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. Neutrons in the moon. [neutron flux and production rate calculations

    Science.gov (United States)

    Kornblum, J. J.; Fireman, E. L.; Levine, M.; Aronson, A.

    1973-01-01

    Neutron fluxes for energies between 15 MeV and thermal at depths of 0 to 300 g/sq cm in the moon are calculated by the discrete ordinate mathod with the ANISN code. With the energy spectrum of Lingenfelter et al. (1972). A total neutron-production rate for the moon of 26 plus or minus neutrons/sq cm sec is determined from the Ar-37 activity measurements in the Apollo 16 drill string, which are found to have a depth dependence in accordance with a neutron source function that decreases exponentially with an attenuation length of 155 g/sq cm.

  4. Air Blast Calculations

    Science.gov (United States)

    2013-07-01

    gauges at this point are still inside the explosive fireball (where the pressure time histories are dramatically influenced by the propagating...outside of the explosive fireball (since no evidence of an air-product interface is detected in any of the pressure time histories at this point

  5. Quantum Monte Carlo Calculations of Neutron Matter

    CERN Document Server

    Carlson, J; Ravenhall, D G

    2003-01-01

    Uniform neutron matter is approximated by a cubic box containing a finite number of neutrons, with periodic boundary conditions. We report variational and Green's function Monte Carlo calculations of the ground state of fourteen neutrons in a periodic box using the Argonne $\\vep $ two-nucleon interaction at densities up to one and half times the nuclear matter density. The effects of the finite box size are estimated using variational wave functions together with cluster expansion and chain summation techniques. They are small at subnuclear densities. We discuss the expansion of the energy of low-density neutron gas in powers of its Fermi momentum. This expansion is strongly modified by the large nn scattering length, and does not begin with the Fermi-gas kinetic energy as assumed in both Skyrme and relativistic mean field theories. The leading term of neutron gas energy is ~ half the Fermi-gas kinetic energy. The quantum Monte Carlo results are also used to calibrate the accuracy of variational calculations ...

  6. Neutronic calculations for a final focus system

    Energy Technology Data Exchange (ETDEWEB)

    Mainardi, E. E-mail: enrico@nuc.berkeley.edu; Premuda, F.; Lee, E

    2001-05-21

    For heavy-ion fusion and for 'liquid-protected' reactor designs such as HYLIFE-II (Moir et al., Fusion Technol. 25 (1994); HYLIFE-II-Progress Report, UCID-21816, 4-82-100), a mixture of molten salts made of F{sup 10}, Li{sup 6}, Li{sup 7}, Be{sup 9} called flibe allows highly compact target chambers. Smaller chambers will have lower costs and will allow the final-focus magnets to be closer to the target with decreased size of the focus spot and of the driver, as well as drastically reduced costs of IFE electricity. Consequently the superconducting coils of the magnets closer to the chamber will suffer higher radiation damage though they can stand only a certain amount of energy deposited before quenching. The scope of our calculations is essentially the total energy deposited on the magnetic lens system by fusion neutrons and induced {gamma}-rays. Such a study is important for the design of the final focus system itself from the neutronic point of view and indicates some guidelines for a design with six magnets in the beam line. The entire chamber consists of 192 beam lines to provide access of heavy ions that will implode the pellet. A 3-D transport calculation of the radiation penetrating through ducts that takes into account the complexity of the system, requires Monte Carlo methods. The development of efficient and precise models for geometric representation and nuclear analysis is necessary. The parameters are optimized thanks to an accurate analysis of six geometrical models that are developed starting from the simplest. Different configurations are examined employing TART 98 (D.E. Cullen, Lawrence Livermore National Laboratory, UCRL-ID-126455, Rev. 1, November, 1997) and MCNP 4B (Briesmeister (Ed.), Version 4B, La-12625-m, March 1997, Los Alamos National Laboratory): two Monte Carlo codes for neutrons and photons. The quantities analyzed include: energy deposited by neutrons and gamma photons, values of the total fluence integrated on the whole

  7. Relativistic calculations of coalescing binary neutron stars

    Indian Academy of Sciences (India)

    Joshua Faber; Phillippe Grandclément; Frederic Rasio

    2004-10-01

    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 quasi-circular orbits of equilibrium solutions. By adding a radiation reaction treatment, we compute the full evolution of a coalescing binary neutron star system. We find that the amount of mass ejected from the system, much less than a per cent, is greatly reduced by the inclusion of relativistic gravitation. The gravity wave energy spectrum shows a clear divergence away from the Newtonian point-mass form, consistent with the form derived from relativistic quasi-equilibrium fluid sequences.

  8. Neutron dosimetry and radiation damage calculations for HFBR

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-03-01

    Neutron dosimetry measurements have been conducted for various positions of the High Flux Beam Reactor (HFBR) at Brookhaven National Laboratory (BNL) in order to measure the neutron flux and energy spectra. Neutron dosimetry results and radiation damage calculations are presented for positions V10, V14, and V15.

  9. Improvement of Neutronics Calculation Methods for Fast Reactors

    OpenAIRE

    Takeda, Toshikazu

    2011-01-01

    To accurately estimate neutronics properties of fast reactors, particularly Japan Sodium-cooled Fast Reactor of1,500 MW electric, calculational methods are being improved in Japan.This paper describes the planning and the ongoing development of the neutronics calculation methods in the fieldof 1) assembly calculations including the calculations of effective cross sections, 2) core calculations and 3) uncertaintyevaluation and uncertainty reduction.

  10. Calculating and measuring thermal neutrons exiting from neutron diffractometers collimators

    CERN Document Server

    Tafazolee, K

    2000-01-01

    process, effectiveness of them are studied for the enhancement of the available system. Final conclusion from the simulation process, indicates that the heavy water with the thickness of 50 to 60 cm. is the best moderator for gaining the better thermal neutrons flux for enhancement of P.N.D. in the T.R.R. Powder Neutron Diffractometer y (P.N.D.) is relatively good and practical way for identification of the 3 dimensional construction of materials. In order to exploit the capabilities of this method, in one of the neutron beam of the Tehran Research Reactor (T.R.R.), a collimator embedded inside the concrete wall, direct the neutrons produced in the core reactor towards a monochromator e. Neutrons having been monochromated by 2 nd collimator are then directed towards the sample. Then the pattern of diffracted neutrons from the sample are studied. In order to make the best out of it, neutrons coming to sit on the sample must be of the thermal type. That means the number/amount of thermal neutrons flux in compar...

  11. Neutron spectra and dose equivalents calculated in tissue for high-energy radiation therapy

    Energy Technology Data Exchange (ETDEWEB)

    Kry, Stephen F.; Howell, Rebecca M.; Salehpour, Mohammad; Followill, David S. [Department of Radiation Physics, University of Texas M. D. Anderson Cancer Center, 1515 Holcombe Boulevard, Houston, Texas 77030 (United States)

    2009-04-15

    Neutrons are by-products of high-energy radiation therapy and a source of dose to normal tissues. Thus, the presence of neutrons increases a patient's risk of radiation-induced secondary cancer. Although neutrons have been thoroughly studied in air, little research has been focused on neutrons at depths in the patient where radiosensitive structures may exist, resulting in wide variations in neutron dose equivalents between studies. In this study, we characterized properties of neutrons produced during high-energy radiation therapy as a function of their depth in tissue and for different field sizes and different source-to-surface distances (SSD). We used a previously developed Monte Carlo model of an accelerator operated at 18 MV to calculate the neutron fluences, energy spectra, quality factors, and dose equivalents in air and in tissue at depths ranging from 0.1 to 25 cm. In conjunction with the sharply decreasing dose equivalent with increased depth in tissue, the authors found that the neutron energy spectrum changed drastically as a function of depth in tissue. The neutron fluence decreased gradually as the depth increased, while the average neutron energy decreased sharply with increasing depth until a depth of approximately 7.5 cm in tissue, after which it remained nearly constant. There was minimal variation in the quality factor as a function of depth. At a given depth in tissue, the neutron dose equivalent increased slightly with increasing field size and decreasing SSD; however, the percentage depth-dose equivalent curve remained constant outside the primary photon field. Because the neutron dose equivalent, fluence, and energy spectrum changed substantially with depth in tissue, we concluded that when the neutron dose equivalent is being determined at a depth within a patient, the spectrum and quality factor used should be appropriate for depth rather than for in-air conditions. Alternately, an appropriate percent depth-dose equivalent curve

  12. 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.

  13. Calculating fusion neutron energy spectra from arbitrary reactant distributions

    Science.gov (United States)

    Eriksson, J.; Conroy, S.; Andersson Sundén, E.; Hellesen, C.

    2016-02-01

    The Directional Relativistic Spectrum Simulator (DRESS) code can perform Monte-Carlo calculations of reaction product spectra from arbitrary reactant distributions, using fully relativistic kinematics. The code is set up to calculate energy spectra from neutrons and alpha particles produced in the D(d, n)3He and T(d, n)4He fusion reactions, but any two-body reaction can be simulated by including the corresponding cross section. The code has been thoroughly tested. The kinematics calculations have been benchmarked against the kinematics module of the ROOT Data Analysis Framework. Calculated neutron energy spectra have been validated against tabulated fusion reactivities and against an exact analytical expression for the thermonuclear fusion neutron spectrum, with good agreement. The DRESS code will be used as the core of a detailed synthetic diagnostic framework for neutron measurements at the JET and MAST tokamaks.

  14. Calculated characteristics of subcritical assembly with anisotropic transport of neutrons

    Energy Technology Data Exchange (ETDEWEB)

    Gorin, N.V.; Lipilina, E.N.; Lyutov, V.D.; Saukov, A.I. [Zababakhin Russian Federal Nuclear Center - All-Russian Scientific Researching Institute of Technical Physics (Russian Federation)

    2003-07-01

    There was considered possibility of creating enough sub-critical system that multiply neutron fluence from a primary source by many orders. For assemblies with high neutron tie between parts, it is impossible. That is why there was developed a construction consisting of many units (cascades) having weak feedback with preceding cascades. The feedback attenuation was obtained placing layers of slow neutron absorber and moderators between the cascades of fission material. Anisotropy of fast neutron transport through the layers was used. The system consisted of many identical cascades aligning one by another. Each cascade consists of layers of moderator, fissile material and absorber of slow neutrons. The calculations were carried out using the code MCNP.4a with nuclear data library ENDF/B5. In this construction neutrons spread predominantly in one direction multiplying in each next fissile layer, and they attenuate considerably in the opposite direction. In a calculated construction, multiplication factor of one cascade is about 1.5 and multiplication factor of whole construction composed of n cascades is 1.5{sup n}. Calculated keff value is 0.9 for one cascade and does not exceed 0.98 for a system containing any number of cascades. Therefore the assembly is always sub-critical and therefore it is safe in respect of criticality. There was considered using such a sub-critical assembly to create a powerful neutron fluence for neutron boron-capturing therapy. The system merits and demerits were discussed. (authors)

  15. Application of ex-vessel neutron dosimetry combined with in-core measurements for correction of neutron source used for RPV fluence calculations

    Energy Technology Data Exchange (ETDEWEB)

    Borodkin, P.G.; Borodkin, G.I.; Khrennikov, N.N. [Scientific and Engineering Centre for Nuclear and Radiation Safety SEC NRS, Malaya Krasnoselskaya ul., 2/8, Bld. 5, 107140 Moscow (Russian Federation); Konheiser, J. [Helmholz Zentrum Dresden-Rossendorf HZDR, Postfach 510119, D-01314 Dresden (Germany)

    2011-07-01

    This paper deals with calculated and semi-analytical evaluations of VVER-1000 reactor core neutron source distributions and their influence on measurements and calculations of the integral through-vessel neutron leakage. Neutron activation measurements analyzed in the paper were carried out in an ex-vessel air cavity at different nuclear power plant units with VVER-1000 during different fuel cycles. The time-integrated neutron source distributions used for DORT calculations were prepared via two different approaches based on (a) calculated fuel burnup (standard routine procedure) and (b) in-core measurements by means of self-powered detectors (SPDs) and thermocouples (TCs) (new approach). Considering that fuel burnup distributions in operating VVER may be evaluated now by the use of analytical methods (calculations) only, it is necessary to develop new approaches for the testing and correction of calculated evaluations of a neutron source. The results presented in this paper allow one to consider the reverse task of the alternative estimation of fuel burnup distributions. The proposed approach is based on the adjustment (fitting) of time-integrated neutron source distributions, and thus fuel burnup patterns, in some part of the reactor core, taking into account neutron leakage measurements, neutron-physical calculations, and in-core SPD and TC measurement data. (authors)

  16. nxs a program library for neutron cross section calculations

    OpenAIRE

    Boin, M.

    2012-01-01

    A collection of routines for calculating neutron scattering and absorption cross sections on the basis of crystal structure descriptions is presented and implemented in the new and reusable nxs program library. An example program providing a graphical user interface to the nxs functions is created to demonstrate their usage. The flexibility of the library and the possibilities for multiple areas of application are shown by further examples involving Monte Carlo neutron simulations concerned ...

  17. Monte Carlo Calculations for Neutron and Gamma Radiation Fields on a Fast Neutron Irradiation Device

    Science.gov (United States)

    Vieira, A.; Ramalho, A.; Gonçalves, I. C.; Fernandes, A.; Barradas, N.; Marques, J. G.; Prata, J.; Chaussy, Ch.

    We used the Monte Carlo program MCNP to calculate the neutron and gamma fluxes on a fast neutron irradiation facility being installed on the Portuguese Research Reactor (RPI). The purpose of this facility is to provide a fast neutron beam for irradiation of electronic circuits. The gamma dose should be minimized. This is achieved by placing a lead shield preceded by a thin layer of boral. A fast neutron flux of the order of 109 n/cm2s is expected at the exit of the tube, while the gamma radiation is kept below 20 Gy/h. We will present results of the neutron and gamma doses for several locations along the tube and different thickness of the lead shield. We found that the neutron beam is very collimated at the end of the tube with a dominant component on the fast region.

  18. Neutron-deuteron scattering calculation for evaluated neutron data libraries

    Science.gov (United States)

    Svenne, J. P.; Canton, L.; Kozier, K. S.

    2008-12-01

    In the low-energy regime, differential cross sections for n + d elastic scattering are not well described in existing nuclear data libraries, such as ENDF/B-VII.0. Supporting experimental data in this energy region are old, sparse and often inconsistent. We have carried out calculations with the AGS three-body theory and the Bonn-B nucleon-nucleon potential at energies 50 keV to 10.0 MeV.

  19. Computational program to neutron flux calculation; Programa computacional para calculo de fluxo de neutrons

    Energy Technology Data Exchange (ETDEWEB)

    Souza, Maria Ines Silvani; Furieri, Rosanne Cefaly de Aranda Amado [Instituto de Engenharia Nuclear (IEN), Rio de Janeiro, RJ (Brazil)

    2000-07-01

    The absolute value of the neutron flux is of paramount importance in reactor physics and other application on the nuclear field. Due to several corrections which should be done, such as radioactive decay of the produced nuclides, normalization factors between different irradiations, neutron spectrum perturbation, cross section behaviour and growing of the reactor power, among other factors, make the calculation of the neutron flux very cumbersome. the software FLUXO was developed to overcome these inconveniences. It is programmed in FORTRAN language, and was written to calculate the absolute flux of thermal, epithermal and fast neutrons, through the foil activation technique. The magnitude of this activation can be measured by a 4{pi} {beta}-{gamma} coincidence measurement or by gamma spectroscopy alone. The software calculates as well, the absolute activity of radioactive sources, and reactor-irradiated samples. (author)

  20. Graphical User Interface for Simplified Neutron Transport Calculations

    Energy Technology Data Exchange (ETDEWEB)

    Schwarz, Randolph; Carter, Leland L

    2011-07-18

    A number of codes perform simple photon physics calculations. The nuclear industry is lacking in similar tools to perform simplified neutron physics shielding calculations. With the increased importance of performing neutron calculations for homeland security applications and defense nuclear nonproliferation tasks, having an efficient method for performing simple neutron transport calculations becomes increasingly important. Codes such as Monte Carlo N-particle (MCNP) can perform the transport calculations; however, the technical details in setting up, running, and interpreting the required simulations are quite complex and typically go beyond the abilities of most users who need a simple answer to a neutron transport calculation. The work documented in this report resulted in the development of the NucWiz program, which can create an MCNP input file for a set of simple geometries, source, and detector configurations. The user selects source, shield, and tally configurations from a set of pre-defined lists, and the software creates a complete MCNP input file that can be optionally run and the results viewed inside NucWiz.

  1. Microscopic calculations and energy expansions for neutron-rich matter

    Energy Technology Data Exchange (ETDEWEB)

    Drischler, Christian; Soma, Vittorio [Institut fuer Kernphysik, Technische Universitaet Darmstadt (Germany); ExtreMe Matter Institute EMMI, GSI Helmholtzzentrum fuer Schwerionenforschung GmbH (Germany); Schwenk, Achim [ExtreMe Matter Institute EMMI, GSI Helmholtzzentrum fuer Schwerionenforschung GmbH (Germany); Institut fuer Kernphysik, Technische Universitaet Darmstadt (Germany)

    2014-07-01

    We investigate the properties of asymmetric nuclear matter with two- and three-nucleon interactions based on chiral effective field theory. Focusing on neutron-rich matter, we calculate the energy for different proton fractions and include estimates of the theoretical uncertainty. We use our ab-initio results to test the quadratic expansion around symmetric matter with the symmetry energy term, and confirm its validity for highly asymmetric systems. Our calculated energy densities are in remarkable agreement with an empirical parameterization, developed to interpolate between pure neutron and symmetric nuclear matter. These findings are very useful for astrophysical applications and for developing new equations of state.

  2. 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)

  3. New methods for neutron response calculations with MCNP

    Energy Technology Data Exchange (ETDEWEB)

    Hendricks, J.S. [Los Alamos National Lab., NM (United States). Applied Theoretical and Computational Physics Div.

    1997-05-01

    MCNP4B was released for international distribution in February, 1997. The author summarized the new MCNP4B features since the release of MCNP4A over three years earlier and compare some results. Then he describes new methods being developed for future code releases. The focus is methods and applications of ex-core neutron response calculations.

  4. Mcnp calculation of neutron scatter in the Main Bay of the Chadwick Building, NPL

    Energy Technology Data Exchange (ETDEWEB)

    Naismith, O.F.; Thomas, D.J.

    1996-02-01

    The Monte Carlo neutron transport code MCNP has been used to calculate the room and air scattered neutron component at 75 cm from a radionuclide source located at the center of the low-scatter area in the Chadwick Building, Bldg. 47, at National Physical Laboratory (NPL). This is the standard distance used for calibrating personal dosemeters, and the calculation provides information for correcting the response of dosemeters to the scattered radiation. Calculations were performed for both an Am-Be and a (252)Cf source. These measurements revealed that the model used for features within the low-scatter area needs to be refined for calculating scatter at distances further from the source than 75 cm.

  5. Quantum Monte Carlo calculations of two neutrons in finite volume

    CERN Document Server

    Klos, P; Tews, I; Gandolfi, S; Gezerlis, A; Hammer, H -W; Hoferichter, M; Schwenk, A

    2016-01-01

    Ab initio calculations provide direct access to the properties of pure neutron systems that are challenging to study experimentally. In addition to their importance for fundamental physics, their properties are required as input for effective field theories of the strong interaction. In this work, we perform auxiliary-field diffusion Monte Carlo calculations of the ground and first excited state of two neutrons in a finite box, considering a simple contact potential as well as chiral effective field theory interactions. We compare the results against exact diagonalizations and present a detailed analysis of the finite-volume effects, whose understanding is crucial for determining observables from the calculated energies. Using the L\\"uscher formula, we extract the low-energy S-wave scattering parameters from ground- and excited-state energies for different box sizes.

  6. Benchmarking Geant4 for spallation neutron source calculations

    Science.gov (United States)

    DiJulio, Douglas D.; Batkov, Konstantin; Stenander, John; Cherkashyna, Nataliia; Bentley, Phillip M.

    2016-09-01

    Geant4 is becoming increasingly used for radiation transport simulations of spallation neutron sources and related components. Historically, the code has seen little usage in this field and it is of general interest to investigate the suitability of Geant4 for such applications. For this purpose, we carried out Geant4 calculations based on simple spallation source geometries and also with the the European Spallation Source Technical Design Report target and moderator configuration. The results are compared to calculations performed with the Monte Carlo N- Particle extended code. The comparisons are carried out over the full spallation neutron source energy spectrum, from sub-eV energies up to thousands of MeV. Our preliminary results reveal that there is generally good agreement between the simulations using both codes. Additionally, we have also implemented a general weight-window generator for Geant4 based applications and present some results of the method applied to the ESS target model.

  7. OPAL REACTOR: Calculation/Experiment comparison of Neutron Flux Mapping in Flux Coolant Channels

    Energy Technology Data Exchange (ETDEWEB)

    Barbot, L.; Domergue, C.; Villard, J. F.; Destouches, C. [CEA, Paris (France); Braoudakis, G.; Wassink, D.; Sinclair, B.; Osborn, J. C.; Huayou, Wu [ANSTO, Syeney (Australia)

    2013-07-01

    The measurement and calculation of the neutron flux mapping of the OPAL research reactor are presented. Following an investigation of fuel coolant channels using sub-miniature fission chambers to measure thermal neutron flux profiles, neutronic calculations were performed. Comparison between calculation and measurement shows very good agreement.

  8. Influence of the target excitations on neutron cross section calculations

    Energy Technology Data Exchange (ETDEWEB)

    Cabezas, R.; Lubian, J. [Center for Applied Studies to Nuclear Development, Havana (Cuba)

    1994-12-31

    Considerable progress has been made in the use of the nuclear models and methods to derive physically meaningful parameters for model calculations. In this sense, we analyze the influence of the collective modes of excitation of the target nucleus on the cross section calculations in some medium atomic-weight nuclei, where spectroscopic studies show evidence of anharmonic vibrations, nonaxial deformations and so on. Coupled channel calculations for low-energy neutron inelastic scattering to collective states in {sup 48}Ti, {sup 54}Cr and {sup 62}Ni isotopes are made using the Davydov-Chaban model (DCM). It is shown, that the consideration of the structural features of the target nucleus excitation in the cross section calculations guarantees the agreement with the experimental measurements.

  9. Calculation of dosimetry parameters for fast neutron radiotherapy

    Energy Technology Data Exchange (ETDEWEB)

    Wells, A.H.

    1978-05-01

    A computer simulation of the interactions of 50 MeV d/sup +/ on Be and 42 MeV p/sup +/ on Be neutron spectra with ICRU muscle tissue and Shonka A-150 tissue equivalent plastic was performed to allow computation of the charged particle spectra that result. Nuclear data were obtained from the Evaluated Nuclear Data File (ENDF) whenever possible and from the Intranuclear Cascade and Evaporation models otherwise. The dosimetry parameters calculated are: the kerma ratio, K/sub A-150//K/sub tissue/; the energy required to form an ion pair, W; and the stopping power ratio, S/sub g//sup W/.

  10. Characterization of neutron beams for boron neutron capture therapy: in-air radiobiological dosimetry.

    Science.gov (United States)

    Yamamoto, Tetsuya; Matsumura, Akira; Yamamoto, Kazuyoshi; Kumada, Hiroaki; Hori, Naohiko; Torii, Yoshiya; Shibata, Yasushi; Nose, Tadao

    2003-07-01

    The survival curves and the RBE for the dose components generated in boron neutron capture therapy (BNCT) were determined separately in neutron beams at Japan Research Reactor No. 4. The surviving fractions of V79 Chinese hamster cells with or without 10B were obtained using an epithermal neutron beam (ENB), a mixed thermal-epithermal neutron beam (TNB-1), and a thermal (TNB-2) neutron beam; these beams were used or are planned for use in BNCT clinical trials. The cell killing effect of the neutron beam in the presence or absence of 10B was highly dependent on the neutron beam used and depended on the epithermal and fast-neutron content of the beam. The RBEs of the boron capture reaction for ENB, TNB-1 and TNB-2 were 4.07 +/- 0.22, 2.98 +/- 0.16 and 1.42 +/- 0.07, respectively. The RBEs 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 RBEs of the neutron and photon components were 1.22 +/- 0.16, 1.23 +/- 0.16, and 1.21 +/- 0.16 for ENB, TNB-1 and TNB-2, respectively. The approach to the experimental determination of RBEs outlined in this paper allows the RBE-weighted dose calculation for each dose component of the neutron beams and contributes to an accurate inter-beam comparison of the neutron beams at the different facilities employed in ongoing and planned BNCT clinical trials.

  11. 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)

  12. 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.)

  13. Calculation of delayed-neutron energy spectra in a QRPA-Hauser-Feshbach model

    Energy Technology Data Exchange (ETDEWEB)

    Kawano, Toshihiko [Los Alamos National Laboratory; Moller, Peter [Los Alamos National Laboratory; Wilson, William B [Los Alamos National Laboratory

    2008-01-01

    Theoretical {beta}-delayed-neutron spectra are calculated based on the Quasiparticle Random-Phase Approximation (QRPA) and the Hauser-Feshbach statistical model. Neutron emissions from an excited daughter nucleus after {beta} decay to the granddaughter residual are more accurately calculated than in previous evaluations, including all the microscopic nuclear structure information, such as a Gamow-Teller strength distribution and discrete states in the granddaughter. The calculated delayed-neutron spectra agree reasonably well with those evaluations in the ENDF decay library, which are based on experimental data. The model was adopted to generate the delayed-neutron spectra for all 271 precursors.

  14. 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)

  15. 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.)

  16. 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.

  17. Precise calculations in simulations of the interaction of low energy neutrons with nano-dispersed media

    Energy Technology Data Exchange (ETDEWEB)

    Artem’ev, V. A., E-mail: niitm@inbox.ru [Research Institute of Materials Technology (Russian Federation); Nezvanov, A. Yu. [Moscow State Industrial University (Russian Federation); Nesvizhevsky, V. V. [Institut Max von Laue—Paul Langevin (France)

    2016-01-15

    We discuss properties of the interaction of slow neutrons with nano-dispersed media and their application for neutron reflectors. In order to increase the accuracy of model simulation of the interaction of neutrons with nanopowders, we perform precise quantum mechanical calculation of potential scattering of neutrons on single nanoparticles using the method of phase functions. We compare results of precise calculations with those performed within first Born approximation for nanodiamonds with the radius of 2–5 nm and for neutron energies 3 × 10{sup -7}–10{sup -3} eV. Born approximation overestimates the probability of scattering to large angles, while the accuracy of evaluation of integral characteristics (cross sections, albedo) is acceptable. Using Monte-Carlo method, we calculate albedo of neutrons from different layers of piled up diamond nanopowder.

  18. Precise calculations in simulations of the interaction of low energy neutrons with nano-dispersed media

    Science.gov (United States)

    Artem'ev, V. A.; Nezvanov, A. Yu.; Nesvizhevsky, V. V.

    2016-01-01

    We discuss properties of the interaction of slow neutrons with nano-dispersed media and their application for neutron reflectors. In order to increase the accuracy of model simulation of the interaction of neutrons with nanopowders, we perform precise quantum mechanical calculation of potential scattering of neutrons on single nanoparticles using the method of phase functions. We compare results of precise calculations with those performed within first Born approximation for nanodiamonds with the radius of 2-5 nm and for neutron energies 3 × 10-7-10-3 eV. Born approximation overestimates the probability of scattering to large angles, while the accuracy of evaluation of integral characteristics (cross sections, albedo) is acceptable. Using Monte-Carlo method, we calculate albedo of neutrons from different layers of piled up diamond nanopowder.

  19. Development and calculation of an energy dependent normal brain tissue neutron RBE for evaluating neutron fields for BNCT.

    Science.gov (United States)

    Woollard, J E; Blue, T E; Gupta, N; Gahbauer, R A

    2001-06-01

    In Boron Neutron Capture Therapy (BNCT) of malignant brain tumors, the energy dependence of a clinically relevant Relative Biological Effectiveness (RBE) for epithermal neutrons, RBE(En), is important in neutron field design. In the first half of this paper, we present the development of an expression for the energy dependent normal-tissue RBE, RBE(En). We then calculate a reasonable estimate for RBE(En) for adult brain tissue. In the second half of the paper, two separate RBE expressions are developed, one for the RBE of the neutrons that interact in tissue via the 14N(n,p)14C reaction, denoted RBE(N), and one for the RBE of the neutrons which interact in tissue via the 1H(n,n')1H reaction, denoted RBE(H). The absorbed-dose-averaged values of these expressions are calculated for the neutron flux spectrum in phantom for the Brookhaven Medical Research Reactor (BMRR) epithermal neutron beam. The calculated values, [RBE(norm)N] = 3.4 and [RBE(norm)H] = 3.2, are within 6% of being equal, and support the use of equal values for RBEN and RBE(H) by researchers at Brookhaven National Laboratory (BNL). Finally, values of [RBE(norm)N] and [RBE(norm)H], along with the absorbed-dose-averaged RBE for brain, [RBE(norm)b], were calculated as a function of depth along the centerline of an ellipsoidal head phantom using flux spectra calculated for our Accelerator-Based Neutron Source (ABNS). These values remained essentially constant with depth, supporting the use of constant values for RBE, as is done at BNL.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-07-01

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

  1. Transport calculation of neutrons leaked to the surroundings of the facilities by the JCO criticality accident in Tokai-mura.

    Science.gov (United States)

    Imanaka, T

    2001-09-01

    A transport calculation of the neutrons leaked to the environment by the JCO criticality accident was carried out based on three-dimensional geometrical models of the buildings within the JCO territory. Our work started from an initial step to simulate the leakage process of neutrons from the precipitation tank, and proceeded to a step to calculate the neutron propagation throughout the JCO facilities. The total fission number during the accident in the precipitation tank was evaluated to be 2.5 x 10(18) by comparing the calculated neutron-induced activities per 235U fission with the measured values in a stainless-steel net sample taken 2 m from the precipitation tank. Shield effects by various structures within the JCO facilities were evaluated by comparing the present results with a previous calculation using two-dimensional models which suppose a point source of the fission spectrum in the air above the ground without any shield structures. The shield effect by the precipitation tank, itself, was obtained to be a factor of 3. The shield factor by the conversion building varied between 1.1 and 2, depending on the direction from the building. The shield effect by the surrounding buildings within the JCO territory was between I and 5, also depending on the direction.

  2. Calculations to support JET neutron yield calibration: Modelling of neutron emission from a compact DT neutron generator

    Science.gov (United States)

    Čufar, Aljaž; Batistoni, Paola; Conroy, Sean; Ghani, Zamir; Lengar, Igor; Milocco, Alberto; Packer, Lee; Pillon, Mario; Popovichev, Sergey; Snoj, Luka

    2017-03-01

    At the Joint European Torus (JET) the ex-vessel fission chambers and in-vessel activation detectors are used as the neutron production rate and neutron yield monitors respectively. In order to ensure that these detectors produce accurate measurements they need to be experimentally calibrated. A new calibration of neutron detectors to 14 MeV neutrons, resulting from deuterium-tritium (DT) plasmas, is planned at JET using a compact accelerator based neutron generator (NG) in which a D/T beam impinges on a solid target containing T/D, producing neutrons by DT fusion reactions. This paper presents the analysis that was performed to model the neutron source characteristics in terms of energy spectrum, angle-energy distribution and the effect of the neutron generator geometry. Different codes capable of simulating the accelerator based DT neutron sources are compared and sensitivities to uncertainties in the generator's internal structure analysed. The analysis was performed to support preparation to the experimental measurements performed to characterize the NG as a calibration source. Further extensive neutronics analyses, performed with this model of the NG, will be needed to support the neutron calibration experiments and take into account various differences between the calibration experiment and experiments using the plasma as a source of neutrons.

  3. Applying Advanced Neutron Transport Calculations for Improving Fuel Performance Codes

    Energy Technology Data Exchange (ETDEWEB)

    Botazzoli, P.; Luzzi, L. [Politecnico di Milano, Department of Energy, Nuclear Engineering Division - CeSNEF, Milano (Italy); Schubert, A.; Van Uffelen, P. [European Commission, Joint Research Centre, Institute for Transuranium Elements, Karlsruhe (Germany); Haeck, W. [Institute de Radioprotection et de Surete Nucleaire, Fontenay-aux-Roses (France)

    2009-06-15

    TRANSURANUS is a computer code for the thermal and mechanical analysis of fuel rods in nuclear reactors. As part of the code, the TUBRNP model calculates the local concentration of the actinides (U, Pu, Am, Cm), the main fission products (Xe, Kr, Cs and Nd) and {sup 4}He produced during the irradiation as a function of the radial position across a fuel pellet (radial profiles). These local quantities are required for the determination of the local power density, the local burn-up, and the source term of fission products and other inert gases. In previous works the neutronic code ALEPH has been used to validate the models for the actinides and fission products concentrations in UO{sub 2} fuels. A similar approach has been adopted in the present work for verifying the Helium production. The present paper focuses on the modelling of the Helium production in PWR oxide fuels (MOX and UO{sub 2}). A reliable prediction of the Helium production and release in LWR oxide fuels is of great interest in case of increasing burn-up, linear heat generation rates and Plutonium content. The contribution of the Helium released plays a fundamental role in the gap pressure and subsequently in the mechanical behaviour of the fuel rod, in particular during the storage of the high burn-up spent fuel. Helium is produced in oxide fuels by three main paths: (i) alpha decay of the actinides (the main contribution is due to {sup 242}Cm, {sup 238}Pu and {sup 244}Cm); (ii) (n,{alpha}) reactions; and (iii) ternary fission. In the present work, the contributions due to ternary fission and the (n,{alpha}) reaction on {sup 16}O as well as some refinements in the {sup 241}Am burn-up chain have been included in TUBRNP. The VESTA neutronic code has been used for the validation of the He production model. The generic VESTA Monte Carlo depletion interface developed at IRSN allows us to couple different Monte Carlo codes with a depletion module. It currently allows for combining the ORIGEN 2.2 isotope

  4. Calculation of Multisphere Neutron Spectrometer Response Functions in Energy Range up to 20 MeV

    CERN Document Server

    Martinkovic, J

    2005-01-01

    Multisphere neutron spectrometer is a basic instrument of neutron measurements in the scattered radiation field at charged-particles accelerators for radiation protection and dosimetry purposes. The precise calculation of the spectrometer response functions is a necessary condition of the propriety of neutron spectra unfolding. The results of the response functions calculation for the JINR spectrometer with LiI(Eu) detector (a set of 6 homogeneous and 1 heterogeneous moderators, "bare" detector within cadmium cover and without it) at two geometries of the spectrometer irradiation - in uniform monodirectional and uniform isotropic neutron fields - are given. The calculation was carried out by the code MCNP in the neutron energy range 10$^{-8}$-20 MeV.

  5. 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.

  6. Calculation of Prompt Fission Neutron Spectra for ~(235)U (n,f)

    Institute of Scientific and Technical Information of China (English)

    2011-01-01

    The prompt fission neutron spectra for neutron-induced fission of 235U at En<5 MeV are calculated using the nuclear evaporation theory with a semi-empirical model, in which the non-constant temperature and the constant temperature related to the Fermi gas model

  7. 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.

  8. A Neutron Burst Associated with an Extensive Air Shower?

    Science.gov (United States)

    Alves, Mauro; Martin, Inacio; Shkevov, Rumen; Gusev, Anatoly; De Abreu, Alessandro

    2016-07-01

    A portable and compact system based on a He-3 tube (LND, USA; model 25311) with an area of approximately 250 cm² and is used to record neutron count rates at ground level in the energy range of 0.025 eV to 10 MeV, in São José dos Campos, SP, Brazil (23° 12' 45" S, 45° 52' 00" W; altitude, 660m). The detector, power supply, digitizer and other hardware are housed in an air-conditioned room. The detector power supply and digitizer are not connected to the main electricity network; a high-capacity 12-V battery is used to power the detector and digitizer. Neutron counts are accumulated at 1-minute intervals continuously. The data are stored in a PC for further analysis. In February 8, 2015, at 12 h 22 min (local time) during a period of fair weather with minimal cloud cover (extensive air shower that occurred over the detector.

  9. Application of MCNP for neutronic calculations at VR-1 training reactor

    Science.gov (United States)

    Huml, Ondřej; Rataj, Jan; Bílý, Tomáš

    2014-06-01

    The paper presents utilization of Monte Carlo MCNP transport code for neutronic calculations of training reactor VR-1. Results of calculations are compared with results of measurements realized during last few critical experiments with various reactor core configurations. Very good agreement between calculations and measurements is observed.

  10. Monte Carlo calculation of skyshine'' neutron dose from ALS (Advanced Light Source)

    Energy Technology Data Exchange (ETDEWEB)

    Moin-Vasiri, M.

