm particles. The application of equation (1 to calculate the concentration is conditioned, in addition to the recognition of source parameters and meteorological conditions, by the recognition of the particle sedimentation speed, which changes with the size of particle radius (2.For a practical calculation of deposition it is therefore necessary to know the differential distribution function f(r of particle radii, which can be made on the basis

  12. PEGASUS: a preequilibrium and multi-step evaporation code for neutron cross section calculation

    Energy Technology Data Exchange (ETDEWEB)

    Nakagawa, Tsuneo; Sugi, Teruo [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Iijima, Shungo; Nishigori, Takeo

    1999-06-01

    The computer code PEGASUS was developed to calculate neutron-induced reaction cross sections on the basis of the closed form exciton model preequilibrium theory and the multi-step evaporation theory. The cross sections and emitted particle spectra are calculated for the compound elastic scattering, (n,{gamma}), (n,n`), (n,p), (n,{alpha}), (n,d), (n,t), (n,{sup 3}He), (n,2n), (n,n`p), (n,n`{alpha}), (n,n`d), (n,n`t), (n,2p) and (n,3n) reactions. The double differential cross sections of emitted particles are also calculated. The calculated results are written on a magnetic disk in the ENDF format. Parameter files and/or systematics formulas are provided for level densities, mass excess, radiation widths and inverse cross sections so that the input data to the code are made minimum. (author)

  13. Improving Neutron Kinetics and Thermal Hydraulics coupled tools for BEPU calculations

    Energy Technology Data Exchange (ETDEWEB)

    Pericas, R.; Reventós, F.; Batet, Il.

    2015-07-01

    The BEPU methodology is capable of providing a solution in terms of increasing the nuclear power production without compromising the safety margins. This study presents different improvements performed using tools available at UPC in the field of Neutron Kinetics and Thermal Hydraulics coupled systems. The paper describes a comparison between the BEPU methodology and the Conservative Bounding methodology within the framework of the Neutron Kinetics and Thermal Hydraulics coupled systems. To perform such comparison the following tools have been selected: TRACE for thermal-hydraulic system calculations, PARCS for reactor kinetics core simulator code. A Main Steam Line Break (MSLB) in a Pressurized Water Reactor (PWR) is the selected simulated transient to show the improvements performed. (Author)

  14. Calculation of the dynamic air flow resistivity of fibre materials

    DEFF Research Database (Denmark)

    Tarnow, Viggo

    1997-01-01

    The acoustic attenuation of acoustic fiber materials is mainly determined by the dynamic resistivity to an oscillating air flow. The dynamic resistance is calculated for a model with geometry close to the geometry of real fibre material. The model constists of parallel cylinders placed randomly......-consistent procedure gives the same results as the more complicated procedure based on average over Voronoi cells. Graphs of the dynamic resistivity versus frequency are given for fiber densities and diameters typical for acoustic fiber materials........The second procedure is an extension to oscillating air flow of the Brinkman self-consistent procedure for dc flow. The procedures are valid for volume concentrations of cylinders less than 0.1. The calculations show that for the density of fibers of interest for acoustic fibre materials the simple self...

  15. Calculation of Ambient (H*(10)) and Personal (Hp(10)) Dose Equivalent from a 252Cf Neutron Source

    Energy Technology Data Exchange (ETDEWEB)

    Traub, Richard J.

    2010-03-26

    The purpose of this calculation is to calculate the neutron dose factors for the Sr-Cf-3000 neutron source that is located in the 318 low scatter room (LSR). The dose factors were based on the dose conversion factors published in ICRP-21 Appendix 6, and the Ambient dose equivalent (H*(10)) and Personal dose equivalent (Hp(10)) dose factors published in ICRP Publication 74.

  16. Monte Carlo Calculation for Landmine Detection using Prompt Gamma Neutron Activation Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Park, Seungil; Kim, Seong Bong; Yoo, Suk Jae [Plasma Technology Research Center, Gunsan (Korea, Republic of); Shin, Sung Gyun; Cho, Moohyun [POSTECH, Pohang (Korea, Republic of); Han, Seunghoon; Lim, Byeongok [Samsung Thales, Yongin (Korea, Republic of)

    2014-05-15

    Identification and demining of landmines are a very important issue for the safety of the people and the economic development. To solve the issue, several methods have been proposed in the past. In Korea, National Fusion Research Institute (NFRI) is developing a landmine detector using prompt gamma neutron activation analysis (PGNAA) as a part of the complex sensor-based landmine detection system. In this paper, the Monte Carlo calculation results for this system are presented. Monte Carlo calculation was carried out for the design of the landmine detector using PGNAA. To consider the soil effect, average soil composition is analyzed and applied to the calculation. This results has been used to determine the specification of the landmine detector.

  17. Consistent neutron-physical and thermal-physical calculations of fuel rods of VVER type reactors

    Directory of Open Access Journals (Sweden)

    Tikhomirov Georgy

    2017-01-01

    Full Text Available For modeling the isotopic composition of fuel, and maximum temperatures at different moments of time, one can use different algorithms and codes. In connection with the development of new types of fuel assemblies and progress in computer technology, the task makes important to increase accuracy in modeling of the above characteristics of fuel assemblies during the operation. Calculations of neutronphysical characteristics of fuel rods are mainly based on models using averaged temperature, thermal conductivity factors, and heat power density. In this paper, complex approach is presented, based on modern algorithms, methods and codes to solve separate tasks of thermal conductivity, neutron transport, and nuclide transformation kinetics. It allows to perform neutron-physical and thermal-physical calculation of the reactor with detailed temperature distribution, with account of temperature-depending thermal conductivity and other characteristics. It was applied to studies of fuel cell of the VVER-1000 reactor. When developing new algorithms and programs, which should improve the accuracy of modeling the isotopic composition and maximum temperature in the fuel rod, it is necessary to have a set of test tasks for verification. The proposed approach can be used for development of such verification base for testing calculation of fuel rods of VVER type reactors

  18. Neutron Shielding Calculation of a Beam Stopper for the Thermal-TAS at HANARO using MCNP6

    Energy Technology Data Exchange (ETDEWEB)

    Jeon, Byoungil; Kim, Jongsoon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    It can be classified as a virtual source, neutron optics components, Monochromator Shielding Unit (MSU), Sample table and Analyzer and Detector bank. At a monochromator, specific wavelength of neutron beam is selected by the Bragg scattering theory. Because the neutron beam comes from reactor source has high intensity, massive shielding units are placed around monochromator. After monochromator, selected neutron beam goes toward sample table. This is a brief account of the Thermal-TAS experiment. To achieve high quality experimental results, signal noise should be controlled. Signal noise can come from not only electronic hardware system but also un-necessary neutron background. To reduce background, proper shielding components should be placed around optical components. Monochromator, analyzer and detector have their own shielding components but sample table doesn't in current status of Thermal-TAS. To reduce background occurred on sample table, direct neutron beam that comes from monochromator should be blocked. In this paper, basic design and radiation shielding calculation of direct beam stopper shielding component have been conducted. A conceptual design and shielding calculation of a neutron beam stopper for the thermal-TAS at HANARO has been conducted. With this result, a beam stopper segment will be fabricated. After further procedures such as mechanical, electronic design and fabrication of a driving part of the stopper, a neutron beam stopper system will be installed on the thermal-TAS at HANARO.

  19. 3D Monte-Carlo transport calculations of whole slab reactor cores: validation of deterministic neutronic calculation routes

    Energy Technology Data Exchange (ETDEWEB)

    Palau, J.M. [CEA Cadarache, Service de Physique des Reacteurs et du Cycle, Lab. de Projets Nucleaires, 13 - Saint-Paul-lez-Durance (France)

    2005-07-01

    This paper presents how Monte-Carlo calculations (French TRIPOLI4 poly-kinetic code with an appropriate pre-processing and post-processing software called OVNI) are used in the case of 3-dimensional heterogeneous benchmarks (slab reactor cores) to reduce model biases and enable a thorough and detailed analysis of the performances of deterministic methods and their associated data libraries with respect to key neutron parameters (reactivity, local power). Outstanding examples of application of these tools are presented regarding the new numerical methods implemented in the French lattice code APOLLO2 (advanced self-shielding models, new IDT characteristics method implemented within the discrete-ordinates flux solver model) and the JEFF3.1 nuclear data library (checked against JEF2.2 previous file). In particular we have pointed out, by performing multigroup/point-wise TRIPOLI4 (assembly and core) calculations, the efficiency (in terms of accuracy and computation time) of the new IDT method developed in APOLLO2. In addition, by performing 3-dimensional TRIPOLI4 calculations of the whole slab core (few millions of elementary volumes), the high quality of the new JEFF3.1 nuclear data files and revised evaluations (U{sup 235}, U{sup 238}, Hf) for reactivity prediction of slab cores critical experiments has been stressed. As a feedback of the whole validation process, improvements in terms of nuclear data (mainly Hf capture cross-sections) and numerical methods (advanced quadrature formulas accounting validation results, validation of new self-shielding models, parallelization) are suggested to improve even more the APOLLO2-CRONOS2 standard calculation route. (author)

  20. Neutron and gamma spectra measurements and calculations in benchmark spherical iron assemblies with sup 2 sup 5 sup 2 Cf neutron source in the centre

    CERN Document Server

    Jansky, B; Turzik, Z; Kyncl, J; Cvachovec, F; Trykov, L A; Volkov, V S

    2002-01-01

    The neutron and gamma spectra measurements have been made for benchmark iron spherical assemblies with the diameter of 30, 50 and 100 cm. The sup 2 sup 5 sup 2 Cf neutron sources with different emissions were placed into the centre of iron spheres. In the first stage of the project, independent laboratories took part in the leakage spectra measurements. The proton recoil method was used with stilbene crystals and hydrogen proportional counters. The working range of spectrometers for neutrons is in energy range from 0.01 to 16 MeV, and for gamma from 0.40 to 12 MeV. Some adequate calculations have been carried out. The propose to carefully analyse the leakage mixed neutron and gamma spectrum from iron sphere of diameter 50 cm and then adopt that field as standard.

  1. The statistical model calculation of prompt neutron spectra from spontaneous fission of {sup 244}Cm and {sup 246}Cm

    Energy Technology Data Exchange (ETDEWEB)

    Gerasimenko, B.F. [V.G. Khlopin Radium Inst., Saint Peterburg (Russian Federation)

    1997-03-01

    The calculations of integral spectra of prompt neutrons of spontaneous fission of {sup 244}Cm and {sup 246}Cm were carried out. The calculations were done by the Statistical Computer Code Complex SCOFIN applying the Hauser-Feschbach method as applied to the description of the de-excitation of excited fission fragments by means of neutron emission. The emission of dipole gamma-quanta from these fragments was considered as a competing process. The average excitation energy of a fragment was calculated by two-spheroidal model of tangent fragments. The density of levels in an excited fragment was calculated by the Fermi-gas model. The quite satisfactory agreement was reached between theoretical and experimental results obtained in frames of Project measurements. The calculated values of average multiplicities of neutron number were 2,746 for {sup 244}Cm and 2,927 for {sup 246}Cm that was in a good accordance with published experimental figures. (author)

  2. Development of Monte Carlo based real-time treatment planning system with fast calculation algorithm for boron neutron capture therapy.

    Science.gov (United States)

    Takada, Kenta; Kumada, Hiroaki; Liem, Peng Hong; Sakurai, Hideyuki; Sakae, Takeji

    2016-12-01

    We simulated the effect of patient displacement on organ doses in boron neutron capture therapy (BNCT). In addition, we developed a faster calculation algorithm (NCT high-speed) to simulate irradiation more efficiently. We simulated dose evaluation for the standard irradiation position (reference position) using a head phantom. Cases were assumed where the patient body is shifted in lateral directions compared to the reference position, as well as in the direction away from the irradiation aperture. For three groups of neutron (thermal, epithermal, and fast), flux distribution using NCT high-speed with a voxelized homogeneous phantom was calculated. The three groups of neutron fluxes were calculated for the same conditions with Monte Carlo code. These calculated results were compared. In the evaluations of body movements, there were no significant differences even with shifting up to 9mm in the lateral directions. However, the dose decreased by about 10% with shifts of 9mm in a direction away from the irradiation aperture. When comparing both calculations in the phantom surface up to 3cm, the maximum differences between the fluxes calculated by NCT high-speed with those calculated by Monte Carlo code for thermal neutrons and epithermal neutrons were 10% and 18%, respectively. The time required for NCT high-speed code was about 1/10th compared to Monte Carlo calculation. In the evaluation, the longitudinal displacement has a considerable effect on the organ doses. We also achieved faster calculation of depth distribution of thermal neutron flux using NCT high-speed calculation code. Copyright © 2016 Associazione Italiana di Fisica Medica. Published by Elsevier Ltd. All rights reserved.

  3. Neutron-gamma flux and dose calculations in a Pressurized Water Reactor (PWR

    Directory of Open Access Journals (Sweden)

    Brovchenko Mariya

    2017-01-01

    Full Text Available The present work deals with Monte Carlo simulations, aiming to determine the neutron and gamma responses outside the vessel and in the basemat of a Pressurized Water Reactor (PWR. The model is based on the Tihange-I Belgian nuclear reactor. With a large set of information and measurements available, this reactor has the advantage to be easily modelled and allows validation based on the experimental measurements. Power distribution calculations were therefore performed with the MCNP code at IRSN and compared to the available in-core measurements. Results showed a good agreement between calculated and measured values over the whole core. In this paper, the methods and hypotheses used for the particle transport simulation from the fission distribution in the core to the detectors outside the vessel of the reactor are also summarized. The results of the simulations are presented including the neutron and gamma doses and flux energy spectra. MCNP6 computational results comparing JEFF3.1 and ENDF-B/VII.1 nuclear data evaluations and sensitivity of the results to some model parameters are presented.

  4. Calculation of the neutron electric dipole moment with two dynamical flavors of domain wall fermions

    Energy Technology Data Exchange (ETDEWEB)

    F. Berruto; T. Blum; K. Orginos; A. Soni

    2005-12-08

    We present a study of the neutron electric dipole moment ({rvec d}{sub N}) within the framework of lattice QCD with two flavors of dynamical light quarks. The dipole moment is sensitive to the topological structure of the gauge fields, and accuracy can only be achieved by using dynamical, or sea quark, calculations. However, the topological charge evolves slowly in these calculations, leading to a relatively large uncertainty in {rvec d}{sub N}. It is shown, using quenched configurations, that a better sampling of the charge distribution reduces this problem, but because the CP even part of the fermion determinant is absent, both the topological charge distribution and {rvec d}{sub N} are pathological in the chiral limit. We discuss the statistical and systematic uncertainties arising from the topological charge distribution and unphysical size of the quark mass in our calculations and prospects for eliminating them. Our calculations employ the RBC collaboration two flavor domain wall fermion and DBW2 gauge action lattices with inverse lattice spacing a{sup -1} {approx} 1.7 GeV, physical volume V {approx} (2 fm){sup 3}, and light quark mass roughly equal to the strange quark mass (m{sub sea} = 0.03 and 0.04). We determine a value of the electric dipole moment that is zero within (statistical) errors, |{rvec d}{sub N}| = -0.04(20) e-{theta}-fm at the smaller sea quark mass. Satisfactory results for the magnetic and electric form factors of the proton and neutron are also obtained and presented.

  5. [Calculating method for crop water requirement based on air temperature].

    Science.gov (United States)

    Tao, Guo-Tong; Wang, Jing-Lei; Nan, Ji-Qin; Gao, Yang; Chen, Zhi-Fang; Song, Ni

    2014-07-01

    The importance of accurately estimating crop water requirement for irrigation forecast and agricultural water management has been widely recognized. Although it has been broadly adopted to determine crop evapotranspiration (ETc) via meteorological data and crop coefficient, most of the data in whether forecast are qualitative rather than quantitative except air temperature. Therefore, in this study, how to estimate ETc precisely only using air temperature data in forecast was explored, the accuracy of estimation based on different time scales was also investigated, which was believed to be beneficial to local irrigation forecast as well as optimal management of water and soil resources. Three parameters of Hargreaves equation and two parameters of McClound equation were corrected by using meteorological data of Xinxiang from 1970 to 2010, and Hargreaves equation was selected to calculate reference evapotranspiration (ET0) during the growth period of winter wheat. A model of calculating crop water requirement was developed to predict ETc at time scales of 1, 3, and 7 d intervals through combining Hargreaves equation and crop coefficient model based on air temperature. Results showed that the correlation coefficients between measured and predicted values of ETc reached 0.883 (1 d), 0.933 (3 d), and 0.959 (7 d), respectively. The consistency indexes were 0.94, 0.95 and 0.97, respectively, which showed that forecast error decreased with the increasing time scales. Forecasted accuracy with an error less than 1 mm x d(-1) was more than 80%, and that less than 2 mm x d(-1) was greater than 90%. This study provided sound basis for irrigation forecast and agricultural management in irrigated areas since the forecasted accuracy at each time scale was relatively high.

  6. Integrated system for production of neutronics and photonics calculational constants. Neutron-induced interactions: index of experimental data

    Energy Technology Data Exchange (ETDEWEB)

    MacGregor, M.H.; Cullen, D.E.; Howerton, R.J.; Perkins, S.T.

    1976-07-04

    Indexes to the neutron-induced interaction data in the Experimental Cross Section Information Library (ECSIL) as of July 4, 1976 are tabulated. The tabulation has two arrangements: isotope (ZA) order and reaction-number order.

  7. Analytical calculations of neutron slowing down and transport in the constant-cross-section problem

    Energy Technology Data Exchange (ETDEWEB)

    Cacuci, D.G.

    1978-04-01

    Aspects of the problem of neutron slowing down and transport in an infinite medium consisting of a single nuclide that scatters elastically and isotropically and has energy-independent cross sections were investigated. The method of singular eigenfunctions was applied to the Boltzmann Equation governing the Laplace transform (with respect to the lethargy variable) of the neutron flux. A new sufficient condition for the convergence of the coefficients of the expansion of the scattering kernel in Legendre polynomials was rigorously derived for this energy-dependent problem. Formulas were obtained for the lethargy-dependent spatial moments of the scalar flux that are valid for medium to large lethargies. Use was made of the well-known connection between the spatial moments of the Laplace-transformed scalar flux and the moments of the flux in the ''eigenvalue space.'' The calculations were aided by the construction of a closed general expression for these ''eigenvalue space'' moments. Extensive use was also made of the methods of combinatorial analysis and of computer evaluation of complicated sequences of manipulations. For the case of no absorption it was possible to obtain for materials of any atomic weight explicit corrections to the age-theory formulas for the spatial moments M/sub 2n/(u) of the scalar flux that are valid through terms of the order of u/sup -5/. The evaluation of the coefficients of the powers of n, as explicit functions of the nuclear mass, represent one of the end products of this investigation. In addition, an exact expression for the second spatial moment, M/sub 2/(u), valid for arbitrary (constant) absorption, was derived. It is now possible to calculate analytically and rigorously the ''age'' for the constant-cross-section problem for arbitrary (constant) absorption and nuclear mass. 5 figures, 1 table.

  8. ALPHN: A computer program for calculating ({alpha}, n) neutron production in canisters of high-level waste

    Energy Technology Data Exchange (ETDEWEB)

    Salmon, R.; Hermann, O.W.

    1992-10-01

    The rate of neutron production from ({alpha}, n) reactions in canisters of immobilized high-level waste containing borosilicate glass or glass-ceramic compositions is significant and must be considered when estimating neutron shielding requirements. The personal computer program ALPHA calculates the ({alpha}, n) neutron production rate of a canister of vitrified high-level waste. The user supplies the chemical composition of the glass or glass-ceramic and the curies of the alpha-emitting actinides present. The output of the program gives the ({alpha}, n) neutron production of each actinide in neutrons per second and the total for the canister. The ({alpha}, n) neutron production rates are source terms only; that is, they are production rates within the glass and do not take into account the shielding effect of the glass. For a given glass composition, the user can calculate up to eight cases simultaneously; these cases are based on the same glass composition but contain different quantities of actinides per canister. In a typical application, these cases might represent the same canister of vitrified high-level waste at eight different decay times. Run time for a typical problem containing 20 chemical species, 24 actinides, and 8 decay times was 35 s on an IBM AT personal computer. Results of an example based on an expected canister composition at the Defense Waste Processing Facility are shown.

  9. ALPHN: A computer program for calculating ([alpha], n) neutron production in canisters of high-level waste

    Energy Technology Data Exchange (ETDEWEB)

    Salmon, R.; Hermann, O.W.

    1992-10-01

    The rate of neutron production from ([alpha], n) reactions in canisters of immobilized high-level waste containing borosilicate glass or glass-ceramic compositions is significant and must be considered when estimating neutron shielding requirements. The personal computer program ALPHA calculates the ([alpha], n) neutron production rate of a canister of vitrified high-level waste. The user supplies the chemical composition of the glass or glass-ceramic and the curies of the alpha-emitting actinides present. The output of the program gives the ([alpha], n) neutron production of each actinide in neutrons per second and the total for the canister. The ([alpha], n) neutron production rates are source terms only; that is, they are production rates within the glass and do not take into account the shielding effect of the glass. For a given glass composition, the user can calculate up to eight cases simultaneously; these cases are based on the same glass composition but contain different quantities of actinides per canister. In a typical application, these cases might represent the same canister of vitrified high-level waste at eight different decay times. Run time for a typical problem containing 20 chemical species, 24 actinides, and 8 decay times was 35 s on an IBM AT personal computer. Results of an example based on an expected canister composition at the Defense Waste Processing Facility are shown.

  10. The design of a multisource americium-beryllium (Am-Be) neutron irradiation facility using MCNP for the neutronic performance calculation.

    Science.gov (United States)

    Sogbadji, R B M; Abrefah, R G; Nyarko, B J B; Akaho, E H K; Odoi, H C; Attakorah-Birinkorang, S

    2014-08-01

    The americium-beryllium neutron irradiation facility at the National Nuclear Research Institute (NNRI), Ghana, was re-designed with four 20 Ci sources using Monte Carlo N-Particle (MCNP) code to investigate the maximum amount of flux that is produced by the combined sources. The results were compared with a single source Am-Be irradiation facility. The main objective was to enable us to harness the maximum amount of flux for the optimization of neutron activation analysis and to enable smaller sample sized samples to be irradiated. Using MCNP for the design construction and neutronic performance calculation, it was realized that the single-source Am-Be design produced a thermal neutron flux of (1.8±0.0007)×10(6) n/cm(2)s and the four-source Am-Be design produced a thermal neutron flux of (5.4±0.0007)×10(6) n/cm(2)s which is a factor of 3.5 fold increase compared to the single-source Am-Be design. The criticality effective, k(eff), of the single-source and the four-source Am-Be designs were found to be 0.00115±0.0008 and 0.00143±0.0008, respectively. Copyright © 2014 Elsevier Ltd. All rights reserved.

  11. Σ-admixture in neutron-rich Li Λ hypernuclei in a microscopic shell-model calculation

    Directory of Open Access Journals (Sweden)

    Harada T.

    2010-04-01

    Full Text Available We systematically investigate the structures of the Λ hypernuclei ΛLi with the mass number A = 7–10 in shell-model calculations considering the ΛN -ΣN coupling in the first-order perturbation method. We find that the calculated Σ-mixing probabilities and energy shifts due to the ΛN -ΣN coupling increase with the neutron number. The Fermi-type and Gamow-Teller-type couplings, which are related to the β-transition properties of the nuclear core state, coherently contribute to the energy shift in neutron-rich hypernuclei.

  12. Neutron capture cross section measurements and theoretical calculation for the {sup 186}W(n,γ){sup 187}W reaction

    Energy Technology Data Exchange (ETDEWEB)

    Al-abyad, Mogahed; Mohamed, Gehan Y. [Atomic Energy Authority, Cairo (Egypt). Experimental Nuclear Physics Dept.

    2017-08-01

    Neutron capture cross section (σ{sub 0}) and resonance integral (I{sub 0}) of the reaction {sup 186}W(n,γ){sup 187}W were measured experimentally using the research reactor (ETRR-2) and an Am-Be neutron source, also calculated using TALYS-1.6 code. The present results of σ{sub 0} are (39.08±2.6, 38.75±0.98 and 38.33 barn) and I{sub 0} are (418.5±74, 439.3±36 and 445.5 barn) by using the reactor, neutron source and TALYS-1.6, respectively. The present results are in acceptable agreement with most of the previous experimental and evaluated data as well as the theoretical calculations.

  13. Measurement and calculations of long-lived radionuclide activity forming in the fast neutron field in some ITER construction steels

    Energy Technology Data Exchange (ETDEWEB)

    Pohorecki, W., E-mail: poho@agh.edu.pl [AGH University of Science and Technology, Faculty of Energy and Fuels, Al. Mickiewicza 30, 30-059 Krakow (Poland); Jodłowski, P. [AGH University of Science and Technology, Faculty of Physics and Applied Computer Science, Al. Mickiewicza 30, 30-059 Krakow (Poland); Pytel, K.; Prokopowicz, R. [National Centre for Nuclear Research, ul. Sołtana 7, 05-400 Otwock-Świerk (Poland)

    2014-10-15

    Highlights: • Measurement and calculations of long-lived gamma-emitting radionuclide activity forming in the fission reactor fast neutron field were done, in some ITER construction steels. • The neutron flux density was measured by means of activation foil method and unfolding technique. • Activity calculations were done by means of FISPACT-II code using the activation libraries EAF-2010 and TALYS-2011. • The activity measurements were done by means of gamma-ray spectrometry. - Abstract: Measurement and calculations of long-lived gamma-emitting radionuclide activity forming in the fission reactor fast neutron field were done, for some ITER construction steels. The activation was conducted in fast neutron irradiation channel of the MARIA research fission reactor (Poland). The dimensions of steel samples were 10 mm × 10 mm × 1 mm and mass was approximately 0.8 g. The neutron flux density was measured by means of activation foil method and unfolding technique; fraction of neutrons above 1 keV was 95%. The activation lasted 242 h and cooling took 100 days; the mean neutron flux density was 2.9E12 n/(cm{sup 2} s) (neutrons above 500 keV are 53% of total) whereas total fluency 2.53E18 cm{sup −2}. The activity measurements were done by means of gamma-ray spectrometry. Activity calculations were done by means of FISPACT-II code using the activation libraries EAF-2010 and TENDL-2011 and experimentally determined neutron flux. Measured activity of long-lived gamma emitting radionuclides was, in average, about 6.3 MBq/g 100 days after activation; the dominant radionuclides were {sup 58}Co and {sup 54}Mn (about 81% and 14% of total activity respectively). The C/E ratio differs for particular radionuclides and is in the range 0.86–0.92 for {sup 51}Cr, 0.93–1.21 for {sup 54}Mn, 0.77–0.98 for {sup 57}Co, 0.91–1.21 for {sup 58}Co, 1.17–1.27 for {sup 59}Fe, and 1.75–2.44 for {sup 60}Co.

  14. Estimates of neutron leakage through penetrations of the CERN intersecting storage rings by Monte Carlo albedo calculations

    CERN Document Server

    Routti, J T

    1975-01-01

    The monokinetic and multigroup Monte Carlo albedo methods applicable to estimating neutron leakage through penetrations in the shielding of high-energy accelerators are reviewed. They are used to calculate attenuation factors and dose levels in the tunnels of the CERN intersecting storage rings. (28 refs).

  15. Gamma-Ray Emission Spectra as a Constraint on Calculations of 234 , 236 , 238U Neutron-Capture Cross Sections

    Science.gov (United States)

    Ullmann, J. L.; Krticka, M.; Kawano, T.; Bredeweg, T. A.; Baramsai, B.; Couture, A.; Haight, R. C.; Jandel, M.; Mosby, S.; O'Donnell, J. M.; Rundberg, R. S.; Vieira, D. J.; Wilhelmy, J. B.; Becker, J. A.; Wu, C. Y.; Chyzh, A.

    2015-10-01

    Calculations of the neutron-capture cross section at low neutron energies (10 eV through 100's of keV) are very sensitive to the nuclear level density and radiative strength function. These quantities are often poorly known, especially for radioactive targets, and actual measurements of the capture cross section are usually required. An additional constraint on the calculation of the capture cross section is provided by measurements of the cascade gamma spectrum following neutron capture. Recent measurements of 234 , 236 , 238U(n, γ) emission spectra made using the DANCE 4 π BaF2 array at the Los Alamos Neutron Science Center will be presented. Calculations of gamma-ray spectra made using the DICEBOX code and of the capture cross section made using the CoH3 code will also be presented. These techniques may be also useful for calculations of more unstable nuclides. This work was performed with the support of the U.S. Department of Energy, National Nuclear Security Administration by Los Alamos National Security, LLC (Contract DE-AC52-06NA25396) and Lawrence Livermore National Security, LLC (Contract DE-AC52-07NA2734).

  16. Verification of Monte Carlo calculations of the neutron flux in typical irradiation channels of the TRIGA reactor, Ljubljana

    NARCIS (Netherlands)

    Jacimovic, R; Maucec, M; Trkov, A

    2003-01-01

    An experimental verification of Monte Carlo neutron flux calculations in typical irradiation channels in the TRIGA Mark II reactor at the Jozef Stefan Institute is presented. It was found that the flux, as well as its spectral characteristics, depends rather strongly on the position of the

  17. Measurement and calculation of neutron leakage spectra from slab samples of beryllium, gallium and tungsten irradiated with 14.8 MeV neutrons

    Science.gov (United States)

    Nie, Y. B.; Ruan, X. C.; Ren, J.; Zhang, S.; Han, R.; Bao, J.; Huang, H. X.; Ding, Y. Y.; Wu, H. C.; Liu, P.; Zhou, Z. Y.

    2017-09-01

    In order to make benchmark validation of the nuclear data for gallium (Ga), tungsten (W) and beryllium (Be) in existing modern evaluated nuclear data files, neutron leakage spectra in the range from 0.8 to 15 MeV from slab samples were measured by time-of-flight technique with a BC501 scintillation detector. The measurements were performed at China Institute of Atomic Energy (CIAE) using a D-T neutron source. The thicknesses of the slabs were 0.5 to 2.5 mean free path for 14.8 MeV neutrons, and the measured angles were chosen to be 60∘ and 120∘. The measured spectra were compared with those calculated by the continuous energy Monte-Carlo transport code MCNP, using the data from the CENDL-3.1, ENDF/B-VII.1 and JENDL-4.0 nuclear data files, the comparison between the experimental and calculated results show that: The results from all three libraries significantly underestimate the cross section in energy range of 10-13 MeV for Ga; For W, the calculated spectra using data from CENDL-3.1 and JENDL-4.0 libraries show larger discrepancies with the measured ones, especially around 8.5-13.5 MeV; and for Be, all the libraries led to underestimation below 3 MeV at 120∘.

  18. Relativistic collision rate calculations for electron-air interactions

    Energy Technology Data Exchange (ETDEWEB)

    Graham, G. [EG and G Energy Measurements, Inc., Los Alamos, NM (United States); Roussel-Dupre, R. [Los Alamos National Lab., NM (United States)

    1993-12-01

    The most recent data available on differential cross sections for electron-air interactions are used to calculate the avalanche, momentum transfer, and energy loss rates that enter into the fluid equations. Data for the important elastic, inelastic, and ionizing processes are generally available out to electron energies of 1--10 keV. Prescriptions for extending these cross sections to the relativistic regime are presented. The angular dependence of the cross sections is included where data are available as is the doubly differential cross section for ionizing collisions. The collision rates are computed by taking moments of the Boltzmann collision integrals with the assumption that the electron momentum distribution function is given by the Juettner distribution function which satisfies the relativistic H- theorem and which reduces to the familiar Maxwellian velocity distribution in the nonrelativistic regime. The distribution function is parameterized in terms of the electron density, mean momentum, and thermal energy and the rates are therefore computed on a two dimensional grid as a function of mean kinetic energy and thermal energy.

  19. Neutron imaging of diabatic two-phase flows relevant to air conditioning

    Energy Technology Data Exchange (ETDEWEB)

    Geoghegan, Patrick J [ORNL; Sharma, Vishaldeep [ORNL

    2017-01-01

    The design of the evaporator of an air conditioning system relies heavily on heat transfer coefficients and pressure drop correlations that predominantly involve an estimate of the changing void fraction and the underlying two-phase flow regime. These correlations dictate whether the resulting heat exchanger is oversized or not and the amount of refrigerant charge necessary to operate. The latter is particularly important when dealing with flammable or high GWP refrigerants. Traditional techniques to measure the void fraction and visualize the flow are either invasive to the flow or occur downstream of the evaporator, where some of the flow distribution will have changed. Neutron imaging has the potential to visualize two-phase flow in-situ where an aluminium heat exchanger structure becomes essentially transparent to the penetrating neutrons. The subatomic particles are attenuated by the passing refrigerant flow. The resulting image may be directly related to the void fraction and the overall picture provides a clear insight into the flow regime present. This work presents neutron images of the refrigerant Isopentane as it passes through the flow channels of an aluminium evaporator at flowrates relevant to air conditioning. The flow in a 4mm square macro channel is compared to that in a 250 m by 750 m rectangular microchannel in terms of void fraction and regime. All neutron imaging experiments were conducted at the High Flux Isotope Reactor, an Oak Ridge National Laboratory facility

  20. Neutron total cross section calculation within the framework of quasi-harmonic approximation

    DEFF Research Database (Denmark)

    Cai, Xiao Xiao; Klinkby, Esben Bryndt

    2017-01-01

    The accuracy of neutron scattering cross sections is the gauge for the realistic outcome of a neutron transport simulation. To improve the traditional harmonic physics model used in such simulations, we revisit the slow neutron transport theory in crystalline materials and aim to develop a unified...... model that has good performance for neutron transport problems in crystals in a wide range of temperatures and pressures. The quasi-harmonic approximation (QHA) correlates phonon evolution explicitly with unit cell volume. Therefore, it is capable of evaluating a variety of material properties at finite...

  1. MC/sup 2/-2: a code to calculate fast neutron spectra and multigroup cross sections. [LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    Henryson, H. II; Toppel, B. J.; Stenberg, C. G.

    1976-06-01

    MC/sup 2/-2 is a program to solve the neutron slowing down problem using basic neutron data derived from the ENDF/B data files. The spectrum calculated by MC/sup 2/-2 is used to collapse the basic data to multigroup cross sections for use in standard reactor neutronics codes. Four different slowing down formulations are used by MC/sup 2/-2: multigroup, continuous slowing down using the Goertzel-Greuling or Improved Goertzel-Greuling moderating parameters, and a hyper-fine-group integral transport calculation. Resolved and unresolved resonance cross sections are calculated accounting for self-shielding, broadening and overlap effects. This document provides a description of the MC/sup 2/-2 program. The physics and mathematics of the neutron slowing down problem are derived and detailed information is provided to aid the MC/sup 2/-2 user in preparing input for the program and implementation of the program on IBM 370 or CDC 7600 computers.

  2. Boron neutron capture therapy design calculation of a 3H(p,n reaction based BSA for brain cancer setup

    Directory of Open Access Journals (Sweden)

    Bassem Elshahat

    2015-09-01

    Full Text Available Purpose: Boron neutron capture therapy (BNCT is a promising technique for the treatment of malignant disease targeting organs of the human body. Monte Carlo simulations were carried out to calculate optimum design parameters of an accelerator based beam shaping assembly (BSA for BNCT of brain cancer setup.Methods: Epithermal beam of neutrons were obtained through moderation of fast neutrons from 3H(p,n reaction in a high density polyethylene moderator and a graphite reflector. The dimensions of the moderator and the reflector were optimized through optimization of epithermal / fast neutron intensity ratio as a function of geometric parameters of the setup. Results: The results of our calculation showed the capability of our setup to treat the tumor within 4 cm of the head surface. The calculated peak therapeutic ratio for the setup was found to be 2.15. Conclusion: With further improvement in the polyethylene moderator design and brain phantom irradiation arrangement, the setup capabilities can be improved to reach further deep-seated tumor.

  3. Calculation of the Chilling Requirement for Air Conditioning in the Excavation Roadway

    Directory of Open Access Journals (Sweden)

    Yueping Qin

    2015-10-01

    Full Text Available To effectively improve the climate conditions of the excavation roadway in coal mine, the calculation of the chilling requirement taking air conditioning measures is extremely necessary. The temperature field of the surrounding rock with moving boundary in the excavation roadway was numerically simulated by using finite volume method. The unstable heat transfer coefficient between the surrounding rock and air flow was obtained via the previous calculation. According to the coupling effects of the air flow inside and outside air duct, the differential calculation mathematical model of air flow temperature in the excavation roadway was established. The chilling requirement was calculated with the selfdeveloped computer program for forecasting the required cooling capacity of the excavation roadway. A good air conditioning effect had been observed after applying the calculated results to field trial, which indicated that the prediction method and calculation procedure were reliable.

  4. Long-Term Calculations with Large Air Pollution Models

    DEFF Research Database (Denmark)

    Ambelas Skjøth, C.; Bastrup-Birk, A.; Brandt, J.

    1999-01-01

    Proceedings of the NATO Advanced Research Workshop on Large Scale Computations in Air Pollution Modelling, Sofia, Bulgaria, 6-10 July 1998......Proceedings of the NATO Advanced Research Workshop on Large Scale Computations in Air Pollution Modelling, Sofia, Bulgaria, 6-10 July 1998...

  5. Design and spectrum calculation of 4H-SiC thermal neutron detectors using FLUKA and TCAD

    Energy Technology Data Exchange (ETDEWEB)

    Huang, Haili; Tang, Xiaoyan; Guo, Hui, E-mail: guohui@mail.xidian.edu.cn; Zhang, Yimen; Zhang, Yimeng; Zhang, Yuming

    2016-10-11

    SiC is a promising material for neutron detection in a harsh environment due to its wide band gap, high displacement threshold energy and high thermal conductivity. To increase the detection efficiency of SiC, a converter such as {sup 6}LiF or {sup 10}B is introduced. In this paper, pulse-height spectra of a PIN diode with a {sup 6}LiF conversion layer exposed to thermal neutrons (0.026 eV) are calculated using TCAD and Monte Carlo simulations. First, the conversion efficiency of a thermal neutron with respect to the thickness of {sup 6}LiF was calculated by using a FLUKA code, and a maximal efficiency of approximately 5% was achieved. Next, the energy distributions of both {sup 3}H and α induced by the {sup 6}LiF reaction according to different ranges of emission angle are analyzed. Subsequently, transient pulses generated by the bombardment of single {sup 3}H or α-particles are calculated. Finally, pulse height spectra are obtained with a detector efficiency of 4.53%. Comparisons of the simulated result with the experimental data are also presented, and the calculated spectrum shows an acceptable similarity to the experimental data. This work would be useful for radiation-sensing applications, especially for SiC detector design.

  6. Integrated system for production of neutronics and photonics calculational constants. Major neutron-induced interactions (Z > 55): graphical, experimental data

    Energy Technology Data Exchange (ETDEWEB)

    Cullen, D.E.; Howerton, R.J.; MacGregor, M.H.; Perkins, S.T.

    1976-07-04

    This report (vol. 7) presents graphs of major neutron-induced interaction cross sections in the Experimental Cross Section Information Library (ECSIL) as of July 4, 1976. It consists primarily of interactions where a single data set contains enough points to show cross section behavior. In contrast, vol. 8 of this UCRL-50400 series consists of interactions where more than one data set is needed to show cross section behavior. Thus, you can find the total, elastic, capture, and fission cross sections (along with the parameters ..nu.. bar, ..cap alpha.., and eta) in vol. 7 and all other reactions in vol. 8. Data are plotted with associated cross section error bars (when given) and compared with the Evaluated Nuclear Data Library (ENDL) as of July 4, 1976. The plots are arranged in ascending order of atomic number (Z) and atomic weight (A). Part A contains the plots for Z = 1 to 55; Part B contains the plots for Z is greater than 55.

  7. 3D Neutronic Analysis in MHD Calculations at ARIES-ST Fusion Reactors Systems

    Science.gov (United States)

    Hançerliogulları, Aybaba; Cini, Mesut

    2013-10-01

    In this study, we developed new models for liquid wall (FW) state at ARIES-ST fusion reactor systems. ARIES-ST is a 1,000 MWe fusion reactor system based on a low aspect ratio ST plasma. In this article, we analyzed the characteristic properties of magnetohydrodynamics (MHD) and heat transfer conditions by using Monte-Carlo simulation methods (ARIES Team et al. in Fusion Eng Des 49-50:689-695, 2000; Tillack et al. in Fusion Eng Des 65:215-261, 2003) . In fusion applications, liquid metals are traditionally considered to be the best working fluids. The working liquid must be a lithium-containing medium in order to provide adequate tritium that the plasma is self-sustained and that the fusion is a renewable energy source. As for Flibe free surface flows, the MHD effects caused by interaction with the mean flow is negligible, while a fairly uniform flow of thick can be maintained throughout the reactor based on 3-D MHD calculations. In this study, neutronic parameters, that is to say, energy multiplication factor radiation, heat flux and fissile fuel breeding were researched for fusion reactor with various thorium and uranium molten salts. Sufficient tritium amount is needed for the reactor to work itself. In the tritium breeding ratio (TBR) >1.05 ARIES-ST fusion model TBR is >1.1 so that tritium self-sufficiency is maintained for DT fusion systems (Starke et al. in Fusion Energ Des 84:1794-1798, 2009; Najmabadi et al. in Fusion Energ Des 80:3-23, 2006).

  8. Uncertainties of the neutronic calculations at core level determined by the KARATE code system and the KIKO3D code

    Energy Technology Data Exchange (ETDEWEB)

    Panka, Istvan; Kereszturi, Andras [Hungarian Academy of Sciences, Budapest (Hungary). Reactor Analysis Dept.

    2013-09-15

    In this paper the uncertainties of the neutronic calculations at core level - originating from the uncertainties of the basic nuclear data - are presented. The investigations have been made for a VVER-1000 core (Kozloduy-6) defined in the frame of the OECD NEA UAM benchmark. In the first part of the paper, the uncertainties of the effective multiplication factor, the assembly-wise radial power distribution, the axial power distribution and the rod worth are shown. After that the preliminary evaluation of the uncertainties of the neutron kinetic calculations are presented for a rod movement transient at HZP (Hot Zero Power) state, where the uncertainties of the time dependent core and assembly powers and the dynamic reactivity were evaluated. In both cases, we will see that the most important quantities - at core level and at HZP state - have a considerable uncertainty which is originating from the uncertainties of the basic cross section library in these investigations. (orig.)

  9. Intercomparison of delayed neutron summation calculations among JEF2.2, ENDF/B-VI and JNDC-V2

    Energy Technology Data Exchange (ETDEWEB)

    Sagisaka, Mitsuyuki [Nagoya Univ. (Japan); Oyamatsu, K.; Kukita, Y.

    1998-03-01

    We perform intercomparison of delayed neutron activities calculated with JEF2.2, ENDF/B-VI and JNDC-V2 with a simple new method. Significant differences are found at t < 20 (s) for major fissioning systems. The differences are found to stem from fission yields or decay data of several nuclides. The list of these nuclides are also given for the future experimental determination of these nuclear data. (author)

  10. 40 CFR 86.166-12 - Method for calculating emissions due to air conditioning leakage.

    Science.gov (United States)

    2010-07-01

    ... to air conditioning leakage. 86.166-12 Section 86.166-12 Protection of Environment ENVIRONMENTAL... for calculating emissions due to air conditioning leakage. This section describes procedures used to determine a refrigerant leakage rate in grams per year from vehicle-based air conditioning units. The...

  11. Whole core pin-by-pin coupled neutronic-thermal-hydraulic steady state and transient calculations using COBAYA3 code

    Energy Technology Data Exchange (ETDEWEB)

    Jimenez, J.; Herrero, J. J.; Cuervo, D.; Aragones, J. M., E-mail: jimenez@din.upm.e [Universidad Politecnica de Madrid, Departamento de Ingenieria Nuclear, Jose Gutierrez Abascal No. 2, 28006 Madrid (Spain)

    2010-10-15

    Nowadays coupled 3-dimensional neutron kinetics and thermal-hydraulic core calculations are performed by applying a radial average channel approach using a meshing of one quarter of assembly in the best case. This approach does not take into account the subchannels effects due to the averaging of the physical fields and the loose of heterogeneity in the thermal-hydraulic model. Therefore the models do not have enough resolution to predict those subchannels effects which are important for the fuel design safety margins, because it is in the local scale, where we can search the hottest pellet or the maximum heat flux. The Polytechnic University of Madrid advanced multi-scale neutron-kinetics and thermal-hydraulics methodologies being implemented in COBAYA3 include domain decomposition by alternate core dissections for the local 3-dimensional fine-mesh scale problems (pin cells/subchannels) and an analytical nodal diffusion solver for the coarse mesh scale coupled with the thermal-hydraulic using a model of one channel per assembly or per quarter of assembly. In this work, we address the domain decomposition by the alternate core dissections methodology applied to solve coupled 3-dimensional neutronic-thermal-hydraulic problems at the fine-mesh scale. The neutronic-thermal-hydraulic coupling at the cell-subchannel scale allows the treatment of the effects of the detailed thermal-hydraulic feedbacks on cross-sections, thus resulting in better estimates of the local safety margins at the pin level. (Author)

  12. Thick activation detectors for neutron spectrometry using different unfolding methods: sensitivity analysis and dose calculation

    Energy Technology Data Exchange (ETDEWEB)

    Medkour Ishak-Boushaki, Ghania, E-mail: gmedkour@yahoo.com [Laboratoire SNIRM-Faculte de Physique, Universite des Sciences et de la Technologie Houari Boumediene, BP 32 El-Alia BabEzzouar, Algiers (Algeria); Boukeffoussa, Khelifa [Laboratoire SNIRM-Faculte de Physique, Universite des Sciences et de la Technologie Houari Boumediene, BP 32 El-Alia BabEzzouar, Algiers (Algeria); Idiri, Zahir [Centre de Recherche Nucleaire d' Alger, 02 Boulevard Frantz-Fanon, BP 399, Algiers (Algeria); Allab, Malika [Laboratoire SNIRM-Faculte de Physique, Universite des Sciences et de la Technologie Houari Boumediene, BP 32 El-Alia BabEzzouar, Algiers (Algeria)

    2012-03-15

    This paper discusses the use of threshold detectors of extended sizes for low intensity neutron fields' characterization. The detectors were tested by the measurement of the neutron spectrum of an {sup 241}Am-Be source. Integral quantities characterizing the neutron field, required for radiological protection, have been derived by unfolding the measured data. A good agreement is achieved between the obtained results and those deduced using Bonner spheres. In addition, a sensitivity analysis of the results to the deconvolution procedure is given. - Highlights: Black-Right-Pointing-Pointer Low intensity neutron fields' characterization using thick threshold detectors. Black-Right-Pointing-Pointer Low activity {sup 241}Am-Be neutron source spectrum measurement. Black-Right-Pointing-Pointer Integral quantities required for radiological protection have been derived. Black-Right-Pointing-Pointer The results are in good agreement with those deduced using Bonner spheres. Black-Right-Pointing-Pointer The results are not very sensitive to the chosen deconvolution procedure.

  13. SOURCES 4A: A Code for Calculating (alpha,n), Spontaneous Fission, and Delayed Neutron Sources and Spectra

    Energy Technology Data Exchange (ETDEWEB)

    Madland, D.G.; Arthur, E.D.; Estes, G.P.; Stewart, J.E.; Bozoian, M.; Perry, R.T.; Parish, T.A.; Brown, T.H.; England, T.R.; Wilson, W.B.; Charlton, W.S.

    1999-09-01

    SOURCES 4A is a computer code that determines neutron production rates and spectra from ({alpha},n) reactions, spontaneous fission, and delayed neutron emission due to the decay of radionuclides. The code is capable of calculating ({alpha},n) source rates and spectra in four types of problems: homogeneous media (i.e., a mixture of {alpha}-emitting source material and low-Z target material), two-region interface problems (i.e., a slab of {alpha}-emitting source material in contact with a slab of low-Z target material), three-region interface problems (i.e., a thin slab of low-Z target material sandwiched between {alpha}-emitting source material and low-Z target material), and ({alpha},n) reactions induced by a monoenergetic beam of {alpha}-particles incident on a slab of target material. Spontaneous fission spectra are calculated with evaluated half-life, spontaneous fission branching, and Watt spectrum parameters for 43 actinides. The ({alpha},n) spectra are calculated using an assumed isotropic angular distribution in the center-of-mass system with a library of 89 nuclide decay {alpha}-particle spectra, 24 sets of measured and/or evaluated ({alpha},n) cross sections and product nuclide level branching fractions, and functional {alpha}-particle stopping cross sections for Z < 106. The delayed neutron spectra are taken from an evaluated library of 105 precursors. The code outputs the magnitude and spectra of the resultant neutron source. It also provides an analysis of the contributions to that source by each nuclide in the problem.

  14. SOURCES 4C : a code for calculating ([alpha],n), spontaneous fission, and delayed neutron sources and spectra.

    Energy Technology Data Exchange (ETDEWEB)

    Wilson, W. B. (William B.); Perry, R. T. (Robert T.); Shores, E. F. (Erik F.); Charlton, W. S. (William S.); Parish, Theodore A.; Estes, G. P. (Guy P.); Brown, T. H. (Thomas H.); Arthur, Edward D. (Edward Dana),; Bozoian, Michael; England, T. R.; Madland, D. G.; Stewart, J. E. (James E.)

    2002-01-01

    SOURCES 4C is a computer code that determines neutron production rates and spectra from ({alpha},n) reactions, spontaneous fission, and delayed neutron emission due to radionuclide decay. The code is capable of calculating ({alpha},n) source rates and spectra in four types of problems: homogeneous media (i.e., an intimate mixture of a-emitting source material and low-Z target material), two-region interface problems (i.e., a slab of {alpha}-emitting source material in contact with a slab of low-Z target material), three-region interface problems (i.e., a thin slab of low-Z target material sandwiched between {alpha}-emitting source material and low-Z target material), and ({alpha},n) reactions induced by a monoenergetic beam of {alpha}-particles incident on a slab of target material. Spontaneous fission spectra are calculated with evaluated half-life, spontaneous fission branching, and Watt spectrum parameters for 44 actinides. The ({alpha},n) spectra are calculated using an assumed isotropic angular distribution in the center-of-mass system with a library of 107 nuclide decay {alpha}-particle spectra, 24 sets of measured and/or evaluated ({alpha},n) cross sections and product nuclide level branching fractions, and functional {alpha}-particle stopping cross sections for Z < 106. The delayed neutron spectra are taken from an evaluated library of 105 precursors. The code provides the magnitude and spectra, if desired, of the resultant neutron source in addition to an analysis of the'contributions by each nuclide in the problem. LASTCALL, a graphical user interface, is included in the code package.

  15. CFD - neutronic coupled calculation of a quarter of a simplified PWR fuel assembly including spacer pressure drop and turbulence enhancement

    Energy Technology Data Exchange (ETDEWEB)

    Pena, C.; Pellacani, F.; Macian Juan, R., E-mail: carlos.pena@ntech.mw.tum.de, E-mail: pellacani@ntech.mw.tum.de, E-mail: macian@ntech.mw.tum.de [Technische Universitaet Muenchen, Garching (Germany). Ntech Lehrstuhl fuer Nukleartechnik; Chiva, S., E-mail: schiva@emc.uji.es [Universitat Jaume I, Castellon de la Plana (Spain). Dept. de Ingenieria Mecanica y Construccion; Barrachina, T.; Miro, R., E-mail: rmiro@iqn.upv.es, E-mail: tbarrachina@iqn.upv.es [Universitat Politecnica de Valencia (ISIRYM/UPV) (Spain). Institute for Industrial, Radiophysical and Environmental Safety

    2011-07-01

    A computational code system based on coupling the 3D neutron diffusion code PARCS v2.7 and the Ansys CFX 13.0 Computational Fluid Dynamics (CFD) code has been developed as a tool for nuclear reactor systems simulations. This paper presents the coupling methodology between the CFD and the neutronic code. The methodology to simulate a 3D-neutronic problem coupled with 1D thermal hydraulics is already a mature technology, being part of the regular calculations performed to analyze different kinds of Reactivity Insertion Accidents (RIA) and asymmetric transients in Nuclear Power Plants, with state-of-the-art coupled codes like TRAC-B/NEM, RELAP5/PARCS, TRACE/PARCS, RELAP3D, RETRAN3D, etc. This work represents one of the first attempts to couple the multiphysics of a nuclear reactor core with a 3D spatial resolution in a computer code. This will open new possibilities regarding the analysis of fuel elements, contributing to a better understanding and design of the heat transfer process and specific fluid dynamics phenomena such as cross flow among fuel elements. The transient simulation of control rod insertion, boron dilution and cold water injection will be made possible with a degree of accuracy not achievable with current methodologies based on the use of system and/or subchannel codes. The transport of neutrons depends on several parameters, like fuel temperature, moderator temperature and density, boron concentration and fuel rod insertion. These data are calculated by the CFD code with high local resolution and used as input to the neutronic code to calculate a 3D nodal power distribution that will be returned and remapped to the CFD code control volumes (cells). Since two different nodalizations are used to discretized the same system, an averaging and interpolating procedure is needed to realize an effective data exchange. These procedures have been developed by means of the Ansys CFX 'User Fortran' interface; a library with several subroutines has

  16. Slow neutron total cross-section, transmission and reflection calculation for poly- and mono-NaCl and PbF{sub 2} crystals

    Energy Technology Data Exchange (ETDEWEB)

    Mansy, Muhammad S., E-mail: mmansy88@asrt.sci.eg [Reactor Physics Department, Nuclear Research Centre, Atomic Energy Authority, Cairo (Egypt); Radioactive Waste Management Unit, Hot Labs Centre, Atomic Energy Authority, Cairo (Egypt); Adib, M.; Habib, N. [Reactor Physics Department, Nuclear Research Centre, Atomic Energy Authority, Cairo (Egypt); Bashter, I.I. [Physics Department, Faculty of Science, Zagazig University (Egypt); Morcos, H.N.; El-Mesiry, M.S. [Reactor Physics Department, Nuclear Research Centre, Atomic Energy Authority, Cairo (Egypt)

    2016-10-01

    Highlights: • Slow neutron cross-section calculation for poly- and mono-crystalline materials. • Monochromatic features of PbF{sub 2} and NaCl mono-crystals. • Characterization of poly- and mono-crystal filters used in neutron diffraction. • Computer code developed calculates neutron cross-section, transmission & reflection. - Abstract: A detailed study about the calculation of total neutron cross-section, transmission and reflection from crystalline materials was performed. The developed computer code is approved to be sufficient for the required calculations, also an excellent agreement has been shown when comparing the code results with the other calculated and measured values. The optimal monochromator and filter parameters were discussed in terms of crystal orientation, mosaic spread, and thickness. Calculations show that 30 cm thick of PbF{sub 2} poly-crystal is an excellent cold neutron filter producing neutron wavelengths longer than 0.66 nm needed for the investigation of magnetic structure experiments. While mono-crystal filter PbF{sub 2} cut along its (1 1 1), having mosaic spread (η = 0.5°) and thickness 10 cm can only transmit thermal neutrons of the desired wavelengths and suppress epithermal and γ-rays forming unwanted background, when it is cooled to liquid nitrogen temperature. NaCl (2 0 0) and PbF{sub 2} (1 1 1) monochromator crystals having mosaic spread (η = 0.5°) and thickness 10 mm shows high neutron reflectivity for neutron wavelengths (λ = 0.114 nm and λ = 0.43 nm) when they used as a thermal and cold neutron monochromators respectively with very low contamination from higher order reflections.

  17. Calculation of self–shielding factor for neutron activation experiments using GEANT4 and MCNP

    Energy Technology Data Exchange (ETDEWEB)

    Romero–Barrientos, Jaime, E-mail: jaromero@ing.uchile.cl [Comisión Chilena de Energía Nuclear, Nueva Bilbao 12501, Las Condes, Santiago (Chile); Universidad de Chile, DFI, Facultad de Ciencias Físicas Y Matemáticas, Avenida Blanco Encalada 2008, Santiago (Chile); Molina, F. [Comisión Chilena de Energía Nuclear, Nueva Bilbao 12501, Las Condes, Santiago (Chile); Aguilera, Pablo [Comisión Chilena de Energía Nuclear, Nueva Bilbao 12501, Las Condes, Santiago (Chile); Universidad de Chile, Depto. de Física, Facultad de Ciencias, Las Palmeras 3425, Ñuñoa, Santiago (Chile); Arellano, H. F. [Universidad de Chile, DFI, Facultad de Ciencias Físicas Y Matemáticas, Avenida Blanco Encalada 2008, Santiago (Chile)

    2016-07-07

    The neutron self–shielding factor G as a function of the neutron energy was obtained for 14 pure metallic samples in 1000 isolethargic energy bins from 1·10{sup −5}eV to 2·10{sup 7}eV using Monte Carlo simulations in GEANT4 and MCNP6. The comparison of these two Monte Carlo codes shows small differences in the final self–shielding factor mostly due to the different cross section databases that each program uses.

  18. Calculations to Support On-line Neutron Spectrum Adjustment by Measurements with Miniature Fission Chambers in the JSI TRIGA Reactor

    Directory of Open Access Journals (Sweden)

    Kaiba Tanja

    2018-01-01

    Full Text Available Preliminary calculations were performed with the aim to establish optimal experimental conditions for the measurement campaign within the collaboration between the Jožef Stefan Institute (JSI and Commissariat à l’Énergie Atomique et aux Énergies Alternatives (CEA Cadarache. The goal of the project is to additionally characterize the neutron spectruminside the JSI TRIGA reactor core with focus on the measurement epi-thermal and fast part of the spectrum. Measurements will be performed with fission chambers containing different fissile materials (235U, 237Np and 242Pu covered with thermal neutron filters (Cd and Gd. The changes in the detected signal and neutron flux spectrum with and without transmission filter were studied. Additional effort was put into evaluation of the effect of the filter geometry (e.g. opening on the top end of the filter on the detector signal. After the analysis of the scoping calculations it was concluded to position the experiment in the outside core ring inside one of the empty fuel element positions.

  19. MCNP modeling of the Swiss LWRs for the calculation of the in- and ex-vessel neutron flux distributions

    Energy Technology Data Exchange (ETDEWEB)

    Pantelias, M.; Volmert, B.; Caruso, S. [National Cooperative for the Disposal of Radioactive Waste Nagra, Hardstrasse 73, 5430, Wettingen (Switzerland); Zvoncek, P. [Laboratory for Nuclear Energy Systems, ETH Zurich, Sonneggstrasse 3, 8092, Zurich (Switzerland); Bitterli, B. [Kernkraftwerk Goesgen-Daeniken AG, 4658 Daeniken (Switzerland); Neukaeter, E.; Nissen, W. [BKW FMB Energie AG-Kernkraftwerk Muehleberg, 3203 Muehleberg (Switzerland); Ledergerber, G. [Kernkraftwerk Leibstadt AG, 5325 Leibstadt (Switzerland); Vielma, R. [Axpo AG-Kernkraftwerk Beznau, 5312 Doettingen (Switzerland)

    2012-07-01

    MCNP models of all Swiss Nuclear Power Plants have been developed by the National Cooperative for the Disposal of Radioactive Waste (Nagra), in collaboration with the utilities and ETH Zurich, for the 2011 decommissioning cost study. The estimation of the residual radionuclide inventories and corresponding activity levels of irradiated structures and components following the NPP shut-down is of crucial importance for the planning of the dismantling process, the waste packaging concept and, consequently, for the estimation of the decommissioning costs. Based on NPP specific data, the neutron transport simulations lead to the best yet knowledge of the neutron spectra necessary for the ensuing activation calculations. In this paper, the modeling concept towards the MCNP-NPPs is outlined and the resulting flux distribution maps are presented. (authors)

  20. Notes on LCW Activation Calculation for Neutron Imaging Operations in the North Cave of Building 194

    Energy Technology Data Exchange (ETDEWEB)

    Anderson, S. G. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2017-09-20

    This note estimates the amount of activation that could be produced in the Facilities-provided Low Conductivity Water (LCW) that is proposed to be used for cooling of electromagnets and beam stops in the Neutron Imaging (NI) accelerator project in the North Cave of Building 194.

  1. Neutronic and thermal-hydraulic calculations for the AP-1000 NPP with the MCNP6 and SERPENT codes

    Energy Technology Data Exchange (ETDEWEB)

    Stefani, Giovanni Laranjo; Maiorino, Jose R.; Santos, Thiago A., E-mail: giovanni.laranjo@ufabc.edu.br, E-mail: joserubens.maiorino@ufabc.edu.br, E-mail: thiago.santos@ufabc.edu.br [Universidade Federal do ABC (CECS/UFABC), Santo Andre, SP (Brazil). Centro de Engenharia, Modelagem e Ciencias Sociais; Rossi, Pedro R., E-mail: pedro.russorossi@gmail.com [FERMIUM - Tecnologia Nuclear, Sao Paulo, SP (Brazil)

    2015-07-01

    The AP-1000 is an evolutionary PWR reactor designed as an evolution of the AP-600 project. The reactor is already pre-licensed by NRC, and is considered to have achieved high standards of safety, possible short construction time and good economic competitiveness. The core is a 17x17 typical assembly using Zirlo as cladding, 3 different enrichment regions, and is controlled by boron, control banks, and burnable poison. The expected fuel final burnup is 62 MWD/ton U and a cycle of 18 months. In this paper we present results for neutronic and thermal-hydraulic calculations for the AP-1000. We use the MCNP6 and SERPENT codes to calculate the first cycle of operation. The calculated parameters are K{sub eff} at BOL and EOL and its variation with burnup and neutron flux, and reactivity coefficients. The production of transuranic elements such as Pu-239 and Pu-241, and burning fuel are calculated over time. In the work a complete reactor was burned for 450 days with no control elements, boron or burnable poison were considered, these results were compared with data provided by the Westinghouse. The results are compared with those reported in the literature. A simple thermal hydraulic analysis allows verification of thermal limits such as fuel and cladding temperatures, and MDNB. (author)

  2. Air and smear sample calculational tool for Fluor Hanford Radiological control

    Energy Technology Data Exchange (ETDEWEB)

    BAUMANN, B.L.

    2003-09-24

    A spreadsheet calculation tool was developed to automate the calculations performed for determining the concentration of airborne radioactivity and smear counting as outlined in HNF-13536, Section 5.2.7, Analyzing Air and smear Samples. This document reports on the design and testing of the calculation tool.

  3. A hybrid method for coupled neutron-ion transport calculations for {sup 10}B and {sup 6}LiF coated and perforated detector efficiencies

    Energy Technology Data Exchange (ETDEWEB)

    Solomon, C.J. [Department of Mechanical and Nuclear Engineering, Kansas State University, Manhattan, KS 66506 (United States)], E-mail: cjs8888@ksu.edu; Shultis, J.K.; McNeil, W.J.; Unruh, T.C.; Rice, B.B.; McGregor, D.S. [Department of Mechanical and Nuclear Engineering, Kansas State University, Manhattan, KS 66506 (United States)

    2007-09-21

    A hybrid method for neutron-ion transport has been developed and applied to modeling coated and perforated neutron semiconductor detectors. The method couples the MCNP transport code developed by Los Alamos National Laboratory and a specialized ion transport code. Output from MCNP is passed to the ion transport code to perform ion energy deposition calculations, and in this manner the two codes are married. The process is controlled by a PERL script. Angle dependent efficiency calculation results for perforated rod detectors are presented, and neutron absorption efficiencies are presented for channel and chevron type perforations to show how problems with rod perforations may be overcome.

  4. ACDOS1: a computer code to calculate dose rates from neutron activation of neutral beamlines and other fusion-reactor components

    Energy Technology Data Exchange (ETDEWEB)

    Keney, G.S.

    1981-08-01

    A computer code has been written to calculate neutron induced activation of neutral-beam injector components and the corresponding dose rates as a function of geometry, component composition, and time after shutdown. The code, ACDOS1, was written in FORTRAN IV to calculate both activity and dose rates for up to 30 target nuclides and 50 neutron groups. Sufficient versatility has also been incorporated into the code to make it applicable to a variety of general activation problems due to neutrons of energy less than 20 MeV.

  5. Neutron spectra measurement and calculations using data libraries CIELO, JEFF-3.2 and ENDF/B-VII.1 in iron benchmark assemblies

    Science.gov (United States)

    Jansky, Bohumil; Rejchrt, Jiri; Novak, Evzen; Losa, Evzen; Blokhin, Anatoly I.; Mitenkova, Elena

    2017-09-01

    The leakage neutron spectra measurements have been done on benchmark spherical assemblies - iron spheres with diameter of 20, 30, 50 and 100 cm. The Cf-252 neutron source was placed into the centre of iron sphere. The proton recoil method was used for neutron spectra measurement using spherical hydrogen proportional counters with diameter of 4 cm and with pressure of 400 and 1000 kPa. The neutron energy range of spectrometer is from 0.1 to 1.3 MeV. This energy interval represents about 85 % of all leakage neutrons from Fe sphere of diameter 50 cm and about of 74% for Fe sphere of diameter 100 cm. The adequate MCNP neutron spectra calculations based on data libraries CIELO, JEFF-3.2 and ENDF/B-VII.1 were done. Two calculations were done with CIELO library. The first one used data for all Fe-isotopes from CIELO and the second one (CIELO-56) used only Fe-56 data from CIELO and data for other Fe isotopes were from ENDF/B-VII.1. The energy structure used for calculations and measurements was 40 gpd (groups per decade) and 200 gpd. Structure 200 gpd represents lethargy step about of 1%. This relatively fine energy structure enables to analyze the Fe resonance neutron energy structure. The evaluated cross section data of Fe were validated on comparisons between the calculated and experimental spectra.

  6. A simple model for calculating air pollution within street canyons

    Science.gov (United States)

    Venegas, Laura E.; Mazzeo, Nicolás A.; Dezzutti, Mariana C.

    2014-04-01

    This paper introduces the Semi-Empirical Urban Street (SEUS) model. SEUS is a simple mathematical model based on the scaling of air pollution concentration inside street canyons employing the emission rate, the width of the canyon, the dispersive velocity scale and the background concentration. Dispersive velocity scale depends on turbulent motions related to wind and traffic. The parameterisations of these turbulent motions include two dimensionless empirical parameters. Functional forms of these parameters have been obtained from full scale data measured in street canyons at four European cities. The sensitivity of SEUS model is studied analytically. Results show that relative errors in the evaluation of the two dimensionless empirical parameters have less influence on model uncertainties than uncertainties in other input variables. The model estimates NO2 concentrations using a simple photochemistry scheme. SEUS is applied to estimate NOx and NO2 hourly concentrations in an irregular and busy street canyon in the city of Buenos Aires. The statistical evaluation of results shows that there is a good agreement between estimated and observed hourly concentrations (e.g. fractional bias are -10.3% for NOx and +7.8% for NO2). The agreement between the estimated and observed values has also been analysed in terms of its dependence on wind speed and direction. The model shows a better performance for wind speeds >2 m s-1 than for lower wind speeds and for leeward situations than for others. No significant discrepancies have been found between the results of the proposed model and that of a widely used operational dispersion model (OSPM), both using the same input information.

  7. EBR-II Static Neutronic Calculations by PHISICS / MCNP6 codes

    Energy Technology Data Exchange (ETDEWEB)

    Paolo Balestra; Carlo Parisi; Andrea Alfonsi

    2016-02-01

    The International Atomic Energy Agency (IAEA) launched a Coordinated Research Project (CRP) on the Shutdown Heat Removal Tests (SHRT) performed in the '80s at the Experimental fast Breeder Reactor EBR-II, USA. The scope of the CRP is to improve and validate the simulation tools for the study and the design of the liquid metal cooled fast reactors. Moreover, training of the next generation of fast reactor analysts is being also considered the other scope of the CRP. In this framework, a static neutronic model was developed, using state-of-the art neutron transport codes like SCALE/PHISICS (deterministic solution) and MCNP6 (stochastic solution). Comparison between both solutions is briefly illustrated in this summary.

  8. Hybrid variational principles and synthesis method for finite element neutron transport calculations

    Energy Technology Data Exchange (ETDEWEB)

    Ackroyd, R.T. (Queen Mary Coll., London (UK). Dept. of Nuclear Engineering); Nanneh, M.M. (Imperial Coll. of Science and Technology, London (UK). Dept. of Mechanical Engineering)

    1990-01-01

    A family of hybrid variational principles is derived using a generalised least squares method. Neutron conservation is automatically satisfied for the hybrid principles employing two trial functions. No interfaces or reflection conditions need to be imposed on the independent even-parity trial function. For some hybrid principles a single trial function can be employed by relating one parity trial function to the other, using one of the parity transport equation in relaxed form. For other hybrid principles the trial functions can be employed sequentially. Synthesis of transport solutions, starting with the diffusion theory approximation, has been used as a way of reducing the scale of the computation that arises with established finite element methods for neutron transport. (author).

  9. Neutron Deep Penetration Calculations in Light Water with Monte Carlo TRIPOLI-4® Variance Reduction Techniques

    OpenAIRE

    Lee Yi-Kang

    2017-01-01

    Nuclear decommissioning takes place in several stages due to the radioactivity in the reactor structure materials. A good estimation of the neutron activation products distributed in the reactor structure materials impacts obviously on the decommissioning planning and the low-level radioactive waste management. Continuous energy Monte-Carlo radiation transport code TRIPOLI-4 has been applied on radiation protection and shielding analyses. To enhance the TRIPOLI-4 application in nuclear decomm...

  10. Calculation of the cluster size distribution functions and small-angle neutron scattering data for C60/N-methylpyrrolidone

    Science.gov (United States)

    Tropin, T. V.; Jargalan, N.; Avdeev, M. V.; Kyzyma, O. A.; Sangaa, D.; Aksenov, V. L.

    2014-01-01

    The aggregate growth in a C60/N-methylpyrrolidone (NMP) solution has been considered in the framework of the approach developed earlier for describing the cluster growth kinetics in fullerene polar solutions. The final cluster size distribution functions in model solutions have been estimated for two fullerene aggregation models including the influence of complex formation on the cluster growth using extrapolations of the characteristics of the cluster state and distribution parameters. Based on the obtained results, the model curves of small-angle neutron scattering have been calculated for a C60/NMP solution at various values of the model parameters.

  11. An unadjusted 25 group neutron cross section set for fast reactor core calculations from JENDL-2 library

    Energy Technology Data Exchange (ETDEWEB)

    Devan, K.; Gopalakrishnan, V.; Lee, S.M. [Nuclear Data Section Indira Ganhi Centre for Atomic Research, Tamilnadu (India)

    1994-12-31

    We have created a 25 group neutron cross section set (IGCJENDL) for nuclides of interest to LMFBRs from the Japanese Evaluated Nuclear Data Library - Version 2 (JENDL-2) in the format of French adjusted Cadarache Version 2 set (1969). The integral validation of IGCJENDL set was done by analyzing nine fast critical assemblies proposed by Cross Section Evaluation Working Group (CSEWG). The calculated integral parameters agreed reasonably well with the reported measured values. It is found that this set predicts the integral parameters, k-eff in particular, close to that predicted by adjusted CARNAVAL IV (French) or BNAB-78 (Russian) sets, for a 1200 MWe theoretical benchmark, representing a large power reactor.

  12. Monte Carlo calculation of large and small-angle electron scattering in air

    Science.gov (United States)

    Cohen, B. I.; Higginson, D. P.; Eng, C. D.; Farmer, W. A.; Friedman, A.; Grote, D. P.; Larson, D. J.

    2017-11-01

    A Monte Carlo method for angle scattering of electrons in air that accommodates the small-angle multiple scattering and larger-angle single scattering limits is introduced. The algorithm is designed for use in a particle-in-cell simulation of electron transport and electromagnetic wave effects in air. The method is illustrated in example calculations.

  13. Calculating emissions into the air. General methodological principles; Calcul des emissions dans l'air. Principes methodologiques generaux

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2000-05-01

    Knowing the quantities of certain substances discharged into the atmosphere is a necessary and fundamental stage in any environmental protection policy to tackle today's problems such as acid rain, the degradation of air quality, global warming and climate change, the depletion of the ozone layer, etc. This quantification, usually known as an 'emission inventory', is built on a set of specific rules which may vary from one inventory to another. This state of affairs presents the enormous disadvantage that the data available are not comparable. At the international level, an attempt at harmonization has been going on for some years between the various international bodies. This work is being pursued in parallel with the improvement of methodologies to estimate discharges from various types of source. To take account of changes in specifications and of improvements in our understanding of phenomena giving rise to atmospheric pollution, the results of inventories of emissions need to be regularly revised, even retrospectively, to maintain a consistent series. CITEPA, which acts as a National Reference Centre, has developed a system of inventories as part of the CORALIE programme with financial help from the French Ministry for Planning and the Environment. (author)

  14. Calculation of Beta Decay Half-Lives and Delayed Neutron Branching Ratio of Fission Fragments with Skyrme-QRPA

    Science.gov (United States)

    Minato, Futoshi

    2016-06-01

    Nuclear β-decay and delayed neutron (DN) emission is important for the r-process nucleosynthesis after the freeze-out, and stable and safe operation of nuclear reactors. Even though radioactive beam facilities have enabled us to measure β-decay and branching ratio of neutron-rich nuclei apart from the stability line in the nuclear chart, there are still a lot of nuclei which one cannot investigate experimentally. In particular, information on DN is rather scarce than that of T1/2. To predict T1/2 and the branching ratios of DN for next JENDL decay data, we have developed a method which comprises the quasiparticle-random-phase-approximation (QRPA) and the Hauser-Feshbach statistical model (HFSM). In this work, we calculate fission fragments with T1/2 ≤ 50 sec. We obtain the rms deviation from experimental half-life of 3:71. Although the result is still worse than GT2 which has been adopted in JENDL decay data, DN spectra are newly calculated. We also discuss further subjects to be done in future for improving the present approach and making next generation of JENDL decay data.

  15. Test tasks for verification of program codes for calculation of neutron-physical characteristics of the BN series reactors

    Science.gov (United States)

    Tikhomirov, Georgy; Ternovikh, Mikhail; Smirnov, Anton; Saldikov, Ivan; Bahdanovich, Rynat; Gerasimov, Alexander

    2017-09-01

    System of test tasks is presented with the fast reactor BN-1200 with nitride fuel as prototype. The system of test tasks includes three test based on different geometric models. Model of fuel element in homogeneous and in heterogeneous form, model of fuel assembly in height-heterogeneous and full heterogeneous form, and modeling of the active core of BN-1200 reactor. Cross-verification of program codes was performed. Transition from simple geometry to more complex one allows to identify the causes of discrepancies in the results during the early stage of cross-verification of codes. This system of tests can be applied for certification of engineering programs based on the method of Monte Carlo to the calculation of full-scale models of the reactor core of the BN series. The developed tasks take into account the basic layout and structural features of the reactor BN-1200. They are intended for study of neutron-physical characteristics, estimation of influence of heterogeneous structure and influence of diffusion approximation. The development of system of test tasks allowed to perform independent testing of programs for calculation of neutron-physical characteristics: engineering programs JARFR and TRIGEX, and codes MCU, TDMCC, and MMK based on the method of Monte Carlo.

  16. Study Neutronic of Small Pb-Bi Cooled Non-Refuelling Nuclear Power Plant Reactor (SPINNOR) with Hexagonal Geometry Calculation

    Science.gov (United States)

    Nur Krisna, Dwita; Su'ud, Zaki

    2017-01-01

    Nuclear reactor technology is growing rapidly, especially in developing Nuclear Power Plant (NPP). The utilization of nuclear energy in power generation systems has been progressing phase of the first generation to the fourth generation. This final project paper discusses the analysis neutronic one-cooled fast reactor type Pb-Bi, which is capable of operating up to 20 years without refueling. This reactor uses Thorium Uranium Nitride as fuel and operating on power range 100-500MWtNPPs. The method of calculation used a computer simulation program utilizing the SRAC. SPINNOR reactor is designed with the geometry of hexagonal shaped terrace that radially divided into three regions, namely the outermost regions with highest percentage of fuel, the middle regions with medium percentage of fuel, and most in the area with the lowest percentage. SPINNOR fast reactor operated for 20 years with variations in the percentage of Uranium-233 by 7%, 7.75%, and 8.5%. The neutronic calculation and analysis show that the design can be optimized in a fast reactor for thermal power output SPINNOR 300MWt with a fuel fraction 60% and variations of Uranium-233 enrichment of 7%-8.5%.

  17. Calculation of Beta Decay Half-Lives and Delayed Neutron Branching Ratio of Fission Fragments with Skyrme-QRPA

    Directory of Open Access Journals (Sweden)

    Minato Futoshi

    2016-01-01

    Full Text Available Nuclear β-decay and delayed neutron (DN emission is important for the r-process nucleosynthesis after the freeze-out, and stable and safe operation of nuclear reactors. Even though radioactive beam facilities have enabled us to measure β-decay and branching ratio of neutron-rich nuclei apart from the stability line in the nuclear chart, there are still a lot of nuclei which one cannot investigate experimentally. In particular, information on DN is rather scarce than that of T1/2. To predict T1/2 and the branching ratios of DN for next JENDL decay data, we have developed a method which comprises the quasiparticle-random-phase-approximation (QRPA and the Hauser-Feshbach statistical model (HFSM. In this work, we calculate fission fragments with T1/2 ≤ 50 sec. We obtain the rms deviation from experimental half-life of 3:71. Although the result is still worse than GT2 which has been adopted in JENDL decay data, DN spectra are newly calculated. We also discuss further subjects to be done in future for improving the present approach and making next generation of JENDL decay data.

  18. Validation of 3D Code KATRIN For Fast Neutron Fluence Calculation of VVER-1000 Reactor Pressure Vessel by Ex-Vessel Measurements and Surveillance Specimens Results

    Directory of Open Access Journals (Sweden)

    Dzhalandinov A.

    2016-01-01

    Full Text Available Usually the synthesis of two-dimensional and one-dimensional discrete ordinate calculations is used to evaluate neutron fluence on VVER-1000 reactor pressure vessel (RPV for prognosis of radiation embrittlement. But there are some cases when this approach is not applicable. For example the latest projects of VVER-1000 have upgraded surveillance program. Containers with surveillance specimens are located on the inner surface of RPV with fast neutron flux maximum. Therefore, the synthesis approach is not suitable enough for calculation of local disturbance of neutron field in RPV inner surface behind the surveillance specimens because of their complicated and heterogeneous structure. In some cases the VVER-1000 core loading consists of fuel assemblies with different fuel height and the applicability of synthesis approach is also ambiguous for these fuel cycles. Also, the synthesis approach is not enough correct for the neutron fluence estimation at the RPV area above core top. Because of these reasons only the 3D neutron transport codes seem to be satisfactory for calculation of neutron fluence on the VVER-1000 RPV. The direct 3D calculations are also recommended by modern regulations.

  19. Air and smear sample calculational tool for Fluor Hanford Radiological control

    Energy Technology Data Exchange (ETDEWEB)

    BAUMANN, B.L.

    2003-07-11

    A spreadsheet calculation tool was developed to automate the calculations performed for determining the concentration of airborne radioactivity and smear counting as outlined in HNF-13536, Section 5.2.7, ''Analyzing Air and Smear Samples''. This document reports on the design and testing of the calculation tool. Radiological Control Technicians (RCTs) will save time and reduce hand written and calculation errors by using an electronic form for documenting and calculating work place air samples. Current expectations are RCTs will perform an air sample and collect the filter or perform a smear for surface contamination. RCTs will then survey the filter for gross alpha and beta/gamma radioactivity and with the gross counts utilize either hand calculation method or a calculator to determine activity on the filter. The electronic form will allow the RCT with a few key strokes to document the individual's name, payroll, gross counts, instrument identifiers; produce an error free record. This productivity gain is realized by the enhanced ability to perform mathematical calculations electronically (reducing errors) and at the same time, documenting the air sample.

  20. Water management in a planar air-breathing fuel cell array using operando neutron imaging

    Science.gov (United States)

    Coz, E.; Théry, J.; Boillat, P.; Faucheux, V.; Alincant, D.; Capron, P.; Gébel, G.

    2016-11-01

    Operando Neutron imaging is used for the investigation of a planar air-breathing array comprising multiple cells in series. The fuel cell demonstrates a stable power density level of 150 mW/cm2. Water distribution and quantification is carried out at different operating points. Drying at high current density is observed and correlated to self-heating and natural convection. Working in dead-end mode, water accumulation at lower current density is largely observed on the anode side. However, flooding mechanisms are found to begin with water condensation on the cathode side, leading to back-diffusion and anodic flooding. Specific in-plane and through-plane water distribution is observed and linked to the planar array design.

  1. Are neutrons responsible for the dose discrepancies between Monte Carlo calculations and measurements in the build-up region for a high-energy photon beam?

    Energy Technology Data Exchange (ETDEWEB)

    Ding, George X. [Medical Physics, Fraser Valley Cancer Centre, British Columbia Cancer Agency, Surrey, BC (Canada)]. E-mail: gding@bccancer.bc.ca; Duzenli, Cheryl; Kalach, Nina I. [Medical Physics, Fraser Valley Cancer Centre, British Columbia Cancer Agency, Surrey, BC (Canada)

    2002-09-07

    This study presents measured neutron dose using a neutron dosimeter in a water phantom and investigates a hypothesis that neutrons in a high-energy photon beam may be responsible for the reported significant dose discrepancies between Monte Carlo calculations and measurements at the build-up region in large fields. Borated polyethylene slabs were inserted between the accelerator head and the phantom in order to remove neutrons generated in the accelerator head. The thickness of the slab ranged from 2.5 cm to 10 cm. A lead slab of 3 mm thickness was also used in the study. The superheated drop neutron dosimeter was used to measure the depth-dose curve of neutrons in a high-energy photon beam and to verify the effectiveness of the slab to remove these neutrons. Total dose measurements were performed in water using a WELLHOFER WP700 beam scanner with an IC-10 ionization chamber. The Monte Carlo code BEAM was used to simulate an 18 MV photon beam from a Varian Clinac-2100EX accelerator. Both EGS4/DOSXYZ and EGSnrc/DOSRZnrc were used in the dose calculations. Measured neutron dose equivalents as a function of depth per unit total dose in water were presented for 10x10 and 40x40 cm{sup 2} fields. The measured results have shown that a 5-10 cm thick borated polyethylene slab can reduce the neutron dose by a factor of 2 when inserted between the accelerator head and the detector. In all cases the measured neutron dose equivalent was less than 0.5% of the photon dose. In order to study if the ion chamber was highly sensitive to the neutron dose, we have investigated the disagreement between the Monte Carlo calculated and measured central-axis depth-dose curves in the build-up region when different shielding materials were used. The result indicated that the IC-10 chamber was not highly sensitive to the neutron dose. Therefore, neutrons present in a high-energy photon beam were unlikely to be responsible for the reported discrepancies in the build-up region for large fields

  2. Are neutrons responsible for the dose discrepancies between Monte Carlo calculations and measurements in the build-up region for a high-energy photon beam?

    Science.gov (United States)

    Ding, George X; Duzenli, Cheryl; Kalach, Nina I

    2002-09-07

    This study presents measured neutron dose using a neutron dosimeter in a water phantom and investigates a hypothesis that neutrons in a high-energy photon beam may be responsible for the reported significant dose discrepancies between Monte Carlo calculations and measurements at the build-up region in large fields. Borated polyethylene slabs were inserted between the accelerator head and the phantom in order to remove neutrons generated in the accelerator head. The thickness of the slab ranged from 2.5 cm to 10 cm. A lead slab of 3 mm thickness was also used in the study. The superheated drop neutron dosimeter was used to measure the depth-dose curve of neutrons in a high-energy photon beam and to verify the effectiveness of the slab to remove these neutrons. Total dose measurements were performed in water using a WELLHOFER WP700 beam scanner with an IC-10 ionization chamber. The Monte Carlo code BEAM was used to simulate an 18 MV photon beam from a Varian Clinac-2100EX accelerator. Both EGS4/DOSXYZ and EGSnrc/DOSRZnrc were used in the dose calculations. Measured neutron dose equivalents as a function of depth per unit total dose in water were presented for 10 x 10 and 40 x 40 cm2 fields. The measured results have shown that a 5-10 cm thick borated polyethylene slab can reduce the neutron dose by a factor of 2 when inserted between the accelerator head and the detector. In all cases the measured neutron dose equivalent was less than 0.5% of the photon dose. In order to study if the ion chamber was highly sensitive to the neutron dose, we have investigated the disagreement between the Monte Carlo calculated and measured central-axis depth-dose curves in the build-up region when different shielding materials were used. The result indicated that the IC-10 chamber was not highly sensitive to the neutron dose. Therefore, neutrons present in a high-energy photon beam were unlikely to be responsible for the reported discrepancies in the build-up region for large fields.

  3. Transport theory calculation for a heterogeneous multi-assembly problem by characteristics method with direct neutron path linking technique

    Energy Technology Data Exchange (ETDEWEB)

    Kosaka, Shinya; Saji, Etsuro [In-Core Fuel Management System Department, Toden Software, Inc., Tokyo (Japan)

    2000-12-01

    A characteristics transport theory code, CHAPLET, has been developed for the purpose of making it practical to perform a whole LWR core calculation with the same level of calculational model and accuracy as that of an ordinary single assembly calculation. The characteristics routine employs the CACTUS algorithm for drawing ray tracing lines, which assists the two key features of the flux solution in the CHAPLET code. One is the direct neutron path linking (DNPL) technique which strictly connects angular fluxes at each assembly interface in the flux solution separated between assemblies. Another is to reduce the required memory storage by sharing the data related to ray tracing among assemblies with the same configuration. For faster computation, the coarse mesh rebalance (CMR) method and the Aitken method were incorporated in the code and the combined use of both methods showed the most promising acceleration performance among the trials. In addition, the parallelization of the flux solution was attempted, resulting in a significant reduction in the wall-clock time of the calculation. By all these efforts, coupled with the results of many verification studies, a whole LWR core heterogeneous transport theory calculation finally became practical. CHAPLET is thought to be a useful tool which can produce the reference solutions for analyses of an LWR (author)

  4. Neutron/gamma coupled library generation and gamma transport calculation with KARMA 1.2

    Energy Technology Data Exchange (ETDEWEB)

    Hong, S. G. [Dept. of Nuclear Engineering, Kyung Hee Univ., 446-701 Deogyeong-daero, GiHeung-gu, Yongin, Gyeonggi-do (Korea, Republic of); Kim, K. S.; Cho, J. Y.; Lee, K. H. [Korea Atomic Energy Research Inst., 305-353 Duckjin-dong, Yuseong-gu, Daejon (Korea, Republic of)

    2012-07-01

    KAERI has developed a lattice transport calculation code KARMA and its multi-group cross section library generation system. Recently, the multi-group cross section library generation system has included a gamma cross section generation capability and KARMA also has been improved to include a gamma transport calculation module. This paper addresses the multi-group gamma cross section generation capability for the KARMA 1.2 code and the preliminary test results of the KARMA 1.2 gamma transport calculations. The gamma transport calculation with KARMA 1.2 gives the gamma flux, gamma smeared power, and gamma energy deposition distributions. The results of the KARMA gamma calculations were compared with those of HELIOS and they showed that KARMA 1.2 gives reasonable gamma transport calculation results. (authors)

  5. COMPUTER PROGRAM FOR CALCULATION MICROCHANNEL HEAT EXCHANGERS FOR AIR CONDITIONING SYSTEMS

    Directory of Open Access Journals (Sweden)

    Olga V. Olshevska

    2016-08-01

    Full Text Available Creating a computer program to calculate microchannel air condensers to reduce design time and carrying out variant calculations. Software packages for thermophysical properties of the working substance and the coolant, the correlation equation for calculating heat transfer, aerodynamics and hydrodynamics, the thermodynamic equations for the irreversible losses and their minimization in the heat exchanger were used in the process of creating. Borland Delphi 7 is used for creating software package.

  6. Simplified geometric model for the calculation of neutron yield in an accelerator of 18 MV for radiotherapy; Modelo geometrico simplificado para el calculo del rendimiento de neutrones en un acelerador de 18 MV para radioterapia

    Energy Technology Data Exchange (ETDEWEB)

    Paredes G, L.C.; Balcazar G, M. [ININ, 52750 La Marquesa, Estado de Mexico (Mexico); Francois L, J.L. [FI-UNAM, 04510 Mexico D.F. (Mexico); Azorin N, J. [UAM-I, 09340 Mexico D.F. (Mexico)

    2008-07-01

    The results of the neutrons yield in different components of the bolster of an accelerator Varian Clinac 2100C of 18 MV for radiotherapy are presented, which contribute to the radiation of flight of neutrons in the patient and bolster planes. For the calculation of the neutrons yield, a simplified geometric model of spherical cell for the armor-plating of the bolster with Pb and W was used. Its were considered different materials for the Bremsstrahlung production and of neutrons produced through the photonuclear reactions and of electro disintegration, in function of the initial energy of the electron. The theoretical result of the total yield of neutrons is of 1.17x10{sup -3} n/e, considering to the choke in position of closed, in the patient plane with a distance source-surface of 100 cm; of which 15.73% corresponds to the target, 58.72% to the primary collimator, 4.53% to the levelled filter of Fe, 4.87% to the levelled filter of Ta and 16.15% to the closed choke. For an initial energy of the electrons of 18 MeV, a half energy of the neutrons of 2 MeV was obtained. The calculated values for radiation of experimental neutrons flight are inferior to the maxima limit specified in the NCRP-102 and IEC-60601-201.Ed.2.0 reports. The absorbed dose of neutrons determined through the measurements with TLD dosemeters in the isocenter to 100 cm of the target when the choke is closed one, is approximately 3 times greater that the calculated for armor-plating of W and 1.9 times greater than an armor-plating of Pb. (Author)

  7. Numerical calculation of transverse coupling impedances: Comparison to Spallation Neutron Source extraction kicker measurements

    Directory of Open Access Journals (Sweden)

    B. Doliwa

    2007-10-01

    Full Text Available The study of beam dynamics and the localization of potential sources of instabilities are important tasks in the design of modern, high-intensity particle accelerators. In the case of synchrotrons and storage rings, coupling impedance data are needed to characterize the parasitic interaction of critical components with the beam. In this article we demonstrate the application of numerical field simulations to the computation of transverse kicker coupling impedances. Based on the 3D simulation results, a parametrized model is developed to incorporate the impedance of an arbitrary pulse-forming network attached to the kicker. Detailed comparisons of numerical results with twin-wire and direct measurements are discussed at the example of the Spallation Neutron Source extraction kicker.

  8. Numerical calculation of transverse coupling impedances: Comparison to Spallation Neutron Source extraction kicker measurements

    Science.gov (United States)

    Doliwa, B.; Arévalo, E.; Weiland, T.

    2007-10-01

    The study of beam dynamics and the localization of potential sources of instabilities are important tasks in the design of modern, high-intensity particle accelerators. In the case of synchrotrons and storage rings, coupling impedance data are needed to characterize the parasitic interaction of critical components with the beam. In this article we demonstrate the application of numerical field simulations to the computation of transverse kicker coupling impedances. Based on the 3D simulation results, a parametrized model is developed to incorporate the impedance of an arbitrary pulse-forming network attached to the kicker. Detailed comparisons of numerical results with twin-wire and direct measurements are discussed at the example of the Spallation Neutron Source extraction kicker.

  9. Inelastic neutron scattering and lattice dynamical calculation of negative thermal expansion in HfW2O8

    Science.gov (United States)

    Mittal, R.; Chaplot, S. L.; Kolesnikov, A. I.; Loong, C.-K.; Mary, T. A.

    2003-08-01

    The compounds ZrW2O8 and HfW2O8 undergo large isotropic negative thermal expansion (NTE) over a wide range of temperatures up to 1443 K and 1050 K, respectively. We have showed previously that large softening of low-energy phonons in ZrW2O8 is responsible for its anomalous thermal expansion behavior. In order to understand the effect of replacing Zr by Hf on NTE behavior we report lattice dynamical calculations and neutron time-of-flight spectroscopic measurements of the phonon density of states for cubic HfW2O8. The calculated phonon spectrum for cubic HfW2O8 is in fair agreement with the experimental data. The phonon spectra in the Zr and Hf compounds differ at low energies largely due to the mass difference. The calculated negative thermal expansion for HfW2O8 is in good agreement with experimental data from the literature. We further report a calculation of the pressure dependence of the detailed phonon dispersion relation which reveals large softening of several phonon branches on compression associated with the NTE.

  10. Experimental validation of calculations of decay heat induced by 14 MeV neutron activation of ITER materials

    Energy Technology Data Exchange (ETDEWEB)

    Taylor, N.P.; Bartels, H.-W.; Saji, G. [ITER Joint Central Team, San Diego, La Jolla (United States); Ikeda, Y.; Maekawa, F. [Japan Atomic Energy Research Institute, Tokai-mura (Japan); Cambi, G. [Physics Department, Bologna University, Bologna (Italy); Cepraga, D.G. [ENEA, INN-FIS-MACO, Bologna (Italy); Cheng, E.T. [TSI Research Inc., Solana Beach, CA (United States); Forrest, R.A.; Sublet, J.-C. [Euratom/UKAEA Fusion Association, Culham Science Centre, Abingdon, Oxfordshire OX14 3DB (United Kingdom); Iida, H.; Santoro, R.T. [ITER Joint Central Team, Garching (Germany); Khater, H.Y.; Sawan, M.E. [Fusion Technology Institute, University of Wisconsin, Madison, WI (United States); Youssef, M.Z. [Mechanical and Aerospace Engineering Department, University of California at Los Angeles, Los Angeles, CA (United States)

    1999-03-01

    In a D-T fusion tokamak, neutron activation of structural material will result in a source of heat from radioactive decay after shutdown of the plasma. Although relatively small, in certain postulated loss-of-cooling accidents this decay heat could drive a temperature transient of modest proportions. In safety studies of the International Thermonuclear Experimental Reactor (ITER), such accident scenarios are analysed for their possible consequences, requiring an accurate knowledge of the level and time-dependence of the decay heat in components in which it is significant. This is calculated by activation modelling using computer codes with libraries of nuclear data. This paper describes an international benchmark activity to validate several of these codes and data, using the results of direct measurements of decay heat in ITER-relevant materials after exposure to a flux of 14 MeV neutrons. The results show consistency between the codes and, when using data from the recently-compiled FENDL/A-2 activation data library, good agreement with the experiments over the time-scales of interest in ITER safety studies. (orig.) 24 refs.

  11. Validation of iron nuclear data for the neutron calculation of nuclear reactors; Validation de donnees nucleaires du fer pour le calcul neutronique des reacteurs nucleaires

    Energy Technology Data Exchange (ETDEWEB)

    Vaglio-Gaudard, C.

    2010-03-29

    The GEN-III and GEN-IV reactors will be equipped with heavy reflectors. However, the existing integral validation of the iron nuclear data in the latest JEFF3 European library in the frame of the neutron calculation of the heavy reflector is very partial: some results exist concerning fast reactors but there is no result corresponding to the LWR heavy reflector. No clear trend on the JEFF3 iron cross sections was brought into evidence up to now for fission reactor calculations. Iron nuclear data were completely re-evaluated in the JEFF3 library. Despite the fact that iron is widely used in the nuclear industry, large uncertainties are still associated with its nuclear data, particularly its inelastic cross section which is very important in the neutron slowing down. A validation of {sup 56}Fe nuclear data was performed on the basis of the analysis of integral experiments. Two major critical experiments, the PERLE experiment and the Gas Benchmark, were interpreted with 3D reference Monte-Carlo calculations and the JEFF3.1.1 library. The PERLE experiment was recently performed in the EOLE zero-power facility (CEA Cadarache). This experiment is dedicated to heavy reflector physics in GEN-III light water reactors. It was especially conceived for the validation of iron nuclear data. The Gas Benchmark is representative of a Gas Fast Reactor with a stainless steel reflector (with no fertile blanket) in the MASURCA facility (CEA Cadarache). Radial traverses of reaction rates were measured to characterize flux attenuation at various energies in the reflector. The results of the analysis of both experiments show good agreement between the calculations and the measurements, which is confirmed by the analysis of complementary experiments (ZR-6M, MISTRAL4, CIRANO-ZONA2B). A process of re-estimating the {sup 56}Fe nuclear data was implemented on the basis of feedback from these two experiments and the RDN code. This code relies on a non-linear regression method using an

  12. Polarized neutron imaging and three-dimensional calculation of magnetic flux trapping in bulk of superconductors

    Science.gov (United States)

    Treimer, Wolfgang; Ebrahimi, Omid; Karakas, Nursel; Prozorov, Ruslan

    2012-05-01

    Polarized neutron radiography was used to study the three-dimensional magnetic flux distribution inside of single-crystal and polycrystalline Pb cylinders with large (cm3) volume and virtually zero demagnetization. Experiments with single crystals being in the Meissner phase (T

  13. Modeling Neutron Star and Galactic Black Hole Emission: The Impact of Proper Extinction Calculations

    Science.gov (United States)

    Smith, Randall K.; Valencic, Lynne A.; Corrales, Lia

    2016-04-01

    Interstellar extinction includes both absorption and scattering of photons from interstellar gas and dust grains, and it has the effect of altering a source's spectrum and its total observed intensity. However, while multiple absorption models exist, there are no useful scattering models in standard X-ray spectrum fitting tools, such as XSPEC. Nonetheless, X-ray halos, created by scattering from dust grains, are detected around even moderately absorbed sources and the impact on an observed source spectrum can be significant, if modest, compared to direct absorption. By convolving the scattering cross section with dust models, we have created a spectral model as a function of energy, type of dust, and extraction region that can be used with models of direct absorption. This will ensure the extinction model is consistent and enable direct connections to be made between a source's X-ray spectral fits and its UV/optical extinction. I will present the model and show its impact on a range of Galactic sources including neutron stars and black holes.

  14. A FIRST APPROXIMATION CALCULATION OF AIR CUSHION CHASSIS WEIGHT OF TRANSPORT AIRPLANE

    Directory of Open Access Journals (Sweden)

    2016-01-01

    Full Text Available This article describes a first approximation of a weighted estimate of air cushion chassis. The algorithm for calculating the weight of air cushion chassis allows not only to estimate the mass of the chassis to a first approximation, but also to conduct a preliminary analysis of the influence of various parameters of the aircraft and the chassis on the weight of the aircraft at the stage of before designing. The algorithm can be expanded to include additional design decisions, such as the transformation of the fuselage, increasing the air cushion chassis canopy due to extensions, center of gravity, etc.

  15. Implementation and training methodology of subcritical reactors neutronic calculations triggered by external neutron source and applications; Implementacao e qualificacao de metodologia de calculos neutronicos em reatores subcriticos acionados por fonte externa de neutrons e aplicacoes

    Energy Technology Data Exchange (ETDEWEB)

    Carluccio, Thiago

    2011-07-01

    This works had as goal to investigate calculational methodologies on subcritical source driven reactor, such as Accelerator Driven Subcritical Reactor (ADSR) and Fusion Driven Subcritical Reactor (FDSR). Intense R and D has been done about these subcritical concepts, mainly due to Minor Actinides (MA) and Long Lived Fission Products (LLFP) transmutation possibilities. In this work, particular emphasis has been given to: (1) complement and improve calculation methodology with neutronic transmutation and decay capabilities and implement it computationally, (2) utilization of this methodology in the Coordinated Research Project (CRP) of the International Atomic Energy Agency Analytical and Experimental Benchmark Analysis of ADS and in the Collaborative Work on Use of Low Enriched Uranium in ADS, especially in the reproduction of the experimental results of the Yalina Booster subcritical assembly and study of a subcritical core of IPEN / MB-01 reactor, (3) to compare different nuclear data libraries calculation of integral parameters, such as k{sub eff} and k{sub src}, and differential distributions, such as spectrum and flux, and nuclides inventories and (4) apply the develop methodology in a study that may help future choices about dedicated transmutation system. The following tools have been used in this work: MCNP (Monte Carlo N particle transport code), MCB (enhanced version of MCNP that allows burnup calculation) and NJOY to process nuclear data from evaluated nuclear data files. (author)

  16. Calculation and comparisons with measurement of fast neutron fluxes in the material testing facilities of the NRU research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Leung, T.C. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

    2012-07-01

    The NRU reactor at Chalk River provides three irradiation facilities to study the effects of fast neutrons (E> 1 MeV) on reactor materials for assessing material damage and deformation. The facilities comprise two types of fast neutron rods (Mark 4 and Mark 7), and a Material Test Bundle (MTB) irradiated in a loop site. This paper describes the neutronic simulation of these testing facilities using the WIMS-AECL and TRIAD codes, and comparisons with the fast neutron flux measurements using iron-wire activation techniques. It also provides comparisons of flux levels, neutron spectra, and size limitations of the experimental cavities between these test facilities. (author)

  17. Calculation of dose coefficients for radionuclides produced in a spallation neutron source utilizing NUBASE and the evaluated nuclear structure data file databases.

    Science.gov (United States)

    Shanahan, J; Eckerman, K; Arndt, A; Gold, C; Patton, P; Rudin, M; Brey, R; Gesell, T; Rusetski, V; Pagava, S

    2006-01-01

    Based on a mercury spallation neutron source target, the UNLV Transmutation Research Program has identified 72 radionuclides with a half-life greater than or equal to a minute as lacking an appropriate reference for a published dose coefficient according to existing radiation safety dose coefficient databases. A method was developed to compare the nuclear data presented in the ENSDF and NUBASE databases for these 72 radionuclides. Due to conflicting or lacking nuclear data in one or more of the databases, internal and external dose coefficient values have been calculated for only 14 radionuclides, which are not currently presented in Federal Guidance Reports Nos. 11, 12, and 13 or Publications 68 and 72 of the International Commission on Radiological Protection. Internal dose coefficient values are reported for inhalation and ingestion of 1 microm and 5 microm AMAD particulates along with the f1 values and absorption types for the adult worker. Internal dose coefficient values are also reported for inhalation and ingestion of 1 microm AMAD particulates as well as the f1 values and absorption types for members of the public. Additionally, external dose coefficient values for air submersion, exposure to contaminated ground surface, and exposure to soil contaminated to an infinite depth are also presented.

  18. Preliminary calculations for the CAFE project (Clean Air For Europe); Calculs preparatoires pour la strategie thematique CAFE (Clean Air For Europe)

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2005-09-15

    The European Commission decided in 2001 an analysis program to reduce the atmospheric emissions. This report presents different limit scenari for France in 2020 (the reference scenari and the MTFR scenari, Maximum Technically Feasible Reduction), optimized scenari calculated by the RAINS model (Regional Air Pollution Information and Simulation), the costs of the scenari calculated with RAINS and the cost-benefit analysis of the strategy CAFE. From the study results, the benefits are higher than the costs, even with the most ambitious scenari. At an european level the emission reduction strategies have no effect on the employment but an impact on the Gross Domestic Product (decrease between 0,04 % and 0,12 % in function of the scenari). (A.L.B.)

  19. Neutron shielding for a {sup 252} Cf source

    Energy Technology Data Exchange (ETDEWEB)

    Vega C, H.R.; Manzanares A, E.; Hernandez D, V.M. [Unidades Academicas de Estudios Nucleares e Ingenieria Electrica, Universidad Autonoma de Zacatecas, C. Cipres 10, Fracc. La Penuela, 98068 Zacatecas (Mexico); Eduardo Gallego, Alfredo Lorente [Depto. de Ingenieria Nuclear, ETS Ingenieros Industriales, Universidad Politecnica de Madrid, C. Jose Gutierrez Abascal 2, 28006 Madrid (Spain)]. e-mail: fermineutron@yahoo.com

    2006-07-01

    To determine the neutron shielding features of water-extended polyester a Monte Carlo study was carried out. Materials with low atomic number are predominantly used for neutron shielding because these materials effectively attenuate neutrons, mainly through inelastic collisions and absorption reactions. During the selection of materials to design a neutron shield, prompt gamma production as well as radionuclide production induced by neutron activation must be considered. In this investigation the Monte Carlo method was used to evaluate the performance of a water-extended polyester shield designed for the transportation, storage, and use of a {sup 252}Cf isotopic neutron source. During calculations a detailed model for the {sup 252}Cf and the shield was utilized. To compare the shielding features of water extended polyester, the calculations were also made for the bare {sup 252}Cf in vacuum, air and the shield filled with water. For all cases the calculated neutron spectra was utilized to determine the ambient equivalent neutron dose at four sites around the shielding. In the case of water extended polyester and water shielding the calculations were extended to include the prompt gamma rays produced during neutron interactions, with this information the Kerma in air was calculated at the same locations where the ambient equivalent neutron dose was determined. (Author)

  20. Calculation of the dynamics of compression plasma flows in a magnetoplasma compressor operating on air

    Science.gov (United States)

    Ananin, S. I.

    1986-03-01

    A two-dimensional model for describing the dynamics of partially ionized air compression plasma flows, generated by a magnetoplasma compressor, is proposed. The model is based on the large-particle method together with the introduction of a magnetic field. The results of calculations of the dynamics and structure of compression plasma flows are studied and compared with experimental data.

  1. Calculation of the dynamics of compression plasma flows in a magnetoplasma compressor operating on air

    Energy Technology Data Exchange (ETDEWEB)

    Ananin, S.I.

    1986-03-01

    This paper presents a two-dimensional model for describing the dynamics of partially ionized air compression plasma flows, generated by a magnetoplasma compressor. The model is based on the large-particle method together with the introduction of a magnetic field. The results of calculations of the dynamics and structure of compression plasma flows are studied and compared with experimental data.

  2. Soil Moisture Estimation Across Scales with Mobile Sensors for Cosmic-Ray Neutrons from the Ground and Air

    Science.gov (United States)

    Schrön, Martin; Köhler, Mandy; Bannehr, Lutz; Köhli, Markus; Fersch, Benjamin; Rebmann, Corinna; Mai, Juliane; Cuntz, Matthias; Kögler, Simon; Schröter, Ingmar; Wollschläger, Ute; Oswald, Sascha; Dietrich, Peter; Zacharias, Steffen

    2016-04-01

    Soil moisture is a key variable for environmental sciences, but its determination at various scales and depths is still an open challenge. Cosmic-ray neutron sensing has become a well accepted and unique method to monitor an effective soil water content, covering tens of hectares in area and tens of centimeters in depth. The technology is famous for its low maintanance, non-invasiveness, continous measurement, and most importantly its large footprint and penetration depth. Beeing more representative than point data, and finer resolved plus deeper penetrating than remote-sensing products, cosmic-ray neutron derived soil moisture products provide unrivaled advantage for agriculture, regional hydrologic and land surface models. The method takes advantage of omnipresent neutrons which are extraordinarily sensitive to hydrogen in soil, plants, snow and air. Unwanted hydrogen sources in the footprint can be excluded by local calibration to extract the pure soil water information. However, this procedure is not feasible for mobile measurements, where neutron detectors are mounted on a car to do catchment-scale surveys. As a solution to that problem, we suggest strategies to correct spatial neutron data with the help of available spatial data of soil type, landuse and vegetation. We further present results of mobile rover campaigns at various scales and conditions, covering small sites from 0.2 km2 to catchments of 100 km2 area, and complex terrain from agricultural fields, urban areas, forests, to snowy alpine sites. As the rover is limited to accessible roads, we further investigated the applicability of airborne measurements. First tests with a gyrocopter at 150 to 200m heights proofed the concept of airborne neutron detection for environmental sciences. Moreover, neutron transport simulations confirm an improved areal coverage during these campaigns. Mobile neutron measurements at the ground or air are a promising tool for the detection of water sources across many

  3. Dose calculation in biological samples in a mixed neutron-gamma field at the TRIGA reactor of the University of Mainz.

    Science.gov (United States)

    Schmitz, Tobias; Blaickner, Matthias; Schütz, Christian; Wiehl, Norbert; Kratz, Jens V; Bassler, Niels; Holzscheiter, Michael H; Palmans, Hugo; Sharpe, Peter; Otto, Gerd; Hampel, Gabriele

    2010-10-01

    To establish Boron Neutron Capture Therapy (BNCT) for non-resectable liver metastases and for in vitro experiments at the TRIGA Mark II reactor at the University of Mainz, Germany, it is necessary to have a reliable dose monitoring system. The in vitro experiments are used to determine the relative biological effectiveness (RBE) of liver and cancer cells in our mixed neutron and gamma field. We work with alanine detectors in combination with Monte Carlo simulations, where we can measure and characterize the dose. To verify our calculations we perform neutron flux measurements using gold foil activation and pin-diodes. Material and methods. When L-α-alanine is irradiated with ionizing radiation, it forms a stable radical which can be detected by electron spin resonance (ESR) spectroscopy. The value of the ESR signal correlates to the amount of absorbed dose. The dose for each pellet is calculated using FLUKA, a multipurpose Monte Carlo transport code. The pin-diode is augmented by a lithium fluoride foil. This foil converts the neutrons into alpha and tritium particles which are products of the (7)Li(n,α)(3)H-reaction. These particles are detected by the diode and their amount correlates to the neutron fluence directly. Results and discussion. Gold foil activation and the pin-diode are reliable fluence measurement systems for the TRIGA reactor, Mainz. Alanine dosimetry of the photon field and charged particle field from secondary reactions can in principle be carried out in combination with MC-calculations for mixed radiation fields and the Hansen & Olsen alanine detector response model. With the acquired data about the background dose and charged particle spectrum, and with the acquired information of the neutron flux, we are capable of calculating the dose to the tissue. Conclusion. Monte Carlo simulation of the mixed neutron and gamma field of the TRIGA Mainz is possible in order to characterize the neutron behavior in the thermal column. Currently we also

  4. A comparison of measured and calculated values of air kerma rates from 137Cs in soil

    Directory of Open Access Journals (Sweden)

    V. P. Ramzaev

    2015-01-01

    Full Text Available In 2010, a study was conducted to determine the air gamma dose rate from 137Cs deposited in soil. The gamma dose rate measurements and soil sampling were performed at 30 reference plots from the south-west districts of the Bryansk region (Russia that had been heavily contaminated as a result of the Chernobyl accident. The 137Cs inventory in the top 20 cm of soil ranged from 260 kBq m–2 to 2800 kBq m–2. Vertical distributions of 137Cs in soil cores (6 samples per a plot were determined after their sectioning into ten horizontal layers of 2 cm thickness. The vertical distributions of 137Cs in soil were employed to calculate air kerma rates, K, using two independent methods proposed by Saito and Jacob [Radiat. Prot. Dosimetry, 1995, Vol. 58, P. 29–45] and Golikov et al. [Contaminated Forests– Recent Developments in Risk Identification and Future Perspective. Kluwer Academic Publishers, 1999. – P. 333–341]. A very good coincidence between the methods was observed (Spearman’s rank coefficient of correlation = 0.952; P<0.01; on average, a difference between the kerma rates calculated with two methods did not exceed 3%. The calculated air kerma rates agreed with the measured dose rates in air very well (Spearman’s coefficient of correlation = 0.952; P<0.01. For large grassland plots (n=19, the measured dose rates were on average 6% less than the calculated kerma rates. The tested methods for calculating the air dose rate from 137Cs in soil can be recommended for practical studies in radiology and radioecology. 

  5. Neutron-gamma flux and dose calculations for feasibility study of DISCOMS instrumentation in case of severe accident in a GEN 3 reactor

    Directory of Open Access Journals (Sweden)

    Brovchenko Mariya

    2017-01-01

    Full Text Available The present paper presents the study carried out in the frame of the DISCOMS project, which stands for “DIstributed Sensing for COrium Monitoring and Safety”. This study concerns the calculation of the neutron and gamma radiations received by the considered instrumentation during the normal reactor operation as well as in case of a severe accident for the EPR reactor, outside the reactor pressure vessel and in the containment basemat. This paper summarizes the methods and hypotheses used for the particle transport simulation outside the vessel during normal reactor operation. The results of the simulations are then presented including the responses for distributed Optical Fiber Sensors (OFS, such as the gamma dose and the fast neutron fluence, and for Self Powered Neutron Detectors (SPNDs, namely the neutron and gamma spectra. Same responses are also evaluated for severe accident situations in order to design the SPNDs being sensitive to the both types of received neutron-gamma radiation. By contrast, fibers, involved as transducers in distributed OFS have to resist to the total radiation gamma dose and neutron fluence received during normal operation and the severe accident.

  6. Neutron-gamma flux and dose calculations for feasibility study of DISCOMS instrumentation in case of severe accident in a GEN 3 reactor

    Science.gov (United States)

    Brovchenko, Mariya; Duhamel, Isabelle; Dechenaux, Benjamin

    2017-09-01

    The present paper presents the study carried out in the frame of the DISCOMS project, which stands for "DIstributed Sensing for COrium Monitoring and Safety". This study concerns the calculation of the neutron and gamma radiations received by the considered instrumentation during the normal reactor operation as well as in case of a severe accident for the EPR reactor, outside the reactor pressure vessel and in the containment basemat. This paper summarizes the methods and hypotheses used for the particle transport simulation outside the vessel during normal reactor operation. The results of the simulations are then presented including the responses for distributed Optical Fiber Sensors (OFS), such as the gamma dose and the fast neutron fluence, and for Self Powered Neutron Detectors (SPNDs), namely the neutron and gamma spectra. Same responses are also evaluated for severe accident situations in order to design the SPNDs being sensitive to the both types of received neutron-gamma radiation. By contrast, fibers, involved as transducers in distributed OFS have to resist to the total radiation gamma dose and neutron fluence received during normal operation and the severe accident.

  7. Improved Accuracy of Density Functional Theory Calculations for CO2 Reduction and Metal-Air Batteries

    DEFF Research Database (Denmark)

    Christensen, Rune; Hansen, Heine Anton; Vegge, Tejs

    2015-01-01

    .e. the electrocatalytic reduction of CO2 and metal-air batteries. In theoretical studies of electrocatalytic CO2 reduction, calculated DFT-level enthalpies of reaction for CO2reduction to various products are significantly different from experimental values[1-3]. In theoretical studies of metal-air battery reactions...... errors in DFT-level computational electrocatalytic CO2reduction is hence identified. The new insight adds increased accuracy e.g., for reaction to formic acid, where the experimental enthalpy of reaction is 0.15 eV. Previously, this enthalpy has been calculated without and with correctional approaches......, Nano Lett., 14, 1016 (2014) [6] J. Wellendorff, K. T. Lundgaard, A. Møgelhøj, V. Petzold, D. D. Landis, J. K. Nørskov, T. Bligaard, and K. W. Jacobsen, Phys. Rev. B, 85, 235149 (2012) Figure 1: Calculated enthalpies of reaction from CO2 to CH3OH (x axis) and HCOOH (y axis). Functional variations...

  8. Li-ion conduction in the LiBH4:LiI system from Density Functional Theory calculations and Quasi-Elastic Neutron Scattering

    DEFF Research Database (Denmark)

    Myrdal, Jon Steinar Gardarsson; Blanchard, Didier; Sveinbjörnsson, Dadi Þorsteinn

    2013-01-01

    to quasi-elastic neutron scattering (QENS) measurements with and without an applied bias potential of 3 V. DFT calculations show that lithium defects such as Frenkel pairs are easily formed at room temperature (formation energy of 0.44 eV) and low energy barriers (0.2 to 0.3 eV) are found between stable...

  9. Reaction Cross Section Calculations in Neutron Induced Reactions and GEANT4 Simulation of Hadronic Interactions for the Reactor Moderator Material BeO

    Directory of Open Access Journals (Sweden)

    Veli ÇAPALI

    2016-05-01

    Full Text Available BeO is one of the most common moderator material for neutron moderation; due to its high density, neutron capture cross section and physical-chemical properties that provides usage at elevated temperatures. As it’s known, for various applications in the field of reactor design and neutron capture, reaction cross–section data are required. The cross–sections of (n,α, (n,2n, (n,t, (n,EL and (n,TOT reactions for 9Be and 16O nuclei have been calculated by using TALYS 1.6 Two Component Exciton model and EMPIRE 3.2 Exciton model in this study. Hadronic interactions of low energetic neutrons and generated isotopes–particles have been investigated for a situation in which BeO was used as a neutron moderator by using GEANT4, which is a powerful simulation software. In addition, energy deposition along BeO material has been obtained. Results from performed calculations were compared with the experimental nuclear reaction data exist in EXFOR.

  10. Calculating osmotic pressure of xylitol solutions from molality according to UNIFAC model and measuring it with air humidity osmometry.

    Science.gov (United States)

    Yu, Lan; Zhan, Tingting; Zhan, Xiancheng; Wei, Guocui; Tan, Xiaoying; Wang, Xiaolan; Li, Chengrong

    2014-11-01

    The osmotic pressure of xylitol solution at a wide concentration range was calculated according to the UNIFAC model and experimentally determined by our newly reported air humidity osmometry. The measurements from air humidity osmometry were compared with UNIFAC model calculations from dilute to saturated solution. Results indicate that air humidity osmometry measurements are comparable to UNIFAC model calculations at a wide concentration range by two one-sided test and multiple testing corrections. The air humidity osmometry is applicable to measure the osmotic pressure and the osmotic pressure can be calculated from the concentration.

  11. Evaluation of lower flammability limits of fuel-air-diluent mixtures using calculated adiabatic flame temperatures.

    Science.gov (United States)

    Vidal, M; Wong, W; Rogers, W J; Mannan, M S

    2006-03-17

    The lower flammability limit (LFL) of a fuel is the minimum composition in air over which a flame can propagate. Calculated adiabatic flame temperatures (CAFT) are a powerful tool to estimate the LFL of gas mixtures. Different CAFT values are used for the estimation of LFL. SuperChems is used by industry to perform flammability calculations under different initial conditions which depends on the selection of a threshold temperature. In this work, the CAFT at the LFL is suggested for mixtures of fuel-air and fuel-air-diluents. These CAFT can be used as the threshold values in SuperChems to calculate the LFL. This paper discusses an approach to evaluate the LFL in the presence of diluents such as N2 and CO2 by an algebraic method and by the application of SuperChems using CAFT as the basis of the calculations. The CAFT for different paraffinic and unsaturated hydrocarbons are presented as well as an average value per family of chemicals.

  12. First Principles Calculations of Electronic and Thermal Properties of AIRE (RE = La, Ce and Pr) Compounds

    Science.gov (United States)

    Srivastava, Vipul; Aynyas, M.; Rajagopalan, M.; Sanyal, S. P.

    2008-04-01

    Electronic properties of non-magnetic cubic B2-type AIRE (RE = La, Ce and Pr) compounds have been derived from self-consistent tight binding linear muffin tin orbital method at ambient pressure. These compounds show metallic behaviour under ambient conditions. While thermal properties like Debye temperature and Grüneisen constant are calculated at T = 0 K within the Debye-Grüneisen model and compared with the others theoretical results. We have also performed a pressure induced variation of Debye temperature. We have found a decrease in Debye temperature around 40 kbar in all the AIRE compounds.

  13. Impact of Air Temperature Distributed Calculation in Glacier Mass Balance Modeling

    Science.gov (United States)

    Dalla Fontana, G.; Carturan, L.; Cazorzi, F.

    2014-12-01

    Distributed models of snow and ice mass balance enable a better understanding of processes involved in glacier hydrology and the prediction of glacier runoff under possible future climatic scenarios. The so-called 'Enhanced Temperature-Index' (ETI) melt models are a good compromise between model simplicity, parsimony of input data, and the capability to account for dominant processes in snow and ice mass balance. Accurate spatial calculation of temperature input data is crucial, given the key role of air temperature in modeling ablation and accumulation processes, further emphasized in ETI models. Compared to ambient conditions, lower temperatures (the so-called glacier cooling effect), and temperature variability (the so-called glacier damping effect) generally occur over glaciers, complicating the extrapolation from off-glacier weather stations. A comprehensive dataset of mass balance measurements and high-altitude meteorological observations was collected on La Mare and Careser glaciers (Ortles-Cevedale, Italian Alps) in 2010 and 2011. This dataset was used to analyze the air temperature distribution and wind regime over the glaciers, and to evaluate the impact of different calculation methods proposed in the literature for calculating on-glacier temperatures from off-glacier data. A general-purpose ETI model (EISModel - Energy Index Snow-and-ice Model) was used for simulating snow and ice accumulation and melt processes. Results indicate that i) none of the existing methods fully accounts for the actual temperature distribution over glaciers, ii) even small deviations in air temperature calculations strongly impact the simulations, and iii) there is an important positive feedback related to glacier shrinking and disintegration. Among the tested methods, the more physically-based procedure of Greuell and Bohm (1998) provided the best overall results. Therefore, it was implemented in EISModel for distributed air temperature calculations over glaciers.