    1990-06-01

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

  11. A Preliminary Assessment of Radiation and Air Activation for the Neutron Science Facility in RAON

    Energy Technology Data Exchange (ETDEWEB)

    Yang, S. C.; Lee, C. W.; Lee, E. J.; Lee, Y. O. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Kim, J. C. [Institute for Basic Science, Daejeon (Korea, Republic of)

    2015-05-15

    The works will stay in the DAQ room during an operation for about 1 month. In order to test the characteristics of the detector, the workers are also possible to access the TOF hall after a shutdown. Therefore, the shielding analysis of the NSF is required to meet the above purpose. In view of this, we performed the calculation of the shielding concrete thickness required for a target room by using MCNPX code with a neutron source obtained from Institute for Basic Science (IBS). In addition, the dose distribution and air activation for the entire space in NSF were evaluated using MCNPX and SP-FISPACT 2010 codes. We have performed the shielding calculation with the neutron source produced from the C(d,n) reactions. The concrete thickness was evaluated for all directions of the target room, and it was confirmed by performing the calculation of dose distribution to the entire space. However, the dose rate for the beam line was high. The radioactivity of radionuclides at TOF hall do not exceeded the air concentration and release limits.

  12. Design basis neutronics calculations for NRU-LOCA experiments

    Energy Technology Data Exchange (ETDEWEB)

    Heaberlin, S.W.; Jenquin, U.P.; McNair, G.W.; Perry, R.T.; Trapp, T.J.; Zimmerman, M.G.

    1979-08-01

    The report describes the neutronics analysis for the LOCA simulation experiments in the NRU reactor. The experimental program will provide greater understanding of nuclear fuel assembly behavior during the heatup, reflood and quench sequence of a hypothetical LOCA. The decay heat and stored heat, which are the energy source in a LOCA will be simulated by fission heat provided by the NRU reactor. The reactor, the test and test operation are described.

  13. Cross correlation calculations and neutron scattering analysis for a portable solid state neutron detection system

    Science.gov (United States)

    Saltos, Andrea

    In efforts to perform accurate dosimetry, Oakes et al. [Nucl. Intrum. Mehods. (2013)] introduced a new portable solid state neutron rem meter based on an adaptation of the Bonner sphere and the position sensitive long counter. The system utilizes high thermal efficiency neutron detectors to generate a linear combination of measurement signals that are used to estimate the incident neutron spectra. The inversion problem associated to deduce dose from the counts in individual detector elements is addressed by applying a cross-correlation method which allows estimation of dose with average errors less than 15%. In this work, an evaluation of the performance of this system was extended to take into account new correlation techniques and neutron scattering contribution. To test the effectiveness of correlations, the Distance correlation, Pearson Product-Moment correlation, and their weighted versions were performed between measured spatial detector responses obtained from nine different test spectra, and the spatial response of Library functions generated by MCNPX. Results indicate that there is no advantage of using the Distance Correlation over the Pearson Correlation, and that weighted versions of these correlations do not increase their performance in evaluating dose. Both correlations were proven to work well even at low integrated doses measured for short periods of time. To evaluate the contribution produced by room-return neutrons on the dosimeter response, MCNPX was used to simulate dosimeter responses for five isotropic neutron sources placed inside different sizes of rectangular concrete rooms. Results show that the contribution of scattered neutrons to the response of the dosimeter can be significant, so that for most cases the dose is over predicted with errors as large as 500%. A possible method to correct for the contribution of room-return neutrons is also assessed and can be used as a good initial estimate on how to approach the problem.

  14. 3D neutronic calculations: CAD-MCNP methodology applied to vessel activation in KOYO-F

    Science.gov (United States)

    Herreras, Y.; Lafuente, A.; Sordo, F.; Cabellos, O.; Perlado, J. M.

    2008-05-01

    This paper presents a methodology for 3D neutronic calculations suitable for complex and extensive geometries. The geometry of the system design is first fully modelled with a CAD program, and subsequently processed through a MCNP-CAD interface in order to generate an MCNP geometry file. Neutronic irradiation results are finally achieved running the MCNPX program, where the geometry input card used is directly the MCNP-CAD interface output. This methodology enables accurate neutronic calculations for complex geometries characterised by high detail levels. This procedure will be applied to the Fast Ignition Fusion Reactor KOYO-F to determine first neutron fluxes calculations along the blanket as well as the material activation in the reduced martensitic 9Cr-1Mo steel vessel.

  15. 3D neutronic calculations: CAD-MCNP methodology applied to vessel activation in KOYO-F

    Energy Technology Data Exchange (ETDEWEB)

    Herreras, Y; Cabellos, O; Perlado, J M [Instituto de Fusion Nuclear (DENIM)/ETSII/Universidad Politecnica, Madrid (Spain); Lafuente, A; Sordo, F [Universidad Politecnica de Madrid (UPM), Madrid (Spain)], E-mail: yuri@denim.upm.es

    2008-05-15

    This paper presents a methodology for 3D neutronic calculations suitable for complex and extensive geometries. The geometry of the system design is first fully modelled with a CAD program, and subsequently processed through a MCNP-CAD interface in order to generate an MCNP geometry file. Neutronic irradiation results are finally achieved running the MCNPX program, where the geometry input card used is directly the MCNP-CAD interface output. This methodology enables accurate neutronic calculations for complex geometries characterised by high detail levels. This procedure will be applied to the Fast Ignition Fusion Reactor KOYO-F to determine first neutron fluxes calculations along the blanket as well as the material activation in the reduced martensitic 9Cr-1Mo steel vessel.

  16. Theoretical study and calculation of the response of a fast neutron dosemeter based on track detection

    Energy Technology Data Exchange (ETDEWEB)

    Decossas, J.L.; Vareille, J.C.; Moliton, J.P.; Teyssier, J.L. (Limoges Univ., 87 (France). Lab. d' Electronique des Polymeres sous Faisceaux Ioniques)

    1983-01-01

    A fast neutron dosemeter is generally composed of a radiator in which n-p elastic scattering occurs and a detector which registers protons. A theoretical study, and the calculation (FORTRAN program) of the response of such a dosemeter is presented involving two steps: 1) The proton flux emerging from a thick radiator on which monoenergetic neutrons are normally incident is studied. This is characterised by its energy spectrum depending on the neutron energy and on the radiator thickness. 2) Proton detection being achieved with a solid state nuclear track detector whose performance is known, the number of registered tracks are calculated. The dosemeter sensitivity (tracks cm/sup -2/. Sv/sup -1/) is deduced. Then, the calculations show that it is possible to optimise the radiator thickness to obtain the smallest variation in sensitivity with neutron energy. The theoretical results are in good agreement with the experimental ones found in the literature.

  17. Monte Carlo and deterministic computational methods for the calculation of the effective delayed neutron fraction

    Science.gov (United States)

    Zhong, Zhaopeng; Talamo, Alberto; Gohar, Yousry

    2013-07-01

    The effective delayed neutron fraction β plays an important role in kinetics and static analysis of the reactor physics experiments. It is used as reactivity unit referred to as "dollar". Usually, it is obtained by computer simulation due to the difficulty in measuring it experimentally. In 1965, Keepin proposed a method, widely used in the literature, for the calculation of the effective delayed neutron fraction β. This method requires calculation of the adjoint neutron flux as a weighting function of the phase space inner products and is easy to implement by deterministic codes. With Monte Carlo codes, the solution of the adjoint neutron transport equation is much more difficult because of the continuous-energy treatment of nuclear data. Consequently, alternative methods, which do not require the explicit calculation of the adjoint neutron flux, have been proposed. In 1997, Bretscher introduced the k-ratio method for calculating the effective delayed neutron fraction; this method is based on calculating the multiplication factor of a nuclear reactor core with and without the contribution of delayed neutrons. The multiplication factor set by the delayed neutrons (the delayed multiplication factor) is obtained as the difference between the total and the prompt multiplication factors. Using Monte Carlo calculation Bretscher evaluated the β as the ratio between the delayed and total multiplication factors (therefore the method is often referred to as the k-ratio method). In the present work, the k-ratio method is applied by Monte Carlo (MCNPX) and deterministic (PARTISN) codes. In the latter case, the ENDF/B nuclear data library of the fuel isotopes (235U and 238U) has been processed by the NJOY code with and without the delayed neutron data to prepare multi-group WIMSD neutron libraries for the lattice physics code DRAGON, which was used to generate the PARTISN macroscopic cross sections. In recent years Meulekamp and van der Marck in 2006 and Nauchi and Kameyama

  18. Gamow's calculation of the neutron star's critical mass revised

    Energy Technology Data Exchange (ETDEWEB)

    Ludwig, Hendrik; Ruffini, Remo [Sapienza Universita di Roma, Rome (Italy); ICRANet, University of Nice-Sophia Antipolis, Nice Cedex (France)

    2014-09-15

    It has at times been indicated that Landau introduced neutron stars in his classic paper of 1932. This is clearly impossible because the discovery of the neutron by Chadwick was submitted more than one month after Landau's work. Therefore, and according to his calculations, what Landau really did was to study white dwarfs, and the critical mass he obtained clearly matched the value derived by Stoner and later by Chandrasekhar. The birth of the concept of a neutron star is still today unclear. Clearly, in 1934, the work of Baade and Zwicky pointed to neutron stars as originating from supernovae. Oppenheimer in 1939 is also well known to have introduced general relativity (GR) in the study of neutron stars. The aim of this note is to point out that the crucial idea for treating the neutron star has been advanced in Newtonian theory by Gamow. However, this pioneering work was plagued by mistakes. The critical mass he should have obtained was 6.9 M, not the one he declared, namely, 1.5 M. Probably, he was taken to this result by the work of Landau on white dwarfs. We revise Gamow's calculation of the critical mass regarding calculational and conceptual aspects and discuss whether it is justified to consider it the first neutron-star critical mass. We compare Gamow's approach to other early and modern approaches to the problem.

  19. 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.

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

    CERN Document Server

    2002-01-01

    This report condenses basic theories and advanced applications of neutron/gamma ray transport calculations in many fields of nuclear energy research. Chapters 1 through 5 treat historical progress of Monte Carlo methods, general issues of variance reduction technique, cross section libraries used in continuous energy Monte Carlo codes. In chapter 6, the following issues are discussed: fusion benchmark experiments, design of ITER, experiment analyses of fast critical assembly, core analyses of JMTR, simulation of pulsed neutron experiment, core analyses of HTTR, duct streaming calculations, bulk shielding calculations, neutron/gamma ray transport calculations of the Hiroshima atomic bomb. Chapters 8 and 9 treat function enhancements of MCNP and MVP codes, and a parallel processing of Monte Carlo calculation, respectively. An important references are attached at the end of this report.

  1. 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)

  2. 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)

  3. Anharmonic effects in neutron cross-section calculation for nuclei in mass range 48 [<=] A [<=] 58

    Energy Technology Data Exchange (ETDEWEB)

    Lubian, J.; Cabezas, R. (Center for Applied Studies to Nuclear Development, Havana (Cuba))

    1993-08-01

    In this paper, a deviation of the target nucleus wavefunction from the harmonic vibrator in the neutron scattering process by medium-mass nuclei at low energies is studied. Two forms of anharmonicities are used: anharmonicities due to the higher-order terms in the Hamiltonians and those due to the different deformation parameters, corresponding to transitions between nuclear states. For calculation of neutron cross sections, combined use of the coupled-channel method and the statistical Hauser-Feshbach-Moldauer theory is applied. It is shown that both kinds of anharmonicities introduced a correction (about 10% in some cases) to the neutron cross sections at low energies. (author).

  4. 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.

  5. 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.

  6. 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)

    1997-04-01

    Neutron fluence measurements and radiation damage calculations are reported for the joint U.S. 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 midplanes was 4.4E+22 n/cm{sup 2} resulting in about 9.0 dpa in type 316 stainless steel.

  7. Neutron dosimetry and damage calculations for the EBRII COBRA-1A irradiations

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-04-01

    Neutron fluence measurements and radiation damage calculations are reported for the joint U.S. and Japanese COBRA-1A1 and 1A2 irradiations in the Experimental Breeder Reactor II. The maximum total neutron fluences at midplane were 2.0E+22 and 7.5E+22 n/cm{sup 2}, for the 1A1 and 1A2 irradiations, respectively, resulting in about 8.0 and 30.3 dpa in stainless steel.

  8. Calculation verification of the utilization of LR-0 for reference neutron spectra

    Science.gov (United States)

    Ján, Milčák; Michal, Košťál; Marie, Švadlenková; Michal, Koleška; Vojtěch, Rypar

    2014-11-01

    Well-defined neutron spectrum is crucial for calibration and testing of detectors for spectrometry and dosimetry purposes. As a possible source of neutrons nuclear reactors can be utilized. In reactor core most of the neutrons are originated from fission and neutron spectra is usually some form of moderated spectra of fast neutrons. The reactor LR-0 is an experimental light-water zero-power pool-type reactor originally designed for research of the VVER type reactor cores, spent-fuel storage lattices and benchmark experiments. The main reactor feature that influences the performance of experiments is the flexible arrangement of the core. Special types of the possible core arrangements on the reactor LR-0 can provide different neutron spectra in special experimental channels. These neutron spectra are modified by inserting different materials around the channel and whole core is driven by standard fuel assemblies. Fast, epithermal or thermal spectra can be simulated using graphite, H2O, D2O insertions, air, Cd foils or fuel with different enrichment.

  9. Xenon poisoning calculation code for miniature neutron source reactor (MNSR)

    Institute of Scientific and Technical Information of China (English)

    2001-01-01

    In line with the actual requirements and based upon the specific char acteristics of MNSR, a revised point-reactor model was adopted to model MNSR's xenon poisoning. The corresponding calculation code, MNSRXPCC (Xenon Poison ing Calculation Code for MNSR), was developed and tested by the Shanghai MNSR data.

  10. Coupled cluster calculations of neutron-rich nuclei

    Science.gov (United States)

    Hagen, Gaute

    2016-09-01

    In this talk I will present recent highlights from ab initio computations of atomic nuclei using coupled-cluster methods with state-of-the-art interactions from chiral effective field theory (EFT). The recent progress in computing nuclei from scratch is based on new optimizations of interactions from chiral EFT, and ab initio methods with a polynomial computational cost together with available super computing resources. The physics advancements I will discuss include: (i) accurate nuclear binding energies and radii of light and medium-mass nuclei, (ii) the neutron distribution and electric dipole polarizability of the nucleus 48Ca, (iii) and the structure of the rare nucleus 78Ni from first principles. All these quantities are currently targeted by precision measurements worldwide.

  11. The use of the neutronic calculation code CORNER for evaluating the protection of fast neutron reactor and CNFC equipment

    Science.gov (United States)

    Shekhanova, M. E.

    2017-01-01

    In this paper we propose a method of using neutronic calculation code CORNER to the analysis of experiments on the protection of fast neutron reactor and CNFC equipment. An example of Winfrith Graphite Benchmark experiment calculation using this approach is presented. This task can be considered as one step in the general theme of the safety analysis of FR with liquid metal coolant, their fuel cycles and related equipment. CORNER implement a solution of the kinetic equation with a source in the three-dimensional hexagonal geometry based on Sn-method. The purpose of this paper is a demonstration of the application of CORNER’s possibilities for the analysis of the actual reactor problems.

  12. Microscopic Calculations of Vortex-Nucleus Interaction in the Neutron Star Crust

    CERN Document Server

    Sekizawa, Kazuyuki; Magierski, Piotr; Bulgac, Aurel; Forbes, Michael McNeil

    2016-01-01

    We investigate the dynamics of a quantized vortex and a nuclear impurity immersed in a neutron superfluid within a fully microscopic time-dependent three-dimensional approach. The magnitude and even the sign of the force between the quantized vortex and the nuclear impurity have been a matter of debate for over four decades. We determine that the vortex and the impurity repel at neutron densities, 0.014 fm$^{-3}$ and 0.031 fm$^{-3}$, which are relevant to the neutron star crust and the origin of glitches, while previous calculations have concluded that the force changes its sign between these two densities and predicted contradictory signs. The magnitude of the force increases with the density of neutron superfluid, while the magnitude of the pairing gap decreases in this density range.

  13. Calculation of Prompt Fission Neutron Spectrum for 233U(n, f) Reaction by Semi-empirical Method

    Institute of Scientific and Technical Information of China (English)

    CHEN; Yong-jing; LIU; Ting-jin; SHU; Neng-chuan

    2013-01-01

    The prompt fission neutron spectra for neutron-induced fission of 233U for low energy neutron(below 6 MeV)are calculated using the nuclear evaporation theory with a semi-empirical method,in which the partition of the total excitation energy between the fission fragments for the nth+233U fission

  14. Measurement of neutron attenuation through thick shields and comparison with calculation

    Energy Technology Data Exchange (ETDEWEB)

    Bull, J.S.; Donahue, J.B.; Burman, R.L.

    1998-12-31

    The large neutrino experiments conducted over the last several years at the Los Alamos Neutron Science Center (LANSCE) have provided the opportunity to measure the effects of neutron attenuation in very thick shields. These experiments have featured detectors with active masses of 6 to 150 tons and shield thicknesses ranging from 3000 to 5280 g/cm{sup 2}. An absolute measurement of the high-energy neutron flux was made from the beam stop in a neutrino cave at ninety degrees and nine meters from the beam stop. Differential neutron shielding measurements in iron were also performed, resulting in an attenuation length of 148 g/cm{sup 2}. These measurements allow for the testing of radiation shielding codes for deep penetration problems. The measured flux and attenuation length is compared to calculations using the LAHET Code System (LCS). These codes incorporate biasing techniques, allowing for direct calculation of deep penetration shielding problems. Calculations of the neutron current and attenuation length are presented and compared with measured values. Results from the shielding codes show good agreement with the measured values.

  15. Structural uncertainty in air mass factor calculation for NO

    NARCIS (Netherlands)

    Lorente Delgado, Alba; Folkert Boersma, K.; Yu, Huan; Dörner, Steffen; Hilboll, Andreas; Richter, Andreas; Liu, Mengyao; Lamsal, Lok N.; Barkley, Michael; Smedt, De Isabelle; Roozendael, Van Michel; Wang, Yang; Wagner, Thomas; Beirle, Steffen; Lin, Jin Tai; Krotkov, Nickolay; Stammes, Piet; Wang, Ping; Eskes, Henk J.; Krol, Maarten

    2017-01-01

    Air mass factor (AMF) calculation is the largest source of uncertainty in NO2 and HCHO satellite retrievals in situations with enhanced trace gas concentrations in the lower troposphere. Structural uncertainty arises when different retrieval methodologies are applied within the scientific community

  16. Calculation of the reactor neutron time of flight spectrum by convolution technique

    Institute of Scientific and Technical Information of China (English)

    Cheng Jin-Xing; Ouyang Xiao-Ping; Zheng Yi; Zhang An-Hui; Ouyang Mao-Jie

    2008-01-01

    It is a very complex and tlme-consuming process to simulate the nuclear reactor neutron spectrum from the reactor core to the export channel by applying a Monte Carlo program. This paper presents a new method to calculate the neutron spectrum by using the convolution technique which considers the channel transportation as a linear system and the transportation scattering as the response function. It also applies Monte Carlo Neutron and Photon Transport Code (MCNP) to simulate the response function numerically. With the application of convolution technique to calculate thespectrum distribution from the core to the channel, the process is then much more convenient only with the simple numerical integral numeration. This saves computer time and reduces some trouble in re-writing of the MCNP program.

  17. 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.)

  18. Calculation of prompt fission neutron spectra for 235U(n,f)

    Institute of Scientific and Technical Information of China (English)

    CHEN Yong-Jing; JIA Min; TAO Xi; QIAN Jing; LIU Ting-Jin; SHU Neng-Chuan

    2012-01-01

    The prompt fission neutron spectra for the neutron-induced fission of 235U at En < 5 MeV are calculated using nuclear evaporation theory with a semi-empirical model,in which the nonconstant and constant temperatures related to the Fermi gas model are taken into account. The calculated prompt fission neutron spectra reproduce the experimental data well.For the n(thermal)+235U reaction,the average nuclear temperature of the fission fragment,and the probability distribution of the nuclear temperature,are discussed and compared with the Los Alamos model.The energy carried away by γ rays emitted from each fragment is also obtained and the results are in good agreement with the existing experimental data.

  19. 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.

  20. The analysis of thermal calculation for air stove drying system

    Directory of Open Access Journals (Sweden)

    Li Xue

    2012-08-01

    Full Text Available This article discusses the existing calculation of heat for a coal-fired hot-blast furnace. By utilizing the standard method of heat calculation for boilers, considering the relation between the theoretical combustion temperature and the excess air coefficient of the boiler, combining some operational parameters of a coal-fired powder hot-blast furnace, the heat calculation of iron ore concentrating dry combustion on a coal-fired hot stove is discussed. It is used to prevent coke and optimize combustion. It also discusses the advantages and disadvantages of flue gas recirculation systems. The conclusion will show the practical applications of this.

  1. Accelerator-driven sub-critical research facility with low-enriched fuel in lead matrix: Neutron flux calculation

    Directory of Open Access Journals (Sweden)

    Avramović Ivana

    2007-01-01

    Full Text Available The H5B is a concept of an accelerator-driven sub-critical research facility (ADSRF being developed over the last couple of years at the Vinča Institute of Nuclear Sciences, Belgrade, Serbia. Using well-known computer codes, the MCNPX and MCNP, this paper deals with the results of a tar get study and neutron flux calculations in the sub-critical core. The neutron source is generated by an interaction of a proton or deuteron beam with the target placed inside the sub-critical core. The results of the total neutron flux density escaping the target and calculations of neutron yields for different target materials are also given here. Neutrons escaping the target volume with the group spectra (first step are used to specify a neutron source for further numerical simulations of the neutron flux density in the sub-critical core (second step. The results of the calculations of the neutron effective multiplication factor keff and neutron generation time L for the ADSRF model have also been presented. Neutron spectra calculations for an ADSRF with an uranium tar get (highest values of the neutron yield for the selected sub-critical core cells for both beams have also been presented in this paper.

  2. 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)

  3. 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. Th

  4. 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 r(1

  5. 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)

  6. Monte Carlo calculation of the neutron and gamma sensitivities of self-powered detectors

    Energy Technology Data Exchange (ETDEWEB)

    Pytel, K.

    1981-01-01

    A calculational model is presented for the self-powered detector response prediction in various radiation environments. The fast beta particles and electron transport is treated by Monte Carlo technique. A new model of electronic processes within the insulator is introduced. Calculated neutron and gamma sensitivities of five detectors (with Rh, V, Co, Ag and Pt emitters) are compared with reported experimental values. The comparison gives a satisfactory agreement for the majority of examined detectors.

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

    Energy Technology Data Exchange (ETDEWEB)

    Smith, L.A.; Gallmeier, F.X. [Oak Ridge Institute for Science and Energy, TN (United States); Gehin, J.C. [Oak Ridge National Lab., TN (United States)] [and others

    1995-05-01

    The FOEHN critical experiment was analyzed to validate the use of multigroup cross sections and Oak Ridge National Laboratory neutronics computer codes in the design of the Advanced Neutron Source. The ANSL-V 99-group master cross section library was used for all the calculations. Three different critical configurations were evaluated using the multigroup KENO Monte Carlo transport code, the multigroup DORT discrete ordinates transport code, and the multigroup diffusion theory code VENTURE. The simple configuration consists of only the fuel and control elements with the heavy water reflector. The intermediate configuration includes boron endplates at the upper and lower edges of the fuel element. The complex configuration includes both the boron endplates and components in the reflector. Cross sections were processed using modules from the AMPX system. Both 99-group and 20-group cross sections were created and used in two-dimensional models of the FOEHN experiment. KENO calculations were performed using both 99-group and 20-group cross sections. The DORT and VENTURE calculations were performed using 20-group cross sections. Because the simple and intermediate configurations are azimuthally symmetric, these configurations can be explicitly modeled in R-Z geometry. Since the reflector components cannot be modeled explicitly using the current versions of these codes, three reflector component homogenization schemes were developed and evaluated for the complex configuration. Power density distributions were calculated with KENO using 99-group cross sections and with DORT and VENTURE using 20-group cross sections. The average differences between the measured values and the values calculated with the different computer codes range from 2.45 to 5.74%. The maximum differences between the measured and calculated thermal flux values for the simple and intermediate configurations are {approx} 13%, while the average differences are < 8%.

  8. Calculation of thermal neutron self-shielding correction factors for aqueous bulk sample prompt gamma neutron activation analysis using the MCNP code

    Energy Technology Data Exchange (ETDEWEB)

    Nasrabadi, M.N. [Department of Physics, Faculty of Science, University of Kashan, Km. 6, Ravand Road, Kashan (Iran, Islamic Republic of)], E-mail: mnnasri@kashanu.ac.ir; Jalali, M. [Isfahan Nuclear Science and Technology Research Institute, Atomic Energy organization of Iran (Iran, Islamic Republic of); Mohammadi, A. [Department of Physics, Faculty of Science, University of Kashan, Km. 6, Ravand Road, Kashan (Iran, Islamic Republic of)

    2007-10-15

    In this work thermal neutron self-shielding in aqueous bulk samples containing neutron absorbing materials is studied using bulk sample prompt gamma neutron activation analysis (BSPGNAA) with the MCNP code. The code was used to perform three dimensional simulations of a neutron source, neutron detector and sample of various material compositions. The MCNP model was validated against experimental measurements of the neutron flux performed using a BF{sub 3} detector. Simulations were performed to predict thermal neutron self-shielding in aqueous bulk samples containing neutron absorbing solutes. In practice, the MCNP calculations are combined with experimental measurements of the relative thermal neutron flux over the sample's surface, with respect to a reference water sample, to derive the thermal neutron self-shielding within the sample. The proposed methodology can be used for the determination of the elemental concentration of unknown aqueous samples by BSPGNAA where knowledge of the average thermal neutron flux within the sample volume is required.

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

    Energy Technology Data Exchange (ETDEWEB)

    Masukawa, Fumihiro; Nakano, Hideo; Nakashima, Hiroshi; Sasamoto, Nobuo [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Tayama, Ryu-ichi; Handa, Hiroyuki; Hayashi, Katsumi [Hitachi Engineering Co., Ltd., Hitachi, Ibaraki (Japan); Hirayama, Hideo [High Energy Accelerator Research Organization, Tsukuba, Ibaraki (Japan); Shin, Kazuo [Kyoto Univ., Kyoto (Japan)

    2003-03-01

    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. (author)

  10. Monte-Carlo calculation of the response functions for two prototype cosmic neutron metrology instruments

    Energy Technology Data Exchange (ETDEWEB)

    Benson, Chris; Joyce, Malcolm J.; Winsby, Andrew [Lancaster University, Engineering Department, Bailrigg, Lancaster LA1 4YR (United Kingdom); Silvie, Jon [BAE SYSTEMS, Barrow-in-Furness, LA14 1AF (United Kingdom)

    2002-08-01

    The response functions for two cosmic neutron detection systems have been calculated using Monte-Carlo computational methods. The detection systems that form the focus of this research are modified Leake detector designs in which a central thermal neutron detector is surrounded by a sphere of high-density polyethylene. In this arrangement, the surrounding polyethylene moderates the incident fast neutrons that are then detected by the central detector; in this case a {sup 3}He-filled proportional counter. In order to extend the response of these detector systems to cater for cosmic neutron environments, a shell of high-Z material has been included in each to promote (n, xn) reactions in the polyethylene moderator. We have used shells of lead and copper for this purpose to bring the high-energy component of the cosmic field, extending up to several GeV, within the capability of the detector systems. In particular, copper has been used in comparison with lead since the former is easier and safer to machine and handle. The overall diameter of the instruments studies in this work is 208 mm. Calculations of the neutron response have been performed with MCNP4C, for the thermal-20 MeV energy range, and with MCNPX 2.1.5/LA150N neutron libraries for the higher-energy cosmic region of the spectrum beyond 20 MeV. The results of these calculations are compared with experimental data that have been recorded with the instruments at the CERN Cosmic Reference Field Facility (CERF), Geneva, Switzerland. This comparison is discussed in respect of the likely applications of these detector systems to high-energy neutron field measurement on-board aircraft and in the vicinity of high-energy particle accelerators. The former application is gaining considerable research attention following the revised estimates of relative biological effectiveness of cosmic neutron fields and the related recommendation that aircrew be regarded occupationally-exposed radiation workers, on behalf of the

  11. Neutron dosimetry and damage calculations for the HFIR-JP-20 irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Greenwood, L.R. [Pacific Northwest National Lab., Richland, WA (United States); Baldwin, C.A. [Oak Ridge National Lab., TN (United States)

    1998-03-01

    Neutron fluence measurements and radiation damage calculations are reported for the joint US-Japanese experiment JP-20, which was conducted in a target position of the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL). The maximum total neutron fluence at midplane was 4.2 {times} 10{sup 22} n/cm{sup 2} (1.0 {times} 10{sup 22} n/cm{sup 2} above 0.1 MeV), resulting in about 8.4 dpa and 388 appm helium in type 316 stainless steel.

  12. Neutron dosimetry and damage calculations for the HFIR-JP-9, -12, and -15 irradiations

    Energy Technology Data Exchange (ETDEWEB)

    Greenwood, L.R. [Pacific Northwest National Lab., Richland, WA (United States); Baldwin, C.A. [Oak Ridge National Lab., TN (United States)

    1998-03-01

    Neutron fluence measurements and radiation damage calculations are reported for the joint US-Japanese experiments JP-9, -12, and -15. These experiments were conducted in target positions of the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL) for a period of nearly four years. The maximum neutron fluence at midplane was 2.6 {times} 10{sup 23} n/cm{sup 2} (7.1 {times} 10{sup 22} n/cm{sup 2} above 0.1 MeV), resulting in about 60 dpa and 3900 appm helium in type 316 stainless steel.

  13. Neutron dosimetry and damage calculations for the HFIR-MFE-200J-1 irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Greenwood, L.R. [Pacific Northwest National Lab., Richland, WA (United States); Baldwin, C.A. [Oak Ridge National Lab., TN (United States)

    1998-03-01

    Neutron fluence measurements and radiation damage calculations are reported for the joint US-Japanese experiment MFE-200-J-, which was conducted in the removable beryllium (RB) position of the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL). The maximum neutron fluence at midplane was 4.1 {times} 10{sup 22} n/cm{sup 2} (1.9 {times} 10{sup 22} n/cm{sup 2} above 0.1 MeV), resulting in about 12 dpa and 28 appm helium in type 316 stainless steel.

  14. Biological shielding assessment and dose rate calculation for a neutron inspection portal

    Science.gov (United States)

    Donzella, A.; Bonomi, G.; Giroletti, E.; Zenoni, A.

    2012-04-01

    With reference to the prototype of neutron inspection portal built and successfully tested in the Rijeka seaport (Croatia) within the EURITRACK (EURopean Illicit Trafficking Countermeasures Kit) project, an assessment of the biological shielding in different set-up configurations of a future portal has been calculated with MCNP Monte Carlo code in the frame of the Eritr@C (European Riposte against Illicit TR@ffiCking) project. In the configurations analyzed the compliance with the dose limits for workers and the population stated by the European legislation is provided by appropriate shielding of the neutron sources and by the delimitation of a controlled area.

  15. Measured and calculated fast neutron spectra in a depleted uranium and lithium hydride shielded reactor

    Science.gov (United States)

    Lahti, G. P.; Mueller, R. A.

    1973-01-01

    Measurements of MeV neutron were made at the surface of a lithium hydride and depleted uranium shielded reactor. Four shield configurations were considered: these were assembled progressively with cylindrical shells of 5-centimeter-thick depleted uranium, 13-centimeter-thick lithium hydride, 5-centimeter-thick depleted uranium, 13-centimeter-thick lithium hydride, 5-centimeter-thick depleted uranium, and 3-centimeter-thick depleted uranium. Measurements were made with a NE-218 scintillation spectrometer; proton pulse height distributions were differentiated to obtain neutron spectra. Calculations were made using the two-dimensional discrete ordinates code DOT and ENDF/B (version 3) cross sections. Good agreement between measured and calculated spectral shape was observed. Absolute measured and calculated fluxes were within 50 percent of one another; observed discrepancies in absolute flux may be due to cross section errors.

  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 matter from chiral two- and three-nucleon calculations up to N$^3$LO

    CERN Document Server

    Drischler, C; Hebeler, K; Schwenk, A

    2016-01-01

    Neutron matter is an ideal laboratory for nuclear interactions derived from chiral effective field theory since all contributions are predicted up to next-to-next-to-next-to-leading order (N$^3$LO) in the chiral expansion. By making use of recent advances in the partial-wave decomposition of three- nucleon (3N) forces, we include for the first time N$^3$LO 3N interactions in many-body perturbation theory (MBPT) up to third order and in self-consistent Green's function theory (SCGF). Using these two complementary many-body frameworks we provide improved predictions for the equation of state of neutron matter at zero temperature and also analyze systematically the many-body convergence for different chiral EFT interactions. Furthermore, we present an extension of the normal-ordering framework to finite temperatures. These developments open the way to improved calculations of neutron-rich matter including estimates of theoretical uncertainties for astrophysical applications.

  18. Neutron effective dose calculation behind concrete shielding of charged particle accelerators with energy up to 100 MeV

    CERN Document Server

    Alejnikov, V E; Krylov, A R

    2002-01-01

    Calculation data of neutron effective dose behind concrete shielding with thickness up to 3 meters is presented. The calculations have been performed by the Monte Carlo and phenomenological methods for monoenergetic neutrons with energy from 5 to 100 MeV as well as for neutron spectra produced by protons with energies of 30 and 72 MeV in thick targets. Comparison between calculations of neutron effective dose behind shielding using phenomenological approach and those by the Monte Carlo method normally shows agreement to within a factor of better than two, i.e. estimation of shielding thickness by those methods shall not exceed one half value layer of neutron effective dose attenuation in shielding. It amounts from 10 to 30 cm of concrete shielding for neutron energies and thickness of shields under consideration

  19. 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 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...

  20. Multi-modal calculations of prompt fission neutrons from 238U(n, f) at low induced energy

    Institute of Scientific and Technical Information of China (English)

    ZHENG Na; ZHONG Chun-Lai; FAN Tie-Shuan

    2011-01-01

    Properties of prompt fission neutrons from 238U(n,f) are calculated for incident neutron energies below 6 MeV using the multi-modal model,including the prompt fission neutron spectrum,the average prompt fission neutron multiplicity,and the prompt fission neutron multiplicity as a function of the fission fragment mass v(A) (usually named “sawtooth” data) The three most dominant fission modes are taken into account.The model parameters are determined on the basis of experimental fission fragment data.The predicted results are in good agreement with the experimental data.

  1. Experimental and calculated calibration of ionization chambers with air circulation

    CERN Document Server

    Peetermans, A

    1972-01-01

    The reports describes the method followed in order to calibrate the different ionization chambers with air circulation, used by the 'Health Physics Group'. The calculations agree more precisely with isotopes cited previously (/sup 11/C, /sup 13/N, /sup 15/O, /sup 41 /Ar, /sup 14/O, /sup 38/Cl) as well as for /sup 85/Kr, /sup 133/Xe, /sup 14/C and tritium which are used for the experimental standardisation of different chambers.

  2. Neutron Flux and Activation Calculations for a High Current Deuteron Accelerator

    CERN Document Server

    Coniglio, Angela; Sandri, Sandro

    2005-01-01

    Neutron analysis of the first Neutral Beam (NB) for the International Thermonuclear Experimental Reactor (ITER) was performed to provide the basis for the study of the following main aspects: personnel safety during normal operation and maintenance, radiation shielding design, transportability of the NB components in the European countries. The first ITER NB is a medium energy light particle accelerator. In the scenario considered for the calculation the accelerated particles are negative deuterium ions with maximum energy of 1 MeV. The average beam current is 13.3 A. To assess neutron transport in the ITER NB structure a mathematical model of the components geometry was implemented into MCNP computer code (MCNP version 4c2. "Monte Carlo N-Particle Transport Code System." RSICC Computer Code Collection. June 2001). The neutron source definition was outlined considering both D-D and D-T neutron production. FISPACT code (R.A. Forrest, FISPACT-2003. EURATOM/UKAEA Fusion, December 2002) was used to assess neutron...