  14. Studies on application of neutron activation analysis -Applied research on air pollution monitoring and development of analytical method of environmental samples

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Yong Sam; Moon, Jong Hwa; Chung, Young Ju; Jeong, Eui Sik; Lee, Sang Mi; Kang, Sang Hun; Cho, Seung Yeon; Kwon, Young Sik; Chung, Sang Wuk; Lee, Kyu Sung; Chun, Ki Hong; Kim, Nak Bae; Lee, Kil Yong; Yoon, Yoon Yeol; Chun, Sang Ki

    1997-09-01

    This research report is written for results of applied research on air pollution monitoring using instrumental neutron activation analysis. For identification and standardization of analytical method, 24 environmental samples are analyzed quantitatively, and accuracy and precision of this method are measured. Using airborne particulate matter and biomonitor chosen as environmental indicators, trace elemental concentrations of sample collected at urban and rural site monthly are determined ant then the calculation of statistics and the factor analysis are carried out for investigation of emission source. Facilities for NAA are installed in a new HANARO reactor, functional test is performed for routine operation. In addition, unified software code for NAA is developed to improve accuracy, precision and abilities of analytical processes. (author). 103 refs., 61 tabs., 19 figs.

  15. Reliability assessment of high energy particle induced radioactivity calculation code DCHAIN-SP 2001 by analysis of integral activation experiments with 14 MeV neutrons

    Energy Technology Data Exchange (ETDEWEB)

    Kai, Tetsuya; Maekawa, Fujio; Kasugai, Yoshimi; Takada, Hiroshi; Ikeda, Yujiro [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Kosako, Kazuaki [Sumitomo Atomic Energy Industries, Ltd., Tokyo (Japan)

    2002-03-01

    Reliability assessment for the high energy particle induced radioactivity calculation code DCHAIN-SP 2001 was carried out through analysis of integral activation experiments with 14-MeV neutrons aiming at validating the cross section and decay data revised from previous version. The following three kinds of experiments conducted at the D-T neutron source facility, FNS, in JAERI were employed: (1) the decay gamma-ray measurement experiment for fusion reactor materials, (2) the decay heat measurement experiment for 32 fusion reactor materials, and (3) the integral activation experiment on mercury. It was found that the calculations with DCHAIN-SP 2001 predicted the experimental data for (1) - (3) within several tens of percent. It was concluded that the cross section data below 20 MeV and the associated decay data as well as the calculation algorithm for solving the Beteman equation that was the master equation of DCHAIN-SP were adequate. (author)

  16. Development of modern methods with respect to neutron transport and uncertainty analyses for reactor core calculations. Interim report; Weiterentwicklung moderner Verfahren zu Neutronentransport und Unsicherheitsanalysen fuer Kernberechnungen. Zwischenbericht

    Energy Technology Data Exchange (ETDEWEB)

    Zwermann, Winfried; Aures, Alexander; Bostelmann, Friederike; Pasichnyk, Ihor; Perin, Yann; Velkov, Kiril; Zilly, Matias

    2016-12-15

    This report documents the status of the research and development goals reached within the reactor safety research project RS1536 ''Development of modern methods with respect to neutron transport and uncertainty analyses for reactor core calculations'' as of the 3{sup rd} quarter of 2016. The superordinate goal of the project is the development, validation, and application of neutron transport methods and uncertainty analyses for reactor core calculations. These calculation methods will mainly be applied to problems related to the core behaviour of light water reactors and innovative reactor concepts, in particular fast reactors cooled by liquid metal. The contributing individual goals are the further optimization and validation of deterministic calculation methods with high spatial and energy resolution, the development of a coupled calculation system using the Monte Carlo method for the neutron transport to describe time-dependent reactor core states, the processing and validation of nuclear data, particularly with regard to covariance data, the development, validation, and application of sampling-based methods for uncertainty and sensitivity analyses, the creation of a platform for performing systematic uncertainty analyses for fast reactor systems, as well as the description of states of severe core damage with the Monte Carlo method. Moreover, work regarding the European NURESAFE project, started in the preceding project RS1503, are being continued and completed.

  17. Calculation of the yields for the primary species formed from the radiolysis of liquid water by fast neutrons at temperatures between 25-350°C.

    Science.gov (United States)

    Butarbutar, Sofia Loren; Sanguanmith, Sunuchakan; Meesungnoen, Jintana; Sunaryo, Geni Rina; Jay-Gerin, Jean-Paul

    2014-06-01

    Monte Carlo simulations were used to calculate the yields for the primary species (e(-)aq, H(•), H2, (•)OH and H2O2) formed from the radiolysis of neutral liquid water by mono-energetic 2 MeV neutrons at temperatures between 25-350°C. The 2 MeV neutron was taken as representative of a fast neutron flux in a reactor. For light water, the moderation of these neutrons generated elastically scattered recoil protons of ∼1.264, 0.465, 0.171 and 0.063 MeV, which at 25°C, had linear energy transfers (LETs) of ∼22, 43, 69 and 76 keV/μm, respectively. Neglecting the radiation effects due to oxygen ion recoils and assuming that the most significant contribution to the radiolysis came from these first four recoil protons, the fast neutron yields could be estimated as the sum of the yields for these protons after allowance was made for the appropriate weightings according to their energy. Yields were calculated at 10(-7), 10(-6) and 10(-5) s after the ionization event at all temperatures, in accordance with the time range associated with the scavenging capacities generally used for fast neutron radiolysis experiments. The results of the simulations agreed reasonably well with the experimental data, taking into account the relatively large uncertainties in the experimental measurements, the relatively small number of reported radiolysis yields, and the simplifications included in the model. Compared with data obtained for low-LET radiation ((60)Co γ rays or fast electrons), our computed yields for fast neutron radiation showed essentially similar temperature dependences over the range of temperatures studied, but with lower values for yields of free radicals and higher values for molecular yields. This general trend is a reflection of the high-LET character of fast neutrons. Although the results of the simulations were consistent with the experiment, more experimental data are required to better describe the dependence of radiolytic yields on temperature and to test

  18. REFRACTIVE NEUTRON LENS

    OpenAIRE

    Petrov, P. V.; Kolchevsky, N. N.

    2013-01-01

    Compound concave refractive lenses are used for focusing neutron beam. Investigations of spectral and focusing properties of a refractive neutron lens are presented. Resolution of the imaging system on the base of refractive neutron lenses depends on material properties and parameters of neutron source. Model of refractive neutron lens are proposed. Results of calculation diffraction resolution and focal depth of refractive neutron lens are discussed.

  19. Development and application of neutron transport methods and uncertainty analyses for reactor core calculations. Technical report; Entwicklung und Einsatz von Neutronentransportmethoden und Unsicherheitsanalysen fuer Reaktorkernberechnungen. Technischer Bericht

    Energy Technology Data Exchange (ETDEWEB)

    Zwermann, W.; Aures, A.; Bernnat, W.; and others

    2013-06-15

    This report documents the status of the research and development goals reached within the reactor safety research project RS1503 ''Development and Application of Neutron Transport Methods and Uncertainty Analyses for Reactor Core Calculations'' as of the 1{sup st} quarter of 2013. The superordinate goal of the project is the development, validation, and application of neutron transport methods and uncertainty analyses for reactor core calculations. These calculation methods will mainly be applied to problems related to the core behaviour of light water reactors and innovative reactor concepts. The contributions of this project towards achieving this goal are the further development, validation, and application of deterministic and stochastic calculation programmes and of methods for uncertainty and sensitivity analyses, as well as the assessment of artificial neutral networks, for providing a complete nuclear calculation chain. This comprises processing nuclear basis data, creating multi-group data for diffusion and transport codes, obtaining reference solutions for stationary states with Monte Carlo codes, performing coupled 3D full core analyses in diffusion approximation and with other deterministic and also Monte Carlo transport codes, and implementing uncertainty and sensitivity analyses with the aim of propagating uncertainties through the whole calculation chain from fuel assembly, spectral and depletion calculations to coupled transient analyses. This calculation chain shall be applicable to light water reactors and also to innovative reactor concepts, and therefore has to be extensively validated with the help of benchmarks and critical experiments.

  20. Cosmic-ray neutron simulations and measurements in Taiwan.

    Science.gov (United States)

    Chen, Wei-Lin; Jiang, Shiang-Huei; Sheu, Rong-Jiun

    2014-10-01

    This study used simulations of galactic cosmic ray in the atmosphere to investigate the neutron background environment in Taiwan, emphasising its altitude dependence and spectrum variation near interfaces. The calculated results were analysed and compared with two measurements. The first measurement was a mobile neutron survey from sea level up to 3275 m in altitude conducted using a car-mounted high-sensitivity neutron detector. The second was a previous measured result focusing on the changes in neutron spectra near air/ground and air/water interfaces. The attenuation length of cosmic-ray neutrons in the lower atmosphere was estimated to be 163 g cm(-2) in Taiwan. Cosmic-ray neutron spectra vary with altitude and especially near interfaces. The determined spectra near the air/ground and air/water interfaces agree well with measurements for neutrons below 10 MeV. However, the high-energy portion of spectra was observed to be much higher than our previous estimation. Because high-energy neutrons contribute substantially to a dose evaluation, revising the annual sea-level effective dose from cosmic-ray neutrons at ground level in Taiwan to 35 μSv, which corresponds to a neutron flux of 5.30 × 10(-3) n cm(-2) s(-1), was suggested. © The Author 2014. Published by Oxford University Press. All rights reserved. For Permissions, please email: journals.permissions@oup.com.

  1. Sassena — X-ray and neutron scattering calculated from molecular dynamics trajectories using massively parallel computers

    Science.gov (United States)

    Lindner, Benjamin; Smith, Jeremy C.

    2012-07-01

    Massively parallel computers now permit the molecular dynamics (MD) simulation of multi-million atom systems on time scales up to the microsecond. However, the subsequent analysis of the resulting simulation trajectories has now become a high performance computing problem in itself. Here, we present software for calculating X-ray and neutron scattering intensities from MD simulation data that scales well on massively parallel supercomputers. The calculation and data staging schemes used maximize the degree of parallelism and minimize the IO bandwidth requirements. The strong scaling tested on the Jaguar Petaflop Cray XT5 at Oak Ridge National Laboratory exhibits virtually linear scaling up to 7000 cores for most benchmark systems. Since both MPI and thread parallelism is supported, the software is flexible enough to cover scaling demands for different types of scattering calculations. The result is a high performance tool capable of unifying large-scale supercomputing and a wide variety of neutron/synchrotron technology. Catalogue identifier: AELW_v1_0 Program summary URL:http://cpc.cs.qub.ac.uk/summaries/AELW_v1_0.html Program obtainable from: CPC Program Library, Queen's University, Belfast, N. Ireland Licensing provisions: GNU General Public License, version 3 No. of lines in distributed program, including test data, etc.: 1 003 742 No. of bytes in distributed program, including test data, etc.: 798 Distribution format: tar.gz Programming language: C++, OpenMPI Computer: Distributed Memory, Cluster of Computers with high performance network, Supercomputer Operating system: UNIX, LINUX, OSX Has the code been vectorized or parallelized?: Yes, the code has been parallelized using MPI directives. Tested with up to 7000 processors RAM: Up to 1 Gbytes/core Classification: 6.5, 8 External routines: Boost Library, FFTW3, CMAKE, GNU C++ Compiler, OpenMPI, LibXML, LAPACK Nature of problem: Recent developments in supercomputing allow molecular dynamics simulations to

  2. Neutron Albedo

    CERN Document Server

    Ignatovich, V K

    2005-01-01

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

  3. GUIDE TO CALCULATING TRANSPORT EFFICIENCY OF AEROSOLS IN OCCUPATIONAL AIR SAMPLING SYSTEMS

    Energy Technology Data Exchange (ETDEWEB)

    Hogue, M.; Hadlock, D.; Thompson, M.; Farfan, E.

    2013-11-12

    This report will present hand calculations for transport efficiency based on aspiration efficiency and particle deposition losses. Because the hand calculations become long and tedious, especially for lognormal distributions of aerosols, an R script (R 2011) will be provided for each element examined. Calculations are provided for the most common elements in a remote air sampling system, including a thin-walled probe in ambient air, straight tubing, bends and a sample housing. One popular alternative approach would be to put such calculations in a spreadsheet, a thorough version of which is shared by Paul Baron via the Aerocalc spreadsheet (Baron 2012). To provide greater transparency and to avoid common spreadsheet vulnerabilities to errors (Burns 2012), this report uses R. The particle size is based on the concept of activity median aerodynamic diameter (AMAD). The AMAD is a particle size in an aerosol where fifty percent of the activity in the aerosol is associated with particles of aerodynamic diameter greater than the AMAD. This concept allows for the simplification of transport efficiency calculations where all particles are treated as spheres with the density of water (1g cm-3). In reality, particle densities depend on the actual material involved. Particle geometries can be very complicated. Dynamic shape factors are provided by Hinds (Hinds 1999). Some example factors are: 1.00 for a sphere, 1.08 for a cube, 1.68 for a long cylinder (10 times as long as it is wide), 1.05 to 1.11 for bituminous coal, 1.57 for sand and 1.88 for talc. Revision 1 is made to correct an error in the original version of this report. The particle distributions are based on activity weighting of particles rather than based on the number of particles of each size. Therefore, the mass correction made in the original version is removed from the text and the calculations. Results affected by the change are updated.

  4. A Monte-Carlo code for neutron efficiency calculations for large volume Gd-loaded liquid scintillation detectors

    Energy Technology Data Exchange (ETDEWEB)

    Trzcinski, A.; Zwieglinski, B. [Soltan Inst. for Nuclear Studies, Warsaw (Poland); Lynen, U. [Gesellschaft fuer Schwerionenforschung mbH, Darmstadt (Germany); Pochodzalla, J. [Max-Planck-Institut fuer Kernphysik, Heidelberg (Germany)

    1998-10-01

    This paper reports on a Monte-Carlo program, MSX, developed to evaluate the performance of large-volume, Gd-loaded liquid scintillation detectors used in neutron multiplicity measurements. The results of simulations are presented for the detector intended to count neutrons emitted by the excited target residue in coincidence with the charged products of the projectile fragmentation following relativistic heavy-ion collisions. The latter products could be detected with the ALADIN magnetic spectrometer at GSI-Darmstadt. (orig.) 61 refs.

  5. DIDEM - An integrated model for comparative health damage costs calculation of air pollution

    Science.gov (United States)

    Ravina, Marco; Panepinto, Deborah; Zanetti, Maria Chiara

    2018-01-01

    Air pollution represents a continuous hazard to human health. Administration, companies and population need efficient indicators of the possible effects given by a change in decision, strategy or habit. The monetary quantification of health effects of air pollution through the definition of external costs is increasingly recognized as a useful indicator to support decision and information at all levels. The development of modelling tools for the calculation of external costs can provide support to analysts in the development of consistent and comparable assessments. In this paper, the DIATI Dispersion and Externalities Model (DIDEM) is presented. The DIDEM model calculates the delta-external costs of air pollution comparing two alternative emission scenarios. This tool integrates CALPUFF's advanced dispersion modelling with the latest WHO recommendations on concentration-response functions. The model is based on the impact pathway method. It was designed to work with a fine spatial resolution and a local or national geographic scope. The modular structure allows users to input their own data sets. The DIDEM model was tested on a real case study, represented by a comparative analysis of the district heating system in Turin, Italy. Additional advantages and drawbacks of the tool are discussed in the paper. A comparison with other existing models worldwide is reported.

  6. APPROXIMATE CALCULATION OF BASIC CHARACTERISTICS OF PLASMA AT THE AIR ELECTRIC EXPLOSION OF METAL CONDUCTOR

    Directory of Open Access Journals (Sweden)

    M. I. Baranov

    2017-12-01

    Full Text Available Purpose. Obtaining approximate calculation correlations for determination of maximal values of temperature Tm, and pressures Pm at a shock wave and speed vm distribution of a shock wave in the plasma products of air electric explosion (EE of metall conductor under act of large impulsive current (LIC. Methodology. Theoretical bases of the electrical engineering, scientific and technical bases of electrophysics, thermal physics and electrophysics bases of powerful high-voltage impulse technique, related to the theory and practice of the phenomenon EE metallic explorer in gas environments under action of LIC. Results. New calculation correlations are got for approximate calculation in a local area of EE in atmospheric air of metallic explorer of maximal values of temperature Tm,, pressures Pm and speeds of vm of shock wave in «metallic plasma» appearing from an explosion under action of LIC of its conducting structure. It is set that numeral values of the sought after sizes of temperature Tm,, pressures Pm and speeds vm as it applies to air EE thin copper conductor under the action of LIC of the microsecond temporal range can arrive at a few ten of thousands of Kelvin, hundreds of technical atmospheres and thousands of meters in a second accordingly. It is shown that similar values of speed vm of shock wave in «metallic plasma» are comparable at speed of detonation wave in hard explosives. An accent is in this connection done on expedience of application air EE thin short metallic conductors at injury of live ammunitions with an ordinary and nuclear explosive. The real technical suggestions are offered on a possible receipt in the discharge circuit of powerful high-voltage generator of LIC of condenser type of «record» (most values of the examined descriptions of «metallic plasma» at air EE thin metallic conductors. Comparison of the obtained results is executed for the probed descriptions of plasma at air EE of the metal conductor with

  7. Calculation and Experimental Validation of Pressure and Temperature Effects on COG-Air Fuel Mixtures

    Directory of Open Access Journals (Sweden)

    Jan Skrinsky

    2018-01-01

    Full Text Available COG have been widely used together with blast furnace gas and blast furnace oxygen gas in the steel industry in Moravian-Silesian region of Czech Republic. COG is a flammable and explosive substance. Most explosion characteristics published so far are valid for pure compounds and limited experimental conditions, mostly ambient. There have been no explosion characteristic exists for COG-air mixtures which cover industrial conditions up to 423 K. Experimental tests have been carried out in a 20-L closed explosion chamber adopted for the explosion tests. The element potential approach in the thermochemical equilibrium calculations applied in the Chemkin subroutine has been used for explosion pressure calculations. Different explosion characteristics have been reported in a range from 298 K up to 423 K and from 0.5 bar(a up to 1.0 bar(a.

  8. Radiolytic yield of ozone in air for low dose neutron and x-ray/gamma-ray radiation

    Science.gov (United States)

    Cole, J.; Su, S.; Blakeley, R. E.; Koonath, P.; Hecht, A. A.

    2015-01-01

    Radiation ionizes surrounding air and produces molecular species, and these localized effects may be used as a signature of, and for quantification of, radiation. Low-level ozone production measurements from radioactive sources have been performed in this work to understand radiation chemical yields at low doses. The University of New Mexico AGN-201 M reactor was used as a tunable radiation source. Ozone levels were compared between reactor-on and reactor-off conditions, and differences (0.61 to 0.73 ppb) well below background levels were measured. Simulations were performed to determine the dose rate distribution and average dose rate to the air sample within the reactor, giving 35 mGy of mixed photon and neutron dose. A radiation chemical yield for ozone of 6.5±0.8 molecules/100 eV was found by a variance weighted average of the data. The different contributions of photons and neutrons to radiolytic ozone production are discussed.

  9. EURISOL-DS Multi-MWatt Hg Target: Neutron flux and fission rate calculations for the MAFF configuration

    CERN Document Server

    Romanets, Y; Vaz, P; Herrera-Martinez, A; Kadi, Y; Kharoua, C; Lettry, J; Lindroos, M

    The EURISOL (The EURopean Isotope Separation On-Line Radioactive Ion Beam) project aims at producing high intensity radioactive ion beams produced by neutron induced fission on a fissile target (235U) surrounding a liquid mercury converter. A proton beam of 1 GeV and 4 MW impinges on the Hg converter generating by spallation reactions high neutron fluxes. In this work the state-of-the-art Monte Carlo codes MCNPX and FLUKA were used to assess the neutronics performance of the system which geometry, inspired from the MAFF concept, allows a versatile manipulation of the fission targets. The objective of the study was to optimize the geometry of the system and the materials used in the fuel and reflector elements of the system, in order to achieve the highest possible fission rate.

  10. Implementation of the optimization for the methodology of the neutronic calculation and thermo-hydraulic in IEA-R1 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Stefani, Giovanni Laranjo de; Conti, Thadeu das Neves; Fedorenko, Giuliana G.; Castro, Vinicius A.; Maio, Mireia F., E-mail: gstefani@ipen.b, E-mail: tnconti@ipen.b, E-mail: g.fedorenko@ipen.b, E-mail: vcastro@ipen.b, E-mail: mfmaio@ipen.b [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil); Santos, Thiago Augusto dos, E-mail: tsantos@ipen.b [Universidade de Sao Paulo (IFUSP), Sao Paulo, SP (Brazil). Inst. de Fisica

    2011-07-01

    This work objective was to create a manager program that would automate the programs and computer codes in use for neutronic calculation and thermo-hydraulic in IEA-R1 reactor thus making the process for calculation of safety parameters and for configuration change up to 98% faster than that used in the reactor today. This process was tested in combination with the reactor operators and is being implemented by the quality department. The main codes and programs involved in the calculations of configuration change are Leopard, Hammier-Technion, Twodb, Citation and Cobra. Calculations of delayed neutron and criticality coefficients given in the process of safety parameters calculation are given by the Hammer-Technion and Citation in a process that involves about eleven repetitions so that it meets all the necessary conditions (such different temperatures of the moderator and fuel). The results are entirely consistent with the expected and absolutely the same as those given by manual process. Thus the work shows its reliability as well the advantage of saving time, once a process that could take up to four hours was turned in one that takes around five minutes when done in a home computer. Much of this advantage is due to the fact that were created subprograms to treat the output of each program used and transform them into the input of the other programs, removing from it the intermediate essential data for this to occur, thus avoiding also a possible human error by handling the various data supplied. (author)

  11. In vitro biological effectiveness of JRR-4 epithermal neutron beam. Experiment under free air beam and in water phantom. Cooperative research

    Energy Technology Data Exchange (ETDEWEB)

    Yamamoto, Tetsuya; Matsumura, Akira; Nose, Tadao [Tsukuba Univ., Tsukuba, Ibaraki (Japan); Yamamoto, Kazuyoshi; Kumada, Hiroaki; Kishi, Toshiaki; Hori, Naohiko; Torii, Yoshiya; Horiguchi, Yoji [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2002-05-01

    The surviving curve and the biological effectiveness factor of dose components generated in boron neutron capture therapy (BNCT) were separately determined in neutron beams at Japan Research Reactor No.4. Surviving fraction of V79 Chinese hamster cell with or without {sup 10}B was obtained using an epithermal neutron beam (ENB), a mixed thermal-epithermal neutron beam (TNB-1), and a thermal neutron beam (TNB-2), which were used or planned to use for BNCT clinical trial. The cell killing effect of these neutron beams with or without the presence of {sup 10}B depended highly on the neutron beam used, according to the epithermal and fast neutron content in the beam. The biological effectiveness factor values of the boron capture reaction for ENB, TNB-1 and TNB-2 were 3.99{+-}0.24, 3.04{+-}0.19 and 1.43{+-}0.08, respectively. The biological effectiveness factor values of the high-LET dose components based on the hydrogen recoils and the nitrogen capture reaction were 2.50{+-}0.32, 2.34{+-}0.30 and 2.17{+-}0.28 for ENB, TNB-1 and TNB-2, respectively. The biological effectiveness factor values of the neutron and photon components were 1.22{+-}0.16, 1.23{+-}0.16 and 1.21{+-}0.16, respectively. The depth function of biological effectiveness factor in water phantom and the difference in biological effectiveness factor among boron compounds were also determined. The experimental determination of biological effectiveness factor outlined in this paper is applicable to the dose calculation for each dose component of the neutron beams and contribute to an accurate biological effectiveness factor as comparison with a neutron beam at a different facility employed in ongoing and planned BNCT clinical trials. (author)

  12. Model calculated global, regional and megacity premature mortality due to air pollution

    Directory of Open Access Journals (Sweden)

    J. Lelieveld

    2013-07-01

    Full Text Available Air pollution by fine particulate matter (PM2.5 and ozone (O3 has increased strongly with industrialization and urbanization. We estimate the premature mortality rates and the years of human life lost (YLL caused by anthropogenic PM2.5 and O3 in 2005 for epidemiological regions defined by the World Health Organization (WHO. This is based upon high-resolution global model calculations that resolve urban and industrial regions in greater detail compared to previous work. Results indicate that 69% of the global population is exposed to an annual mean anthropogenic PM2.5 concentration of >10 μg m−3 (WHO guideline and 33% to > 25 μg m−3 (EU directive. We applied an epidemiological health impact function and find that especially in large countries with extensive suburban and rural populations, air pollution-induced mortality rates have been underestimated given that previous studies largely focused on the urban environment. We calculate a global respiratory mortality of about 773 thousand/year (YLL ≈ 5.2 million/year, 186 thousand/year by lung cancer (YLL ≈ 1.7 million/year and 2.0 million/year by cardiovascular disease (YLL ≈ 14.3 million/year. The global mean per capita mortality caused by air pollution is about 0.1% yr−1. The highest premature mortality rates are found in the Southeast Asia and Western Pacific regions (about 25% and 46% of the global rate, respectively where more than a dozen of the most highly polluted megacities are located.

  13. Calculation of pre-equilibrium effects in neutron-induced cross section on 32,34S isotopes using the EMPIRE 3.2 code

    Directory of Open Access Journals (Sweden)

    Yettou Leila

    2015-01-01

    Full Text Available In this study, a new version EMPIRE 3.2 code was used in the cross section calculations of (n,p reactions and in the calculation of proton emission spectra produced by (n,xp reactions. Exciton model predictions combined with the Kalbach angular distribution systematics were used and some parameters such as those of mean free path, cluster emission in terms of Iwamoto-Harada model, optical model potentials of Morillon for neutrons and protons in the energy range up to 20 MeV, level density for spherical nuclei of Gilbert-Cameron model and width fluctuation correction in terms of compound nucleus have been investigated our calculations. The excitation functions and the proton emission spectra for 32,34S nuclei were calculated, discussed and found in good agreement with available experimental data.

  14. Calculations of multidimensional test reactor neutron diffusion problems with KORAT 3D code and its combination with two-velocity two-temperature thermohydraulic RATEG code

    Energy Technology Data Exchange (ETDEWEB)

    Voronova, O.A.; Grebennikov, A.N.; Zvenigorodskaya, O.A. [Russian Federal Nuclear Center, Arzamas (Russian Federation)] [and others

    1996-09-01

    The paper briefly presents the method for numerical computation of 3-D group neutron diffusion equation implemented in KORAT 3D code within SATURN complex and oriented to 3-D stationary and nonstationary calculations in nuclear reactor physics including those on up-to-date distributed-memory multiprocessors. The computational results are given obtained with the above mentioned code for multidimensional test reactor problems as well as the results of numerical efficiency studies for the code parallelization on multiprocessors Cray T3D and SP-2. Then the setup is given for 3-D dynamic problem simulating a hypothetical mode of WPBER reactor. The computational results are presented for the above problem which includes the neutron-nuclear processes, thermohydraulic processes and the feedback. The computations used two program complexes: KORAT 3D + multichannel two-velocity two-temperature thermohydraulics code RATEG and READY complex. (author)

  15. The NIMO Monte Carlo model for box-air-mass factor and radiance calculations

    Science.gov (United States)

    Hay, Timothy D.; Bodeker, Greg E.; Kreher, Karin; Schofield, Robyn; Liley, J. Ben; Scherer, Martin; McDonald, Adrian J.

    2012-06-01

    A new fully spherical multiple scattering Monte Carlo radiative transfer model named NIMO (NIWA Monte Carlo model) is presented. The ray tracing algorithm is described in detail along with the treatment of scattering and absorption, and the simulation of backward adjoint trajectories. The primary application of NIMO is the calculation of box-air-mass factors (box-AMFs), which are used to convert slant column densities (SCDs) of trace gases, derived from UV-visible multiple axis Differential Optical Absorption Spectroscopy (MAX-DOAS) measurements, into vertical column densities (VCDs). Box-AMFs are also employed as weighting functions for optimal estimation retrievals of vertical trace gas profiles from SCDs. Monte Carlo models are well suited to AMF calculations at high solar zenith angles (SZA) and at low viewing elevation angles where multiple scattering is important. Additionally, the object-oriented structure of NIMO makes it easily extensible to new applications by plugging in objects for new absorbing or scattering species. Box-AMFs and radiances, calculated for various wavelengths, SZAs, viewing elevation and azimuth angles and aerosol scenarios, are compared with results from nine other models using a set of exercises from a recent radiative transfer model intercomparison. NIMO results for these simulations are well within the range of variability of the other models.

  16. Neutron Resonance Parameters of 238U and the Calculated Cross Sections from the Reich-Moore Analysis of Experimental Data in the Neutron Energy Range from 0 keV to 20 keV

    Energy Technology Data Exchange (ETDEWEB)

    Derrien, H

    2005-12-05

    The neutron resonance parameters of {sup 238}U were obtained from a SAMMY analysis of high-resolution neutron transmission measurements and high-resolution capture cross section measurements performed at the Oak Ridge Electron Linear Accelerator (ORELA) in the years 1970-1990, and from more recent transmission and capture cross section measurements performed at the Geel Linear Accelerator (GELINA). Compared with previous evaluations, the energy range for this resonance analysis was extended from 10 to 20 keV, taking advantage of the high resolution of the most recent ORELA transmission measurements. The experimental database and the method of analysis are described in this report. The neutron transmissions and the capture cross sections calculated with the resonance parameters are compared with the experimental data. A description is given of the statistical properties of the resonance parameters and of the recommended values of the average parameters. The new evaluation results in a slight decrease of the effective capture resonance integral and improves the prediction of integral thermal benchmarks by 70 pcm to 200 pcm.

  17. Calculation of the backscattering in water and compared to the values in air; Calculo del factor de retrodispersion en agua y comparativa con los valores en aire

    Energy Technology Data Exchange (ETDEWEB)

    Minano Herrero, J. A.; Sarasa Rubio, A.; Roldan Arjona, J. M.

    2011-07-01

    The purpose of this paper is to calculate values of BSF in water and comparison with data on air 11SF found in the literature. For this simulations have been performed by the Monte Carlo method for calculating values ??kerma water in the presence of a manikin of this material and in the absence thereof. The simulations were performed for monoenergetic beams in order to facilitate the calculation of the BSF for any spectral distribution of those found in the field of radiology.

  18. Neutronic simulation calculations to assess the proliferation resistance of nuclear technologies; Neutronenphysikalische Simulationsrechnungen zur Proliferationsresistenz nuklearer Technologien

    Energy Technology Data Exchange (ETDEWEB)

    Englert, Matthias

    2009-07-13

    This thesis investigates the proliferation resistance of nuclear technologies on the basis of three case studies. After a brief description of the concept of proliferation resistance the utilized computer codes and methods are presented. The first case study investigates the potential of monolithic fuel for the conversion of one-fuel-element high-flux research reactors from highly enriched to low enriched uranium using the example of the german research reactor FRM-II. The second case study assesses the proliferation potential of future tokamak based fusion reactors by using neutronic simulations of a possible plutonium production. The third example investigates the proliferation potential of spallation neutron sources to produce nuclear weapon relevant material and the proliferation resistance of such facilities. (orig.)

  19. EURISOL-DS multi-MW target unit: Neutronics performance and shielding assessment, dose rate and material activation calculations for the MAFF configuration

    CERN Document Server

    Romanets, Y; Kadi, Y; Luis, R; Goncalves, I F; Tecchio, L; Kharoua, C; Vaz, P; Ene, D; David, J C; Rocca, R; Negoita, F

    2010-01-01

    One of the objectives of the EURISOL (EURopean Isotope Separation On-Line Radioactive Ion Beam) Design Study consisted of providing a safe and reliable facility layout and design for the following operational parameters and characteristics: (a) a 4 MW proton beam of 1 GeV energy impinging on a mercury target (the converter); (b) high neutron fluxes (similar to 3 x 10(16) neutrons/s) generated by spallation reactions of the protons impinging in the converter and (c) fission rate on fissile U-235 targets in excess of 10(15) fissions/s. In this work, the state-of-the-art Monte Carlo codes MCNPX (Pelowitz, 2005) and FLUKA (Vlachoudis, 2009; Ferrari et al., 2008) were used to characterize the neutronics performance and to perform the shielding assessment (Herrera-Martinez and Kadi, 2006; Cornell, 2003) of the EURISOLTarget Unit and to provide estimations of dose rate and activation of different components, in view of the radiation safety assessment of the facility. Dosimetry and activation calculations were perfor...

  20. Neutronics Benchmarks for the Utilization of Mixed-Oxide Fuel: Joint US/Russian Progress Report for Fiscal 1997. Volume 3 - Calculations Performed in the Russian Federation

    Energy Technology Data Exchange (ETDEWEB)

    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.

  1. (n,p, (n,2n, (n,d, and (n,α cross-section calculations of 16O with 0-40 MeV energy neutrons

    Directory of Open Access Journals (Sweden)

    Ozdemir Omer Faruk

    2015-01-01

    Full Text Available Oxygen is one of the elements which interacts with emitted neutrons after fission reactions. Oxygen exists abundantly both in nuclear fuel (UO2 and moderators (H2O. Nuclear reactions of oxygen with neutrons are important in terms of stability of nuclear fuel and neutron economy. In this study, equilibrium and pre-equilibrium models have been used to calculate (n,p, (n,d, (n,2n and (n,α nuclear reaction cross-sections of 16O. In these calculations, neutron incident energy has been taken up to 40 MeV. Hybrid and Standard Weisskopf-Ewing Models in ALICE-2011 program, Weisskopf-Ewing and Full Exciton Models in PCROSS program, and Cascade Exciton Model in CEM03.01 program have been utilized. The calculated results have been compared with experimental and theroretical cross-section data which are obtained from libraries of EXFOR and ENDF/B VII.1.

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

    Energy Technology Data Exchange (ETDEWEB)

    White, Morgan C. [Univ. of Florida, Gainesville, FL (United States)

    2000-07-01

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

  3. Optimization study and neutronic and thermal-hydraulic design calculations of a 75 KWTH aqueous homogeneous reactor for medical isotopes production

    Energy Technology Data Exchange (ETDEWEB)

    Perez, Daniel Milian; Lorenzo, Daniel E. Milian; Garcia, Lorena P. Rodriguez; Llanes, Jesus Salomon; Hernandez, Carlos R. Garcia, E-mail: dperez@instec.cu, E-mail: dmilian@instec.cu, E-mail: lorenapilar@instec.cu, E-mail: cgh@instec.cu [Instituto Superior de Tecnologias y Ciencias Aplicadas (InSTEC), La Habana (Cuba); Lira, Carlos A. Brayner de Oliveira, E-mail: cabol@ufpe.br [Universidade Federal de Pernambuco (UFPE), Recife (Brazil); Rodriguez, Manuel Cadavid, E-mail: mcadavid2001@yahoo.com [Tecnologia Nuclear Medica Spa, TNM (Chile)

    2015-07-01

    {sup 99m}Tc is the most common radioisotope used in nuclear medicine. It is a very useful radioisotope, which is used in about 30-40 million procedures worldwide every year. Medical diagnostic imaging techniques using {sup 99m}Tc represent approximately 80% of all nuclear medicine procedures. Although {sup 99m}Tc can be produced directly on a cyclotron or other type of particle accelerator, currently is almost exclusively produced from the beta-decay of its 66-h parent {sup 99}Mo. {sup 99}Mo production system in an Aqueous Homogeneous Reactor (AHR) is potentially advantageous because of its low cost, small critical mass, inherent passive safety, and simplified fuel handling, processing and purification characteristics. In this paper, an AHR conceptual design using Low Enriched Uranium (LEU) is studied and optimized for the production of {sup 99}Mo. Aspects related with the neutronic behavior such as optimal reflector thickness, critical height, medical isotopes production and the reactivity feedback introduced in the solution by the volumetric expansion of the fuel solution due to thermal expansion of the fuel solution and the void volume generated by radiolytic gas bubbles were evaluated. Thermal-hydraulics studies were carried out in order to show that sufficient cooling capacity exists to prevent fuel overheating. The neutronic and thermal-hydraulics calculations have been performed with the MCNPX computational code and the version 14 of ANSYS CFX respectively. The neutronic calculations demonstrated that the reactor is able to produce 370 six-day curies of {sup 99}Mo in 5 days operation cycles and the CFD simulation demonstrated that the heat removal systems provide sufficient cooling capacity to prevent fuel overheating, the maximum temperature reached by the fuel (89.29 deg C) was smaller to the allowable temperature limit (90 deg C). (author)

  4. Concrete enclosure to shield a neutron source

    Energy Technology Data Exchange (ETDEWEB)

    Villagrana M, L. E.; Rivera P, E.; De Leon M, H. A.; Soto B, T. G.; Hernandez D, V. M.; Vega C, H. R., E-mail: emmanuelvillagrana@hotmail.com [Universidad Autonoma de Zacatecas, Unidad Academica de Estudios Nucleares, Apdo. Postal 336, 98000 Zacatecas (Mexico)

    2012-10-15

    In the aim to design a shielding for a {sup 239}PuBe isotopic neutron source several Monte Carlo calculations were carried out using MCNP5 code. First, a point-like source was modeled in vacuum and the neutron spectrum and the ambient dose equivalent were calculated at several distances ranging from 5 up to 150 cm, these calculations were repeated including air, and a 1 x 1 x 1 m{sup 3} enclosure that was shielded with 5, 15, 20, 25, 30, 50 and 80 cm-thick Portland type concrete walls. At all the points located inside the enclosure neutron spectra from 10{sup -8} up 0.5 MeV were the same regardless the distance from the source showing the room-return effect, for energies larger than 0.5 MeV neutron spectra are diminished as the distance increases. Outside the enclosure it was noticed that neutron spectra becomes -softer- as the concrete thickness increases due to reduction of mean neutron energy. With the ambient dose values the attenuation curve in terms of concrete thickness was calculated. (Author)

  5. Gamma-Ray Emission Spectra as a Constraint on Calculations of 234,236,238U Neutron-Capture Cross Sections

    Energy Technology Data Exchange (ETDEWEB)

    Ullmann, John Leonard [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Kawano, Toshihiko [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Bredeweg, Todd Allen [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Baramsai, Bayarbadrakh [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Couture, Aaron Joseph [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Haight, Robert Cameron [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Jandel, Marian [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Mosby, Shea Morgan [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); O' Donnell, John M. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Rundberg, Robert S. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Vieira, David J. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Wilhelmy, Jerry B. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Becker, John A. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Wu, Ching-Yen [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Krticka, Milan [Charles Univ., Prague (Czech Republic)

    2015-05-28

    Neutron capture cross sections in the “continuum” region (>≈1 keV) and gamma-emission spectra are of importance to basic science and many applied fields. Careful measurements have been made on most common stable nuclides, but physicists must rely on calculations (or “surrogate” reactions) for rare or unstable nuclides. Calculations must be benchmarked against measurements (cross sections, gamma-ray spectra, and <Γγ>). Gamma-ray spectrum measurements from resolved resonances were made with 1 - 2 mg/cm2 thick targets; cross sections at >1 keV were measured using thicker targets. The results show that the shape of capture cross section vs neutron energy is not sensitive to the form of the strength function (although the magnitude is); the generalized Lorentzian E1 strength function is not sufficient to describe the shape of observed gamma-ray spectra; MGLO + “Oslo M1” parameters produces quantitative agreement with the measured 238U(n,γ) cross section; additional strength at low energies (~ 3 MeV) -- likely M1-- is required; and careful study of complementary results on low-lying giant resonance strength is needed to consistently describe observations.

  6. Calculation and evaluation of cross-sections and kerma factors for neutrons up to 100 MeV on {sup 16}O and {sup 14}N

    Energy Technology Data Exchange (ETDEWEB)

    Chadwick, M.B. [California Univ., Livermor, CA (United States). Lawrence Livermore National Lab.; Young, P.G.

    1997-03-01

    We present evaluations of the interaction of neutrons with energies between 20 and 100 MeV with oxygen and nitrogen nuclei, which follows on from our previous work on carbon. Our aim is to accurately represent integrated cross sections, inclusive emission spectra, and kerma factors, in a data library which can be used in radiation transport calculations. We apply the FKK-GNASH nuclear model code, which includes Hauser-Feshbach, preequilibrium, and direct reaction mechanisms, and use experimental measurements to optimize the calculations. We determine total, elastic, and nonelastic cross sections, angle-energy correlated emission spectra for light ejectiles with A {<=} 4 and gamma-rays, and average energy depositions. Our results for charged-particle emission spectra agree well with the measurements of Subramanian et al. We compare kerma factors derived from our evaluated cross sections with experimental data, providing an integral benchmarking of our work. (author). 52 refs.

  7. Heat Source Characterization In A TREAT Fuel Particle Using Coupled Neutronics Binary Collision Monte-Carlo Calculations

    Energy Technology Data Exchange (ETDEWEB)

    Schunert, Sebastian; Schwen, Daniel; Ghassemi, Pedram; Baker, Benjamin; Zabriskie, Adam; Ortensi, Javier; Wang, Yaqi; Gleicher, Frederick; DeHart, Mark; Martineau, Richard

    2017-04-01

    This work presents a multi-physics, multi-scale approach to modeling the Transient Test Reactor (TREAT) currently prepared for restart at the Idaho National Laboratory. TREAT fuel is made up of microscopic fuel grains (r ˜ 20µm) dispersed in a graphite matrix. The novelty of this work is in coupling a binary collision Monte-Carlo (BCMC) model to the Finite Element based code Moose for solving a microsopic heat-conduction problem whose driving source is provided by the BCMC model tracking fission fragment energy deposition. This microscopic model is driven by a transient, engineering scale neutronics model coupled to an adiabatic heating model. The macroscopic model provides local power densities and neutron energy spectra to the microscpic model. Currently, no feedback from the microscopic to the macroscopic model is considered. TREAT transient 15 is used to exemplify the capabilities of the multi-physics, multi-scale model, and it is found that the average fuel grain temperature differs from the average graphite temperature by 80 K despite the low-power transient. The large temperature difference has strong implications on the Doppler feedback a potential LEU TREAT core would see, and it underpins the need for multi-physics, multi-scale modeling of a TREAT LEU core.

  8. Investigation of Isfahan miniature neutron source reactor (MNSR for boron neutron capture therapy by MCNP simulation

    Directory of Open Access Journals (Sweden)

    S.Z Kalantari

    2015-01-01

    Full Text Available One of the important neutron sources for Boron Neutron Capture Therapy (BNCT is a nuclear reactor. It needs a high flux of epithermal neutrons. The optimum conditions of the neutron spectra for BNCT are provided by the International Atomic Energy Agency (IAEA. In this paper, Miniature Neutron Source Reactor (MNSR as a neutron source for BNCT was investigated. For this purpose, we designed a Beam Shaping Assembly (BSA for the reactor and the neutron transport from the core of the reactor to the output windows of BSA was simulated by MCNPX code. To optimize the BSA performance, two sets of parameters should be evaluated, in-air and in-phantom parameters. For evaluating in-phantom parameters, a Snyder head phantom was used and biological dose rate and dose-depth curve were calculated in brain normal and tumor tissues. Our calculations showed that the neutron flux of the MNSR reactor can be used for BNCT, and the designed BSA in optimum conditions had a good therapeutic characteristic for BNCT.

  9. Evaluation of cross sections and calculation of kerma factors for neutrons up to 80 MeV on {sup 12}C

    Energy Technology Data Exchange (ETDEWEB)

    Harada, M.; Watanabe, Y. [Kyushu Univ., Fukuoka (Japan); Chiba, S.; Fukahori, T.

    1997-03-01

    We have evaluated the cross sections for neutrons with incident energies from 20 to 80 MeV on {sup 12}C for the JENDL high-energy file. The total cross sections were determined by a generalized least-squares method with available experimental data. The cross sections of elastic and inelastic scattering to the first 2{sup +} were evaluated with the theoretical calculations. The optical potentials necessary for these calculations were derived using a microscopic approach by Jeukenne-Lejeune-Mahaux. For the evaluation of double differential emission cross sections (DDXs), we have developed a code system SCINFUL/DDX in which total 35 reactions including the 3-body simultaneous breakup process (n+{sup 12}C {yields} n+{alpha}+{sup 8}Be) can be taken into consideration in terms of a Monte Carlo method, and have calculated the DDXs of all light-emissions (A{<=}4) and heavier reaction products. The results for protons, deuterons, and alphas showed overall good agreement with experimental data. The code is also applicable for calculations of total and partial kerma factors. Total kerma factors calculated for energies from 20 to 80 MeV were compared with the measurements and the other latest evaluations from the viewpoints of medical application and nuclear heating estimation. (author)

  10. Calculation methods for air supply design in industrial facilities. Literature review

    Energy Technology Data Exchange (ETDEWEB)

    Hagstroem, K.; Siren, K.; Zhivov, A.M.

    1999-09-01

    The objectives of air distribution systems for warm air heating, ventilating, and air-conditioning are to create the proper thermal environment conditions in the occupied zone (combination of temperature, humidity, and air movement), and to control vapor and air born particle concentration within the target levels set by the process requirements and/or threshold limit values based on health effects, fire and explosion prevention, or other considerations. HVAC systems designs are constrained by existing codes, standards, and guidelines, which specify some minimum requirements for the HVAC system elements, occupant`s and process environmental quality and safety. There is a variety of different methods consulting engineers use to design room air diffusion and to select and size air diffusers, such as assumption of perfect mixing, design methods employing the empirical relations determined through research, such as the air diffusion performance index (ADPI), air jet theory and computational fluid dynamics (CFD) codes. Air supplied into the room through the various types of outlets (grills, ceiling mounted air diffusers, perforated panels etc.), is distributed by turbulent air jets. In mixing type air distribution systems, these air jets are the primary factor affecting room air motion. Numerous theoretical and experimental studies that developed a solid base for turbulent air jets theory were conducted concurrently in different countries (Germany, Sweden, Russia, U.K., USA) from the 1930`s through the 1980`s. Design methods based on air jet theory allows for the prediction of extreme values of air velocities and air temperatures in the occupied zone of empty spaces. Current air jet theory techniques account for the effects of buoyancy, confinement, jets interaction. For many conditions of jet discharge, it is possible to analyze jet performance and determine: the angle of divergence of the jet boundary; the velocity patterns along heated or chilled the jet axis; the

  11. Benchmark experiments on neutron streaming through JET Torus Hall penetrations

    Science.gov (United States)

    Batistoni, P.; Conroy, S.; Lilley, S.; Naish, J.; Obryk, B.; Popovichev, S.; Stamatelatos, I.; Syme, B.; Vasilopoulou, T.; contributors, JET

    2015-05-01

    Neutronics experiments are performed at JET for validating in a real fusion environment the neutronics codes and nuclear data applied in ITER nuclear analyses. In particular, the neutron fluence through the penetrations of the JET torus hall is measured and compared with calculations to assess the capability of state-of-art numerical tools to correctly predict the radiation streaming in the ITER biological shield penetrations up to large distances from the neutron source, in large and complex geometries. Neutron streaming experiments started in 2012 when several hundreds of very sensitive thermo-luminescence detectors (TLDs), enriched to different levels in 6LiF/7LiF, were used to measure the neutron and gamma dose separately. Lessons learnt from this first experiment led to significant improvements in the experimental arrangements to reduce the effects due to directional neutron source and self-shielding of TLDs. Here we report the results of measurements performed during the 2013-2014 JET campaign. Data from new positions, at further locations in the South West labyrinth and down to the Torus Hall basement through the air duct chimney, were obtained up to about a 40 m distance from the plasma neutron source. In order to avoid interference between TLDs due to self-shielding effects, only TLDs containing natural Lithium and 99.97% 7Li were used. All TLDs were located in the centre of large polyethylene (PE) moderators, with natLi and 7Li crystals evenly arranged within two PE containers, one in horizontal and the other in vertical orientation, to investigate the shadowing effect in the directional neutron field. All TLDs were calibrated in the quantities of air kerma and neutron fluence. This improved experimental arrangement led to reduced statistical spread in the experimental data. The Monte Carlo N-Particle (MCNP) code was used to calculate the air kerma due to neutrons and the neutron fluence at detector positions, using a JET model validated up to the

  12. Fullerene films and fullerene-dodecylamine adduct monolayers at air-water interfaces studied by neutron and x-ray reflection

    DEFF Research Database (Denmark)

    Wang, J.Y.; Vaknin, D.; Uphaus, R.A.

    1994-01-01

    Neutron and X-ray reflection measurements and surface pressure isotherms of spread films of the fullerene-dodecylamine adduct C60-[NH2(CH2)11CH3]x all indicate that this material may form monomolecular layers on water surfaces. The reflection data sets (neutron on both H2O and D2O) can be accounted...... stratum and the remainder alkyl tails are located close to both the air and the water interfaces. The alkyl moieties close to the aqueous substrate are hydrated. The reflection experiments and the isotherms suggest that on average 8 +/- 3 dodecylamine molecules are present per fullerene, consistent within...... error, with elemental analysis 5 +/- 2. By contrast, neutron reflection and surface pressure vs. area data of spread films of C60 fullerenes on aqueous surfaces indicate the formation of inhomogeneous multilayer films with a thickness and a surface roughness exceeding the molecular size...

  13. Calculations of Compound Nucleus Spin-Parity Distributions Populated via the (p,t Reaction in Support of Surrogate Neutron Capture Measurements

    Directory of Open Access Journals (Sweden)

    Benstead J.

    2016-01-01

    Full Text Available The surrogate reaction method may be used to determine the cross section for neutron induced reactions not accessible through standard experimental techniques. This is achieved by creating the same compound nucleus as would be expected in the desired reaction, but through a different incident channel, generally a direct transfer reaction. So far, the surrogate technique has been applied with reasonable success to determine the fission cross section for a number of actinides, but has been less successful when applied to other reactions, e.g. neutron capture, due to a ‘spin-parity mismatch’. This mismatch, between the spin and parity distributions of the excited levels of the compound nucleus populated in the desired and surrogate channels, leads to differing decay probabilities and hence reduces the validity of using the surrogate method to infer the cross section in the desired channel. A greater theoretical understanding of the expected distribution of levels excited in both the desired and surrogate channels is therefore required in order to attempt to address this mismatch and allow the method to be utilised with greater confidence. Two neutron transfer reactions, e.g. (p,t, which allow the technique to be utilised for isotopes further removed from the line of stability, are the subject of this study. Results are presented for the calculated distribution of compound nucleus states populated in 90Zr, via the 90Zr(p,t90Zr reaction, and are compared against measured data at an incident proton energy of 28.56 MeV.

  14. Hot-wire air flow meter for gasoline fuel-injection system. Calculation of air mass in cylinder during transient condition; Gasoline funsha system yo no netsusenshiki kuki ryuryokei. Kato untenji no cylinder juten kukiryo no keisan

    Energy Technology Data Exchange (ETDEWEB)

    Oyama, Y. [Hitachi Car Engineering, Ltd., Tokyo (Japan); Nishimura, Y.; Osuga, M.; Yamauchi, T. [Hitachi, Ltd., Tokyo (Japan)

    1997-10-01

    Air flow characteristics of hot-wire air flow meters for gasoline fuel-injection systems with supercharging and exhaust gas recycle during transient conditions were investigated to analyze a simple method for calculating air mass in cylinder. It was clarified that the air mass in cylinder could be calculated by compensating for the change of air mass in intake system by using aerodynamic models of intake system. 3 refs., 6 figs., 1 tab.

  15. Calculation of the ultracold neutron upscattering loss probability in fluid walled storage bottles using experimental measurements of the liquid thermomechanical properties of fomblin

    Science.gov (United States)

    Lamoreaux, S. K.; Golub, R.

    2002-10-01

    Presently, the most accurate values of the free neutron beta-decay lifetime result from measurements using fluid-coated ultacold neutron (UCN) storage bottles. The purpose of this work is to investigate the temperature-dependent UCN loss rate from these storage systems. To verify that the surface properites of fomblin films are the same as the bulk properties, we present experimental measurements of the properties of a liquid ``fomblin'' surface obtained by the quasielastic scattering of laser light. The properties include the surface tension and viscosity as functions of temperature. The results are compared to measurements of the bulk fluid properties. We then calculate the upscattering rate of UCNs from thermally excited surface capillary waves on the liquid surface and compare the results to experimental measurements of the UCN lifetime in fomblin-fluid-walled UCN storage bottles, and show that the excess storage loss rate for UCN energies near the fomblin potential can be explained. The rapid temperature dependence of the fomblin storage lifetime is explained by our analysis.

  16. A method for calculation of forces acting on air cooled gas turbine blades based on the aerodynamic theory

    Directory of Open Access Journals (Sweden)

    Grković Vojin R.

    2013-01-01

    Full Text Available The paper presents the mathematical model and the procedure for calculation of the resultant force acting on the air cooled gas turbine blade(s based on the aerodynamic theory and computation of the circulation around the blade profile. In the conducted analysis was examined the influence of the cooling air mass flow expressed through the cooling air flow parameter λc, as well as, the values of the inlet and outlet angles β1 and β2, on the magnitude of the tangential and axial forces. The procedure and analysis were exemplified by the calculation of the tangential and axial forces magnitudes. [Projekat Ministarstva nauke Republike Srbije: Development and building the demonstrative facility for combined heat and power with gasification

  17. Area balance method for calculation of air interchange in fire-resesistance testing laboratory for building products and constructions

    Directory of Open Access Journals (Sweden)

    Sargsyan Samvel Volodyaevich

    2014-09-01

    Full Text Available Fire-resistance testing laboratory for building products and constructions is a production room with a substantial excess heat (over 23 W/m . Significant sources of heat inside the aforementioned laboratory are firing furnace, designed to simulate high temperature effects on structures and products of various types in case of fire development. The excess heat production in the laboratory during the tests is due to firing furnaces. The laboratory room is considered as an object consisting of two control volumes (CV, in each of which there may be air intake and air removal, pollutant absorption or emission. In modeling air exchange conditions the following processes are being considered: the processes connected with air movement in the laboratory room: the jet stream in a confined space, distribution of air parameters, air motion and impurity diffusion in the ventilated room. General upward ventilation seems to be the most rational due to impossibility of using local exhaust ventilation. It is connected with the peculiarities of technological processes in the laboratory. Air jets spouted through large-perforated surface mounted at the height of 2 m from the floor level, "flood" the lower control volume, entrained by natural convective currents from heat sources upward and removed from the upper area. In order to take advantage of the proposed method of the required air exchange calculation, you must enter additional conditions, taking into account the provision of sanitary-hygienic characteristics of the current at the entrance of the service (work area. Exhaust air containing pollutants (combustion products, is expelled into the atmosphere by vertical jet discharge. Dividing ventilated rooms into two control volumes allows describing the research process in a ventilated room more accurately and finding the air exchange in the lab room during the tests on a more reasonable basis, allowing to provide safe working conditions for the staff without

  18. Hindered rotational energy barriers of BH4- tetrahedra in β-Mg(BH4)2 from quasielastic neutron scattering and DFT calculations

    DEFF Research Database (Denmark)

    Blanchard, Didier; Maronsson, Jon Bergmann; Riktor, M.D.

    2012-01-01

    In this work, hindered rotations of the BH4- tetrahedra in Mg(BH4)2 were studied by quasielastic neutron scattering, using two instruments with different energy resolution, in combination with density functional theory (DFT) calculations. Two thermally activated reorientations of the BH4- units......, around the 2-fold (C2) and 3-fold (C3) axes were observed at temperatures from 120 to 440 K. The experimentally obtained activation energies (EaC2 = 39 and 76 meV and EaC3 = 214 meV) and mean residence times between reorientational jumps are comparable with the energy barriers obtained from DFT...... calculations. A linear dependency of the energy barriers for rotations around the C2 axis parallel to the Mg-Mg axis with the distance between these two axes was revealed by the DFT calculations. At the lowest temperature (120 K) only 15% of the BH4- units undergo rotational motion and from comparison with DFT...

  19. Correction factors for the INER-improved free-air ionization chambers calculated with the Monte Carlo method.

    Science.gov (United States)

    Lin, Uei-Tyng; Chu, Chien-Hau

    2006-05-01

    Monte Carlo method was used to simulate the correction factors for electron loss and scattered photons for two improved cylindrical free-air ionization chambers (FACs) constructed at the Institute of Nuclear Energy Research (INER, Taiwan). The method is based on weighting correction factors for mono-energetic photons with X-ray spectra. The newly obtained correction factors for the medium-energy free-air chamber were compared with the current values, which were based on a least-squares fit to experimental data published in the NBS Handbook 64 [Wyckoff, H.O., Attix, F.H., 1969. Design of free-air ionization chambers. National Bureau Standards Handbook, No. 64. US Government Printing Office, Washington, DC, pp. 1-16; Chen, W.L., Su, S.H., Su, L.L., Hwang, W.S., 1999. Improved free-air ionization chamber for the measurement of X-rays. Metrologia 36, 19-24]. The comparison results showed the agreement between the Monte Carlo method and experimental data is within 0.22%. In addition, mono-energetic correction factors for the low-energy free-air chamber were calculated. Average correction factors were then derived for measured and theoretical X-ray spectra at 30-50 kVp. Although the measured and calculated spectra differ slightly, the resulting differences in the derived correction factors are less than 0.02%.

  20. Neutrons characterization of the nuclear reactor Ian-R1 of Colombia; Caracterizacion de los neutrones del reactor nuclear IAN-R1 de Colombia

    Energy Technology Data Exchange (ETDEWEB)

    Gonzalez P, L. X.; Martinez O, S. A. [Universidad Pedagogica y Tecnologica de Colombia, Grupo de Fisica Nuclear Aplicada y Simulacion, Carretera Central del Norte Km. 1, Via Paipa, 150003 Tunja, Boyaca (Colombia); Vega C, H. R., E-mail: s.agustin.martinez@uptc.edu.co [Universidad Autonoma de Zacatecas, Unidad Academica de Estudios Nucleares, Cipres No. 10, Fracc. La Penuela, 98068 Zacatecas (Mexico)

    2014-08-15

    By means of Monte Carlo methods, with the code MCNPX, the neutron characteristics of the research nuclear reactor Ian-R1 of Colombia, in power off but with the neutrons source in their start position, have been valued. The neutrons spectra, the total flow and their average power were calculated in the irradiation spaces inside the graphite reflector, as well as in the cells with air. Also the spectra, the total flow and the absorbed dose were calculated in several places distributed along the radial shaft inside the water moderator. The neutrons total flow was also considered to the long of the axial shaft. The characteristics of the neutrons spectra vary depending on their position regarding the source and the material that surrounds to the cell where the calculation was made. (Author)

  1. Evaluating methods for estimating space-time paths of individuals in calculating long-term personal exposure to air pollution

    Science.gov (United States)

    Schmitz, Oliver; Soenario, Ivan; Vaartjes, Ilonca; Strak, Maciek; Hoek, Gerard; Brunekreef, Bert; Dijst, Martin; Karssenberg, Derek

    2016-04-01

    Air pollution is one of the major concerns for human health. Associations between air pollution and health are often calculated using long-term (i.e. years to decades) information on personal exposure for each individual in a cohort. Personal exposure is the air pollution aggregated along the space-time path visited by an individual. As air pollution may vary considerably in space and time, for instance due to motorised traffic, the estimation of the spatio-temporal location of a persons' space-time path is important to identify the personal exposure. However, long term exposure is mostly calculated using the air pollution concentration at the x, y location of someone's home which does not consider that individuals are mobile (commuting, recreation, relocation). This assumption is often made as it is a major challenge to estimate space-time paths for all individuals in large cohorts, mostly because limited information on mobility of individuals is available. We address this issue by evaluating multiple approaches for the calculation of space-time paths, thereby estimating the personal exposure along these space-time paths with hyper resolution air pollution maps at national scale. This allows us to evaluate the effect of the space-time path and resulting personal exposure. Air pollution (e.g. NO2, PM10) was mapped for the entire Netherlands at a resolution of 5×5 m2 using the land use regression models developed in the European Study of Cohorts for Air Pollution Effects (ESCAPE, http://escapeproject.eu/) and the open source software PCRaster (http://www.pcraster.eu). The models use predictor variables like population density, land use, and traffic related data sets, and are able to model spatial variation and within-city variability of annual average concentration values. We approximated space-time paths for all individuals in a cohort using various aggregations, including those representing space-time paths as the outline of a persons' home or associated parcel

  2. A Comparison of Bulk Aerodynamic Methods for Calculating Air-Sea Flux

    National Research Council Canada - National Science Library

    Eleuterio, Daniel

    1998-01-01

    The Louis et al. (1982) bulk aerodynamic method for air-sea flux estimates is currently used in mesoscale models such as COAMPS, while the TOGA-COARE method is a state of the art flux parameterization involving recent...

  3. Cooling load calculations of radiant and all-air systems for commercial buildings

    OpenAIRE

    Bourdakis, Eleftherios; Bauman, Fred; Schiavon, Stefano; Raftery, Paul; Olesen, Bjarne W.

    2017-01-01

    The authors simulated in TRNSYS three radiant systems coupled with a 50% sized variable air volume (VAV) system and a 50% sized all-air VAV system with night ventilation. The objective of this study was to identify the differences in the cooling load profiles of the examined systems when they are sized based on different levels of the maximum cooling demand. The authors concluded that for high thermal mass radiant system nocturnal operation was adequate for providing an acceptable thermal env...

  4. Neutron spectra calculation and doses in a subcritical nuclear reactor based on thorium; Calculo de espectros de neutrones y dosis en un reactor nuclear subcritico a base de Torio

    Energy Technology Data Exchange (ETDEWEB)

    Medina C, D.; Hernandez A, P. L.; Hernandez D, V. M.; Vega C, H. R. [Universidad Autonoma de Zacatecas, Unidad Academica de Estudios Nucleares, Cipres No. 10, Fracc. La Penuela, 98068 Zacatecas, Zac. (Mexico); Sajo B, L., E-mail: dmedina_c@hotmail.com [Universidad Simon Bolivar, Laboratorio de Fisica Nuclear, Apdo. Postal 89000, Caracas 1080A (Venezuela, Bolivarian Republic of)

    2015-10-15

    This paper describes a heterogeneous subcritical nuclear reactor with molten salts based on thorium, with graphite moderator and a source of {sup 252}Cf, whose dose levels in the periphery allows its use in teaching and research activities. The design was done by the Monte Carlo method with the code MCNP5 where the geometry, dimensions and fuel was varied in order to obtain the best design. The result is a cubic reactor of 110 cm side with graphite moderator and reflector. In the central part they have 9 ducts that were placed in the direction of axis Y. The central duct contains the source of {sup 252}Cf, of 8 other ducts, are two irradiation ducts and the other six contain a molten salt ({sup 7}LiF - BeF{sub 2} - ThF{sub 4} - UF{sub 4}) as fuel. For design the k{sub eff}, neutron spectra and ambient dose equivalent was calculated. In the first instance the above calculation for a virgin fuel was called case 1, then a percentage of {sup 233}U was used and the percentage of Th was decreased and was called case 2. This with the purpose to compare two different fuels working inside the reactor. In the case 1 a value was obtained for the k{sub eff} of 0.13 and case 2 of 0.28, maintaining the subcriticality in both cases. In the dose levels the higher value is in case 2 in the axis Y with a value of 3.31 e-3 ±1.6% p Sv/Q this value is reported in for one. With this we can calculate the exposure time of personnel working in the reactor. (Author)

  5. Calculation of the radiation environment caused by galactic cosmic rays for determining air crew exposure

    CERN Document Server

    Ferrari, A; Rancati, T

    2001-01-01

    The spectra of secondary particles resulting from interactions of primary galactic cosmic rays with the nuclei in the atmosphere have been calculated using the Monte Carlo transport code FLUKA. The simulations have been carried out at solar minimum and solar maximum activity, for several values of the vertical geomagnetic cut-off. The effective dose rate and the ambient dose equivalent rate as a function of geomagnetic cut-off and altitude have been obtained using appropriate sets of conversion coefficients. The calculated results are discussed and compared with experimental data and other calculations. A simple method is proposed to calculate the radiation exposure at aircraft altitudes. (55 refs).

  6. Extensive air shower Monte Carlo modeling at the ground and aircraft flight altitude in the South Atlantic Magnetic Anomaly and comparison with neutron measurements

    Science.gov (United States)

    Pazianotto, M. T.; Cortés-Giraldo, M. A.; Federico, C. A.; Hubert, G.; Gonçalez, O. L.; Quesada, J. M.; Carlson, B. V.

    2017-02-01

    Modeling cosmic-ray-induced particle fluxes in the atmosphere is very important for developing many applications in aeronautics, space weather and on ground experimental arrangements. There is a lack of measurements and modeling at flight altitude and on ground in the South Atlantic Magnetic Anomaly. In this work we have developed an application based on the Geant4 toolkit called gPartAt that is aimed at the analysis of extensive air shower particle spectra. Another application has been developed using the MCNPX code with the same approach in order to evaluate the models and nuclear data libraries used in each application. Moreover, measurements were performed to determine the ambient dose equivalent rate of neutrons at flight altitude in different regions and dates in the Brazilian airspace; these results were also compared with the simulations. The results from simulations of the neutron spectra at ground level were also compared to data from a neutron spectrometer in operation since February 2015 at the Pico dos Dias Observatory in Brazil, at 1864 m above sea level, as part of a collaboration between the Institute for Advanced Studies (IEAv) and the French Aerospace Lab (ONERA). This measuring station is being operated with support from the National Astrophysics Laboratory (LNA). The modeling approaches were also compared to the AtmoRad computational platform, QARM, EXPACS codes and with measurements of the neutron spectrum taken in 2009 at the Pico dos Dias Observatory.

  7. Calculation and measurement of a neutral air flow velocity impacting a high voltage capacitor with asymmetrical electrodes

    Directory of Open Access Journals (Sweden)

    M. Malík

    2014-01-01

    Full Text Available This paper deals with the effects surrounding phenomenon of a mechanical force generated on a high voltage asymmetrical capacitor (the so called Biefeld-Brown effect. A method to measure this force is described and a formula to calculate its value is also given. Based on this the authors derive a formula characterising the neutral air flow velocity impacting an asymmetrical capacitor connected to high voltage. This air flow under normal circumstances lessens the generated force. In the following part this velocity is measured using Particle Image Velocimetry measuring technique and the results of the theoretically calculated velocity and the experimentally measured value are compared. The authors found a good agreement between the results of both approaches.

  8. Cooling load calculations of radiant and all-air systems for commercial buildings

    DEFF Research Database (Denmark)

    Bourdakis, Eleftherios; Bauman, Fred; Schiavon, Stefano

    The authors simulated in TRNSYS three radiant systems coupled with a 50% sized variable air volume (VAV) system and a 50% sized all-air VAV system with night ventilation. The objective of this study was to identify the differences in the cooling load profiles of the examined systems when they are......The authors simulated in TRNSYS three radiant systems coupled with a 50% sized variable air volume (VAV) system and a 50% sized all-air VAV system with night ventilation. The objective of this study was to identify the differences in the cooling load profiles of the examined systems when...... they are sized based on different levels of the maximum cooling demand. The authors concluded that for high thermal mass radiant system nocturnal operation was adequate for providing an acceptable thermal environment even when the radiant system was sized based on the 50% of the maximum cooling demand. The 50......% all-air system alone was able to provide comfort if night cooling was implemented. On the other hand, radiant cooling panels (low thermal mass) should be operating during the occupancy period. When sizing a high thermal mass radiant cooling system, the effect of thermal inertia and the response time...

  9. A method for development of efficient 3D models for neutronic calculations of ASTRA critical facility using experimental information

    Energy Technology Data Exchange (ETDEWEB)

    Balanin, A. L.; Boyarinov, V. F.; Glushkov, E. S.; Zimin, A. A.; Kompaniets, G. V.; Nevinitsa, V. A., E-mail: Neviniza-VA@nrcki.ru; Moroz, N. P.; Fomichenko, P. A.; Timoshinov, A. V. [National Research Center Kurchatov Institute (Russian Federation); Volkov, Yu. N. [National Research Nuclear University MEPhI (Russian Federation)

    2016-12-15

    The application of experimental information on measured axial distributions of fission reaction rates for development of 3D numerical models of the ASTRA critical facility taking into account azimuthal asymmetry of the assembly simulating a HTGR with annular core is substantiated. Owing to the presence of the bottom reflector and the absence of the top reflector, the application of 2D models based on experimentally determined buckling is impossible for calculation of critical assemblies of the ASTRA facility; therefore, an alternative approach based on the application of the extrapolated assembly height is proposed. This approach is exemplified by the numerical analysis of experiments on measurement of efficiency of control rods mockups and protection system (CPS).

  10. Calculation of the neutron flux and fluence in the covering of the nucleus and the vessel of a BWR; Calculo del flujo neutronico y fluencia en la envolvente del nucleo y la vasija de un reactor nuclear BWR

    Energy Technology Data Exchange (ETDEWEB)

    Martinez C, E.; Del Valle G, E. [IPN, Escuela Superior de Fisica y Matematicas, U. P. Adolfo Lopez Mateos, Col. Lindavista, 07738 Mexico D. F. (Mexico); Longoria G, L. C., E-mail: evalle@esfm.ipn.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2011-11-15

    One of the main objectives related with the safety in any nuclear power plant, including the nuclear power plant of Laguna Verde, is to guarantee the structural integrity of the pressure vessel of the reactor. To identify and quantifying the damage caused be neutron irradiation in the vessel of any nuclear reactor, is necessary to know as much the neutron flux as the fluence that it has been receiving during their time of operation life, since the observables damages by means of tests mechanics are products of micro-structural effects, induced by neutron irradiation, therefore, is important the study and prediction of the neutron flux to have a better knowledge of the damage that are receiving these materials. In our calculation the code DORT was used, which solves the transport equation in discreet coordinates and in two dimensions (x-y, r-{theta} and r-z), in accord to the regulator guide, it requires to make and approach of the neutron flux in three dimensions by means of the Synthesis Method. Whit this method is possible to achieve a representation of the flux in 3D combining or synthesizing the calculated fluxes by DORT code in r-{theta}, r-z and r. In this work the application of the Synthesis Method is presented, according to the Regulator Guide 1.190, to determine the fluxes 3D in the interns of a BWR using three different space meshes. (Author)

  11. An object-oriented 3D nodal finite element solver for neutron transport calculations in the Descartes project

    Energy Technology Data Exchange (ETDEWEB)

    Akherraz, B.; Lautard, J.J. [CEA Saclay, Dept. Modelisation de Systemes et Structures, Serv. d' Etudes des Reacteurs et de Modelisation Avancee (DMSS/SERMA), 91 - Gif sur Yvette (France); Erhard, P. [Electricite de France (EDF), Dir. de Recherche et Developpement, Dept. Sinetics, 92 - Clamart (France)

    2003-07-01

    In this paper we present two applications of the Nodal finite elements developed by Hennart and del Valle, first to three-dimensional Cartesian meshes and then to two-dimensional Hexagonal meshes. This work has been achieved within the framework of the DESCARTES project, which is a co-development effort by the 'Commissariat a l'Energie Atomique' (CEA) and 'Electricite de France' (EDF) for the development of a toolbox for reactor core calculations based on object oriented programming. The general structure of this project is based on the object oriented method. By using a mapping technique proposed in Schneider's thesis and del Valle, Mund, we show how this structuration allows us an easy implementation of the hexagonal case from the Cartesian case. The main attractiveness of this methodology is the possibility of a pin-by-pin representation by division of each lozenge into smaller ones. Furthermore, we will explore the use of non structured quadrangles to treat the circular geometry within a hexagon. It remains nevertheless, in the hexagonal case, the implementation of the acceleration of the internal iterations by the DSA (Diffusion Synthetic Acceleration) or the TSA. (authors)

  12. Matching Calculation and Performance Analysis for Aero Engine Starting Air Feeding System

    Science.gov (United States)

    2016-03-30

    APU-Pipeline-ATS” system is built based on air bleeding model of APU[11][12]. Influence of turbine guide inlet pressure limit on Inlet Guide Vane (IGV...Zhao, X.C. and Huang, X.H.. “Research on Inlet Guide Vane Regulating Rules and its Influence on Performance of Auxiliary Power Unit”. Journal of...computational accuracy of modeling method. Base on the cooperation system model, the influence of guide inlet pressure limit on cooperation points is

  13. The sensitivity of single air parcel trajectory calculations to starting elevation.

    Science.gov (United States)

    Saunders, Rolando O; Scotty, Erica; Kahl, Jonathan D W

    2013-10-01

    Trajectory models are frequently used to characterize the atmospheric transport pathways for airborne gases and aerosols. Users of these models must specify a starting elevation for their calculations. The variation of wind with altitude causes trajectory models to be sensitive to the starting elevation, particularly when single trajectories rather than Lagrangian particle dispersion simulations are used to characterize atmospheric transport. In this work we systematically investigate and quantify the sensitivity of single trajectory calculations to the starting elevation. The analysis was based on an eight-year database of daily, 48-h back-trajectories calculated for ten sites. Trajectories were calculated at four different starting elevations, and the horizontal difference between endpoints was determined for five upwind travel times. Trajectory model calculations were found to be strongly sensitive to starting elevation. A 500 m difference in starting elevation leads to an average horizontal separation of 326 km after 48 h. Mean horizontal separations of 627 km and 886 km were found for starting elevation differences of 1000 m and 1500 m, respectively. A seasonal dependence of the sensitivity was found, with the smallest separations occurring during the summer, the largest during winter, and intermediate values during the fall and spring. A linear relationship was observed between trajectory model sensitivity and difference in starting elevation. Empirical equations were presented to approximate this relationship. Copyright © 2013 Elsevier B.V. All rights reserved.

  14. Atlas of neutron resonances

    CERN Document Server

    Mughabghab, Said

    2018-01-01

    Atlas of Neutron Resonances: Resonance Properties and Thermal Cross Sections Z= 1-60, Sixth Edition, contains an extensive list of detailed individual neutron resonance parameters for Z=1-60, as well as thermal cross sections, capture resonance integrals, average resonance parameters and a short survey of the physics of thermal and resonance neutrons. The long introduction contains: nuclear physics formulas aimed at neutron physicists; topics of special interest such as valence neutron capture, nuclear level density parameters, and s-, p-, and d-wave neutron strength functions; and various comparisons of measured quantities with the predictions of nuclear models, such as the optical model. As in the last edition, additional features have been added to appeal to a wider spectrum of users. These include: spin-dependent scattering lengths that are of interest to solid-state physicists, nuclear physicists and neutron evaluators; calculated and measured Maxwellian average 5-keV and 30-keV capture cross sections o...

  15. Calculation of conversion coefficients using Chinese adult reference phantoms for air submersion and ground contamination.

    Science.gov (United States)

    Lu, Wei; Qiu, Rui; Wu, Zhen; Li, Chunyan; Yang, Bo; Liu, Huan; Ren, Li; Li, Junli

    2017-03-21

    The effective and organ equivalent dose coefficients have been widely used to provide assessment of doses received by adult members of the public and by workers exposed to environmental radiation from nuclear facilities under normal or accidental situations. Advancements in phantom types, weighting factors, decay data, etc, have led to the publication of newer results in this regard. This paper presents a new set of conversion coefficients for air submersion and ground contamination (with the use of Geant4) for photons from 15 keV to 10 MeV using the Chinese and International Commission on Radiological Protection (ICRP) adult reference male and female phantoms. The radiation fields, except for energy spectrum at low energies, were validated by the data obtained from the Monte Carlo code YURI. The effective dose coefficients of monoenergetic photons, obtained for the ICRP adult reference phantoms, agree well with recently published data for air submersion and ground contamination with a plane source at a depth of 0.5 g cm-2 in soil, but an average difference of 36.5% is observed for ground surface contamination with the abovementioned radiation field. The average differences in organ equivalent dose coefficients between the Chinese and the ICRP adult reference phantoms are within 6% for most organs, but noticeable differences of up to 70% or even higher are found at photon energies below 30 keV under air submersion. The effective dose coefficients obtained with the Chinese adult reference phantoms are greater than those of the ICRP adult reference phantoms above 30 keV and 0.5 MeV for ground contamination and air submersion, respectively; the average differences from the Chinese adult reference phantoms are about 3.6% and 0.4% in the whole energy range with maximum differences of 31.8% and 27.6% at 15 keV for air submersion and ground contamination respectively. These differences are attributed to anatomical discrepancies in overlying tissue mass of an

  16. Regional Contrasts of the Warming Rate over Land Significantly Depend on the Calculation Methods of Mean Air Temperature.

    Science.gov (United States)

    Wang, Kaicun; Zhou, Chunlüe

    2015-07-22

    Global analyses of surface mean air temperature (T(m)) are key datasets for climate change studies and provide fundamental evidences for global warming. However, the causes of regional contrasts in the warming rate revealed by such datasets, i.e., enhanced warming rates over the northern high latitudes and the "warming hole" over the central U.S., are still under debate. Here we show these regional contrasts depend on the calculation methods of T(m). Existing global analyses calculate T(m) from daily minimum and maximum temperatures (T2). We found that T2 has a significant standard deviation error of 0.23 °C/decade in depicting the regional warming rate from 2000 to 2013 but can be reduced by two-thirds using T(m) calculated from observations at four specific times (T4), which samples diurnal cycle of land surface air temperature more often. From 1973 to 1997, compared with T4, T2 significantly underestimated the warming rate over the central U.S. and overestimated the warming rate over the northern high latitudes. The ratio of the warming rate over China to that over the U.S. reduces from 2.3 by T2 to 1.4 by T4. This study shows that the studies of regional warming can be substantially improved by T4 instead of T2.

  17. Regional Contrasts of the Warming Rate over Land Significantly Depend on the Calculation Methods of Mean Air Temperature

    Science.gov (United States)

    Wang, Kaicun; Zhou, Chunlüe

    2015-01-01

    Global analyses of surface mean air temperature (Tm) are key datasets for climate change studies and provide fundamental evidences for global warming. However, the causes of regional contrasts in the warming rate revealed by such datasets, i.e., enhanced warming rates over the northern high latitudes and the “warming hole” over the central U.S., are still under debate. Here we show these regional contrasts depend on the calculation methods of Tm. Existing global analyses calculate Tm from daily minimum and maximum temperatures (T2). We found that T2 has a significant standard deviation error of 0.23 °C/decade in depicting the regional warming rate from 2000 to 2013 but can be reduced by two-thirds using Tm calculated from observations at four specific times (T4), which samples diurnal cycle of land surface air temperature more often. From 1973 to 1997, compared with T4, T2 significantly underestimated the warming rate over the central U.S. and overestimated the warming rate over the northern high latitudes. The ratio of the warming rate over China to that over the U.S. reduces from 2.3 by T2 to 1.4 by T4. This study shows that the studies of regional warming can be substantially improved by T4 instead of T2. PMID:26198976

  18. Activity-based Calculation Models for the Brazilian Air Force Cellular Unit of Intendancy

    Science.gov (United States)

    2013-03-01

    during the Operation "Goal 2006", to rescue 154 crash victims of a Boeing 737 of the company Gol Air Lines in the Amazon Rainforest. The team of...responsibility of the CUI, specified in the Manual of CUI (MMA 400-3/1976), namely: a) finance; b) providing supplies, classes: I-Material of Subsistence...for troops support, according to the doctrinal definition of the Cellular Unit of Intendancy Manual (MMA400-3/1976): “The Cellular Unit of

  19. Calculation and analysis of the mobility and diffusion coefficient of thermal electrons in methane/air premixed flames

    KAUST Repository

    Bisetti, Fabrizio

    2012-12-01

    Simulations of ion and electron transport in flames routinely adopt plasma fluid models, which require transport coefficients to compute the mass flux of charged species. In this work, the mobility and diffusion coefficient of thermal electrons in atmospheric premixed methane/air flames are calculated and analyzed. The electron mobility is highest in the unburnt region, decreasing more than threefold across the flame due to mixture composition effects related to the presence of water vapor. Mobility is found to be largely independent of equivalence ratio and approximately equal to 0.4m 2V -1s -1 in the reaction zone and burnt region. The methodology and results presented enable accurate and computationally inexpensive calculations of transport properties of thermal electrons for use in numerical simulations of charged species transport in flames. © 2012 The Combustion Institute.

  20. Validation of absolute axial neutron flux distribution calculations with MCNP with 197Au(n,γ)198Au reaction rate distribution measurements at the JSI TRIGA Mark II reactor.

    Science.gov (United States)

    Radulović, Vladimir; Štancar, Žiga; Snoj, Luka; Trkov, Andrej

    2014-02-01

    The calculation of axial neutron flux distributions with the MCNP code at the JSI TRIGA Mark II reactor has been validated with experimental measurements of the (197)Au(n,γ)(198)Au reaction rate. The calculated absolute reaction rate values, scaled according to the reactor power and corrected for the flux redistribution effect, are in good agreement with the experimental results. The effect of different cross-section libraries on the calculations has been investigated and shown to be minor. Copyright © 2013 Elsevier Ltd. All rights reserved.

  1. Properties of planetary ices in the NH3 + CO2 ± H2O ternary system using neutron diffraction and ab initio calculations

    Science.gov (United States)

    Howard, C. M.; Wood, I. G.; Fortes, A. D.; Vocadlo, L.

    2016-12-01

    BackgroundInteractions between simple molecules are of fundamental interest across diverse areas of the physical sciences, and the ternary system NH3 + CO2 ± H2O is no exception. In the outer solar system, interaction of CO2 with aqueous ammonia is likely to occur, synthesizing `rock-forming' minerals [1], with CO2 perhaps playing a role in ammonia-water oceans and cryomagmas inside icy planetary bodies - the discovery of ammonium carbonates in a crater of Pluto's moon Charon [2] adds weight to CO2 occuring in these planetary environments. In the same context, ammonium carbonates may have some astrobiological relevance, since removal of water leads to the formation of urea. On Earth, combination of CO2 with aqueous ammonia has relevance to carbon capture schemes [3], and there is interest in using such materials for hydrogen storage in fuel cells [4]. Consequently, from earthly matters of climate change to the study of extraterrestrial ices, understanding the structures and properties of ammonium carbonates are important. Despite this, our knowledge of ammonium carbonates is limited under ambient conditions of pressure and temperature and is entirely absent at the higher pressures, severely limiting our ability to model the behaviour of NH3 + CO2 ± H2O solids and fluids in planetary environments. ResultsWe report the results of several experiments using variable pressure and temperature neutron diffraction work on ammonium carbonate monohydrate, ammonium bicarbonate and ammonium carbamate, with complementary Density Functional Theory (DFT) calculations. The excellent agreement between experiments and DFT calculations obtained so far adds weight to the accuracy of calculated material properties of ammonium sesquicarbonate monohydrate and several polymorphs of urea where little empirical data exists. These experimental and computational studies provide the structural, thermoelastic and vibrational information required for accurate planetary modelling and remote

  2. Health effects from air pollution - calculation prices; Sundhedseffekter af luftforurening - beregningspriser

    Energy Technology Data Exchange (ETDEWEB)

    Skou Andersen, Mikael; Seested Nielsen, Jytte; Borgen Soerensen, Peter [DMU, Afdeling for Systemanalyse, Roskilde (Denmark); Frohn, Lise Marie; Solvang Jensen, Steen; Hertel, Ole; Brandt, Joergen; Christensen, Jesper [DMU, Afdeling for Atmosfaerisk Miljoe, Roskilde (Denmark)

    2004-10-01

    The basic aim of research focusing on the accounting of external effects is to provide estimates for the possible benefits of environmental policy projects. There is a demand within the field of environmental economic analysis for evaluations and estimates of the economic benefits arising from pollution reductions. Where benefits arising out of reduced emissions of harmful substances are concerned, these can be estimated according to the avoided damage costs associated with negative pollution impacts. A scientifically based method designed for this purpose has been developed in the pan-European ExternE project via the EU research programmes. The ExternE project has been implemented with the aim of estimating monetary values for externalities attached to air pollution accruing from energy production and transport. For this purpose, complex model-based environmental impact estimations have been coupled with corresponding monetary valuations via the modelling tool, Ecosense. A problem arising in relation to Ecosense methodology has been that externalities are often of local nature - pollution impact depending on the particular locality of the emission and dispersal characteristics in the surrounding environment. Therefore, the economic estimation of the impacts of pollution is based on a location-specific modelling system. (BA)

  3. Core calculations of JMTR

    Energy Technology Data Exchange (ETDEWEB)

    Nagao, Yoshiharu [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment

    1998-03-01

    In material testing reactors like the JMTR (Japan Material Testing Reactor) of 50 MW in Japan Atomic Energy Research Institute, the neutron flux and neutron energy spectra of irradiated samples show complex distributions. It is necessary to assess the neutron flux and neutron energy spectra of an irradiation field by carrying out the nuclear calculation of the core for every operation cycle. In order to advance core calculation, in the JMTR, the application of MCNP to the assessment of core reactivity and neutron flux and spectra has been investigated. In this study, in order to reduce the time for calculation and variance, the comparison of the results of the calculations by the use of K code and fixed source and the use of Weight Window were investigated. As to the calculation method, the modeling of the total JMTR core, the conditions for calculation and the adopted variance reduction technique are explained. The results of calculation are shown. Significant difference was not observed in the results of neutron flux calculations according to the difference of the modeling of fuel region in the calculations by K code and fixed source. The method of assessing the results of neutron flux calculation is described. (K.I.)

  4. Measured Neutron Spectra and Dose Equivalents From a Mevion Single-Room, Passively Scattered Proton System Used for Craniospinal Irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Howell, Rebecca M., E-mail: rhowell@mdanderson.org [Department of Radiation Physics, The University of Texas M. D. Anderson Cancer Center, Houston, Texas (United States); Burgett, Eric A.; Isaacs, Daniel [Department of Nuclear Engineering, Idaho State University, Pocatello, Idaho (United States); Price Hedrick, Samantha G.; Reilly, Michael P.; Rankine, Leith J.; Grantham, Kevin K.; Perkins, Stephanie; Klein, Eric E. [Department of Radiation Oncology, Washington University, St. Louis, Missouri (United States)

    2016-05-01

    Purpose: To measure, in the setting of typical passively scattered proton craniospinal irradiation (CSI) treatment, the secondary neutron spectra, and use these spectra to calculate dose equivalents for both internal and external neutrons delivered via a Mevion single-room compact proton system. Methods and Materials: Secondary neutron spectra were measured using extended-range Bonner spheres for whole brain, upper spine, and lower spine proton fields. The detector used can discriminate neutrons over the entire range of the energy spectrum encountered in proton therapy. To separately assess internally and externally generated neutrons, each of the fields was delivered with and without a phantom. Average neutron energy, total neutron fluence, and ambient dose equivalent [H* (10)] were calculated for each spectrum. Neutron dose equivalents as a function of depth were estimated by applying published neutron depth–dose data to in-air H* (10) values. Results: For CSI fields, neutron spectra were similar, with a high-energy direct neutron peak, an evaporation peak, a thermal peak, and an intermediate continuum between the evaporation and thermal peaks. Neutrons in the evaporation peak made the largest contribution to dose equivalent. Internal neutrons had a very low to negligible contribution to dose equivalent compared with external neutrons, largely attributed to the measurement location being far outside the primary proton beam. Average energies ranged from 8.6 to 14.5 MeV, whereas fluences ranged from 6.91 × 10{sup 6} to 1.04 × 10{sup 7} n/cm{sup 2}/Gy, and H* (10) ranged from 2.27 to 3.92 mSv/Gy. Conclusions: For CSI treatments delivered with a Mevion single-gantry proton therapy system, we found measured neutron dose was consistent with dose equivalents reported for CSI with other proton beamlines.

  5. Calculation of the Arc Velocity Along the Polluted Surface of Short Glass Plates Considering the Air Effect

    Directory of Open Access Journals (Sweden)

    Tao Yuan

    2012-03-01

    Full Text Available To investigate the microphysics mechanism and the factors that influence arc development along a polluted surface, the arc was considered as a plasma fluid. Based on the image method and the collision ionization theory, the electric field of the arc needed to maintain movement with different degrees of pollution was calculated. According to the force of the charged particle in an arc plasma stressed under an electric field, a calculation model of arc velocity, which is dependent on the electric field of the arc head that incorporated the effects of airflow around the electrode and air resistance is presented. An experiment was carried out to measure the arc velocity, which was then compared with the calculated value. The results of the experiment indicated that the lighter the pollution is, the larger the electric field of the arc head and arc velocity is; when the pollution is heavy, the effect of thermal buoyancy that hinders arc movement increases, which greatly reduces the arc velocity.

  6. Neutronic reactor

    Science.gov (United States)

    Wende, Charles W. J.; Babcock, Dale F.; Menegus, Robert L.

    1983-01-01

    A nuclear reactor includes an active portion with fissionable fuel and neutron moderating material surrounded by neutron reflecting material. A control element in the active portion includes a group of movable rods constructed of neutron-absorbing material. Each rod is movable with respect to the other rods to vary the absorption of neutrons and effect control over neutron flux.

  7. Neutronic reactor

    Energy Technology Data Exchange (ETDEWEB)

    Babcock, D.F.; Menegus, R.L.; Wende, C.W.

    1983-01-04

    A nuclear reactor includes an active portion with fissionable fuel and neutron moderating material surrounded by neutron reflecting material. A control element in the active portion includes a group of movable rods constructed of neutron-absorbing material. Each rod is movable with respect to the other rods to vary the absorption of neutrons and effect control over neutron flux.

  8. Method for calculating steady-state waves in an air cushion vehicle. Part 2; Air cushion vehicle no teijo zoha keisanho ni tsuite. 2

    Energy Technology Data Exchange (ETDEWEB)

    Eguchi, T. [Mitsui Engineering and Shipbuilding Co. Ltd., Tokyo (Japan)

    1997-10-01

    Discussions were given on a method to estimate resistance constituents in wave resistance made in an air chamber of an air cushion vehicle (ACV). An orthogonal coordinate system is considered, which uses the center of a hull as the zero point and is made dimensionless by using cushion length. Flow around the ACV is supposed as an ideal flow, whereas speed potential is defined in the flow field. Then, a linear free surface condition is hypothesized on water surface Z = 0. Number and density of waves were used to introduce a condition to be satisfied by the speed potential. A numerical calculation method arranged a blow-out panel on the water surface, and used a panel shift type Rankine source method which satisfies the free surface condition at Z = 0. Cushion pressure distribution becomes a step-like discontinuous function, and mathematical infinity is generated in the differentiation values. Under an assumption that the pressure rises per one panel where pressure jump is present, the distribution was approximated by providing one panel with inclination of the finite quantity therein. Estimation on wave height distribution in the cushion chamber showed a tendency of qualitatively agreeing with the experimental result, but the wave heights shown in the experiment had the average level decreased as it goes toward the rear of the hull. 5 refs., 5 figs.

  9. Radiation doses from radiation sources of neutrons and photons by different computer calculation; Tecniche di calcolo di intensita` di dose da sorgenti di radiazione neutronica e fotonica con l`uso di codici basati su metodologie diverse

    Energy Technology Data Exchange (ETDEWEB)

    Siciliano, F.; Lippolis, G.; Bruno, S.G. [ENEA, Centro Ricerche Trisaia, Rotondella (Italy)

    1995-11-01

    In the present paper the calculation technique aspects of dose rate from neutron and photon radiation sources are covered with reference both to the basic theoretical modeling of the MERCURE-4, XSDRNPM-S and MCNP-3A codes and from practical point of view performing safety analyses of irradiation risk of two transportation casks. The input data set of these calculations -regarding the CEN 10/200 HLW container and dry PWR spent fuel assemblies shipping cask- is frequently commented as for as connecting points of input data and understanding theoretic background are concerned.

  10. A neutron spectrum unfolding code based on iterative procedures

    Energy Technology Data Exchange (ETDEWEB)

    Ortiz R, J. M.; Vega C, H. R., E-mail: morvymm@yahoo.com.mx [Universidad Autonoma de Zacatecas, Unidad Academica de Ingenieria Electrica, Apdo. Postal 336, 98000 Zacatecas (Mexico)

    2012-10-15

    In this work, the version 3.0 of the neutron spectrum unfolding code called Neutron Spectrometry and Dosimetry from Universidad Autonoma de Zacatecas (NSDUAZ), is presented. This code was designed in a graphical interface under the LabVIEW programming environment and it is based on the iterative SPUNIT iterative algorithm, using as entrance data, only the rate counts obtained with 7 Bonner spheres based on a {sup 6}Lil(Eu) neutron detector. The main features of the code are: it is intuitive and friendly to the user; it has a programming routine which automatically selects the initial guess spectrum by using a set of neutron spectra compiled by the International Atomic Energy Agency. Besides the neutron spectrum, this code calculates the total flux, the mean energy, H(10), h(10), 15 dosimetric quantities for radiation protection porpoises and 7 survey meter responses, in four energy grids, based on the International Atomic Energy Agency compilation. This code generates a full report in html format with all relevant information. In this work, the neutron spectrum of a {sup 241}AmBe neutron source on air, located at 150 cm from detector, is unfolded. (Author)

  11. Dose calculation in biological samples in a mixed neutron-gamma field at the TRIGA reactor of the University of Mainz

    DEFF Research Database (Denmark)

    Schmitz, T.; Blaickner, M.; Schütz, C.

    2010-01-01

    FLUKA, a multipurpose Monte Carlo transport code. The pin-diode is augmented by a lithium fluoride foil. This foil converts the neutrons into alpha and tritium particles which are products of the (7)Li(n,α)(3)H-reaction. These particles are detected by the diode and their amount correlates...

  12. Measurement of neutron dose in the compensator IMRT treatment.

    Science.gov (United States)

    Rezaian, Abbas; Nedaie, Hassan Ali; Banaee, Nooshin

    2017-10-01

    A radiation treatment delivery technique, intensity modulated radiation therapy (IMRT), has found widespread use in the treatment of cancers. One of IMRT implementing methods is IMRT compensator based, which the modulation are done by high Z materials. When photons with energies higher than 8MV interact with high Z material in path, Photoneutrons are produced. In this study, the effect of compensator on photoneutron production was investigated. The Monte Carlo code MCNPX was used to calculate the neutron dose equivalent as a function of the depth in phantom with and without compensator. Measurements were made using CR-39 track-etched detectors. CR-39 detectors, were cut in dimensions of 2.5×2.5 cm 2 by laser, placed in different depths of slab phantom and then irradiated by 18MV photons. Same procedure was performed with the compensator present and absent. The measured data were compared with MCNP calculations. In both experimental and simulation results, neutron dose equivalent when compensator used, was less than non-compensator field. The calculated neutron dose equivalent was maximum at surface and decreased exponentially by increasing depth, but in experimental data, the neutron dose equivalent reached a maximum at approximately 3cm depth in the phantom and beyond which decreased with depth.CR-39 calibration was carried out in air, by considering that neutron energy spectrum changes toward thermal neutrons by depth in phantom increasing, it is suggested that for measuring equivalent neutron dose at phantom depth, should have proper neutron calibration in terms of energy spectrum. Copyright © 2017 Elsevier Ltd. All rights reserved.

  13. Calculation of air pollutant emissions for episodes in Europe. Final report; Ermittlung von Luftschadstoffemissionen fuer Episoden in Europa. Abschlussbericht

    Energy Technology Data Exchange (ETDEWEB)

    Friedrich, R.; Lenhart, L.; Heck, T.

    1996-11-01

    For the further investigation of the large scale transport and chemical transformation of air pollutants with atmospheric models reliable emission data with a high temporal, spatial and species resolution are required for Europe. Within the GENEMIS project methods and models have been developed to generate such data, starting from annual emission data of the CORINAIR and LOTOS inventories. The main objective of converting annual emission data into reliable hourly values has been reached on the basis of numerous socio-economical, meteorological and technical indicator data. The models which were developed within this framework have been used to generate european emission inventories for different base years and selected episods. The analysis of the results showed significant spatial and large temporal variations of the emissions. It has been shown that former simplifying assumptions concerning the temporal behaviour of emissions were not sufficient to describe real situations. The emission data calculated within GENEMIS have been used directly as an input for air quality modeling purposes. (orig.) [Deutsch] Zur weiteren Erforschung des grossraeumigen atmosphaerischen Transports und der chemischen Umwandlung von Luftverunreinigungen mit Hilfe von entsprechenden Simulationsmodellen werden zeitlich, raeumlich und stofflich hochaufgeloeste Emissionsdaten fuer Europa benoetigt. Im Rahmen des Forschungsvorhabens GENEMIS wurden Methoden und Modelle entwickelt, mit denen - ausgehend von den jaehrlichen Emissionskatastern von CORINAIR und LOTOS - hochaufgeloeste Emissionsdaten in Europa ermittelt werden koennen. Die im Vordergrund stehende zeitliche Disaggregierung in stuendliche Emissionswerte wurde dabei auf der Grundlage zahlreicher sozio-oekonomischer, meteorologischer und technischer Indikatordaten vorgenommen. Die entwickelten Modelle wurden zur Erstellung von hochaufgeloesten Emissionsdatensaetzen fuer verschiedene Bezugsjahre und -zeitraeume angewendet. In der

  14. Measurement of the energy spectrum of cosmic-ray induced neutrons aboard an ER-2 high-altitude airplane

    CERN Document Server

    Goldhagen, P E; Kniss, T; Reginatto, M; Singleterry, R C; Van Steveninck, W; Wilson, J W

    2002-01-01

    Crews working on present-day jet aircraft are a large occupationally exposed group with a relatively high average effective dose from galactic cosmic radiation. Crews of future high-speed commercial aircraft flying at higher altitudes would be even more exposed. To help reduce the significant uncertainties in calculations of such exposures, the atmospheric ionizing radiation (AIR) project, an international collaboration of 15 laboratories, made simultaneous radiation measurements with 14 instruments on five flights of a NASA ER-2 high-altitude aircraft. The primary AIR instrument was a highly sensitive extended-energy multisphere neutron spectrometer with lead and steel shells placed within the moderators of two of its 14 detectors to enhance response at high energies. Detector responses were calculated for neutrons and charged hadrons at energies up to 100 GeV using MCNPX. Neutron spectra were unfolded from the measured count rates using the new MAXED code. We have measured the cosmic-ray neutron spectrum (t...

  15. On the neutron bursts origin.

    CERN Document Server

    Stenkin, Yu V

    2002-01-01

    The origin of the neutron bursts in Extensive Air Showers (EAS) is explained using results of the experiments and CORSIKA based Monte-Carlo simulations. It is shown that events with very high neutron multiplicity observed last years in neutron monitors as well as in surrounding detectors, are caused by usual EAS core with primary energies > 1 PeV. No exotic processes were needed for the explanation.

  16. Air

    Science.gov (United States)

    ... and your health: Green living Sun Water Air Health effects of air pollution How to protect yourself from air pollution Chemicals Noise Quizzes Links to more information girlshealth glossary girlshealth. ...

  17. The Effect of Air/Sea Exchange and Mixing on Organic Carbon Export Calculated from an Oxygen Mass Balance

    Science.gov (United States)

    Hamme, R. C.; Emerson, S. R.

    2002-12-01

    The production of organic carbon, and its export from the upper ocean, is a major control on the CO2 concentration in the atmosphere and thus an important determinant of the earth's climate. The flux of organic carbon from the euphotic zone can be calculated from an upper ocean oxygen mass balance if the rates of physical processes that influence oxygen can be constrained. We present an intensive one-year dataset of oxygen, nitrogen, argon and neon measurements collected at the Hawaii Ocean Time-series (HOT) from July 2000 to June 2001. Oxygen is supersaturated in the surface waters during the entire year due to a combination of biological and physical effects, such as heating and bubble-mediated gas exchange. We use a one dimensional dynamic mixed layer (PWP) model, driven by local heat flux and wind speed estimates, to examine the processes that control gas concentrations. The observed inert gas measurements are used to constrain the rates of bubble-mediated gas exchange by different bubble mechanisms and vertical mixing within the model. Diffusive gas exchange is calculated from wind speed. The model-derived rates of the physical processes combined with the observed oxygen concentrations yield a net biological oxygen production of 1.6 +/- 0.8 mol O2/m2/yr (1.1 +/- 0.6 mol C/m2/yr). Refinements to this provisional estimate and an analysis of its sensitivity to the rates of air/sea exchange and mixing will be presented at the meeting.

  18. A tool for calculation of 7Li(p,n)7Be neutron source spectra below the three-body break-up reaction threshold

    Science.gov (United States)

    Pachuau, Rebecca; Lalremruata, B.; Otuka, N.; Hlondo, L. R.; Punte, L. R. M.; Thanga, H. H.

    2017-09-01

    We developed a new deterministic neutron source spectrum code EPEN - Energy of Proton Energy of Neutron - for a given lithium target thickness, sample angular coverage and proton energy from the reaction threshold to the three-body break-up threshold. The angular differential cross sections of the 7Li(p,n0)7Be and 7Li(p,n1)7Be reactions evaluated by Liskien and Paulsen were adopted above 1.95 MeV while the functional form suggested by Macklin and Gibbons was adopted for the 7Li(p,n0)7Be reaction cross section near threshold. The spectra obtained by EPEN are validated by the experimental spectra and also compared with the spectra predicted by two Monte Carlo codes, SimLiT and PINO. The results of comparison are discussed in detail.

  19. INFLUENCE OF SCATTERED NEUTRON RADIATION ON METROLOGICAL CHARACTERISTICS OF АТ140 NEUTRON CALIBRATION FACILITY

    Directory of Open Access Journals (Sweden)

    D. I. Komar

    2017-01-01

    Full Text Available Today facilities with collimated radiation field are widely used as reference in metrological support of devices for neutron radiation measurement. Neutron fields formed by radionuclide neutron sources. The aim of this research was to study characteristics of experimentally realized neutron fields geometries on АТ140 Neutron Calibration Facility using Monte Carlo method.For calibration, we put a device into neutron field with known flux density or ambient equivalent dose rate. We can form neutron beam from radionuclide fast-neutron source in different geometries. In containercollimator of АТ140 Neutron Calibration Facility we can install special inserts to gather fast-neutron geometry or thermal-neutron geometry. We need to consider neutron scattering from air and room’s walls. We can conduct measurements of neutron field characteristics in several points and get the other using Monte Carlo method.Thermal neutron collimator forms a beam from radionuclide source with a significant amount of neutrons with thermal energies. From found relationship between full neutron flux and distance to neutron source we see that inverse square law is violated. Scattered radiation contribution into total flux increases when we are moving away from neutron source and significantly influences neutron fields characteristics. While source is exposed in shadow-cone geometry neutron specter has pronounced thermal component from wall scattering.In this work, we examined main geometry types used to acquire reference neutron radiation using radionuclide sources. We developed Monte Carlo model for 238Pu-Be neutron source and АТ140 Neutron Calibration Facility’s container-collimator. We have shown the most significant neutron energy distribution factor to be scattered radiation from room’s walls. It leads to significant changes of neutron radiation specter at a distance from the source. When planning location, and installing the facility we should consider

  20. Measurement and Calculation of High-Energy Neutron Spectra behind Shielding at the CERF 120 GeV/c Hadron Beam Facility

    CERN Document Server

    Nakao, N; Roesler, S; Brugger, M; Hagiwara, M; Vincke, H; Khater, H; Prinz, A A; Rokni, S H; Kosako, K

    2008-01-01

    Neutron energy spectra were measured behind the lateral shield of the CERF (CERN-EU High Energy Reference Field) facility at CERN with a 120 GeV/c positive hadron beam (a mixture of mainly protons and pions) on a cylindrical copper target (7-cm diameter by 50-cm long). An NE213 organic liquid scintillator (12.7-cm diameter by 12.7-cm long) was located at various longitudinal positions behind shields of 80- and 160-cm thick concrete and 40-cm thick iron. The measurement locations cover an angular range with respect to the beam axis between 13 and 133 degrees. Neutron energy spectra in the energy range between 32 MeV and 380 MeV were obtained by unfolding the measured pulse height spectra with the detector response functions which have been verified in the neutron energy range up to 380 MeV in separate experiments. Since the source term and experimental geometry in this experiment are well characterized and simple, and results are given in the form of energy spectra, these experimental results are very useful a...

  1. Neutron energy measurement for practical applications

    Indian Academy of Sciences (India)

    M V Roshan

    2018-02-07

    orbit coupling and ... High-energy neutron spectrometry from fusion reactors, plasma focus devices, and the accelerators with .... Rigorous calculations based on physics concepts prove that neutron energies can be obtained by ...

  2. Neutron resonance spectroscopy

    Energy Technology Data Exchange (ETDEWEB)

    Gunsing, F

    2005-06-15

    The present document has been written in order to obtain the diploma 'Habilitation a Diriger des Recherches'. Since this diploma is indispensable to supervise thesis students, I had the intention to write a document that can be useful for someone starting in the field of neutron resonance spectroscopy. Although the here described topics are already described elsewhere, and often in more detail, it seemed useful to have most of the relevant information in a single document. A general introduction places the topic of neutron-nucleus interaction in a nuclear physics context. The large variations of several orders of magnitude in neutron-induced reaction cross sections are explained in terms of nuclear level excitations. The random character of the resonances make nuclear model calculation predictions impossible. Then several fields in physics where neutron-induced reactions are important and to which I have contributed in some way or another, are mentioned in a first synthetic chapter. They concern topics like parity nonconservation in certain neutron resonances, stellar nucleosynthesis by neutron capture, and data for nuclear energy applications. The latter item is especially important for the transmutation of nuclear waste and for alternative fuel cycles. Nuclear data libraries are also briefly mentioned. A second chapter details the R-matrix theory. This formalism is the foundation of the description of the neutron-nucleus interaction and is present in all fields of neutron resonance spectroscopy. (author)

  3. Neutronics Benchmarks for the Utilization of Mixed-Oxide Fuel: Joint U.S./Russian Progress Report for Fiscal Year 1997 Volume 2-Calculations Performed in the United States

    Energy Technology Data Exchange (ETDEWEB)

    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.

  4. Application of Laplace transform for determination of albedo type boundary conditions for neutronic calculations; Aplicacao da transformada de Laplace para determinacao de condicoes de contorno tipo albedo para calculos neutronicos

    Energy Technology Data Exchange (ETDEWEB)

    Petersen, Claudio Zen

    2008-07-01

    In this dissertation we use the Laplace transform to derive expressions for nonstandard albedo boundary conditions for one and two non-multiplying regions at the ends of one dimensional domains. In practice, the fuel regions of reactor cores are surrounded by reflector regions that reduce neutron leakage. In order to exclude the reflector regions from the calculations, we introduce a reflection coefficient or albedo. We use the present albedo boundary conditions to solve numerically slab-geometry monoenergetic and multigroup diffusion equations using the conventional finite difference method. Numerical results are generated for fixed source and eigenvalue diffusion problems in slab geometry(author)

  5. Passive neutron assay of irradiated nuclear fuels

    Energy Technology Data Exchange (ETDEWEB)

    Hsue, S.T.; Stewart, J.E.; Kaieda, K.; Halbig, J.K.; Phillips, J.R.; Lee, D.M.; Hatcher, C.R.

    1979-02-01

    Passive neutron assay of irradiated nuclear fuel has been investigated by calculations and experiments as a simple, complementary technique to the gamma assay. From the calculations it was found that the neutron emission arises mainly from the curium isotopes, the neutrons exhibit very good penetrability of the assemblies, and the neutron multiplication is not affected by the burnup. From the experiments on BWR and PWR assemblies, the neutron emission rate is proportional to burnup raised to 3.4 power. The investigations indicate that the passive neutron assay is a simple and useful technique to determine the consistency of burnups between assemblies.

  6. Hyperons in neutron stars

    Directory of Open Access Journals (Sweden)

    Tetsuya Katayama

    2015-07-01

    Full Text Available Using the Dirac–Brueckner–Hartree–Fock approach, the properties of neutron-star matter including hyperons are investigated. In the calculation, we consider both time and space components of the vector self-energies of baryons as well as the scalar ones. Furthermore, the effect of negative-energy states of baryons is partly taken into account. We obtain the maximum neutron-star mass of 2.08M⊙, which is consistent with the recently observed, massive neutron stars. We discuss a universal, repulsive three-body force for hyperons in matter.

  7. Nanostructure of polymer monolayer and polyelectrolyte brush at air/water interface by X-ray and neutron reflectometry

    CERN Document Server

    Matsuoka, H; Matsumoto, K

    2003-01-01

    The nanostructure of amphiphilic diblock copolymer monolayer on water was directly investigated by in situ X-ray and neutron reflectivity techniques. The diblock copolymer consists of polysilacyclobutane, which is very flexible, as a hydrophobic block and polymethacrylic acid, an anionic polymer, as a hydrophilic block. The polymers with shorter hydrophilic segment formed a very smooth and uniform monolayer with hydrophobic layer on water and dense hydrophilic layer under the water. But the longer hydrophilic segment polymer formed three-layered monolayer with polyelectrolyte brush in addition to hydrophobic and dense hydrophilic layers. The dense hydrophilic layer is thought to be formed to avoid a contact between hydrophobic polymer layer and water. Its role is something like a 'carpet'. An additional interesting information is that the thickness of the 'carpet layer' is almost 15A, independent the surface pressure and hydrophilic polymer length. Highly quantitative information was obtained about the nanost...

  8. Calculation of the inventory and near-field release rates of radioactivity from neutron-activated metal parts discharged from the high flux isotope reactor and emplaced in solid waste storage area 6 at Oak Ridge National Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Kelmers, A.D.; Hightower, J.R.

    1987-05-01

    Emplacement of contaminated reactor components involves disposal in lined and unlined auger holes in soil above the water table. The radionuclide inventory of disposed components was calculated. Information on the composition and weight of the components, as well as reasonable assumptions for the neutron flux fueling use, the time of neutron exposure, and radioactive decay after discharge, were employed in the inventory calculation. Near-field release rates of /sup 152/Eu, /sup 154/Eu, and /sup 155/Eu from control plates and cylinders were calculated for 50 years after emplacement. Release rates of the europium isotopes were uncertain. Two release-rate-limiting models were considered and a range of reasonable values were assumed for the time-to-failure of the auger-hole linear and aluminum cladding and europium solubility in SWSA-6 groundwater. The bounding europium radionuclide near-field release rates peaked at about 1.3 Ci/year total for /sup 152,154,155/Eu in 1987 for the lower bound, and at about 420 Ci/year in 1992 for the upper bound. The near-field release rates of /sup 55/Fe, /sup 59/Ni, /sup 60/Co, and /sup 63/Ni from stainless steel and cobalt alloy components, as well as of /sup 10/Be, /sup 41/Ca, and /sup 55/Fe from beryllium reflectors, were calculated for the next 100 years, assuming bulk waste corrosion was the release-rate-limiting step. Under the most conservative assumptions for the reflectors, the current (1986) total radionuclide release rate was calculated to be about 1.2 x 10/sup -4/ Ci/year, decreasing by 1992 to a steady release of about 1.5 x 10/sup -5/ Ci/year due primarily to /sup 41/Ca. 50 refs., 13 figs., 8 tabs.

  9. Measurement and calculation of high-energy neutron spectra behind shielding at the CERF 120 GeV/c hadron beam facility

    Energy Technology Data Exchange (ETDEWEB)

    Nakao, N. [Stanford Linear Accelerator Center (SLAC), Stanford, CA 94309 (United States)], E-mail: nakao@fnal.gov; Taniguchi, S. [Japan Synchrotron Radiation Research Institute (JASRI), Hyogo 679-5198 (Japan); Roesler, S.; Brugger, M. [CERN, CH-1211 Geneva 23 (Switzerland); Hagiwara, M. [Cyclotron Radioisotope Center (CYRIC), Tohoku University, Sendai 980-8579 (Japan); Vincke, H. [Stanford Linear Accelerator Center (SLAC), Stanford, CA 94309 (United States); CERN, CH-1211 Geneva 23 (Switzerland); Khater, H.; Prinz, A.A.; Rokni, S.H. [Stanford Linear Accelerator Center (SLAC), Stanford, CA 94309 (United States); Kosako, K. [Shimizu Corporation, Tokyo (Japan)

    2008-01-15

    Neutron energy spectra were measured behind the lateral shield of the CERF (CERN-EU High Energy Reference Field) facility at CERN with a 120 GeV/c positive hadron beam (a mixture of mainly protons and pions) on a cylindrical copper target (7-cm diameter by 50-cm long). An NE213 organic liquid scintillator (12.7-cm diameter by 12.7-cm long) was located at various longitudinal positions behind shields of 80- and 160-cm thick concrete and 40-cm thick iron. The measurement locations cover an angular range with respect to the beam axis between 13 and 133{sup o}. Neutron energy spectra in the energy range between 32 MeV and 380 MeV were obtained by unfolding the measured pulse height spectra with the detector response functions which have been verified in the neutron energy range up to 380 MeV in separate experiments. Since the source term and experimental geometry in this experiment are well characterized and simple and results are given in the form of energy spectra, these experimental results are very useful as benchmark data to check the accuracies of simulation codes and nuclear data. Monte Carlo simulations of the experimental set up were performed with the FLUKA, MARS and PHITS codes. Simulated spectra for the 80-cm thick concrete often agree within the experimental uncertainties. On the other hand, for the 160-cm thick concrete and iron shield differences are generally larger than the experimental uncertainties, yet within a factor of 2. Based on source term simulations, observed discrepancies among simulations of spectra outside the shield can be partially explained by differences in the high-energy hadron production in the copper target.

  10. Measurement and calculation of high-energy neutron spectra behind shielding at the CERF 120 GeV/c hadron beam facility

    Science.gov (United States)

    Nakao, N.; Taniguchi, S.; Roesler, S.; Brugger, M.; Hagiwara, M.; Vincke, H.; Khater, H.; Prinz, A. A.; Rokni, S. H.; Kosako, K.

    2008-01-01

    Neutron energy spectra were measured behind the lateral shield of the CERF (CERN-EU High Energy Reference Field) facility at CERN with a 120 GeV/c positive hadron beam (a mixture of mainly protons and pions) on a cylindrical copper target (7-cm diameter by 50-cm long). An NE213 organic liquid scintillator (12.7-cm diameter by 12.7-cm long) was located at various longitudinal positions behind shields of 80- and 160-cm thick concrete and 40-cm thick iron. The measurement locations cover an angular range with respect to the beam axis between 13 and 133°. Neutron energy spectra in the energy range between 32 MeV and 380 MeV were obtained by unfolding the measured pulse height spectra with the detector response functions which have been verified in the neutron energy range up to 380 MeV in separate experiments. Since the source term and experimental geometry in this experiment are well characterized and simple and results are given in the form of energy spectra, these experimental results are very useful as benchmark data to check the accuracies of simulation codes and nuclear data. Monte Carlo simulations of the experimental set up were performed with the FLUKA, MARS and PHITS codes. Simulated spectra for the 80-cm thick concrete often agree within the experimental uncertainties. On the other hand, for the 160-cm thick concrete and iron shield differences are generally larger than the experimental uncertainties, yet within a factor of 2. Based on source term simulations, observed discrepancies among simulations of spectra outside the shield can be partially explained by differences in the high-energy hadron production in the copper target.

  11. Neutron kerma factors, and water equivalence of some tissue substitutes

    Energy Technology Data Exchange (ETDEWEB)

    Singh, V. P.; Badiger, N. M. [Karnatak University, Department of Physics, Dharwad, 580003 (India); Vega C, H. R., E-mail: kudphyvps@rediffmail.com [Universidad Autonoma de Zacatecas, Unidad Academica de Estudios Nucleares, Cipres No. 10, Fracc. La Penuela, 98068 Zacatecas (Mexico)

    2014-08-15

    The kerma factors and kerma relative to air and water of 24 compounds used as tissue substitutes were calculated for neutron energy from 2.53 x 10{sup -8} up to 29 MeV. The kerma ratio of the tissue substitutes relative to air and water were calculated by the ratio of kerma factors of the tissue substitute to air and water respectively. The water equivalence of the selected tissue substitutes was observed above neutron energies 100 eV. Kerma ratio relative to the air for Poly-vinylidene fluoride and Teflon are found to be nearest to unity in very low energy (up to 1 eV) and above 63 eV respectively. It was found that the natural rubber as a water equivalent tissue substitute compound. The results of the kerma factors in our investigation shows a very good agreement with those published in ICRU-44. We found that at higher neutron energies, the kerma factors and kerma ratios of the selected tissue substitute compounds are approximately same, but differences are large for energies below 100 eV. (Author)

  12. Measurements and calculations of air activation in the NuMI neutrino production facility at Fermilab with the 120-GeV proton beam on target

    Energy Technology Data Exchange (ETDEWEB)

    Rakhno, I. L. [Fermilab; Hylen, J. [Fermilab; Kasper, P. [Fermilab; Mokhov, N. V. [Fermilab; Quinn, M. [Fermilab; Striganov, S. I. [Fermilab; Vaziri, K. [Fermilab

    2017-09-18

    Measurements and calculations of the air activation at a high-energy proton accelerator are described. The quantity of radionuclides released outdoors depends on operation scenarios including details of the air exchange inside the facility. To improve the prediction of the air activation levels, the MARS15 Monte Carlo code radionuclide production model was modified to be used for these studies. Measurements were done to benchmark the new model and verify its use in optimization studies for the new DUNE experiment at the Long Baseline Neutrino Facility (LBNF) at Fermilab. The measured production rates for the most important radionuclides – 11C, 13N, 15O and 41Ar – are in a good agreement with those calculated with the improved MARS15 code.

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

    Energy Technology Data Exchange (ETDEWEB)

    Pinchedez, K

    1999-06-01

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

  14. Calculations of reactivity based in the solution of the Neutron transport equation in X Y geometry and Lineal perturbation theory; Calculos de reactividad basados en la solucion de la Ecuacion de transporte de neutrones en geometria XY y Teoria de perturbacion lineal

    Energy Technology Data Exchange (ETDEWEB)

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

    2005-07-01

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

  15. Improvement of Sodium Neutronic Nuclear Data for the Computation of Generation IV Reactors; Contribution a l'amelioration des donnees nucleaires neutroniques du sodium pour le calcul des reacteurs de generation IV

    Energy Technology Data Exchange (ETDEWEB)

    Archier, P.

    2011-09-14

    The safety criteria to be met for Generation IV sodium fast reactors (SFR) require reduced and mastered uncertainties on neutronic quantities of interest. Part of these uncertainties come from nuclear data and, in the particular case of SFR, from sodium nuclear data, which show significant differences between available international libraries (JEFF-3.1.1, ENDF/B-VII.0, JENDL-4.0). The objective of this work is to improve the knowledge on sodium nuclear data for a better calculation of SFR neutronic parameters and reliable associated uncertainties. After an overview of existing {sup 23}Na data, the impact of the differences is quantified, particularly on sodium void reactivity effects, with both deterministic and stochastic neutronic codes. Results show that it is necessary to completely re-evaluate sodium nuclear data. Several developments have been made in the evaluation code Conrad, to integrate new nuclear reactions models and their associated parameters and to perform adjustments with integral measurements. Following these developments, the analysis of differential data and the experimental uncertainties propagation have been performed with Conrad. The resolved resonances range has been extended up to 2 MeV and the continuum range begins directly beyond this energy. A new {sup 23}Na evaluation and the associated multigroup covariances matrices were generated for future uncertainties calculations. The last part of this work focuses on the sodium void integral data feedback, using methods of integral data assimilation to reduce the uncertainties on sodium cross sections. This work ends with uncertainty calculations for industrial-like SFR, which show an improved prediction of their neutronic parameters with the new evaluation. (author) [French] Les criteres de surete exiges pour les reacteurs rapides au sodium de Generation IV (RNR-Na) se traduisent par la necessite d'incertitudes reduites et maitrisees sur les grandeurs neutroniques d'interet. Une part

  16. Neutrons from multifragmentation reactions

    CERN Document Server

    Trautmann, W; Brzychczyk, J; Buyukcizmeci, N; Mishustin, I N; Pawlowski, P

    2011-01-01

    The neutron emission in the fragmentation of stable and radioactive Sn and La projectiles of 600 MeV per nucleon has been studied with the Large Neutron Detector LAND coupled to the ALADIN forward spectrometer at SIS. A cluster-recognition algorithm is used to identify individual particles within the hit distributions registered with LAND. The obtained momentum distributions are extrapolated over the full phase space occupied by the neutrons from the projectile-spectator source. The mean multiplicities of spectator neutrons reach values of up to 12 and depend strongly on the isotopic composition of the projectile. An effective source temperature of T approx. 3 - 4 MeV is deduced from the transverse momentum distributions. For the interpretation of the data, calculations with the Statistical Multifragmentation Model for a properly chosen ensemble of excited sources were performed. The possible modification of the liquid-drop parameters of the fragment description in the hot environment is studied, and a signif...