  3. Calculation of kinetic parameters of Caliban metallic core experimental reactor from stochastic neutron measurements

    Energy Technology Data Exchange (ETDEWEB)

    Casoli, P.; Authier, N.; Baud, J. [Commissariat a l' energie Atomique, Centre de Valduc, 21120 Is-sur-Tille (France)

    2009-07-01

    Several experimental devices are operated by the Criticality and Neutron Science Research Department of the CEA Valduc Laboratory. One of these is the metallic core reactor Caliban. The knowledge of the fundamental kinetic parameters of the reactor is very useful, indeed necessary, to the operator. The purpose of this study was to develop and perform experiments allowing to determinate some of these parameters. The prompt neutron decay constant and particularly its value at criticality can be measured with reactor noise techniques such as the interval-distribution, the Feynman variance-to-mean, and the Rossi-{alpha} methods. By introducing the Nelson number, the effective delayed neutron fraction and the average neutron lifetime can also be calculated with the Rossi-{alpha} method. Subcritical, critical, and even supercritical experiments were performed. With the Rossi-{alpha} technique, it was found that the prompt neutron decay constant at criticality was (6.02*10{sup 5} {+-} 9%). Experiments also brought out the limitations of the used experimental parameters. (authors)

  4. Reference dosimetry calculations for neutron capture therapy with comparison of analytical and voxel models.

    Science.gov (United States)

    Goorley, J T; Kiger, W S; Zamenhof, R G

    2002-02-01

    As clinical trials of Neutron Capture Therapy (NCT) are initiated in the U.S. and other countries, new treatment planning codes are being developed to calculate detailed dose distributions in patient-specific models. The thorough evaluation and comparison of treatment planning codes is a critical step toward the eventual standardization of dosimetry, which, in turn, is an essential element for the rational comparison of clinical results from different institutions. In this paper we report development of a reference suite of computational test problems for NCT dosimetry and discuss common issues encountered in these calculations to facilitate quantitative evaluations and comparisons of NCT treatment planning codes. Specifically, detailed depth-kerma rate curves were calculated using the Monte Carlo radiation transport code MCNP4B for four different representations of the modified Snyder head phantom, an analytic, multishell, ellipsoidal model, and voxel representations of this model with cubic voxel sizes of 16, 8, and 4 mm. Monoenergetic and monodirectional beams of 0.0253 eV, 1, 2, 10, 100, and 1000 keV neutrons, and 0.2, 0.5, 1, 2, 5, and 10 MeV photons were individually simulated to calculate kerma rates to a statistical uncertainty of neutron beam with a broad neutron spectrum, similar to epithermal beams currently used or proposed for NCT clinical trials, was computed for all models. The thermal neutron, fast neutron, and photon kerma rates calculated with the 4 and 8 mm voxel models were within 2% and 4%, respectively, of those calculated for the analytical model. The 16 mm voxel model produced unacceptably large discrepancies for all dose components. The effects from different kerma data sets and tissue compositions were evaluated. Updating the kerma data from ICRU 46 to ICRU 63 data produced less than 2% difference in kerma rate profiles. The depth-dose profile data, Monte Carlo code input, kerma factors, and model construction files are available

  5. 中子剂量测量及估算方法%The measurement and calculation method for neutron dose

    Institute of Scientific and Technical Information of China (English)

    向剑; 戴光复; 苑淑渝; 丁艳秋; 张良安

    2008-01-01

    Company with the development of science,the neutron is used more and more widely,for example,neutron therapy cancer,neutron logging,neutron photograph and so on.The most wide application on medical treatment with neutron is boron neutron capture therapy.But it also brings some problems when it is in use.When the operator perform with the neutron,it may receive neutron irradiation.So the measurement and calculation for neutron dose become important.At home the research of neutron dose need to be advanced research.So the measurement and calculation method of neutron dose are conformed and summarized in this article for advance research.%随着科技的发展,中子在许多行业得到越来越广泛的应用,在医疗上应用最广泛的是硼中子俘获治疗.但在使用中子辐射的过程中,操作人员可能会受到中子辐射,因此中子剂量的测量和估算问题也就变得重要起来.目前,国内关于中子剂量的研究在有些方面还不是很深人,因此对中子剂量的测量和估算方法进行了归纳和阐述.

  6. EA-MC Neutronic Calculations on IAEA ADS Benchmark 3.2

    Energy Technology Data Exchange (ETDEWEB)

    Dahlfors, Marcus [Uppsala Univ. (Sweden). Dept. of Radiation Sciences; Kadi, Yacine [CERN, Geneva (Switzerland). Emerging Energy Technologies

    2006-01-15

    The neutronics and the transmutation properties of the IAEA ADS benchmark 3.2 setup, the 'Yalina' experiment or ISTC project B-70, have been studied through an extensive amount of 3-D Monte Carlo calculations at CERN. The simulations were performed with the state-of-the-art computer code package EA-MC, developed at CERN. The calculational approach is outlined and the results are presented in accordance with the guidelines given in the benchmark description. A variety of experimental conditions and parameters are examined; three different fuel rod configurations and three types of neutron sources are applied to the system. Reactivity change effects introduced by removal of fuel rods in both central and peripheral positions are also computed. Irradiation samples located in a total of 8 geometrical positions are examined. Calculations of capture reaction rates in {sup 129}I, {sup 237}Np and {sup 243}Am samples and of fission reaction rates in {sup 235}U, {sup 237}Np and {sup 243}Am samples are presented. Simulated neutron flux densities and energy spectra as well as spectral indices inside experimental channels are also given according to benchmark specifications. Two different nuclear data libraries, JAR-95 and JENDL-3.2, are applied for the calculations.

  7. Calculation of energy deposition, photon and neutron production in proton therapy of thyroid gland using MCNPX.

    Science.gov (United States)

    Mowlavi, Ali Asghar; Fornasie, Maria Rosa; de Denaro, Mario

    2011-01-01

    In this study, the MCNPX code has been used to simulate a proton therapy in thyroid gland, in order to calculate the proton energy deposition in the target region. As well as, we have calculated the photon and neutron production spectra due to proton interactions with the tissue. We have considered all the layers of tissue, from the skin to the thyroid gland, and an incident high energy pencil proton beam. The results of the simulation show that the best proton energy interval, to cover completely the thyroid tissue, is from 42 to 54 MeV, assuming that the thyroid gland has a 14 mm thickness and is located 11.2mm under the skin surface. The most percentage of deposited energy (78%) is related to the 54 MeV proton energy beam. Total photon and neutron production are linear and polynomial second order functions of the proton energy, respectively.

  8. Fission Product Decay Heat Calculations for Neutron Fission of 232Th

    Science.gov (United States)

    Son, P. N.; Hai, N. X.

    2016-06-01

    Precise information on the decay heat from fission products following times after a fission reaction is necessary for safety designs and operations of nuclear-power reactors, fuel storage, transport flasks, and for spent fuel management and processing. In this study, the timing distributions of fission products' concentrations and their integrated decay heat as function of time following a fast neutron fission reaction of 232Th were exactly calculated by the numerical method with using the DHP code.

  9. Lattice dynamics of wurtzite CdS: Neutron scattering and ab-initio calculations

    Science.gov (United States)

    Debernardi, A.; Pyka, N. M.; Göbel, A.; Ruf, T.; Lauck, R.; Kramp, S.; Cardona, M.

    1997-08-01

    We have measured the phonon dispersion of wurtzite CdS by inelastic neutron scattering in a single crystal made from the nonabsorbing isotope 114Cd. One of the two silent B 1-modes occurs at 3.96 THz ( k = 0 ). It is significantly lower and less dispersive than so far assumed. Previous semiempirical lattice dynamical models need to be reanalyzed. However, the observed dispersion branches compare favorably with an ab-initio calculation.

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

    OpenAIRE

    Bassem Elshahat; Akhtar Naqvi; Nabil Maalej

    2015-01-01

    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 dimensio...

  11. Neutronics calculations for the Oak Ridge National Laboratory Tokamak Reactor Studies

    Energy Technology Data Exchange (ETDEWEB)

    Santoro, R.T.; Baker, V.C.; Barnes, J.M.

    1976-01-01

    Neutronics calculations have been carried out to analyze the nuclear performance of conceptual blanket and shield designs for the Tokamak Experimental Power Reactor (EPR) and the Tokamak Demonstration Reactor Plant (DRP) being considered at the Oak Ridge National Laboratory. These reactor designs represent a sequence in the commercialization of fusion-generated electrical power. All of the calculations were carried out using the one-dimensional discrete ordinates code ANISN and the latest available ENDF/B-IV coupled neutron-gamma-ray transport cross-section data, fluence-to-kerma conversion factors, and radiation damage cross-section data. The calculations include spatial and integral heating-rate estimates in the reactor with emphasis on the recovery of fusion neutron energy in the blanket and limiting the heat-deposition rate in the superconducting toroidal field coils. Radiation damage due to atomic displacements and gas production produced in the reactor structural material and in the toroidal field coil windings were also estimated. The tritium-breeding ratio when natural lithium is used as the fertile material in the DRP blanket and in the experimental breeding modules in the EPR is also given.

  12. Transmission of fast neutrons along cylindrical air-filled ducts pierced in ilmenite concrete

    Energy Technology Data Exchange (ETDEWEB)

    Megahid, R.M.; Bashter, I.I.

    1980-01-01

    The variation of flux of fast neutrons along air filled ducts passing through ilmenite concrete of density 4.6 gm. cm/sup -3/ was measured. Ducts of diameter 2.9, 5.8 and 10 cm were used. Measurements were carried out at different distances (up to 120 cm) along the duct axis. The source of neutrons was a collimated beam of reactor neutrons emitted from one of the horizontal channels of ET-RR-1 reactor. All measurements were performed using phosphorus activation detectors. The data obtained show the dependence of flux values on duct length and diameter.

  13. Field-theory calculation of the electric dipole moment of the neutron and paramagnetic atoms

    Science.gov (United States)

    Griffith, Joel; Blundell, Steven; Sapirstein, Jonathan

    2013-04-01

    Electric dipole moments (edms) of bound states that arise from the constituents having edms are studied with field-theoretic techniques. The systems treated are the neutron and a set of paramagnetic atoms. In the latter case it is well known that the atomic edm differs greatly from the electron edm when the internal electric fields of the atom are taken into account. In the nonrelativistic limit these fields lead to a complete suppression, but for heavy atoms large enhancement factors are present. A general bound-state field theory approach applicable to both the neutron and paramagnetic atoms is set up. It is applied first to the neutron, treating the quarks as moving freely in a confining spherical well. It is shown that the effect of internal electric fields is small in this case. The atomic problem is then revisited using field-theory techniques in place of the usual Hamiltonian methods, and the atomic enhancement factor is shown to be consistent with previous calculations. Possible application of bound-state techniques to other sources of the neutron edm is discussed.

  14. French PWR 900 MWe pressure vessel surveillance neutron field characteristics TRIPOLI-3 calculations and experimental determination

    Energy Technology Data Exchange (ETDEWEB)

    Nimal, J.C.; Bourdet, L.; Zheng, S.H.; Vergnaud, T.; Kodeli, I. [CEA Centre d`Etudes de Saclay, 91 - Gif-sur-Yvette (France). Dept. de Mecanique et de Technologie; Lloret, R.; Bevilacqua, A. [CEA Centre d`Etudes de Saclay, 91 - Gif-sur-Yvette (France). Dept. des Reacteurs Experimentaux; Lefebvre, J.C. [Electricite de France (EDF), 69 - Villeurbanne (France)

    1994-12-31

    This paper presents an overview of the studies performed by CEA and EDF in the scope of the pressure vessel surveillance of the French nuclear power plants. The power plants are equipped with surveillance capsules, attached to the thermal shield. They contain the dosimeters and vessel material specimens for monitoring the effects of irradiation on the pressure vessel material. The Monte Carlo code TRIPOLI-3 is used with two nuclear data libraries to calculate the neutron flux, the steel damage and the dosimeter reaction rates, and takes into account the results of sensitivity/uncertainty calculations. 2 figs., 7 tabs., 10 refs.

  15. Calculation of Prompt Fission Neutron from 233U(n, f) Reaction by Multi-Modal Los Alamos Model%Calculation of Prompt Fission Neutron from 233U(n, f) Reaction by Multi-Modal Los Alamos Model

    Institute of Scientific and Technical Information of China (English)

    郑娜; 钟春来; 樊铁栓

    2012-01-01

    An attempt is made to improve the evaluation of the prompt fission neutron emis- sion from 233U(n, f) reaction for incident neutron energies below 6 MeV. The multi-modal fission approach is applied to the improved version of Los Alamos model and the point by point model. The prompt fission neutron spectra and the prompt fission neutron as a function of fragment mass (usually named "sawtooth" data) v(A) are calculated independently for the three most dominant fission modes (standard I, standard II and superlong), and the total spectra and v(A) are syn- thesized. The multi-modal parameters are determined on the basis of experimental data of fission fragment mass distributions. The present calculation results can describe the experimental data very well, and the proposed treatment is thus a useful tool for prompt fission neutron emission prediction.

  16. 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%.

  17. Calculations of the thermal and fast neutron fluxes in the Syrian miniature neutron source reactor using the MCNP-4C code.

    Science.gov (United States)

    Khattab, K; Sulieman, I

    2009-04-01

    The MCNP-4C code, based on the probabilistic approach, was used to model the 3D configuration of the core of the Syrian miniature neutron source reactor (MNSR). The continuous energy neutron cross sections from the ENDF/B-VI library were used to calculate the thermal and fast neutron fluxes in the inner and outer irradiation sites of MNSR. The thermal fluxes in the MNSR inner irradiation sites were also measured experimentally by the multiple foil activation method ((197)Au (n, gamma) (198)Au and (59)Co (n, gamma) (60)Co). The foils were irradiated simultaneously in each of the five MNSR inner irradiation sites to measure the thermal neutron flux and the epithermal index in each site. The calculated and measured results agree well.

  18. PWR neutron ex-vessel detection calculations using three-dimensional codes; Calculs de detection neutronique externe dans un rep

    Energy Technology Data Exchange (ETDEWEB)

    Dekens, O.; Lefebvre, J.C.; Rohart, M. [Electricite de France (EDF), 69 -Villeurbanne (France); Chiron, M. [CEA Centre d`Etudes de Saclay, 91 -Gif-sur-Yvette (France). Direction des Reacteurs Nucleaires; Wouters, R. de [TRACTEBEL, Brussels (Belgium)

    1997-10-01

    During the accident of TM12, the signal delivered by source detectors was exceptionally high. This phenomenon was found out to be due to the water inventory in the primary system. Thus, in their research activity, Electricite de France (EdF) and Commissariat a l`Energie Atomique (CEA) have jointly launched a programme, whose aim was to determine to what extent the response of ex-vessel neutron detectors are representative of reactor water level (or sources positions) in a French 900 MWe PWR. In this framework, both partners developed the methods needed for each step of the calculation chain. Finally, a simulation of a LOCA indicates that the loss of coolant can be detected by existing monitoring system, and could be more efficiently found by changing the position of the source range detectors. (authors). 11 refs.

  19. {sup 33}S for Neutron Capture Therapy: Nuclear Data for Monte Carlo Calculations

    Energy Technology Data Exchange (ETDEWEB)

    Porras, I., E-mail: porras@ugr.es [Departamento de Física Atómica, Molecular y Nuclear, Facultad de Ciencias, Universidad de Granada, E-18071 Granada (Spain); Sabaté-Gilarte, M.; Praena, J.; Quesada, J.M. [Departamento de Física Atómica, Molecular y Nuclear, Facultad de Física, Universidad de Sevilla, E-41012 Sevilla (Spain); Esquinas, P.L. [Departament of Physics and Astronomy, University of British Columbia, Vancouver, BC (Canada)

    2014-06-15

    A study of the nuclear data required for the Monte Carlo simulation of boron neutron capture therapy including the {sup 33}S isotope as an enhancer of the dose at small depths has been performed. In particular, the controversy on the available data for the {sup 33}S(n, α) cross section will be shown, which motivates new measurements. In addition to this, kerma factors for the main components of tissue are calculated with the use of fitting functions. Finally, we have applied these data to a potential neutron capture treatment with boron and sulfur addition to tissue in which part of the hydrogen atoms are replaced by deuterium, which improves the procedure.

  20. IRACM : A code system to calculate induced radioactivity produced by ions and neutrons

    Energy Technology Data Exchange (ETDEWEB)

    Tanaka, Susumu; Fukuda, Mitsuhiro; Nishimura, Koichi [Japan Atomic Energy Research Inst., Takasaki, Gunma (Japan). Takasaki Radiation Chemistry Research Establishment; Watanabe, Hiromasa; Yamano, Naoki

    1997-05-01

    It is essential to estimate of radioactivity induced in accelerator components and samples bombarded by energetic ion beams and the secondary neutrons of high-energy accelerator facilities in order to reduce the amount of radioactive wastes and to minimize radiation exposure to personnel. A computer code system IRACM has been developed to estimate product nuclides and induced radioactivity in various radiation environments of accelerator facilities. Nuclide transmutation with incident particles of neutron, proton, deuteron, alpha, {sup 12}C, {sup 14}N, {sup 16}O, {sup 20}Ne and {sup 40}Ar can be computed for arbitrary multi-layer target system in a one-dimensional geometry. The code system consists of calculation modules and libraries including activation cross sections, decay data and photon emission data. The system can be executed in both FACOM-M780 mainframe and DEC workstations. (author)

  1. System of Modelling and Calculation Analysis of Neutron- Physical Experiments at Fast Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Moiseyev, A.V. [SSC RF - IPPE, 1 Bondarenko Square, Obninsk, Kaluga Region 249033 (Russian Federation)

    2008-07-01

    There is an actual task on storage, processing and analysis of the unique experimental data received on power fast reactors for their subsequent use in projects of fast reactors of new (4.) generation. For modeling and carrying out analysis of experiments the integrated computing system MODEXSYS has been developed. In this system the mechanism for consecutive calculation of a fast reactor states with the detailed description of its components is created. The system includes the database describing fast reactor states, results of neutron-physical characteristics measurements at fast reactor, calculation and benchmark models of experiments and calculation results. In system convenient search means and the special graphics shell are provided. It has Interfaces for processing of calculation results and their analysis. MODEXSYS system has been applied for analysis of three types of experiments at fast reactor: k{sub eff}, control rod worth and energy release distribution. The most important results of this analysis are described. Application of MODEXSYS system will raise accuracy and reliability of forecasting of fast reactors neutron-physical characteristics; for BN-600 reactor recommended level of accuracy is resulted. (authors)

  2. 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.

  3. Neutron-deuteron scattering calculations with W-matrix representation of the two-body input

    Energy Technology Data Exchange (ETDEWEB)

    Bartnik, E.A.; Haberzettl, H.; Januschke, T.; Kerwath, U.; Sandhas, W.

    1987-11-01

    Employing the W-matrix representation of the partial-wave T matrix introduced by Bartnik, Haberzettl, and Sandhas, we show for the example of the Malfliet-Tjon potentials I and III that the single-term separable part of the W-matrix representation, when used as input in three-nucleon neutron-deuteron scattering calculations, is fully capable of reproducing the exact results obtained by Kloet and Tjon. This approximate two-body input not only satisfies the two-body off-shell unitarity relation but, moreover, it also contains a parameter which may be used in optimizing the three-body data. We present numerical evidence that there exists a variational (minimum) principle for the determination of the three-body binding energy which allows one to choose this parameter also in the absence of an exact reference calculation. Our results for neutron-deuteron scattering show that it is precisely this choice of the parameter which provides optimal scattering data. We conclude that the W-matrix approach, despite its simplicity, is a remarkably efficient tool for high-quality three-nucleon calculations.

  4. Neutron-deuteron scattering calculations with W-matrix representation of the two-body input

    Energy Technology Data Exchange (ETDEWEB)

    Bartnik, E.A.; Haberzettl, H.; Januschke, T.; Kerwath, U.; Sandhas, W.

    1987-05-01

    Employing the W-matrix representation of the partial-wave T matrix introduced by Bartnik, Haberzettl, and Sandhas, we show for the example of the Malfliet-Tjon potentials I and III that the single-term separable part of the W-matrix representation, when used as input in three-nucleon neutron-deuteron scattering calculations, is fully capable of reproducing the exact results obtained by Kloet and Tjon. This approximate two-body input not only satisfies the two-body off-shell unitarity relation but, moreover, it also contains a parameter which may be used in optimizing the three-body data. We present numerical evidence that there exists a variational (minimum) principle for the determination of the three-body binding energy which allows one to choose this parameter also in the absence of an exact reference calculation. Our results for neutron-deuteron scattering show that it is precisely this choice of the parameter which provides optimal scattering data. We conclude that the W-matrix approach, despite its simplicity, is a remarkably efficient tool for high-quality three-nucleon calculations.

  5. 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)

  6. Microscopic calculation of the spin-dependent neutron scattering lengths on 3He

    CERN Document Server

    Hofmann, H M

    2003-01-01

    We report on the spin.dependent neutron scattering length on 3He from a microscopic calculation of p-3H, n-3He, and d-2H scattering employing the Argonne v18 nucleon-nucleon potential with and without additional three-nucleon force. The results and that of a comprehensive R-matrix analysis are compared to a recent measurement. The overall agreement for the scattering lengths is quite good. The imaginary parts of the scattering lengths are very sensitive to the inclusion of three-nucleon forces, whereas the real parts are almost insensitive.

  7. Monte Carlo modeling of proton therapy installations: a global experimental method to validate secondary neutron dose calculations

    Science.gov (United States)

    Farah, J.; Martinetti, F.; Sayah, R.; Lacoste, V.; Donadille, L.; Trompier, F.; Nauraye, C.; De Marzi, L.; Vabre, I.; Delacroix, S.; Hérault, J.; Clairand, I.

    2014-06-01

    Monte Carlo calculations are increasingly used to assess stray radiation dose to healthy organs of proton therapy patients and estimate the risk of secondary cancer. Among the secondary particles, neutrons are of primary concern due to their high relative biological effectiveness. The validation of Monte Carlo simulations for out-of-field neutron doses remains however a major challenge to the community. Therefore this work focused on developing a global experimental approach to test the reliability of the MCNPX models of two proton therapy installations operating at 75 and 178 MeV for ocular and intracranial tumor treatments, respectively. The method consists of comparing Monte Carlo calculations against experimental measurements of: (a) neutron spectrometry inside the treatment room, (b) neutron ambient dose equivalent at several points within the treatment room, (c) secondary organ-specific neutron doses inside the Rando-Alderson anthropomorphic phantom. Results have proven that Monte Carlo models correctly reproduce secondary neutrons within the two proton therapy treatment rooms. Sensitive differences between experimental measurements and simulations were nonetheless observed especially with the highest beam energy. The study demonstrated the need for improved measurement tools, especially at the high neutron energy range, and more accurate physical models and cross sections within the Monte Carlo code to correctly assess secondary neutron doses in proton therapy applications.

  8. Monte Carlo modeling of proton therapy installations: a global experimental method to validate secondary neutron dose calculations.

    Science.gov (United States)

    Farah, J; Martinetti, F; Sayah, R; Lacoste, V; Donadille, L; Trompier, F; Nauraye, C; De Marzi, L; Vabre, I; Delacroix, S; Hérault, J; Clairand, I

    2014-06-07

    Monte Carlo calculations are increasingly used to assess stray radiation dose to healthy organs of proton therapy patients and estimate the risk of secondary cancer. Among the secondary particles, neutrons are of primary concern due to their high relative biological effectiveness. The validation of Monte Carlo simulations for out-of-field neutron doses remains however a major challenge to the community. Therefore this work focused on developing a global experimental approach to test the reliability of the MCNPX models of two proton therapy installations operating at 75 and 178 MeV for ocular and intracranial tumor treatments, respectively. The method consists of comparing Monte Carlo calculations against experimental measurements of: (a) neutron spectrometry inside the treatment room, (b) neutron ambient dose equivalent at several points within the treatment room, (c) secondary organ-specific neutron doses inside the Rando-Alderson anthropomorphic phantom. Results have proven that Monte Carlo models correctly reproduce secondary neutrons within the two proton therapy treatment rooms. Sensitive differences between experimental measurements and simulations were nonetheless observed especially with the highest beam energy. The study demonstrated the need for improved measurement tools, especially at the high neutron energy range, and more accurate physical models and cross sections within the Monte Carlo code to correctly assess secondary neutron doses in proton therapy applications.

  9. 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)

  10. Study of calculated and measured time dependent delayed neutron yields. [TX, for calculating delayed neutron yields; MATINV, for matrix inversion; in FORTRAN for LSI-II minicomputer

    Energy Technology Data Exchange (ETDEWEB)

    Waldo, R.W.

    1980-05-01

    Time-dependent delayed neutron emission is of interest in reactor design, reactor dynamics, and nuclear physics studies. The delayed neutrons from neutron-induced fission of /sup 232/U, /sup 237/Np, /sup 238/Pu, /sup 241/Am, /sup 242m/Am, /sup 245/Cm, and /sup 249/Cf were studied for the first time. The delayed neutron emission from /sup 232/Th, /sup 233/U, /sup 235/U, /sup 238/U, /sup 239/Pu, /sup 241/Pu, and /sup 242/Pu were measured as well. The data were used to develop an empirical expression for the total delayed neutron yield. The expression gives accurate results for a large variety of nuclides from /sup 232/Th to /sup 252/Cf. The data measuring the decay of delayed neutrons with time were used to derive another empirical expression predicting the delayed neutron emission with time. It was found that nuclides with similar mass-to-charge ratios have similar decay patterns. Thus the relative decay pattern of one nuclide can be established by any measured nuclide with a similar mass-to-charge ratio. A simple fission product yield model was developed and applied to delayed neutron precursors. It accurately predicts observed yield and decay characteristics. In conclusion, it is possible to not only estimate the total delayed neutron yield for a given nuclide but the time-dependent nature of the delayed neutrons as well. Reactors utilizing recycled fuel or burning actinides are likely to have inventories of fissioning nuclides that have not been studied until now. The delayed neutrons from these nuclides can now be incorporated so that their influence on the stability and control of reactors can be delineated. 8 figures, 39 tables.

  11. Development of NRESP98 Monte Carlo codes for the calculation of neutron response functions of neutron detectors. Calculation of the response function of spherical BF{sub 3} proportional counter

    Energy Technology Data Exchange (ETDEWEB)

    Hashimoto, M.; Saito, K.; Ando, H. [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center

    1998-05-01

    The method to calculate the response function of spherical BF{sub 3} proportional counter, which is commonly used as neutron dose rate meter and neutron spectrometer with multi moderator system, is developed. As the calculation code for evaluating the response function, the existing code series NRESP, the Monte Carlo code for the calculation of response function of neutron detectors, is selected. However, the application scope of the existing NRESP is restricted, the NRESP98 is tuned as generally applicable code, with expansion of the geometrical condition, the applicable element, etc. The NRESP98 is tested with the response function of the spherical BF{sub 3} proportional counter. Including the effect of the distribution of amplification factor, the detailed evaluation of the charged particle transportation and the effect of the statistical distribution, the result of NRESP98 calculation fit the experience within {+-}10%. (author)

  12. Calculations of the Efficiency of Registration of Thermal Neutrons by Complex Converters Constructed on the Basis of Gadolinium Foils

    CERN Document Server

    Abdushukurov, D A; Muminov, Kh Kh; Chistyakov, D Yu

    2007-01-01

    We consider the results of modeling of the efficiency of registration of thermal neutrons by the converters, which are made from natural gadolinium and its 157 isotope foils. Efficiency for a case of falling of neutrons under various angles to a plane of converters is calculated. It is shown, that at small angles of falling of neutrons to a plane of converters it is possible to receive the efficiency of registration close to a theoretical limit. Efficiency of the complex converter made of kapton supporting film with gadolinium converters layered on both its sides is considered. All calculations are carried out for four fixed neutron energies, which correspond to the wavelengths of 1, 1.8, 3 and 4 $\\AA$.

  13. The fast neutron fluence and the activation detector activity calculations using the effective source method and the adjoint function

    Energy Technology Data Exchange (ETDEWEB)

    Hep, J.; Konecna, A.; Krysl, V.; Smutny, V. [Calculation Dept., Skoda JS plc, Orlik 266, 31606 Plzen (Czech Republic)

    2011-07-01

    This paper describes the application of effective source in forward calculations and the adjoint method to the solution of fast neutron fluence and activation detector activities in the reactor pressure vessel (RPV) and RPV cavity of a VVER-440 reactor. Its objective is the demonstration of both methods on a practical task. The effective source method applies the Boltzmann transport operator to time integrated source data in order to obtain neutron fluence and detector activities. By weighting the source data by time dependent decay of the detector activity, the result of the calculation is the detector activity. Alternatively, if the weighting is uniform with respect to time, the result is the fluence. The approach works because of the inherent linearity of radiation transport in non-multiplying time-invariant media. Integrated in this way, the source data are referred to as the effective source. The effective source in the forward calculations method thereby enables the analyst to replace numerous intensive transport calculations with a single transport calculation in which the time dependence and magnitude of the source are correctly represented. In this work, the effective source method has been expanded slightly in the following way: neutron source data were performed with few group method calculation using the active core calculation code MOBY-DICK. The follow-up neutron transport calculation was performed using the neutron transport code TORT to perform multigroup calculations. For comparison, an alternative method of calculation has been used based upon adjoint functions of the Boltzmann transport equation. Calculation of the three-dimensional (3-D) adjoint function for each required computational outcome has been obtained using the deterministic code TORT and the cross section library BGL440. Adjoint functions appropriate to the required fast neutron flux density and neutron reaction rates have been calculated for several significant points within the RPV

  14. An approximated method of calculation of neutron spectra in reactor cells; Um metodo aproximado de calculo do espectro de neutrons em celulas de reatores

    Energy Technology Data Exchange (ETDEWEB)

    Caldeira, Alexandre D. [Centro Tecnico Aeroespacial (CTA-IEAv), Sao Jose dos Campos, SP (Brazil). Inst. de Estudos Avancados

    2000-07-01

    This work deals with the cell neutron spectra calculated with the transport equation for an infinite medium applied to the homogenized cell. Considering a radioisotope production reactor fuel cell, as a sample case, the maximum deviation found between the approximated and the S{sub N} methods was 13%. (author)

  15. An algorithm for calculating fresh air age in central ventilation system

    Institute of Scientific and Technical Information of China (English)

    LI; Xianting; (李先庭); Li; Dongning; (李冬宁); DOU; Chunpeng; (窦春鹏)

    2003-01-01

    Fresh air age is an important index to evaluate indoor environment. The conventional method for measuring or calculating fresh air age is only suitable for simple ventilation systems and not for central ventilation systems. In this paper, an algorithm for calculating fresh air age in central ventilation system is presented, which is based on the analysis of air flow in duct and air mixing. An example is given to illustrate the algorithm. The fresh air age in every ventilated room and duct can be acquired after all rooms and duct are directly calculated in turn without iteration. The algorithm is suitable for different central ventilation systems.

  16. Calculation of intermediate neutron flux in the radial reflectors of graphite reactors, comparison with experiments; Calcul du flux de neutrons intermediaires dans les reflecteurs lateraux des piles a graphite. Comparaison avec l'experience

    Energy Technology Data Exchange (ETDEWEB)

    Brisbois, J.; Vergnaud, T.; Oceraies, Y

    1967-12-01

    In a graphite pile, EDF or Inca type reactor, it is necessary to know the value of the intermediate neutron flux at the output of the lateral reflector in order to determine more precisely the neutron flux at the level of ionisation chambers. A sub critical pile of graphite and natural uranium was built, allowing to reconstitute the geometry of the radiation sources and the disposition of inferior and lateral protections of these piles. This pile is supplied with thermal neutrons coming from the Nereide light water type reactor. Some measurements of intermediate neutron flux have been made in this pile in order to establish a formalism for neutron flux calculation in slowing down in a whole core-lateral reflector, from the distribution of the thermal neutrons flux in the core. The flux calculation is done by age theory in three dimensions, in two homogenous media, separated by an axially semi infinite and laterally finite plane. One of these media includes a distribution of source. The constants are modified in order to take into account the presence of empty channels in the stacking. These calculations are done by the Malaga code. The checking of the formalism has been made in a greater complex geometry of these reactors that introduces an uncertainty factor in the comparison of results. We can however tell that we estimate correctly the variation of the intermediate neutrons flux in the core as well as its descending in a holed lateral reflector. The ratio between the calculation and the experiment is inferior to 2 or 3. Most of the time to a factor 2. [French] Dans une pile a graphite, du type EdF ou Inca, il est necessaire de connaitre la valeur du flux de neutrons intermediaires a la sortie du reflecteur lateral, afin de determiner avec plus de precision le flux de neutrons au niveau des chambres d'ionisation. Il a ete construit un empilement sous-critique, graphite uranium naturel, qui permet de reconstituer la geometrie des sources de rayonnement et la

  17. 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.

  18. Comparisons of Neutron Cross Sections and Isotopic Composition Calculations for Fission-Product Evaluations

    Science.gov (United States)

    Kim, Do Heon; Gil, Choong-Sup; Chang, Jonghwa; Lee, Yong-Deok

    2005-05-01

    The neutron absorption cross sections for 18 fission products evaluated within the framework of the KAERI (Korea Atomic Energy Research Institute)-BNL (Brookhaven National Laboratory) international collaboration have been compared with ENDF/B-VI.7. Also, the influence of the new evaluations on the isotopic composition calculations of the fission products has been estimated through the OECD/NEA burnup credit criticality benchmarks (Phase 1B) and the LWR/Pu recycling benchmarks. These calculations were performed by WIMSD-5B with the 69-group libraries prepared from three evaluated nuclear data libraries: ENDF/B-VI.7, ENDF/B-VI.8 including the new evaluations in the resonance region covering the thermal region, and the expected ENDF/B-VII including those in the upper resonance region up to 20 MeV. For Xe-131, the composition calculated with ENDF/B-VI.8 shows a maximum difference of 5.02% compared to ENDF/B-VI.7. However, the isotopic compositions of all the fission products calculated with the expected ENDF/B-VII show no differences when compared to ENDF/B-VI.7 for the thermal reactor benchmark cases.

  19. Atomic calculation for the atmospheres of strongly-magnetized neutron stars

    CERN Document Server

    Mori, K; Mori, Kaya; Hailey, Charles J.

    2001-01-01

    Complete modeling of radiative transfer in neutron star atmospheres is in progress, taking into account the anisotropy induced by magnetic fields, non-ideal effects and general relativity. As part of our modeling, we present a novel atomic calculation method producing an extensive atomic data set including energy values and oscillator strengths in the so-called Landau regime ($B > 4.7\\times10^9Z^2$ G). Conventional atmosphere models for B=0 are not applicable to typical field strengths of cooling neutron stars ($B \\sim10^{12}-10^{13}$ G), since an atom no longer keeps its spherical shape. The elemental composition and the configuration of the magnetic field in the atmosphere are presently unknown, so that atomic data must be produced for ground and excited states of several ions as a function of magnetic field. To accomplish this efficiently, we minimized the iterations in the Hartree equation and treated exchange terms and higher Landau states by perturbation methods. This method has the effect of reducing t...

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

    CERN Document Server

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

    2013-01-01

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

  1. Neutron dose measurements of Varian and Elekta linacs by TLD600 and TLD700 dosimeters and comparison with MCNP calculations.

    Science.gov (United States)

    Nedaie, Hassan Ali; Darestani, Hoda; Banaee, Nooshin; Shagholi, Negin; Mohammadi, Kheirollah; Shahvar, Arjang; Bayat, Esmaeel

    2014-01-01

    High-energy linacs produce secondary particles such as neutrons (photoneutron production). The neutrons have the important role during treatment with high energy photons in terms of protection and dose escalation. In this work, neutron dose equivalents of 18 MV Varian and Elekta accelerators are measured by thermoluminescent dosimeter (TLD) 600 and TLD700 detectors and compared with the Monte Carlo calculations. For neutron and photon dose discrimination, first TLDs were calibrated separately by gamma and neutron doses. Gamma calibration was carried out in two procedures; by standard 60Co source and by 18 MV linac photon beam. For neutron calibration by (241)Am-Be source, irradiations were performed in several different time intervals. The Varian and Elekta linac heads and the phantom were simulated by the MCNPX code (v. 2.5). Neutron dose equivalent was calculated in the central axis, on the phantom surface and depths of 1, 2, 3.3, 4, 5, and 6 cm. The maximum photoneutron dose equivalents which calculated by the MCNPX code were 7.06 and 2.37 mSv.Gy(-1) for Varian and Elekta accelerators, respectively, in comparison with 50 and 44 mSv.Gy(-1) achieved by TLDs. All the results showed more photoneutron production in Varian accelerator compared to Elekta. According to the results, it seems that TLD600 and TLD700 pairs are not suitable dosimeters for neutron dosimetry inside the linac field due to high photon flux, while MCNPX code is an appropriate alternative for studying photoneutron production.

  2. Calculation of Maxwellian-averaged cross sections and their uncertainties using ENDF/B-VII.1 evaluated neutron library

    Science.gov (United States)

    Pritychenko, Boris

    2011-10-01

    Present contribution represents a first application of ENDF/B-VII.1 neutron library for calculation of Maxwellian-averaged cross sections and astrophysical reaction rates. Recent improvements in neutron cross section evaluations and more extensive utilization of covariance files, by the CSEWG collaboration, allowed us to perform complete calculations and provide additional insights on all currently available neutron-induced reaction data. Nuclear reaction calculations using ENDF libraries and current Java technologies will be discussed and new results will be presented. This work was sponsored by the Office of Nuclear Physics, Office of Science of the U.S. Department of Energy, under Contract No. DE-AC02-98CH10886 with Brookhaven Science Associates, LLC.

  3. Calculation of neutron-decay-proton trapping in the Jovian magnetosphere

    Science.gov (United States)

    Thomas, J. R.; Doherty, W. R.

    1972-01-01

    Investigations on the possible proton fluxes resulting from cosmic ray albedo neutron decay (CRAND) are summarized. It is recalled that experimental data on the earth's belts are reasonably consistent with a CRAND source for most of the protons with energies 50 MeV in the heart of the inner belt. It is assumed that the CRAND and cosmic ray flux and spectrum for Jupiter are the same as for the Earth, with the significant difference that considerably higher energy particles are needed to reach the Jupiter atmosphere at the same magnetic latitude. The ratio of the CRAND source averaged over an L shell on Jupiter to that on Earth at L = 1.3 is derived from a geometrical calculation. The ratio of average loss rate on Jupiter to that for the Earth is also discussed. The flux around Jupiter is obtained by scaling according to the ratio of source to loss from the flux observed in the Earth's inner belt.