  17. Automation on computer of the partial area method in the analysis of resonances induced by 'S' neutrons 2. with an interference term and extension of the method to the treatment of multi resonances (1963); Automatisation sur ordinateur de la methode des aires partielles dans l'analyse des resonances induites par les neutrons ''S''. 2, avec terme d'interference et extension de la methode au traitement des multiresonances (1963)

    Energy Technology Data Exchange (ETDEWEB)

    Bianchi, G.; Corge, C.R. [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1963-07-01

    This report deals with the numerical analysis on an I.B.M. 7090 computer of transmission resonances induced by 's' wave neutrons in time of flight experiments. The analysis method used is the partial area one. In this second part the interference term is taken into account. Modifications have been made in the programs and subroutines described in the first part, to determine the resonant transmissions from experimental raw data, and the relating partial areas. Also programs and subroutines are thoroughly described, which estimate the resonance parameters. The field of the partial area method has been extended to cover the case where several resonances have to be treated simultaneously, provided they do not interfere. (authors) [French] Le pretent rapport a pour objet l'analyse numerique sur ordinateur I.B.M. 7090 des resonances dues aux neutrons ''s'' dans les experiences de transmission par temps de vol, la methode d'analyse utilisee etant la methode dea aires partielles. Dans cette deuxieme partie il a ete tenu compte du terme d'interference. On y trouvera une description des amenagements apportes aux programmes et sous-programmes decrits dans la premiere partie pour determiner les transmissions interfero-resonnantes a partir des donnees experimentales brutes et les aires partielles afferentes. Sont egalement decrits les programmes et sous-programmes necessaires au calcul des parametres caracteristiques des resonances. Le domaine d'application de la methode a ete etendu au traitement simultane de plusieurs resonances groupees n'interferant pas entre elles. (auteurs)

  18. Neutron beam design for low intensity neutron and gamma-ray radioscopy using small neutron sources

    CERN Document Server

    Matsumoto, T

    2003-01-01

    Two small neutron sources of sup 2 sup 5 sup 2 Cf and sup 2 sup 4 sup 1 Am-Be radioisotopes were used for design of neutron beams applicable to low intensity neutron and gamma ray radioscopy (LINGR). In the design, Monte Carlo code (MCNP) was employed to generate neutron and gamma ray beams suited to LINGR. With a view to variable neutron spectrum and neutron intensity, various arrangements were first examined, and neutron-filter, gamma-ray shield and beam collimator were verified. Monte Carlo calculations indicated that with a suitable filter-shield-collimator arrangement, thermal neutron beam of 3,900 ncm sup - sup 2 s sup - sup 1 with neutron/gamma ratio of 7x10 sup 7 , and 25 ncm sup - sup 2 s sup - sup 1 with very large neutron/gamma ratio, respectively, could be produced by using sup 2 sup 5 sup 2 Cf(122 mu g) and a sup 2 sup 4 sup 1 Am-Be(37GBq)radioisotopes at the irradiation port of 35 cm from the neutron sources.

  19. Use of the Neutron Activation Analysis, k0-standardized method, in the evaluation of the air quality in critical points of Belo Horizonte

    Energy Technology Data Exchange (ETDEWEB)

    Moura, Igor Felipe S.; Cruz, Ananda B.; Cesar, Raisa H.S.; Oliveira, Aline F.G. de; Barreto, Alberto A.; Menezes, Maria Ângela de B.C., E-mail: igorfelipedx@yahoo.com.br, E-mail: abc@cdtn.br, E-mail: raisa.santana@hotmail.com, E-mail: afgo@cdtn.br, E-mail: aab@cdtn.br, E-mail: menezes@cdtn.br [Universidade Federal de Minas Gerais (DEN/UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear; Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN), Belo Horizonte, MG (Brazil). Divisão de Meio Ambiente

    2017-07-01

    Belo Horizonte, the capital of Minas Gerais, is one of the largest cities in Brazil. It is geographically centered on a region with intense mining and industrial activity and a fleet of more than 1.5 million vehicles. The numerous sources of atmospheric emissions of particulate matter in this capital and its vicinity offer a considerable risk to the degradation of air quality. In order to contribute with elemental composition of particulate matter, the Nuclear Technology Development Center / Brazilian Commission for Nuclear Energy, and the Department of Nuclear Engineering of the Federal University of Minas Gerais has developed a research based on environmental samples. This work aims to present the sampling methodology, the concentration values of sampled particulate matter, the methodology of analysis of elemental concentration using the nuclear technique of neutron activation analysis, k{sub 0}-standardized method, and interpreted through statistical techniques and values of the elemental concentrations more representative of the samplings carried out. The results pointed out the vehicle emissions as the major source that contributes to airborne particulate matter followed by soil resuspension. (author)

  20. Pacific Northwest National Laboratory Site Dose-per-Unit-Release Factors for Use in Calculating Radionuclide Air Emissions Potential-to-Emit Doses

    Energy Technology Data Exchange (ETDEWEB)

    Barnett, J. Matthew; Rhoads, Kathleen

    2008-09-29

    This report documents assumptions and inputs used to prepare the dose-per-unit-release factors for the Pacific Northwest National Laboratory (PNNL) Site (including the buildings that make up the Physical Sciences Facility [PSF] as well as the Environmental Molecular Sciences Laboratory [EMSL]) calculated using the EPA-approved Clean Air Act Assessment Package 1988–Personal Computer (CAP88-PC) Version 3 software package. The dose-per-unit-release factors are used to prepare dose estimates for a maximum public receptor (MPR) in support of Radioactive Air Pollutants Notice of Construction (NOC) applications for the PNNL Site.

  1. Pacific Northwest National Laboratory Site Dose-per-Unit-Release Factors for Use in Calculating Radionuclide Air Emissions Potential-to-Emit Doses

    Energy Technology Data Exchange (ETDEWEB)

    Barnett, J. Matthew; Rhoads, Kathleen

    2009-06-11

    This report documents assumptions and inputs used to prepare the dose-per-unit-release factors for the Pacific Northwest National Laboratory (PNNL) Site (including the buildings that make up the Physical Sciences Facility [PSF] as well as the Environmental Molecular Sciences Laboratory [EMSL]) calculated using the EPA-approved Clean Air Act Assessment Package 1988–Personal Computer (CAP88-PC) Version 3 software package. The dose-per-unit-release factors are used to prepare dose estimates for a maximum public receptor (MPR) in support of Radioactive Air Pollutants Notice of Construction (NOC) applications for the PNNL Site.

  2. Development of a computational model for the calculation of neutron dose equivalent in laminated primary barriers of radiotherapy rooms; Desenvolvimento de um modelo computacional para calculo do equivalente de dose de neutrons em barreiras primarias laminadas de salas de radioterapia

    Energy Technology Data Exchange (ETDEWEB)

    Rezende, Gabriel Fonseca da Silva

    2015-06-01

    Many radiotherapy centers acquire 15 and 18 MV linear accelerators to perform more effective treatments for deep tumors. However, the acquisition of these equipment must be accompanied by an additional care in shielding planning of the rooms that will house them. In cases where space is restricted, it is common to find primary barriers made of concrete and metal. The drawback of this type of barrier is the photoneutron emission when high energy photons (e.g. 15 and 18 MV spectra) interact with the metallic material of the barrier. The emission of these particles constitutes a problem of radiation protection inside and outside of radiotherapy rooms, which should be properly assessed. A recent work has shown that the current model underestimate the dose of neutrons outside the treatment rooms. In this work, a computational model for the aforementioned problem was created from Monte Carlo Simulations and Artificial Intelligence. The developed model was composed by three neural networks, each being formed of a pair of material and spectrum: Pb18, Pb15 and Fe18. In a direct comparison with the McGinley method, the Pb18 network exhibited the best responses for approximately 78% of the cases tested; the Pb15 network showed better results for 100% of the tested cases, while the Fe18 network produced better answers to 94% of the tested cases. Thus, the computational model composed by the three networks has shown more consistent results than McGinley method. (author)

  3. Neutron sources and its dosimetric characteristics; Fuentes de neutrones y sus caracteristicas dosimetricas

    Energy Technology Data Exchange (ETDEWEB)

    Vega C, H.R.; Manzanares A, E.; Hernandez D, V.M.; Mercado S, G.A. [Universidad Autonoma de Zacatecas, A.P. 336, 98000 Zacatecas (Mexico); Gallego D, E.; Lorente F, A. [Universidad Politecnica de Madrid, C/Jose Gutierrez Abascal 2, E-28006 Madrid (Spain)

    2005-07-01

    By means of Monte Carlo methods the spectra of the produced neutrons {sup 252} Cf, {sup 252} Cf/D{sub 2}O, {sup 241} Am Be, {sup 239} Pu Be, {sup 140} La Be, {sup 239} Pu{sup 18}O{sub 2} and {sup 226} Ra Be have been calculated. With the information of the spectrum it was calculated the average energy of the neutrons of each source. By means of the fluence coefficients to dose it was determined, for each one of the studied sources, the fluence factors to dose. The calculated doses were H, H{sup *}(10), H{sub p,sIab} (10, 0{sup 0}), E{sub AP} and E{sub ISO}. During the phase of the calculations the sources were modeled as punctual and their characteristics were determined to 100 cm in the hole. Also, for the case of the sources of {sup 239} Pu Be and {sup 241} Am Be, were carried out calculations modeling the sources with their respective characteristics and the dosimetric properties were determined in a space full with air. The results of this last phase of the calculations were compared with the experimental results obtained for both sources. (Author)

  4. Neutron background estimates in GESA

    Directory of Open Access Journals (Sweden)

    Fernandes A.C.

    2014-01-01

    Full Text Available The SIMPLE project looks for nuclear recoil events generated by rare dark matter scattering interactions. Nuclear recoils are also produced by more prevalent cosmogenic neutron interactions. While the rock overburden shields against (μ,n neutrons to below 10−8 cm−2 s−1, it itself contributes via radio-impurities. Additional shielding of these is similar, both suppressing and contributing neutrons. We report on the Monte Carlo (MCNP estimation of the on-detector neutron backgrounds for the SIMPLE experiment located in the GESA facility of the Laboratoire Souterrain à Bas Bruit, and its use in defining additional shielding for measurements which have led to a reduction in the extrinsic neutron background to ∼ 5 × 10−3 evts/kgd. The calculated event rate induced by the neutron background is ∼ 0,3 evts/kgd, with a dominant contribution from the detector container.

  5. Neutron streaming studies along JET shielding penetrations

    OpenAIRE

    Stamatelatos Ion E.; Vasilopoulou Theodora; Batistoni Paola; Obryk Barbara; Popovichev Sergey; Naish Jonathan

    2017-01-01

    Neutronic benchmark experiments are carried out at JET aiming to assess the neutronic codes and data used in ITER analysis. Among other activities, experiments are performed in order to validate neutron streaming simulations along long penetrations in the JET shielding configuration. In this work, neutron streaming calculations along the JET personnel entrance maze are presented. Simulations were performed using the MCNP code for Deuterium-Deuterium and Deuterium- Tritium plasma sources. The ...

  6. Procedure of calculation of the spatial distribution of temperatures and heat fluxes in the steam generator of a nuclear power installation with an RBEC fast-neutron reactor

    Science.gov (United States)

    Frolov, A. A.; Sedov, A. A.

    2016-08-01

    A method for combined 3D/1D-modeling of thermohydraulics of a once-through steam generator (SG) based on the joint analysis of three-dimensional thermo- and hydrodynamics of a single-phase heating coolant in the intertube space and one-dimensional thermohydraulics of steam-generating channels (tubes) with the use of well-known friction and heat-transfer correlations under various boiling conditions is discussed. This method allows one to determine the spatial distribution of temperatures and heat fluxes of heat-exchange surfaces of SGs with a single-phase heating coolant in the intertube space and with steam generation within tubes. The method was applied in the analytical investigation of typical operation of a once-through SG of a nuclear power installation with an RBEC fast-neutron heavy-metal reactor that is being designed by Kurchatov Institute in collaboration with OKB GIDROPRESS and Leipunsky Institute of Physics and Power Engineering. Flow pattern and temperature fields were obtained for the heavy-metal heating coolant in the intertube space. Nonuniformities of heating of the steam-water coolant in different heat-exchange tubes and nonuniformities in the distribution of heat fluxes at SG heat-exchange surfaces were revealed.

  7. Measurements of H*(10) in reference neutron fields using Bonner sphere spectrometry and LET spectrometry

    CERN Document Server

    Golnik, N; Králik, M

    2002-01-01

    A Bonner sphere spectrometer and the REM-2 recombination chamber were used for inter-comparison measurements of the neutron component of ambient dose equivalent, H sub n *(10) in reference neutron fields. The sup 2 sup 4 sup 1 Am-Be and sup 2 sup 5 sup 2 Cf neutron sources were exposed either free-in-air or placed in iron or paraffin filters. The REM-2 recombination chamber was used as a LET spectrometer. The agreement of H sub n *(10) values measured with both the methods was within experimental uncertainties of few percent. The determined neutron spectra were used for calculations of the REM-2 chamber response to H*(10).

  8. Application of neural networks for the calculation of technical losses of electric energy in air power lines 6-35 kV

    OpenAIRE

    Володимир Леонідович Бакулевський

    2015-01-01

    A model for the calculation of technical losses of electricity in the air lines with voltage of 6-35 kV based on neural networks with due regard to meteorological factors has been worked out; the main components of the model have been considered and researched; the best ones being selected, that is: a set of input variables, volume of excerpts (training, control and testing), architecture and network activation function, network learning algorithm was proposed. Simulation was conducted in OS ...

  9. Neutron collimator for neutron radiography applications at tangential port of the TRIGA RC-1 reactor

    Science.gov (United States)

    Rosa, R.; Andreoli, F.; Mattoni, M.; Palomba, M.

    2009-06-01

    At the ENEA TRIGA research reactor (Casaccia Research Center, Rome) a new neutron collimator has been designed and installed at the neutron tangential channel. This collimator, that is part of a neutron/X-ray facility for NDT analysis, was experimentally characterized and optimized in terms of thermal neutron fluence rate, spatial/energetic distribution, photon air KERMA and effective beam diameter. This paper shows the methodologies and the results of the experimental analysis that were carried out.

  10. Characterization of source neutrons produced by Frascati neutron generator

    Energy Technology Data Exchange (ETDEWEB)

    Pillon, M.; Angelone, M.; Martone, M.; Rado, V. [Associazione EURATOM-ENEA sulla Fusione, Frascati (Italy)

    1994-11-01

    The characterization of the 14 MeV neutron source of the Frascati neutron generator (FNG) has been carried out using an absolutely calibrated U{sup Nat} micro-fission chamber with plane electrodes and {sup 93}Nb(n,2n){sup 92m}Nb activation reaction. The reaction rates have been measured at different angles with respect to the beam direction. The experimental results are compared with the corresponding quantities calculated by using a Monte Carlo code with an ad hoc source model. The calculated neutron emission reproduces satisfactory well the measured one. Improving of the used source model is discussed.

  11. In vitro biological effectiveness of JRR-4 epithermal neutron beam. Experiment under free air beam and in water phantom. Cooperative research

    CERN Document Server

    Yamamoto, T; Horiguchi, Y; Kishi, T; Kumada, H; Matsumura, A; Nose, T; Torii, Y; Yamamoto, K

    2002-01-01

    The surviving curve and the biological effectiveness factor of dose components generated in boron neutron capture therapy (BNCT) were separately determined in neutron beams at Japan Research Reactor No.4. Surviving fraction of V79 Chinese hamster cell with or without sup 1 sup 0 B was obtained using an epithermal neutron beam (ENB), a mixed thermal-epithermal neutron beam (TNB-1), and a thermal neutron beam (TNB-2), which were used or planned to use for BNCT clinical trial. The cell killing effect of these neutron beams with or without the presence of sup 1 sup 0 B depended highly on the neutron beam used, according to the epithermal and fast neutron content in the beam. The biological effectiveness factor values of the boron capture reaction for ENB, TNB-1 and TNB-2 were 3.99+-0.24, 3.04+-0.19 and 1.43+-0.08, respectively. The biological effectiveness factor values of the high-LET dose components based on the hydrogen recoils and the nitrogen capture reaction were 2.50+-0.32, 2.34+-0.30 and 2.17+-0.28 for EN...

  12. Neutron detector

    Science.gov (United States)

    Stephan, Andrew C [Knoxville, TN; Jardret,; Vincent, D [Powell, TN

    2011-04-05

    A neutron detector has a volume of neutron moderating material and a plurality of individual neutron sensing elements dispersed at selected locations throughout the moderator, and particularly arranged so that some of the detecting elements are closer to the surface of the moderator assembly and others are more deeply embedded. The arrangement captures some thermalized neutrons that might otherwise be scattered away from a single, centrally located detector element. Different geometrical arrangements may be used while preserving its fundamental characteristics. Different types of neutron sensing elements may be used, which may operate on any of a number of physical principles to perform the function of sensing a neutron, either by a capture or a scattering reaction, and converting that reaction to a detectable signal. High detection efficiency, an ability to acquire spectral information, and directional sensitivity may be obtained.

  13. Fusion neutronics

    CERN Document Server

    Wu, Yican

    2017-01-01

    This book provides a systematic and comprehensive introduction to fusion neutronics, covering all key topics from the fundamental theories and methodologies, as well as a wide range of fusion system designs and experiments. It is the first-ever book focusing on the subject of fusion neutronics research. Compared with other nuclear devices such as fission reactors and accelerators, fusion systems are normally characterized by their complex geometry and nuclear physics, which entail new challenges for neutronics such as complicated modeling, deep penetration, low simulation efficiency, multi-physics coupling, etc. The book focuses on the neutronics characteristics of fusion systems and introduces a series of theories and methodologies that were developed to address the challenges of fusion neutronics, and which have since been widely applied all over the world. Further, it introduces readers to neutronics design’s unique principles and procedures, experimental methodologies and technologies for fusion systems...

  14. Neutron matter, neutron pairing, and neutron drops based on chiral effective field theory interactions

    Energy Technology Data Exchange (ETDEWEB)

    Krueger, Thomas

    2016-10-19

    The physics of neutron-rich systems is of great interest in nuclear and astrophysics. Precise knowledge of the properties of neutron-rich nuclei is crucial for understanding the synthesis of heavy elements. Infinite neutron matter determines properties of neutron stars, a final stage of heavy stars after a core-collapse supernova. It also provides a unique theoretical laboratory for nuclear forces. Strong interactions are determined by quantum chromodynamics (QCD). However, QCD is non-perturbative at low energies and one presently cannot directly calculate nuclear forces from it. Chiral effective field theory circumvents these problems and connects the symmetries of QCD to nuclear interactions. It naturally and systematically includes many-nucleon forces and gives access to uncertainty estimates. We use chiral interactions throughout all calculation in this thesis. Neutron stars are very extreme objects. The densities in their interior greatly exceed those in nuclei. The exact composition and properties of neutron stars is still unclear but they consist mainly of neutrons. One can explore neutron stars theoretically with calculations of neutron matter. In the inner core of neutron stars exist very high densities and thus maybe exotic phases of matter. To investigate whether there exists a phase transition to such phases even at moderate densities we study the chiral condensate in neutron matter, the order parameter of chiral symmetry breaking, and find no evidence for a phase transition at nuclear densities. We also calculate the more extreme system of spin-polarised neutron matter. With this we address the question whether there exists such a polarised phase in neutron stars and also provide a benchmark system for lattice QCD. We find spin-polarised neutron matter to be an almost non-interacting Fermi gas. To understand the cooling of neutron stars neutron pairing is of great importance. Due to the high densities especially triplet pairing is of interest. We

  15. Neutron Stars

    Science.gov (United States)

    Cottam, J.

    2007-01-01

    Neutron stars were discovered almost 40 years ago, and yet many of their most fundamental properties remain mysteries. There have been many attempts to measure the mass and radius of a neutron star and thereby constrain the equation of state of the dense nuclear matter at their cores. These have been complicated by unknown parameters such as the source distance and burning fractions. A clean, straightforward way to access the neutron star parameters is with high-resolution spectroscopy. I will present the results of searches for gravitationally red-shifted absorption lines from the neutron star atmosphere using XMM-Newton and Chandra.

  16. Development of the neutron-transport code TransRay and studies on the two- and three-dimensional calculation of effective group cross sections; Entwicklung des Neutronentransportcodes TransRay und Untersuchungen zur zwei- und dreidimensionalen Berechnung effektiver Gruppenwirkungsquerschnitte

    Energy Technology Data Exchange (ETDEWEB)

    Beckert, C.

    2007-12-19

    Conventionally the data preparation of the neutron cross sections for reactor-core calculations pursues with 2D cell codes. Aim of this thesis was, to develop a 3D cell code, to study with this code 3D effects, and to evaluate the necessarity of a 3D data preparation of the neutron cross sections. For the calculation of the neutron transport the method of the first-collision probabilities, which are calculated with the ray-tracing method, was chosen. The mathematical algorithms were implemented in the 2D/3D cell code TransRay. For the geometry part of the program the geometry module of a Monte Carlo code was used.The ray tracing in 3D was parallelized because of the high computational time. The program TransRay was verified on 2D test problems. For a reference pressured-water reactor following 3D problems were studied: A partly immersed control rod and void (vacuum or steam) around a fuel rod as model of a steam void. All problems were for comparison calculated also with the programs HELIOS(2D) and MCNP(3D). The dependence of the multiplication factor and the averaged two-group cross section on the immersion depth of the control rod respectively of the height of the steam void were studied. The 3D-calculated two-group cross sections were compared with three conventional approximations: Linear interpolation, interpolation with flux weighting, and homogenization, At the 3D problem of the control rod it was shown that the interpolation with flux weighting is a good approximation. Therefore here a 3D data preparation is not necessary. At the test case of the single control rod, which is surrounded by the void, the three approximation for the two-group cross sections were proved as unsufficient. Therefore a 3D data preparation is necessary. The single fuel-rod cell with void can be considered as the limiting case of a reactor, in which a phase interface has been formed. [German] Standardmaessig erfolgt die Datenaufbereitung der Neutronenwirkungsquerschnitte fuer

  17. Secondary neutron spectrum from 250-MeV passively scattered proton therapy: Measurement with an extended-range Bonner sphere system

    Energy Technology Data Exchange (ETDEWEB)

    Howell, Rebecca M., E-mail: rhowell@mdanderson.org [Department of Radiation Physics, The University of Texas MD Anderson Cancer Center, Houston, Texas 77030 (United States); Burgett, E. A. [Department of Nuclear Engineering, Idaho State University, Pocatello, Idaho 83201 (United States)

    2014-09-15

    Purpose: Secondary neutrons are an unavoidable consequence of proton therapy. While the neutron dose is low compared to the primary proton dose, its presence and contribution to the patient dose is nonetheless important. The most detailed information on neutrons includes an evaluation of the neutron spectrum. However, the vast majority of the literature that has reported secondary neutron spectra in proton therapy is based on computational methods rather than measurements. This is largely due to the inherent limitations in the majority of neutron detectors, which are either not suitable for spectral measurements or have limited response at energies greater than 20 MeV. Therefore, the primary objective of the present study was to measure a secondary neutron spectrum from a proton therapy beam using a spectrometer that is sensitive to neutron energies over the entire neutron energy spectrum. Methods: The authors measured the secondary neutron spectrum from a 250-MeV passively scattered proton beam in air at a distance of 100 cm laterally from isocenter using an extended-range Bonner sphere (ERBS) measurement system. Ambient dose equivalent H*(10) was calculated using measured fluence and fluence-to-ambient dose equivalent conversion coefficients. Results: The neutron fluence spectrum had a high-energy direct neutron peak, an evaporation peak, a thermal peak, and an intermediate energy continuum between the thermal and evaporation peaks. The H*(10) was dominated by the neutrons in the evaporation peak because of both their high abundance and the large quality conversion coefficients in that energy interval. The H*(10) 100 cm laterally from isocenter was 1.6 mSv per proton Gy (to isocenter). Approximately 35% of the dose equivalent was from neutrons with energies ≥20 MeV. Conclusions: The authors measured a neutron spectrum for external neutrons generated by a 250-MeV proton beam using an ERBS measurement system that was sensitive to neutrons over the entire

  18. Secondary neutron spectrum from 250-MeV passively scattered proton therapy: measurement with an extended-range Bonner sphere system.

    Science.gov (United States)

    Howell, Rebecca M; Burgett, E A

    2014-09-01

    Secondary neutrons are an unavoidable consequence of proton therapy. While the neutron dose is low compared to the primary proton dose, its presence and contribution to the patient dose is nonetheless important. The most detailed information on neutrons includes an evaluation of the neutron spectrum. However, the vast majority of the literature that has reported secondary neutron spectra in proton therapy is based on computational methods rather than measurements. This is largely due to the inherent limitations in the majority of neutron detectors, which are either not suitable for spectral measurements or have limited response at energies greater than 20 MeV. Therefore, the primary objective of the present study was to measure a secondary neutron spectrum from a proton therapy beam using a spectrometer that is sensitive to neutron energies over the entire neutron energy spectrum. The authors measured the secondary neutron spectrum from a 250-MeV passively scattered proton beam in air at a distance of 100 cm laterally from isocenter using an extended-range Bonner sphere (ERBS) measurement system. Ambient dose equivalent H*(10) was calculated using measured fluence and fluence-to-ambient dose equivalent conversion coefficients. The neutron fluence spectrum had a high-energy direct neutron peak, an evaporation peak, a thermal peak, and an intermediate energy continuum between the thermal and evaporation peaks. The H*(10) was dominated by the neutrons in the evaporation peak because of both their high abundance and the large quality conversion coefficients in that energy interval. The H*(10) 100 cm laterally from isocenter was 1.6 mSv per proton Gy (to isocenter). Approximately 35% of the dose equivalent was from neutrons with energies ≥20 MeV. The authors measured a neutron spectrum for external neutrons generated by a 250-MeV proton beam using an ERBS measurement system that was sensitive to neutrons over the entire energy range being measured, i.e., thermal to

  19. Monte Carlo calculations on transmutation of trans-uranic nuclear waste isotopes using spallation neutrons difference of lead and graphite moderators

    CERN Document Server

    Hashemi-Nezhad, S R; Brandt, R; Krivopustov, M I; Kulakov, B A; Odoj, R; Sosnin, A N; Wan, J S; Westmeier, W

    2002-01-01

    Transmutation rates of sup 2 sup 3 sup 9 Pu and some minor actinides ( sup 2 sup 3 sup 7 Np, sup 2 sup 4 sup 1 Am, sup 2 sup 4 sup 5 Cm and sup 2 sup 4 sup 6 Cm), in two accelerator-driven systems (ADS) with lead or graphite moderating environments, were calculated using the LAHET code system. The ADS that were used had a large volume (approx 32 m sup 3) and contained no fissile material, except for a small amount of fissionable waste nuclei that existed in some cases. Calculations were performed at an incident proton energy of 1.5 GeV and the spallation target was lead. Also breeding rates of sup 2 sup 3 sup 9 Pu and sup 2 sup 3 sup 3 U as well as the transmutation rates of two long-lived fission products sup 9 sup 9 Tc and sup 1 sup 2 sup 9 I were calculated at different locations in the moderator. It is shown that an ADS with graphite moderator is a much more effective transmuter than that with lead moderator.

  20. Model calculations of the effects of present and future emissions of air pollutants from shipping in the Baltic Sea and the North Sea

    Directory of Open Access Journals (Sweden)

    J. E. Jonson

    2015-01-01

    Full Text Available Land-based emissions of air pollutants in Europe have steadily decreased over the past two decades, and this decrease is expected to continue. Within the same time span emissions from shipping have increased in EU ports and in the Baltic Sea and the North Sea, defined as SECAs (sulfur emission control areas, although recently sulfur emissions, and subsequently particle emissions, have decreased. The maximum allowed sulfur content in marine fuels in EU ports is now 0.1%, as required by the European Union sulfur directive. In the SECAs the maximum fuel content of sulfur is currently 1% (the global average is about 2.4%. This will be reduced to 0.1% from 2015, following the new International Maritime Organization (IMO rules. In order to assess the effects of ship emissions in and around the Baltic Sea and the North Sea, regional model calculations with the EMEP air pollution model have been made on a 1/4° longitude × 1/8° latitude resolution, using ship emissions in the Baltic Sea and the North Sea that are based on accurate ship positioning data. The effects on depositions and air pollution and the resulting number of years of life lost (YOLLs have been calculated by comparing model calculations with and without ship emissions in the two sea areas. In 2010 stricter regulations for sulfur emissions were implemented in the two sea areas, reducing the maximum sulfur content allowed in marine fuels from 1.5 to 1%. In addition ships were required to use fuels with 0.1 % sulfur in EU harbours. The calculations have been made with emissions representative of 2009 and 2011, i.e. before and after the implementation of the stricter controls on sulfur emissions from 2010. The calculations with present emissions show that per person, an additional 0.1–0.2 years of life lost is estimated in areas close to the major ship tracks with current emission levels. Comparisons of model calculations with emissions before and after the implementation of stricter

  1. Fast neutron activation analysis by means of low voltage neutron generator

    Science.gov (United States)

    Medhat, M. E.

    A description of D-T neutron generator (NG) is presented. This machine can be used for fast neutron activation analysis applied to determine some selected elements, especially light elements, in different materials. Procedure of neutron flux determination and efficiency calculation is described. Examples of testing some Egyptian natural cosmetics are given.

  2. Physics of solar neutron production: Questionable detection of neutrons from the 31 December 2007 flare

    National Research Council Canada - National Science Library

    Share, Gerald H; Murphy, Ronald J; Tylka, Allan J; Kozlovsky, Benz; Ryan, James M; Gwon, Chul

    2011-01-01

    ...–10 MeV solar neutrons. We discuss the cross sections for neutron production in solar flares and calculate the escaping neutron spectra for monoenergetic and power law particle spectra at the Sun and at the distance (0.48 AU...

  3. Effects of bone- and air-tissue inhomogeneities on the dose distributions of the Leksell Gamma Knife (registered) calculated with PENELOPE

    Energy Technology Data Exchange (ETDEWEB)

    Al-Dweri, Feras M O [Departamento de Fisica Moderna, Universidad de Granada, E-18071 Granada (Spain); Department of Physics, Applied Science Private University, Amman (Jordan); Rojas, E Leticia [Departamento de Fisica Moderna, Universidad de Granada, E-18071 Granada (Spain); Instituto Nacional de Investigaciones Nucleares, Carretera Mexico-Toluca, km 36.5, Ocoyoacac, C.P. 52045 (Mexico); Lallena, Antonio M [Departamento de Fisica Moderna, Universidad de Granada, E-18071 Granada (Spain)

    2005-12-07

    Monte Carlo simulation with PENELOPE (version 2003) is applied to calculate Leksell Gamma Knife (registered) dose distributions for heterogeneous phantoms. The usual spherical water phantom is modified with a spherical bone shell simulating the skull and an air-filled cube simulating the frontal or maxillary sinuses. Different simulations of the 201 source configuration of the Gamma Knife have been carried out with a simplified model of the geometry of the source channel of the Gamma Knife recently tested for both single source and multisource configurations. The dose distributions determined for heterogeneous phantoms including the bone- and/or air-tissue interfaces show non-negligible differences with respect to those calculated for a homogeneous one, mainly when the Gamma Knife isocentre approaches the separation surfaces. Our findings confirm an important underdosage ({approx}10%) nearby the air-tissue interface, in accordance with previous results obtained with the PENELOPE code with a procedure different from ours. On the other hand, the presence of the spherical shell simulating the skull produces a few per cent underdosage at the isocentre wherever it is situated.

  4. Benchmark integral neutron experiments for Fe, Be and C with DT neutron by liquid scintillation detector.

    Science.gov (United States)

    Tie, He; Pu, Zheng; Jun, Xiao; Chuanxin, Zhu; Jian, Yang; Haiping, Guo

    2017-06-01

    The measurements of iron, beryllium and carbon sphere neutron leakage spectra using BC501A detector by DT neutron source are presented. The experiments were carried out in Institute of Nuclear Physics and Chemistry (INPC), China. Neutron leakage spectra in a wide energy range at angles of 0° and 30° from the direction of incident deuteron beam were obtained. The results show the leakage neutron flux decreases notably with the spherical shell increasing when neutron energy >10MeV. When neutron energy calculation was made using the MCNP5 Code with the ENDF/B-VII.1 nuclear data files. Copyright © 2017. Published by Elsevier Ltd.

  5. Neutron tubes

    Science.gov (United States)

    Leung, Ka-Ngo; Lou, Tak Pui; Reijonen, Jani

    2008-03-11

    A neutron tube or generator is based on a RF driven plasma ion source having a quartz or other chamber surrounded by an external RF antenna. A deuterium or mixed deuterium/tritium (or even just a tritium) plasma is generated in the chamber and D or D/T (or T) ions are extracted from the plasma. A neutron generating target is positioned so that the ion beam is incident thereon and loads the target. Incident ions cause D-D or D-T (or T-T) reactions which generate neutrons. Various embodiments differ primarily in size of the chamber and position and shape of the neutron generating target. Some neutron generators are small enough for implantation in the body. The target may be at the end of a catheter-like drift tube. The target may have a tapered or conical surface to increase target surface area.

  6. Neutron stars as cosmic hadron physics laboratories

    Science.gov (United States)

    Pines, D.

    1985-05-01

    Extensive observations of Her-1 with the Exosat satellite have led to a new understanding of both the dynamics of neutron-star superfluids and the free precession of neutron stars. Detailed microscopic calculations on neutron matter and the properties of the pinned crustal superfluid are provided to serve as a basis for comparing theory with observation on neutron stars. Topics discussed include the Hadron matter equation of state, neutron star structure, Hadron superfluids, the vortex creep theory, Vela pulsar glitches, astrophysical constraints on neutron matter energy gaps, the 35 day periodicity of Her-1, and the neutron matter equation of state. It is concluded that since the post-glitch fits and the identification of the 35th periodicity in Her X-1 as stellar wobble require a rigid neutron matter equation of state, the astrophysical evidence for such an equation seems strong, as well as that for an intermediate Delta(rho) curve.

  7. Cabin air temperature of parked vehicles in summer conditions: life-threatening environment for children and pets calculated by a dynamic model

    Science.gov (United States)

    Horak, Johannes; Schmerold, Ivo; Wimmer, Kurt; Schauberger, Günther

    2017-10-01

    In vehicles that are parked, no ventilation and/or air conditioning takes place. If a vehicle is exposed to direct solar radiation, an immediate temperature rise occurs. The high cabin air temperature can threaten children and animals that are left unattended in vehicles. In the USA, lethal heat strokes cause a mean death rate of 37 children per year. In addition, temperature-sensitive goods (e.g. drugs in ambulances and veterinary vehicles) can be adversely affected by high temperatures. To calculate the rise of the cabin air temperature, a dynamic model was developed that is driven by only three parameters, available at standard meteorological stations: air temperature, global radiation and wind velocity. The transition from the initial temperature to the constant equilibrium temperature depends strongly on the configuration of the vehicle, more specifically on insulation, window area and transmission of the glass, as well as on the meteorological conditions. The comparison of the model with empirical data showed good agreement. The model output can be applied to assess the heat load of children and animals as well as temperature-sensitive goods, which are transported and/or stored in a vehicle.

  8. A refined protocol for calculating air flow rate of naturally ventilated broiler barns based on CO2 mass balance

    NARCIS (Netherlands)

    Mendes, L.; Tinoco, I.; Ogink, N.W.M.; Osorio-Hernandez, R.; Saraz, J.

    2014-01-01

    This study was conducted to evaluate relatively simple protocols for monitoring ventilation rate (VR) in naturally-ventilated barns (NVB). The test protocols were first applied to a mechanically-ventilated broiler barn (MVB), where VR was estimated more accurately and then were used to calculate VR

  9. Thermalization of monoenergetic neutrons in a concrete room

    Energy Technology Data Exchange (ETDEWEB)

    Vega C, H.R.; Manzanares A, E.; Hernandez D, V.M.; Mercado, G.A. [UAZ, A.P. 336, 98000 Zacatecas (Mexico); Iniguez, M.P.; Martin M, A. [Universidad de Valladolid, (Spain)

    2006-07-01

    The thermalization of neutrons from monoenergetic neutron sources in a concrete room has been studied. During calibration of neutron detectors it is mandatory to make corrections due to neutron scattering produced by the room walls, therefore this factor must be known in advance. The scattered neutrons are thermalized and produce a neutron field that is directly proportional to source strength and inversely proportional to room total wall-surfaces, the proportional coefficient has been calculated for neutrons whose energy goes from 1 eV to 20 MeV. This coefficient was calculated using Monte Carlo methods for 150, 200 and 300 cm-radius spherical cavity, where monoenergetic neutrons were located at the center, along the spherical cavity radius neutron spectra were calculated at several source-to-detector distances inside the cavity. The obtained coefficient is almost three times larger than the factor normally utilized. (Author)

  10. New Geometrical Approach for the Air-gap Reluctance Calculation for the Design of the Machines by the Flow Lines Method

    Directory of Open Access Journals (Sweden)

    BENATIA B Mostefa

    2012-10-01

    Full Text Available The method of the lines of flow remains always a design method which could be applied by a very broad population of manufacturers for the design of the electric machines. It is an easy method to implement, founded on basic electromagnetism principles. Our article present in general a procedure of calculation of the reluctances of variable reluctance machines (6/4 starting from the use of the analyticalrelations which generate the electromagnetic principles of the machine. A new method based on the simple geometry and the trigonometrically one was used for the calculation of the reluctances of the lines which through the air-gap of the machine. The results of the design obtained by the flow lines method with the application of this new approach were checked by the finite elements method. Analyzes and the comparison of the margins of error remain acceptable.

  11. Synovectomy by Neutron capture; Sinovectomia por captura de neutrones

    Energy Technology Data Exchange (ETDEWEB)

    Vega C, H.R.; Torres M, C. [Centro Regional de Estudios Nucleares, Universidad Autonoma de Zacatecas, C. Cipres 10, Fracc. La Penuela, 98000 Zacatecas (Mexico)

    1998-12-31

    The Synovectomy by Neutron capture has as purpose the treatment of the rheumatoid arthritis, illness which at present does not have a definitive curing. This therapy requires a neutron source for irradiating the articulation affected. The energy spectra and the intensity of these neutrons are fundamental since these neutrons induce nuclear reactions of capture with Boron-10 inside the articulation and the freely energy of these reactions is transferred at the productive tissue of synovial liquid, annihilating it. In this work it is presented the neutron spectra results obtained with moderator packings of spherical geometry which contains in its center a Pu{sup 239} Be source. The calculations were realized through Monte Carlo method. The moderators assayed were light water, heavy water base and the both combination of them. The spectra obtained, the average energy, the neutron total number by neutron emitted by source, the thermal neutron percentage and the dose equivalent allow us to suggest that the moderator packing more adequate is what has a light water thickness 0.5 cm (radius 2 cm) and 24.5 cm heavy water (radius 26.5 cm). (Author)

  12. Engineering analyses and design calculations of NASA, Langley Research Center hydrogen-air-vitiated heater with oxygen replenishment

    Science.gov (United States)

    1973-01-01

    The technical basis is presented for the design of the hydrogen-air-vitiated heater. The heater liner is subjected to a maximum thermal environment at a specified condition, where the combustion gas temperature, pressure and flow rate are 5000 F, 750 psia, and 11.0 lb/sec, respectively, and results in a heat flux of the order of 275 BTU/sec-sq ft. Cooling and stress analyses indicate that water is the logical choice for cooling of the combustor liner. A mixing analysis was undertaken to establish a good combination of combustor length and injector configuration. The analysis, using a conservative analytical approach, indicates a combustor length of the order of 5 ft combined with discrete fuel and oxidizer injection at an approximate 2-1/2 inch radial combustor position, and results in uniform combustion products at the heater exit for all specified envelope conditions.

  13. Short-term impacts of air pollutants in Switzerland: Preliminary scenario calculations for selected Swiss energy systems

    Energy Technology Data Exchange (ETDEWEB)

    Andreani-Aksoyoglu, S.; Keller, J. [Paul Scherrer Inst. (PSI), Villigen (Switzerland)

    1999-08-01

    In the frame of the comprehensive assessment of Swiss energy systems, air quality simulations were performed by using a 3-dimensional photo-chemical dispersion model. The objective is to investigate the impacts of pollutants in Switzerland for future options of Swiss energy systems. Four scenarios were investigated: Base Case: simulations with the projected emissions for the year 2030, Scenario 1) all nuclear power plants were replaced by oil-driven combined cycle plants (CCP), Scenarios 2 to 4) traffic emissions were reduced in whole Switzerland as well as in the cities and on the highways separately. Changes in the pollutant concentrations and depositions, and the possible short-term impacts are discussed on the basis of exceedences of critical levels for plants and limits given to protect the public health. (author) 2 figs., 7 refs.

  14. Neutron star structure from QCD

    CERN Document Server

    Fraga, Eduardo S; Vuorinen, Aleksi

    2016-01-01

    In this review article, we argue that our current understanding of the thermodynamic properties of cold QCD matter, originating from first principles calculations at high and low densities, can be used to efficiently constrain the macroscopic properties of neutron stars. In particular, we demonstrate that combining state-of-the-art results from Chiral Effective Theory and perturbative QCD with the current bounds on neutron star masses, the Equation of State of neutron star matter can be obtained to an accuracy better than 30% at all densities.

  15. Development of An Epi-thermal Neutron Field for Fundamental Researches for BNCT with A DT Neutron Source

    Science.gov (United States)

    Osawa, Yuta; Imoto, Shoichi; Kusaka, Sachie; Sato, Fuminobu; Tanoshita, Masahiro; Murata, Isao

    2017-09-01

    Boron Neutron Capture Therapy (BNCT) is known to be a new promising cancer therapy suppressing influence against normal cells. In Japan, Accelerator Based Neutron Sources (ABNS) are being developed for BNCT. For the spread of ABNS based BNCT, we should characterize the neutron field beforehand. For this purpose, we have been developing a low-energy neutron spectrometer based on 3He position sensitive proportional counter. In this study, a new intense epi-thermal neutron field was developed with a DT neutron source for verification of validity of the spectrometer. After the development, the neutron field characteristics were experimentally evaluated by using activation foils. As a result, we confirmed that an epi-thermal neutron field was successfully developed suppressing fast neutrons substantially. Thereafter, the neutron spectrometer was verified experimentally. In the verification, although a measured detection depth distribution agreed well with the calculated distribution by MCNP, the unfolded spectrum was significantly different from the calculated neutron spectrum due to contribution of the side neutron incidence. Therefore, we designed a new neutron collimator consisting of a polyethylene pre-collimator and boron carbide neutron absorber and confirmed numerically that it could suppress the side incident neutrons and shape the neutron flux to be like a pencil beam.

  16. Development of An Epi-thermal Neutron Field for Fundamental Researches for BNCT with A DT Neutron Source

    Directory of Open Access Journals (Sweden)

    Osawa Yuta

    2017-01-01

    Full Text Available Boron Neutron Capture Therapy (BNCT is known to be a new promising cancer therapy suppressing influence against normal cells. In Japan, Accelerator Based Neutron Sources (ABNS are being developed for BNCT. For the spread of ABNS based BNCT, we should characterize the neutron field beforehand. For this purpose, we have been developing a low-energy neutron spectrometer based on 3He position sensitive proportional counter. In this study, a new intense epi-thermal neutron field was developed with a DT neutron source for verification of validity of the spectrometer. After the development, the neutron field characteristics were experimentally evaluated by using activation foils. As a result, we confirmed that an epi-thermal neutron field was successfully developed suppressing fast neutrons substantially. Thereafter, the neutron spectrometer was verified experimentally. In the verification, although a measured detection depth distribution agreed well with the calculated distribution by MCNP, the unfolded spectrum was significantly different from the calculated neutron spectrum due to contribution of the side neutron incidence. Therefore, we designed a new neutron collimator consisting of a polyethylene pre-collimator and boron carbide neutron absorber and confirmed numerically that it could suppress the side incident neutrons and shape the neutron flux to be like a pencil beam.

  17. Simplified Two-Time Step Method for Calculating Combustion Rates and Nitrogen Oxide Emissions for Hydrogen/Air and Hydorgen/Oxygen

    Science.gov (United States)

    Molnar, Melissa; Marek, C. John

    2005-01-01

    A simplified single rate expression for hydrogen combustion and nitrogen oxide production was developed. Detailed kinetics are predicted for the chemical kinetic times using the complete chemical mechanism over the entire operating space. These times are then correlated to the reactor conditions using an exponential fit. Simple first order reaction expressions are then used to find the conversion in the reactor. The method uses a two-time step kinetic scheme. The first time averaged step is used at the initial times with smaller water concentrations. This gives the average chemical kinetic time as a function of initial overall fuel air ratio, temperature, and pressure. The second instantaneous step is used at higher water concentrations (> 1 x 10(exp -20) moles/cc) in the mixture which gives the chemical kinetic time as a function of the instantaneous fuel and water mole concentrations, pressure and temperature (T4). The simple correlations are then compared to the turbulent mixing times to determine the limiting properties of the reaction. The NASA Glenn GLSENS kinetics code calculates the reaction rates and rate constants for each species in a kinetic scheme for finite kinetic rates. These reaction rates are used to calculate the necessary chemical kinetic times. This time is regressed over the complete initial conditions using the Excel regression routine. Chemical kinetic time equations for H2 and NOx are obtained for H2/air fuel and for the H2/O2. A similar correlation is also developed using data from NASA s Chemical Equilibrium Applications (CEA) code to determine the equilibrium temperature (T4) as a function of overall fuel/air ratio, pressure and initial temperature (T3). High values of the regression coefficient R2 are obtained.

  18. Measuring neutron spectra in radiotherapy using the nested neutron spectrometer.

    Science.gov (United States)

    Maglieri, Robert; Licea, Angel; Evans, Michael; Seuntjens, Jan; Kildea, John

    2015-11-01

    Out-of-field neutron doses resulting from photonuclear interactions in the head of a linear accelerator pose an iatrogenic risk to patients and an occupational risk to personnel during radiotherapy. To quantify neutron production, in-room measurements have traditionally been carried out using Bonner sphere systems (BSS) with activation foils and TLDs. In this work, a recently developed active detector, the nested neutron spectrometer (NNS), was tested in radiotherapy bunkers. The NNS is designed for easy handling and is more practical than the traditional BSS. Operated in current-mode, the problem of pulse pileup due to high dose-rates is overcome by measuring current, similar to an ionization chamber. In a bunker housing a Varian Clinac 21EX, the performance of the NNS was evaluated in terms of reproducibility, linearity, and dose-rate effects. Using a custom maximum-likelihood expectation-maximization algorithm, measured neutron spectra at various locations inside the bunker were then compared to Monte Carlo simulations of an identical setup. In terms of dose, neutron ambient dose equivalents were calculated from the measured spectra and compared to bubble detector neutron dose equivalent measurements. The NNS-measured spectra for neutrons at various locations in a treatment room were found to be consistent with expectations for both relative shape and absolute magnitude. Neutron fluence-rate decreased with distance from the source and the shape of the spectrum changed from a dominant fast neutron peak near the Linac head to a dominant thermal neutron peak in the moderating conditions of the maze. Monte Carlo data and NNS-measured spectra agreed within 30% at all locations except in the maze where the deviation was a maximum of 40%. Neutron ambient dose equivalents calculated from the authors' measured spectra were consistent (one standard deviation) with bubble detector measurements in the treatment room. The NNS may be used to reliably measure the neutron

  19. Composite boron nitride neutron detectors

    Science.gov (United States)

    Roth, M.; Mojaev, E.; Khakhan, O.; Fleider, A.; Dul`kin, E.; Schieber, M.

    2014-09-01

    Single phase polycrystalline hexagonal boron nitride (BN) or mixed with boron carbide (BxC) embedded in an insulating polymeric matrix acting as a binder and forming a composite material as well as pure submicron size polycrystalline BN has been tested as a thermal neutron converter in a multilayer thermal neutron detector design. Metal sheet electrodes were covered with 20-50 μm thick layers of composite materials and assembled in a multi-layer sandwich configuration. High voltage was applied to the metal electrodes to create an interspacing electric field. The spacing volume could be filled with air, nitrogen or argon. Thermal neutrons were captured in converter layers due to the presence of the 10B isotope. The resulting nuclear reaction produced α-particles and 7Li ions which ionized the gas in the spacing volume. Electron-ion pairs were collected by the field to create an electrical signal proportional to the intensity of the neutron source. The detection efficiency of the multilayer neutron detectors is found to increase with the number of active converter layers. Pixel structures of such neutron detectors necessary for imaging applications and incorporation of internal moderator materials for field measurements of fast neutron flux intensities are discussed as well.

  20. The Fate of Merging Neutron Stars

    Science.gov (United States)

    Kohler, Susanna

    2017-08-01

    A rapidly spinning, highly magnetized neutron star is one possible outcome when two smaller neutron stars merge. [Casey Reed/Penn State University]When two neutron stars collide, the new object that they make can reveal information about the interior physics of neutron stars. New theoretical work explores what we should be seeing, and what it can teach us.Neutron Star or Black Hole?So far, the only systems from which weve detected gravitational waves are merging black holes. But other compact-object binaries exist and are expected to merge on observable timescales in particular, binary neutron stars. When two neutron stars merge, the resulting object falls into one of three categories:a stable neutron star,a black hole, ora supramassive neutron star, a large neutron star thats supported by its rotation but will eventually collapse to a black hole after it loses angular momentum.Histograms of the initial (left) and final (right) distributions of objects in the authors simulations, for five different equations of state. Most cases resulted primarily in the formation of neutron stars (NSs) or supramassive neutron stars (sNSs), not black holes (BHs). [Piro et al. 2017]Whether a binary-neutron-star merger results in another neutron star, a black hole, or a supramassive neutron star depends on the final mass of the remnant and what the correct equation of state is that describes the interiors of neutron stars a longstanding astrophysical puzzle.In a recent study, a team of scientists led by Anthony Piro (Carnegie Observatories) estimated which of these outcomes we should expect for mergers of binary neutron stars. The teams results along with future observations of binary neutron stars may help us to eventually pin down the equation of state for neutron stars.Merger OutcomesPiro and collaborators used relativistic calculations of spinning and non-spinning neutron stars to estimate the mass range that neutron stars would have for several different realistic equations of

  1. Nuclear structure of tellurium 133 via beta decay and shell model calculations in the doubly magic tin 132 region. [J,. pi. , transition probabilities, neutron and proton separation, g factors

    Energy Technology Data Exchange (ETDEWEB)

    Lane, S.M.

    1979-08-01

    An experimental investigation of the level structure of /sup 133/Te was performed by spectroscopy of gamma-rays following the beta-decay of 2.7 min /sup 133/Sb. Multiscaled gamma-ray singles spectra and 2.5 x 10/sup 7/ gamma-gamma coincidence events were used in the assignment of 105 of the approximately 400 observed gamma-rays to /sup 133/Sb decay and in the construction of the /sup 133/Te level scheme with 29 excited levels. One hundred twenty-two gamma-rays were identified as originating in the decay of other isotopes of Sb or their daughter products. The remaining gamma-rays were associated with the decay of impurity atoms or have as yet not been identified. A new computer program based on the Lanczos tridiagonalization algorithm using an uncoupled m-scheme basis and vector manipulations was written. It was used to calculate energy levels, parities, spins, model wavefunctions, neutron and proton separation energies, and some electromagnetic transition probabilities for the following nuclei in the /sup 132/Sn region: /sup 128/Sn, /sup 129/Sn, /sup 130/Sn, /sup 131/Sn, /sup 130/Sb, /sup 131/Sb, /sup 132/Sb, /sup 133/Sb, /sup 132/Te, /sup 133/Te, /sup 134/Te, /sup 134/I, /sup 135/I, /sup 135/Xe, and /sup 136/Xe. The results are compared with experiment and the agreement is generally good. For non-magic nuclei: the lg/sub 7/2/, 2d/sub 5/2/, 2d/sub 3/2/, 1h/sub 11/2/, and 3s/sub 1/2/ orbitals are available to valence protons and the 2d/sub 5/2/, 2d/sub 3/2/, 1h/sub 11/2/, and 3s/sub 1/2/ orbitals are available to valence neutron holes. The present CDC7600 computer code can accommodate 59 single particle states and vectors comprised of 30,000 Slater determinants. The effective interaction used was that of Petrovich, McManus, and Madsen, a modification of the Kallio-Kolltveit realistic force. Single particle energies, effective charges and effective g-factors were determined from experimental data for nuclei in the /sup 132/Sn region. 116 references.

  2. Metrology and quality of radiation therapy dosimetry of electron, photon and epithermal neutron beams

    Energy Technology Data Exchange (ETDEWEB)

    Kosunen, A

    1999-08-01

    In radiation therapy using electron and photon beams the dosimetry chain consists of several sequential phases starting by the realisation of the dose quantity in the Primary Standard Dosimetry Laboratory and ending to the calculation of the dose to a patient. A similar procedure can be described for the dosimetry of epithermal neutron beams in boron neutron capture therapy (BNCT). To achieve the required accuracy of the dose delivered to a patient the quality of all steps in the dosimetry procedure has to be considered. This work is focused on two items in the dosimetry chains: the determination of the dose in the reference conditions and the evaluation of the accuracy of dose calculation methods. The issues investigated and discussed in detail are: a)the calibration methods of plane parallel ionisation chambers used in electron beam dosimetry, (b) the specification of the critical dosimetric parameter i.e. the ratio of stopping powers for water to air, (S I ?){sup water} {sub air}, in photon beams, (c) the feasibility of the twin ionization chamber technique for dosimetry in epithermal neutron beams applied to BNCT and (d) the determination accuracy of the calculated dose distributions in phantoms in electron, photon, and epithermal neutron beams. The results demonstrate that up to a 3% improvement in the consistency of dose determinations in electron beams is achieved by the calibration of plane parallel ionisation chambers in high energy electron beams instead of calibrations in {sup 60}Co gamma beams. In photon beam dosimetry (S I ?){sup water} {sub air} can be determined with an accuracy of 0.2% using the percentage dose at the 10 cm depth, %dd(10), as a beam specifier. The use of %odd(10) requires the elimination of the electron contamination in the photon beam. By a twin ionisation chamber technique the gamma dose can be determined with uncertainty of 6% (1 standard deviation) and the total neutron dose with an uncertainty of 15 to 20% (1 standard deviation

  3. Neutron dosimetry in solid water phantom

    Energy Technology Data Exchange (ETDEWEB)

    Benites-Rengifo, Jorge Luis, E-mail: jlbenitesr@prodigy.net.mx [Centro Estatal de Cancerologia de Nayarit, Calzada de la Cruz 118 Sur, Tepic Nayarit, Mexico and Instituto Tecnico Superior de Radiologia, ITEC, Calle Leon 129, Tepic Nayarit (Mexico); Vega-Carrillo, Hector Rene, E-mail: fermineutron@yahoo.com [Universidad Autonoma de Zacatecas, Unidad Academica de Estudios Nucleares, Apdo. postal 336, 98000, Zacatecas, Zac. (Mexico)

    2014-11-07

    The neutron spectra, the Kerma and the absorbed dose due to neutrons were estimated along the incoming beam in a solid water phantom. Calculations were carried out with the MCNP5 code, where the bunker, the phantom and the model of the15 MV LINAC head were modeled. As the incoming beam goes into the phantom the neutron spectrum is modified and the dosimetric values are reduced.

  4. Fast-neutron interaction with niobium

    Energy Technology Data Exchange (ETDEWEB)

    Smith, A.B.; Guenther, P.T.; Poenitz, W.P.; Smith, D.L.; Whalen, J.F.; Howerton, R.J.

    1985-01-01

    Results of a comprehensive study of the interaction of fast neutrons with niobium are presented, including measurement and interpretation of neutron total, differential-scattering and radiative-capture cross sections. The experimental results are interpreted in the context of the optical-statistical model, with attention to the Fermi-surface anomaly. Experimental results, physical interpretations and rigorous statistical methods are used to provide a comprehensive evaluated nuclear data file suitable for use in a wide range of applied neutronic calculations.

  5. Optimization of neutron source

    Energy Technology Data Exchange (ETDEWEB)

    Hooper, E.B.

    1993-11-09

    I consider here the optimization of the two component neutron source, allowing beam species and energy to vary. A simple model is developed, based on the earlier publications, that permits the optimum to be obtained simply. The two component plasma, with one species of hot ion (D{sup +} or T{sup +}) and the complementary species of cold ion, is easy to analyze in the case of a spatially uniform cold plasma, as to good approximation the total number of hot ions is important but not their spatial distribution. Consequently, the optimization can ignore spatial effects. The problem of a plasma with both types of hot ions and cold ions is rather more difficult, as the neutron production by hot-hot interactions is sensitive to their spatial distributions. Consequently, consideration of this problem will be delayed to a future memorandum. The basic model is that used in the published articles on the two-component, beam-plasma mirror source. I integrate the Fokker-Planck equation analytically, obtaining good agreement with previous numerical results. This simplifies the optimization, by providing a functional form for the neutron production. The primary result is expressed in terms of the power efficiency: watts of neutrons/watts of primary power. The latter includes the positive ion neutralization efficiency. At 150 keV, the present model obtains an efficiency of 0.66%, compared with 0.53% of the earlier calculation.

  6. Report to the DOE nuclear data committee. [EV RANGE 10-100; CROSS SECTIONS; PHOTONEUTRONS; NEUTRONS; GAMMA RADIATION; COUPLED CHANNEL THEORY; DIFFERENTIAL CROSS SECTIONS; MEV RANGE 01-10; ; CAPTURE; GAMMA SPECTRA; THERMAL NEUTRONS; COMPUTER CALCULATIONS; DECAY; FISSION PRODUCTS; FISSION YIELD; SHELL MODELS; NUCLEAR DATA COLLECTIONS

    Energy Technology Data Exchange (ETDEWEB)

    Struble, G.L.; Haight, R.C.

    1981-03-01

    Topics covered include: studies of (n, charged particle) reactions with 14 to 15 MeV neutrons; photoneutron cross sections for /sup 15/N; neutron radiative capture; Lane-model analysis of (p,p) and (n,n) scattering on the even tin isotopes; neutron scattering cross sections for /sup 181/Ta, /sup 197/Au, /sup 209/Bi, /sup 232/Th, and /sup 238/U inferred from proton scattering and charge exchange cross sections; neutron-induced fission cross sections of /sup 245/Cm and /sup 242/Am; fission neutron multiplicities for /sup 245/Cm and /sup 242/Am; the transport of 14 MeV neutrons through heavy materials 150 < A < 208; /sup 249/Cm energy levels from measurement of thermal neutron capture gamma rays; /sup 231/Th energy levels from neutron capture gamma ray and conversion electron spectroscopy; new measurements of conversion electron binding energies in berkelium and californium; nuclear level densities; relative importance of statistical vs. valence neutron capture in the mass-90 region; determination of properties of short-lived fission products; fission yield of /sup 87/Br and /sup 137/I from 15 nuclei ranging from /sup 232/Th to /sup 249/Cf; evaluation of charged particle data for the ECPL library; evaluation of secondary charged-particle energy and angular distributions for ENDL; and evaluated nuclear structure libraries derived from the table of isotopes. (GHT)

  7. Field calibration of PADC track etch detectors for local neutron dosimetry in man using different radiation qualities

    Energy Technology Data Exchange (ETDEWEB)

    Haelg, Roger A., E-mail: rhaelg@phys.ethz.ch [Institute for Radiotherapy, Radiotherapie Hirslanden AG, Hirslanden Medical Center, Rain 34, CH-5000 Aarau (Switzerland); Besserer, Juergen [Institute for Radiotherapy, Radiotherapie Hirslanden AG, Hirslanden Medical Center, Rain 34, CH-5000 Aarau (Switzerland); Boschung, Markus; Mayer, Sabine [Division for Radiation Safety and Security, Paul Scherrer Institut, CH-5232 Villigen (Switzerland); Clasie, Benjamin [Department of Radiation Oncology, Massachusetts General Hospital, 30 Fruit Street, Boston, MA 02114 (United States); Kry, Stephen F. [Department of Radiation Physics, The University of Texas M.D. Anderson Cancer Center, 1515 Holcombe Blvd., Houston, TX 77030 (United States); Schneider, Uwe [Institute for Radiotherapy, Radiotherapie Hirslanden AG, Hirslanden Medical Center, Rain 34, CH-5000 Aarau (Switzerland); Vetsuisse Faculty, University of Zurich, Winterthurerstrasse 204, CH-8057 Zurich (Switzerland)

    2012-12-01

    In order to quantify the dose from neutrons to a patient for contemporary radiation treatment techniques, measurements inside phantoms, representing the patient, are necessary. Published reports on neutron dose measurements cover measurements performed free in air or on the surface of phantoms and the doses are expressed in terms of personal dose equivalent or ambient dose equivalent. This study focuses on measurements of local neutron doses inside a radiotherapy phantom and presents a field calibration procedure for PADC track etch detectors. An initial absolute calibration factor in terms of H{sub p}(10) for personal dosimetry is converted into neutron dose equivalent and additional calibration factors are derived to account for the spectral changes in the neutron fluence for different radiation therapy beam qualities and depths in the phantom. The neutron spectra used for the calculation of the calibration factors are determined in different depths by Monte Carlo simulations for the investigated radiation qualities. These spectra are used together with the energy dependent response function of the PADC detectors to account for the spectral changes in the neutron fluence. The resulting total calibration factors are 0.76 for a photon beam (in- and out-of-field), 1.00 (in-field) and 0.84 (out-of-field) for an active proton beam and 1.05 (in-field) and 0.91 (out-of-field) for a passive proton beam, respectively. The uncertainty for neutron dose measurements using this field calibration method is less than 40%. The extended calibration procedure presented in this work showed that it is possible to use PADC track etch detectors for measurements of local neutron dose equivalent inside anthropomorphic phantoms by accounting for spectral changes in the neutron fluence.

  8. Aerial Neutron Detection: Neutron Signatures for Nonproliferation and Emergency Response Applications

    Energy Technology Data Exchange (ETDEWEB)

    Maurer, Richard J.; Stampahar, Thomas G.; Smith, Ethan X.; Mukhopadhyay, Sanjoy; Wolff, Ronald S.; Rourke, Timothy J.; LeDonne, Jeffrey P.; Avaro, Emanuele; Butler, D. Andre; Borders, Kevin L.; Stampahar, Jezabel; Schuck, William H.; Selfridge, Thomas L.; McKissack, Thomas M.; Duncan, William W.; Hendricks, Thane J.

    2012-10-17

    From 2007 to the present, the Remote Sensing Laboratory has been conducting a series of studies designed to expand our fundamental understanding of aerial neutron detection with the goal of designing an enhanced sensitivity detection system for long range neutron detection. Over 35 hours of aerial measurements in a helicopter were conducted for a variety of neutron emitters such as neutron point sources, a commercial nuclear power reactor, nuclear reactor spent fuel in dry cask storage, depleted uranium hexafluoride and depleted uranium metal. The goals of the project were to increase the detection sensitivity of our instruments such that a 5.4 × 104 neutron/second source could be detected at 100 feet above ground level at a speed of 70 knots and to enhance the long-range detection sensitivity for larger neutron sources, i.e., detection ranges above 1000 feet. In order to increase the sensitivity of aerial neutron detection instruments, it is important to understand the dynamics of the neutron background as a function of altitude. For aerial neutron detection, studies have shown that the neutron background primarily originates from above the aircraft, being produced in the upper atmosphere by galactic cosmic-ray interactions with air molecules. These interactions produce energetic neutrons and charged particles that cascade to the earth’s surface, producing additional neutrons in secondary collisions. Hence, the neutron background increases as a function of altitude which is an impediment to long-range neutron detection. In order to increase the sensitivity for long range detection, it is necessary to maintain a low neutron background as a function of altitude. Initial investigations show the variation in the neutron background can be decreased with the application of a cosmic-ray shield. The results of the studies along with a representative data set are presented.

  9. Application of neural networks for the calculation of technical losses of electric energy in air power lines 6-35 kV

    Directory of Open Access Journals (Sweden)

    Володимир Леонідович Бакулевський

    2015-11-01

    Full Text Available A model for the calculation of technical losses of electricity in the air lines with voltage of 6-35 kV based on neural networks with due regard to meteorological factors has been worked out; the main components of the model have been considered and researched; the best ones being selected, that is: a set of input variables, volume of excerpts (training, control and testing, architecture and network activation function, network learning algorithm was proposed. Simulation was conducted in OS STATISTICA Neural Networks. Input variables are: transmission line (TL active load, transmission line rated voltage, transmission line cross section and length of wire, average air temperature, wind speed, rainfall availability; output variable – that is technical losses in electric transmission line. To select the optimal input vector model the data selection methods were used: variables testing using trial and error method, variables stepped inclusion and exclusion algorithm. It has been proved that the most important variables are TL active load and average air temperature. all input variables under review should be included in the created artificial neural network (ANN. It was determined that the optimal volume for ANN training set given parameters made 250 observations, control and test excerpts volume were respectively 250 and 332 observations. It has been proved that the best type of architecture is multilayer perceptron ANN that being compared to radial basis functions and generalized regression network is characterized by minimal errors and complexity of the network. ANN of the following architecture: multilayer perceptron, 7 neurons in the input layer, 5 neurons in the hidden layer, 1 output neuron, logistics as activation function – has been taken optimal

  10. Neutron diffraction

    Energy Technology Data Exchange (ETDEWEB)

    Heger, G. [Rheinisch-Westfaelische Technische Hochschule Aachen, Inst. fuer Kristallographie, Aachen (Germany)

    1996-12-31

    X-ray diffraction using conventional laboratory equipment and/or synchrotron installations is the most important method for structure analyses. The purpose of this paper is to discuss special cases, for which, in addition to this indispensable part, neutrons are required to solve structural problems. Even though the huge intensity of modern synchrotron sources allows in principle the study of magnetic X-ray scattering the investigation of magnetic structures is still one of the most important applications of neutron diffraction. (author) 15 figs., 1 tab., 10 refs.

  11. Neutronics of pulsed spallation neutron sources

    CERN Document Server

    Watanabe, N

    2003-01-01

    Various topics and issues on the neutronics of pulsed spallation neutron sources, mainly for neutron scattering experiments, are reviewed to give a wide circle of readers a better understanding of these sources in order to achieve a high neutronic performance. Starting from what neutrons are needed, what the spallation reaction is and how to produce slow-neutrons more efficiently, the outline of the target and moderator neutronics are explained. Various efforts with some new concepts or ideas have already been devoted to obtaining the highest possible slow-neutron intensity with desired pulse characteristics. This paper also reviews the recent progress of such efforts, mainly focused on moderator neutronics, since moderators are the final devices of a neutron source, which determine the source performance. Various governing parameters for neutron-pulse characteristics such as material issues, geometrical parameters (shape and dimensions), the target-moderator coupling scheme, the ortho-para-hydrogen ratio, po...

  12. Measurements of secondary neutrons from cosmic radiation with a Bonner sphere spectrometer at 79 degrees N.

    Science.gov (United States)

    Rühm, Werner; Mares, V; Pioch, C; Weitzenegger, E; Vockenroth, R; Paretzke, H G

    2009-04-01

    Air crew members and airline passengers are continuously exposed to cosmic radiation during their flights. Particles ejected by the sun during so-called solar particle events (SPEs) in periods of high solar activity can contribute to this exposure. In rare cases the dose from a single SPE might even exceed the annual dose limit of 1 mSv above which dose monitoring of air crews is legally required in Germany. Measurements performed by means of neutron monitors have already shown that the relative intensity of secondary neutrons from cosmic radiation is enhanced during an SPE, particularly at regions close to the magnetic poles of the Earth where shielding of the cosmic radiation by the geomagnetic field is low. Here we describe a Bonner sphere spectrometer installed at the Koldewey station at 79 degrees N, i.e. about 1,000 km from the geographic North pole, which is designed to provide first experimental data on the time-dependent energy spectrum of neutrons produced in the atmosphere during an SPE. This will be important to calculate doses from these neutrons to air crew members. The system is described in detail and first results are shown that were obtained during quiet periods of sun activity.

  13. Antiproton radiotherapy: peripheral dose from secondary neutrons

    DEFF Research Database (Denmark)

    Fahimian, Benjamin P.; DeMarco, John J.; Keyes, Roy

    2009-01-01

    is the normal tissue dose resulting from secondary neutrons produced in the annihilation of antiprotons on the nucleons of the target atoms. Here we present the first organ specific Monte Carlo calculations of normal tissue equivalent neutron dose in antiproton therapy through the use of a segmented CT...

  14. Energy measurement of prompt fission neutrons in 239Pu(n,f) for incident neutron energies from 1 to 200 MeV

    CERN Document Server

    Chatillon, A; Granier, Th; Laurent, B; Taïeb, J; Noda, S; Haight, R C; Devlin, M; Nelson, R O; O’Donnell, J M

    2010-01-01

    Prompt fission neutron spectra in the neutron-induced fission of 239Pu have been measured for incident neutron energies from 1 to 200 MeV at the Los Alamos Neutron Science Center. Preliminary results are discussed and compared to theoretical model calculation.

  15. Methods for absorbing neutrons

    Science.gov (United States)

    Guillen, Donna P [Idaho Falls, ID; Longhurst, Glen R [Idaho Falls, ID; Porter, Douglas L [Idaho Falls, ID; Parry, James R [Idaho Falls, ID

    2012-07-24

    A conduction cooled neutron absorber may include a metal matrix composite that comprises a metal having a thermal neutron cross-section of at least about 50 barns and a metal having a thermal conductivity of at least about 1 W/cmK. Apparatus for providing a neutron flux having a high fast-to-thermal neutron ratio may include a source of neutrons that produces fast neutrons and thermal neutrons. A neutron absorber positioned adjacent the neutron source absorbs at least some of the thermal neutrons so that a region adjacent the neutron absorber has a fast-to-thermal neutron ratio of at least about 15. A coolant in thermal contact with the neutron absorber removes heat from the neutron absorber.

  16. Statistical elements in calculations procedures for air quality control; Elementi di statistica nelle procedure di calcolo per il controllo della qualita' dell'aria

    Energy Technology Data Exchange (ETDEWEB)

    Mura, M.C. [Istituto Superiore di Sanita' , Laboratorio di Igiene Ambientale, Rome (Italy)

    2001-07-01

    The statistical processing of data resulting from the monitoring of chemical atmospheric pollution aimed at air quality control is presented. The form of procedural models may offer a practical instrument to the operators in the sector. The procedural models are modular and can be easily integrated with other models. They include elementary calculation procedures and mathematical methods for statistical analysis. The calculation elements have been developed by probabilistic induction so as to relate them to the statistical analysis. The calculation elements have been developed by probabilistic induction so as to relate them to the statistical models, which are the basis of the methods used for the study and the forecast of atmospheric pollution. This report is part of the updating and training activity that the Istituto Superiore di Sanita' has been carrying on for over twenty years, addressed to operators of the environmental field. [Italian] Il processo di elaborazione statistica dei dati provenienti dal monitoraggio dell'inquinamento chimico dell'atmosfera, finalizzato al controllo della qualita' dell'aria, e' presentato in modelli di procedure al fine di fornire un sintetico strumento di lavoro agli operatori del settore. I modelli di procedure sono modulari ed integrabili. Includono gli elementi di calcolo elementare ed i metodi statistici d'analisi. Gli elementi di calcolo sono sviluppati con metodo d'induzione probabilistica per collegarli ai modelli statistici, che sono alla base dei metodi d'analisi nello studio del fenomeno dell'inquinamento atmosferico anche a fini previsionali. Il rapporto si inserisce nell'attivita' di aggiornamento e di formazione che fin dagli anni ottanta l'Istituto Superiore di Sanita' indirizza agli operatori del settore ambientale.

  17. Approach to the calculation of energy deposition in a container of fuel irradiated by the neutronic codes coupling fluid-dynamics; Aprpoximacion al calculo de la deposicion energetica en un contenedor de combustible irradiado mediante el acoplamiento de codigos neutronico fluido-dinamicos

    Energy Technology Data Exchange (ETDEWEB)

    Hueso, C.; Aleman, A.; Colomer, C.; Fabbri, M.; Martin, M.; Saellas, J.

    2013-07-01

    In this work identifies a possible area of improvement through the creation of a code of coupling between deposition energy codes which calculate neutron (MCNP), and data from heading into fluid dynamics (ANSYS-Fluent) or codes thermomechanical, called MAFACS (Monte Carlo ANSYS Fluent Automatic Coupling Software), being possible to so summarize the process by shortening the needs of computing time, increasing the precision of the results and therefore improving the design of the components.

  18. Basic of Neutron NDA

    Energy Technology Data Exchange (ETDEWEB)

    Trahan, Alexis Chanel [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-09-15

    The objectives of this presentation are to introduce the basic physics of neutron production, interactions and detection; identify the processes that generate neutrons; explain the most common neutron mechanism, spontaneous and induced fission and (a,n) reactions; describe the properties of neutron from different sources; recognize advantages of neutron measurements techniques; recognize common neutrons interactions; explain neutron cross section measurements; describe the fundamental of 3He detector function and designs; and differentiate between passive and active assay techniques.

  19. Functional renormalization group approach to neutron matter

    Directory of Open Access Journals (Sweden)

    Matthias Drews

    2014-11-01

    Full Text Available The chiral nucleon-meson model, previously applied to systems with equal number of neutrons and protons, is extended to asymmetric nuclear matter. Fluctuations are included in the framework of the functional renormalization group. The equation of state for pure neutron matter is studied and compared to recent advanced many-body calculations. The chiral condensate in neutron matter is computed as a function of baryon density. It is found that, once fluctuations are incorporated, the chiral restoration transition for pure neutron matter is shifted to high densities, much beyond three times the density of normal nuclear matter.

  20. Neutron streaming studies along JET shielding penetrations

    Directory of Open Access Journals (Sweden)

    Stamatelatos Ion E.

    2017-01-01

    Full Text Available Neutronic benchmark experiments are carried out at JET aiming to assess the neutronic codes and data used in ITER analysis. Among other activities, experiments are performed in order to validate neutron streaming simulations along long penetrations in the JET shielding configuration. In this work, neutron streaming calculations along the JET personnel entrance maze are presented. Simulations were performed using the MCNP code for Deuterium-Deuterium and Deuterium- Tritium plasma sources. The results of the simulations were compared against experimental data obtained using thermoluminescence detectors and activation foils.

  1. Velocity-space sensitivity of neutron spectrometry measurements

    DEFF Research Database (Denmark)

    Jacobsen, Asger Schou; Salewski, Mirko; Eriksson, J.

    2015-01-01

    Neutron emission spectrometry (NES) measures the energies of neutrons produced in fusion reactions. Here we present velocity-space weight functions for NES and neutron yield measurements. Weight functions show the sensitivity as well as the accessible regions in velocity space for a given range...... of the neutron energy spectrum. Combined with a calculated fast-ion distribution function, they determine the part of the distribution function producing detectable neutrons in a given neutron energy range. Furthermore, we construct a forward model based on weight functions capable of rapidly calculating neutron...... energy spectra. This forward model can be inverted and could thereby be used to directly measure the fast-ion phase-space distribution functions, possibly in combination with other fast-ion diagnostics. The presented methods and results can be applied to neutron energy spectra measured by any kind...

  2. Neutron cross sections for fusion. [Review

    Energy Technology Data Exchange (ETDEWEB)

    Haight, R.C.

    1979-10-01

    First generation fusion reactors will most likely be based on the /sup 3/H(d,n)/sup 4/He reaction, which produces 14-MeV neutrons. In these reactors, both the number of neutrons and the average neutron energy will be significantly higher than for fission reactors of the same power. Accurate neutron cross section data are therefore of great importance. They are needed in present conceptual designs to calculate neutron transport, energy deposition, nuclear transmutation including tritium breeding and activation, and radiation damage. They are also needed for the interpretation of radiation damage experiments, some of which use neutrons up to 40 MeV. In addition, certain diagnostic measurements of plasma experiments require nuclear cross sections. The quality of currently available data for these applications will be reviewed and current experimental programs will be outlined. The utility of nuclear models to provide these data also will be discussed. 65 references.

  3. Effective mass of free neutrons in neutron star crust

    Energy Technology Data Exchange (ETDEWEB)

    Chamel, Nicolas [Copernicus Astronomical Center (CAMK), Polish Academy of Sciences, ul. Bartycka 18, 00-716 Warsaw (Poland) and LUTH, Paris Observatory, 5 place Jules Janssen, 92195 Meudon (France)]. E-mail: nchamel@camk.edu.pl

    2006-07-24

    The inner layers of a neutron star crust, composed of a Coulomb lattice of neutron rich nuclear clusters immersed in a sea of 'free' superfluid neutrons, are closely analogous to periodic condensed matter systems such as electronic, photonic or phononic crystals. Applying methods from solid state physics to the neutron star context, it has been recently shown that Bragg scattering leads to a strong renormalization of the neutron mass in the bottom layers of the crust. Knowledge of this effective mass throughout the crust is essential in order to understand the dynamical properties of the neutron superfluid and the origin of pulsar glitches. The purpose of the present work is to evaluate this effective mass in the outermost layers of the inner crust, near the drip point {rho}{sub drip}{approx}4x10{sup 11} g-bar cm{sup -3}. The results are compared with the case of electrons in ordinary solids and as an example the corresponding effective electron mass in copper is calculated.

  4. Effective delayed neutron fraction and prompt neutron lifetime of Tehran research reactor mixed-core

    Science.gov (United States)

    Lashkari, A.; Khalafi, H.; Kazeminejad, H.

    2013-01-01

    In this work, kinetic parameters of Tehran research reactor (TRR) mixed cores have been calculated. The mixed core configurations are made by replacement of the low enriched uranium control fuel elements with highly enriched uranium control fuel elements in the reference core. The MTR_PC package, a nuclear reactor analysis tool, is used to perform the analysis. Simulations were carried out to compute effective delayed neutron fraction and prompt neutron lifetime. Calculation of kinetic parameters is necessary for reactivity and power excursion transient analysis. The results of this research show that effective delayed neutron fraction decreases and prompt neutron lifetime increases with the fuels burn-up. Also, by increasing the number of highly enriched uranium control fuel elements in the reference core, the prompt neutron lifetime increases, but effective delayed neutron fraction does not show any considerable change. PMID:24976672

  5. Characterization of the source neutrons produced by the Frascati Neutron Generator

    Energy Technology Data Exchange (ETDEWEB)

    Pillon, M. [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); Rado, V. [Associazione EURATOM-ENEA sulla Fusione, Frascati (Italy)

    1995-03-01

    Integral experiments with a neutron source based on the T(d,n){alpha} reaction require a good knowledge of the source neutron distribution. The source neutrons are energy-angle distributed with an energy spread of a few mega-electron-volts and anisotropy emission of a few percent.The characterization of the 14-MeV neutron source of the Frascati Neutron Generator (FNG) has been carried out using an absolutely calibrated U{sup Nat} microfission chamber with plane electrodes and {sup 93}Nb(n,2n){sup 92m}Nb activation reaction. The reaction rates have been measured at different angles with respect to the beam direction. The experimental results are compared with the corresponding quantities calculated by using a Monte Carlo code (MCNP) with an ``ad hoc`` source model. The calculated neutron emission reproduces the measured one satisfactorily well. Improvement of the source model used is discussed. (orig.).

  6. A feasibility study of a deuterium-deuterium neutron generator-based boron neutron capture therapy system for treatment of brain tumors.

    Science.gov (United States)

    Hsieh, Mindy; Liu, Yingzi; Mostafaei, Farshad; Poulson, Jean M; Nie, Linda H

    2017-02-01

    Boron neutron capture therapy (BNCT) is a binary treatment modality that uses high LET particles to achieve tumor cell killing. Deuterium-deuterium (DD) compact neutron generators have advantages over nuclear reactors and large accelerators as the BNCT neutron source, such as their compact size, low cost, and relatively easy installation. The purpose of this study is to design a beam shaping assembly (BSA) for a DD neutron generator and assess the potential of a DD-based BNCT system using Monte Carlo (MC) simulations. The MC model consisted of a head phantom, a DD neutron source, and a BSA. The head phantom had tally cylinders along the centerline for computing neutron and photon fluences and calculating the dose as a function of depth. The head phantom was placed at 4 cm from the BSA. The neutron source was modeled to resemble the source of our current DD neutron generator. A BSA was designed to moderate and shape the 2.45-MeV DD neutrons to the epithermal (0.5 eV to 10 keV) range. The BSA had multiple components, including moderator, reflector, collimator, and filter. Various materials and configurations were tested for each component. Each BSA layout was assessed in terms of the in-air and in-phantom parameters. The maximum brain dose was limited to 12.5 Gray-Equivalent (Gy-Eq) and the skin dose to 18 Gy-Eq. The optimized BSA configuration included 30 cm of lead for reflector, 45 cm of LiF, and 10 cm of MgF2 for moderator, 10 cm of lead for collimator, and 0.1 mm of cadmium for thermal neutron filter. Epithermal flux at the beam aperture was 1.0 × 105  nepi /cm2 -s; thermal-to-epithermal neutron ratio was 0.05; fast neutron dose per epithermal was 5.5 × 10-13  Gy-cm2 /φepi , and photon dose per epithermal was 2.4 × 10-13  Gy-cm2 /φepi . The AD, AR, and the advantage depth dose rate were 12.1 cm, 3.7, and 3.2 × 10-3  cGy-Eq/min, respectively. The maximum skin dose was 0.56 Gy-Eq. The DD neutron yield that is needed to irradiate in

  7. Computational characterization and experimental validation of the thermal neutron source for neutron capture therapy research at the University of Missouri

    Energy Technology Data Exchange (ETDEWEB)

    Broekman, J. D. [University of Missouri, Research Reactor Center, 1513 Research Park Drive, Columbia, MO 65211-3400 (United States); Nigg, D. W. [Idaho National Laboratory, PO Box 1625, Idaho Falls, ID 83415 (United States); Hawthorne, M. F. [University of Missouri, International Institute of Nano and Molecular Medicine, 1514 Research Park Dr., Columbia, MO 65211-3450 (United States)

    2013-07-01

    Parameter studies, design calculations and neutronic performance measurements have been completed for a new thermal neutron beamline constructed for neutron capture therapy cell and small-animal radiobiology studies at the University of Missouri Research Reactor. The beamline features the use of single-crystal silicon and bismuth sections for neutron filtering and for reduction of incident gamma radiation. The computational models used for the final beam design and performance evaluation are based on coupled discrete-ordinates and Monte Carlo techniques that permit detailed modeling of the neutron transmission properties of the filtering crystals with very few approximations. Validation protocols based on neutron activation spectrometry measurements and rigorous least-square adjustment techniques show that the beam produces a neutron spectrum that has the anticipated level of thermal neutron flux and a somewhat higher than expected, but radio-biologically insignificant, epithermal neutron flux component. (authors)

  8. Metrology and quality of radiation therapy dosimetry of electron, photon and epithermal neutron beams

    Science.gov (United States)

    Kosunen, Antti

    In radiation therapy using electron and photon beams the dosimetry chain consists of several sequential phases starting by the realization of the dose quantity in the Primary Standard Dosimetry Laboratory and ending to the calculation of the dose to a patient. A similar procedure can be described for the dosimetry of epithermal neutron beams in boron neutron capture therapy (BNCT). This work is focused on two items in the dosimetry chains: the determination of the dose in the reference conditions and the evaluation of the accuracy of dose calculation methods. The issues investigated and discussed in detail are: (a)the calibration methods of plane parallel ionization chambers used in electron beam dosimetry, (b)the specification of the critical dosimetric parameter i.e. the ratio of stopping powers for water to air, (S/r) waterair , in photon beams, (c)the feasibility of the twin ionization chamber technique for dosimetry in epithermal neutron beams applied to BNCT and (d)the determination accuracy of the calculated dose distributions in phantoms in electron, photon, and epithermal neutron beams. The results demonstrate that UP to a 3% improvement in the consistency of dose determinations in electron beams is achieved by the calibration of plane parallel ionization chambers in high energy electron beams instead of calibrations in 60Co gamma beams. In photon beam dosimetry (S/r) waterair can be determined with an accuracy of 0.2% using the percentage dose at the 10 cm depth, %dd(10), as a beam specifier. By a twin ionization chamber technique accuracy the gamma dose can be determined with uncertainty of 6% (1 standard deviation) and the total neutron dose with an uncertainty of 15 to 20% (1 standard deviation). The general accuracy achieved by treatment planning systems is approximately 4% for photons and 5 to 7% for electrons. Large (>10%) deviations in calculated doses are possible even when relatively modern calculation approaches are used.

  9. Black Hole - Neutron Star Binary Mergers

    Data.gov (United States)

    National Aeronautics and Space Administration — Gravitational radiation waveforms for black hole-neutron star coalescence calculations. The physical input is Newtonian physics, an ideal gas equation of state with...

  10. Neutron-star matter within the energy-density functional theory and neutron-star structure

    Energy Technology Data Exchange (ETDEWEB)

    Fantina, A. F.; Chamel, N.; Goriely, S. [Institut d' Astronomie et d' Astrophysique, CP226, Université Libre de Bruxelles (ULB), 1050 Brussels (Belgium); Pearson, J. M. [Dépt. de Physique, Université de Montréal, Montréal (Québec), H3C 3J7 (Canada)

    2015-02-24

    In this lecture, we will present some nucleonic equations of state of neutron-star matter calculated within the nuclear energy-density functional theory using generalized Skyrme functionals developed by the Brussels-Montreal collaboration. These equations of state provide a consistent description of all regions of a neutron star. The global structure of neutron stars predicted by these equations of state will be discussed in connection with recent astrophysical observations.

  11. Neutron spectrum for neutron capture therapy in boron; Espectro de neutrones para terapia por captura de neutrones en boro

    Energy Technology Data Exchange (ETDEWEB)

    Medina C, D.; Soto B, T. G. [Universidad Autonoma de Zacatecas, Unidad Academica de Estudios Nucleares, Programa de Doctorado en Ciencias Basicas, 98068 Zacatecas, Zac. (Mexico); Baltazar R, A. [Universidad Autonoma de Zacatecas, Unidad Academica de Ingenieria Electrica, Programa de Doctorado en Ingenieria y Tecnologia Aplicada, 98068 Zacatecas, Zac. (Mexico); Vega C, H. R., E-mail: dmedina_c@hotmail.com [Universidad Autonoma de Zacatecas, Unidad Academica de Estudios Nucleares, Cipres No. 10, Fracc. La Penuela, 98068 Zacatecas, Zac. (Mexico)

    2016-10-15

    Glioblastoma multiforme is the most common and aggressive of brain tumors and is difficult to treat by surgery, chemotherapy or conventional radiation therapy. One treatment alternative is the Neutron Capture Therapy in Boron, which requires a beam modulated in neutron energy and a drug with {sup 10}B able to be fixed in the tumor. When the patients head is exposed to the neutron beam, they are captured by the {sup 10}B and produce a nucleus of {sup 7}Li and an alpha particle whose energy is deposited in the cancer cells causing it to be destroyed without damaging the normal tissue. One of the problems associated with this therapy is to have an epithermal neutrons flux of the order of 10{sup 9} n/cm{sup 2}-sec, whereby irradiation channels of a nuclear research reactor are used. In this work using Monte Carlo methods, the neutron spectra obtained in the radial irradiation channel of the TRIGA Mark III reactor are calculated when inserting filters whose position and thickness have been modified. From the arrangements studied, we found that the Fe-Cd-Al-Cd polyethylene filter yielded a ratio between thermal and epithermal neutron fluxes of 0.006 that exceeded the recommended value (<0.05), and the dose due to the capture gamma rays is lower than the dose obtained with the other arrangements studied. (Author)

  12. Study of neutron focusing at the Texas Cold Neutron Source. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Wehring, B.W.; Uenlue, K.

    1996-12-19

    The goals of this three-year study were: (1) design a neutron focusing system for use with the Texas Cold Neutron Source (TCNS) to produce an intense beam of cold neutrons appropriate for prompt gamma activation analysis (PGAA); (2) orchestrate the construction of the focusing system, integrate it into the TCNS neutron guide complex, and measure its performance; and (3) design, setup, and test a cold-neutron PGAA system which utilizes the guided focused cold neutron beam. During the first year of the DOE grant, a new procedure was developed and used to design a focusing converging guide consisting of truncated rectangular cone sections. Detailed calculations were performed using a 3-D Monte Carlo code which the authors wrote to trace neutrons through the curved guide of the TCNS into the proposed converging guide. Using realistic reflectivities for Ni-Ti supermirrors, the authors obtained gains of 3 to 5 for 4 different converging guide geometries. During the second year of the DOE grant, the subject of this final report, Ovonic Synthetic Materials Company was contracted to build a converging neutron guide focusing system to the specifications. Considerable time and effort were spent working with Ovonics on selecting the materials for the converging neutron guide system. The major portion of the research on the design of a cold-neutron PGAA system was also completed during the second year. At the beginning of the third year of the grant, a converging neutron guide focusing system had been ordered, and a cold-neutron PGAA system had been designed. Since DOE did not fund the third year, there was no money to purchase the required equipment for the cold-neutron PGAA system and no money to perform tests of either the converging neutron guide or the cold-neutron PGAA system. The research already accomplished would have little value without testing the systems which had been designed. Thus the project was continued at a pace that could be sustained with internal funding.

  13. On the significance of the energy correlations of spallation neutrons on the neutron fluctuations in accelerator-driven subcritical systems

    CERN Document Server

    Pázsit, I; Fhager, V

    2000-01-01

    Studies of neutron fluctuations in spallation-driven subcritical systems require the use of energy-dependent master equations. In particular, calculation of the second moment of the neutron distribution requires knowledge on the energy correlations (two-point distributions) of the source particles. It is shown here that such correlations will exist even if the energies of all neutrons, generated in any single spallation event, are independent, provided that the energy distribution of the neutrons for separate spallation events is dependent on the number of neutrons generated. A simple model of number dependence of the energy spectrum is constructed, and the arising energy correlations are calculated. The error in calculating the second moment of the neutron distribution, arising when assuming zero correlations (i.e. using only one-particle energy spectra), is estimated in a simple model of neutron slowing down.

  14. Neutron resonances in planar waveguides

    Energy Technology Data Exchange (ETDEWEB)

    Kozhevnikov, S. V., E-mail: kozhevn@nf.jinr.ru, E-mail: kzh-sv@mail.ru; Ignatovich, V. K.; Petrenko, A. V. [Joint Institute for Nuclear Research, Neutron Physics Laboratory (Russian Federation); Radu, F. [Helmholtz-Zentrum Berlin für Materialen und Energie (Germany)

    2016-12-15

    We report on the results of the experimental investigation of the spectral width of neutron resonances in planar waveguides using the time-of-flight method and recording the microbeam emerging from the waveguide end. Experimental data are compared with the results of theoretical calculations.

  15. Peninsulas of the neutron stability of nuclei in the vicinity of neutron magic numbers

    Energy Technology Data Exchange (ETDEWEB)

    Tarasov, V. N., E-mail: vtarasov@kipt.kharkov.ua [Kharkov Institute of Physics and Technology, National Science Center (Ukraine); Gridnev, K. A. [St. Petersburg State University, Fock Institute of Physics (Russian Federation); Greiner, W.; Gridnev, D. K. [Johann Wolfgang Goethe-Universitaet, Institut fuer Theoretische Physik (Germany); Kuprikov, V. I.; Tarasov, D. V. [Kharkov Institute of Physics and Technology, National Science Center (Ukraine); Vinas, X. [Universitat de Barcelona (Spain)

    2012-01-15

    On the basis of the Hartree-Fock method as implemented with Skyrme forces (Ska, SkM*, Sly4, and SkI2) and with allowance for an axial deformation and nucleon pairing in the Bardeen-Cooper-Schrieffer approximation, the properties of extremely neutron-rich even-even nuclei were calculated beyond the neutron drip line known earlier from theoretical calculations. It was shown that the chains of isotopes beyond the neutron drip line that contain N = 32, 58, 82, 126, and 184 neutrons form peninsulas of nuclei stable against the emission of one neutron and, in some cases, peninsulas of nuclei stable against the emission of two neutrons. The neutron- and proton-density distributions in nuclei forming stability peninsulas were found to be spherically symmetric. A mechanism via which the stability of nuclei might be restored beyond the neutron drip line was discussed. A comparison with the results of calculations by the Hartree-Fock-Bogolyubov method was performed for long chains of sulfur and gadolinium isotopes up to the neutron drip line.

  16. Humidity coefficient correction in the calculation equations of air refractive index by He-Ne laser based on phase step interferometry.

    Science.gov (United States)

    Chen, Qianghua; Liu, Jinghai; He, Yongxi; Luo, Huifu; Luo, Jun; Wang, Feng

    2015-02-10

    The refractive index of air (RIA) is an important parameter in precision measurement. The revisions to Edlen's equations by Boensch and Potulski [Metrologia 35, 133 (1998)] are mostly used to calculate the RIA at present. Since the humidity correction coefficients in the formulas were performed with four wavelengths of a Cd(114) lamp (644.0, 508.7, 480.1, and 467.9 nm) and at the temperature range of 19.6°C-20.1°C, the application is restricted when an He-Ne laser is used as the light source, which is mostly applied in optical precision measurement, and the environmental temperature is far away from 20°C as well. To solve this problem, a measurement system based on phase step interferometry for measuring the effect of the humidity to the RIA is presented, and a corresponding humidity correction equation is derived. The analysis and comparison results show that the uncertainty of the presented equation is better than that of Boensch and Potulski's. It is more suitable in present precision measurements by He-Ne laser, and the application temperature range extends to 14.6°C-24.0°C as well.

  17. Spallation Neutron Source (SNS)

    Data.gov (United States)

    Federal Laboratory Consortium — The SNS at Oak Ridge National Laboratory is a next-generation spallation neutron source for neutron scattering that is currently the most powerful neutron source in...

  18. Neutron scattering. Lectures

    Energy Technology Data Exchange (ETDEWEB)

    Brueckel, Thomas; Heger, Gernot; Richter, Dieter; Roth, Georg; Zorn, Reiner (eds.)

    2010-07-01

    The following topics are dealt with: Neutron sources, neutron properties and elastic scattering, correlation functions measured by scattering experiments, symmetry of crystals, applications of neutron scattering, polarized-neutron scattering and polarization analysis, structural analysis, magnetic and lattice excitation studied by inelastic neutron scattering, macromolecules and self-assembly, dynamics of macromolecules, correlated electrons in complex transition-metal oxides, surfaces, interfaces, and thin films investigated by neutron reflectometry, nanomagnetism. (HSI)

  19. NUCLEAR HEATING IN LIF DOSEMETERS IN A FUSION NEUTRON FIELD, TRIAL OF DIRECT COMPARISON OF EXPERIMENTAL AND SIMULATED RESULTS.

    Science.gov (United States)

    Pohorecki, Wladyslaw; Obryk, Barbara

    2017-09-29

    The results of nuclear heating measured by means of thermoluminescent dosemeters (TLD-LiF) in a Cu block irradiated by 14 MeV neutrons are presented. The integral Cu experiment relevant for verification of copper nuclear data at neutron energies characteristic for fusion facilities was performed in the ENEA FNG Laboratory at Frascati. Five types of TLDs were used: highly photon sensitive LiF:Mg,Cu,P (MCP-N), 7LiF:Mg,Cu,P (MCP-7) and standard, lower sensitivity LiF:Mg,Ti (MTS-N), 7LiF:Mg,Ti (MTS-7) and 6LiF:Mg,Ti (MTS-6). Calibration of the detectors was performed with gamma rays in terms of air-kerma (10 mGy of 137Cs air-kerma). Nuclear heating in the Cu block was also calculated with the use of MCNP transport code Nuclear heating in Cu and air in TLD's positions was calculated as well. The nuclear heating contribution from all simulated by MCNP6 code particles including protons, deuterons, alphas tritons and heavier ions produced by the neutron interactions were calculated. A trial of the direct comparison between experimental results and results of simulation was performed. © The Author 2017. Published by Oxford University Press. All rights reserved. For Permissions, please email: journals.permissions@oup.com.

  20. Neutron Therapy Facility

    Data.gov (United States)

    Federal Laboratory Consortium — The Neutron Therapy Facility provides a moderate intensity, broad energy spectrum neutron beam that can be used for short term irradiations for radiobiology (cells)...

  1. Comparison of Bonner sphere responses calculated by different Monte Carlo codes at energies between 1 MeV and 1 GeV – Potential impact on neutron dosimetry at energies higher than 20 MeV

    CERN Document Server

    Rühm, W; Pioch, C; Agosteo, S; Endo, A; Ferrarini, M; Rakhno, I; Rollet, S; Satoh, D; Vincke, H

    2014-01-01

    Bonner Spheres Spectrometry in its high-energy extended version is an established method to quantify neutrons at a wide energy range from several meV up to more than 1 GeV. In order to allow for quantitative measurements, the responses of the various spheres used in a Bonner Sphere Spectrometer (BSS) are usually simulated by Monte Carlo (MC) codes over the neutron energy range of interest. Because above 20 MeV experimental cross section data are scarce, intra-nuclear cascade (INC) and evaporation models are applied in these MC codes. It was suspected that this lack of data above 20 MeV may translate to differences in simulated BSS response functions depending on the MC code and nuclear models used, which in turn may add to the uncertainty involved in Bonner Sphere Spectrometry, in particular for neutron energies above 20 MeV. In order to investigate this issue in a systematic way, EURADOS (European Radiation Dosimetry Group) initiated an exercise where six groups having experience in neutron transport calcula...

  2. Variation in lunar neutron dose estimates.

    Science.gov (United States)

    Slaba, Tony C; Blattnig, Steve R; Clowdsley, Martha S

    2011-12-01

    The radiation environment on the Moon includes albedo neutrons produced by primary particles interacting with the lunar surface. In this work, HZETRN2010 is used to calculate the albedo neutron contribution to effective dose as a function of shielding thickness for four different space radiation environments and to determine to what extent various factors affect such estimates. First, albedo neutron spectra computed with HZETRN2010 are compared to Monte Carlo results in various radiation environments. Next, the impact of lunar regolith composition on the albedo neutron spectrum is examined, and the variation on effective dose caused by neutron fluence-to-effective dose conversion coefficients is studied. A methodology for computing effective dose in detailed human phantoms using HZETRN2010 is also discussed and compared. Finally, the combined variation caused by environmental models, shielding materials, shielding thickness, regolith composition and conversion coefficients on the albedo neutron contribution to effective dose is determined. It is shown that a single percentage number for characterizing the albedo neutron contribution to effective dose can be misleading. In general, the albedo neutron contribution to effective dose is found to vary between 1-32%, with the environmental model, shielding material and shielding thickness being the driving factors that determine the exact contribution. It is also shown that polyethylene or other hydrogen-rich materials may be used to mitigate the albedo neutron exposure.

  3. Impact of nuclear data on fast neutron therapy

    Energy Technology Data Exchange (ETDEWEB)

    Hartmann Siantar, C.L.; Chandler, W.P.; Rathkopf, J.A.; Resler, D.A.; Cox, L.J.; Chadwick, M.B.; White, R.M.

    1994-05-12

    By combining a new, all-particle Monte Carlo radiation transport code, PEREGRINE, with the Lawrence Livermore National Laboratory (LLNL) nuclear data base, we have studied the importance of various neutron reactions on dose distributions in biological materials. Monte Carlo calculations have been performed for 5--20 MeV neutron pencil beams incident on biologically relevant materials arranged in several simple geometries. Results highlight the importance of nuclear data used for calculating dose distributions resulting from fast neutron therapy.

  4. Overview of Spallation Neutron Source Physics

    Science.gov (United States)

    Russell, G. J.; Pitcher, E. J.; Muhrer, G.; Mezei, F.; Ferguson, P. D.

    source. We have also calculated the neutron intensity exiting a neutron guide at 10 m from the Lujan Center partially-coupled liquid H2, moderator. This data can be used to estimate the feasibility of performing Fundamental Physics with Pulsed Neutron Beams at a spallation neutron source such as the Lujan Center at Los Alamos.

  5. Monte Carlo Simulation of the Time-Of-Flight Technique for the Measurement of Neutron Cross-section in the Pohang Neutron Facility

    Energy Technology Data Exchange (ETDEWEB)

    An, So Hyun; Lee, Young Ouk; Lee, Cheol Woo [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Lee, Young Seok [National Fusion Research Institute, Daejeon (Korea, Republic of)

    2007-10-15

    It is essential that neutron cross sections are measured precisely for many areas of research and technique. In Korea, these experiments have been performed in the Pohang Neutron Facility (PNF) with the pulsed neutron facility based on the 100 MeV electron linear accelerator. In PNF, the neutron energy spectra have been measured for different water levels inside the moderator and compared with the results of the MCNPX calculation. The optimum size of the water moderator has been determined on the base of these results. In this study, Monte Carlo simulations for the TOF technique were performed and neutron spectra of neutrons were calculated to predict the measurements.

  6. Low dimensional neutron moderators for enhanced source brightness

    DEFF Research Database (Denmark)

    Mezei, Ferenc; Zanini, Luca; Takibayev, Alan

    2014-01-01

    In a recent numerical optimization study we have found that liquid para-hydrogen coupled cold neutron moderators deliver 3–5 times higher cold neutron brightness at a spallation neutron source if they take the form of a flat, quasi 2-dimensional disc, in contrast to the conventional more voluminous...... for cold neutrons. This model leads to the conclusions that the optimal shape for high brightness para-hydrogen neutron moderators is the quasi 1-dimensional tube and these low dimensional moderators can also deliver much enhanced cold neutron brightness in fission reactor neutron sources, compared...... to the much more voluminous liquid D2 or H2 moderators currently used. Neutronic simulation calculations confirm both of these theoretical conclusions....

  7. Ionizing radiation transport in air

    Energy Technology Data Exchange (ETDEWEB)

    Klimanov, V.A.; Konovalov, S.A.; Kochanov, V.A.; Ksenofontov, A.I.; Kukhtevich, V.I.; Mashkovich, V.P.; Meshcherin, B.N.; Panchenko, A.M.; Sukhoruchkin, A.K.; Trubnikov, A.I.

    1979-01-01

    The book considers the problems of gamma-ray and neutron transport in air and near the air-ground and air-vacuum interface. New modifications of the Monte-Carlo method for solving these problems are presented. Differential and integral characteristics of gamma-ray, neutron, and secondary gamma-ray transport are given and the time characteristics of radiation sources are analyzed.

  8. Spin distribution in neutron induced preequilibrium reactions

    Energy Technology Data Exchange (ETDEWEB)

    Dashdorj, D; Kawano, T; Chadwick, M; Devlin, M; Fotiades, N; Nelson, R O; Mitchell, G E; Garrett, P E; Agvaanluvsan, U; Becker, J A; Bernstein, L A; Macri, R; Younes, W

    2005-10-04

    The preequilibrium reaction mechanism makes an important contribution to neutron-induced reactions above E{sub n} {approx} 10 MeV. The preequilibrium process has been studied exclusively via the characteristic high energy neutrons produced at bombarding energies greater than 10 MeV. They are expanding the study of the preequilibrium reaction mechanism through {gamma}-ray spectroscopy. Cross-section measurements were made of prompt {gamma}-ray production as a function of incident neutron energy (E{sub n} = 1 to 250 MeV) on a {sup 48}Ti sample. Energetic neutrons were delivered by the Los Alamos National Laboratory spallation neutron source located at the Los Alamos Neutron Science Center facility. The prompt-reaction {gamma} rays were detected with the large-scale Compton-suppressed Germanium Array for Neutron Induced Excitations (GEANIE). Neutron energies were determined by the time-of-flight technique. The {gamma}-ray excitation functions were converted to partial {gamma}-ray cross sections taking into account the dead-time correction, target thickness, detector efficiency and neutron flux (monitored with an in-line fission chamber). Residual state population was predicted using the GNASH reaction code, enhanced for preequilibrium. The preequilibrium reaction spin distribution was calculated using the quantum mechanical theory of Feshback, Kerman, and Koonin (FKK). The multistep direct part of the FKK theory was calculated for a one-step process. The FKK preequilibrium spin distribution was incorporated into the GNASH calculations and the {gamma}-ray production cross sections were calculated and compared with experimental data. The difference in the partial {gamma}-ray cross sections using spin distributions with and without preequilibrium effects is significant.

  9. Optimization of the geometry and composition of a neutron system for treatment by Boron Neutron Capture Therapy

    Directory of Open Access Journals (Sweden)

    Rohollah Gheisari

    2015-01-01

    Full Text Available Background: In the field of the treatment by Boron Neutron Capture Therapy (BNCT, an optimized neutron system was proposed. This study (simulation was conducted to optimize the geometry and composition of neutron system and increase the epithermal neutron flux for the treatment of deep tumors is performed. Materials and Methods: A neutron system for BNCT was proposed. The system included 252Cf neutron source, neutron moderator/reflector arrangement, filter and concrete. To capture fast neutrons, different neutron filters Fe, Pb, Ni and PbF2 with various thicknesses were simulated and studied. Li (with 1 mm thick was used for filtering of thermal neutrons. Bi with thickness of 1 cm was used to minimize the intensity of gamma rays. Monte Carlo simulation code MCNPX 2.4.0 was used for design of the neutron system and calculation of the neutron components at the output port of the system. Results: For different thicknesses of the filters, the fast neutron flux, the epithermal and thermal flux were calculated at the output port of the system. The spatial distribution of the fast neutron flux, the epithermal flux and gamma flux in human head phantom with the presence of 40 ppm of 10B were obtained. The present calculations showed that Pb filter (about 1 cm at the output port is suitable for fast neutron capture. The thickness of Li filter was determined due to its high absorption cross-section in thermal region. Bi was used as a gamma filter by the reason of it is good for shielding gamma rays, while having high transmission epithermal neutrons. Conclusion: The epithermal neutron flux has enhanced about 38 percent at the output port of the present system, compared with recent system proposed by Ghassoun et al. At 2 cm depth inside the head phantom, the neutron flux reaches a maximum value about . At this depth, the ratio of the thermal neutron flux to the epithermal flux is about three times, that suggests such a neutron system to treat tumors in the

  10. Superfluid neutron stars

    OpenAIRE

    Langlois, David

    2001-01-01

    Neutron stars are believed to contain (neutron and proton) superfluids. I will give a summary of a macroscopic description of the interior of neutron stars, in a formulation which is general relativistic. I will also present recent results on the oscillations of neutron stars, with superfluidity explicitly taken into account, which leads in particular to the existence of a new class of modes.

  11. Neutron scattering. Lectures

    Energy Technology Data Exchange (ETDEWEB)

    Brueckel, Thomas; Heger, Gernot; Richter, Dieter; Roth, Georg; Zorn, Reiner (eds.)

    2010-07-01

    The following topics are dealt with: Neutron sources, symmetry of crystals, diffraction, nanostructures investigated by small-angle neutron scattering, the structure of macromolecules, spin dependent and magnetic scattering, structural analysis, neutron reflectometry, magnetic nanostructures, inelastic scattering, strongly correlated electrons, dynamics of macromolecules, applications of neutron scattering. (HSI)

  12. Calculation of neutron activation of control rods of a nuclear reactor, using MCNP5; Calculo de activacion neutronica de barras de control de un reactor nuclear, utilizando MCNP5

    Energy Technology Data Exchange (ETDEWEB)

    Pena V, J.D.

    2016-07-01

    The control rods of a nuclear reactor are activated by neutron irradiation. The generated activity produces a dose around the rod which is irrelevant inside the reactor, but significant when the rod is withdrawn and placed in a storage pool, because this dose is a potential risk to the surrounding personnel. On the other hand, most of the activation occurs in the stainless steel components of the rod. The Monte Carlo model can reliably determine the activation produced in a stainless steel part exposed to a neutron flux in a reactor and the dose measurement around this part. This thesis presents the Monte Carlo models developed for the activation of the control rods of the TRIGA Mark III reactor of Instituto Nacional de Investigaciones Nucleares (ININ) when only standard fuel was available. Therefore, the validations of the Monte Carlo models are reliable. (Author)

  13. From nuclear structure to neutron stars

    Directory of Open Access Journals (Sweden)

    Gandolfi Stefano

    2014-03-01

    Full Text Available Recent progress in quantum Monte Carlo with modern nucleon-nucleon interactions have enabled the successful description of properties of light nuclei and neutronrich matter. As a demonstration, we show that the agreement between theoretical calculations of the charge form factor of 12C and the experimental data is excellent. Applying similar methods to isospin-asymmetric systems allows one to describe neutrons confined in an external potential and homogeneous neutron-rich matter. Of particular interest is the nuclear symmetry energy, the energy cost of creating an isospin asymmetry. Combining these advances with recent observations of neutron star masses and radii gives insight into the equation of state of neutron-rich matter near and above the saturation density. In particular, neutron star radius measurements constrain the derivative of the symmetry energy.

  14. Neutron dosimetry at the reactor facility VENUS

    Energy Technology Data Exchange (ETDEWEB)

    Deboodt, P.; Vermeersch, F.; Vanhavere, F.; Minsart, G. [SCK-CEN, Belgian Nuclear Research Centre, Mol (Belgium)

    1997-09-01

    The reactor VENUS is a zero-power research reactor mainly devoted to studies on light water fuels. The need for undertaking a neutron spectrometric and dosimetric study became apparent when locally high neutron dose rates were measured. The spectrometric study is based on two approaches. The first is an experimental one in which the neutron spectrum was measured at three positions around the facility. The second is a theoretical one in which a numerical modelling of the neutron transport at the reactor site was performed in order to determine neutron spectra and fluence rates at different positions around the site. The measured and calculated spectra are interpreted in terms of the responses of different individual and environmental dosemeters. These responses are confronted with the in situ measurements. The impact of the ICRP 60 recommendations on the determined dose rates is also studied. (author).

  15. Nuclear reactor neutron shielding

    Science.gov (United States)

    Speaker, Daniel P; Neeley, Gary W; Inman, James B

    2017-09-12

    A nuclear reactor includes a reactor pressure vessel and a nuclear reactor core comprising fissile material disposed in a lower portion of the reactor pressure vessel. The lower portion of the reactor pressure vessel is disposed in a reactor cavity. An annular neutron stop is located at an elevation above the uppermost elevation of the nuclear reactor core. The annular neutron stop comprises neutron absorbing material filling an annular gap between the reactor pressure vessel and the wall of the reactor cavity. The annular neutron stop may comprise an outer neutron stop ring attached to the wall of the reactor cavity, and an inner neutron stop ring attached to the reactor pressure vessel. An excore instrument guide tube penetrates through the annular neutron stop, and a neutron plug comprising neutron absorbing material is disposed in the tube at the penetration through the neutron stop.

  16. NEUTRONIC REACTORS

    Science.gov (United States)

    Wigner, E.P.; Young, G.J.

    1958-10-14

    A method is presented for loading and unloading rod type fuel elements of a neutronic reactor of the heterogeneous, solld moderator, liquid cooled type. In the embodiment illustrated, the fuel rods are disposed in vertical coolant channels in the reactor core. The fuel rods are loaded and unloaded through the upper openings of the channels which are immersed in the coolant liquid, such as water. Unloading is accomplished by means of a coffer dam assembly having an outer sleeve which is placed in sealing relation around the upper opening. A radiation shield sleeve is disposed in and reciprocable through the coffer dam sleeve. A fuel rod engaging member operates through the axial bore in the radiation shield sleeve to withdraw the fuel rod from its position in the reactor coolant channel into the shield, the shield snd rod then being removed. Loading is accomplished in the reverse procedure.

  17. Utilization of low voltage D-T neutron generators in neutron physics studies

    Energy Technology Data Exchange (ETDEWEB)

    Singkarat, S.

    1995-08-01

    In a small nuclear laboratory of a developing country a low voltage D-T neutron generator can be a very useful scientific apparatus. Such machines have been used successfully for more than 40 years in teaching and scientific research. The original continuous mode 150-kV D-T neutron generator has been modified to have also a capability of producing 2-ns pulsed neutrons. Together with a carefully designed 10 m long flight path collimator and shielding of a 25 cm diameter {center_dot} 10 cm thick BC-501 neutron detector, the pulsing system was successfully used for measuring the double differential cross-section (DDX) of natural iron for 14.1-MeV neutron from the angle of 30 deg to 150 deg in 10 deg steps. In order to extend the utility of the generator, two methods for converting the almost monoenergetic 14-MeV neutrons to monoenergetic neutrons of lower energy were proposed and tested. The first method uses a pulsed neutron generator and the second method uses an ordinary continuous mode generator. The latter method was successfully used to measure the scintillation light output of a 1.4 cm diameter spherical NE-213 scintillation detector. The neutron generator has also been used in the continuous search for improved neutron detection techniques. There is a proposal, based on Monte Carlo calculations, of using a scintillation fiber for a fast neutron spectrometer. Due to the slender shape of the fiber, the pattern of produced light gives a peak in the pulse height spectrum instead of the well-known rectangular-like distribution, when the fiber is bombarded end-on by a beam of 14-MeV neutrons. Experimental investigations were undertaken. Detailed investigations on the light transportation property of a short fiber were performed. The predicted peak has not yet been found but the fiber detector may be developed as a directional discrimination fast neutron detector. 18 refs.

  18. Peeling Off Neutron Skins from Neutron-Rich Nuclei: Constraints on the Symmetry Energy from Neutron-Removal Cross Sections

    Science.gov (United States)

    Aumann, T.; Bertulani, C. A.; Schindler, F.; Typel, S.

    2017-12-01

    An experimentally constrained equation of state of neutron-rich matter is fundamental for the physics of nuclei and the astrophysics of neutron stars, mergers, core-collapse supernova explosions, and the synthesis of heavy elements. To this end, we investigate the potential of constraining the density dependence of the symmetry energy close to saturation density through measurements of neutron-removal cross sections in high-energy nuclear collisions of 0.4 to 1 GeV /nucleon . We show that the sensitivity of the total neutron-removal cross section is high enough so that the required accuracy can be reached experimentally with the recent developments of new detection techniques. We quantify two crucial points to minimize the model dependence of the approach and to reach the required accuracy: the contribution to the cross section from inelastic scattering has to be measured separately in order to allow a direct comparison of experimental cross sections to theoretical cross sections based on density functional theory and eikonal theory. The accuracy of the reaction model should be investigated and quantified by the energy and target dependence of various nucleon-removal cross sections. Our calculations explore the dependence of neutron-removal cross sections on the neutron skin of medium-heavy neutron-rich nuclei, and we demonstrate that the slope parameter L of the symmetry energy could be constrained down to ±10 MeV by such a measurement, with a 2% accuracy of the measured and calculated cross sections.

  19. A Dosimetry Study of Deuterium-Deuterium Neutron Generator-based In Vivo Neutron Activation Analysis.

    Science.gov (United States)

    Sowers, Daniel; Liu, Yingzi; Mostafaei, Farshad; Blake, Scott; Nie, Linda H

    2015-12-01

    A neutron irradiation cavity for in vivo neutron activation analysis (IVNAA) to detect manganese, aluminum, and other potentially toxic elements in human hand bone has been designed and its dosimetric specifications measured. The neutron source is a customized deuterium-deuterium neutron generator that produces neutrons at 2.45 MeV by the fusion reaction 2H(d, n)3He at a calculated flux of 7 × 10(8) ± 30% s(-1). A moderator/reflector/shielding [5 cm high density polyethylene (HDPE), 5.3 cm graphite and 5.7 cm borated (HDPE)] assembly has been designed and built to maximize the thermal neutron flux inside the hand irradiation cavity and to reduce the extremity dose and effective dose to the human subject. Lead sheets are used to attenuate bremsstrahlung x rays and activation gammas. A Monte Carlo simulation (MCNP6) was used to model the system and calculate extremity dose. The extremity dose was measured with neutron and photon sensitive film badges and Fuji electronic pocket dosimeters (EPD). The neutron ambient dose outside the shielding was measured by Fuji NSN3, and the photon dose was measured by a Bicron MicroREM scintillator. Neutron extremity dose was calculated to be 32.3 mSv using MCNP6 simulations given a 10-min IVNAA measurement of manganese. Measurements by EPD and film badge indicate hand dose to be 31.7 ± 0.8 mSv for neutrons and 4.2 ± 0.2 mSv for photons for 10 min; whole body effective dose was calculated conservatively to be 0.052 mSv. Experimental values closely match values obtained from MCNP6 simulations. These are acceptable doses to apply the technology for a manganese toxicity study in a human population.

  20. Bicubic spline function approximation of the solution of the fast neutron transport equation

    CERN Document Server

    Chamayou, J M F

    1975-01-01

    The numerical method of approximation of the fast neutron stationary transport equation by means of bicubic cardinal splines is investigated in order to calculate the neutron flux in the one- dimensional plane geometry. A numerical example is given. (5 refs).

  1. Weapons Neutron Research Facility (WNR)

    Data.gov (United States)

    Federal Laboratory Consortium — The Weapons Neutron Research Facility (WNR) provides neutron and proton beams for basic, applied, and defense-related research. Neutron beams with energies ranging...

  2. Nested neutron microfocusing optics on SNAP

    Science.gov (United States)

    Ice, G. E.; Choi, J.-Y.; Takacs, P. Z.; Khounsary, A.; Puzyrev, Y.; Molaison, J. J.; Tulk, C. A.; Andersen, K. H.; Bigault, T.

    2010-06-01

    The high source intensity of the Spallation Neutron Source (SNS), together with efficient detectors and large detector solid angles, now makes possible neutron experiments with much smaller sample volumes than previously were practical. Nested Kirkpatrick-Baez supermirror optics provide a promising and efficient way to further decrease the useable neutron sample size by focusing polychromatic neutrons into microbeams. Because the optics are nondispersive, they are ideal for spallation sources and for polychromatic and wide bandpass experiments on reactor sources. Theoretical calculations indicate that nested mirrors can preserve source brilliance at the sample for small beams and for modest divergences that are appropriate for diffraction experiments. Although the flux intercepted by a sample can be similar with standard beam-guided approaches, the signal-to-background is much improved with small beams on small samples. Here we describe the design, calibration and performance of a nested neutron mirror pair for the Spallation Neutrons At Pressure (SNAP) beamline at the SNS. High-pressure neutron diffraction is but one example of a large class of neutron experiments that will benefit from spatially-resolved microdiffraction.

  3. Neutron intensity of fast reactor spent fuel

    Energy Technology Data Exchange (ETDEWEB)

    Takamatsu, Misao; Aoyama, Takafumi [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center

    1998-03-01

    Neutron intensity of spent fuel of the JOYO Mk-II core with a burnup of 62,500 MWd/t and cooling time of 5.2 years was measured at the spent fuel storage pond. The measured data were compared with the calculated values based on the JOYO core management code system `MAGI`, and the average C/E approximately 1.2 was obtained. It was found that the axial neutron intensity didn`t simply follow the burnup distribution, and the neutron intensity was locally increased at the bottom end of the fuel region due to an accumulation of {sup 244}Cm. (author)

  4. Neutron scattering instrumentation for biology at spallation neutron sources

    Energy Technology Data Exchange (ETDEWEB)

    Pynn, R. [Los Alamos National Laboratory, NM (United States)

    1994-12-31

    Conventional wisdom holds that since biological entities are large, they must be studied with cold neutrons, a domain in which reactor sources of neutrons are often supposed to be pre-eminent. In fact, the current generation of pulsed spallation neutron sources, such as LANSCE at Los Alamos and ISIS in the United Kingdom, has demonstrated a capability for small angle scattering (SANS) - a typical cold- neutron application - that was not anticipated five years ago. Although no one has yet built a Laue diffractometer at a pulsed spallation source, calculations show that such an instrument would provide an exceptional capability for protein crystallography at one of the existing high-power spoliation sources. Even more exciting is the prospect of installing such spectrometers either at a next-generation, short-pulse spallation source or at a long-pulse spallation source. A recent Los Alamos study has shown that a one-megawatt, short-pulse source, which is an order of magnitude more powerful than LANSCE, could be built with today`s technology. In Europe, a preconceptual design study for a five-megawatt source is under way. Although such short-pulse sources are likely to be the wave of the future, they may not be necessary for some applications - such as Laue diffraction - which can be performed very well at a long-pulse spoliation source. Recently, it has been argued by Mezei that a facility that combines a short-pulse spallation source similar to LANSCE, with a one-megawatt, long-pulse spallation source would provide a cost-effective solution to the global shortage of neutrons for research. The basis for this assertion as well as the performance of some existing neutron spectrometers at short-pulse sources will be examined in this presentation.