  4. Calculation of flow distribution in air reverse circulation bit interior fluid field by simplifying air flow model

    Institute of Scientific and Technical Information of China (English)

    Shuqing HAO; Hongwei HUANG; Kun YIN

    2007-01-01

    By simplifying the characters in the air reverse circulation bit interior fluid field, the authors used air dynamics and fluid mechanics to calculate the air distribution in the bit and obtained an equation of flow distribution with a unique resolution. This study will provide help for making certain the bit parameters of the bit structure effectively and study the air reverse circulation bit interior fluid field character deeply.

  5. 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.

  6. New Measurements and Calculations to Characterize the Caliban Pulsed Reactor Cavity Neutron Spectrum by the Foil Activation Method

    Energy Technology Data Exchange (ETDEWEB)

    Jacquet, X.; Casoli, P.; Authier, N.; Rousseau, G. [CEA, Centre de Valduc, 21120 Is-sur-Tille (France); Barsu, C. [Pl. de la fontaine, 25410 Corcelles-Ferrieres (France)

    2011-07-01

    Caliban is a cylindrical metallic core reactor mainly composed of uranium 235. It is operated by the Criticality and Neutron Science Research Laboratory located at the French Atomic Energy Commission research center in Valduc. As with other fast burst reactors, Caliban is used extensively for determining the responses of electronic parts or other objects and materials to neutron-induced displacements. Therefore, Caliban's irradiation characteristics, and especially its central cavity neutron spectrum, have to be very accurately evaluated. The foil activation method has been used in the past by the Criticality and Neutron Science Research Laboratory to evaluate the neutron spectrum of the different facilities it operated, and in particular to characterize the Caliban cavity spectrum. In order to strengthen and to improve our knowledge of the Caliban cavity neutron spectrum and to reduce the uncertainties associated with the available evaluations, new measurements have been performed on the reactor and interpreted by the foil activation method. A sensor set has been selected to sample adequately the studied spectrum. Experimental measured reaction rates have been compared to the results from UMG spectrum unfolding software and to values obtained with the activation code Fispact. Experimental and simulation results are overall in good agreement, although gaps exist for some sensors. UMG software has also been used to rebuild the Caliban cavity neutron spectrum from activation measurements. For this purpose, a default spectrum is needed, and one has been calculated with the Monte-Carlo transport code Tripoli 4 using the benchmarked Caliban description. (authors)

  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. Test case specifications for coupled neutronics-thermal hydraulics calculation of Gas-cooled Fast Reactor

    Science.gov (United States)

    Osuský, F.; Bahdanovich, R.; Farkas, G.; Haščík, J.; Tikhomirov, G. V.

    2017-01-01

    The paper is focused on development of the coupled neutronics-thermal hydraulics model for the Gas-cooled Fast Reactor. It is necessary to carefully investigate coupled calculations of new concepts to avoid recriticality scenarios, as it is not possible to ensure sub-critical state for a fast reactor core under core disruptive accident conditions. Above mentioned calculations are also very suitable for development of new passive or inherent safety systems that can mitigate the occurrence of the recriticality scenarios. In the paper, the most promising fuel material compositions together with a geometry model are described for the Gas-cooled fast reactor. Seven fuel pin and fuel assembly geometry is proposed as a test case for coupled calculation with three different enrichments of fissile material in the form of Pu-UC. The reflective boundary condition is used in radial directions of the test case and vacuum boundary condition is used in axial directions. During these condition, the nuclear system is in super-critical state and to achieve a stable state (which is numerical representation of operational conditions) it is necessary to decrease the reactivity of the system. The iteration scheme is proposed, where SCALE code system is used for collapsing of a macroscopic cross-section into few group representation as input for coupled code NESTLE.

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

    CERN Document Server

    Berruto, F; Orginos, K; Soni, A

    2005-01-01

    We present a study of the neutron electric dipole moment ($\\vec d_N$) within the framework of lattice QCD with two flavors of dynamical lig ht quarks. The dipole moment is sensitive to the topological structure of the gaug e fields, and accuracy can only be achieved by using dynamical, or sea quark, calc ulations. However, the topological charge evolves slowly in these calculations, le ading to a relatively large uncertainty in $\\vec d_N$. It is shown, using quenched configurations, that a better sampling of the charge d istribution reduces this problem, but because the CP even part of the fermion determinant is absent, both the topological charge dis tribution and $\\vec d_N$ are pathological in the chiral limit. We discuss the statistical and systematic uncertainties arising from the topological charge distr ibution and unphysical size of the quark mass in our calculations and prospects fo r eliminating them. Our calculations employ the RBC collaboration two flavor domain wall fermion and DBW2 gauge action l...

  12. Structural and theoretical investigations of short hydrogen bonds: neutron diffraction and plane-wave DFT calculations of urea phosphoric acid

    Science.gov (United States)

    Wilson, Chick C.; Morrison, Carole A.

    2002-08-01

    Low temperature neutron diffraction and high level computational methods have been applied to investigate the short hydrogen bond in urea-phosphoric acid. It is found that isolated molecule calculations predict a `normal' O-H⋯O hydrogen bond, in strong disagreement with the very short, 3 c-4 e hydrogen bond found from the neutron diffraction. Extending these calculations into a periodic environment using plane-wave DFT methods give much improved agreement with experiment, with a much shorter, stronger hydrogen bond, and significant elongation of the O-H `covalent' bond.

  13. Large-scale QRPA calculation of E1-strength and its impact on the neutron capture cross section

    CERN Document Server

    Goriely, S

    2002-01-01

    Large-scale QRPA calculations of the E1-strength are performed as a first attempt to microscopically derive the radiative neutron capture cross sections for the whole nuclear chart. A folding procedure is applied to the QRPA strength distribution to take the damping of the collective motion into account. It is shown that the resulting E1-strength function based on the SLy4 Skyrme force is in close agreement with photoabsorption data as well as the available experimental E1-strength at low energies. The increase of the E1-strength at low energies for neutron-rich nuclei is qualitatively analyzed and shown to affect the corresponding radiative neutron capture cross section significantly. A complete set of E1-strength function is made available for practical applications in a table format for all 7neutron drip lines.

  14. Long-Term Calculations with Large Air Pollution Models

    DEFF Research Database (Denmark)

    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...

  15. Calculating Hurst exponent and neutron monitor data in a single parallel algorithm

    Science.gov (United States)

    Kussainov, A. S.; Kussainov, S. G.

    2015-09-01

    We implemented an algorithm for simultaneous parallel calculation of the Hurst exponent H and the fractal dimension D for the time series of interest. Parallel programming environment was provided by OpenMPI library installed on three machines networked in the virtual cluster and operated by Debian Wheeze operating system. We applied our program for a comparative analysis of week and a half long, one minute resolution, six channels data from neutron monitor. To ensure a faultless functioning of the written code we applied it to analysis of the random Gaussian noise signal and time series with manually introduced self-affinity features. Both of them have the well-known values of H and D. All results are in good correspondence with each other and supported by the modern theories on signal processing thus confirming the validity of the implemented algorithms. Our code could be used as a standalone tool for the different time series data analysis as well as for the further work on development and optimization of the parallel algorithms for the time series parameters calculations.

  16. Measured and calculated effective delayed neutron fraction of the IPR-R1 Triga reactor

    Energy Technology Data Exchange (ETDEWEB)

    Souza, Rose Mary G.P.; Dalle, Hugo M.; Campolina, Daniel A.M., E-mail: souzarm@cdtn.b, E-mail: dallehm@cdtn.b, E-mail: campolina@cdtn.b [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2011-07-01

    The effective delayed neutron fraction, {beta}{sub eff}, one of the most important parameter in reactor kinetics, was measured for the 100 kW IPR-R1 TRIGA Mark I research reactor, located at the Nuclear Technology Development Center - CDTN, Belo Horizonte, Brazil. The current reactor core has 63 fuel elements, containing about 8.5% and 8% by weight of uranium enriched to 20% in U{sup 235}. The core has cylindrical configuration with an annular graphite reflector. Since the first criticality of the reactor in November 1960, the core configuration and the number of fuel elements have been changed several times. At that time, the reactor power was 30 kW, there were 56 fuel elements in the core, and the {beta}{sub eff} value for the reactor recommended by General Atomic (manufacturer of TRIGA) was 790 pcm. The current {beta}{sub eff} parameter was determined from experimental methods based on inhour equation and on the control rod drops. The estimated values obtained were (774 {+-} 38) pcm and (744 {+-} 20) pcm, respectively. The {beta}{sub eff} was calculated by Monte Carlo transport code MCNP5 and it was obtained 747 pcm. The calculated and measured values are in good agreement, and the relative percentage error is -3.6% for the first case, and 0.4% for the second one. (author)

  17. 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.

  18. Good manufacturing practice for modelling air pollution: Quality criteria for computer models to calculate air pollution

    Science.gov (United States)

    Dekker, C. M.; Sliggers, C. J.

    To spur on quality assurance for models that calculate air pollution, quality criteria for such models have been formulated. By satisfying these criteria the developers of these models and producers of the software packages in this field can assure and account for the quality of their products. In this way critics and users of such (computer) models can gain a clear understanding of the quality of the model. Quality criteria have been formulated for the development of mathematical models, for their programming—including user-friendliness, and for the after-sales service, which is part of the distribution of such software packages. The criteria have been introduced into national and international frameworks to obtain standardization.

  19. 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).

  20. A method based on potential theory for calculating air cavity formation of an air cavity resistance reduction ship

    Institute of Scientific and Technical Information of China (English)

    LI Yun-bo; WU Xiao-yu; MA Yong; WANG Jin-guang

    2008-01-01

    This research is intended to provide academic reference and design guidance for further studies to determine the most effective means to reduce a ship's resistance through an air-cavity.On the basis of potential theory and on the assumption of an ideal and irrotational fluid,this paper drives a method for calculating air cavity formation using slender ship theory then points out the parameters directly related to the formation of air cavities and their interrelationships.Simulations showed that the formation of an air cavity is affected by cavitation number,velocity,groove geometry and groove size.When the ship's velocity and groove structure are given,the cavitation number must be within range to form a steady air cavity.The interface between air and water forms a wave shape and could be adjustedby an air injection system.

  1. 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.

  2. Calculation of the absolute detection efficiency of a moderated /sup 235/U neutron detector on the Tokamak Fusion Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ku, L.P.; Hendel, H.W.; Liew, S.L.

    1989-02-01

    Neutron transport simulations have been carried out to calculate the absolute detection efficiency of a moderated /sup 235/U neutron detector which is used on the TFTR as a part of the primary fission detector diagnostic system for measuring fusion power yields. Transport simulations provide a means by which the effects of variations in various shielding and geometrical parameters can be explored. These effects are difficult to study in calibration experiments. The calculational model, benchmarked against measurements, can be used to complement future detector calibrations, when the high level of radioactivity resulting from machine operation may severely restrict access to the tokamak. We present a coupled forward-adjoint algorithm, employing both the deterministic and Monte Carlo sampling methods, to model the neutron transport in the complex tokamak and detector geometries. Sensitivities of the detector response to the major and minor radii, and angular anisotropy of the neutron emission are discussed. A semi-empirical model based on matching the calculational results with a small set of experiments produces good agreement (+-15%) for a wide range of source energies and geometries. 20 refs., 6 figs., 4 tabs.

  3. Optimum design of a moderator system based on dose calculation for an accelerator driven Boron Neutron Capture Therapy.

    Science.gov (United States)

    Inoue, R; Hiraga, F; Kiyanagi, Y

    2014-06-01

    An accelerator based BNCT has been desired because of its therapeutic convenience. However, optimal design of a neutron moderator system is still one of the issues. Therefore, detailed studies on materials consisting of the moderator system are necessary to obtain the optimal condition. In this study, the epithermal neutron flux and the RBE dose have been calculated as the indicators to look for optimal materials for the filter and the moderator. As a result, it was found that a combination of MgF2 moderator with Fe filter gave best performance, and the moderator system gave a dose ratio greater than 3 and an epithermal neutron flux over 1.0×10(9)cm(-2)s(-1).

  4. Calculation of Cross Section of Radiative Halo-Neutron Capture by 12C at Stellar Energy with the Asymptotic Normalization Coefficient Method

    Institute of Scientific and Technical Information of China (English)

    WU Kai-Su; CHEN Yong-Shou; LIU Zu-Hua; LIN Cheng-Jian; ZHANG Huan-Qiao

    2003-01-01

    The cross section of the direct neutron capture reaction 12C(n,7)13C(l/2+) is calculated with the asymptotic normalization coefficient method. The result is in good agreement with a recent experiment at low energy. An enormous enhancement of cross section is found for this direct neutron capture in which a p-wave neutron is captured into an 2?i/2 orbit with neutron halo. The possible effect of the neutron halo structure presented in this reaction on the s-process in astrophysics is discussed in general.

  5. Measurement and calculation of fast neutron and gamma spectra in well defined cores in LR-0 reactor.

    Science.gov (United States)

    Košťál, Michal; Matěj, Zdeněk; Cvachovec, František; Rypar, Vojtěch; Losa, Evžen; Rejchrt, Jiří; Mravec, Filip; Veškrna, Martin

    2017-02-01

    A well-defined neutron spectrum is essential for many types of experimental topics and is also important for both calibration and testing of spectrometric and dosimetric detectors. Provided it is well described, such a spectrum can also be employed as a reference neutron field that is suitable for validating selected cross sections. The present paper aims to compare calculations and measurements of such a well-defined spectra in geometrically similar cores of the LR-0 reactor with fuel containing slightly different enrichments (2%, 3.3% and 3.6%). The common feature to all cores is a centrally located dry channel which can be used for the insertion of studied materials. The calculation of neutron and gamma spectra was realized with the MCNP6 code using ENDF/B-VII.0, JEFF-3.1, JENDL-3.3, ROSFOND-2010 and CENDL-3.1 nuclear data libraries. Only minor differences in neutron and gamma spectra were found in the comparison of the presented reactor cores with different fuel enrichments. One exception is the gamma spectrum in the higher energy region (above 8MeV), where more pronounced variations could be observed.

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

    Science.gov (United States)

    Huang, Haili; Tang, Xiaoyan; Guo, Hui; Zhang, Yimen; Zhang, Yimeng; Zhang, Yuming

    2016-10-01

    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 6LiF or 10B is introduced. In this paper, pulse-height spectra of a PIN diode with a 6LiF 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 6LiF was calculated by using a FLUKA code, and a maximal efficiency of approximately 5% was achieved. Next, the energy distributions of both 3H and α induced by the 6LiF reaction according to different ranges of emission angle are analyzed. Subsequently, transient pulses generated by the bombardment of single 3H 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.

  7. Spatial and energy distributions of skyshine neutron and gamma radiation from nuclear reactors on the ground-air boundary

    Energy Technology Data Exchange (ETDEWEB)

    Orlov, Y.; Netecha, M.E.; Vasiliev, A.P.; Avaev, V.N.; Vasiliev, G.A. [Research and Development Institute of Power Engineering, Moscow (Russian Federation); Zelensky, D.I.; Istomin, Y.L.; Cherepnin, Y.S. [Institute of Atomic Energy of the National Nuclear Center of the Republic of Kazakhstan, Semipalatinsk-21 (Kazakhstan); Nomura, Y. [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2000-03-01

    A set of measurements on skyshine radiation was conducted at two special research reactors. A broad range of detectors was used in the measurements to record neutron and gamma radiations. Dosimetric and radiometric field measurements of the neutrons and gamma quanta of the radiation scattered in the air were performed at distances of 50 to 1000 m from the reactor during different weather conditions. The neutron spectra in the energy range of 1 eV to 10 MeV and the gamma quanta spectra in the range of 0.1-10 MeV were measured. (author)

  8. ANALISIS KANDUNGAN UNSUR ESENSIAL DAN TOKSIK DALAM TEH DAN AIR SEDUHANNYA DENGAN AKTIVASI NEUTRON

    Directory of Open Access Journals (Sweden)

    Th. Rina Mulyaningsih

    2015-04-01

    Full Text Available Kadar unsur logam K, Ca, Mn, Mg, Fe, Na, Zn, Rb, Br, Cr, Cs, La,Sc dan Co dalam 14 sampel teh hijau, teh hitam, teh hitam dengan aroma melati, aroma vanila, bunga rosella dan air seduhan teh telah ditentukan dengan analisis aktivasi neutron. Sampel teh dipilih dari produksi dalam negeri dan diperoleh dari Pasar Swalayan di daerah Serpong. Iradiasi neutron sampel dilakukan di Fasilitas Iradiasi reaktor RSG-GAS pada fluks neutron thermal sekitar sekitar 1013 ncm-2s-1. Prosedur kerja menggunakan SOP yang dikeluarkan oleh FNCA. Sebagai kontrol mutu digunakan SRM-NIST 1573a Tomato leaves dan NIST 1547 Peach leaves. Hasil analisis menunjukkan bahwa konsentrasi semua unsur bervariasi tergantung jenis teh. Konsentrasi Ca, K, Mg dan Mn dalam teh cukup tinggi > 100 mg/kg . Konsentrasi Ca dan K memiliki rentang nilai antara 1135,36-9123,21 dan 1064,41-2473,12 mg/kg serta Mg 2725,6-5528,5; dan Mn 95,38-815,48 mg/kg. Unsur mikroesensial Na, Fe, Co, La, Cr, Br, Sc, Cs, Rb dan Zn memiliki konsentrasi 100 mg/kg. Concentration of Ca and K have values in a range of 1135.36-9123.21 and 1064.41-2473.12 mg/kg as well as Mg of 2725.6-5528.5; and Mn of 95.38-815.48 mg/kg.Concentration of Na, Fe, Co, La, Cr, Br, Sc, Cs, Rb and Zn <100 mg/kg. Most elements in these tea were released into the infusions at defferent percentages in a range of 27.89-68.94% depending on the sort of the tea. There were not detected toxical elements Hg, Cd and As except Cr with low concentration. Therefore tea drink sare adequately good enough as essential elements source and content no toxic elements. Keywords: elemental analysis, essential, toxic, tea, neutron activation.

  9. 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.)

  10. 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)

  11. A study on calculation method for mechanical impedance of air spring

    Science.gov (United States)

    Changgeng, SHUAI; Penghui, LI; Rustighi, Emiliano

    2016-09-01

    This paper proposes an approximate analytic method of obtaining the mechanical impedance of air spring. The sound pressure distribution in cylindrical air spring is calculated based on the linear air wave theory. The influences of different boundary conditions on the acoustic pressure field distribution in cylindrical air spring are analysed. A 1-order ordinary differential matrix equation for the state vector of revolutionary shells under internal pressure is derived based on the non-moment theory of elastic thin shell. Referring to the transfer matrix method, a kind of expanded homogeneous capacity high precision integration method is introduced to solve the non-homogeneous matrix differential equation. Combined the solved stress field of shell with the calculated sound pressure field in air spring under the displacement harmonic excitation, the approximate analytical expression of the input and transfer mechanical impedance for the air spring can be achieved. The numerical simulation with the Comsol Multiphysics software verifies the correctness of theoretical analysis result.

  12. SU-E-T-569: Neutron Shielding Calculation Using Analytical and Multi-Monte Carlo Method for Proton Therapy Facility

    Energy Technology Data Exchange (ETDEWEB)

    Cho, S; Shin, E H; Kim, J; Ahn, S H; Chung, K; Kim, D-H; Han, Y; Choi, D H [Samsung Medical Center, Seoul (Korea, Republic of)

    2015-06-15

    Purpose: To evaluate the shielding wall design to protect patients, staff and member of the general public for secondary neutron using a simply analytic solution, multi-Monte Carlo code MCNPX, ANISN and FLUKA. Methods: An analytical and multi-Monte Carlo method were calculated for proton facility (Sumitomo Heavy Industry Ltd.) at Samsung Medical Center in Korea. The NCRP-144 analytical evaluation methods, which produced conservative estimates on the dose equivalent values for the shielding, were used for analytical evaluations. Then, the radiation transport was simulated with the multi-Monte Carlo code. The neutron dose at evaluation point is got by the value using the production of the simulation value and the neutron dose coefficient introduced in ICRP-74. Results: The evaluation points of accelerator control room and control room entrance are mainly influenced by the point of the proton beam loss. So the neutron dose equivalent of accelerator control room for evaluation point is 0.651, 1.530, 0.912, 0.943 mSv/yr and the entrance of cyclotron room is 0.465, 0.790, 0.522, 0.453 mSv/yr with calculation by the method of NCRP-144 formalism, ANISN, FLUKA and MCNP, respectively. The most of Result of MCNPX and FLUKA using the complicated geometry showed smaller values than Result of ANISN. Conclusion: The neutron shielding for a proton therapy facility has been evaluated by the analytic model and multi-Monte Carlo methods. We confirmed that the setting of shielding was located in well accessible area to people when the proton facility is operated.

  13. Monte Carlo Calculation of Core Reactivity and Fluxes for the Development of the BNCT Neutron Source at the Kyiv Research Reactor

    Science.gov (United States)

    Gritzay, Olena; Kalchenko, Oleksandr; Klimova, Nataliya; Razbudey, Volodymyr; Sanzhur, Andriy; Binney, Stephen

    2005-05-01

    The presented results show our consecutive steps in developing a neutron source with parameters required by Boron Neutron Capture Therapy (BNCT) at the Kyiv Research Reactor (KRR). The main goal of this work was to analyze the influence of installation of different types of uranium converters close to the reactor core on neutron beam characteristics and on level of reactor safety. The general Monte Carlo radiation transport code MCNP, version 4B, has been used for these calculations.

  14. Model calculations of the age of firn air across the Antarctic continent

    OpenAIRE

    2004-01-01

    The age of firn air in Antarctica at pore close-off depth is only known for a few specific sites where firn air has been sampled for analyses. We present a model that calculates the age of firn air at pore close-off depth for the entire Antarctic continent. The model basically uses four meteorological parameters as input (surface temperature, pressure, accumulation rate and wind speed). Using parameterisations for surface snow density, pore close-off density and tortuosity, ...

  15. Comparison of fresh fuel experimental measurements to MCNPX calculations using self-interrogation neutron resonance densitometry

    Energy Technology Data Exchange (ETDEWEB)

    LaFleur, Adrienne M., E-mail: alafleur@lanl.gov [Nuclear Nonproliferation Division, Los Alamos National Laboratory, P.O. Box 1663 MS E540, Los Alamos, NM 87545 (United States); Charlton, William S., E-mail: wcharlton@tamu.edu [Nuclear Security Science and Policy Institute, Texas A and M University, 3473 TAMU, College Station, TX 77843 (United States); Menlove, Howard O., E-mail: hmenlove@lanl.gov [Nuclear Nonproliferation Division, Los Alamos National Laboratory, P.O. Box 1663 MS E540, Los Alamos, NM 87545 (United States); Swinhoe, Martyn T., E-mail: swinhoe@lanl.gov [Nuclear Nonproliferation Division, Los Alamos National Laboratory, P.O. Box 1663 MS E540, Los Alamos, NM 87545 (United States)

    2012-07-11

    A new non-destructive assay technique called Self-Interrogation Neutron Resonance Densitometry (SINRD) is currently being developed at Los Alamos National Laboratory (LANL) to improve existing nuclear safeguards measurements for Light Water Reactor (LWR) fuel assemblies. SINRD consists of four {sup 235}U fission chambers (FCs): bare FC, boron carbide shielded FC, Gd covered FC, and Cd covered FC. Ratios of different FCs are used to determine the amount of resonance absorption from {sup 235}U in the fuel assembly. The sensitivity of this technique is based on using the same fissile materials in the FCs as are present in the fuel because the effect of resonance absorption lines in the transmitted flux is amplified by the corresponding (n,f) reaction peaks in the fission chamber. In this work, experimental measurements were performed in air with SINRD using a reference Pressurized Water Reactor (PWR) 15 Multiplication-Sign 15 low enriched uranium (LEU) fresh fuel assembly at LANL. The purpose of this experiment was to assess the following capabilities of SINRD: (1) ability to measure the effective {sup 235}U enrichment of the PWR fresh LEU fuel assembly and (2) sensitivity and penetrability to the removal of fuel pins from an assembly. These measurements were compared to Monte Carlo N-Particle eXtended transport code (MCNPX) simulations to verify the accuracy of the MCNPX model of SINRD. The reproducibility of experimental measurements via MCNPX simulations is essential to validating the results and conclusions obtained from the simulations of SINRD for LWR spent fuel assemblies. - Highlights: Black-Right-Pointing-Pointer Development of new measurement technique called SINRD to improve LWR safeguards. Black-Right-Pointing-Pointer Performed SINRD experiment to measure {sup 235}U and pin diversions in PWR fresh assembly. Black-Right-Pointing-Pointer Excellent agreement of MCNPX and measured results confirmed accuracy of SINRD model. Black-Right-Pointing-Pointer SINRD

  16. 40 CFR 86.166-12 - Method for calculating emissions due to air conditioning leakage.

    Science.gov (United States)

    2010-07-01

    ... 40 Protection of Environment 18 2010-07-01 2010-07-01 false Method for calculating emissions due to air conditioning leakage. 86.166-12 Section 86.166-12 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) AIR PROGRAMS (CONTINUED) CONTROL OF EMISSIONS FROM NEW AND IN-USE HIGHWAY VEHICLES AND ENGINES Emission Regulations for...

  17. Development of additional module to neutron-physic and thermal-hydraulic computer codes for coolant acoustical characteristics calculation

    Energy Technology Data Exchange (ETDEWEB)

    Proskuryakov, K.N.; Bogomazov, D.N.; Poliakov, N. [Moscow Power Engineering Institute (Technical University), Moscow (Russian Federation)

    2007-07-01

    The new special module to neutron-physic and thermal-hydraulic computer codes for coolant acoustical characteristics calculation is worked out. The Russian computer code Rainbow has been selected for joint use with a developed module. This code system provides the possibility of EFOCP (Eigen Frequencies of Oscillations of the Coolant Pressure) calculations in any coolant acoustical elements of primary circuits of NPP. EFOCP values have been calculated for transient and for stationary operating. The calculated results for nominal operating were compared with results of measured EFOCP. For example, this comparison was provided for the system: 'pressurizer + surge line' of a WWER-1000 reactor. The calculated result 0.58 Hz practically coincides with the result of measurement (0.6 Hz). The EFOCP variations in transients are also shown. The presented results are intended to be useful for NPP vibration-acoustical certification. There are no serious difficulties for using this module with other computer codes.

  18. A Study of Neutronics Effects of the Spacer Grids in a Typical PWR via Monte Carlo Calculation

    Energy Technology Data Exchange (ETDEWEB)

    Bach, Tran Xuan; Cho, Nam Zin [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

    2014-10-15

    Neutronics analysis, the spacer grids which support fuel rods are not explicitly described, but they are homogenized with coolant. However, the effects of neglecting or simplifying the spacer grids are not reported in the literature to the best of our knowledge. In this paper, to investigate the effects of spacer grids in neutronics analysis, a detailed description of spacer grids is added to the KAIST benchmark problem 1B. Then, the effective multiplication factor, spatial distributions of neutron flux, and its energy spectrum are obtained for the two cases (with and without spacer grids). Numerical results show that the effects of spacer grids are not negligible. In this paper, to investigate the effect of spacer grids, the spacer grid geometry is described in detail in the Monte Carlo calculation. In the numerical test, the two cases are compared in the context of a modified KAIST benchmark problem 1B. Case 1 does not have spacer grids, while the space is filled by coolant instead. Case 2 includes the spacer grids. The difference in neutron flux spectra is also observed. Thus, the effect of the spacer grids should be considered in the whole-core reactor analysis. In practice, the spacer grids are homogenized into coolant to consider its effect. As a further study, therefore, it would be worthwhile to investigate the differences between the homogenization and the explicit description of the spacer grids.

  19. 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.

  20. 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.

  1. 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

  2. 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)

  3. Using Neutron Radiography to Quantify Water Transport and the Degree of Saturation in Entrained Air Cement Based Mortar

    Science.gov (United States)

    Lucero, Catherine L.; Bentz, Dale P.; Hussey, Daniel S.; Jacobson, David L.; Weiss, W. Jason

    Air entrainment is commonly added to concrete to help in reducing the potential for freeze thaw damage. It is hypothesized that the entrained air voids remain unsaturated or partially saturated long after the smaller pores fill with water. Small gel and capillary pores in the cement matrix fill quickly on exposure to water, but larger pores (entrapped and entrained air voids) require longer times or other methods to achieve saturation. As such, it is important to quantitatively determine the water content and degree of saturation in air entrained cementitious materials. In order to further investigate properties of cement-based mortar, a model based on Beer's Law has been developed to interpret neutron radiographs. This model is a powerful tool for analyzing images acquired from neutron radiography. A mortar with a known volume of aggregate, water to cement ratio and degree of hydration can be imaged and the degree of saturation can be estimated.

  4. A generic data translation scheme for the coupling of high-fidelity fusion neutronics and CFD calculations

    Energy Technology Data Exchange (ETDEWEB)

    Qiu, Yuefeng, E-mail: yuefeng.qiu@kit.edu [Association KIT-Euratom, Karlsruhe Institute of Technology (KIT), Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen, Karlsruhe (Germany); Lu, Peng [Association KIT-Euratom, Karlsruhe Institute of Technology (KIT), Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen, Karlsruhe (Germany); School of Nuclear Science and Technology, University of Science and Technology of China, Hefei, Anhui 230026 (China); Fischer, Ulrich; Pereslavtsev, Pavel; Kecskes, Szabolcs [Association KIT-Euratom, Karlsruhe Institute of Technology (KIT), Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen, Karlsruhe (Germany)

    2014-10-15

    Highlights: • A data translation scheme has been developed for coupling Monte Carlo neutronics and CFD simulations. • It contains a generic data translation kernel, and interfaces for the MCNP, CFX and Fluent code. • A blanket test case model was investigated for validation and verification purposes. • Results of the so-called Inversion Check are very close to MCNP calculated results. - Abstract: The design of fusion device components is achieved through iterative coupled neutronics and thermal hydraulics analyses. A translation scheme has been developed for transferring the nuclear heating data from Monte Carlo (MC) neutronic calculations to CFD simulations. It contains a generic data translation kernel which supports the high-fidelity data mapping of MC meshes on CFD meshes, and provides interfaces for processing the nuclear response data on the meshes for CFD codes. This translation scheme has been implemented in the open-source pre- and post-processing platform SALOME to extend its capabilities on data manipulations and visualizations. For verification purposes, a blanket test case based on the Helium Cooled Pebble Bed Test Blanket Module was investigated. The processing of the heating distribution data was validated through a so-called Inversion Check comparing the inverted heating field with the original MC tally distribution. The results of the verification have been discussed in detail, and the reliability of the data translation scheme is concluded.

  5. Applying full multigroup cell characteristics from MCU code to finite difference calculations of neutron field in VVER core

    Energy Technology Data Exchange (ETDEWEB)

    Gorodkov, S.S.; Kalugin, M.A. [Nuclear Research Centre ' ' Kurchatov Institute' ' , Moscow (Russian Federation)

    2015-09-15

    Up to now core calculations with Monte Carlo provided only average cross-sections of mesh cells for further use either in finite difference calculations or as benchmark ones for approximate spectral algorithms. Now MCU code is capable to handle functions, which may be interpreted as average diffusion coefficients. Subsequently the results of finite difference calculations with cells characteristic sets obtained in such a way can be compared with Monte Carlo results as benchmarks, giving reliable information on quality of production code under consideration. As an example of such analysis, the results of mesh calculations with 1-, 2-, 4-, 8- and 12 neutron groups of some model VVER fuel assembly are presented in comparison with the exact Monte Carlo solution. As a second example, an analysis is presented of water gap approximate enlargement between fuel assemblies, allowing VVER core region be covered by regular mesh.

  6. AFWL (Air Force Weapons Laboratory) HULL (Hydrodynamics Unlimited) calculations of air blast over a dam slope. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Fry, M.A.; Needham, C.E.; Stucker, M.; Chambers, B.S.; Ganong, G.P.

    1976-10-01

    This laboratory performed Hydrodynamics Unlimited (HULL) calculations of the air blast over a dam for two yields and two pressure regions. A 5th calculation included a rigid blockhouse at the foot of the dam. Although the shielding effect of the dam reduced the incident blast wave overpressure, reflection of the shock from the valley floor raised the peak overpressure up to at least 40% of the free air value. In almost every case, the overpressure impulses near the foot of the dam were greater than or equal to free air values. The rigid blockhouse experienced the most severe overpressure environments. The assumption of a 50-psi hard blockhouse is reasonable. During collapse of the blockhouse, it appears to be rigid to the air flow, since it responds slowly to the rapid air blast. Although there may be other reasons to detonate the weapon on the surface of the reservoir, the best way to destroy the blockhouse and any related structures with air blast, probably would be to detonate the device downstream of the blockhouse.

  7. Probing TeV scale physics via ultra cold neutron decays and calculating non-standard baryon matrix elements

    CERN Document Server

    Gupta, Rajan; Joseph, Anosh; Lin, Huey-Wen; Cohen, Saul D

    2012-01-01

    We motivate undertaking precision analyses of neutron decays to look for signatures of new scalar and tensor interactions that can arise in extensions of the Standard Model at the TeV scale. The key ingrediant needed to connect experimental data with theoretical analysis are high-precision calculations of matrix elements of isovector bilinear operators between the decaying neutron and final state proton. We describe the status of our Lattice QCD program of using valence clover fermions on dynamical N_f=2+1+1 HISQ configurations generated by the MILC Collaboration. On the theoretical side we use the effective field theory method and provide both model independent and dependent analyses to obtain bounds on possible scalar and tensor interactions, both from low energy experiments and LHC data.

  8. Calculation of transport coefficients of air-water vapor mixtures thermal plasmas used in circuit breakers

    Directory of Open Access Journals (Sweden)

    KOHIO Niéssan

    2014-12-01

    Full Text Available In this paper we calculate the transport coefficients of plasmas formed by air and water vapor mixtures. The calculation, which assume local thermodynamic equilibrium (LTE are performed in the temperature range from 500 to 12000 K. We use the Gibbs free energy minimization method to determine the equilibrium composition of the plasmas, which is necessary to calculate the transport coefficients. We use the Chapman-Enskog method to calculate the transport coefficients. The results are presented and discussed according to the rate of water vapor. The results of the total thermal conductivity and electrical conductivity show in particular that the increasing of the rate of water vapor in air can be interesting for power cut. This could be improve the performance of plasma during current breaking in air contaminate by the water vapor.

  9. Computer program calculates gamma ray source strengths of materials exposed to neutron fluxes

    Science.gov (United States)

    Heiser, P. C.; Ricks, L. O.

    1968-01-01

    Computer program contains an input library of nuclear data for 44 elements and their isotopes to determine the induced radioactivity for gamma emitters. Minimum input requires the irradiation history of the element, a four-energy-group neutron flux, specification of an alloy composition by elements, and selection of the output.

  10. A new interpretation of the proton-neutron bound state The calculation of the binding energy

    CERN Document Server

    Mandache, N

    1996-01-01

    We treat the old problem of the proton-neutron bound state (the deuteron). Using a new concept of incomplete (partial) annihilation process we derive a formula for the binding energy of the deuteron, which does not contain any new constant. Some implications of this new approach are discussed.

  11. 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.

  12. Implications for clinical treatment from the micrometer site dosimetric calculations in boron neutron capture therapy

    Energy Technology Data Exchange (ETDEWEB)

    Nichols, Trent L. [Department of Physics and Astronomy, University of Tennessee, Knoxville, TN 37901 (United States)], E-mail: tnichol2@utk.edu; Kabalka, George W. [Department of Chemistry, University of Tennessee, Knoxville, TN 37901 (United States); Miller, Laurence F. [Department of Nuclear and Radiological Engineering, University of Tennessee, Knoxville, TN 37901 (United States); McCormack, Michael T. [Department of Medicine, University of Tennessee Graduate School of Medicine, Knoxville, TN 37920 (United States); Johnson, Andrew [Rush University Medical Center, Chicago, IL 60612 (United States)

    2009-07-15

    Boron neutron capture therapy has now been used for several malignancies. Most clinical trials have addressed its use for the treatment of glioblastoma multiforme. A few trials have focused on the treatment of malignant melanoma with brain metastases. Trial results for the treatment of glioblastoma multiforme have been encouraging, but have not achieved the success anticipated. Results of trials for the treatment of malignant melanoma have been very promising, though with too few patients for conclusions to be drawn. Subsequent to these trials, regimens for undifferentiated thyroid carcinoma, hepatic metastases from adenocarcinoma of the colon, and head and neck malignancies have been developed. These tumors have also responded well to boron neutron capture therapy. Glioblastoma is an infiltrative tumor with distant individual tumor cells that might create a mechanism for therapeutic failure though recurrences are often local. The microdosimetry of boron neutron capture therapy can provide an explanation for this observation. Codes written to examine the micrometer scale energy deposition in boron neutron capture therapy have been used to explore the effects of near neighbor cells. Near neighbor cells can contribute a significantly increased dose depending on the geometric relationships. Different geometries demonstrate that tumors which grow by direct extension have a greater near neighbor effect, whereas infiltrative tumors lose this near neighbor dose which can be a significant decrease in dose to the cells that do not achieve optimal boron loading. This understanding helps to explain prior trial results and implies that tumors with small, closely packed cells that grow by direct extension will be the most amenable to boron neutron capture therapy.