  5. Prompt Neutron Lifetime for the NBSR Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hanson, A.L.; Diamond, D.

    2012-06-24

    In preparation for the proposed conversion of the National Institute of Standards and Technology (NIST) research reactor (NBSR) from high-enriched uranium (HEU) to low-enriched uranium (LEU) fuel, certain point kinetics parameters must be calculated. We report here values of the prompt neutron lifetime that have been calculated using three independent methods. All three sets of calculations demonstrate that the prompt neutron lifetime is shorter for the LEU fuel when compared to the HEU fuel and longer for the equilibrium end-of-cycle (EOC) condition when compared to the equilibrium startup (SU) condition for both the HEU and LEU fuels.

  6. SU-E-T-598: Parametric Equation for Quick and Reliable Estimate of Stray Neutron Doses in Proton Therapy and Application for Intracranial Tumor Treatments

    Energy Technology Data Exchange (ETDEWEB)

    Bonfrate, A; Farah, J; Sayah, R; Clairand, I [Institut de Radioprotection et de Surete Nucleaire (IRSN), Fontenay-aux-roses (France); De Marzi, L; Delacroix, S [Institut Curie Centre de Protontherapie d Orsay (CPO), Orsay (France); Herault, J [Centre Antoine Lacassagne (CAL) Cyclotron biomedical, Nice (France); Lee, C [National Cancer Institute, Rockville, MD (United States); Bolch, W [Univ Florida, Gainesville, FL (United States)

    2015-06-15

    Purpose: Development of a parametric equation suitable for a daily use in routine clinic to provide estimates of stray neutron doses in proton therapy. Methods: Monte Carlo (MC) calculations using the UF-NCI 1-year-old phantom were exercised to determine the variation of stray neutron doses as a function of irradiation parameters while performing intracranial treatments. This was done by individually changing the proton beam energy, modulation width, collimator aperture and thickness, compensator thickness and the air gap size while their impact on neutron doses were put into a single equation. The variation of neutron doses with distance from the target volume was also included in it. Then, a first step consisted in establishing the fitting coefficients by using 221 learning data which were neutron absorbed doses obtained with MC simulations while a second step consisted in validating the final equation. Results: The variation of stray neutron doses with irradiation parameters were fitted with linear, polynomial, etc. model while a power-law model was used to fit the variation of stray neutron doses with the distance from the target volume. The parametric equation fitted well MC simulations while establishing fitting coefficients as the discrepancies on the estimate of neutron absorbed doses were within 10%. The discrepancy can reach ∼25% for the bladder, the farthest organ from the target volume. Finally, the validation showed results in compliance with MC calculations since the discrepancies were also within 10% for head-and-neck and thoracic organs while they can reach ∼25%, again for pelvic organs. Conclusion: The parametric equation presents promising results and will be validated for other target sites as well as other facilities to go towards a universal method.

  7. A review of reaction rates and thermodynamic and transport properties for the 11-species air model for chemical and thermal nonequilibrium calculations to 30000 K

    Science.gov (United States)

    Gupta, Roop N.; Yos, Jerrold M.; Thompson, Richard A.

    1989-01-01

    Reaction rate coefficients and thermodynamic and transport properties are provided for the 11-species air model which can be used for analyzing flows in chemical and thermal nonequilibrium. Such flows will likely occur around currently planned and future hypersonic vehicles. Guidelines for determining the state of the surrounding environment are provided. Approximate and more exact formulas are provided for computing the properties of partially ionized air mixtures in such environments.

  8. Lithium neutron producing target for BINP accelerator-based neutron source

    Energy Technology Data Exchange (ETDEWEB)

    Bayanov, B.; Belov, V.; Kindyuk, V.; Oparin, E.; Taskaev, S. E-mail: taskaev@inp.nsk.su

    2004-11-01

    Pilot innovative accelerator-based neutron source for neutron capture therapy is under construction now at the Budker Institute of Nuclear Physics, Novosibirsk, Russia. One of the main elements of the facility is lithium target, that produces neutrons via threshold {sup 7}Li(p,n){sup 7}Be reaction at 25 kW proton beam with energies 1.915 or 2.5 MeV. In the present report, the results of experiments on neutron producing target prototype are presented, the results of calculations of hydraulic resistance for heat carrier flow and lithium layer temperature are shown. Calculation showed that the lithium target could run up to 10 mA proton beam before melting. Choice of target variant is substantiated. Program of immediate necessary experiments is described. Target design for neutron source constructed at BINP is presented. Manufacturing the neutron producing target up to the end of 2004 and obtaining a neutron beam on BINP accelerator-based neutron source are planned during 2005.

  9. Lithium neutron producing target for BINP accelerator-based neutron source.

    Science.gov (United States)

    Bayanov, B; Belov, V; Kindyuk, V; Oparin, E; Taskaev, S

    2004-11-01

    Pilot innovative accelerator-based neutron source for neutron capture therapy is under construction now at the Budker Institute of Nuclear Physics, Novosibirsk, Russia. One of the main elements of the facility is lithium target, that produces neutrons via threshold (7)Li(p,n)(7)Be reaction at 25 kW proton beam with energies 1.915 or 2.5 MeV. In the present report, the results of experiments on neutron producing target prototype are presented, the results of calculations of hydraulic resistance for heat carrier flow and lithium layer temperature are shown. Calculation showed that the lithium target could run up to 10 mA proton beam before melting. Choice of target variant is substantiated. Program of immediate necessary experiments is described. Target design for neutron source constructed at BINP is presented. Manufacturing the neutron producing target up to the end of 2004 and obtaining a neutron beam on BINP accelerator-based neutron source are planned during 2005.

  10. Neutron spectra and dosimetric features of isotopic neutron sources: a review

    Energy Technology Data Exchange (ETDEWEB)

    Vega C, H. R. [Universidad Autonoma de Zacatecas, Unidad Academica de Estudios Nucleares, Cipres No. 10, Fracc. La Penuela, 98060 Zacatecas, Zac. (Mexico); Martinez O, S. A., E-mail: fermineutron@yahoo.com [Universidad Pedagogica y Tecnologica de Colombia, Grupo de Fisica Nuclear Aplicada y Simulacion, Av. Central del Norte 39-115, 150003 Tunja, Boyaca (Colombia)

    2015-10-15

    A convenient way to produce neutrons is the isotopic neutron source, where the production is through (α, n), (γ, n), and spontaneous fission reactions. Isotopic neutron sources are small, easy to handle, and have a relative low cost. On the other hand the neutron yield is small and mostly of them produces neutrons with a wide energy distribution. In this work, a review is carried out about the the main features of {sup 24}NaBe, {sup 24}NaD{sub 2}O, {sup 116}InBe, {sup 140}LaBe, {sup 238}PuLi, {sup 239}PuBe, {sup 241}AmB, {sup 241}AmBe, {sup 241}AmF, {sup 241}AmLi, {sup 242}CmBe, {sup 210}PoBe, {sup 226}RaBe, {sup 252}Cf and {sup 252}Cf/D{sub 2}O isotopic neutron source. Also, using Monte Carlo methods, the neutron spectra in 31 energy groups, the neutron mean energy; the Ambient dose equivalent, the Personal dose equivalent and the Effective dose were calculated for these isotopic neutron sources. (Author)

  11. Neutron reactions in accreting neutron stars: a new pathway to efficient crust heating.

    Science.gov (United States)

    Gupta, Sanjib S; Kawano, Toshihiko; Möller, Peter

    2008-12-05

    In our calculation of neutron star crust heating we include several key new model features. In earlier work electron capture (EC) only allowed neutron emission from the daughter ground state; here we calculate, in a deformed quasi-random-phase approximation (QRPA) model, EC decay rates to all states in the daughter that are allowed by Gamow-Teller selection rules and energetics. The subsequent branching ratios between the 1n,...,xn channels and the competing gamma decay are calculated in a Hauser-Feshbach model. In our multicomponent plasma model a single (EC, xn) reaction step can produce several neutron-deficient nuclei, each of which can further decay by (EC, xn). Hence, the neutron emission occurs more continuously with increasing depth as compared to that in a one-component plasma model.

  12. Average neutron detection efficiency for DEMON detectors

    Science.gov (United States)

    Zhang, S.; Lin, W.; Rodrigues, M. R. D.; Huang, M.; Wada, R.; Liu, X.; Zhao, M.; Jin, Z.; Chen, Z.; Keutgen, T.; Kowalski, S.; Hagel, K.; Barbui, M.; Bonasera, A.; Bottosso, C.; Materna, T.; Natowitz, J. B.; Qin, L.; Sahu, P. K.; Schmidt, K. J.; Wang, J.

    2013-05-01

    The neutron detection efficiency of a DEMON detector, averaged over the whole volume, was calculated using GEANT and applied to determine neutron multiplicities in an intermediate heavy ion reaction. When a neutron source is set at a distance of about 1 m from the front surface of the detector, the average efficiency, ɛav, is found to be significantly lower (20-30%) than the efficiency measured at the center of the detector, ɛ0. In the GEANT simulation the ratio R=ɛav/ɛ0 was calculated as a function of neutron energy. The experimental central efficiency multiplied by R was then used to determine the average efficiency. The results were applied to a study of the 64Zn+112Sn reaction at 40 A MeV which employed 16 DEMON detectors. The neutron multiplicity was extracted using a moving source fit. The derived multiplicities are compared well with those determined using the neutron ball in the NIMROD detector array in a separate experiment. Both are in good agreement with multiplicities predicted by a transport model calculation using an antisymmetric molecular dynamics (AMD) model code.

  13. Dark neutron stars

    Science.gov (United States)

    Jones, P. B.

    2017-06-01

    There is good evidence that electron-positron pair formation is not present in that section of the pulsar open magnetosphere, which is the source of coherent radio emission, but the possibility of two-photon pair creation in an outer gap remains. Calculation of transition rates for this process based on measured whole-surface temperatures, combined with a survey of γ-ray, X-ray and optical luminosities, expressed per primary beam lepton, shows that few Fermi-LAT pulsars have significant outer-gap pair creation. For radio-loud pulsars with positive polar-cap corotational charge density and an ion-proton plasma, there must be an outward flow of electrons from some other part of the magnetosphere to maintain a constant net charge on the star. In the absence of pair creation, it is likely that this current is the source of GeV γ-emission observed by the Fermi-LAT and its origin is in the region of the outer gap. With negative polar-cap corotational charge density, the compensating current in the absence of pair creation can consist only of ions or protons. These neutron stars are likely to be radio-quiet, have no observable γ-emission, and hence can be described as dark neutron stars.

  14. Layered semiconductor neutron detectors

    Science.gov (United States)

    Mao, Samuel S; Perry, Dale L

    2013-12-10

    Room temperature operating solid state hand held neutron detectors integrate one or more relatively thin layers of a high neutron interaction cross-section element or materials with semiconductor detectors. The high neutron interaction cross-section element (e.g., Gd, B or Li) or materials comprising at least one high neutron interaction cross-section element can be in the form of unstructured layers or micro- or nano-structured arrays. Such architecture provides high efficiency neutron detector devices by capturing substantially more carriers produced from high energy .alpha.-particles or .gamma.-photons generated by neutron interaction.

  15. Neutron streak camera

    Science.gov (United States)

    Wang, Ching L.

    1983-09-13

    Apparatus for improved sensitivity and time resolution of a neutron measurement. The detector is provided with an electrode assembly having a neutron sensitive cathode which emits relatively low energy secondary electrons. The neutron sensitive cathode has a large surface area which provides increased sensitivity by intercepting a greater number of neutrons. The cathode is also curved to compensate for differences in transit time of the neutrons emanating from the point source. The slower speeds of the secondary electrons emitted from a certain portion of the cathode are matched to the transit times of the neutrons impinging thereupon.

  16. Neutron in biology

    Energy Technology Data Exchange (ETDEWEB)

    Niimura, Nobuo [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1997-11-01

    Neutron in biology can provide an experimental method of directly locating relationship of proteins and DNA. However, there are relatively few experimental study of such objects since it takes a lot of time to collect a sufficient number of Bragg reflections and inelastic spectra due to the low flux of neutron illuminating the sample. Since a next generation neutron source of JAERI will be 5MW spallation neutron source and its effective neutron flux will be 10{sup 2} to 10{sup 3} times higher than the one of JRR-3M, neutron in biology will open a completely new world for structural biology. (author)

  17. Transmission Bragg edge spectroscopy measurements at ORNL Spallation Neutron Source

    Science.gov (United States)

    Tremsin, A. S.; McPhate, J. B.; Vallerga, J. V.; Siegmund, O. H. W.; Feller, W. B.; Bilheux, H. Z.; Molaison, J. J.; Tulk, C. A.; Crow, L.; Cooper, R. G.; Penumadu, D.

    2010-11-01

    Results of neutron transmission Bragg edge spectroscopic experiments performed at the SNAP beamline of the Spallation Neutron Source are presented. A high resolution neutron counting detector with a neutron sensitive microchannel plate and Timepix ASIC readout is capable of energy resolved two dimensional mapping of neutron transmission with spatial accuracy of ~55 μm, limited by the readout pixel size, and energy resolution limited by the duration of the initial neutron pulse. A two dimensional map of the Fe 110 Bragg edge position was obtained for a bent steel screw sample. Although the neutron pulse duration corresponded to ~30 mÅ energy resolution for 15.3 m flight path, the accuracy of the Bragg edge position in our measurements was improved by analytical fitting to a few mÅ level. A two dimensional strain map was calculated from measured Bragg edge values with an accuracy of ~few hundreds μistrain for 300s of data acquisition time.

  18. Neutronic Reactor Design to Reduce Neutron Loss

    Science.gov (United States)

    Miles, F. T.

    1961-05-01

    A nuclear reactor construction is described in which an unmoderated layer of the fissionable material is inserted between the moderated portion of the reactor core and the core container steel wall. The wall is surrounded by successive layers of pure fertile material and moderator containing fertile material. The unmoderated layer of the fissionable material will insure that a greater portion of fast neutrons will pass through the steel wall than would thermal neutrons. Since the steel has a smaller capture cross section for the fast neutrons, greater nunnbers of neutrons will pass into the blanket, thereby increasing the over-all efficiency of the reactor. (AEC)

  19. Basics of Neutrons for First Responders

    Energy Technology Data Exchange (ETDEWEB)

    Rees, Brian G. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2018-02-05

    These are slides from a presentation on the basics of neutrons. A few topics covered are: common origins of terrestrial neutron radiation, neutron sources, neutron energy, interactions, detecting neutrons, gammas from neutron interactions, neutron signatures in gamma-ray spectra, neutrons and NaI, neutron fluence to dose (msV), instruments' response to neutrons.

  20. Calculation of liquid-cooled finplate tube heat exchangers for the dehumidification of air with an enhanced water load; Berechnung fluessigkeitsgekuehlter Lamellenrohr-Waermeuebertrager zur Entfeuchtung von Luft mit hoher Wasserbeladung

    Energy Technology Data Exchange (ETDEWEB)

    Sievers, Martin [Stuttgart Univ. (Germany). Inst. fuer Thermodynamik der Luft- und Raumfahrt; Sievers, Uwe [Hochschule fuer Angewandte Wissenschaften Hamburg (Germany). Inst. fuer Energiesysteme und Brennstoffzellentechnik

    2011-07-01

    Demoisturizing of air or combustion gases is a process step in many air conditioners, energy engineering plants and chemical plants, e.g. high-efficiency boilers, fuel cell systems, and sea water desalination systems. By cooling a gas-steam mixture down to below condensation temperature, the steam will be partly condensed, and the condensate will then be separated from the gaseous gas-steam mixture. Sea water desalination plants ave higher temperatures and higher air moisture at the demoisturizer inlet, so some of the common assumptions for calculation of combined heat and mass transfer are not applicable. The publication presents equations for calculating the demoisturizing of air with high water content, as in sea water desalination systems or fuel cell systems, for liquid-cooled finned-tube heat exchangers. The coupled heat and mass transfer is calculated without the common simplifying assumptions, and the calculation method based on the cell method is validated on the basis of measurements. [German] Die Entfeuchtung von Luft oder Verbrennungsgasen ist ein Prozessschritt, der in verschiedenen Anlagen der Klimatechnik, der Energietechnik und der Verfahrenstechnik auftritt. Beispiele sind Klimaanlagen in Gebaeuden und Fahrzeugen, Brennwertkessel, Brennstoffzellenanlagen und Meerwasserentsalzungsanlagen. In diesen wird durch Abkuehlung eines Gas-Dampf-Gemisches unter seine Taupunkttemperatur der Dampf teilweise kondensiert und anschliessend die Kondensatphase von dem gasfoermig verbleibenden Gas-Dampf-Gemisch abgetrennt. In Meerwasserentsalzungsanlagen zur Gewinnung von Trink- oder Brauchwasser, die nach dem Luftbefeuchtungs-Entfeuchtungs-(HDH)-Prinzip arbeiten, liegen die Temperatur und die Wasserbeladung der feuchten Luft am Eintritt in den Entfeuchter deutlich hoeher als in herkoemmlichen Klimaanlagen. Deshalb treffen hier einige uebliche vereinfachende Annahmen zur Berechnung des gekoppelten Waerme- und Stoffuebergangs nicht zu. Fuer fluessigkeitsgekuehlte

  1. Neutron field inside a PET Cyclotron vault room

    Energy Technology Data Exchange (ETDEWEB)

    Vega C, H.R. [UAZ, C. Cipres 10, 98068 Zacatecas (Mexico); Mendez, R. [CIEMAT, Ave. Complutense 22, 28040 Madrid (Spain); Iniguez, M.P. [Universidad de Valladolid, Po Prado de la Magdalena s/n, 47011 Valladolid (Spain); Climent, J.M.; Penuelas, I. [Servicio de Medicina Nuclear de la Clinica Universitaria de Navarra, Pamplona (Spain); Barquero, R. [Hospital Universitario Rio Hortega, Valladolid (Spain)]. e-mail: fermineutron@yahoo.com

    2006-07-01

    The neutron field around a Positron Emission Tomography cyclotron was investigated during {sup 18} F radioisotope production with an 18 MeV proton beam. In this study the Ion Beam Application cyclotron, model Cyclone 18/9, was utilized. Measurements were carried out with a Bonner sphere neutron spectrometer with pairs of thermoluminescent dosemeters (TLD600 and TLD700) as thermal neutron detector. The TLDs readouts were utilized to unfold the neutron spectra at three different positions inside the cyclotron's vault room. With the spectra the Ambient dose equivalent was calculated. Neutron spectra unfolding were performed with the BUNKIUT code and the UTA4 response matrix. Neutron spectra were also determined by Monte Carlo calculations using a detailed model of cyclotron and vault room. (Author)

  2. Microtron MT 25 as a source of neutrons

    Energy Technology Data Exchange (ETDEWEB)

    Kralik, M.; Solc, J. [Czech Metrology Institute, CZ-102 00 Prague 10 (Czech Republic); Chvatil, D.; Krist, P.; Turek, K. [Nuclear Physics Institute, p.r.i., AS CR, CZ-250 68 Rez (Czech Republic); Granja, C. [Institute of Experimental and Applied Physics, Horska 3a/22, CZ-128 00 Prague 2 (Czech Republic)

    2012-08-15

    The objective was to describe Microtron MT25 as a source of neutrons generated by bremsstrahlung induced photonuclear reactions in U and Pb targets. Bremsstrahlung photons were produced by electrons accelerated at energy 21.6 MeV. Spectral fluence of the generated neutrons was calculated with MCNPX code and then experimentally determined at two positions by means of a Bonner spheres spectrometer in which the detector of thermal neutrons was replaced by activation Mn tablets or track detectors CR-39 with a {sup 10}B radiator. The measured neutron spectral fluence and the calculated anisotropy served for the estimation of neutron yield from the targets and for the determination of ambient dose equivalent rate at the place of measurement. Microtron MT25 is intended as one of the sources for testing neutron sensitive devices which will be sent into the space.

  3. Determine Neutron Yield From Beryllium Compounds Bombarding With Alpha Particles

    Directory of Open Access Journals (Sweden)

    Osamah Nawfal Oudah

    2015-04-01

    Full Text Available Abstract The chemical combination of Beryllium used in the amp945n reaction as a target materials. wherefore Beryllium has generally been the material of select when manufacturing a neutron source. Beryllium compounds has a significant impact on the scheme and manufacture of the 238Cm-9Be neutron source. In present work the Beryllium chemical combinations were considered as a producer of neutrons with different percentage of mass and the neutron yields were likewise determined using the outcomes of the ASTAR calculations per unit incident charge. The neutron yields of Beryllium Oxide Beryllium Carbide Beryllium Hydride Beryllium Hydroxide and Beryllium Nitride were calculated as 3.1006e-005 3.0365e-005 2.2866e-005 2.0464e-005 and 3.0386e-005respectively. The results show that the Beryllium Oxide is a suitable material to use in the 238Cm-9Be neutron source.

  4. Study of the seismic behaviour of the fast reactor cores; Etude du comportement sismique des coeurs de reacteurs a neutrons rapides

    Energy Technology Data Exchange (ETDEWEB)

    Cerqueira, E

    1998-12-31

    This work studies the seismic behaviour of fast neutrons reactor cores. It consists in analyzing the tests made on the models Rapsodie and Symphony by using the calculation code Castem 2000. Te difficulty is in the description of connections of the system and the effects of the fluid (calculation in water). The results for the programme Rapsodie are near the experimental results. For the programme Symphony, the calculations in air have allowed to represent the behaviour of fuel assemblies in a satisfying way. It is still to analyze the tests Symphony in water. (N.C.)

  5. Studies on application of radiation and radioisotopes -Studies on application of neutron activation analysis-

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Yong Sam; Jung, Yung Joo; Jung, Eui Sik; Lee, Sang Mee [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of); Kim, Nak Bae [Korea Institute of Geology, Mining and Materials, Taejon (Korea, Republic of)

    1995-07-01

    To apply Neutron activation analysis to routine analysis of environmental samples utilizing the research reactor (TRIGA MK-III), improving effects of analytical sensitivity have been investigated using both of thermal and epithermal neutron irradiating technique. Identification and development of analytical procedure was carried out using three kinds of standard reference materials (urban particulate matter, coal fly ash, soil). In addition, the confidence of this method was established by participation in collaborative research for the training and apply of international credit of analytical procedure. Practical studies on air dust samples have also been carried out regionally and seasonally. For the investigation on emission source of special element, enrichment factor was calculated in urban and rural area. Besides, a suitable process of biological sample (pine needle) analyses has been established by carrying out identification of uncertainty using standard reference material. The concentration of elements in practical samples were also determined regionally and seasonally. 14 figs, 26 tabs, 67 refs. (Author).

  6. The energy spectrum of delayed neutrons from thermal neutron induced fission of sup 2 sup 3 sup 5 U and its analytical approximation

    CERN Document Server

    Doroshenko, A Y; Tarasko, M Z

    2001-01-01

    The energy spectrum of the delayed neutrons is the poorest known of all input data required in the calculation of the effective delayed neutron fractions. In addition to delayed neutron spectra based on the aggregate spectrum measurements there are two different approaches for deriving the delayed neutron energy spectra. Both of them are based on the data related to the delayed neutron spectra from individual precursors of delayed neutrons. In present work these two different data sets were compared with the help of an approximation by gamma-function. The choice of this approximation function instead of the Maxwellian or evaporation type of distribution is substantiated.

  7. Superfluid Density of Neutrons in the Inner Crust of Neutron Stars: New Life for Pulsar Glitch Models.

    Science.gov (United States)

    Watanabe, Gentaro; Pethick, C J

    2017-08-11

    Calculations of the effects of band structure on the neutron superfluid density in the crust of neutron stars made under the assumption that the effects of pairing are small [N. Chamel, Phys. Rev. C 85, 035801 (2012)PRVCAN0556-2813] lead to moments of inertia of superfluid neutrons so small that the crust alone is insufficient to account for the magnitude of neutron star glitches. Inspired by earlier work on ultracold atomic gases in an optical lattice, we investigate fermions with attractive interactions in a periodic lattice in the mean-field approximation. The effects of band structure are suppressed when the pairing gap is of order or greater than the strength of the lattice potential. By applying the results to the inner crust of neutron stars, we conclude that the reduction of the neutron superfluid density is considerably less than previously estimated and, consequently, it is premature to rule out models of glitches based on neutron superfluidity in the crust.

  8. GEANT4 used for neutron beam design of a neutron imaging facility at TRIGA reactor in Morocco

    Science.gov (United States)

    Ouardi, A.; Machmach, A.; Alami, R.; Bensitel, A.; Hommada, A.

    2011-09-01

    Neutron imaging has a broad scope of applications and has played a pivotal role in visualizing and quantifying hydrogenous masses in metallic matrices. The field continues to expand into new applications with the installation of new neutron imaging facilities. In this scope, a neutron imaging facility for computed tomography and real-time neutron radiography is currently being developed around 2.0MW TRIGA MARK-II reactor at Maamora Nuclear Research Center in Morocco (Reuscher et al., 1990 [1]; de Menezes et al., 2003 [2]; Deinert et al., 2005 [3]). The neutron imaging facility consists of neutron collimator, real-time neutron imaging system and imaging process systems. In order to reduce the gamma-ray content in the neutron beam, the tangential channel was selected. For power of 250 kW, the corresponding thermal neutron flux measured at the inlet of the tangential channel is around 3×10 11 ncm 2/s. This facility will be based on a conical neutron collimator with two circular diaphragms with diameters of 4 and 2 cm corresponding to L/D-ratio of 165 and 325, respectively. These diaphragms' sizes allow reaching a compromise between good flux and efficient L/D-ratio. Convergent-divergent collimator geometry has been adopted. The beam line consists of a gamma filter, fast neutrons filter, neutron moderator, neutron and gamma shutters, biological shielding around the collimator and several stages of neutron collimator. Monte Carlo calculations by a fully 3D numerical code GEANT4 were used to design the neutron beam line ( http://www.info.cern.ch/asd/geant4/geant4.html[4]). To enhance the neutron thermal beam in terms of quality, several materials, mainly bismuth (Bi) and sapphire (Al 2O 3) were examined as gamma and neutron filters respectively. The GEANT4 simulations showed that the gamma and epithermal and fast neutron could be filtered using the bismuth (Bi) and sapphire (Al 2O 3) filters, respectively. To get a good cadmium ratio, GEANT 4 simulations were used to

  9. Quantifying moisture transport in cementitious materials using neutron radiography

    Science.gov (United States)

    Lucero, Catherine L.

    . It has been found through this study that small pores, namely voids created by chemical shrinkage, gel pores, and capillary pores, ranging from 0.5 nm to 50 microm, fill quickly through capillary action. However, large entrapped and entrained air voids ranging from 0.05 to 1.25 mm remain empty during the initial filling process. In mortar exposed to calcium chloride solution, a decrease in sorptivity was observed due to an increase in viscosity and surface tension of the solution as proposed by Spragg et al 2011. This work however also noted a decrease in the rate of absorption due to a reaction between the salt and matrix which results in the filling of the pores in the concrete. The results from neutron imaging can help in the interpretation of standard absorption tests. ASTM C1585 test results can be further analyzed in several ways that could give an accurate indication of the durability of the concrete. Results can be reported in depth of penetration versus the square root of time rather than mm3 of fluid per mm2 of exposed surface area. Since a known fraction of pores are initially filling before reaching the edge of the sample, the actual depth of penetration can be calculated. This work is compared with an 'intrinsic sorptivity' that can be used to interpret mass measurements. Furthermore, the influence of shrinkage reducing admixtures (SRAs) on drying was studied. Neutron radiographs showed that systems saturated in water remain "wetter" than systems saturated in 5% SRA solution. The SRA in the system reduces the moisture diffusion coefficient due an increase in viscosity and decrease in surface tension. Neutron radiography provided spatial information of the drying front that cannot be achieved using other methods.

  10. Standard Guide for Benchmark Testing of Light Water Reactor Calculations

    CERN Document Server

    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 ...

  11. Neutron anatomy

    Energy Technology Data Exchange (ETDEWEB)

    Bacon, G.E. [Univ. of Sheffield (United Kingdom)

    1994-12-31

    The familiar extremes of crystalline material are single-crystals and random powders. In between these two extremes are polycrystalline aggregates, not randomly arranged but possessing some preferred orientation and this is the form taken by constructional materials, be they steel girders or the bones of a human or animal skeleton. The details of the preferred orientation determine the ability of the material to withstand stress in any direction. In the case of bone the crucial factor is the orientation of the c-axes of the mineral content - the crystals of the hexagonal hydroxyapatite - and this can readily be determined by neutron diffraction. In particular it can be measured over the volume of a piece of bone, utilizing distances ranging from 1mm to 10mm. The major practical problem is to avoid the intense incoherent scattering from the hydrogen in the accompanying collagen; this can best be achieved by heat-treatment and it is demonstrated that this does not affect the underlying apatite. These studies of bone give leading anatomical information on the life and activities of humans and animals - including, for example, the life history of the human femur, the locomotion of sheep, the fracture of the legs of racehorses and the life-styles of Neolithic tribes. We conclude that the material is placed economically in the bone to withstand the expected stresses of life and the environment. The experimental results are presented in terms of the magnitude of the 0002 apatite reflection. It so happens that for a random powder the 0002, 1121 reflections, which are neighboring lines in the powder pattern, are approximately equal in intensity. The latter reflection, being of manifold multiplicity, is scarcely affected by preferred orientation so that the numerical value of the 0002/1121 ratio serves quite accurately as a quantitative measure of the degree of orientation of the c-axes in any chosen direction for a sample of bone.

  12. Neutrons from Antiproton Irradiation

    DEFF Research Database (Denmark)

    Bassler, Niels; Holzscheiter, Michael; Petersen, Jørgen B.B.

    the neutron spectrum. Additionally, we used a cylindrical polystyrene loaded with several pairs of thermoluminescent detectors containing Lithium-6 and Lithium-7, which effectively detects thermalized neutrons. The obtained results are compared with FLUKA imulations. Results: The results obtained...

  13. Energy saving screen materials : measurement method of radiation exchange, air permeability and humidity transport and a calculation method for energy saving

    NARCIS (Netherlands)

    Hemming, Silke; Baeza Romero, Esteban; Mohammadkhani, Vida; Breugel, van Bram

    2017-01-01

    The objectives of the project were the quantifi cation of the greenhouse screen properties (emissivity and transmissivity for thermal infrared radiation, air permeability and humidity transport) and the determination of the total energy saving under defi ned conditions in order to be able to compare

  14. Neutron dosimetry based on nuclear track etched detectors

    Energy Technology Data Exchange (ETDEWEB)

    Bouassoule, T.; Fernandez, F.; Marin, M.; Tomas, M. [Grup de Fisica de les Radiacions. Departament de Fisica, Universitat Autonoma de Barcelona, E-08193 Bellaterra (Spain)

    1999-07-01

    In this work, the response of a neutron dosimeter based on plastic track detectors has been studied. The detector geometry used consists on a C R-39 detector 500 m thick plus either a Makrofol converter 300 {mu} m thick or air used as converter, for the study of the response to fast or thermal neutrons respectively. The possibility of using Makrofol as a high energy neutron dosemeter has also been studied. In order to validate the results obtained from Monte Carlo simulations, a set of irradiations to monoenergetic neutron beams has been performed at the Ptb and to realistic fields at Cadarache neutron irradiation facilities. An excellent agreement has been found between the simulated and the experimental values. The lower detection limit value found for C R-39 and fast neutrons was 60 {mu} Sv. (Author)

  15. Development of a mid-head radiation dose response function. [Phantom determinations of neutron and. gamma. absorbed doses in mid-brain for military applications

    Energy Technology Data Exchange (ETDEWEB)

    Trubey, D. K.; Knight, J. R.; Bartine, D. E.; Pace, III, J. V.

    1979-02-01

    Calculations have been made of the incident neutron and gamma-ray absorbed dose response as a function of energy in the mid-head position of a phantom model. The calculations were performed with the DOT discrete ordinates transport code in the adjoint mode using co-axial cylinders to represent the head and torso. Results, given in a coupled 37-neutron-group, 21-gamma-ray-group structure (37/21) and a 22-neutron-group, 18-gamma-ray-group structure (22/18), are compared with previously obtained results. The mid-head response is less than the conventional radiation protection fluence-to-dose factors which are based on maximum phantom values. In the case of a fission source in air the neutron dose is about a factor of 4 less, and the secondary gamma-ray dose is about a factor of 1.5 less. For a fusion source the neutron dose ratio varies from about 1.9 at close range to about 3. The gamma-ray dose ratio is about the same as for the fission source. Tables of the various response functions are presented in the Appendix A.

  16. UCN Source at an External Beam of Thermal Neutrons

    Directory of Open Access Journals (Sweden)

    E. V. Lychagin

    2015-01-01

    Full Text Available We propose a new method for production of ultracold neutrons (UCNs in superfluid helium. The principal idea consists in installing a helium UCN source into an external beam of thermal or cold neutrons and in surrounding this source with a solid methane moderator/reflector cooled down to ~4 K. The moderator plays the role of an external source of cold neutrons needed to produce UCNs. The flux of accumulated neutrons could exceed the flux of incident neutrons due to their numerous reflections from methane; also the source size could be significantly larger than the incident beam diameter. We provide preliminary calculations of cooling of neutrons. These calculations show that such a source being installed at an intense source of thermal or cold neutrons like the ILL or PIK reactor or the ESS spallation source could provide the UCN density 105 cm−3, the production rate 107 UCN/s−1. Main advantages of such an UCN source include its low radiative and thermal load, relatively low cost, and convenient accessibility for any maintenance. We have carried out an experiment on cooling of thermal neutrons in a methane cavity. The data confirm the results of our calculations of the spectrum and flux of neutrons in the methane cavity.

  17. International Neutron Radiography Newsletter

    DEFF Research Database (Denmark)

    Domanus, Joseph Czeslaw

    1986-01-01

    At the First World Conference on Neutron Radiography i t was decided to continue the "Neutron Radiography Newsletter", published previously by J.P. Barton, as the "International Neutron Radiography Newsletter" (INRNL), with J.C. Doraanus as editor. The British Journal of Non-Destructive Testing...

  18. Transport coefficients in superfluid neutron stars

    Energy Technology Data Exchange (ETDEWEB)

    Tolos, Laura [Instituto de Ciencias del Espacio (IEEC/CSIC) Campus Universitat Autònoma de Barcelona, Facultat de Ciències, Torre C5, E-08193 Bellaterra (Barcelona) (Spain); Frankfurt Institute for Advances Studies. Johann Wolfgang Goethe University, Ruth-Moufang-Str. 1, 60438 Frankfurt am Main (Germany); Manuel, Cristina [Instituto de Ciencias del Espacio (IEEC/CSIC) Campus Universitat Autònoma de Barcelona, Facultat de Ciències, Torre C5, E-08193 Bellaterra (Barcelona) (Spain); Sarkar, Sreemoyee [Tata Institute of Fundamental Research, Homi Bhaba Road, Mumbai-400005 (India); Tarrus, Jaume [Physik Department, Technische Universität München, D-85748 Garching (Germany)

    2016-01-22

    We study the shear and bulk viscosity coefficients as well as the thermal conductivity as arising from the collisions among phonons in superfluid neutron stars. We use effective field theory techniques to extract the allowed phonon collisional processes, written as a function of the equation of state and the gap of the system. The shear viscosity due to phonon scattering is compared to calculations of that coming from electron collisions. We also comment on the possible consequences for r-mode damping in superfluid neutron stars. Moreover, we find that phonon collisions give the leading contribution to the bulk viscosities in the core of the neutron stars. We finally obtain a temperature-independent thermal conductivity from phonon collisions and compare it with the electron-muon thermal conductivity in superfluid neutron stars.

  19. Neutrino Processes in Neutron Stars

    Directory of Open Access Journals (Sweden)

    Kolomeitsev E.E.

    2010-10-01

    Full Text Available The aim of these lectures is to introduce basic processes responsible for cooling of neutron stars and to show how to calculate the neutrino production rate in dense strongly interacting nuclear medium. The formalism is presented that treats on equal footing one-nucleon and multiple-nucleon processes and reactions with virtual bosonic modes and condensates. We demonstrate that neutrino emission from dense hadronic component in neutron stars is subject of strong modifications due to collective effects in the nuclear matter. With the most important in-medium processes incorporated in the cooling code an overall agreement with available soft X ray data can be easily achieved. With these findings the so-called “standard” and “non-standard” cooling scenarios are replaced by one general “nuclear medium cooling scenario” which relates slow and rapid neutron star coolings to the star masses (interior densities. The lectures are split in four parts. Part I: After short introduction to the neutron star cooling problem we show how to calculate neutrino reaction rates of the most efficient one-nucleon and two-nucleon processes. No medium effects are taken into account in this instance. The effects of a possible nucleon pairing are discussed. We demonstrate that the data on neutron star cooling cannot be described without inclusion of medium effects. It motivates an assumption that masses of the neutron stars are different and that neutrino reaction rates should be strongly density dependent. Part II: We introduce the Green’s function diagram technique for systems in and out of equilibrium and the optical theorem formalism. The latter allows to perform calculations of production rates with full Green’s functions including all off-mass-shell effects. We demonstrate how this formalism works within the quasiparticle approximation. Part III: The basic concepts of the nuclear Fermi liquid approach are introduced. We show how strong

  20. Fundamental Problems of Neutron Physics at the Spallation Neutron Source at the ORNL

    Energy Technology Data Exchange (ETDEWEB)

    Vladimir Gudkov

    2008-07-16

    We propose to provide theoretical support for the experimental program in fundamental neutron physics at the SNS. This includes the study of neutron properties, neutron beta-decay, parity violation effects and time reversal violation effects. The main purpose of the proposed research is to work on theoretical problems related to experiments which have a high priority at the SNS. Therefore, we will make a complete analysis of beta-decay process including calculations of radiative corrections and recoil corrections for angular correlations for polarized neutron decay, with an accuracy better that is supposed to be achieved in the planning experiments. Based on the results of the calculations, we will provide analysis of sensitivity of angular correlations to be able to search for the possible extensions of the Standard model. Also we will help to plan other experiments to address significant problems of modern physics and will work on their theoretical support.

  1. Neutron flux assessment of a neutron irradiation facility based on inertial electrostatic confinement fusion.

    Science.gov (United States)

    Sztejnberg Gonçalves-Carralves, M L; Miller, M E

    2015-12-01

    Neutron generators based on inertial electrostatic confinement fusion were considered for the design of a neutron irradiation facility for explanted organ Boron Neutron Capture Therapy (BNCT) that could be installed in a health care center as well as in research areas. The chosen facility configuration is "irradiation chamber", a ~20×20×40 cm(3) cavity near or in the center of the facility geometry where samples to be irradiated can be placed. Neutron flux calculations were performed to study different manners for improving scattering processes and, consequently, optimize neutron flux in the irradiation position. Flux distributions were assessed through numerical simulations of several models implemented in MCNP5 particle transport code. Simulation results provided a wide spectrum of combinations of net fluxes and energy spectrum distributions. Among them one can find a group that can provide thermal neutron fluxes per unit of production rate in a range from 4.1·10(-4) cm(-2) to 1.6·10(-3) cm(-2) with epithermal-to-thermal ratios between 0.3% and 13% and fast-to-thermal ratios between 0.01% to 8%. Neutron generators could be built to provide more than 10(10) n s(-1) and, consequently, with an arrangement of several generators appropriate enough neutron fluxes could be obtained that would be useful for several BNCT-related irradiations and, eventually, for clinical practice. Copyright © 2015 Elsevier Ltd. All rights reserved.

  2. A review of reaction rates and thermodynamic and transport properties for an 11-species air model for chemical and thermal nonequilibrium calculations to 30000 K

    Science.gov (United States)

    Gupta, Roop N.; Yos, Jerrold M.; Thompson, Richard A.; Lee, Kam-Pui

    1990-01-01

    Reaction rate coefficients and thermodynamic and transport properties are reviewed and supplemented for the 11-species air model which can be used for analyzing flows in chemical and thermal nonequilibrium up to temperatures of 3000 K. Such flows will likely occur around currently planned and future hypersonic vehicles. Guidelines for determining the state of the surrounding environment are provided. Curve fits are given for the various species properties for their efficient computation in flowfield codes. Approximate and more exact formulas are provided for computing the properties of partially ionized air mixtures in a high energy environment. Limitations of the approximate mixing laws are discussed for a mixture of ionized species. An electron number-density correction for the transport properties of the charged species is obtained. This correction has been generally ignored in the literature.

  3. Neutronic analysis of a high power density hybrid reactor using ...

    Indian Academy of Sciences (India)

    Natural lithium is also utilized for comparison to these two innovative coolants. Neutron transport calculations are performed on a simple experimental hybrid blanket with cylindrical geometry with the help of the SCALE 4·3 System by solving the Boltzmann transport equation with the XSDRNPM code in 238 neutron groups ...

  4. Evaluation of response matrix of a multisphere neutron spectrometer ...

    Indian Academy of Sciences (India)

    For calculating neutron energy over 20 MeV, the library HIGH was used. This library is an. NJOY generated neutron data library containing cross-sections for some required elements evaluated with the GNASH code up to 100 MeV [8]. 2. ..... An improvement of the energy resolution in the epithermal and intermediate region.

  5. Background neutron in the endcap and barrel regions of resistive ...

    Indian Academy of Sciences (India)

    The detector response calculations taken as a function of the neutron energy in the range of 0.01 eV–1 GeV have been simulated through RPC set-up. In order to evaluate the response of detector in the LHC background environment, the neutron spectrum expected in the CMS muon endcap and barrel region were taken ...

  6. Does mass accretion lead to field decay in neutron stars?

    Science.gov (United States)

    Shibazaki, N.; Murakami, T.; Shaham, J.; Nomoto, K.

    1989-01-01

    Adopting the hypothesis of accretion-induced magnetic field decay in neutron stars, the consequent evolution of a neutron star's spin and magnetic field are calculated. The results are consistent with observations of binary and millisecond radio pulsars. Thermomagnetic effects could provide a possible physical mechanism for such accretion-induced field decay.

  7. An empirical fit to estimated neutron emission cross sections from ...

    Indian Academy of Sciences (India)

    Neutron emission cross section for various elements from 9Be to 209Bi have been calculated using the hybrid model code ALICE-91 for proton induced reactions in the energy range 25 MeV to 105 MeV. An empirical expression relating neutron emission cross section to target mass number and incident proton energy has ...

  8. Grazing Incidence Neutron Optics

    Science.gov (United States)

    Gubarev, Mikhail V. (Inventor); Ramsey, Brian D. (Inventor); Engelhaupt, Darell E. (Inventor)

    2013-01-01

    Neutron optics based on the two-reflection geometries are capable of controlling beams of long wavelength neutrons with low angular divergence. The preferred mirror fabrication technique is a replication process with electroform nickel replication process being preferable. In the preliminary demonstration test an electroform nickel optics gave the neutron current density gain at the focal spot of the mirror at least 8 for neutron wavelengths in the range from 6 to 20.ANG.. The replication techniques can be also be used to fabricate neutron beam controlling guides.

  9. Semiconductor neutron detector

    Science.gov (United States)

    Ianakiev, Kiril D [Los Alamos, NM; Littlewood, Peter B [Cambridge, GB; Blagoev, Krastan B [Arlington, VA; Swinhoe, Martyn T [Los Alamos, NM; Smith, James L [Los Alamos, NM; Sullivan, Clair J [Los Alamos, NM; Alexandrov, Boian S [Los Alamos, NM; Lashley, Jason Charles [Santa Fe, NM

    2011-03-08

    A neutron detector has a compound of lithium in a single crystal form as a neutron sensor element. The lithium compound, containing improved charge transport properties, is either lithium niobate or lithium tantalate. The sensor element is in direct contact with a monitor that detects an electric current. A signal proportional to the electric current is produced and is calibrated to indicate the neutrons sensed. The neutron detector is particularly useful for detecting neutrons in a radiation environment. Such radiation environment may, e.g. include gamma radiation and noise.

  10. Neutron scattering. Experiment manuals

    Energy Technology Data Exchange (ETDEWEB)

    Brueckel, Thomas; Heger, Gernot; Richter, Dieter; Roth, Georg; Zorn, Reiner (eds.)

    2010-07-01

    The following topics are dealt with: The thermal triple axis spectrometer PUMA, the high-resolution powder diffractometer SPODI, the hot single-crystal diffractometer HEiDi for structure analysis with neutrons, the backscattering spectrometer SPHERES, neutron polarization analysis with tht time-of-flight spectrometer DNS, the neutron spin-echo spectrometer J-NSE, small-angle neutron scattering with the KWS-1 and KWS-2 diffractometers, the very-small-angle neutron scattering diffractrometer with focusing mirror KWS-3, the resonance spin-echo spectrometer RESEDA, the reflectometer TREFF, the time-of-flight spectrometer TOFTOF. (HSI)

  11. Neutron scatter camera

    Science.gov (United States)

    Mascarenhas, Nicholas; Marleau, Peter; Brennan, James S.; Krenz, Kevin D.

    2010-06-22

    An instrument that will directly image the fast fission neutrons from a special nuclear material source has been described. This instrument can improve the signal to background compared to non imaging neutron detection techniques by a factor given by ratio of the angular resolution window to 4.pi.. In addition to being a neutron imager, this instrument will also be an excellent neutron spectrometer, and will be able to differentiate between different types of neutron sources (e.g. fission, alpha-n, cosmic ray, and D-D or D-T fusion). Moreover, the instrument is able to pinpoint the source location.

  12. Broad Energy Range Neutron Spectroscopy using a Liquid Scintillator and a Proportional Counter: Application to a Neutron Spectrum Similar to that from an Improvised Nuclear Device.

    Science.gov (United States)

    Xu, Yanping; Randers-Pehrson, Gerhard; Marino, Stephen A; Garty, Guy; Harken, Andrew; Brenner, David J

    2015-09-11

    A novel neutron irradiation facility at the Radiological Research Accelerator Facility (RARAF) has been developed to mimic the neutron radiation from an Improvised Nuclear Device (IND) at relevant distances (e.g. 1.5 km) from the epicenter. The neutron spectrum of this IND-like neutron irradiator was designed according to estimations of the Hiroshima neutron spectrum at 1.5 km. It is significantly different from a standard reactor fission spectrum, because the spectrum changes as the neutrons are transported through air, and it is dominated by neutron energies from 100 keV up to 9 MeV. To verify such wide energy range neutron spectrum, detailed here is the development of a combined spectroscopy system. Both a liquid scintillator detector and a gas proportional counter were used for the recoil spectra measurements, with the individual response functions estimated from a series of Monte Carlo simulations. These normalized individual response functions were formed into a single response matrix for the unfolding process. Several accelerator-based quasi-monoenergetic neutron source spectra were measured and unfolded to test this spectroscopy system. These reference neutrons were produced from two reactions: T(p,n)3He and D(d,n)3He, generating neutron energies in the range between 0.2 and 8 MeV. The unfolded quasi-monoenergetic neutron spectra indicated that the detection system can provide good neutron spectroscopy results in this energy range. A broad-energy neutron spectrum from the 9Be(d,n) reaction using a 5 MeV deuteron beam, measured at 60 degrees to the incident beam was measured and unfolded with the evaluated response matrix. The unfolded broad neutron spectrum is comparable with published time-of-flight results. Finally, the pair of detectors were used to measure the neutron spectrum generated at the RARAF IND-like neutron facility and a comparison is made to the neutron spectrum of Hiroshima.

  13. DRC2: A code with specialized applications for coupling localized Monte Carlo adjoint calculations with fluences from two-dimensional R-Z discrete ordinates air-over-ground calculations

    Energy Technology Data Exchange (ETDEWEB)

    Slater, C.O.

    1992-01-01

    The DRC2 code, which couples MASH or MASHX adjoint leakages with DORT 2-D discrete ordinates forward directional fluences, is described. The forward fluences are allowed to vary both axially and radially over the coupling surface, as opposed to the strictly axial variation allowed by the predecessor DRC code. Input instructions are presented along with descriptions and results from several sample problems. Results from the sample problems are used to compare DRC2 with DRC, DRC2 with DORT, and DRC2 with itself for the case of x-y dependence versus no x-y dependence of the forward fluence. The test problems demonstrate that for small systems DRC and DRC2 give essentially the same results. Some significant differences are noted for larger systems. Additionally, DRC2 results with no x-y dependence of the forward directional fluences are practically the same as those calculated by DRC.

  14. Benchmark experiment on vanadium assembly with D-T neutrons. Leakage neutron spectrum measurement

    Energy Technology Data Exchange (ETDEWEB)

    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)

  15. Single-Neutron Structure of Neutron-Rich Nuclei near N=50 and N=82

    Energy Technology Data Exchange (ETDEWEB)

    Cizewski, J. A. [Rutgers University; Jones, K. L. [University of Tennessee, Knoxville (UTK); Kozub, R. L. [Tennessee Technological University; Pain, S. D. [Rutgers University; Bardayan, Daniel W [ORNL; Blackmon, Jeff C [ORNL; Adekola, Aderemi S [ORNL; Chae, K. Y. [University of Tennessee, Knoxville (UTK); Chipps, K. [Colorado School of Mines, Golden; Erikson, Luke [Colorado School of Mines, Golden; Gaddis, A. L. [Furman University; Harlin, Christopher W [ORNL; Hatarik, Robert [Rutgers University; Howard, Joshua A [ORNL; Kaplan, Ron [ORNL; Krolas, W. [University of Warsaw; Liang, J Felix [ORNL; Livesay, Jake [ORNL; Ma, Zhanwen [ORNL; Matei, Catalin [Oak Ridge Associated Universities (ORAU); Moazen, Brian [University of Tennessee, Knoxville (UTK); Nesaraja, Caroline D [ORNL; O' Malley, Patrick [Rutgers University; Patterson, N. P. [University of Surrey, UK; Paulauskas, Stanley [University of Tennessee, Knoxville (UTK); Shapira, Dan [ORNL; Shriner, Jr., John F [ORNL; Sissom, D. J. [Tennessee Technological University; Smith, Michael Scott [ORNL; Swan, T. P. [University of Surrey, UK; Thomas, J. S. [Rutgers University; Wilson, Gemma L [ORNL

    2009-01-01

    The 82Ge, 84Se, 132Sn, 130Sn, and 134Te (d,p) reactions have been measured with {approx}4-5-MeV-A rare isotope beams and CD2 targets at the HRIBF at ORNL. Energies and spectroscopic strengths have been measured for excitations in 83Ge and 85Se. Direct neutron capture calculations on 82Ge are presented. Preliminary results for single-neutron excitations in 131Sn, 133Sn, and 135Te are reported.

  16. Neutron structural biology

    Energy Technology Data Exchange (ETDEWEB)

    Niimura, Nobuo [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1998-03-01

    Neutron diffraction provides an experimental method of directly locating hydrogen atoms in protein which play important roles in physiological functions. However, there are relatively few examples of neutron crystallography in biology since it takes a lot of time to collect a sufficient number of Bragg reflections due to the low flux of neutrons illuminating the sample. In order to overcome the flux problem, we have successfully developed the neutron IP, where the neutron converter, {sup 6}Li or Gd, was mixed with a photostimulated luminescence material on flexible plastic support. Neutron Laue diffraction 2A data from tetragonal lysozyme were collected for 10 days with neutron imaging plates, and 960 hydrogen atoms in the molecule and 157 bound water molecules were identified. These results explain the proposed hydrolysis mechanism of the sugar by the lysozyme molecule and that lysozyme is less active at pH7.0. (author)

  17. The neutron-deuteron elastic scattering angular distribution at 95 MeV

    Energy Technology Data Exchange (ETDEWEB)

    Mermod, Philippe

    2004-04-01

    The neutron-deuteron elastic scattering differential cross section has been measured at 95 MeV incident neutron energy, with the Medley setup at TSL in Uppsala. The neutron-proton differential cross section has also been measured for normalization purposes. The data are compared with theoretical calculations to investigate the role of three-nucleon force effects.

  18. Declination Calculator

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — Declination is calculated using the current International Geomagnetic Reference Field (IGRF) model. Declination is calculated using the current World Magnetic Model...

  19. Artificial neural networks in neutron dosimetry

    Energy Technology Data Exchange (ETDEWEB)

    Vega C, H.R.; Hernandez D, V.M.; Manzanares A, E.; Mercado, G.A.; Perales M, W.A.; Robles R, J.A. [Unidades Academicas de Estudios Nucleares, UAZ, A.P. 336, 98000 Zacatecas (Mexico); Gallego, E.; Lorente, A. [Depto. de Ingenieria Nuclear, Universidad Politecnica de Madrid, (Spain)

    2005-07-01

    An artificial neural network has been designed to obtain the neutron doses using only the Bonner spheres spectrometer's count rates. Ambient, personal and effective neutron doses were included. 187 neutron spectra were utilized to calculate the Bonner count rates and the neutron doses. The spectra were transformed from lethargy to energy distribution and were re-binned to 31 energy groups using the MCNP 4C code. Re-binned spectra, UTA4 response matrix and fluence-to-dose coefficients were used to calculate the count rates in Bonner spheres spectrometer and the doses. Count rates were used as input and the respective doses were used as output during neural network training. Training and testing was carried out in Mat lab environment. The artificial neural network performance was evaluated using the {chi}{sup 2}- test, where the original and calculated doses were compared. The use of Artificial Neural Networks in neutron dosimetry is an alternative procedure that overcomes the drawbacks associated in this ill-conditioned problem. (Author)

  20. Neutronic analysis for bolometers in ITER

    Energy Technology Data Exchange (ETDEWEB)

    Suarez, A., E-mail: alejandro.suarez@iter.org [CIEMAT, Avda. Complutense 40, 28040 Madrid (Spain); Reichle, R.; Loughlin, M.; Polunovskiy, E.; Walsh, M. [ITER Organization, Route de Vinon sur Verdon, 13115, St. Paul lez Durance (France)

    2013-10-15

    Highlights: ► Radiation damage calculations for the bolometers in ITER. ► Redesign of the bolometric diagnostic in EPP01. ► New bolometer radiation damage values in EPP01 in the safe zone. -- Abstract: Neutronic considerations in ITER have such importance that they drive the design of many diagnostics and components of the machine, and bolometers are not an exception. Bolometer cameras will be installed on the vacuum vessel, viewing the plasma through the gaps between blanket modules, divertor, equatorial and upper port plugs. The ITER reference bolometer sensors are of a resistive type. For this study it is assumed that they are composed of a thin silicon nitride carrier film and platinum resistors disposed in a Wheatstone bridge configuration. Their assumed radiation hardness is 0.1 dpa. Neutronic calculations were performed with the Monte Carlo program MCNP5, the FENDL 2.1 nuclear data library and the latest B-lite ITER neutronic model with the appropriate modifications using the CAD to MCNP converter MCAM. A complete characterization of the neutron fluxes in all the bolometer locations and the calculation of neutron damage were performed. Values above the failure threshold damage were obtained for some of the bolometers, leading to a complete redesign of some parts of the bolometric system in order to extend its lifetime.

  1. Pressure Vessel Calculations for VVER-440 Reactors

    Science.gov (United States)

    Hordósy, G.; Hegyi, Gy.; Keresztúri, A.; Maráczy, Cs.; Temesvári, E.; Vértes, P.; Zsolnay, É.

    2003-06-01

    Monte Carlo calculations were performed for a selected cycle of the Paks NPP Unit II to test a computational model. In the model the source term was calculated by the core design code KARATE and the neutron transport calculations were performed by the MCNP. Different forms of the source specification were examined. The calculated results were compared with measurements and in most cases fairly good agreement was found.

  2. Simulation of resistive plate chamber sensitivity to neutrons

    CERN Document Server

    Altieri, S; Bruno, G; Merlo, M; Ratti, S P; Riccardi, C; Torre, P; Vitulo, P; Abbrescia, M; Colaleo, A; Iaselli, Giuseppe; Loddo, F; Maggi, M; Marangelli, B; Natali, S; Nuzzo, S; Pugliese, G; Ranieri, A; Romano, F

    2001-01-01

    The Resistive Plate Chambers (RPCs) sensitivity to neutrons has been simulated using GEANT code with MICAP and FLUKA interfaces. The calculations have been performed as a function of the neutrons energy in the range 0.02 eV-1 GeV. To evaluate the response of the detector in the IHC background environment, the neutron energy spectrum expected in the CMS muon barrel has been taken into account; a hit rate due to neutrons of about 0.6 Hz cm/sup -2/ has been estimated for a 250*250 cm/sup 2/ RPC in the RB1 station. (2 refs).

  3. Dispersion and decay of collective modes in neutron star cores

    Science.gov (United States)

    Kobyakov, D. N.; Pethick, C. J.; Reddy, S.; Schwenk, A.

    2017-08-01

    We calculate the frequencies of collective modes of neutrons, protons, and electrons in the outer core of neutron stars. The neutrons and protons are treated in a hydrodynamic approximation and the electrons are regarded as collisionless. The coupling of the nucleons to the electrons leads to Landau damping of the collective modes and to significant dispersion of the low-lying modes. We investigate the sensitivity of the mode frequencies to the strength of entrainment between neutrons and protons, which is not well characterized. The contribution of collective modes to the thermal conductivity is evaluated.

  4. Eurados trial performance test for neutron personal dosimetry

    DEFF Research Database (Denmark)

    Bordy, J.M.; Stadtmann, H.; Ambrosi, P.

    2001-01-01

    This paper reports on the results of a neutron trial performance test sponsored by the European Commission and organised by EURADOS. As anticipated, neutron dosimetry results were very dependent on the dosemeter type and the dose calculation algorithm. Fast neutron fields were generally well...... measured, but particular problems were noted in the determination of intermediate energy fields and large incident angles, demonstrating the difficulties of neutron personal dosimetry. Of particular concern from a radiological protection point of view was the large number of results underestimating...... personal dose equivalent. A considerable over-response was noted in a few cases....

  5. Background neutron spectrum at 2420 m above sea level

    Science.gov (United States)

    Vega-Carrillo, Hector Rene; Manzanares-Acuña, Eduardo

    2004-05-01

    The ambient neutron spectrum was measured in-doors at ground level in Zacatecas Mexico at 2420 m above sea level. A Bonner sphere spectrometer with a 6LiI(Eu) scintillator was used to obtain the neutron spectrum. With the spectrum the ambient dose equivalent was calculated using the ICRP 74 neutron fluence-to-dose conversion factors. The neutron fluence rate was 65±3 cm -2 h -1, producing 13.7±0.6 nSv h -1 due to ambient dose equivalent.

  6. Preparation of Radioactive Gold Nanoparticle by Neutron Activation

    OpenAIRE

    Rohadi Awaludin

    2010-01-01

    It was reported that gold nanoparticle could be used for cancer therapy using thermal effect. It is possible to kill cancer cells using radiation of radioisotope. Study on preparation of radioactive gold by neutron activation at central irradiation position (CIP) of G.A. Siwabessy reactor with neutron flux 1.26 x 1014 neutron s-1cm-2 has been carried out. It was revealed that a radioisotop of gold (198Au) was produced by neutron activation from natural gold. Calculation re...

  7. NEUTRON DENSITY CONTROL IN A NEUTRONIC REACTOR

    Science.gov (United States)

    Young, G.J.

    1959-06-30

    The method and means for controlling the neutron density in a nuclear reactor is described. It describes the method and means for flattening the neutron density distribution curve across the reactor by spacing the absorbing control members to varying depths in the central region closer to the center than to the periphery of the active portion of the reactor to provide a smaller neutron reproduction ratio in the region wherein the members are inserted, than in the remainder of the reactor thereby increasing the over-all potential power output.