  13. Experimental investigation of neutronic characteristics of the IR-8 reactor to confirm the results of calculations by MCU-PTR code

    Energy Technology Data Exchange (ETDEWEB)

    Surkov, A. V., E-mail: surkov.andrew@gmail.com; Kochkin, V. N.; Pesnya, Yu. E.; Nasonov, V. A.; Vihrov, V. I.; Erak, D. Yu. [National Research Center Kurchatov Institute (Russian Federation)

    2015-12-15

    A comparison of measured and calculated neutronic characteristics (fast neutron flux and fission rate of {sup 235}U) in the core and reflector of the IR-8 reactor is presented. The irradiation devices equipped with neutron activation detectors were prepared. The determination of fast neutron flux was performed using the {sup 54}Fe (n, p) and {sup 58}Ni (n, p) reactions. The {sup 235}U fission rate was measured using uranium dioxide with 10% enrichment in {sup 235}U. The determination of specific activities of detectors was carried out by measuring the intensity of characteristic gamma peaks using the ORTEC gamma spectrometer. Neutron fields in the core and reflector of the IR-8 reactor were calculated using the MCU-PTR code.

  14. 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.

  15. Slow neutron total cross-section, transmission and reflection calculation for poly- and mono-NaCl and PbF2 crystals

    Science.gov (United States)

    Mansy, Muhammad S.; Adib, M.; Habib, N.; Bashter, I. I.; Morcos, H. N.; El-Mesiry, M. S.

    2016-10-01

    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 PbF2 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 PbF2 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 PbF2 (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.

  16. Calculation of neutron and gamma fluxes in support to the interpretation of measuring devices irradiated in the core periphery of the OSIRIS Material Testing Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Malouch, Fadhel [Alternative Energies and Atomic Energy Commission - CEA, Saclay Center, DEN/DANS/DM2S/SERMA, F-91191 Gif-sur-Yvette Cedex (France)

    2015-07-01

    Technological irradiations carried out in material testing reactors (MTRs) are used to study the behavior of materials under irradiation conditions required by different types of nuclear power plants (NPPs). For MTRs, specific instrumentation is required for the experiment monitoring and for the characterization of irradiation conditions, in particular the flux of neutrons and photons. To measure neutron and photon flux in experimental locations, different sensors can be used, such as SPNDs (self-powered neutron detectors), SPGDs (self-powered gamma detectors) and ionization chambers. These sensors involve interactions producing ultimately a measurable electric current. Various sensors have been recently tested in the core periphery of the OSIRIS reactor (located at the CEA-Saclay center) in order to qualify their responses to the neutron and the photon flux. One of the key input data for this qualification is to have a relevant evaluation of neutron and gamma fluxes at the irradiation location. The objective of this work is to evaluate the neutron and the gamma flux in the core periphery of the OSIRIS reactor. With this intention, specific neutron-photonic three-dimensional calculations have been performed and are mainly based on the TRIPOLI-4{sup R} three-dimensional continuous-energy Monte Carlo code, developed by CEA (Saclay Center) and extensively validated against reactor dosimetry benchmarks. In the case of the OSIRIS reactor, TRIPOLI-4{sup R} code has been validated against experimental results based on neutron flux and nuclear heating measurements performed in ex-core and in-core experiments. In this work, simultaneous contribution of neutrons and gamma photons in the core periphery is considered using neutron-photon coupled transport calculations. Contributions of prompt and decay photons have been taken into account for the gamma flux calculation. Specific depletion codes are used upstream to provide the decay-gamma sources required by TRIPOLI-4

  17. Test problem for thermal-hydraulics and neutronic coupled calculation fore ALFREAD reactor core

    Science.gov (United States)

    Filip, A.; Darie, G.; Saldikov, I. S.; Smirnov, A. D.; Tikhomirov, G. V.

    2017-01-01

    The beginning of a new era of nuclear reactor requires technological advances and also multiples studies. The European Liquid metal cooled Fast breeder Reactor is one of the designs for the generation IV nuclear reactor, selected by ENEA. A pioneer of its time, ELFR needs a demonstrator in order to prove the feasibility of this project and to acquire more data and experience in operating a LFR. For this reason the ALFRED project was started and it is expected to be under operation by the year 2030. This paper has the objective of analyzing the neutronic and thermohydraulics of the ALFRED core by the means of a coupled scheme. The selected code for neutronic simulation is MCNP and the selected code for thermohydraulics is ANSYS.

  18. 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.

  19. Small-angle neutron scattering investigation of polyurethane aged in dry and wet air

    Directory of Open Access Journals (Sweden)

    Q. Tian

    2014-05-01

    Full Text Available The microstructures of Estane 5703 aged at 70°C in dry and wet air have been studied by small-angle neutron scattering. The samples were swollen in deuterated toluene for enhancing the contrast. The scattering data show the characteristic domain structure of polyurethanes consisting of soft and hard segments. Debye-Anderson-Brumberger function used with hard sphere structure factor, and the Teubner-Strey model are used to analyze the two-phase domain structure of the polymer. The combined effects of temperature and humidity have a strong disruption effect on the microstructures of Estane. For the sample aged at 70°C in wet air for 1 month, the domain size, described by the correlation length, increases from 2.3 to 3.8 nm and their distance, expressed by hard-sphere interaction radius, increases from 8.4 to 10.6 nm. The structure development is attributed to degradation of polymer chains as revealed by gel permeation chromatography. The hydrolysis of ester links on polymer backbone at 70°C in the presence of water humidity is the main reason for the changes of the microstructure. These findings can contribute to developing predictive models for the safety, performance, and lifetime of polyurethanes.

  20. A Computer Program to Calculate the Supersonic Flow over a Solid Cone in Air or Water.

    Science.gov (United States)

    1984-06-01

    ix air or water. The rain objective is to calculate the ccne semi-vertei angle given prescribed initial ccndi- tions. The program is written in...tc the motion of the metal jet frcm an explczive shaped-charge fired underwater. A tiical result for supersonic flow over a ccne in water is as follcws...the ccne semi-vertex angle is calculated to be 7.23 degrees. Gene rally, pressures invclved in water flow are much larger than for air flow, and the

  1. Analytic Calculation of Radio Emission from Extensive Air Showers subjected to Atmospheric Electric Fields

    CERN Document Server

    Scholten, Olaf; de Vries, Krijn D; van Sloten, Lucas

    2016-01-01

    We have developed a code that semi-analytically calculates the radio footprint (intensity and polarization) of an extensive air shower subject to atmospheric electric fields. This can be used to reconstruct the height dependence of atmospheric electric field from the measured radio footprint. The various parameterizations of the spatial extent of the induced currents are based on the results of Monte-Carlo shower simulations. The calculated radio footprints agree well with microscopic CoREAS simulations.

  2. 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.

  3. Self-consistent calculations of the strength function and radiative neutron capture cross section for stable and unstable tin isotopes

    CERN Document Server

    Goriely, S; Krewald, S

    2011-01-01

    The E1 strength function for 15 stable and unstable Sn even-even isotopes from A=100 till A=176 are calculated using the self-consistent microscopic theory which, in addition to the standard (Q)RPA approach, takes into account the single-particle continuum and the phonon coupling. Our analysis shows two distinct regions for which the integral characteristics of both the giant and pygmy resonances behave rather differently. For neutron-rich nuclei, starting from $^{132}$Sn, we obtain a giant E1 resonance which significantly deviates from the widely-used systematics extrapolated from experimental data in the $\\beta$-stability valley. We show that the inclusion of the phonon coupling is necessary for a proper description of the low-energy pygmy resonances and the corresponding transition densities for $A132$ region the influence of phonon coupling is significantly smaller. The radiative neutron capture cross sections leading to the stable $^{124}$Sn and unstable $^{132}$Sn and $^{150}$Sn nuclei are calculated wi...

  4. 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.

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

    Science.gov (United States)

    Dzhalandinov, A.; Tsofin, V.; Kochkin, V.; Panferov, P.; Timofeev, A.; Reshetnikov, A.; Makhotin, D.; Erak, D.; Voloschenko, A.

    2016-02-01

    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.

  6. Sensitivity to Nuclear Data and Neutron Source Type in Calculations of Transmutation Capabilities of the Energy Amplifier Demonstration Facility

    Energy Technology Data Exchange (ETDEWEB)

    Dahlfors, Marcus

    2003-05-01

    This text is a summary of two studies the author has performed within the field of 3-D Monte Carlo calculations of Accelerator Driven Systems (ADS) for transmutation of nuclear waste. The simulations were carried out with the state-of-the-art computer code package EA-MC, developed by C. Rubbia and his group at CERN. The concept studied is ANSALDOs 80 MWth Energy Amplifier Demonstration Facility based on classical MOX-fuel technology and on molten Lead-Bismuth Eutectic cooling. A review of neutron cross section sensitivity in numerical calculations of an ADS and a comparative assessment relevant to the transmutation efficiency of plutonium and minor actinides in fusion/fission hybrids and ADS are presented.

  7. 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.)

  8. Implementation of a method for calculating output factors in air for irregular fields

    Energy Technology Data Exchange (ETDEWEB)

    Suero Rodrigo, M. A.; Marques Frguela, E.

    2011-07-01

    The concept of output factor in air (Sc) was introduced to characterize the variation of the incident photon fluence per unit monitor with different settings of the collimator. The objective of this work is the implementation of the method proposed by Zhu et al. (2004) to calculate both as FCSc Sc and verification with the measurements performed in mini-mannequin.

  9. Chasing air masses in the Arctic vortex: An evaluation of trajectory calculations using an active Match

    Science.gov (United States)

    Wegner, T.; Grooss, J.; Mueller, R.; Stroh, F.; Lehmann, R.; Volk, C.; Hösen, E.; Vom Scheidt, M.; Wintel, J.; Riediger, O.; Schlager, H.; Scheibe, M.; Stock, P.; Ravegnani, F.; Ulanovsky, A.; Yushkov, V. A.; von Hobe, M.

    2010-12-01

    During the RECONCILE campaign in the Arctic winter 2009/10, an active Match experiment was performed sampling the same air masses up to three times during two consecutive flights of the high-altitude research aircraft M55-Geophysica from Kiruna (67.83 N, 20.42 E). The first flight was westbound and its flightpath designed to resample the air masses from the outbound leg during the return to Kiruna with a time difference of up to 3 hours. Another match was attempted during a second flight 72 hours later when the air masses had moved into the Geophysica's range again. Flightplans were designed using trajectory calculations driven by ECMWF wind fields. In situ measurements of N2O and NOy revealed strong gradients inside the vortex thus allowing us to examine the accuracy of such trajectory calculations with wind fields in different spatial and temporal resolution.

  10. 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)

  11. THE METHODICAL BASIS OF INTEGRAL INDEX CALCULATION OF AIR COMPANY STABLE FINANCIAL DEVELOPMENT

    Directory of Open Access Journals (Sweden)

    Tatjana S. Rotar

    2015-01-01

    Full Text Available Based on the earlier published materials of authors about company’s stable development and pointing out the methods of integral index calculation of stable development, the following article is devoted to analysis of different approaches of financial stability definition.The review of the chosen methods of financial stability evaluation shows that all authors suggest the same set of indicators for evaluation, but at the same time each scientist offers his own methods for their calculation.The work also includes calculation of stable financial development indexes on the example of three Russian leading air companies.

  12. Direct calculation of acoustic streaming including the boundary layer phenomena in an ultrasonic air pump

    Science.gov (United States)

    Wada, Yuji; Koyama, Daisuke; Nakamura, Kentaro

    2012-05-01

    Direct finite difference fluid simulation of acoustic streaming on the fine-meshed three-dimensiona model by graphics processing unit (GPU)-oriented calculation array is discussed. Airflows due to the acoustic traveling wave are induced when an intense sound field is generated in a gap between a bending transducer and a reflector. Calculation results showed good agreement with the measurements in the pressure distribution. In addition to that, several flow-vortices were observed near the boundary of the reflector and the transducer, which have been often discussed in acoustic tube near the boundary, and have never been observed in the calculation in the ultrasonic air pump of this type.

  13. PTRAC File Utilization for Calculation of Free-Air Ionization Chamber Correction Factors by MCNPX

    Science.gov (United States)

    Šolc, Jaroslav; Sochor, Vladimír

    2014-06-01

    A free-air ionization chamber is used as a standard of photon air-kerma. Several correction factors are applied to the air-kerma value. Correction factors for electron loss (kloss) and for additional ionization current caused by photon scatter (ksc), photon fluorescence (kfl), photon transmission through diaphragm edge (kdtr), and photon scatter from the surface of the diaphragm aperture (kdsc) were determined by the MCNPX code utilizing information stored in Particle Track (PTRAC) output files. Individual steps of the procedure are described and the calculated values of the correction factors are presented. The values are in agreement with the correction factors published in a literature for similar free-air chambers.

  14. 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. 

  15. 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)

  16. 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......, but compared to determine patterns in functional dependence. The method is exemplified by ensemble comparison of reaction enthalpy to methanol and formic acid depicted in Figure 1. The functional dependence on the calculated reaction enthalpy to methanol is twice as large as that to formic acid. This suggests...... 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...

  17. Preequilibrium and statistical model calculations for neutron activation cross sections on titanium isotopes

    Energy Technology Data Exchange (ETDEWEB)

    Ivascu, M.; Avrigeanu, M.; Avrigeanu, V.

    1986-01-01

    Calculated data are presented on (n,p), (n,n'p) and (n,2n) reaction cross sections for stable titanium isotopes in the energy range from threshold up to 20 MeV. An improved preequilibrium approach, allowing a unitary use of preequilibrium and equilibrium emission parameters, has increased the agreement between calculated and experimental cross sections.

  18. Statistical model calculations of pre-scission neutron multiplicity for the heavy ion induced fusion-fission reactions with actinide target 232Th

    Directory of Open Access Journals (Sweden)

    Thakur Meenu

    2015-01-01

    Full Text Available The reaction mechanism of 19F + 232Th and 28Si + 232Th systems populating the near-super-heavy compound nuclei 251Es and 260Rf respectively are investigated using neutron multiplicity as a probe. The prescission neutron multiplicities of these compound nuclei are calculated at different excitation energies using a statistical model code. These calculations are performed using the Bohr-Wheeler transition state fission width as well as the dissipative dynamical fission width based on the Kramers’ prescription. For 19F + 232Th system, the measured yield of pre-scission is compared with the statistical model calculations for the decay of a compound nucleus in the excitation energy range of 54-90 MeV. The comparison between the measured and the calculated values indicates that the Bohr-Wheeler fission width underestimates the pre-scission neutron yield and a large amount of dissipation strength is required to reproduce the experimental pre-scission neutron multiplicities. The excitation energy dependence of the fitted values of the dissipation coefficient is also discussed. In addition, exploratory statistical model calculations of pre-scission neutron multiplicity for the 28Si + 232Th system are presented in the above range of excitation energy.

  19. Study of the accumulation of air pollution by the biological indicators, using 14 MeV neutron activation

    Science.gov (United States)

    Senhou, A.; Khoukhi, T. El; Chouak, A.; Cherkaoui, R. El Moursili; Yahiaoui, A. El; Lferde, M.

    2001-06-01

    14 MeV neutron activation analysis was used to determine air polluting elements in samples of mosses, lichens and tree barks, collected from different regions in Morocco. The analysis of spectra shows clearly that the elements Mg, Al, Si, Cl, J, Ca, Ti and Fe can easily be determined by 14 NAA with good precision, while results for Zn, Rb, Sr, Ba and La are less precise. Curves showing correlation between Al and Mg concentrations are given for different sites.

  20. 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

  1. Methodology for uncertainty calculation of net total cooling effect estimation for rating room air conditioners and packaged terminal air conditioners

    Energy Technology Data Exchange (ETDEWEB)

    Fonseca Diaz, Nestor [Universidad Tecnologica de Pereira, Facultad de Ingenieria Mecanica, Pereira (Colombia); University of Liege, Campus du Sart Tilman, Bat: B49, P33, B-4000 Liege (Belgium)

    2009-09-15

    This article presents the general procedure for uncertainty calculation of net total cooling effect estimation for rating room air conditioners and packaged terminal air conditioners, by means of measurements carried out in a test bench specially designed for this purpose. The uncertainty analysis presented in this work looks for establishing a confidence degree or certainty of experimental results. It is particularly important considering that international standards related to this type of analysis are too ambiguous when treating this subject. The uncertainty analysis is on the other hand an indispensable requirement to international standard ISO 17025 [ISO, 2005. International Standard. 17025. General Requirement to Test and Calibration Laboratories Competences. International Organization for Standardization, Geneva.], which must be applied to obtain the required quality levels according to the Word Trade Organization WTO. (author)

  2. 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.

  3. Reaction Matrix Calculations in Neutron Matter with Alternating-Layer-Spin Structure under π0 Condensation. II ---Numerical Results---

    Science.gov (United States)

    Tamiya, K.; Tamagaki, R.

    1981-10-01

    Results obtained by applying a formulation based on the reaction matrix theory developed in I are given. Calculations by making use of a modified realistic potential, the Reid soft-core potential with the OPEP-part enhanced due to the isobar (Δ)-mixing, show that the transition to the [ALS] phase of quasi-neutrons corresponding to a typical π0 condensation occurs in the region of (2 ˜ 3) times the nuclear density. The most important ingredients responsible for this transition are the growth of the attractive 3P2 + 3F2 contribution mainly from the spin-parallel pairs in the same leyers and the reduction of the repulsive 3P1 contribution mainly from the spin-antiparallel pairs in the nearest layers; these mainfest themselves as the [ALS]-type localization develops. Properties of the matter under the new phase thus obtained such as the shape of the Fermi surface and the effective mass are discussed.

  4. Integrated doses calculation in evacuation scenarios of the neutron generator facility at Missouri S&T

    Science.gov (United States)

    Sharma, Manish K.; Alajo, Ayodeji B.

    2016-08-01

    Any source of ionizing radiations could lead to considerable dose acquisition to individuals in a nuclear facility. Evacuation may be required when elevated levels of radiation is detected within a facility. In this situation, individuals are more likely to take the closest exit. This may not be the most expedient decision as it may lead to higher dose acquisition. The strategy followed in preventing large dose acquisitions should be predicated on the path that offers least dose acquisition. In this work, the neutron generator facility at Missouri University of Science and Technology was analyzed. The Monte Carlo N-Particle (MCNP) radiation transport code was used to model the entire floor of the generator's building. The simulated dose rates in the hallways were used to estimate the integrated doses for different paths leading to exits. It was shown that shortest path did not always lead to minimum dose acquisition and the approach was successful in predicting the expedient path as opposed to the approach of taking the nearest exit.

  5. Integrated doses calculation in evacuation scenarios of the neutron generator facility at Missouri S&T

    Energy Technology Data Exchange (ETDEWEB)

    Sharma, Manish K.; Alajo, Ayodeji B., E-mail: alajoa@mst.edu

    2016-08-11

    Any source of ionizing radiations could lead to considerable dose acquisition to individuals in a nuclear facility. Evacuation may be required when elevated levels of radiation is detected within a facility. In this situation, individuals are more likely to take the closest exit. This may not be the most expedient decision as it may lead to higher dose acquisition. The strategy followed in preventing large dose acquisitions should be predicated on the path that offers least dose acquisition. In this work, the neutron generator facility at Missouri University of Science and Technology was analyzed. The Monte Carlo N-Particle (MCNP) radiation transport code was used to model the entire floor of the generator's building. The simulated dose rates in the hallways were used to estimate the integrated doses for different paths leading to exits. It was shown that shortest path did not always lead to minimum dose acquisition and the approach was successful in predicting the expedient path as opposed to the approach of taking the nearest exit.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-05-17

    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

  7. Status of benchmark calculations of the neutron characteristics of the cascade molten salt ADS for the nuclear waste incineration

    Energy Technology Data Exchange (ETDEWEB)

    Dudnikov, A.A.; Alekseev, P.N.; Subbotin, S.A.; Vasiliev, A.V.; Abagyan, L.P.; Alexeyev, N.I.; Gomin, E.A.; Ponomarev, L.I.; Kolyaskin, O.E.; Men' shikov, L.I. [Russian Research Centre Kurchatov Inst., Moscow (Russian Federation); Kolesov, V.F.; Ivanin, I.A.; Zavialov, N.V. [Russian Federal Nuclear Center, RFNC-VNIIEF, Nizhnii Novgorod region (Russian Federation)

    2001-07-01

    The facility for incineration of long-lived minor actinides and some dangerous fission products should be an important feature of the future nuclear power (NP). For many reasons the liquid-fuel reactor driven by accelerator can be considered as the perspective reactor- burner for radioactive waste. The fuel of such reactor is the fluoride molten salt composition with minor actinides (Np, Cm, Am) and some fission products ({sup 99}Tc, {sup 129}I, etc.). Preliminary analysis shows that the values of keff, calculated with different codes and nuclear data differ up to several percents for such fuel compositions. Reliable critical and subcritical benchmark experiments with molten salt fuel compositions with significant quantities of minor actinides are absent. One of the main tasks for the numerical study of this problem is the estimation of nuclear data for such fuel compositions and verification of the different numerical codes used for the calculation of keff, neutron spectra and reaction rates. It is especially important for the resonance region where experimental data are poor or absent. The calculation benchmark of the cascade subcritical molten salt reactor is developed. For the chosen nuclear fuel composition the comparison of the results obtained by three different Monte-Carlo codes (MCNP4A, MCU, and C95) using three different nuclear data libraries are presented. This report concerns the investigation of subcritical molten salt reactor unit main peculiarities carried out at the beginning of ISTC project 1486. (author)

  8. 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.

  9. Calculation of Temperature Distribution in Capsule for Neutron Exposure of the Cold Moderator Materials

    CERN Document Server

    Ro Du Min

    2004-01-01

    Methods and results of the numerical calculation of temperature distribution in the spherical segmented small capsule filled with heat-generating substance are presented. Variable finite-difference method allowed one to evaluate a small drop of temperature near the boundary between the filling substance and the thermocouple installed inside the capsule, which originates from the difference in thermal conductivity.

  10. 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)

  11. Two-dimensional DORT discrete ordinates X-Y geometry neutron flux calculations for the Halden Heavy Boiling Water Reactor core configurations

    Energy Technology Data Exchange (ETDEWEB)

    Slater, C.O.

    1990-07-01

    Results are reported for two-dimensional discrete ordinates, X-Y geometry calculations performed for seven Halden Heavy Boiling Water Reactor core configurations. The calculations were performed in support of an effort to reassess the neutron fluence received by the reactor vessel. Nickel foil measurement data indicated considerable underprediction of fluences by the previously used multigroup removal- diffusion method. Therefore, calculations by a more accurate method were deemed appropriate. For each core configuration, data are presented for (1) integral fluxes in the core and near the vessel wall, (2) neutron spectra at selected locations, (3) isoflux contours superimposed on the geometry models, (4) plots of the geometry models, and (5) input for the calculations. The initial calculations were performed with several mesh sizes. Comparisons of the results from these calculations indicated that the uncertainty in the calculated fluxes should be less than 10%. However, three-dimensional effects (such as axial asymmetry in the fuel loading) could contribute to much greater uncertainty in the calculated neutron fluxes. 7 refs., 22 figs., 11 tabs.

  12. Calculated irradiation response of materials using fission reactor (HFIR, ORR, and EBR-II) neutron spectra

    Energy Technology Data Exchange (ETDEWEB)

    Gabriel, T.A.; Bishop, B.L.; Wiffen, F.W.

    1979-08-01

    In order to plan radiation damage experiments in fission reactors keyed toward fusion reactor applications, it is necessary to have available for these facilities displacement per atom (dpa) and gas production rates for many potential materials. This report supplies such data for the elemental constituents of alloys of interest to the United States fusion reactor alloy development program. The calculations are presented for positions of interest in the HFIR, ORR, and EBR-II reactors. DPA and gas production rates in alloys of interest can be synthesized from these results.

  13. Prediction of in-phantom dose distribution using in-air neutron beam characteristics for BNCS

    Energy Technology Data Exchange (ETDEWEB)

    Verbeke, Jerome M.

    1999-12-14

    A monoenergetic neutron beam simulation study is carried out to determine the optimal neutron energy range for treatment of rheumatoid arthritis using radiation synovectomy. The goal of the treatment is the ablation of diseased synovial membranes in joints, such as knees and fingers. This study focuses on human knee joints. Two figures-of-merit are used to measure the neutron beam quality, the ratio of the synovium absorbed dose to the skin absorbed dose, and the ratio of the synovium absorbed dose to the bone absorbed dose. It was found that (a) thermal neutron beams are optimal for treatment, (b) similar absorbed dose rates and therapeutic ratios are obtained with monodirectional and isotropic neutron beams. Computation of the dose distribution in a human knee requires the simulation of particle transport from the neutron source to the knee phantom through the moderator. A method was developed to predict the dose distribution in a knee phantom from any neutron and photon beam spectra incident on the knee. This method was revealed to be reasonably accurate and enabled one to reduce by a factor of 10 the particle transport simulation time by modeling the moderator only.

  14. 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

  15. 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.

  16. Calculation of the upper flammability limit of methane/air mixtures at elevated pressures and temperatures.

    Science.gov (United States)

    Van den Schoor, F; Verplaetsen, F; Berghmans, J

    2008-05-30

    Four different numerical methods to calculate the upper flammability limit of methane/air mixtures at initial pressures up to 10 bar and initial temperatures up to 200 degrees C are evaluated by comparison with experimental data. Planar freely propagating flames are calculated with the inclusion of a radiation heat loss term in the energy conservation equation to numerically obtain flammability limits. Three different reaction mechanisms are used in these calculations. At atmospheric pressure, the results of these calculations are satisfactory. At elevated pressures, however, large discrepancies are found. The spherically expanding flame calculations only show a marginal improvement compared with the planar flame calculations. On the other hand, the application of a limiting burning velocity with a pressure dependence Su,lim approximately p(-1/2) is found to predict the pressure dependence of the upper flammability limit very well, whereas the application of a constant limiting flame temperature is found to slightly underestimate the temperature dependence of the upper flammability limit.

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

    Directory of Open Access Journals (Sweden)

    J. Lelieveld

    2013-03-01

    Full Text Available Air pollution by fine particulate matter (PM2.5 and ozone (O3 has increased strongly with industrialization and urbanization. We estimated 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. We carried out high-resolution global model calculations to resolve urban and industrial regions in greater detail compared to previous work. We applied a health impact function to estimate premature mortality for people of 30 yr and older, using parameters derived from epidemiological cohort studies. Our results suggest that especially in large countries with extensive suburban and rural populations, air pollution-induced mortality rates have previously been underestimated. We calculate a global respiratory mortality of about 773 thousand yr−1 (YLL ≈ 5.2 million yr−1, 186 thousand yr−1 by lung cancer (YLL ≈ 1.7 million yr−1 and 2.0 million yr−1 by cardiovascular disease (YLL ≈ 14.3 million yr−1. 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.

  18. Model calculations of the age of firn air across the Antarctic continent

    Directory of Open Access Journals (Sweden)

    K. A. Kaspers

    2004-01-01

    Full Text Available The age of firn air in Antarctica at pore close-off depth is only known for a few specific sites where firn air has been sampled for analyses. We present a model that calculates the age of firn air at pore close-off depth for the entire Antarctic continent. The model basically uses four meteorological parameters as input (surface temperature, pressure, accumulation rate and wind speed. Using parameterisations for surface snow density, pore close-off density and tortuosity, in combination with a density-depth model and data of a regional atmospheric climate model, distribution of pore close-off depth for the entire Antarctic continent is determined. The deepest pore close-off depth was found for the East Antarctic Plateau near 72° E, 82° S, at 150±15 m (2σ. A firn air diffusion model was applied to calculate the age of CO2 at pore close-off depth. The results predict that the oldest firn gas (CO2 age is located between Dome Fuji, Dome Argos and Vostok at 43° E, 78° S being 148±23 (1σ or 38 for 2σ years old. At this location an atmospheric trace gas record should be obtained. In this study we show that methyl chloride could be recorded with a predicted length of 125 years as an example for trace gas records at this location. The longest record currently available from firn air is derived at South Pole, being 80 years. Sensitivity tests reveal that the locations with old firn air (East Antarctic Plateau have an estimated uncertainty (2σ for the modelled CO2 age at pore close-off depth of 30% and of about 40% for locations with younger firn air (CO2 age typically 40 years. Comparing the modelled age of CO2 at pore close-off depth with directly determined ages at seven sites yielded a correlation coefficient of 0.90 and a slope close to 1, suggesting a high level of confidence for the modelled results in spite of considerable remaining uncertainties.

  19. The power distribution and neutron fluence measurements and calculations in the VVER-1000 Mock-Up on the LR-0 research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kostal, M.; Juricek, V.; Rypar, V.; Svadlenkova, M. [Research Center Rez Ltd., 250 68 Husinec-Rez 130 (Czech Republic); Cvachovec, F. [Univ. of Defence, Kounicova 65, 662 10 Brno (Czech Republic)

    2011-07-01

    The power density distribution in a reactor has significant influence on core structures and pressure vessel mechanical resistance, as well as on the physical characteristics of nuclear fuel. This quantity also has an effect on the leakage neutron and photon field. This issue has become of increasing importance, as it touches on actual questions of the VVER nuclear power plant life time extension. This paper shows the comparison of calculated and experimentally determined pin by pin power distributions. The calculation has been performed with deterministic and Monte Carlo approaches. This quantity is accompanied by the neutron and photon flux density calculation and measurements at different points of the light water zero-power (LR-0) research reactor mock-up core, reactor built-in component (core barrel), and reactor pressure vessel and model. The effect of the different data libraries used for calculation is discussed. (authors)

  20. 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.

  1. Monte Carlo calculation for the development of a BNCT neutron source (1eV-10KeV) using MCNP code.

    Science.gov (United States)

    El Moussaoui, F; El Bardouni, T; Azahra, M; Kamili, A; Boukhal, H

    2008-09-01

    Different materials have been studied in order to produce the epithermal neutron beam between 1eV and 10KeV, which are extensively used to irradiate patients with brain tumors such as GBM. For this purpose, we have studied three different neutrons moderators (H(2)O, D(2)O and BeO) and their combinations, four reflectors (Al(2)O(3), C, Bi, and Pb) and two filters (Cd and Bi). Results of calculation showed that the best obtained assembly configuration corresponds to the combination of the three moderators H(2)O, BeO and D(2)O jointly to Al(2)O(3) reflector and two filter Cd+Bi optimize the spectrum of the epithermal neutron at 72%, and minimize the thermal neutron to 4% and thus it can be used to treat the deep tumor brain. The calculations have been performed by means of the Monte Carlo N (particle code MCNP 5C). Our results strongly encourage further studying of irradiation of the head with epithermal neutron fields.

  2. Consistency of neutron cross-section data, S /SUB N/ calculations, and measured tritium production for a 14-MeV neutron-driven sphere of natural lithium deuteride

    Energy Technology Data Exchange (ETDEWEB)

    Reupke, W.A.; Davidson, J.N.; Muir, D.W.

    1982-12-01

    The authors present algorithms, describe a computer program, and gives a computational procedure for the statistical consistency analysis of neutron cross-section data, S /SUB N/ calculations, and measured tritium production in 14-MeV neutron-driven integral assemblies. Algorithms presented include a reduced matrix manipulation technique suitable for manygroup, 14-MeV neutron transport calculations. The computer program incorporates these algorithms and is expanded and improved to facilitate analysis of such integral experiments. Details of the computational procedure are given for a natural lithium deuteride experiment performed at the Los Alamos National Laboratory. Results are explained in terms of calculated cross-section sensitivities and uncertainty estimates. They include a downward adjustment of the /sup 7/Li(n,xt) 14-MeV cross section from 328 + or - 22 to 284 + or - 24 mb, which is supported by the trend of recent differential and integral measurements. It is concluded that with appropriate refinements, the techniques of consistency analysis can be usefully applied to the analysis of 14-MeV neutron-driven tritium production integral experiments.

  3. Neutron shielding performance of water-extended polyester

    Energy Technology Data Exchange (ETDEWEB)

    Vega Carrillo, H.R.; Manzanares-Acuna, E.; Hernandez-Davila, V.M. [Zacatecas Univ. Autonoma, Nuclear Studies (Mexico); Vega Carrillo, H.R.; Hernandez-Davila, V.M. [Zacatecas Univ. Autonoma, Electric Engineering Academic Units (Mexico); Gallego, E.; Lorente, A. [Madrid Univ. Politecnica, cNuclear Engineering Department (Mexico)

    2006-07-01

    A Monte Carlo study to determine the shielding features to neutrons of water-extended polyester (WEP) was carried out. Materials with low atomic number are predominantly used for neutron shielding because these materials effectively attenuate neutrons, mainly through elastic and inelastic collisions. In addition to neutron attenuation properties, other desirable properties for neutron shielding materials include mechanical strength, stability, low cost, and ease of handling. During the selection of materials to design a neutron shield, prompt gamma production as well as radionuclide induced by neutron activation must be considered. In this investigation the Monte Carlo method (MCNP code) 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, for comparison the calculations were extended to water shielding, the bare source in vacuum and in air. (authors)

  4. Thermodynamic characteristics of air masses along the Guadalquivir valley determined through the calculation of trajectories

    Directory of Open Access Journals (Sweden)

    M. A. Hernández-Ceballos

    2011-01-01

    Full Text Available The Guadalquivir valley favors the channeling of air masses from coastal areas to inland Andalusia. This paper presents a first approximation of the spatial variation along the Guadalquivir valley in some of the representative thermodynamic properties of air masses. We have selected three representative sites of its lower, middle and high course, analyzing all of them on their daily trajectories and hourly records of potential temperature, specific humidity and wind speed during the period 2000-2007. The set of trajectories has been calculated using the HYSPLIT model (Hybrid Single-Particle Lagrangian Integrated Trajectory, establishing 12 UTC as the arrivaltime, a duration of 120 hours and a final height of incidence of 500 m. The cluster analysis has allowed the selection of ten different types of air masses, and those with a clear origin from the west were selected from this group. Analysis in the three sites of the daily cycles of potential temperature show a gradual cooling (3-4 K during the cold period (November-February of the year and warming during the warm period (June-September in the range of 5-6 K between the ends of the valley. The specific humidity experiences a drop, regardless of the period and type of air mass, as the air mass travels through the valley, being more intense during the warm period with up to 8 g kg-1 instead of the 1-2 g kg-1 in the cold period. The wind speed cycles show a progressive drop of intensity along the valley, more marked in the final section with a reduction of up to 3 m s-1 per 100 km, the more intense values being recorded during the warm period of the year with average values of up to 4 m s-1.

  5. Reaction Matrix Calculations in Neutron Matter with Alternating-Layer-Spin Structure under π0 Condensation. I ---Formulation---

    Science.gov (United States)

    Tamiya, K.; Tamagaki, R.

    1981-09-01

    Based on the viewpoint that a typical π0 condensation is realized with the [ALS] (Alternating-Layer-Spin) structure of nucleon system, a framework to calculate the energy of neutron matter under such a new phase is presented in the reaction matrix theory. This enables us to treat both effects on equal footing; the long-range effect dominated by the OPEP tensor component with the enhancement due to the mixing of Δ(1236MeV) and the sort-range effect much influenced by repulsive core and spin-orbit force. Starting with the [ALS] model wave function constructed on the Bloch basis which assures to take the limit of no localization, we have the expressions for energy quantities expressed by the partial-wave contributions. This scheme provides a way to understand the mechanism of energy gain in the new phase, by making use of the notions of the ordinary unclear matter theory such as the potential picture and the partial waves. Some numerical examples are shown.

  6. 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)

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

    Science.gov (United States)

    Lelieveld, J.; Barlas, C.; Giannadaki, D.; Pozzer, A.

    2013-07-01

    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.

  8. 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.

  9. 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.

  10. Design calculations of an epithermal neutron beam and development of a treatment planning system for the renovation of thor for boron neutron capture therapy

    Energy Technology Data Exchange (ETDEWEB)

    Liu, Y-W H.; Teng, Y.H.; Liao, M.Z. [National Tsing Hua Univ., Department of Engineering and System Science, Taiwan (China)

    2000-10-01

    Tsing Hua University was recently granted by National Science Council a five-year project to renovate its Open-Pool reactor (THOR) for boron neutron capture therapy. With this support, the whole graphite blocks in the original thermal column region can be removed for redesigning and constructing a better epithermal neutron beam. THOR is a 1 MW research reactor. The cross section area of the core facing the thermal column is 60 cm x 50 cm. By using 60 cm FLUENTAL plus 10 cm Pb, with cross section area of 70 cm x 60 cm and surrounded by 6 cm thick PbF{sub 2} reflector, the epithermal neutron flux at the filter/moderator exit can reach {approx}8.5 x 10{sup 9} n/cm{sup 2}/s. When the collimator is added, the epithermal neutron beam intensity at the beam exit is reduced to 3 x 10{sup 9} n/cm{sup 2}/sec, but is still six times higher than the previous beam. Facing the clinical trials scheduled 3 and half years from now, a preliminary version of treatment planning system is developed. It includes a pre-processor to read CT scan and post-processors to display dose distributions. (author)

  11. 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.