  8. Bias in Dynamic Monte Carlo Alpha Calculations

    Energy Technology Data Exchange (ETDEWEB)

    Sweezy, Jeremy Ed [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Nolen, Steven Douglas [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Adams, Terry R. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Trahan, Travis John [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2015-02-06

    A 1/N bias in the estimate of the neutron time-constant (commonly denoted as α) has been seen in dynamic neutronic calculations performed with MCATK. In this paper we show that the bias is most likely caused by taking the logarithm of a stochastic quantity. We also investigate the known bias due to the particle population control method used in MCATK. We conclude that this bias due to the particle population control method is negligible compared to other sources of bias.

  9. Referential calculation of particulate matter in the air as a factor of environmental pollution in the urban area of the city of Pujilí

    Directory of Open Access Journals (Sweden)

    Paola Vallejo Choez

    2016-06-01

    Full Text Available This is a preliminary investigation on the environmental quality of the city of Pujilí, made from the collection of samples of particulate matter and vehicular traffic counts on six points of the city. The methodology is based on the provisions of the Unified Text of Secondary Environmental Legislation for measuring atmospheric particulate matter, and the use of count tables for vehicle registration. The results reflect the impact of vehicular traffic, the characteristics of the rolling road layer, soil erosion, and climate on air pollution and its impact on the health of the population.

  10. Study of EAS neutron component temporal structure

    Science.gov (United States)

    Gromushkin, D. M.; Petrukhin, A. A.; Stenkin, Yu. V.; Yashin, I. I.

    2011-04-01

    The neutron component of Extensive Air Showers (EAS) carries information about the primary cosmic ray flux as well as about parameters of hadronic interactions at ultra-high energies. We present here the data obtained with the "Neutron" array which is a prototype of a novel type EAS array PRISMA (Stenkin, 2009). The prototype consists of 5 large area scintillator detectors (0.75 m2 each) placed in the corners and in the center of 5 m side square. The scintillator consisting of an alloy of ZnS(Ag) and 6LiF is shaped as a thin layer of grains covered with thin transparent plastic film.

  11. Measurement of in-phantom neutron flux and gamma dose in Tehran research reactor boron neutron capture therapy beam line.

    Science.gov (United States)

    Bavarnegin, Elham; Sadremomtaz, Alireza; Khalafi, Hossein; Kasesaz, Yaser

    2016-01-01

    Determination of in-phantom quality factors of Tehran research reactor (TRR) boron neutron capture therapy (BNCT) beam. The doses from thermal neutron reactions with 14N and 10B are calculated by kinetic energy released per unit mass approach, after measuring thermal neutron flux using neutron activation technique. Gamma dose is measured using TLD-700 dosimeter. Different dose components have been measured in a head phantom which has been designed and constructed for BNCT purpose in TRR. Different in-phantom beam quality factors have also been determined. This study demonstrates that the TRR BNCT beam line has potential for treatment of superficial tumors.

  12. Contribution to the algorithmic and efficient programming of new parallel architectures including accelerators for neutron physics and shielding computations; Contribution a l'algorithmique et a la programmation efficace des nouvelles architectures paralleles comportant des accelerateurs de calcul dans le domaine de la neutronique et de la radioprotection

    Energy Technology Data Exchange (ETDEWEB)

    Dubois, J.

    2011-10-13

    In science, simulation is a key process for research or validation. Modern computer technology allows faster numerical experiments, which are cheaper than real models. In the field of neutron simulation, the calculation of eigenvalues is one of the key challenges. The complexity of these problems is such that a lot of computing power may be necessary. The work of this thesis is first the evaluation of new computing hardware such as graphics card or massively multi-core chips, and their application to eigenvalue problems for neutron simulation. Then, in order to address the massive parallelism of supercomputers national, we also study the use of asynchronous hybrid methods for solving eigenvalue problems with this very high level of parallelism. Then we experiment the work of this research on several national supercomputers such as the Titane hybrid machine of the Computing Center, Research and Technology (CCRT), the Curie machine of the Very Large Computing Centre (TGCC), currently being installed, and the Hopper machine at the Lawrence Berkeley National Laboratory (LBNL). We also do our experiments on local workstations to illustrate the interest of this research in an everyday use with local computing resources. (author) [French] Les travaux de cette these concernent dans un premier temps l'evaluation des nouveaux materiels de calculs tels que les cartes graphiques ou les puces massivement multicoeurs, et leur application aux problemes de valeurs propres pour la neutronique. Ensuite, dans le but d'utiliser le parallelisme massif des supercalculateurs, nous etudions egalement l'utilisation de methodes hybrides asynchrones pour resoudre des problemes a valeur propre avec ce tres haut niveau de parallelisme. Nous experimentons ensuite le resultat de ces recherches sur plusieurs supercalculateurs nationaux tels que la machine hybride Titane du Centre de Calcul, Recherche et Technologies (CCRT), la machine Curie du Tres Grand Centre de Calcul (TGCC) qui

  13. Neutron data evaluation of {sup 238}U

    Energy Technology Data Exchange (ETDEWEB)

    Maslov, V.M.; Porodzinskij, Y.V.; Hasegawa, Akira; Shibata, Keiichi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1998-08-01

    Cross sections for neutron-induced reactions on {sup 238}U are calculated by using the Hauser-Feshbach-Moldauer theory, the coupled channel model and the double-humped fission barrier model. The direct excitation of ground state band levels is calculated with a rigid-rotator model. The direct excitation of vibrational octupole and K = 2{sup +} quadrupole bands is included using a soft (deformable) rotator model. The competition of inelastic scattering to fission reaction is shown to be sensitive to the target nucleus level density at excitations above the pairing gap. As for fission, (n,2n), (n,3n), and (n,4n) reactions, secondary neutron spectra data are consistently reproduced. Pre-equilibrium emission of first neutron is included. Shell effects in the level densities are shown to be important for estimation of energy dependence of non-emissive fission cross section. (author). 105 refs.

  14. Neutron interactions with biological tissue. Final report

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-02-17

    This program was aimed at creating a quantitative physical description, at the micrometer and nanometer levels, of the physical interactions of neutrons with tissue through the ejected secondary charged particles. The authors used theoretical calculations whose input includes neutron cross section data; range, stopping power, ion yield, and straggling information; and geometrical properties. Outputs are initial and slowing-down spectra of charged particles, kerma factors, average values of quality factors, microdosimetric spectra, and integral microdosimetric parameters such as {bar y}{sub F}, {bar y}{sub D}, y{sup *}. Since it has become apparent that nanometer site sizes are also relevant to radiobiological effects, the calculations of event size spectra and their parameters were extended to these smaller diameters. This information is basic to radiological physics, radiation biology, radiation protection of workers, and standards for neutron dose measurement.

  15. Detection of fast neutrons with the Medipix-2 pixel detector

    Science.gov (United States)

    Uher, J.; Jakubek, J.; Koster, U.; Lebel, C.; Leroy, C.; Pospisil, S.; Skoda, R.; Vykydal, Z.

    2008-06-01

    Neutron radiography using thermal and fast neutrons is becoming increasingly important for scientific, technical, security and other applications. In the past, the Medipix-2 imaging detector was successfully adapted for thermal neutron imaging by our group by adding a 6LiF neutron converter covering the surface of the active part of the detector. Recently, the Medipix-2 detector was also adapted and used for fast neutron imaging. Fast neutrons are detected through proton recoil from a 1 mm thick polyethylene layer placed on the detector surface. Basic detection and imaging properties of the detector were calculated, simulated and measured. The measurements were done using an AmBe neutron source, as well as external neutron beams of the Sparrow reactor at the Czech Technical University in Prague and of the high-flux reactor at the Institut Laue-Langevin (ILL) in Grenoble. The detector provided a reasonable signal-to-background ratio of about eight at the neutron beam of ILL's Neutrograph neutron radiography and tomography station. The estimated spatial resolution was at a level of 100 μm.

  16. Résolutions rapides et fiables pour les solveurs d'algèbre linéaire numérique en calcul haute performance.

    OpenAIRE

    Baboulin, Marc

    2012-01-01

    In this "Habilitation à Diriger des Recherches" (HDR), we present our research in high-performance scientific computing over the recent years. Our work has been mainly related to parallel algorithms for numerical linear algebra solvers and their parallel implementation in public domain software libraries. We illustrate in this manuscript how these calculations can be accelerated using innovative algorithms and be made reliable using specific quantities in error analysis. First we explain how ...

  17. Status of spallation neutron source

    Energy Technology Data Exchange (ETDEWEB)

    Oyama, Yukio [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1997-03-01

    Existing and planned facilities using proton accelerator driven spallation neutron source are reviewed. These include new project of neutron science proposed from Japan Atomic Energy Research Institute. The present status of facility requirement and accelerator technology leads us to new era of neutron science such as neutron scattering research and nuclear transmutation study using very intense neutron source. (author)

  18. Fast Neutron Dosimeter for the Space Environment Project

    Data.gov (United States)

    National Aeronautics and Space Administration — Model calculations and risk assessment estimates indicate that secondary neutrons, with energies ranging between 0.5 to >150 MeV, make a significant contribution...

  19. Geant4 beam model for boron neutron capture therapy: investigation of neutron dose components.

    Science.gov (United States)

    Moghaddasi, Leyla; Bezak, Eva

    2018-01-23

    Boron neutron capture therapy (BNCT) is a biochemically-targeted type of radiotherapy, selectively delivering localized dose to tumour cells diffused in normal tissue, while minimizing normal tissue toxicity. BNCT is based on thermal neutron capture by stable [Formula: see text]B nuclei resulting in emission of short-ranged alpha particles and recoil [Formula: see text]Li nuclei. The purpose of the current work was to develop and validate a Monte Carlo BNCT beam model and to investigate contribution of individual dose components resulting of neutron interactions. A neutron beam model was developed in Geant4 and validated against published data. The neutron beam spectrum, obtained from literature for a cyclotron-produced beam, was irradiated to a water phantom with boron concentrations of 100 μg/g. The calculated percentage depth dose curves (PDDs) in the phantom were compared with published data to validate the beam model in terms of total and boron depth dose deposition. Subsequently, two sensitivity studies were conducted to quantify the impact of: (1) neutron beam spectrum, and (2) various boron concentrations on the boron dose component. Good agreement was achieved between the calculated and measured neutron beam PDDs (within 1%). The resulting boron depth dose deposition was also in agreement with measured data. The sensitivity study of several boron concentrations showed that the calculated boron dose gradually converged beyond 100 μg/g boron concentration. This results suggest that 100μg/g tumour boron concentration may be optimal and above this value limited increase in boron dose is expected for a given neutron flux.

  20. Neutron spectrometry with artificial neural networks

    Energy Technology Data Exchange (ETDEWEB)

    Vega C, H.R.; Hernandez D, V.M.; Manzanares A, E.; Rodriguez, J.M.; Mercado S, G.A. [Universidad Autonoma de Zacatecas, A.P. 336, 98000 Zacatecas (Mexico); Iniguez de la Torre Bayo, M.P. [Universidad de Valladolid, Valladolid (Spain); Barquero, R. [Hospital Universitario Rio Hortega, Valladolid (Spain); Arteaga A, T. [Envases de Zacatecas, S.A. de C.V., Zacatecas (Mexico)]. e-mail: rvega@cantera.reduaz.mx

    2005-07-01

    An artificial neural network has been designed to obtain the neutron spectra from the Bonner spheres spectrometer's count rates. The neural network was trained using 129 neutron spectra. These include isotopic neutron sources; reference and operational spectra from accelerators and nuclear reactors, spectra from mathematical functions as well as few energy groups and monoenergetic spectra. The spectra were transformed from lethargy to energy distribution and were re-bin ned to 31 energy groups using the MCNP 4C code. Re-binned spectra and UTA4 response matrix were used to calculate the expected count rates in Bonner spheres spectrometer. These count rates were used as input and the respective spectrum was used as output during neural network training. After training the network was tested with the Bonner spheres count rates produced by a set of neutron spectra. This set contains data used during network training as well as data not used. Training and testing was carried out in the Mat lab program. To verify the network unfolding performance the original and unfolded spectra were compared using the {chi}{sup 2}-test and the total fluence ratios. The use of Artificial Neural Networks to unfold neutron spectra in neutron spectrometry is an alternative procedure that overcomes the drawbacks associated in this ill-conditioned problem. (Author)

  1. Neutron spectrometry using artificial neural networks

    Energy Technology Data Exchange (ETDEWEB)

    Vega-Carrillo, Hector Rene [Unidad Academica de Estudios Nucleares, Universidad Autonoma de Zacatecas, Apdo. Postal 336, 98000 Zacatecas, Zac. (Mexico)]|[Unidad Academica de Ing. Electrica, Universidad Autonoma de Zacatecas, Apdo. Postal 336, 98000 Zacatecas, Zac. (Mexico)]|[Unidad Academica de Matematicas, Universidad Autonoma de Zacatecas, Apdo. Postal 336, 98000 Zacatecas, Zac. (Mexico)]. E-mail: fermineutron@yahoo.com; Martin Hernandez-Davila, Victor [Unidad Academica de Estudios Nucleares, Universidad Autonoma de Zacatecas, Apdo. Postal 336, 98000 Zacatecas, Zac. (Mexico)]|[Unidad Academica de Ing. Electrica, Universidad Autonoma de Zacatecas, Apdo. Postal 336, 98000 Zacatecas, Zac. (Mexico); Manzanares-Acuna, Eduardo [Unidad Academica de Estudios Nucleares, Universidad Autonoma de Zacatecas, Apdo. Postal 336, 98000 Zacatecas, Zac. (Mexico); Mercado Sanchez, Gema A. [Unidad Academica de Matematicas, Universidad Autonoma de Zacatecas, Apdo. Postal 336, 98000 Zacatecas, Zac. (Mexico); Pilar Iniguez de la Torre, Maria [Depto. Fisica Teorica, Molecular y Nuclear, Universidad de Valladolid, Valladolid (Spain); Barquero, Raquel [Hospital Universitario Rio Hortega, Valladolid (Spain); Palacios, Francisco; Mendez Villafane, Roberto [Depto. Fisica Teorica, Molecular y Nuclear, Universidad de Valladolid, Valladolid (Spain)]|[Universidad Europea Miguel de Cervantes, C. Padre Julio Chevalier No. 2, 47012 Valladolid (Spain); Arteaga Arteaga, Tarcicio [Unidad Academica de Estudios Nucleares, Universidad Autonoma de Zacatecas, Apdo. Postal 336, 98000 Zacatecas, Zac. (Mexico)]|[Envases de Zacatecas, SA de CV, Parque Industrial de Calera de Victor Rosales, Zac. (Mexico); Manuel Ortiz Rodriguez, Jose [Unidad Academica de Estudios Nucleares, Universidad Autonoma de Zacatecas, Apdo. Postal 336, 98000 Zacatecas, Zac. (Mexico)]|[Unidad Academica de Ing. Electrica, Universidad Autonoma de Zacatecas, Apdo. Postal 336, 98000 Zacatecas, Zac. (Mexico)

    2006-04-15

    An artificial neural network has been designed to obtain neutron spectra from Bonner spheres spectrometer count rates. The neural network was trained using 129 neutron spectra. These include spectra from isotopic neutron sources; reference and operational spectra from accelerators and nuclear reactors, spectra based on mathematical functions as well as few energy groups and monoenergetic spectra. The spectra were transformed from lethargy to energy distribution and were re-binned to 31 energy groups using the MCNP 4C code. The re-binned spectra and the UTA4 response matrix were used to calculate the expected count rates in Bonner spheres spectrometer. These count rates were used as input and their respective spectra were used as output during the neural network training. After training, the network was tested with the Bonner spheres count rates produced by folding a set of neutron spectra with the response matrix. This set contains data used during network training as well as data not used. Training and testing was carried out using the Matlab{sup (R)} program. To verify the network unfolding performance, the original and unfolded spectra were compared using the root mean square error. The use of artificial neural networks to unfold neutron spectra in neutron spectrometry is an alternative procedure that overcomes the drawbacks associated with this ill-conditioned problem.

  2. Design considerations for neutron activation and neutron source strength monitors for ITER

    Energy Technology Data Exchange (ETDEWEB)

    Barnes, C.W. [Los Alamos National Lab., NM (United States); Jassby, D.L.; LeMunyan, G.; Roquemore, A.L. [Princeton Univ., NJ (United States). Plasma Physics Lab.; Walker, C. [ITER Joint Central Team, Garching (Germany)

    1997-12-31

    The International Thermonuclear Experimental Reactor will require highly accurate measurements of fusion power production in time, space, and energy. Spectrometers in the neutron camera could do it all, but experience has taught us that multiple methods with redundancy and complementary uncertainties are needed. Previously, conceptual designs have been presented for time-integrated neutron activation and time-dependent neutron source strength monitors, both of which will be important parts of the integrated suite of neutron diagnostics for this purpose. The primary goals of the neutron activation system are: to maintain a robust relative measure of fusion energy production with stability and wide dynamic range; to enable an accurate absolute calibration of fusion power using neutronic techniques as successfully demonstrated on JET and TFTR; and to provide a flexible system for materials testing. The greatest difficulty is that the irradiation locations need to be close to plasma with a wide field of view. The routing of the pneumatic system is difficult because of minimum radius of curvature requirements and because of the careful need for containment of the tritium and activated air. The neutron source strength system needs to provide real-time source strength vs. time with {approximately}1 ms resolution and wide dynamic range in a robust and reliable manner with the capability to be absolutely calibrated by in-situ neutron sources as done on TFTR, JT-60U, and JET. In this paper a more detailed look at the expected neutron flux field around ITER is folded into a more complete design of the fission chamber system.

  3. Neutron sources and applications

    Energy Technology Data Exchange (ETDEWEB)

    Price, D.L. [ed.] [Argonne National Lab., IL (United States); Rush, J.J. [ed.] [National Inst. of Standards and Technology, Gaithersburg, MD (United States)

    1994-01-01

    Review of Neutron Sources and Applications was held at Oak Brook, Illinois, during September 8--10, 1992. This review involved some 70 national and international experts in different areas of neutron research, sources, and applications. Separate working groups were asked to (1) review the current status of advanced research reactors and spallation sources; and (2) provide an update on scientific, technological, and medical applications, including neutron scattering research in a number of disciplines, isotope production, materials irradiation, and other important uses of neutron sources such as materials analysis and fundamental neutron physics. This report summarizes the findings and conclusions of the different working groups involved in the review, and contains some of the best current expertise on neutron sources and applications.

  4. Isolated Neutron Stars

    Directory of Open Access Journals (Sweden)

    Popov S.

    2010-10-01

    Full Text Available Several aspects related to astrophysics of isolated neutron stars are discussed. We start with an introduction into the “new zoo” of young isolated neutron stars. In addition to classical radio pulsars, now we know several species (soft gamma-ray repeators, anomalous X-ray pulsars, central compact objects in supernova remnants, close-by cooling neutron stars - aka “Magnificent seven”, - RRATs, and some others. All these types are briefly discussed. In the second lecture a description of magneto-rotational evolution of neutron stars is given. Finally, in the third lecture we discuss population synthesis of isolated neutron stars. In some details we discuss population synthesis of young isolated radio pulsars and young close-by cooling neutron stars.

  5. Prototype Stilbene Neutron Collar

    Energy Technology Data Exchange (ETDEWEB)

    Prasad, M. K. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Shumaker, D. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Snyderman, N. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Verbeke, J. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Wong, J. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2016-10-26

    A neutron collar using stilbene organic scintillator cells for fast neutron counting is described for the assay of fresh low enriched uranium (LEU) fuel assemblies. The prototype stilbene collar has a form factor similar to standard He-3 based collars and uses an AmLi interrogation neutron source. This report describes the simulation of list mode neutron correlation data on various fuel assemblies including some with neutron absorbers (burnable Gd poisons). Calibration curves (doubles vs 235U linear mass density) are presented for both thermal and fast (with Cd lining) modes of operation. It is shown that the stilbene collar meets or exceeds the current capabilities of He-3 based neutron collars. A self-consistent assay methodology, uniquely suited to the stilbene collar, using triples is described which complements traditional assay based on doubles calibration curves.

  6. Simulation of a high energy neutron irradiation facility at beamline 11 of the China Spallation Neutron Source

    Energy Technology Data Exchange (ETDEWEB)

    Tairan, Liang [School of Physics and Electronic Information Inner Mongolia University for the Nationalities, Tongliao 028043 (China); Zhiduo, Li [Dongguan Branch, Institute of High Energy Physics, CAS, Beijing 100049 (China); Wen, Yin, E-mail: wenyin@aphy.iphy.ac.cn [Dongguan Branch, Institute of High Energy Physics, CAS, Beijing 100049 (China); Institute of Physics, CAS, P.O. Box 603, Beijing 100190 (China); Fei, Shen [Dongguan Branch, Institute of High Energy Physics, CAS, Beijing 100049 (China); Quanzhi, Yu [Dongguan Branch, Institute of High Energy Physics, CAS, Beijing 100049 (China); Institute of Physics, CAS, P.O. Box 603, Beijing 100190 (China); Tianjiao, Liang [Dongguan Branch, Institute of High Energy Physics, CAS, Beijing 100049 (China)

    2017-07-11

    The China Spallation Neutron Source (CSNS) will accommodate 20 neutron beamlines at its first target station. These beamlines serve different purposes, and beamline 11 is designed to analyze the degraded models and damage mechanisms, such as Single Event Effects in electronic components and devices for aerospace electronic systems. This paper gives a preliminary discussion on the scheme of a high energy neutron irradiation experiment at the beamline 11 shutter based on the Monte Carlo simulation method. The neutron source term is generated by calculating the neutrons scattering into beamline 11 with a model that includes the target-moderator-reflector area. Then, the neutron spectrum at the sample position is obtained. The intensity of neutrons with energy of hundreds of MeV is approximately 1E8 neutron/cm{sup 2}/s, which is useful for experiments. The displacement production rate and gas productions are calculated for common materials such as tungsten, tantalum and SS316. The results indicate that the experiment can provide irradiation dose rate ranges from 1E-5 to 1E-4 dpa per operating year. The residual radioactivity is also calculated for regular maintenance work. These results give the basic reference for the experimental design.

  7. Simulation of a high energy neutron irradiation facility at beamline 11 of the China Spallation Neutron Source

    Science.gov (United States)

    Tairan, Liang; Zhiduo, Li; Wen, Yin; Fei, Shen; Quanzhi, Yu; Tianjiao, Liang

    2017-07-01

    The China Spallation Neutron Source (CSNS) will accommodate 20 neutron beamlines at its first target station. These beamlines serve different purposes, and beamline 11 is designed to analyze the degraded models and damage mechanisms, such as Single Event Effects in electronic components and devices for aerospace electronic systems. This paper gives a preliminary discussion on the scheme of a high energy neutron irradiation experiment at the beamline 11 shutter based on the Monte Carlo simulation method. The neutron source term is generated by calculating the neutrons scattering into beamline 11 with a model that includes the target-moderator-reflector area. Then, the neutron spectrum at the sample position is obtained. The intensity of neutrons with energy of hundreds of MeV is approximately 1E8 neutron/cm2/s, which is useful for experiments. The displacement production rate and gas productions are calculated for common materials such as tungsten, tantalum and SS316. The results indicate that the experiment can provide irradiation dose rate ranges from 1E-5 to 1E-4 dpa per operating year. The residual radioactivity is also calculated for regular maintenance work. These results give the basic reference for the experimental design.

  8. Radiative Neutron Capture on Lithium-7

    NARCIS (Netherlands)

    Rupak, Gautam; Renato, Higa

    2011-01-01

    The radiative neutron capture on lithium-7 is calculated model independently using a low energy halo effective field theory. The cross section is expressed in terms of scattering parameters directly related to the S-matrix element. The cross section depends on the poorly known p-wave effective range

  9. The DIORAMA Neutron Emitter

    Energy Technology Data Exchange (ETDEWEB)

    Terry, James Russell [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2016-05-05

    Emission of neutrons in a given event is modeled by the DioramaEmitterNeutron object, a subclass of the abstract DioramaEmitterModule object. The GenerateEmission method of this object is the entry point for generation of a neutron population for a given event. Shown in table 1, this method requires a number of parameters to be defined in the event definition.

  10. The thermal neutron facility HOTNES: theoretical design.

    Science.gov (United States)

    Bedogni, R; Pietropaolo, A; Gomez-Ros, J M

    2017-09-01

    HOTNES (HOmogeneous Thermal NEutron Source) is a thermal neutron irradiation facility with extended and very uniform irradiation area. A (241)Am-B radionuclide neutron source with nominal strenght 3.5×10(6) s(-1) is located on bottom of a large cylindrical cavity (30cm diameter, 70cm in height) delimited by polyethylene walls. The upper part of this volume (30cm diameter, 40cm in height) is used to irradiate samples. A polyethylene cylinder, acting as shadowing object, prevents fast neutrons to directly reach the irradiation volume. Indeed neutrons can only reach the irradiation volume after multiple scattering with the cavity walls. The facility was designed trough extensive calculations with MCNPX. Irradiation planes are disks with 30cm diameter, centred on the cavity axis, and parallel to the cavity bottom. The value of thermal fluence in a given irradiation plane is as uniform as 1-2%. The value of thermal fluence rate simply depends on the height from the cavity bottom. Values of thermal fluence rate in the range 700-1000cm(-2)s(-1) are available, depending on the irradiation plane chosen. The fraction of thermal neutrons is in the order of 90%, also depending on the irradiation plane. The angular distribution of thermal neutrons is roughly isotropic. Taking advantage of the HOTNES design, even large devices can be uniformly irradiated. This work presents HOTNES's design and describes the neutron field in the irradiation volume in terms of spatial, energy and direction distributions. Copyright © 2017 Elsevier Ltd. All rights reserved.

  11. Time reversal invariance in polarized neutron decay

    Energy Technology Data Exchange (ETDEWEB)

    Wasserman, Eric G. [Harvard Univ., Cambridge, MA (United States)

    1994-03-01

    An experiment to measure the time reversal invariance violating (T-violating) triple correlation (D) in the decay of free polarized neutrons has been developed. The detector design incorporates a detector geometry that provides a significant improvement in the sensitivity over that used in the most sensitive of previous experiments. A prototype detector was tested in measurements with a cold neutron beam. Data resulting from the tests are presented. A detailed calculation of systematic effects has been performed and new diagnostic techniques that allow these effects to be measured have been developed. As the result of this work, a new experiment is under way that will improve the sensitivity to D to 3 x 10-4 or better. With higher neutron flux a statistical sensitivity of the order 3 x 10-5 is ultimately expected. The decay of free polarized neutrons (n → p + e + $\\bar{v}$e) is used to search for T-violation by measuring the triple correlation of the neutron spin polarization, and the electron and proton momenta (σn • pp x pe). This correlation changes sign under reversal of the motion. Since final state effects in neutron decay are small, a nonzero coefficient, D, of this correlation indicates the violation of time reversal invariance. D is measured by comparing the numbers of coincidences in electron and proton detectors arranged symmetrically about a longitudinally polarized neutron beam. Particular care must be taken to eliminate residual asymmetries in the detectors or beam as these can lead to significant false effects. The Standard Model predicts negligible T-violating effects in neutron decay. Extensions to the Standard Model include new interactions some of which include CP-violating components. Some of these make first order contributions to D.

  12. Neutron structural biology

    Energy Technology Data Exchange (ETDEWEB)

    Niimura, Nobuo [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1999-03-01

    Neutron structural biology will be one of the most important fields in the life sciences which will interest human beings in the 21st century because neutrons can provide not only the position of hydrogen atoms in biological macromolecules but also the dynamic molecular motion of hydrogen atoms and water molecules. However, there are only a few examples experimentally determined at present because of the lack of neutron source intensity. Next generation neutron source scheduled in JAERI (Performance of which is 100 times better than that of JRR-3M) opens the life science of the 21st century. (author)

  13. Neutron-emission measurements at a white neutron source

    Energy Technology Data Exchange (ETDEWEB)

    Haight, Robert C [Los Alamos National Laboratory

    2010-01-01

    Data on the spectrum of neutrons emittcd from neutron-induced reactions are important in basic nuclear physics and in applications. Our program studies neutron emission from inelastic scattering as well as fission neutron spectra. A ''white'' neutron source (continuous in energy) allows measurements over a wide range of neutron energies all in one experiment. We use the tast neutron source at the Los Alamos Neutron Science Center for incident neutron energies from 0.5 MeV to 200 MeV These experiments are based on double time-of-flight techniques to determine the energies of the incident and emitted neutrons. For the fission neutron measurements, parallel-plate ionization or avalanche detectors identify fission in actinide samples and give the required fast timing pulse. For inelastic scattering, gamma-ray detectors provide the timing and energy spectroscopy. A large neutron-detector array detects the emitted neutrons. Time-of-flight techniques are used to measure the energies of both the incident and emitted neutrons. Design considerations for the array include neutron-gamma discrimination, neutron energy resolution, angular coverage, segmentation, detector efficiency calibration and data acquisition. We have made preliminary measurements of the fission neutron spectra from {sup 235}U, {sup 238}U, {sup 237}Np and {sup 239}Pu. Neutron emission spectra from inelastic scattering on iron and nickel have also been investigated. The results obtained will be compared with evaluated data.

  14. Consistency of differential and integral thermonuclear neutronics data

    Energy Technology Data Exchange (ETDEWEB)

    Reupke, W.A.

    1978-01-01

    To increase the accuracy of the neutronics analysis of nuclear reactors, physicists and engineers have employed a variety of techniques, including the adjustment of multigroup differential data to improve consistency with integral data. Of the various adjustment strategies, a generalized least-squares procedure which adjusts the combined differential and integral data can significantly improve the accuracy of neutronics calculations compared to calculations employing only differential data. This investigation analyzes 14 MeV neutron-driven integral experiments, using a more extensively developed methodology and a newly developed computer code, to extend the domain of adjustment from the energy range of fission reactors to the energy range of fusion reactors.

  15. Does mass accretion lead to field decay in neutron stars

    Science.gov (United States)

    Shibazaki, N.; Murakami, T.; Shaham, Jacob; Nomoto, K.

    1989-01-01

    The recent discovery of cyclotron lines from gamma-ray bursts indicates that the strong magnetic fields of isolated neutron stars might not decay. The possible inverse correlation between the strength of the magnetic field and the mass accreted by the neutron star suggests that mass accretion itself may lead to the decay of the magnetic field. The spin and magnetic field evolution of the neutron star was calculated under the hypothesis of the accretion-induced field decay. It is shown that the calculated results are consistent with the observations of binary and millisecond radio pulsars.

  16. Thermal neutron scattering evaluation framework

    Science.gov (United States)

    Chapman, Chris; Leal, Luiz; Rahnema, Farzad; Danon, Yaron; Arbanas, Goran

    2017-09-01

    A neutron scattering kernel data evaluation framework for computation of model-dependent predictions and their uncertainties is outlined. In this framework, model parameters are fitted to double-differential cross section measurements and their uncertainties. For convenience, the initial implementation of this framework uses the molecular dynamics model implemented in the GROMACS code. It is applied to light water using the TIP4P/2005f interaction model. These trajectories computed by GROMACS are then processed using nMOLDYN to compute the density of states, which is then used to calculate the scattering kernel using the Gaussian approximation. Double differential cross sections computed from the scattering kernel are then fitted to double-differential scattering data measured at the Spallation Neutron Source detector at Oak Ridge National Laboratory. The fitting procedure is designed to yield optimized model-parameters and their uncertainties in the form of a covariance matrix, from which new evaluations of thermal neutron scattering kernel will be generated. The Unified Monte Carlo method will be used to fit the simulation data to the experimental data.

  17. Calculations of the giant-dipole-resonance photoneutrons using a coupled EGS4-morse code

    Energy Technology Data Exchange (ETDEWEB)

    Liu, J.C.; Nelson, W.R.; Kase, K.R.; Mao, X.S.

    1995-10-01

    The production and transport of the photoneutrons from the giant-dipoleresonance reaction have been implemented in a coupled EGS4-MORSE code. The total neutron yield (including both the direct neutron and evaporation neutron components) is calculated by folding the photoneutron yield cross sections with the photon track length distribution in the target. Empirical algorithms based on the measurements have been developed to estimate the fraction and energy of the direct neutron component for each photon. The statistical theory in the EVAP4 code, incorporated as a MORSE subroutine, is used to determine the energies of the evaporation neutrons. These represent major improvements over other calculations that assumed no direct neutrons, a constant fraction of direct neutrons, monoenergetic direct neutron, or a constant nuclear temperature for the evaporation neutrons. It was also assumed that the slow neutrons (< 2.5 MeV) are emitted isotropically and the fast neutrons are emitted anisotropically in the form of 1+Csin{sup 2}{theta}, which have a peak emission at 900. Comparisons between the calculated and the measured photoneutron results (spectra of the direct, evaporation and total neutrons; nuclear temperatures; direct neutron fractions) for materials of lead, tungsten, tantalum and copper have been made. The results show that the empirical algorithms, albeit simple, can produce reasonable results over the interested photon energy range.

  18. Stability of interlinked neutron vortex and proton flux tube arrays in a neutron star: equilibrium configurations

    Science.gov (United States)

    Drummond, L. V.; Melatos, A.

    2017-12-01

    Three-dimensional, Gross-Pitaevskii equation (GPE) simulations are presented of the interaction between neutron superfluid vortices and proton superconductor flux tubes in a rotating, harmonic trap, representing an idealized model of the outer core of a neutron star. Low-energy states of the neutron condensate are calculated by evolving the GPE in imaginary time in the presence of a prescribed, static, rectilinear flux tube array. The calculations are carried out as a function of the angle between the global magnetic and rotation axes, and the amplitude and sign of the current-current and density couplings between the neutron and proton condensates. It is found that the system is frustrated by the competition between vortex-vortex repulsion and vortex-flux-tube attraction (pinning), leading to the formation of vortex tangles and 'glassy' behaviour characterized by multiple metastable states spaced closely in energy. The dimensionless parameters in the simulations are ordered as one expects in a neutron star, but the dynamic range is many orders of magnitude smaller than in reality, so caution must be exercised when assessing the astrophysical implications. Nevertheless, the results suggest that tangled vorticity may be endemic in neutron star outer cores.

  19. Influence of density and chemical composition of soils in the neutrons probes answer; Influencia da densidade e da composicao quimica dos solos na resposta de sondas de neutrons

    Energy Technology Data Exchange (ETDEWEB)

    Crispino, Marcos Luiz; Antonino, Antonio Celso Dantas; Dall`Olio, Attilio; Oliveira Lira, Carlos Alberto Brayner de [Pernambuco Univ., Recife, PE (Brazil). Dept. de Energia Nuclear; Carneiro, Clemente J. Gusmao [Minas Gerais Univ., Belo Horizonte, MG (Brazil). Dept. de Engenharia Nuclear

    1996-08-01

    The determination of soil humidity with neutron probes is based in the measure of the thermal neutron flux intensity and its behavior with the soil depend: soil`s chemical composition; soils physical parameters; neutrons` energetic spectrum and neutron-source detector geometry.The objective of this paper is to apply the multigroup function theory to calculate a neutron probe calibration curve utilizing representatives parameters and coefficients of soils horizons in a experimental station in Zona da Mata, Pernambuco, Brazil 2 tabs., 3 figs.

  20. Neutron Stars: Formation and Structure

    OpenAIRE

    Kutschera, Marek

    1998-01-01

    A short introduction is given to astrophysics of neutron stars and to physics of dense matter in neutron stars. Observed properties of astrophysical objects containing neutron stars are discussed. Current scenarios regarding formation and evolution of neutron stars in those objects are presented. Physical principles governing the internal structure of neutron stars are considered with special emphasis on the possible spin ordering in the neutron star matter.

  1. Correction factor K calculation for Americium-Beryllium neutron sources measured in a manganese sulfate bath; Calculo do fator de correcao K para fontes de neutrons de Americio-Berilio medida no banho de sulfato de manganes do LNMRI/IRD

    Energy Technology Data Exchange (ETDEWEB)

    Leite, Sandro P.; Fonseca, Evaldo S. da; Patrao, Karla C.S.; Goncalves, Marcello G. [Instituto de Radioprotecao e Dosimetria (IRD/CNEN-RJ), Rio de Janeiro, RJ (Brazil). Lab. de Neutrons; Pereira, Walsan W. [Instituto de Radioprotecao e Dosimetria (IRD/CNEN-RJ), Rio de Janeiro, RJ (Brazil). Lab. Nacional de Metrologia das Radiacoes Ionizantes (LNMRI)

    2005-03-15

    This paper simulates a manganese sulfate bath at the Ionizing Radiation Metrology National Laboratory for the calculation of K correction factor for the neutro emission ratio in some Americium-Beryllium sources.

  2. An Analytical Model of Leakage Neutron Equivalent Dose for Passively-Scattered Proton Radiotherapy and Validation with Measurements

    Energy Technology Data Exchange (ETDEWEB)

    Schneider, Christopher; Newhauser, Wayne, E-mail: newhauser@lsu.edu [Department of Physics and Astronomy, Louisiana State University and Agricultural and Mechanical College, 202 Nicholson Hall, Baton Rouge, LA 70803 (United States); Mary Bird Perkins Cancer Center, 4950 Essen Lane, Baton Rouge, LA 70809 (United States); Farah, Jad [Institut de Radioprotection et de Sûreté Nucléaire, Service de Dosimétrie Externe, BP-17, 92262 Fontenay-aux-Roses (France)

    2015-05-18

    Exposure to stray neutrons increases the risk of second cancer development after proton therapy. Previously reported analytical models of this exposure were difficult to configure and had not been investigated below 100 MeV proton energy. The purposes of this study were to test an analytical model of neutron equivalent dose per therapeutic absorbed dose (H/D) at 75 MeV and to improve the model by reducing the number of configuration parameters and making it continuous in proton energy from 100 to 250 MeV. To develop the analytical model, we used previously published H/D values in water from Monte Carlo simulations of a general-purpose beamline for proton energies from 100 to 250 MeV. We also configured and tested the model on in-air neutron equivalent doses measured for a 75 MeV ocular beamline. Predicted H/D values from the analytical model and Monte Carlo agreed well from 100 to 250 MeV (10% average difference). Predicted H/D values from the analytical model also agreed well with measurements at 75 MeV (15% average difference). The results indicate that analytical models can give fast, reliable calculations of neutron exposure after proton therapy. This ability is absent in treatment planning systems but vital to second cancer risk estimation.

  3. An Analytical Model of Leakage Neutron Equivalent Dose for Passively-Scattered Proton Radiotherapy and Validation with Measurements

    Directory of Open Access Journals (Sweden)

    Christopher Schneider

    2015-05-01

    Full Text Available Exposure to stray neutrons increases the risk of second cancer development after proton therapy. Previously reported analytical models of this exposure were difficult to configure and had not been investigated below 100 MeV proton energy. The purposes of this study were to test an analytical model of neutron equivalent dose per therapeutic absorbed dose  at 75 MeV and to improve the model by reducing the number of configuration parameters and making it continuous in proton energy from 100 to 250 MeV. To develop the analytical model, we used previously published H/D values in water from Monte Carlo simulations of a general-purpose beamline for proton energies from 100 to 250 MeV. We also configured and tested the model on in-air neutron equivalent doses measured for a 75 MeV ocular beamline. Predicted H/D values from the analytical model and Monte Carlo agreed well from 100 to 250 MeV (10% average difference. Predicted H/D values from the analytical model also agreed well with measurements at 75 MeV (15% average difference. The results indicate that analytical models can give fast, reliable calculations of neutron exposure after proton therapy. This ability is absent in treatment planning systems but vital to second cancer risk estimation.

  4. Optimization of Neutron Spectrum in Northwest Beam Tube of Tehran Research Reactor for BNCT, by MCNP Code

    Energy Technology Data Exchange (ETDEWEB)

    Zamani, M. [National Radiation Protection Department - NRPD, Atomic Energy Organization of Iran - AEOI, Tehran (Iran, Islamic Republic of); End of North Kargar st, Atomic Energy Organization of Iran, P.O. Box: 14155-1339, Tehran (Iran, Islamic Republic of); Kasesaz, Y.; Khalafi, H.; Shayesteh, M. [Radiation Application School, Nuclear Science and Technology Research Institute, AEOI, Tehran (Iran, Islamic Republic of)

    2015-07-01

    In order to gain the neutron spectrum with proper components specification for BNCT, it is necessary to design a Beam Shape Assembling (BSA), include of moderator, collimator, reflector, gamma filter and thermal neutrons filter, in front of the initial radiation beam from the source. According to the result of MCNP4C simulation, the Northwest beam tube has the most optimized neuron flux between three north beam tubes of Tehran Research Reactor (TRR). So, it has been chosen for this purpose. Simulation of the BSA has been done in four above mentioned phases. In each stage, ten best configurations of materials with different length and width were selected as the candidates for the next stage. The last BSA configuration includes of: 78 centimeters of air as an empty space, 40 centimeters of Iron plus 52 centimeters of heavy-water as moderator, 30 centimeters of water or 90 centimeters of Aluminum-Oxide as a reflector, 1 millimeters of lithium (Li) as thermal neutrons filter and finally 3 millimeters of Bismuth (Bi) as a filter of gamma radiation. The result of Calculations shows that if we use this BSA configuration for TRR Northwest beam tube, then the best neutron flux and spectrum will be achieved for BNCT. (authors)

  5. Cold neutron interferometry

    Science.gov (United States)

    Kitaguchi, Masaaki

    2009-10-01

    Neutron interferometry is a powerful technique for studying fundamental physics. A large dimensional interferometer for long wavelength neutrons is extremely important in order to investigate problems of fundamental physics, including tests of quantum measurement theories and searches for non-Newtonian effects of gravitation, since the sensitivity of interferometer depends on the wavelength and the interaction length. Neutron multilayer mirrors enable us to develop the large scale interferometer for long wavelength neutrons. The multilayer mirror is one of the most useful devices in cold neutron optics. A multilayer of two materials with different potentials is understood as a one-dimensional crystal, which is suitable for Bragg reflection of long wavelength neutrons. Cold and very cold neutrons can be utilized for the interferometer by using the multilayer mirrors with the proper lattice constants. Jamin-type interferometer by using beam splitting etalons (BSEs) has shown the feasibility of the development of large scale interferometer, which enables us to align the four independent mirrors within required precision. The BSE contains two parallel multilayer mirrors. A couple of the BSEs in the Jamin-type interferometer separates and recombines the two paths spatially. Although the path separation was small at the first test, now we have already demonstrated the interferometer with perfectly separated paths. This has confirmed that the multilayer mirrors cause no serious distortion of wave front to compose a interferometer. Arranging such mirrors, we are capable of establishing even a Mach-Zehnder type with much larger size. The interferometer using supermirrors, which reflects the wide range of the wavelength of neutrons, can increase the neutron counts for high precision measurements. We are planning the experiments using the interferometer both for the very cold neutrons and for the pulsed neutrons including J-PARC.

  6. Eleventh DOE workshop on personnel neutron dosimetry

    Energy Technology Data Exchange (ETDEWEB)

    1991-12-31

    Since its formation, the Office of Health (EH-40) has stressed the importance of the exchange of information related to and improvements in neutron dosimetry. This Workshop was the eleventh in the series sponsored by the Department of Energy (DOE). It provided a forum for operational personnel at DOE facilities to discuss current issues related to neutron dosimetry and for leading investigators in the field to discuss promising approaches for future research. A total of 26 papers were presented including the keynote address by Dr. Warren K. Sinclair, who spoke on, ``The 1990 Recommendations of the ICRP and their Biological Background.`` The first several papers discussed difficulties in measuring neutrons of different energies and ways of compensating or deriving correction factors at individual facilities. Presentations were also given by the US Navy and Air Force. Current research in neutron dosimeter development was the subject of the largest number of papers. These included a number on the development of neutron spectrometers. Selected papers were processed separately for inclusion in the Energy Science and Technology Database.

  7. Neutronics calculation of an heterogeneous compact and thermal core by means of deterministic and stochastic transport theory. Application to the experimental reactor of the University of Strasbourg; Modelisation neutronique d`un coeur thermique compact et heterogene en theorie du transport deterministe et probabiliste. Application au reacteur experimental de l`Universite de Strasbourg

    Energy Technology Data Exchange (ETDEWEB)

    Jammes, Ch

    1997-11-28

    The aim of this work is to create, validate theoretically and experimentally a calculation route for a thermal irradiation reactor. This is the research reactor of the University of Strasbourg, which presents all of characteristics of this reactor-type: compact and heterogeneous core, slab-type fuel with a high 235-uranium enrichment. This calculation route is based on the first use of the following two modern transport methods: the TDT method and the Monte Carlo method. The former, programmed within the APOLLO2 code, is a two dimensional collision probabilities method. The later, used by the TRIPOLI4 code, is a stochastic method. Both can be applied to complex geometries. After a few theoretical reminders about transport codes, a set of integral experiments is described which have been realized within the research reactor of the University of Strasbourg. One of them has been performed for this study. At the beginning of the theoretical part, significant errors are apparent due to the use of calculation route based on homogenization, condensation and the diffusion approximation. An extensive comparison between the discrete ordinates method and the TDT method carries out that the use of the TDT method is relevant for the studied reactor. The treatment of axial leakage with this method is the only disadvantage. Therefore, the use of the code TRIPOLI4 is recommended for a more accurate study of leakage within a reflector. By means of the experimental data, the ability of our calculation route is confirmed for essential neutronics questions such as the critical mass determination, the power distribution and the fuel management. (author)

  8. MEMS Calculator

    Science.gov (United States)

    SRD 166 MEMS Calculator (Web, free access)   This MEMS Calculator determines the following thin film properties from data taken with an optical interferometer or comparable instrument: a) residual strain from fixed-fixed beams, b) strain gradient from cantilevers, c) step heights or thicknesses from step-height test structures, and d) in-plane lengths or deflections. Then, residual stress and stress gradient calculations can be made after an optical vibrometer or comparable instrument is used to obtain Young's modulus from resonating cantilevers or fixed-fixed beams. In addition, wafer bond strength is determined from micro-chevron test structures using a material test machine.

  9. Enhancement of a 252Cf-based neutron beam via subcritical multiplication for neutron capture therapy.

    Science.gov (United States)

    Wang, C K; Zino, J F; Kessler, G

    2000-01-01

    Previous studies indicated that an epithermal-neutron beam based on bare 252Cf is not feasible for neutron capture therapy (NCT). It was reported that a clinically useful epithermal-neutron beam requires a minimum of 1.0 g of 252Cf, which is more than twice the US current annual supply. However, it was reasoned that the required quantity of 252Cf could be dramatically reduced when used with a subcritical multiplying assembly (SMA). This reasoning is based on the assumption that the epithermal-neutron beam intensity for NCT is directly proportional to the fission neutron population, and that the neutron multiplying factor of the SMA can be estimated by 1/(1 - k(eff)). We have performed detailed Monte Carlo calculations to investigate the validity of the above reasoning. Our results show that 1/(1 - k(eff)) grossly overestimates the beam enhancement factor for NCT. For example, Monte Carlo calculations predict a beam enhancement factor of 6.0 for an optimized SMA geometry with k(eff) = 0.968. This factor is much less than 31 predicted by 1/(1 - k(eff)). The overestimation is due to the fact that most of the neutrons produced in the SMA are self-shielded, whereas self-shielding is negligible in a bare 252Cf source. Since the beam intensity of a 0.1 g 252Cf with the optimized SMA enhancement is still more than an order of magnitude too low compared to the existing reactor beams, we conclude that the enhancement via an SMA for a 252Cf-based epithermal-neutron beam is inadequate for NCT.

  10. The impact of air pollutant and methane emission controls on tropospheric ozone and radiative forcing: CTM calculations for the period 1990-2030

    Directory of Open Access Journals (Sweden)

    F. Dentener

    2005-01-01

    Full Text Available To explore the relationship between tropospheric ozone and radiative forcing with changing emissions, we compiled two sets of global scenarios for the emissions of the ozone precursors methane (CH4, carbon monoxide (CO, non-methane volatile organic compounds (NMVOC and nitrogen oxides (NOx up to the year 2030 and implemented them in two global Chemistry Transport Models. The 'Current Legislation' (CLE scenario reflects the current perspectives of individual countries on future economic development and takes the anticipated effects of presently decided emission control legislation in the individual countries into account. In addition, we developed a 'Maximum technically Feasible Reduction' (MFR scenario that outlines the scope for emission reductions offered by full implementation of the presently available emission control technologies, while maintaining the projected levels of anthropogenic activities. Whereas the resulting projections of methane emissions lie within the range suggested by other greenhouse gas projections, the recent pollution control legislation of many Asian countries, requiring introduction of catalytic converters for vehicles, leads to significantly lower growth in emissions of the air pollutants NOx, NMVOC and CO than was suggested by the widely used and more pessimistic IPCC (Intergovernmental Panel on Climate Change SRES (Special Report on Emission Scenarios scenarios (Nakicenovic et al., 2000, which made Business-as-Usual assumptions regarding emission control technology. With the TM3 and STOCHEM models we performed several long-term integrations (1990-2030 to assess global, hemispheric and regional changes in CH4, CO, hydroxyl radicals, ozone and the radiative climate forcings resulting from these two emission scenarios. Both models reproduce broadly the observed trends in CO, and CH4 concentrations from 1990 to 2002. For the 'current legislation' case, both models indicate an increase of the annual average ozone

  11. Contributions to the qualification of the ''CRISTAL'' criticality calculi scheme: interpretation of critical experiments. Elaboration of a characterization system of neutronic configurations; Contributions a la qualification du schema de calcul de criticite ''cristal'': interpretation d'experiences critiques. Elaboration d'un systeme de caracterisation des configurations neutroniques

    Energy Technology Data Exchange (ETDEWEB)

    Gagnier, E

    1999-06-01

    This thesis work is about the validation of the new criticality-safety package CRISTAL and contributes to the modernization and the improvement of the computational tools. The first part presents neutronic elements, the objectives of safety criticality studies and the package CRISTAL. Then, the validation work concerned two series of experiments involving uranyl solutions (UO{sub 2}F{sub 2}) and UO{sub 2} powders. For these experiments, the differences between the computation results and the experimental results were analysed. It was highlighted interesting physical phenomena such of the compensations of errors between the approximate representation by the 99 energy group structure on the first resonance of oxygen and the anisotropy of the diffusion simulation as well as the influence of uranium 234 in high enriched solutions in uranium 235. Once the work of the experimental qualification carried out, raises the question of the use the base of qualification and the ''calculation-experiment'' variations which are referred to it. It is often difficult to establish the link between the ''studied configuration'' and the experiments of the base of qualification. The presented characterisation system proposes to answer in a way automatic and quantified this difficulty: - in bringing an answer on the package qualification for the studied configuration, - in giving an estimate of the package bias. To answer these points, it was defined a set of 35 characteristic neutronic parameters representing the behaviour of the medium. To process the information brought by these parameters and to use it to answer the objectives of the system, we called upon statistical methods (Principal Components Analysis and Sliced Inverse Regression). The results obtained in the feasibility studies showed the relevance of these methods for the considered objectives. (author)

  12. Response of a neutron monitor area with TLDs pairs

    Energy Technology Data Exchange (ETDEWEB)

    Guzman G, K. A.; Borja H, C. G.; Valero L, C.; Hernandez D, V. M.; Vega C, H. R. [Universidad Autonoma de Zacatecas, Unidad Academica de Estudios Nucleares, Calle Cipres No. 10, Fracc. La Penuela, 98068 Zacatecas (Mexico); Gallego, E.; Lorente, A., E-mail: ing_karen_guzman@yahoo.com.mx [Universidad Politecnica de Madrid, Departamento de Ingenieria Nuclear, Jose Gutierrez Abascal 2, E-28006 Madrid (Spain)

    2011-10-15

    The response of a passive neutron monitor area has been calculated using the Monte Carlo code MCNP5. The response was the amount of n({sup 6}Li, T){alpha} reactions occurring in a TLD-600 located at the center of a cylindrical polyethylene moderator. Fluence, (n, a) and H*(10) responses were calculated for 47 monoenergetic neutron sources. The H*(10) relative response was compared with responses of commercially available neutron monitors being alike. Due to {sup 6}Li cross section (n, {alpha}) reactions are mainly produced by thermal neutrons, however TLD-600 is sensitive to gamma-rays; to eliminate the signal due to photons monitor area was built to hold 2 pairs of TLD-600 and 2 pairs of TLD-700, thus from the difference between TLD-600 and TLD-700 readouts the net signal due to neutrons is obtained. The monitor area was calibrated at the Universidad Politecnica de Madrid using a {sup 241}AmBe neutron source; net TLD readout was compared with the H*(10) measured with a Bert hold Lb-6411. Performance of the neutron monitor area was determined through two independent experiments, in both cases the H*(10) was statistically equal to H*(10) measured with a Bert hold Lb-6411. Neutron monitor area with TLDs pairs can be used in working areas with intense, mixed and pulsed radiation fields. (Author)

  13. Holographic Quark Matter and Neutron Stars.

    Science.gov (United States)

    Hoyos, Carlos; Jokela, Niko; Rodríguez Fernández, David; Vuorinen, Aleksi

    2016-07-15

    We use a top-down holographic model for strongly interacting quark matter to study the properties of neutron stars. When the corresponding equation of state (EOS) is matched with state-of-the-art results for dense nuclear matter, we consistently observe a first-order phase transition at densities between 2 and 7 times the nuclear saturation density. Solving the Tolman-Oppenheimer-Volkov equations with the resulting hybrid EOSs, we find maximal stellar masses in excess of two solar masses, albeit somewhat smaller than those obtained with simple extrapolations of the nuclear matter EOSs. Our calculation predicts that no quark matter exists inside neutron stars.

  14. Neutron spectrometry and dosimetry with ANNs

    Energy Technology Data Exchange (ETDEWEB)

    Vega C, H. R.; Hernandez D, V. M. [Unidad Academica de Estudios Nucleares, Universidad Autonoma de Zacatecas, Cipres 10, Fracc. La Penuela, 98068 Zacatecas (Mexico); Gallego, E.; Lorente, A. [Departamento de Ingenieria Nuclear, Universidad Politecnica de Madrid, Jose Gutierrez Abascal 2, 28006 Madrid (Spain)], e-mail: fermineutron@yahoo.com

    2009-10-15

    Artificial neural networks technology has been applied to unfold the neutron spectra and to calculate the effective dose, the ambient equivalent dose, and the personal dose equivalent for {sup 252}Cf and {sup 241}AmBe neutron sources. A Bonner sphere spectrometry with a {sup 6}LiI(Eu) scintillator was utilized to measure the count rates of the spheres that were utilized as input in two artificial neural networks, one for spectrometry and another for dosimetry. Spectra and the ambient dose equivalent were also obtained with BUNKIUT code and the UTA4 response matrix. With both procedures spectra and ambient dose equivalent agrees in less than 10%. (author)

  15. Passive neutron dosemeter with activation detector

    Energy Technology Data Exchange (ETDEWEB)

    Valero L, C.; Banuelos F, A.; Guzman G, K. A.; Borja H, C. G.; Hernandez D, V. M.; Vega C, H. R., E-mail: fermineutron@yahoo.com [Universidad Autonoma de Zacatecas, Unidad Academica de Estudios Nucleares, Calle Cipres No. 10, Fracc. La Penuela, 98068 Zacatecas (Mexico)

    2011-10-15

    A passive neutron dosemeter with {sup 197}Au activation detector has been developed. The area dosemeter was made as a 20.5 {phi} x 20.5 cm{sup 2} polyethylene moderator, with a polyethylene pug where a {sup 197}Au foil can be located either parallel or perpendicular to moderator axis. Using Monte Carlo methods, with the MCNP5 code. With the fluence response and the fluence-to-equivalent dose conversion coefficients from ICRP-74, responses to H*(10) were also calculated, these were compared against responses of commercially available neutron area monitors and dosemeters. (Author)

  16. Neutron elastic and inelastic cross section measurements for 28Si

    Science.gov (United States)

    Derdeyn, E. C.; Lyons, E. M.; Morin, T.; Hicks, S. F.; Vanhoy, J. R.; Peters, E. E.; Ramirez, A. P. D.; McEllistrem, M. T.; Mukhopadhyay, S.; Yates, S. W.

    2017-09-01

    Neutron elastic and inelastic cross sections are critical for design and implementation of nuclear reactors and reactor equipment. Silicon, an element used abundantly in fuel pellets as well as building materials, has little to no experimental cross sections in the fast neutron region to support current theoretical evaluations, and thus would benefit from any contribution. Measurements of neutron elastic and inelastic differential scattering cross sections for 28Si were performed at the University of Kentucky Accelerator Laboratory for incident neutron energies of 6.1 MeV and 7.0 MeV. Neutrons were produced by accelerated deuterons incident on a deuterium gas cell. These nearly mono-energetic neutrons then scattered off a natural Si sample and were detected using liquid deuterated benzene scintillation detectors. Scattered neutron energy was deduced using time-of-flight techniques in tandem with kinematic calculations for an angular distribution. The relative detector efficiency was experimentally determined over a neutron energy range from approximately 0.5 to 7.75 MeV prior to the experiment. Yields were corrected for multiple scattering and neutron attenuation in the sample using the forced-collision Monte Carlo correction code MULCAT. Resulting cross sections will be presented along with comparisons to various data evaluations. Research is supported by USDOE-NNSA-SSAP: NA0002931, NSF: PHY-1606890, and the Donald A. Cowan Physics Institute at the University of Dallas.

  17. Neutron spectra due (13)N production in a PET cyclotron.

    Science.gov (United States)

    Benavente, J A; Vega-Carrillo, H R; Lacerda, M A S; Fonseca, T C F; Faria, F P; da Silva, T A

    2015-05-01

    Monte Carlo and experimental methods have been used to characterize the neutron radiation field around PET (Positron Emission Tomography) cyclotrons. In this work, the Monte Carlo code MCNPX was used to estimate the neutron spectra, the neutron fluence rates and the ambient dose equivalent (H*(10)) in seven locations around a PET cyclotron during (13)N production. In order to validate these calculations, H*(10) was measured in three sites and were compared with the calculated doses. All the spectra have two peaks, one above 0.1MeV due to the evaporation neutrons and another in the thermal region due to the room-return effects. Despite the relatively large difference between the measured and calculated H*(10) for one point, the agreement was considered good, compared with that obtained for (18)F production in a previous work. Copyright © 2015 Elsevier Ltd. All rights reserved.

  18. Sensitivity Analysis of Core Neutronic Parameters in Electron Accelerator-driven Subcritical Advanced Liquid Metal Reactor

    OpenAIRE

    Ebrahimkhani, Marziye; Hassanzadeh, Mostafa; Feghhi, Sayed Amier Hossian; Masti, Darush

    2016-01-01

    Calculation of the core neutronic parameters is one of the key components in all nuclear reactors. In this research, the energy spectrum and spatial distribution of the neutron flux in a uranium target have been calculated. In addition, sensitivity of the core neutronic parameters in accelerator-driven subcritical advanced liquid metal reactors, such as electron beam energy (Ee) and source multiplication coefficient (ks), has been investigated. A Monte Carlo code (MCNPX_2.6) has been used to ...

  19. Neutron Multiplicity Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Frame, Katherine Chiyoko [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-06-28

    Neutron multiplicity measurements are widely used for nondestructive assay (NDA) of special nuclear material (SNM). When combined with isotopic composition information, neutron multiplicity analysis can be used to estimate the spontaneous fission rate and leakage multiplication of SNM. When combined with isotopic information, the total mass of fissile material can also be determined. This presentation provides an overview of this technique.

  20. Project of the URAN array for registration of atmospheric neutrons

    Science.gov (United States)

    Gromushkin, D. M.; Barbashina, N. S.; Bogdanov, F. A.; Kokoulin, R. P.; Ovchinnikov, V. V.; Petrukhin, A. A.; Stenkin, Yu V.; Khokhlov, S. S.; Shulzhenko, I. A.; Yashin, I. I.

    2016-02-01

    The project of a new setup is directed at the registration of atmospheric neutrons (URAN) generated by hadronic component of extensive air showers (EAS). The setup includes 72 en-detector which simultaneously register two major EAS components: electromagnetic by the group passage of charged particles and hadron component by the thermal neutrons. The neutrons and charged particles are detected using a specialized scintillation composition made of granulated alloy of crystals based on the ZnS(Ag) powder with an admixture of B2O3.