  12. Recoil proton, alpha particle, and heavy ion impacts on microdosimetry and RBE of fast neutrons: analysis of kerma spectra calculated by Monte Carlo simulation

    Energy Technology Data Exchange (ETDEWEB)

    Pignol, J.-P. [Toronto-Sunnybrook Regional Cancer Centre, Radiotherapy Dept., Toronto, Ontario (Canada); Slabbert, J. [National Accelerator Centre, Faure (South Africa)

    2001-02-01

    Fast neutrons (FN) have a higher radio-biological effectiveness (RBE) compared with photons, however the mechanism of this increase remains a controversial issue. RBE variations are seen among various FN facilities and at the same facility when different tissue depths or thicknesses of hardening filters are used. These variations lead to uncertainties in dose reporting as well as in the comparisons of clinical results. Besides radiobiology and microdosimetry, another powerful method for the characterization of FN beams is the calculation of total proton and heavy ion kerma spectra. FLUKA and MCNP Monte Carlo code were used to simulate these kerma spectra following a set of microdosimetry measurements performed at the National Accelerator Centre. The calculated spectra confirmed major classical statements: RBE increase is linked to both slow energy protons and alpha particles yielded by (n,{alpha}) reactions on carbon and oxygen nuclei. The slow energy protons are produced by neutrons having an energy between 10 keV and 10 MeV, while the alpha particles are produced by neutrons having an energy between 10 keV and 15 MeV. Looking at the heavy ion kerma from <15 MeV and the proton kerma from neutrons <10 MeV, it is possible to anticipate y* and RBE trends. (author)

  13. Recoil proton, alpha particle, and heavy ion impacts on microdosimetry and RBE of fast neutrons: analysis of kerma spectra calculated by Monte Carlo simulation.

    Science.gov (United States)

    Pignol, J P; Slabbert, J

    2001-02-01

    Fast neutrons (FN) have a higher radio-biological effectiveness (RBE) compared with photons, however the mechanism of this increase remains a controversial issue. RBE variations are seen among various FN facilities and at the same facility when different tissue depths or thicknesses of hardening filters are used. These variations lead to uncertainties in dose reporting as well as in the comparisons of clinical results. Besides radiobiology and microdosimetry, another powerful method for the characterization of FN beams is the calculation of total proton and heavy ion kerma spectra. FLUKA and MCNP Monte Carlo code were used to simulate these kerma spectra following a set of microdosimetry measurements performed at the National Accelerator Centre. The calculated spectra confirmed major classical statements: RBE increase is linked to both slow energy protons and alpha particles yielded by (n,alpha) reactions on carbon and oxygen nuclei. The slow energy protons are produced by neutrons having an energy between 10 keV and 10 MeV, while the alpha particles are produced by neutrons having an energy between 10 keV and 15 MeV. Looking at the heavy ion kerma from neutrons <10 MeV, it is possible to anticipate y* and RBE trends.

  14. 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.

  15. Neutron Albedo

    CERN Document Server

    Ignatovich, V K

    2005-01-01

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

  16. Preliminary Analysis of the Multisphere Neutron Spectrometer

    Science.gov (United States)

    Goldhagen, P.; Kniss, T.; Wilson, J. W.; Singleterry, R. C.; Jones, I. W.; VanSteveninck, W.

    2003-01-01

    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 (thermal to greater than 10 GeV), total neutron fluence rate, and neutron effective dose and dose equivalent rates and their dependence on altitude and geomagnetic cutoff. The measured cosmic-ray neutron spectra have almost no thermal neutrons, a large "evaporation" peak near 1 MeV and a second broad peak near 100 MeV which contributes about 69% of the neutron effective dose. At high altitude, geomagnetic latitude has very little effect on the shape of the spectrum, but it is the dominant variable affecting neutron fluence rate, which was 8 times higher at the northernmost measurement location than it was at the southernmost. The shape of the spectrum varied only slightly with altitude from 21 km down to 12 km (56 - 201 grams per square centimeter atmospheric depth), but was significantly different on the ground. In all cases, ambient dose equivalent was greater than effective dose for cosmic-ray neutrons.

  17. Quasi-Elastic Electron-Deuteron Scattering and Calculation of Neutron Electromagnetic Form Factors at Q2 = 1.75 to 4.00 (GeV/c)2

    Institute of Scientific and Technical Information of China (English)

    N. Ghahramany; M. Vaez zadeh Asadi; G.R. Boroun

    2003-01-01

    Electric and Magnetic form factors of neutron are calculated via electron-deuteron scattering at 1.511 ~5.507 GeV energy using SLAC group data. Our results show that the neutron electric form factor is not equal to zero;rather it has a small value, indicating that in spite of the fact that total charge is almost neutral, there is a nonuniformcharge distribution within the neutron, and that magnetic form factor follows the dipole fit.

  18. Large-scale shell-model calculations for unnatural-parity high-spin states in neutron-rich Cr and Fe isotopes

    CERN Document Server

    Togashi, Tomoaki; Utsuno, Yutaka; Otsuka, Takaharu; Honma, Michio

    2014-01-01

    We investigate unnatural-parity high-spin states in neutron-rich Cr and Fe isotopes using large-scale shell-model calculations. These shell-model calculations are carried out within the model space of $fp$-shell + $0g_{9/2}$ + $1d_{5/2}$ orbits with the truncation allowing $1\\hbar\\omega$ excitation of a neutron. The effective Hamiltonian consists of GXPF1Br for $fp$-shell orbits and $V_{\\rm MU}$ with a modification for the other parts. The present shell-model calculations can describe and predict the energy levels of both natural- and unnatural-parity states up to the high-spin states in Cr and Fe isotopes with $N\\le35$. The total energy surfaces present the prolate deformations on the whole and indicate that the excitation of one neutron into the $0g_{9/2}$ orbit plays the role of enhancing the prolate deformation. For the positive(unnatural)-parity states in odd-mass Cr and Fe isotopes, their energy levels and prolate deformations indicate the decoupling limit of the particle-plus-rotor model. We have found...

  19. Structural uncertainty in air mass factor calculation for NO2 and HCHO satellite retrievals

    Science.gov (United States)

    Lorente, Alba; Folkert Boersma, K.; Yu, Huan; Dörner, Steffen; Hilboll, Andreas; Richter, Andreas; Liu, Mengyao; Lamsal, Lok N.; Barkley, Michael; De Smedt, Isabelle; Van Roozendael, Michel; Wang, Yang; Wagner, Thomas; Beirle, Steffen; Lin, Jin-Tai; Krotkov, Nickolay; Stammes, Piet; Wang, Ping; Eskes, Henk J.; Krol, Maarten

    2017-03-01

    Air mass factor (AMF) calculation is the largest source of uncertainty in NO2 and HCHO satellite retrievals in situations with enhanced trace gas concentrations in the lower troposphere. Structural uncertainty arises when different retrieval methodologies are applied within the scientific community to the same satellite observations. Here, we address the issue of AMF structural uncertainty via a detailed comparison of AMF calculation methods that are structurally different between seven retrieval groups for measurements from the Ozone Monitoring Instrument (OMI). We estimate the escalation of structural uncertainty in every sub-step of the AMF calculation process. This goes beyond the algorithm uncertainty estimates provided in state-of-the-art retrievals, which address the theoretical propagation of uncertainties for one particular retrieval algorithm only. We find that top-of-atmosphere reflectances simulated by four radiative transfer models (RTMs) (DAK, McArtim, SCIATRAN and VLIDORT) agree within 1.5 %. We find that different retrieval groups agree well in the calculations of altitude resolved AMFs from different RTMs (to within 3 %), and in the tropospheric AMFs (to within 6 %) as long as identical ancillary data (surface albedo, terrain height, cloud parameters and trace gas profile) and cloud and aerosol correction procedures are being used. Structural uncertainty increases sharply when retrieval groups use their preference for ancillary data, cloud and aerosol correction. On average, we estimate the AMF structural uncertainty to be 42 % over polluted regions and 31 % over unpolluted regions, mostly driven by substantial differences in the a priori trace gas profiles, surface albedo and cloud parameters. Sensitivity studies for one particular algorithm indicate that different cloud correction approaches result in substantial AMF differences in polluted conditions (5 to 40 % depending on cloud fraction and cloud pressure, and 11 % on average) even for low

  20. Shielding Calculation of Neutron Guide Tube in Scatter Hall%散射大厅内中子导管屏蔽计算

    Institute of Scientific and Technical Information of China (English)

    孙勇; 霍合勇; 曹超

    2013-01-01

    中子导管将冷中子束从冷源引出至散射大厅,为保证大厅工作人员的安全,提供低本底实验环境,必须设计相应的屏蔽体进行屏蔽.在已有中子导管屏蔽体初步结构设计方案的条件下,联合McStas、MCNP,采用分段计算的方法对其进行了屏蔽计算,得到了散射大厅内中子导管周围不同位置处的辐射剂量率,验证了中子导管屏蔽体结构设计方案的有效性,为进一步开展工程设计提供了依据.%The cold neutrons are guided to the scatter hall from the cold neutron source by the neutron guide tube. Designing a shielding system of the neutron guide tube is necessary for the safety of the workers and providing a low background experiment environment in the scatter hall. The primary design of the shielding system was completed. In this paper, the calculated shielding effects were presented by McStas and MCNP with the method of dividing the whole system into several sects. The results indicate that the primary design scheme of the shielding system is feasible.

  1. The inelastic neutron scattering spectra of [alpha]-3-amino-5-nitro-1,2,4-2H-triazole: Experiment and DFT calculations

    Science.gov (United States)

    Ciezak, Jennifer A.; Trevino, S. F.

    2005-02-01

    The inelastic neutron scattering (INS) spectrum of α-3-amino-5-nitro-1,2,4-triazole is presented through 1200 cm -1. A comparison of the INS spectrum with an isolated molecule B3LYP/6-311G** calculation reveals generally good frequency and intensity agreement with two notable differences in intensity. Periodic density functional theory (DFT) calculations are employed to determine whether the intermolecular hydrogen bonding is the origin of these differences between the B3LYP/6-311G** and INS spectrum.

  2. 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)

  3. 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.

  4. Calculations of neutron and photon source terms and attenuation profiles for the generic design of the SPEAR3 storage ring shield.

    Science.gov (United States)

    Rokni, S H; Khater, H; Liu, J C; Mao, S; Vincke, H

    2005-01-01

    The FLUKA Monte Carlo particle generation and transport code was used to calculate shielding requirements for the 3 GeV, 500 mA SPEAR3 storage ring at the Stanford Synchrotron Radiation Laboratory. The photon and neutron dose equivalent source term data were simulated for a 3 GeV electron beam interacting with two typical target/shielding geometries in the ring. The targets simulated are a rectangular block of 0.7 cm thick copper and a 5 cm thick iron block, both tilted at 1 degree relative to the beam direction. Attenuation profiles for neutrons and photons in concrete and lead as a function of angle at different shield thicknesses were calculated. The first, second and equilibrium attenuation lengths of photons and neutrons in the shield materials are derived from the attenuation profiles. The source term data and the attenuation lengths were then used to evaluate the shielding requirements for the ratchet walls of all front-ends of the SPEAR3 storage ring.

  5. Neutronic modelling of the reflector for the calculation of pressurized water reactors: application to EPR; Modelisation neutronique du reflecteur pour le calcul des coeurs des reacteurs nucleaires a eau pressurisee: application a l'EPR

    Energy Technology Data Exchange (ETDEWEB)

    Sandrin, Ch.

    2010-04-15

    This PhD Thesis aims to achieve a method for the modelling of the reflector surrounding the core for neutronics core calculations. This method should consider the EPR reactor specificities (steel reflector) and the increased demand in precision. In neutronics core calculations, the reflector can be represented either by albedos boundary conditions (current ratios) or by one or several media, surrounding the core, characterised by homogenized parameters. Those parameters (cross sections and diffusion coefficients) should be obtained using equivalence so that they allow a good reproduction of the reference albedos in a representative situation. During this PhD, such an equivalence method has been developed in the APOLLO-2 code with the minimization of a functional of the differences between the reference albedos and those computed with the equivalent parameters. Because of the positiveness constraints, a local minimization, such as Newton-like methods, is not always possible and we have therefore also implemented a Particle Swarm Optimization Algorithm for more than two energy groups' problems. The parameters obtained have been used in two dimensions EPR core calculations with the CRONOS-2 code for various fuel loadings in two to eight groups diffusion. Those core calculation have been validated against reference Monte-Carlo calculations and against core calculations with albedos boundary conditions. In addition to the increased easiness of utilization, the implemented equivalence method has yielded an improvement of the results for the two groups calculation. With a higher energy groups number, the use of a unique equivalent reflector does not account correctly for the two dimensions effects; a modelling with different reflector meshes has improved the results. The modelling of the reflector by two dimensions albedos boundary conditions is the more suited for the representation of the boundary conditions and, therefore, should the two dimensions albedos

  6. Comparison and Physical Interpretation of MCNP and TART Neutron and Gamma Monte Carlo Shielding Calculations for a Heavy-Ion ICF System

    Energy Technology Data Exchange (ETDEWEB)

    Mainardi, E.; Premuda, F.; Lee, E.

    2002-07-01

    For heavy-ion beam driven inertial fusion ''liquid-protected'' reactor designs such as HYLIFE-II, a mixture of molten salts made of F{sup 10}, Li{sup -6}, Li{sup 7} and Be{sup 9} (called flibe) allows small chambers and final-focus magnets closer to the target with superconducting coils suffering higher radiation damage, though they can stand only a certain amount of energy deposited before quenching. This work has been primarily focusing on verifying that total energy deposited by fusion neutrons and induced gamma rays remain under such limit values and the final purpose is the optimization of the shielding of the magnetic lens system from the points of view of the geometrical configuration and of the physical nature of the materials adopted. The system is analyzed in terms of six geometrical models going from simplified up to much more realistic representations of a system of 192 beam lines, each focused by six magnets. A 3-D transport calculation of the radiation penetrating through ducts, that takes into account the complexity of the system, requires Monte Carlo methods. The quantities analyzed, using the two codes MCNP and TART include: neutron mean free path and total path length dependence on energy, energy deposited by neutrons and gamma photons, values of the total fluence integrated in the whole energy range, and the neutron spectrum in different zones of the system. The technical nature of the design problem and the methodology followed were presented in a previous paper by summarizing briefly the results for the deposited energy distribution on the six focal magnets. Now a much more extensive comparison of the performances of the two codes for different configurations of the system is discussed, separating the n and {gamma} contributions, in the light of the physical interpretation of the results in terms of first flight and of scattered neutron fluxes, of primary {gamma} and of secondary {gamma} generated by inelastically scattered

  7. Neutron dosimetry; Dosimetria de neutrons

    Energy Technology Data Exchange (ETDEWEB)

    Fratin, Luciano

    1993-12-31

    A neutron irradiation facility was designed and built in order to establish a procedure for calibrating neutron monitors and dosemeters. A 185 GBq {sup 241} Am Be source of known is used as a reference source. The irradiation facility using this source in the air provides neutron dose rates between 9 nSv s{sup -1} and 0,5 {sup {mu}}Sv s{sup -1}. A calibrated 50 nSv s{sup -1} thermal neutron field is obtained by using a specially designed paraffin block in conjunction with the {sup 241} Am Be source. A Bonner multisphere spectrometer was calibrated, using a procedure based on three methods proposed by international standards. The unfold {sup 241} Am Be neutron spectrum was determined from the Bonner spheres data and resulted in a good agreement with expected values for fluence rate, dose rate and mean energy. A dosimetric system based on the electrochemical etching of CR-39 was developed for personal dosimetry. The dosemeter badge using a (n,{alpha}) converter, the etching chamber and high frequency power supply were designed and built specially for this project. The electrochemical etching (ECE) parameters used were: a 6N KOH solution, 59 deg C, 20 kV{sub pp} cm{sup -1}, 2,0 kHz, 3 hours of ECE for thermal and intermediate neutrons and 6 hours for fast neutrons. The calibration factors for thermal, intermediate and fast neutrons were determined for this personal dosemeter. The sensitivities determined for the developed dosimetric system were (1,46{+-} 0,09) 10{sup 4} tracks cm{sup -2} mSv{sup -1} for thermal neutrons, (9{+-}3) 10{sup 2} tracks cm{sup -2} mSV{sup -1} for intermediate neutrons and (26{+-}4) tracks cm{sup -2} mSv{sup -1} for fast neutrons. The lower and upper limits of detection were respectively 0,002 mSv and 0,6 mSv for thermal neutrons, 0,04 mSv and 8 mSv for intermediate neutrons and 1 mSv and 12 mSv for fast neutrons. In view of the 1990`s ICRP recommendations, it is possible to conclude that the personal dosemeter described in this work is

  8. Integrated system for production of neutronics and photonics calculational constants. Supplemental neutron-induced interactions (Z > 35): graphical, experimental data. [61,671 data points

    Energy Technology Data Exchange (ETDEWEB)

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

    1976-07-04

    This report (vol. 8) presents graphs of supplemental neutron-induced cross sections in the Experimental Cross Section Information Library (ECSIL) as of July 4, 1976. It consists of interactions where more than one data set is needed to show cross-section behavior. In contrast, vol. 7 of this UCRL-50400 series consists primarily of interactions where a single data set contains enough points to show cross-section behavior. Vol. 7 contains total, elastic, capture, and fission cross sections (along with the parameters anti ..nu.., ..cap alpha.., and eta). Volume 8 contains all other reactions. 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 35; Part B contains the plots for Z greater than 35.

  9. Integrated system for production of neutronics and photonics calculational constants. Supplemental neutron-induced interactions (Z less than or equal to 35): 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. 8) presents graphs of supplemental neutron-induced cross sections in the Experimental Cross Section Information Library (ECSIL) as of July 4, 1976. It consists of interactions where more than one data set is needed to show cross-section behavior. In contrast, Vol. 7 of this UCRL-50400 series consists primarily of interactions where a single data set contains enough points to show cross-section behavior. In Vol. 7 can be found the total, elastic, capture, and fission cross sections (along with the parameters anti ..nu.., ..cap alpha.., and eta). Volume 8 contains all other reactions. 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 35; Part B contains the plots for Z greater than 35.

  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. 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.

  12. Detection of thermal neutrons with the PRISMA-YBJ array in extensive air showers selected by the ARGO-YBJ experiment

    Science.gov (United States)

    Bartoli, B.; Bernardini, P.; Bi, X. J.; Cao, Z.; Catalanotti, S.; Chen, S. Z.; Chen, T. L.; Cui, S. W.; Dai, B. Z.; D'Amone, A.; Danzengluobu; De Mitri, I.; D'Ettorre Piazzoli, B.; Di Girolamo, T.; Di Sciascio, G.; Feng, C. F.; Feng, Zhaoyang; Feng, Zhenyong; Gou, Q. B.; Guo, Y. Q.; He, H. H.; Hu, Haibing; Hu, Hongbo; Iacovacci, M.; Iuppa, R.; Jia, H. Y.; Labaciren; Li, H. J.; Liu, C.; Liu, J.; Liu, M. Y.; Lu, H.; Ma, L. L.; Ma, X. H.; Mancarella, G.; Mari, S. M.; Marsella, G.; Mastroianni, S.; Montini, P.; Ning, C. C.; Perrone, L.; Pistilli, P.; Salvini, P.; Santonico, R.; Shen, P. R.; Sheng, X. D.; Shi, F.; Surdo, A.; Tan, Y. H.; Vallania, P.; Vernetto, S.; Vigorito, C.; Wang, H.; Wu, C. Y.; Wu, H. R.; Xue, L.; Yang, Q. Y.; Yang, X. C.; Yao, Z. G.; Yuan, A. F.; Zha, M.; Zhang, H. M.; Zhang, L.; Zhang, X. Y.; Zhang, Y.; Zhao, J.; Zhaxiciren; Zhaxisangzhu; Zhou, X. X.; Zhu, F. R.; Zhu, Q. Q.; Stenkin, Yu. V.; Alekseenko, V. V.; Aynutdinov, V.; Cai, Z. Y.; Guo, X. W.; Liu, Y.; Rulev, V.; Shchegolev, O. B.; Stepanov, V.; Volchenko, V.; Zhang, H.

    2016-08-01

    We report on a measurement of thermal neutrons, generated by the hadronic component of extensive air showers (EAS), by means of a small array of EN-detectors developed for the PRISMA project (PRImary Spectrum Measurement Array), novel devices based on a compound alloy of ZnS(Ag) and 6LiF. This array has been operated within the ARGO-YBJ experiment at the high altitude Cosmic Ray Observatory in Yangbajing (Tibet, 4300 m a.s.l.). Due to the tight correlation between the air shower hadrons and thermal neutrons, this technique can be envisaged as a simple way to estimate the number of high energy hadrons in EAS. Coincident events generated by primary cosmic rays of energies greater than 100 TeV have been selected and analyzed. The EN-detectors have been used to record simultaneously thermal neutrons and the air shower electromagnetic component. The density distributions of both components and the total number of thermal neutrons have been measured. The correlation of these data with the measurements carried out by ARGO-YBJ confirms the excellent performance of the EN-detector.

  13. Detection of thermal neutrons with the PRISMA-YBJ array in Extensive Air Showers selected by the ARGO-YBJ experiment

    CERN Document Server

    Bartoli, B; Bi, X J; Cao, Z; Catalanotti, S; Chen, S Z; Chen, T L; Cui, S W; Dai, B Z; D'Amone, A; Danzengluobu,; De Mitri, I; Piazzoli, B D'Ettorre; Di Girolamo, T; Di Sciascio, G; Feng, C F; Feng, Zhaoyang; Feng, Zhenyong; Gou, Q B; Guo, Y Q; He, H H; Hu, Haibing; Hu, Hongbo; Iacovacci, M; Iuppa, R; Jia, H Y; Labaciren,; Li, H J; Liu, C; Liu, J; Liu, M Y; Lu, H; Ma, L L; Ma, X H; Mancarella, G; Mari, S M; Marsella, G; Mastroianni, S; Montini, P; Ning, C C; Perrone, L; Pistilli, P; Salvini, P; Santonico, R; Shen, P R; Sheng, X D; Shi, F; Surdo, A; Tan, Y H; Vallania, P; Vernetto, S; Vigorito, C; Wang, H; Wu, C Y; Wu, H R; Xue, L; Yang, Q Y; Yang, X C; Yao, Z G; Yuan, A F; Zha, M; Zhang, H M; Zhang, L; Zhang, X Y; Zhang, Y; Zhao, J; Zhaxiciren,; Zhaxisangzhu,; Zhou, X X; Zhu, F R; Zhu, Q Q; Stenkin, Yu V; Alekseenko, V V; Aynutdinov, V; Cai, Z Y; Guo, X W; Liu, Y; Rulev, V; Shchegolev, O B; Stepanov, V; Volchenko, V; Zhang, H

    2015-01-01

    We report on a measurement of thermal neutrons, generated by the hadronic component of extensive air showers (EAS), by means of a small array of EN-detectors developed for the PRISMA project (PRImary Spectrum Measurement Array), novel devices based on a compound alloy of ZnS(Ag) and 6LiF. This array has been operated within the ARGO-YBJ experiment at the high altitude Cosmic Ray Observatory in Yangbajing (Tibet, 4300 m a.s.l.). Due to the tight correlation between the air shower hadrons and thermal neutrons, this technique can be envisaged as a simple way to get information on the EAS hadronic component, avoiding the use of huge calorimeters. Coincident events generated by primary cosmic rays of energies greater than 100 TeV have been selected and analyzed. The EN-detectors have been used to record simultaneously thermal neutrons and the air shower electromagnetic component. The density distribution of both components and the total number of thermal neutrons have been measured. The correlation of these data w...

  14. 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.

  15. Turbulent Transfer Coefficients and Calculation of Air Temperature inside Tall Grass Canopies in Land Atmosphere Schemes for Environmental Modeling.

    Science.gov (United States)

    Mihailovic, D. T.; Alapaty, K.; Lalic, B.; Arsenic, I.; Rajkovic, B.; Malinovic, S.

    2004-10-01

    A method for estimating profiles of turbulent transfer coefficients inside a vegetation canopy and their use in calculating the air temperature inside tall grass canopies in land surface schemes for environmental modeling is presented. The proposed method, based on K theory, is assessed using data measured in a maize canopy. The air temperature inside the canopy is determined diagnostically by a method based on detailed consideration of 1) calculations of turbulent fluxes, 2) the shape of the wind and turbulent transfer coefficient profiles, and 3) calculation of the aerodynamic resistances inside tall grass canopies. An expression for calculating the turbulent transfer coefficient inside sparse tall grass canopies is also suggested, including modification of the corresponding equation for the wind profile inside the canopy. The proposed calculations of K-theory parameters are tested using the Land Air Parameterization Scheme (LAPS). Model outputs of air temperature inside the canopy for 8 17 July 2002 are compared with micrometeorological measurements inside a sunflower field at the Rimski Sancevi experimental site (Serbia). To demonstrate how changes in the specification of canopy density affect the simulation of air temperature inside tall grass canopies and, thus, alter the growth of PBL height, numerical experiments are performed with LAPS coupled with a one-dimensional PBL model over a sunflower field. To examine how the turbulent transfer coefficient inside tall grass canopies over a large domain represents the influence of the underlying surface on the air layer above, sensitivity tests are performed using a coupled system consisting of the NCEP Nonhydrostatic Mesoscale Model and LAPS.

  16. Development of a computer code for neutronic calculations of a hexagonal lattice of nuclear reactor using the flux expansion nodal method

    Directory of Open Access Journals (Sweden)

    Mohammadnia Meysam

    2013-01-01

    Full Text Available The flux expansion nodal method is a suitable method for considering nodalization effects in node corners. In this paper we used this method to solve the intra-nodal flux analytically. Then, a computer code, named MA.CODE, was developed using the C# programming language. The code is capable of reactor core calculations for hexagonal geometries in two energy groups and three dimensions. The MA.CODE imports two group constants from the WIMS code and calculates the effective multiplication factor, thermal and fast neutron flux in three dimensions, power density, reactivity, and the power peaking factor of each fuel assembly. Some of the code's merits are low calculation time and a user friendly interface. MA.CODE results showed good agreement with IAEA benchmarks, i. e. AER-FCM-101 and AER-FCM-001.

  17. Calculation of Tissue-Air Ratios(TAR) in Irregularly shaped Field for Co-60 Gamma Radiation

    Energy Technology Data Exchange (ETDEWEB)

    Ji, Young Hoon [Dept. of Therapetic Radiology, Kangnam General Hospital, Seoul (Korea, Republic of)

    1989-05-15

    In order to calculate the dose on each interest point in five types of irregularly shaped fields used commonly in radiotherapy, the tissue-air ratios (TAR) in these fields for Go-60 gamma radiation were calculated using the newly devised SAR-chart. The TARs calculated from newly method of using the SAR-chart, computer method and approximation method at the interest point were compared to the TARs obtained from measurement. The result are as follows; In case of the interest points on central axis the calculated TARs in irregularly shaped fields by the above mentioned methods were well agreed within the error of , whereas for the interest points on off-axis the calculated TARs were resulted in the maximum errors of and respectively. From these results, the accuracy of calculation method of using the SAR-chart was confirmed.

  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. Benchmarking of Decay Heat Measured Values of ITER Materials Induced by 14 MeV Neutron Activation with Calculated Results by ACAB Activation Code

    Energy Technology Data Exchange (ETDEWEB)

    Tore, C.; Ortego, P.; Rodriguez Rivada, A.

    2014-07-01

    The aim of this paper is the comparison between the calculated and measured decay heat of material samples which were irradiated at the Fusion Neutron Source of JAERI in Japan with D-T production of 14MeV neutrons. In the International Thermonuclear Experimental Reactor (ITER) neutron activation of the structural material will result in a source of heat after shutdown of the reactor. The estimation of decay heat value with qualified codes and nuclear data is an important parameter for the safety analyses of fusion reactors against lost of coolant accidents. When a loss of coolant and/or flow accident happen plasma facing components are heated up by decay heat. If the temperature of the components exceeds the allowable temperature, the accident would expand to loose the integrity of ITER. Uncertainties associated with decay prediction less than 15% are strongly requested by the ITER designers. Additionally, accurate decay heat prediction is required for making reasonable shutdown scenarios of ITER. (Author)

  1. 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

  2. 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.

  3. 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

  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. 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

  6. 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.

  7. 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)

  8. 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.

  9. Neutron dosimetry, damage calculations, and helium measurements for the HFIR-MFE-60J-1 and MFE-330J-1 spectral tailoring experiments

    Energy Technology Data Exchange (ETDEWEB)

    Greenwood, L.R. [Pacific Northwest Laboratory, Richland, WA (United States); Baldwin, C.A. [Oak Ridge National Lab., TN (United States); Oliver, B.M.

    1995-04-01

    The objective is to provide dosimetry and damage analysis for fusion materials irradiation experiments. Neutron fluence measurements and radiation damage calculations are reported for the joint US -Japanese MFE-60J-1 and MFE-330J-1 experiments in the hafnium-lined removable beryllium (RB{sup *}) position of the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory. These experiments were continuations of the ORR-6J and 7J irradiations performed in the Oak Ridge Research Reactor. The combination of irradiations was designed to tailor the neutron spectrum in order to achieve fusion reactor helium/dpa levels in stainless steel. These experiments produced maximum helium (appm)/dpa(displacement per atom) levels of 10.2 at 18.5 dpa for the ORR-6J and HFIR-MFE-60J-1 combination and 11.8 at 19.0 dpa for the ORR-7J and HFIR-MFE-330J-1 combination. A helium measurement in one JPCA sample was in good agreement with helium calculations.

  10. 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.

  11. 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

  12. Effects of Fission Yield Data in the Calculation of Antineutrino Spectra for 235U (n ,fission) at Thermal and Fast Neutron Energies

    Science.gov (United States)

    Sonzogni, A. A.; McCutchan, E. A.; Johnson, T. D.; Dimitriou, P.

    2016-04-01

    Fission yields form an integral part of the prediction of antineutrino spectra generated by nuclear reactors, but little attention has been paid to the quality and reliability of the data used in current calculations. Following a critical review of the thermal and fast ENDF/B-VII.1 235U 235 fission yields, deficiencies are identified and improved yields are obtained, based on corrections of erroneous yields, consistency between decay and fission yield data, and updated isomeric ratios. These corrected yields are used to calculate antineutrino spectra using the summation method. An anomalous value for the thermal fission yield of 86Ge generates an excess of antineutrinos at 5-7 MeV, a feature which is no longer present when the corrected yields are used. Thermal spectra calculated with two distinct fission yield libraries (corrected ENDF/B and JEFF) differ by up to 6% in the 0-7 MeV energy window, allowing for a basic estimate of the uncertainty involved in the fission yield component of summation calculations. Finally, the fast neutron antineutrino spectrum is calculated, which at the moment can only be obtained with the summation method and may be relevant for short baseline reactor experiments using highly enriched uranium fuel.

  13. Effects of Fission Yield Data in the Calculation of Antineutrino Spectra for ^{235}U(n,fission) at Thermal and Fast Neutron Energies.

    Science.gov (United States)

    Sonzogni, A A; McCutchan, E A; Johnson, T D; Dimitriou, P

    2016-04-01

    Fission yields form an integral part of the prediction of antineutrino spectra generated by nuclear reactors, but little attention has been paid to the quality and reliability of the data used in current calculations. Following a critical review of the thermal and fast ENDF/B-VII.1 ^{235}U fission yields, deficiencies are identified and improved yields are obtained, based on corrections of erroneous yields, consistency between decay and fission yield data, and updated isomeric ratios. These corrected yields are used to calculate antineutrino spectra using the summation method. An anomalous value for the thermal fission yield of ^{86}Ge generates an excess of antineutrinos at 5-7 MeV, a feature which is no longer present when the corrected yields are used. Thermal spectra calculated with two distinct fission yield libraries (corrected ENDF/B and JEFF) differ by up to 6% in the 0-7 MeV energy window, allowing for a basic estimate of the uncertainty involved in the fission yield component of summation calculations. Finally, the fast neutron antineutrino spectrum is calculated, which at the moment can only be obtained with the summation method and may be relevant for short baseline reactor experiments using highly enriched uranium fuel.

  14. 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)

  15. Authentic Assessment in the Geometry Classroom: Calculating the Classroom Air-Exchange Rate.

    Science.gov (United States)

    Erich, David J.

    2002-01-01

    Introduces a room air-exchange activity designed to assess student understanding of the concept of volume. Lists materials for the activity and its procedures. Includes the lesson plan and a student worksheet. (KHR)

  16. Improved Modeling of Residential Air Conditioners and Heat Pumps for Energy Calculations

    Energy Technology Data Exchange (ETDEWEB)

    Cutler, D.; Winkler, J.; Kruis, N.; Christensen, C.; Brendemuehl, M.

    2013-01-01

    This report presents improved air conditioner and heat pump modeling methods in the context of whole-building simulation tools, with the goal of enabling more accurate evaluation of cost effective equipment upgrade opportunities and efficiency improvements in residential buildings.

  17. Contact angle and adsorption energies of nanoparticles at the air-liquid interface determined by neutron reflectivity and molecular dynamics

    Science.gov (United States)

    Reguera, Javier; Ponomarev, Evgeniy; Geue, Thomas; Stellacci, Francesco; Bresme, Fernando; Moglianetti, Mauro

    2015-03-01

    Understanding how nanomaterials interact with interfaces is essential to control their self-assembly as well as their optical, electronic, and catalytic properties. We present here an experimental approach based on neutron reflectivity (NR) that allows the in situ measurement of the contact angles of nanoparticles adsorbed at fluid interfaces. Because our method provides a route to quantify the adsorption and interfacial energies of the nanoparticles in situ, it circumvents problems associated with existing indirect methods, which rely on the transport of the monolayers to substrates for further analysis. We illustrate the method by measuring the contact angle of hydrophilic and hydrophobic gold nanoparticles, coated with perdeuterated octanethiol (d-OT) and with a mixture of d-OT and mercaptohexanol (MHol), respectively. The contact angles were also calculated via atomistic molecular dynamics (MD) computations, showing excellent agreement with the experimental data. Our method opens the route to quantify the adsorption of complex nanoparticle structures adsorbed at fluid interfaces featuring different chemical compositions.Understanding how nanomaterials interact with interfaces is essential to control their self-assembly as well as their optical, electronic, and catalytic properties. We present here an experimental approach based on neutron reflectivity (NR) that allows the in situ measurement of the contact angles of nanoparticles adsorbed at fluid interfaces. Because our method provides a route to quantify the adsorption and interfacial energies of the nanoparticles in situ, it circumvents problems associated with existing indirect methods, which rely on the transport of the monolayers to substrates for further analysis. We illustrate the method by measuring the contact angle of hydrophilic and hydrophobic gold nanoparticles, coated with perdeuterated octanethiol (d-OT) and with a mixture of d-OT and mercaptohexanol (MHol), respectively. The contact angles were

  18. 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.

  19. Comparison of direct and quasi-static methods for neutron kinetic calculations with the EDF R and D COCAGNE code

    Energy Technology Data Exchange (ETDEWEB)

    Girardi, E.; Guerin, P. [Electricite de France - RandD, 1 av. du General de Gaulle, 92141, Clamart (France); Dulla, S.; Nervo, M.; Ravetto, P. [Dipartimento di Energetica, Politecnico di Torino, 24, c.so Duca degli Abruzzi, 10129, Torino (Italy)

    2012-07-01

    Quasi-Static (QS) methods are quite popular in the reactor physics community and they exhibit two main advantages. First, these methods overcome both the limits of the Point Kinetic (PK) approach and the issues of the computational effort related to the direct discretization of the time-dependent neutron transport equation. Second, QS methods can be implemented in such a way that they can be easily coupled to very different external spatial solvers. In this paper, the results of the coupling between the QS methods developed by Politecnico di Torino and the EDF R and D core code COCAGNE are presented. The goal of these activities is to evaluate the performances of QS methods (in term of computational cost and precision) with respect to the direct kinetic solver (e.g. {theta}-scheme) already available in COCAGNE. Additionally, they allow to perform an extensive cross-validation of different kinetic models (QS and direct methods). (authors)

  20. Development code for sensitivity and uncertainty analysis of input on the MCNPX for neutronic calculation in PWR core

    Science.gov (United States)

    Hartini, Entin; Andiwijayakusuma, Dinan

    2014-09-01

    This research was carried out on the development of code for uncertainty analysis is based on a statistical approach for assessing the uncertainty input parameters. In the butn-up calculation of fuel, uncertainty analysis performed for input parameters fuel density, coolant density and fuel temperature. This calculation is performed during irradiation using Monte Carlo N-Particle Transport. The Uncertainty method based on the probabilities density function. Development code is made in python script to do coupling with MCNPX for criticality and burn-up calculations. Simulation is done by modeling the geometry of PWR terrace, with MCNPX on the power 54 MW with fuel type UO2 pellets. The calculation is done by using the data library continuous energy cross-sections ENDF / B-VI. MCNPX requires nuclear data in ACE format. Development of interfaces for obtaining nuclear data in the form of ACE format of ENDF through special process NJOY calculation to temperature changes in a certain range.

  1. Development code for sensitivity and uncertainty analysis of input on the MCNPX for neutronic calculation in PWR core

    Energy Technology Data Exchange (ETDEWEB)

    Hartini, Entin, E-mail: entin@batan.go.id; Andiwijayakusuma, Dinan, E-mail: entin@batan.go.id [Center for Development of Nuclear Informatics - National Nuclear Energy Agency, PUSPIPTEK, Serpong, Tangerang, Banten (Indonesia)

    2014-09-30

    This research was carried out on the development of code for uncertainty analysis is based on a statistical approach for assessing the uncertainty input parameters. In the butn-up calculation of fuel, uncertainty analysis performed for input parameters fuel density, coolant density and fuel temperature. This calculation is performed during irradiation using Monte Carlo N-Particle Transport. The Uncertainty method based on the probabilities density function. Development code is made in python script to do coupling with MCNPX for criticality and burn-up calculations. Simulation is done by modeling the geometry of PWR terrace, with MCNPX on the power 54 MW with fuel type UO2 pellets. The calculation is done by using the data library continuous energy cross-sections ENDF / B-VI. MCNPX requires nuclear data in ACE format. Development of interfaces for obtaining nuclear data in the form of ACE format of ENDF through special process NJOY calculation to temperature changes in a certain range.

  2. 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.

  3. Development and deployment of AQUIS: A PC-based emission inventory calculator and air information management system

    Energy Technology Data Exchange (ETDEWEB)

    Smith, A.E.; Tschanz, J.; Monarch, M.; Narducci, P.; Bormet, S.

    1995-06-01

    The Air Quality Utility Information System (AQUIS) is a database management system. AQUIS assists users in calculation emissions, both traditional and toxic, and tracking and reporting emissions and source information. With some facilities having over 1200 sources and AQUIS calculating as many as 125 pollutants for a single source, tracking and correlating this information involve considerable effort. Originally designed for use at seven facilities of the Air Force Material Command, the user community has expanded to over 50 facilities since last reported at the 1993 Air and Waste Management Association (AWMA) annual meeting. This expansion in the user community has provided an opportunity to test the system under expanded operating conditions and in applications not anticipated during original system design. User feedback is used to determine needed enhancements and features and to prioritize the content of new releases. In responding to evolving user needs and new emission calculation procedures, it has been necessary to reconfigure AQUIS several times. Reconfigurations have ranged from simple to complex. These changes have necessitated augmenting quality assurance (QA) and validation procedures.

  4. Neutron Monitors and muon detectors for solar modulation studies: Interstellar flux, yield function, and assessment of critical parameters in count rate calculations

    CERN Document Server

    Maurin, D; Derome, L; Ghelfi, A; Hubert, G

    2014-01-01

    Particles count rates at given Earth location and altitude result from the convolution of (i) the interstellar (IS) cosmic-ray fluxes outside the solar cavity, (ii) the time-dependent modulation of IS into Top-of-Atmosphere (TOA) fluxes, (iii) the rigidity cut-off (or geomagnetic transmission function) and grammage at the counter location, (iv) the atmosphere response to incoming TOA cosmic rays (shower development), and (v) the counter response to the various particles/energies in the shower. Count rates from neutron monitors or muon counters are therefore a proxy to solar activity. In this paper, we review all ingredients, discuss how their uncertainties impact count rate calculations, and how they translate into variation/uncertainties on the level of solar modulation $\\phi$ (in the simple Force-Field approximation). The main uncertainty for neutron monitors is related to the yield function. However, many other effects have a significant impact, at the 5-10% level on $\\phi$ values. We find no clear ranking...

  5. Structure-Dependent Vibrational Dynamics of Mg(BH4)2 Polymorphs Probed with Neutron Vibrational Spectroscopy and First-Principles Calculations

    Energy Technology Data Exchange (ETDEWEB)

    Dimitrievska, Mirjana; White, James L.; Zhou, Wei; Stavila, Vitalie; Klebanoff, Leonard E.; Udovic, Terrence J.

    2016-09-28

    The structure-dependent vibrational properties of different Mg(BH4)2 polymorphs (..alpha.., ..beta.., ..gamma.., and ..delta.. phases) were investigated with a combination of neutron vibrational spectroscopy (NVS) measurements and density functional theory (DFT) calculations, with emphasis placed on the effects of the local structure and orientation of the BH4- anions. DFT simulations closely match the neutron vibrational spectra. The main bands in the low-energy region (20-80 meV) are associated with the BH4- librational modes. The features in the intermediate energy region (80-120 meV) are attributed to overtones and combination bands arising from the lower-energy modes. The features in the high-energy region (120-200 meV) correspond to the BH4- symmetric and asymmetric bending vibrations, of which four peaks located at 140, 142, 160, and 172 meV are especially intense. There are noticeable intensity distribution variations in the vibrational bands for different polymorphs. This is explained by the differences in the spatial distribution of BH4- anions within various structures. An example of the possible identification of products after the hydrogenation of MgB2, using NVS measurements, is presented. These results provide fundamental insights of benefit to researchers currently studying these promising hydrogen-storage materials.

  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. Improved Modeling of Residential Air Conditioners and Heat Pumps for Energy Calculations

    Energy Technology Data Exchange (ETDEWEB)

    Cutler, D. [National Renewable Energy Lab. (NREL), Golden, CO (United States); Winkler, J. [National Renewable Energy Lab. (NREL), Golden, CO (United States); Kruis, N. [National Renewable Energy Lab. (NREL), Golden, CO (United States); Christensen, C. [National Renewable Energy Lab. (NREL), Golden, CO (United States); Brandemuehl, M. [Univ. of Colorado, Boulder, CO (United States)

    2013-01-01

    This report presents improved air conditioner and heat pump modeling methods in the context of whole-building simulation tools, with the goal of enabling more accurate evaluation of cost-effective equipment upgrade opportunities and efficiency improvements in residential buildings.

  8. Calculation of Generation Rate of Electron Ion Pairs Ionized by Radioactive Nuclide in Air

    Institute of Scientific and Technical Information of China (English)

    WU; Xiu-feng; ZHANG; Li-feng; LUO; Zhi-fu

    2015-01-01

    Alpha and beta nuclides are widely employed in industrial production and life for the ability of ionization.Static eliminator,ionization smoke detector,electron capture detector and radioactive lightning rod are some typical examples.Alpha/beta rays produce electrons by ionizing the air,and then the charge is transferred during

  9. Methodology of Fuel Burn Up Fitting in VVER-1000 Reactor Core by Using New Ex-Vessel Neutron Dosimetry and In-Core Measurements and its Application for Routine Reactor Pressure Vessel Fluence Calculations

    Directory of Open Access Journals (Sweden)

    Borodkin Pavel

    2016-01-01

    Full Text Available Paper describes the new approach of fitting axial fuel burn-up patterns in peripheral fuel assemblies of VVER-1000 type reactors, on the base of ex-core neutron leakage measurements, neutron-physical calculations and in-core SPND measured data. The developed approach uses results of new ex-vessel measurements on different power units through different reactor cycles and their uncertainties to clear the influence of a fitted fuel burn-up profile to the RPV neutron fluence calculations. The new methodology may be recommended to be included in the routine fluence calculations used in RPV lifetime management and may be taken into account during VVER-1000 core burn-up pattern correction.

  10. Methodology of Fuel Burn Up Fitting in VVER-1000 Reactor Core by Using New Ex-Vessel Neutron Dosimetry and In-Core Measurements and its Application for Routine Reactor Pressure Vessel Fluence Calculations

    Science.gov (United States)

    Borodkin, Pavel; Borodkin, Gennady; Khrennikov, Nikolay

    2016-02-01

    Paper describes the new approach of fitting axial fuel burn-up patterns in peripheral fuel assemblies of VVER-1000 type reactors, on the base of ex-core neutron leakage measurements, neutron-physical calculations and in-core SPND measured data. The developed approach uses results of new ex-vessel measurements on different power units through different reactor cycles and their uncertainties to clear the influence of a fitted fuel burn-up profile to the RPV neutron fluence calculations. The new methodology may be recommended to be included in the routine fluence calculations used in RPV lifetime management and may be taken into account during VVER-1000 core burn-up pattern correction.

  11. Comparison and physical interpretation of MCNP and TART neutron and {gamma} Monte Carlo shielding calculations for a heavy-ion ICF system

    Energy Technology Data Exchange (ETDEWEB)

    Mainardi, E. E-mail: enrico@nuc.berkeley.edu; Premuda, F.; Lee, E

    2004-01-01

    Inertial confinement fusion (ICF) aims to induce implosions of D-T pellets to obtain a extremely dense and hot plasma with lasers or heavy-ion beams. For heavy-ion fusion (HIF), recent research has focused on 'liquid-protected' designs that allow highly compact target chambers. In the design of a reactor such as HYLIFE-II [Fus. Techol. 25 (1984); HYLIFE-II Progress Report, UCID-21816, 4.82-100], the liquid used is a molten salt made of F{sup 10}, Li{sup 6}, Li{sup 7}, Be{sup 9} (called flibe). Flibe allows the final-focus magnets to be closer to the target, which helps to reduce the focus spot size and in turn the size of the driver, with a large reduction of the cost of HIF electricity. Consequently the superconducting coils of the magnets closer to the D-T neutron source will potentially suffer higher damage though they can stand only a certain amount of energy deposited before quenching. This work has been primarily focusing on verifying that total energy deposited by fusion neutrons and induced {gamma} rays remain under such limit values and the final purpose is the optimization of the shielding of the magnetic lens system from the points of view of the geometrical configuration and of the physical nature of the materials adopted. The system is analyzed in terms of six geometrical models going from simplified up to much more realistic representations of a system of 192 beam lines, each focused by six magnets. A 3-D transport calculation of the radiation penetrating through ducts, that takes into account the complexity of the system, requires Monte Carlo methods. The technical nature of the design problem and the methodology followed were presented in a previous paper [Nucl. Instr. and Meth. A 464 (2001) 410] by summarizing briefly the results for the deposited energy distribution on the six focal magnets of a beam line. Now a comparison of the performances of the two codes TART98 [TART98: A Coupled Neutron-Photon 3-D Combinational Geometry Monte Carlo

  12. Qualification of coupled 3D neutron kinetic/thermal hydraulic code systems by the calculation of a VVER-440 benchmark. Re-connection of an isolated loop

    Energy Technology Data Exchange (ETDEWEB)

    Kotsarev, Alexander; Lizorkin, Mikhail [National Research Centre ' Kurchatov Institute' , Moscow (Russian Federation); Bencik, Marek; Hadek, Jan [UJV Rez, a.s., Rez (Czech Republic); Kozmenkov, Yaroslav; Kliem, Soeren [Helmholtz-Zentrum Dresden-Rossendorf (HZDR) e.V., Dresden (Germany)

    2016-09-15

    The 7th AER dynamic benchmark is a continuation of the efforts to validate the codes systematically for the estimation of the transient behavior of VVER type nuclear power plants. The main part of the benchmark is the simulation of the re-connection of an isolated circulation loop with low temperature in a VVER-440 plant. This benchmark was calculated by the National Research Centre ''Kurchatov Institute'' (with the code ATHLET/BIPR-VVER), UJV Rez (with the code RELAP5-3D {sup copyright}) and HZDR (with the code DYN3D/ATHLET). The paper gives an overview of the behavior of the main thermal hydraulic and neutron kinetic parameters in the provided solutions.

  13. Measurement and calculation of the fast-neutron and photon spectra from the core boundary to the biological shielding in the WWER-1000 reactor model.

    Science.gov (United States)

    Osmera, B; Cvachovec, F; Kyncl, J; Smutný, V

    2005-01-01

    The fast-neutron and photon space-energy distributions have been measured in an axially (1.25 m active height) and azimuthally (60 degree symmetry sector) shortened model of the WWER-1000 reactor assembled in the LR-0 experimental reactor. The space-energy distributions have been calculated with the stochastic code MCNP and the deterministic three-dimensional code TORT. Selected results are presented and discussed in the paper. This work has been done in the frame of the EU 5th FW project REDOS REDOS, Reactor Dosimetry: Accurate determination and benchmarking of radiation field parameters, relevant for reactor pressure vessel monitoring. EURATOM Programme, Call 2000/C 294/04). All geometry and material composition data of the model as well as the available experimental data were carefully checked and revised.

  14. 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)

  15. Validation of deterministic and Monte Carlo codes for neutronics calculation of the IRT-type research reactor

    Science.gov (United States)

    Shchurovskaya, M. V.; Alferov, V. P.; Geraskin, N. I.; Radaev, A. I.

    2017-01-01

    The results of the validation of a research reactor calculation using Monte Carlo and deterministic codes against experimental data and based on code-to-code comparison are presented. The continuous energy Monte Carlo code MCU-PTR and the nodal diffusion-based deterministic code TIGRIS were used for full 3-D calculation of the IRT MEPhI research reactor. The validation included the investigations for the reactor with existing high enriched uranium (HEU, 90 w/o) fuel and low enriched uranium (LEU, 19.7 w/o, U-9%Mo) fuel.

  16. Calculation of Resolution Function for Triple-Axis Neutron Spectrometer%中子三轴谱仪分辨函数的模拟计算

    Institute of Scientific and Technical Information of China (English)

    刘丽鹃; 谢超美; 徐家云; 阳剑; 谢雷

    2009-01-01

    The theory of triple-axis neutron spectrometer (TAS) was described briefly. With MCSTAS code based on Monte-Carlo method, the resolution function was calculated at elastic scattering. The energy resolution, the scan of vanadium sample scattering experiment and the resolution function shape of different transferred energy were simulated. The relation between the choice of measuring position in real experiment and the shape of the resolution ellipse was analysed. The result shows that simulation calculation can calculate the shape of resolution function rapidly.%阐述了中子三轴谱仪的工作原理,利用MCSTAS程序模拟计算了分辨函数,分别计算了在弹性散射下能量分辨率、钒样品散射实验的转移能量的扫描和不同转移能量下的分辨函数形态.分析了实验中测量位置的选取与分辨椭圆状态的关系.结果表明,模拟计算能快速地计算出分辨函数形态.

  17. 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...

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

    Science.gov (United States)

    2013-03-01

    readiness (Operations “Aghata” and “Cruzex”), humanitarian missions (support to combat dengue in the city of Rio de Janeiro ; support of the military...disinfection, sanitary and barber shop; u) water supply ; v) water treatment; x) providing electrical power. 1.8 Organization This research will be...recreation, and water and electricity supply . 2.1.1 Cellular Unit of Intendancy. The CUI provides the Air Force with the necessary mobility to

  19. Air renewal times and ventilation rate calculations for underground workings using radioactive measurement

    Institute of Scientific and Technical Information of China (English)

    Ayman A. El-Abnoudy⇑; Sayed F. Hassan

    2016-01-01

    Potential alpha emitters are of prime concern to the ventilation engineer due to their rapid concentration increasing once radon released in the mine atmosphere, causing tissue irradiation and lung cancer. Studying of the time based variations of the natural ventilation in tunnels and their relationship to the external parameters contribute to the air circulation assessment. Due to the continuous and high fluctu-ation of the meteorological conditions affecting the air circulation and intensity through the underground workings, there is a difficulty in the natural ventilation assessment by only the ordinary meteorological measurements. So, in this paper, the possibility of using the radioactive measurements, allowing for the air aging and ventilation quality to be qualified, is investigated through three different underground structures. Referring to the most confined structure of them, results show that one structure has a better exchange rate by a factor 1.8 and the other has the best rate by a factor 2.1. This parameter can be linked to the operating costs and size of a future ventilation system.

  20. DESIGNING AND CALCULATING OF THE AIR CONVEYING CHUTE%空气输送流槽的设计计算

    Institute of Scientific and Technical Information of China (English)

    李志华

    2001-01-01

    The air conveying chute is an equipment used to convey materials by compressed air. It is installed slopingly. The conyeying principle and structure is introduced, the designing and calculating way is pointed out. It is thought that it should be popularized and applied energetically in the industry of ruber and plastics.%空气输送流槽是一种利用空气和物料混合以压缩空气作为动力来输送物料的装置,倾斜安装。介绍了空气输送流槽的输送原理和结构。提出了设计计算空气输送流槽的方法。认为在橡胶、塑料工业中应大力推广应用空气输送流槽。

  1. Application of a prospective model for calculating worker exposure due to the air pathway for operations in a laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Grimbergen, T.W.M. [Department of Occupational Health, Safety and Environment, Vrije Universiteit, Amsterdam (Netherlands); NRG Radiation and Environment, PO Box 9034, 6800 ES Arnhem (Netherlands); Wiegman, M.M. [VU Medisch Centrum, Department of Nuclear Medicine and PET-Research, Amsterdam (Netherlands)

    2007-07-01

    In order to arrive at recommendations for guidelines on maximum allowable quantities of radioactive material in laboratories, a proposed mathematical model was used for the calculation of transfer fractions for the air pathway. A set of incident scenarios was defined, including spilling, leakage and failure of the fume hood. For these 'common incidents', dose constraints of 1 mSv and 0.1 mSv are proposed in case the operations are being performed in a controlled area and supervised area, respectively. In addition, a dose constraint of 1 {mu}Sv is proposed for each operation under regular working conditions. Combining these dose constraints and the transfer fractions calculated with the proposed model, maximum allowable quantities were calculated for different laboratory operations and situations. Provided that the calculated transfer fractions can be experimentally validated and the dose constraints are acceptable, it can be concluded from the results that the dose constraint for incidents is the most restrictive one. For non-volatile materials this approach leads to quantities much larger than commonly accepted. In those cases, the results of the calculations in this study suggest that limitation of the quantity of radioactive material, which can be handled safely, should be based on other considerations than the inhalation risks. Examples of such considerations might be the level of external exposure, uncontrolled spread of radioactive material by surface contamination, emissions in the environment and severe accidents like fire. (authors)

  2. Application of a prospective model for calculating worker exposure due to the air pathway for operations in a laboratory.

    Science.gov (United States)

    Grimbergen, T W M; Wiegman, M M

    2007-01-01

    In order to arrive at recommendations for guidelines on maximum allowable quantities of radioactive material in laboratories, a proposed mathematical model was used for the calculation of transfer fractions for the air pathway. A set of incident scenarios was defined, including spilling, leakage and failure of the fume hood. For these 'common incidents', dose constraints of 1 mSv and 0.1 mSv are proposed in case the operations are being performed in a controlled area and supervised area, respectively. In addition, a dose constraint of 1 microSv is proposed for each operation under regular working conditions. Combining these dose constraints and the transfer fractions calculated with the proposed model, maximum allowable quantities were calculated for different laboratory operations and situations. Provided that the calculated transfer fractions can be experimentally validated and the dose constraints are acceptable, it can be concluded from the results that the dose constraint for incidents is the most restrictive one. For non-volatile materials this approach leads to quantities much larger than commonly accepted. In those cases, the results of the calculations in this study suggest that limitation of the quantity of radioactive material, which can be handled safely, should be based on other considerations than the inhalation risks. Examples of such considerations might be the level of external exposure, uncontrolled spread of radioactive material by surface contamination, emissions in the environment and severe accidents like fire.

  3. PENENTUAN KANDUNGAN UNSUR PADA INSTALASI PENGOLAHAN AIR LIMBAH (IPAL RSUP DR. SOERADJI TIRTONEGORO KLATEN DENGAN METODE ANALISIS AKTIVASI NEUTRON REAKTOR KARTINI

    Directory of Open Access Journals (Sweden)

    - Niati

    2012-01-01

    Full Text Available Limbah cair hasil aktivasi manusia misalnya di Rumah Sakit harus diolah terlebih dahulu sebelum dialirkan ke lingkungan. Pengolahan limbah cair ini dilakukan untuk mengantisipasi adanya suatu hal yang berbahaya atau tidak aman bagi lingkungan. Permasalahan yang dikaji adalah dengan mengetahui jenis unsur dan kadarnya apakah melebihi dari batas kadar baku mutu limbah dan air minum. Metode  Analisis Aktivasi Neutron (AAN untuk analisis kualitatif yaitu mengetahui jenis unsur dan analisis kuantitatif yaitu menghitung kadar dari jenis unsur tersebut. Sampel limbah cair diaktivasi menggunakan sumber neutron dari Reaktor Kartini, kemudian dicacah menggunakan Spektrometri-γ, barulah analisis kualitatif dan kuantitatif dapat dilakukan. Hasil penelitian sampel air sumur dan limbah cair RS secara kualitatif terdapat jenis unsur dengan waktu peluruhan pendek seperti : Fe, Cl, dan Al dan waktu peluruhan panjang terdapat jenis unsur Br dan Na. Secara kuantitatif untuk waktu peluruhan pendek dengan evaporasi kadar Cl antara (0,0849-3,01E-06 ppm, kadar Al antara (2,3197-3,9841E-07 ppm; tanpa evaporasi kadar Cl antara (0,65785-2,3197E-07 ppm, kadar Al antara (2,5113-2,7761E-09 ppm. Untuk waktu peluruhan panjang dengan evaporasi kadar Br antara (0,069846-1,9147E-04 ppm, kadar Na antara (0,8058-3,2544E-05 ppm; tanpa evaporasi kadar Br 5,031E-06 ppm, kadar Na antara (6,7857-8,3285E-07 ppm. Dari hasil penelitian tersebut dapat disimpulkan bahwa waktu peluruhan dan perbedaan perlakuan sampel mengakibatkan jenis unsur dan kadar unsur yang dihasilkan juga berbeda-beda. Berdasarkan penghitungan kadar jenis unsur dan setelah dibandingkan dengan kadar baku mutu limbah dan mutu air maka limbah cair RSUP Dr. Soeradji Tirtonegoro Klaten dalam batas aman apabila dibuang ke lingkungan dan air sumur tersebut juga aman untuk dikonsumsi. Kata kunci : Instalasi Pengolahan Air Limbah (IPAL, AAN, Reaktor Kartini

  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 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.

  6. Energy Payback Time Calculation for a Building Integrated Semitransparent Thermal (BISPVT System with Air Duct

    Directory of Open Access Journals (Sweden)

    Kanchan Mudgil

    2013-07-01

    Full Text Available This paper evaluates the energy payback time (EPBT of building integrated photovoltaic thermal (BISPVT system for Srinagar, India. Three different photovoltaic (PV modules namely mono crystalline silicon (m-Si, poly crystalline silicon (p-Si, and amorphous silicon (a-Si have been considered for calculation of EPBT. It is found that, the EPBT is lowest in m-Si. Hence, integration of m-Si PV modules on the roof of a room is economical.

  7. Accuracy of Spencer-Attix cavity theory and calculations of fluence correction factors for the air kerma formalism.

    Science.gov (United States)

    La Russa, D J; Rogers, D W O

    2009-09-01

    EGSnrc calculations of ion chamber response and Spencer-Attix (SA) restricted stopping-power ratios are used to test the assumptions of the SA cavity theory and to assess the accuracy of this theory as it applies to the air kerma formalism for 60Co beams. Consistent with previous reports, the EGSnrc calculations show that the SA cavity theory, as it is normally applied, requires a correction for the perturbation of the charged particle fluence (K(fl)) by the presence of the cavity. The need for K(fl) corrections arises from the fact that the standard prescription for choosing the low-energy threshold delta in the SA restricted stopping-power ratio consistently underestimates the values of delta needed if no perturbation to the fluence is assumed. The use of fluence corrections can be avoided by appropriately choosing delta, but it is not clear how delta can be calculated from first principles. Values of delta required to avoid K(fl) corrections were found to be consistently higher than delta values obtained using the conventional approach and are also observed to be dependent on the composition of the wall in addition to the cavity size. Values of K(fl) have been calculated for many of the graphite-walled ion chambers used by the national metrology institutes around the world and found to be within 0.04% of unity in all cases, with an uncertainty of about 0.02%.

  8. Standing steady-state wave-making calculation method for air cushion vehicles; Air cushion vehicle no teijo zoha keisanho ni tsuite

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-04-10

    The pulse-height distribution of a cushion room of air cushion vehicle (ACV) has been tried to be approached by means of the panel shift type Rankine source method. When using this method, it was not required to introduce the pressure distribution model simulating the fall-off effect for the step-formed cushion pressure distribution. The wave form and wave making resistance could be estimated precisely by assigning the pressure gradient to one longitudinal direction panel in the calculation. The waveform shape within the cushion room could be calculated rather precisely by comparing with the analytic solution. This calculation method did have an ability providing the pulse-height information in the cushion room of ACV for seal design and configuration of ships. The analytic solution using for the comparison was sufficient for determining the pulse-height in the high speed region. However, it was hard to respond to non-linear problems or optional shape problems. It was pointed out to be further improved. 5 refs., 8 figs.

  9. European transboundary acidifying air pollution. Ten years calculated fields and budgets to the end of the first Sulphur Protocol

    Energy Technology Data Exchange (ETDEWEB)

    Barrett, K.; Seland, Oe.; Foss, A.; Mylona, S.; Sandnes, H.; Styve, H.; Tarrason, L.

    1995-07-01

    The Cooperative Programme for the Monitoring and Evaluation of the Long Range Transmission and Air Pollutants in Europe, EMEP, plays an integral part in data collection and scientific cooperation for implementation of the 1979 Geneva Convention on Long Range Transboundary Air Pollution. Within EMEP, the Meteorological Synthesizing Centre - West (MSC-W) is an international technical centre. The purpose of the MSC-W, focusing in part on acidifying substances, is to estimate the concentrations of relevant sulphur and nitrogen pollutants across Europe on the basis of emission information and meteorological data, and to estimate the transboundary fluxes of these substances. Responding to these specific obligations, the report presents calculations of sulphur and nitrogen concentrations and depositions and of their transboundary fluxes. The calculations are performed by the receptor oriented one layer trajectory (Lagrangian) acid deposition model, which during 1995 has been used to estimate acidifying pollutant fluxes for the ten year period 1985-1994. This corresponds to the period between initial signing and conclusion of the first Sulphur Protocol, signed in Helsinki in 1985. 90 refs., 42 figs., 43 tabs.

  10. A fast parallel code for calculating energies and oscillator strengths of many-electron atoms at neutron star magnetic field strengths in adiabatic approximation

    Science.gov (United States)

    Engel, D.; Klews, M.; Wunner, G.

    2009-02-01

    We have developed a new method for the fast computation of wavelengths and oscillator strengths for medium-Z atoms and ions, up to iron, at neutron star magnetic field strengths. The method is a parallelized Hartree-Fock approach in adiabatic approximation based on finite-element and B-spline techniques. It turns out that typically 15-20 finite elements are sufficient to calculate energies to within a relative accuracy of 10-5 in 4 or 5 iteration steps using B-splines of 6th order, with parallelization speed-ups of 20 on a 26-processor machine. Results have been obtained for the energies of the ground states and excited levels and for the transition strengths of astrophysically relevant atoms and ions in the range Z=2…26 in different ionization stages. Catalogue identifier: AECC_v1_0 Program summary URL:http://cpc.cs.qub.ac.uk/summaries/AECC_v1_0.html Program obtainable from: CPC Program Library, Queen's University, Belfast, N. Ireland Licensing provisions: Standard CPC licence, http://cpc.cs.qub.ac.uk/licence/licence.html No. of lines in distributed program, including test data, etc.: 3845 No. of bytes in distributed program, including test data, etc.: 27 989 Distribution format: tar.gz Programming language: MPI/Fortran 95 and Python Computer: Cluster of 1-26 HP Compaq dc5750 Operating system: Fedora 7 Has the code been vectorised or parallelized?: Yes RAM: 1 GByte Classification: 2.1 External routines: MPI/GFortran, LAPACK, PyLab/Matplotlib Nature of problem: Calculations of synthetic spectra [1] of strongly magnetized neutron stars are bedevilled by the lack of data for atoms in intense magnetic fields. While the behaviour of hydrogen and helium has been investigated in detail (see, e.g., [2]), complete and reliable data for heavier elements, in particular iron, are still missing. Since neutron stars are formed by the collapse of the iron cores of massive stars, it may be assumed that their atmospheres contain an iron plasma. Our objective is to fill the gap

  11. Neutron flux calculation and fluence in the encircling of the core and vessel of a reactor BWR; Calculo de flujo neutronico y fluencia en la envolvente del nucleo y la vasija de un reactor BWR

    Energy Technology Data Exchange (ETDEWEB)

    Martinez C, E.

    2011-07-01

    One of the main objectives related to the safety of any nuclear power plant, including the nuclear power plant of Laguna Verde is to ensure the structural integrity of reactor pressure vessel. To identify and quantify the damage caused by neutron irradiation in the vessel of any nuclear reactor, it is necessary to know both the neutron flux and the neutron fluence that the vessel has been receiving during its operation lifetime, and that the damage observed by mechanical testing are products of microstructural effects induced by neutron irradiation; therefore, it is important the study and prediction of the neutron flux in order to have a better understanding of the damage that these materials are receiving. The calculation here described uses the DORT code, which solves the neutron transport equation in discrete ordinates in two dimensions (x-y, r-{theta} and r-z), according to a regulatory guide, it should make an approximation of the neutron flux in three dimensions by the so called synthesis method. It is called in that way because it achieves a representation of 3 Dimensional neutron flux combining or summarizing the fluxes calculated by DORT r-{theta}, r-z and r. This work presents the application of synthesis method, according to Regulatory Guide 1190, to determine the 3 Dimensional fluxes in internal BWR reactor using three different spatial meshes. The results of the neutron flux and fluence, using three different meshes in the directions r, {theta} and z were compared with results reported in the literature obtaining a difference not larger than 9.61%, neutron flux reached its maximum, 1.58 E + 12 n/cm{sup 2}s, at a height H 4 (239.07 cm) and angle 32.236{sup o} in the core shroud and 4.00 E + 09 n/cm{sup 2}s at a height H 4 and angle 35.27{sup o} in the inner wall of the reactor vessel, positions that are consistent to within {+-}10% over the ones reported in the literature. (Author)

  12. The PSIMECX medium-energy neutron activation cross-section library. Part II: Calculational methods for light to medium mass nuclei

    Energy Technology Data Exchange (ETDEWEB)

    Atchison, F.

    1998-09-01

    The PSIMECX library contains calculated nuclide production cross-sections from neutron-induced reactions in the energy range about 2 to 800 MeV in the following 72 stable isotopes of 24 elements: {sup 12}C, {sup 13}C, {sup 16}O, {sup 17}O, {sup 18}O, {sup 23}Na, {sup 24}Mg, {sup 25}Mg, {sup 26}Mg, {sup 27}Al, {sup 28}Si, {sup 29}Si, {sup 30}Si, {sup 31}P, {sup 32}S, {sup 33}S, {sup 34}S, {sup 36}S, {sup 35}Cl, {sup 37}Cl, {sup 39}K, {sup 40}K, {sup 41}K, {sup 40}Ca, {sup 42}Ca, {sup 43}Ca, {sup 44}Ca, {sup 46}Ca, {sup 48}Ca, {sup 46}Ti, {sup 47}Ti, {sup 48}Ti, {sup 49}Ti, {sup 50}Ti, {sup 50}V, {sup 51}V, {sup 50}Cr, {sup 52}Cr, {sup 53}Cr, {sup 54}Cr, {sup 55}Mn, {sup 54}Fe, {sup 56}Fe, {sup 57}Fe, {sup 58}Fe, {sup 58}Ni, {sup 60}Ni, {sup 61}Ni, {sup 62}Ni, {sup 64}Ni, {sup 63}Cu, {sup 65}Cu, {sup 64}Zn, {sup 66}Zn, {sup 67}Zn, {sup 68}Zn, {sup 70}Zn, {sup 92}Mo, {sup 94}Mo, {sup 95}Mo, {sup 96}Mo, {sup 97}Mo, {sup 98}Mo, {sup 100}Mo, {sup 121}Sb, {sup 123}Sb, {sup 204}Pb, {sup 206}Pb, {sup 207}Pb, {sup 208}Pb, {sup 232}Th and {sup 238}U. The energy range covers essentially all transmutation channels other than capture. The majority of the selected elements are principal constituents of normal materials of construction used in and around accelerator facilities and the library is, first and foremost, designed to be a tool for the estimation of their activation in wide-band neutron fields. This second report, of a series of three, describes and discusses the calculational methods used for the stable isotopes up to and including {sup 123}Sb. The library itself has been described in the first report of the series and the treatment for the heavy nuclei is given in the third. (author)

  13. 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)

  14. 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)

  15. 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.

  16. ANALISIS LAJU DOSIS NEUTRON REAKTOR PLTN PWR 1000 MWe MENGGUNAKAN PROGRAM MCNP

    Directory of Open Access Journals (Sweden)

    Amir Hamzah

    2015-03-01

    the reactor core of 1000 MWe PWR performed using MCNP program. The calculation model performed in 9 zones: reactor core, water, baffle, water, barrel, pressure vessel, concrete and the outside air. Determination of the distribution of neutron flux and spectra made to the radial direction to the outside of concrete shield with an accuracy between 10% to 30% in each energy group of 1 and 50 groups. The analysis results of neutron dose rate at the surface of the reactor biological shield of 1000 MWe PWR reactor at full power condition is lower than safety limit value. In terms of neutron radiation exposure, it can be concluded that the two meter thick concrete radiation shielding meets the safety requirements. Key words: PWR NPP, neutron flux, shielding, neutron dose rate, MCNP.

  17. 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.

  18. Neutron spectrum and yield of the Hiroshima A-bomb deduced from radionuclide measurements at one location.

    Science.gov (United States)

    Rühm, W; Kato, K; Korschinek, G; Morinaga, H; Nolte, E

    1995-07-01

    In this paper measurements of the radionuclides of 36Cl, 41Ca, 60Co, 152Eu and 154Eu in samples from Hiroshima, which were exposed to neutrons of the A-bomb explosion, are interpreted. In order to calculate the neutron spectrum at the sample site, neutron transport calculations using Monte Carlo techniques were carried out. Activation profiles in a granite mock-up irradiated with reactor neutrons could be reproduced by this method using DS86 input parameters. The calculated neutron spectrum at the sample site for non-thermal neutrons is identical to that obtained in DS86, but contains some 50% more thermal neutrons. The influence of parameters like soil composition, source terms and air humidity on the activation of these radioisotopes is discussed. The granite-covered earth at the sample site, for example, hardens the spectrum in comparison with DS86 values. Even when using a fission spectrum pointing downward and neglecting air humidity one cannot explain our 36Cl measurements. If the effective thermal neutron fluences, that have a similar ratio of resonance integral to thermal neutron capture cross sections obtained from 36Cl, 41Ca and 152Eu, are averaged, a bomb yield of about 16 kt is deduced in agreement with a bomb yield of (15 +/- 3) kt estimated in DS86.

  19. Formation of solid thorium monoxide at near-ambient conditions as observed by neutron reflectometry and interpreted by screened hybrid functional calculations

    Science.gov (United States)

    He, Heming; Majewski, Jaroslaw; Allred, David D.; Wang, Peng; Wen, Xiaodong; Rector, Kirk D.

    2017-04-01

    Oxidation of a ∼1000 Å sputter-deposited thorium thin film at 150 °C in 100 ppm of flowing oxygen in argon produces the long-sought solid form of thorium monoxide. Changes in the scattering length density (SLD) distribution in the film over the 700-min experiment measured by in-situ, dynamic neutron reflectometry (NR) shows the densities, compositions and thickness of the various thorium oxides layers formed. Screened, hybrid density-functional theory calculations of potential thorium oxides aid interpretation, providing atomic-level picture and energetics for understanding oxygen migration. NR provided evidence of the formation of substoichiometric thorium oxide, ThOy (y thorium metal and its dioxide overcoat which grows inward, consuming the thorium at a rate of 2.1 Å/min while y increases until reaching 1:1 oxygen-to-thorium. Its presence indicates that kinetically-favored solid-phase ThO can be preferentially generated as a majority phase under the thermodynamically-favored ThO2 top layer at conditions close to ambient.

  20. Close shell interactions in 3-ethoxycarbonyl-4-hydroxy-6-methoxymethyleneoxy-1-methyl-2-quinolone: 100 K single crystal neutron diffraction study and ab initio calculations

    Science.gov (United States)

    Pozzi, C. G.; Fantoni, A. C.; Goeta, A. E.; Wilson, C. C.; Autino, J. C.; Punte, G.

    2005-10-01

    The molecular and crystal structures of the title compound have been determined from a single crystal neutron diffraction experiment at 100 K. A comparison between the main geometrical features and related properties of the in-crystal and the ab initio optimized free molecule structures has shown that crystal packing induces a significant distortion in the molecular geometry. Packing instead would only have a moderate effect on the observed intramolecular resonance assisted hydrogen bond. Supermolecular ab initio molecular orbital calculations have been performed on the six different dimers one molecule forms with its nine nearest neighbours. The obtained results clearly show that dispersion contributions dominate in the most strongly interacting dimers, in good qualitative accord with the predictions made by using different empirical potentials. A qualitative description of the most prominent inductive effects determined from the electron density deformation upon dimer formation is presented. Topological analyses of the dimers charge densities have been performed in the framework of the Bader's AIM theory and all the intermolecular bond critical points have been identified. An attempt to determine some of the interaction energies only from topological quantities made evident the practical limitations of such an approach.

  1. On the crystal energy and structure of A2TinO2n+1 (A=Li, Na, K) titanates by DFT calculations and neutron diffraction

    Science.gov (United States)

    Catti, Michele; Pinus, Ilya; Scherillo, Antonella

    2013-09-01

    First-principles quantum-mechanical calculations (CRYSTAL09 code, B3LYP functional) were performed on alkali titanates A2TinO2n+1 with layered structure (n=3,4,6). Monoclinic structural types with unshifted (P21/m) and with shifted (C2/m) layers were considered. Crystal energies and full structural details were obtained for all Li, Na, and K phases. Neutron diffraction data were collected on powder samples of P21/m-Li2Ti3O7 (a=9.3146(3), b=3.7522(1), c=7.5447(3) Å, β=97.611(4)°) and C2/m-K2Ti4O9 (a=18.2578(8), b=3.79160(9), c=12.0242(4) Å, β=106.459(4)°) and their structures were Rietveld-refined. Computed energies show the P21/m arrangement as favoured over the C2/m one for n=3, and the opposite holds for n=6. In the n=4 case the P21/m configuration is predicted to be more stable for Li and Na, and the C2/m one for K titanates. Analysis of Li-O and K-O crystal-chemical environments from experiment and theory shows that the alkali atom bonding is stabilized/destabilized in the different phases consistently with the energy trend.

  2. Neutron Thermal Cross Sections, Westcott Factors, Resonance Integrals, Maxwellian Averaged Cross Sections and Astrophysical Reaction Rates Calculated from the ENDF/B-VII.1, JEFF-3.1.2, JENDL-4.0, ROSFOND-2010, CENDL-3.1 and EAF-2010 Evaluated Data Libraries

    Science.gov (United States)

    Pritychenko, B.; Mughabghab, S. F.

    2012-12-01

    We present calculations of neutron thermal cross sections, Westcott factors, resonance integrals, Maxwellian-averaged cross sections and astrophysical reaction rates for 843 ENDF materials using data from the major evaluated nuclear libraries and European activation file. Extensive analysis of newly-evaluated neutron reaction cross sections, neutron covariances, and improvements in data processing techniques motivated us to calculate nuclear industry and neutron physics quantities, produce s-process Maxwellian-averaged cross sections and astrophysical reaction rates, systematically calculate uncertainties, and provide additional insights on currently available neutron-induced reaction data. Nuclear reaction calculations are discussed and new results are presented. Due to space limitations, the present paper contains only calculated Maxwellian-averaged cross sections and their uncertainties. The complete data sets for all results are published in the Brookhaven National Laboratory report.

  3. Anisotropic lattice thermal expansion of PbFeBO{sub 4}: A study by X-ray and neutron diffraction, Raman spectroscopy and DFT calculations

    Energy Technology Data Exchange (ETDEWEB)

    Murshed, M. Mangir, E-mail: murshed@uni-bremen.de [Chemische Kristallographie fester Stoffe, Institut für Anorganische Chemie, Universität Bremen, Leobener Straße, D-28359 Bremen (Germany); Mendive, Cecilia B.; Curti, Mariano [Departamento de Química, Facultad de Ciencias Exactas y Naturales, Universidad Nacional de Mar del Plata, Dean Funes 3350, B7600AYL, Mar del Plata (Argentina); Nénert, Gwilherm [Institut Laue-Langevin, 6 rue Jules Horowitz, 38042 Grenoble (France); Kalita, Patricia E. [Department of Physics and Astronomy and High-Pressure Science and Engineering Center, University of Nevada Las Vegas, Box 4002, Las Vegas, NV 89154-4002 (United States); Lipinska, Kris [Department of Mechanical Engineering, University of Nevada Las Vegas, 4505 Maryland Parkway, Box 454009, Las Vegas, NV 89154-4009 (United States); Cornelius, Andrew L. [Department of Physics and Astronomy and High-Pressure Science and Engineering Center, University of Nevada Las Vegas, Box 4002, Las Vegas, NV 89154-4002 (United States); Huq, Ashfia [Chemical and Engineering Materials Division, Oak Ridge National Laboratory, Oak Ridge, TN 37831-6475 (United States); Gesing, Thorsten M. [Chemische Kristallographie fester Stoffe, Institut für Anorganische Chemie, Universität Bremen, Leobener Straße, D-28359 Bremen (Germany)

    2014-11-15

    Highlights: • Mullite-type PbFeBO{sub 4} shows uni-axial negative coefficient of thermal expansion. • Anisotropic thermal expansion of the metric parameters was modeled using modified Grüneisen approximation. • The model includes harmonic, quasi-harmonic and intrinsic anharmonic contributions to the internal energy. • DFT calculation, temperature- and pressure-dependent Raman spectra help understand the phonon decay and associated anharmonicity. - Abstract: The lattice thermal expansion of mullite-type PbFeBO{sub 4} is presented in this study. The thermal expansion coefficients of the metric parameters were obtained from composite data collected from temperature-dependent neutron and X-ray powder diffraction between 10 K and 700 K. The volume thermal expansion was modeled using extended Grüneisen first-order approximation to the zero-pressure equation of state. The additive frame of the model includes harmonic, quasi-harmonic and intrinsic anharmonic potentials to describe the change of the internal energy as a function of temperature. The unit-cell volume at zero-pressure and 0 K was optimized during the DFT simulations. Harmonic frequencies of the optical Raman modes at the Γ-point of the Brillouin zone at 0 K were also calculated by DFT, which help to assign and crosscheck the experimental frequencies. The low-temperature Raman spectra showed significant anomaly in the antiferromagnetic regions, leading to softening or hardening of some phonons. Selected modes were analyzed using a modified Klemens model. The shift of the frequencies and the broadening of the line-widths helped to understand the anharmonic vibrational behaviors of the PbO{sub 4}, FeO{sub 6} and BO{sub 3} polyhedra as a function of temperature.

  4. Experimental magnetic form factors in Co3V2O8 : A combined study of ab initio calculations, magnetic Compton scattering, and polarized neutron diffraction

    Science.gov (United States)

    Qureshi, N.; Zbiri, M.; Rodríguez-Carvajal, J.; Stunault, A.; Ressouche, E.; Hansen, T. C.; Fernández-Díaz, M. T.; Johnson, M. R.; Fuess, H.; Ehrenberg, H.; Sakurai, Y.; Itou, M.; Gillon, B.; Wolf, Th.; Rodríguez-Velamazan, J. A.; Sánchez-Montero, J.

    2009-03-01

    We present a combination of ab initio calculations, magnetic Compton scattering, and polarized neutron experiments, which elucidate the density distribution of unpaired electrons in the kagome staircase system Co3V2O8 . Ab initio wave functions were used to calculate the spin densities in real and momentum spaces, which show good agreement with the respective experiments. It has been found that the spin polarized orbitals are equally distributed between the t2g and the eg levels for the spine (s) Co ions while the eg orbitals of the cross-tie (c) Co ions only represent 30% of the atomic spin density. Furthermore, the results reveal that the magnetic moments of the cross-tie Co ions, which are significantly smaller than those of the spine Co ions in the zero-field ferromagnetic structure, do not saturate by applying an external magnetic field of 2 T along the easy axis a . In turn, the increasing bulk magnetization, which can be observed by field dependent macroscopic measurements, originates from induced magnetic moments on the O and V sites. The refined individual magnetic moments are μ(Coc)=1.54(4)μB , μ(Cos)=2.87(3)μB , μ(V)=0.41(4)μB , μ(O1)=0.05(5)μB , μ(O2)=0.35(5)μB , and μ(O3)=0.36(5)μB combining to the same macroscopic magnetization value, which was previously only attributed to the Co ions.

  5. A highly optimized code for calculating atomic data at neutron star magnetic field strengths using a doubly self-consistent Hartree-Fock-Roothaan method

    Science.gov (United States)

    Schimeczek, C.; Engel, D.; Wunner, G.

    2014-05-01

    Our previously published code for calculating energies and bound-bound transitions of medium-Z elements at neutron star magnetic field strengths [D. Engel, M. Klews, G. Wunner, Comp. Phys. Comm. 180, 3-2-311 (2009)] was based on the adiabatic approximation. It assumes a complete decoupling of the (fast) gyration of the electrons under the action of the magnetic field and the (slow) bound motion along the field under the action of the Coulomb forces. For the single-particle orbitals this implied that each is a product of a Landau state and an (unknown) longitudinal wave function whose B-spline coefficients were determined self-consistently by solving the Hartree-Fock equations for the many-electron problem on a finite-element grid. In the present code we go beyond the adiabatic approximation, by allowing the transverse part of each orbital to be a superposition of Landau states, while assuming that the longitudinal part can be approximated by the same wave function in each Landau level. Inserting this ansatz into the energy variational principle leads to a system of coupled equations in which the B-spline coefficients depend on the weights of the individual Landau states, and vice versa, and which therefore has to be solved in a doubly self-consistent manner. The extended ansatz takes into account the back-reaction of the Coulomb motion of the electrons along the field direction on their motion in the plane perpendicular to the field, an effect which cannot be captured by the adiabatic approximation. The new code allows for the inclusion of up to 8 Landau levels. This reduces the relative error of energy values as compared to the adiabatic approximation results by typically a factor of three (1/3 of the original error) and yields accurate results also in regions of lower neutron star magnetic field strengths where the adiabatic approximation fails. Further improvements in the code are a more sophisticated choice of the initial wave functions, which takes into

  6. Calculating Air Quality and Climate Co-Benefits Metrics from Adjoint Elasticities in Chemistry-Climate Models

    Science.gov (United States)

    Spak, S.; Henze, D. K.; Carmichael, G. R.

    2013-12-01

    The science and policy communities both need common metrics that clearly, comprehensively, and intuitively communicate the relative sensitivities of air quality and climate to emissions control strategies, include emissions and process uncertainties, and minimize the range of error that is transferred to the metric. This is particularly important because most emissions control policies impact multiple short-lived climate forcing agents, and non-linear climate and health responses in space and time limit the accuracy and policy value of simple emissions-based calculations. Here we describe and apply new second-order elasticity metrics to support the direct comparison of emissions control policies for air quality and health co-benefits analyses using adjoint chemical transport and chemistry-climate models. Borrowing an econometric concept, the simplest elasticities in the atmospheric system are the percentage changes in concentrations due to a percentage change in the emissions. We propose a second-order elasticity metric, the Emissions Reduction Efficiency, which supports comparison across compounds, to long-lived climate forcing agents like CO2, and to other air quality impacts, at any temporal or spatial scale. These adjoint-based metrics (1) possess a single uncertainty range; (2) allow for the inclusion of related health and other impacts effects within the same framework; (3) take advantage of adjoint and forward sensitivity models; and (4) are easily understood. Using global simulations with the adjoint of GEOS-Chem, we apply these metrics to identify spatial and sectoral variability in the climate and health co-benefits of sectoral emissions controls on black carbon, sulfur dioxide, and PM2.5. We find spatial gradients in optimal control strategies on every continent, along with differences among megacities.

  7. Reactivity of aldehydes at the air-water interface. Insights from molecular dynamics simulations and ab initio calculations.

    Science.gov (United States)

    Martins-Costa, Marilia T C; García-Prieto, Francisco F; Ruiz-López, Manuel F

    2015-02-14

    Understanding the influence of solute-solvent interactions on chemical reactivity has been a subject of intense research in the last few decades. Theoretical studies have focused on bulk solvation phenomena and a variety of models and methods have been developed that are now widely used by both theoreticians and experimentalists. Much less attention has been paid, however, to processes that occur at liquid interfaces despite the important role such interfaces play in chemistry and biology. In this study, we have carried out sequential molecular dynamics simulations and quantum mechanical calculations to analyse the influence of the air-water interface on the reactivity of formaldehyde, acetaldehyde and benzaldehyde, three simple aldehydes of atmospheric interest. The calculated free-energy profiles exhibit a minimum at the interface, where the average reactivity indices may display large solvation effects. The study emphasizes the role of solvation dynamics, which are responsible for large fluctuations of some molecular properties. We also show that the photolysis rate constant of benzaldehyde in the range 290-308 nm increases by one order of magnitude at the surface of a water droplet, from 2.7 × 10(-5) s(-1) in the gas phase to 2.8 × 10(-4) s(-1) at the air-water interface, and we discuss the potential impact of this result on the chemistry of the troposphere. Experimental data in this domain are still scarce and computer simulations like those presented in this work may provide some insights that can be useful to design new experiments.

  8. Calculation and structural analysis for the rigidity of air spindle in the single point diamond turning lathe

    Science.gov (United States)

    An, Chenhui; Xu, Qiao; Zhang, Feihu; Zhang, Jianfeng

    2007-12-01

    Ultra-precision machining for optical lens is a key subject in the field of modern optics machining, the focus of which is the higher demands for profile precision and surface roughness. As a kind of deterministic machining, the single point diamond turning lathe is widely used in the optical field, thus higher stabilization for the turning lathe is required with small amplitude of vibrations in a broad frequency-domain. The single point diamond turning lathe now boast its various forms both at home and abroad, and the vertical flying cutting milling style is an important branch. This kind of lathe is characterized with low guide rail velocity and main errors of this part are the alignment error of guide rail, the disturbance evolved by driving components, and the low velocity crawl. Such errors are presented as low-frequency profile error on the workpiece surface, and often relate to the guide rail velocity. The rotate speed of the spindle is higher comparatively, and the system is composed as a vibration element with mass, air-rigidity, air-damping and the periodicity impact vibration. As a result, this vibration can copy to the work piece by the tool nose in machining process, so we must manage to reduce the vibration for high machining precision. This paper is to deduce the proper dynamic parameter for reducing the spindle vibration and optimize the spindle structure via dynamic calculation for the diamond turning lathes used and bring forward the reformative idea for the lathes.

  9. Source and replica calculations

    Energy Technology Data Exchange (ETDEWEB)

    Whalen, P.P.

    1994-02-01

    The starting point of the Hiroshima-Nagasaki Dose Reevaluation Program is the energy and directional distributions of the prompt neutron and gamma-ray radiation emitted from the exploding bombs. A brief introduction to the neutron source calculations is presented. The development of our current understanding of the source problem is outlined. It is recommended that adjoint calculations be used to modify source spectra to resolve the neutron discrepancy problem.

  10. A highly optimized code for calculating atomic data at neutron star magnetic field strengths using a doubly self-consistent Hartree-Fock-Roothaan method

    Science.gov (United States)

    Schimeczek, C.; Engel, D.; Wunner, G.

    2012-07-01

    Our previously published code for calculating energies and bound-bound transitions of medium-Z elements at neutron star magnetic field strengths [D. Engel, M. Klews, G. Wunner, Comput. Phys. Comm. 180 (2009) 302-311] was based on the adiabatic approximation. It assumes a complete decoupling of the (fast) gyration of the electrons under the action of the magnetic field and the (slow) bound motion along the field under the action of the Coulomb forces. For the single-particle orbitals this implied that each is a product of a Landau state and an (unknown) longitudinal wave function whose B-spline coefficients were determined self-consistently by solving the Hartree-Fock equations for the many-electron problem on a finite-element grid. In the present code we go beyond the adiabatic approximation, by allowing the transverse part of each orbital to be a superposition of Landau states, while assuming that the longitudinal part can be approximated by the same wave function in each Landau level. Inserting this ansatz into the energy variational principle leads to a system of coupled equations in which the B-spline coefficients depend on the weights of the individual Landau states, and vice versa, and which therefore has to be solved in a doubly self-consistent manner. The extended ansatz takes into account the back-reaction of the Coulomb motion of the electrons along the field direction on their motion in the plane perpendicular to the field, an effect which cannot be captured by the adiabatic approximation. The new code allows for the inclusion of up to 8 Landau levels. This reduces the relative error of energy values as compared to the adiabatic approximation results by typically a factor of three (1/3 of the original error), and yields accurate results also in regions of lower neutron star magnetic field strengths where the adiabatic approximation fails. Further improvements in the code are a more sophisticated choice of the initial wave functions, which takes into

  11. Influence of the delta ray production threshold on water-to-air stopping power ratio calculations for carbon ion beam radiotherapy.

    Science.gov (United States)

    Sánchez-Parcerisa, D; Gemmel, A; Jäkel, O; Rietzel, E; Parodi, K

    2013-01-07

    Previous calculations of the water-to-air stopping power ratio (s(w,)(air)) for carbon ion beams did not involve tracking of delta ray electrons, even though previous calculations with protons predict an effect up to 1%. We investigate the effect of the delta ray production threshold in s(w,)(air) calculations and propose an empirical expression which takes into account the effect of the delta ray threshold as well as the uncertainty in the mean ionization potentials (I-values) of air and water. The formula is derived from the results of Monte Carlo calculations using the most up-to-date experimental data for I-values and a delta ray production threshold of 10 keV. It allows us to reduce the standard uncertainty in s(w,)(air) below 0.8%, instead of the current 2% given in international protocols, which results in a reduction of the overall uncertainty for absolute dosimetry based on air-filled ionization chambers.

  12. 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.)

  13. Nano-metrology of porous structures — I Comparison of measured neutron scattering with calculated scattering to access pore lattice, diameter, and wall parameters, using models of extended arrays of regular or randomised pores

    Energy Technology Data Exchange (ETDEWEB)

    Webber, J. Beau W., E-mail: J.B.W.Webber@kent.ac.uk

    2013-05-15

    Neutron scattering offers a length-scale-independent method of probing structured matter on an atomic scale through nano-scale to meso-scale. A protocol is presented that provides a versatile method of determining structure, by comparison of measured and calculated neutron scattering, for any structural distribution that can be described algebraically or numerically, requiring no particular model other than the model of the structure, and needing no adjustable parameters other than the scale and other parameters describing the physical model. The method enables the direct comparison of measured and calculated scattering from structured matter: from simple finite and infinite bodies, from extended regular array of pores, or from extended arrays of pores with a partially randomised character. Examples are given for the radial distributions of a range of regular bodies, of large arrays of highly ordered porous materials such as templated SBA-15 and MCM-41 silicas, as well as for more disordered materials such as sol–gel silicas. Monte Carlo integration of the calculated scattering for ensembles of up to about 100,000 pores has been studied using these techniques. The method enables the calculation of the solid–solid density correlation function G(r) for model systems, and hence, by Fourier transformation, the expected scattering. Example measured scattering is compared with the calculated scattering, with further data presented in a related paper. The technique allows the direct calculation and comparison with measurement of all three of the main pore structural parameters: lattice spacing, pore diameter, and pore-wall thickness. Example SBA-15 wide and small angle neutron scattering (SANS) data, measured on NIMROD (the Near and InterMediate Range Order Diffractometer at ISIS), is used as an initial evaluation of the applicability of the techniques. The method is also applicable to determining structure by comparing calculating with measured diffraction broadening

  14. Neutron detection efficiency determinations for the TUNL neutron-neutron and neutron-proton scattering-length measurements

    Energy Technology Data Exchange (ETDEWEB)

    Trotter, D.E. Gonzalez [Department of Physics, Duke University and Triangle Universities Nuclear Laboratory, Durham, NC 27708-0308 (United States)], E-mail: crowell@tunl.duke.edu; Meneses, F. Salinas [Department of Physics, Duke University and Triangle Universities Nuclear Laboratory, Durham, NC 27708-0308 (United States); Tornow, W. [Department of Physics, Duke University and Triangle Universities Nuclear Laboratory, Durham, NC 27708-0308 (United States)], E-mail: tornow@tunl.duke.edu; Crowell, A.S.; Howell, C.R. [Department of Physics, Duke University and Triangle Universities Nuclear Laboratory, Durham, NC 27708-0308 (United States); Schmidt, D. [Physikalisch-Technische Bundesanstalt, D-38116, Braunschweig (Germany); Walter, R.L. [Department of Physics, Duke University and Triangle Universities Nuclear Laboratory, Durham, NC 27708-0308 (United States)

    2009-02-11

    The methods employed and the results obtained from measurements and calculations of the detection efficiency for the neutron detectors used at Triangle Universities Nuclear Laboratory (TUNL) in the simultaneous determination of the {sup 1}S{sub 0} neutron-neutron and neutron-proton scattering lengths a{sub nn} and a{sub np}, respectively, are described. Typical values for the detector efficiency were 0.3. Very good agreement between the different experimental methods and between data and calculation has been obtained in the neutron energy range below E{sub n}=13MeV.

  15. The dose comparison between the THOR and HFR epithermal neutron beams

    Energy Technology Data Exchange (ETDEWEB)

    Lin, Yi-Chun [Biomedical Engineering and Environmental Sciences Department, National Tsing Hua University, Hsinchu, Taiwan (China); Roca, Antoaneta [Institute for Energy, Joint Research Centre, European Commission (Netherlands); Faculty of Physics, University of Bucharest, Bucuresti-Magurele (Romania); Liu, Yuan-Hao, E-mail: yhl.taiwan@gmail.co [Health Physics Division, Nuclear Science and Technology Development Center, National Tsing Hua University, No. 101, Section 2, Kuang-Fu Road, Hsinchu 30013, Taiwan (China); Tsai, Pi-En [Health Physics Division, Nuclear Science and Technology Development Center, National Tsing Hua University, No. 101, Section 2, Kuang-Fu Road, Hsinchu 30013, Taiwan (China); Nievaart, Sander [Institute for Energy, Joint Research Centre, European Commission (Netherlands); Liu, Hong-Ming [Health Physics Division, Nuclear Science and Technology Development Center, National Tsing Hua University, No. 101, Section 2, Kuang-Fu Road, Hsinchu 30013, Taiwan (China); Moss, Ray [Institute for Energy, Joint Research Centre, European Commission (Netherlands); Chou, Wen-Tsae [Biomedical Engineering and Environmental Sciences Department, National Tsing Hua University, Hsinchu, Taiwan (China); Jiang, Shiang-Huei [Institute of Nuclear Engineering and Science, National Tsing Hua University, Hsinchu, Taiwan (China)

    2010-12-15

    This study is a part of the beam comparison campaign, inter-center dose comparison, between boron neutron capture therapy facilities at the Tsing Hua Open-pool Reactor and the High Flux Reactor. The clinical information exchange can improve the dosimetry uncertainty for medical physics in a mixed field. The method of paired Mg(Ar) and TE(TE) ionization chambers was used to determine the gamma-ray and neutron dose rates. Furthermore, activation foils, including gold, copper, and manganese, were employed to estimate the thermal and epithermal neutron fluxes. Measurements were performed free in air and also in a PMMA phantom. All the chambers were calibrated using a {sup 60}Co primary standard source at the Institute of Nuclear Energy Research, Taiwan. Spectrum dependent neutron sensitivity of TE(TE) chamber is one of the important parameters to evaluate dose components. The requested neutron spectra were calculated by the Monte Carlo code MCNP. The measured thermal neutron fluxes, gamma-ray and neutron dose rates of the THOR beam in the phantom were 2.6, 2.2, and 2.1 times of the HFR beam at 2.5-cm depth, respectively. The higher thermal neutron flux and neutron and gamma-ray dose rates are due to the higher epithermal neutron beam intensity of the THOR.

  16. 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.

  17. Use of neutron activation analysis for the control of air pollution of Algiers; Utilisation de l'analyse par activation neutronique pour le controle de la pollution de l'air d'Alger

    Energy Technology Data Exchange (ETDEWEB)

    Belamri, M.; Benrachedi, K. [Universite M' hamed Bouguarra, Lab. de Technologie Alimentaire, Boumerdes (Algeria)

    2010-07-15

    The urban zone needs clean air to assure public health. To achieve this goal several filter samples were collected in different sites in Algiers city. Toxic elements such as: Na, Mg, Cl, Sc, Cr, Ti, V, Fe, Co, Cu, Zn, Se, Br, Ag, Sb, Ce, La, Hf, Ta and Hg have been measured in the filters using neutron activation analysis technique. Irradiation of filter samples and standards were carried out in Es-Salem reactor. The experimental procedure and the results are discussed. We noted during this work that the upper limit values for suspended dusts and the high concentrations for some toxic elements found are due to the weather conditions and intense road traffic around collecting sites. (authors)

  18. Improvement effect on the depth-dose distribution by CSF drainage and air infusion of a tumour-removed cavity in boron neutron capture therapy for malignant brain tumours

    Science.gov (United States)

    Sakurai, Yoshinori; Ono, Koji; Miyatake, Shin-ichi; Maruhashi, Akira

    2006-03-01

    Boron neutron capture therapy (BNCT) without craniotomy for malignant brain tumours was started using an epi-thermal neutron beam at the Kyoto University Reactor in June 2002. We have tried some techniques to overcome the treatable-depth limit in BNCT. One of the effective techniques is void formation utilizing a tumour-removed cavity. The tumorous part is removed by craniotomy about 1 week before a BNCT treatment in our protocol. Just before the BNCT irradiation, the cerebro-spinal fluid (CSF) in the tumour-removed cavity is drained out, air is infused to the cavity and then the void is made. This void improves the neutron penetration, and the thermal neutron flux at depth increases. The phantom experiments and survey simulations modelling the CSF drainage and air infusion of the tumour-removed cavity were performed for the size and shape of the void. The advantage of the CSF drainage and air infusion is confirmed for the improvement in the depth-dose distribution. From the parametric surveys, it was confirmed that the cavity volume had good correlation with the improvement effect, and the larger effect was expected as the cavity volume was larger.

  19. 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)

  20. 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

    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...... 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...

  1. 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...

  2. 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...

  3. Effects of bone- and air-tissue inhomogeneities on the dose distributions of the Leksell Gamma Knife$^{\\circledR}$ calculated with PENELOPE

    CERN Document Server

    Al-Dweri, F M O; Rojas, E L; Al-Dweri, Feras M.O.; Lallena, Antonio M.

    2005-01-01

    Monte Carlo simulation with PENELOPE (v.~2003) is applied to calculate Leksell Gamma Knife$^{\\circledR}$ 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 isocenter approaches the separation surfaces. Our findings confirm an important underdosage ($\\sim$10%) nearby the air-tissue interface, in accordance with previous results obtained with PENELOPE code with a procedure different to ours....

  4. Finite difference calculation of acoustic streaming including the boundary layer phenomena in an ultrasonic air pump on graphics processing unit array

    Science.gov (United States)

    Wada, Yuji; Koyama, Daisuke; Nakamura, Kentaro

    2012-09-01

    Direct finite difference fluid simulation of acoustic streaming on the fine-meshed threedimension model by graphics processing unit (GPU)-oriented calculation array is discussed. Airflows due to the acoustic traveling wave are induced when an intense sound field is generated in a gap between a bending transducer and a reflector. Calculation results showed good agreement with the measurements in the pressure distribution. In addition to that, several flow-vortices were observed near the boundary of the reflector and the transducer, which have been often discussed in acoustic tube near the boundary, and have not yet been observed in the calculation in the ultrasonic air pump of this type.

  5. Isodose Curves and Treatment Planning for Boron Neutron Capture Therapy.

    Science.gov (United States)

    Liu, Hungyuan B.

    The development of Boron Neutron Capture Therapy (BNCT) has been progressing in both ^{10 }B compound development and testing and neutron beam delivery. Animal tests are now in progress with several ^{10}B compounds and once the results of these animal tests are promising, patient trials can be initiated. The objective of this study is to create a treatment planning method based on the dose calculations by a Monte Carlo code of a mixed radiation field to provide linkage between phantom dosimetry and patient irradiation. The research started with an overall review of the development of BNCT. Three epithermal neutron facilities are described, including the operating Brookhaven Medical Research Reactor (BMRR) beam, the designed Missouri University Research Reactor (MURR) beam, and a designed accelerator based neutron source. The flux and dose distributions in a head model have been calculated for irradiation by these neutron beams. Different beam parameters were inter -compared for effectiveness. Dosimetric measurements in an elliptical lucite phantom and a cylindrical water phantom were made and compared to the MCNP calculations for irradiation by the BMRR beam. Repeated measurements were made and show consistent. To improve the statistical results calculated by MCNP, a neutron source plane was designed to start neutrons at the BMRR irradiation port. The source plane was used with the phantoms for dosimetric calculations. After being verified by different phantom dosimetry and in-air flux measurements at the irradiation port, the source plane was used to calculate the flux and dose distributions in the head model. A treatment planning program was created for use on a PC which uses the MCNP calculated results as input. This program calculates the thermal neutron flux and dose distributions of each component of radiation in the central coronal section of the head model for irradiation by a neutron beam. Different combinations of head orientations and irradiation

  6. 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

    2014-08-01

    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, although recently sulphur emissions, and subsequently particle emissions, have decreased in EU ports and in the Baltic Sea and the North Sea, defined as SECAs (Sulphur Emission Control Areas. The maximum allowed sulphur content in marine fuels in EU ports is now 0.1%, as required by the European Union sulphur directive. In the SECAs the maximum fuel content of sulphur is currently 1% (the global average is about 2.4%. This will be reduced to 0.1% from 2015, following the new IMO rules (International Maritime Organisation. 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 (YOLL have been calculated by comparing model calculations with and without ship emissions in the two sea areas. The calculations have been made with emissions representative of 2009 and 2011, i.e. before and after the implementation of stricter controls on sulphur emissions from mid 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 present emission levels. Comparisons of model calculations with emissions before and after the implementation of stricter emission control on sulphur show a general decrease in calculated particle concentration. At the same time, however, an increase in ship activity has resulted in higher emissions and subsequently air concentrations, in particular of NOx

  7. Thermal calculation method for quad-sectional regenerative air preheater%四分仓回转式空气预热器热力计算方法

    Institute of Scientific and Technical Information of China (English)

    陈昌贤; 孙奉仲; 李飞; 吴艳艳

    2014-01-01

    A new thermal calculation model for quad-sectional regenerative air preheater was investigated based on the analysis of operating mode and heat transfer characteristics of quad-sectional air preheaters.The model included two cal-culating steps:the average outlet flue gas temperature and the weighted average outlet air temperature were obtained through the first step calculation;and based on the first step calculation results, the average outlet air temperatures were obtained in right-secondary airside, primary air side and left-secondary airside respectively by the second step calcula-tion.The results showed that the maximum and minimum relative deviations between calculated values and design val-ues were 2.27% and 0.21% respectively.For a 300 MW circulating fluidized bed boiler unit with quad-sectional air preheater, the maximum and average relative deviations between calculated values and actual operation data were 4%and 1.8%respectively, the accuracy and reliability of this thermal calculation model were verified.%基于四分仓空气预热器的运行方式和传热特点的分析,建立了先合仓再分仓的热力计算模型。模型包含合仓和分仓两步计算,通过合仓计算获得平均出口烟温和有加权的平均出口风温;利用合仓计算结果进行分仓计算,获得右二次风仓、一次风仓和左二次风仓的平均出口风温。结果表明,计算值与设计值的最大相对偏差为2.27%,最小相对偏差为0.21%。以某300 MW循环流化床锅炉机组的四分仓空气预热器为实例,在4个常用工况下,计算值与实际运行数据的最大相对偏差为4%,平均相对偏差为1.8%,验证了此计算方法的准确性和可靠性。

  8. 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

    CERN Document Server

    White, M C

    2000-01-01

    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 tran...

  9. 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...

  10. 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.

  11. 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)

  12. 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)

  13. Calculations of Neutron and y Transport through Rare-earth Doped Polymer%中子、γ射线在稀土-高分子材料中的输运

    Institute of Scientific and Technical Information of China (English)

    呼延雪莹; 胡碧涛

    2011-01-01

    A series of shielding analyses have been performed to estimate the material composition and optimum thickness required for a new radiation shield with various rare-earth doped polymer and heavy metal mixtures.The neutron and y photon fluxes have been calculated by Monte Carlo N-Particle(MCNP) transport code.The results indicate that the relative fluxes of y photon and neutron in both traditional and new composite materials follow an exponential decay rule with the distance of penetration.It can be seen that the composite material consisting of rare-earth doped polymer and heavy metal has stronger neutron shielding performance than lead-boron polyethylene,but weaker y shielding effectiveness than W-Ni alloy.It isalso found that materials with more components of rare earth elements don't always provide better neutron shielding performance.%为研究新型复合屏蔽材料的最佳厚度与各种成分最佳配比,用MCNP计算了中子、γ射线在稀土-高分子与重金属复合材料中的通量.对中子、γ射线在屏蔽体中变化规律进行了深入探索,同传统复合屏蔽材料的屏蔽性能进行了对比.结果表明,中子和γ射线通过屏蔽体时,其强度遵循指数衰减规律.新型屏蔽材料对中子的屏蔽效果均优于铅硼聚乙烯,对γ射线的屏蔽效果均劣于W-Ni合金,且并非稀土含量越高,材料对中子辐射屏蔽能力越强.

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

    Science.gov (United States)

    Goldhagen, P; Reginatto, M; Kniss, T; Wilson, J W; Singleterry, R C; Jones, I W; Van Steveninck, 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 (thermal to >10 GeV), total neutron fluence rate, and neutron effective dose and dose equivalent rates and their dependence on altitude and geomagnetic cutoff. The measured cosmic-ray neutron spectra have almost no thermal neutrons, a large "evaporation" peak near 1 MeV and a second broad peak near 100 MeV which contributes about 69% of the neutron effective dose. At high altitude, geomagnetic latitude has very little effect on the shape of the spectrum, but it is the dominant variable affecting neutron fluence rate, which was eight times higher at the northernmost measurement location than it was at the southernmost. The shape of the spectrum varied only slightly with altitude from 21 km down to 12 km (56-201 g cm-2 atmospheric depth), but was significantly different on the ground. In all cases, ambient dose equivalent was greater than effective dose for cosmic-ray neutrons.

  15. 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)

  16. Neutronic studies of the coupled moderators for spallation neutron sources

    Institute of Scientific and Technical Information of China (English)

    Yin Wen; Liang Jiu-Qing

    2005-01-01

    We investigate the neutronic performance of coupled moderators to be implemented in spallation neutron sources by Monte-Carlo simulation and give the slow neutron spectra for the cold and thermal moderators. CH4 moderator can provide slow neutrons with highly desirable characteristics and will be used in low-power spallation neutron soureces. The slow neutron intensity extracted from different angles has been calculated. The capability of moderation of liquid H2 is lower than H2O and liquid CH4 due to lower atomic number density of hydrogen but we can compensate for this disadvantage by using a premoderator. The H2O premoderator of 2cm thickness can reduce the heat deposition in the cold moderator by about 33% without spoiling the neutron pulse.

  17. 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.

  18. Simple calculation method for outdoor air specific enthalpy in Shanghai%上海地区室外空气比焓简捷计算方法

    Institute of Scientific and Technical Information of China (English)

    黄晨; 张倩茹

    2014-01-01

    分析了上海地区典型气象年数据,发现室外空气比焓可仅采用室外干球温度确定,经过曲线拟合获得了上海地区室外空气比焓对于温度的四次多项式函数。讨论了典型年数据集中度,采用2种实时数据进行了验证,结果表明可以采用该简捷计算方法计算室外空气比焓,标准差小于8 kJ/kg。采用该简捷计算方法分析了利用室外自然冷量的节能潜力。提出其他地区可采用该地区或代表地区的室外空气比焓简捷计算公式求解室外空气比焓。%Analyses the data of the typical meteorological year in Shanghai.Finds that the outdoor air specific enthalpy can be calculated only by the outdoor air dry-bulb temperature.The four order polynomial function of the temperature is obtained by curve-fitting method.Discusses the concentration degree of the typical meteorological year data,and verifies the data with two kinds of real-time data.The results show that the simple method can be applied to calculate outdoor air specific enthalpy,and the standard deviation is less than 8 kJ/kg.Analyses the energy-saving potential of free cooling with the outdoor air using this simple calculation method.Proposes that other areas can adopt the function of this area or a representative area to make simple calculation.

  19. Calculation of conversion coefficients for air kerma to ambient dose equivalent using transmitted spectra of megavoltage X-rays through concrete.

    Science.gov (United States)

    Cordeiro, T P V; Silva, A X

    2012-12-01

    With the fast advancement of technology, (60)Co teletherapy units are largely being replaced with medical linear accelerators. In most cases, the linear accelerator tends to be installed in the same room in which the (60)Co teletherapy unit was previously placed. If in-depth structural remodelling is out of the question, high-density concrete is usually used to improve shielding against primary, scatter and leakage radiation originating in the new equipment. This work presents a study based on Monte Carlo simulations of the transmission of some clinical photon spectra (from 6, 10, 15, 18 and 25 MV accelerators) through concrete, considering two different densities. Concrete walls with thickness ranging from 0.70 to 2.0 m were irradiated with 30 cm×30 cm primary beam spectra. The results show that the thickness of the barrier decreases up to ∼65 % when barite (high-density concrete) is used instead of ordinary concrete. The average energies of primary and transmitted beam spectra were also calculated. In addition, conversion coefficients from air kerma to ambient dose equivalent, H*(d)/K(air), and air kerma to effective dose, E/K(air), for photon spectra from the transmitted spectra were calculated and compared. The results suggest that the 10-mm depth is not the best choice to represent the effective dose.

  20. Thermal energy recovery of air conditioning system--heat recovery system calculation and phase change materials development

    Energy Technology Data Exchange (ETDEWEB)

    Gu Zhaolin; Liu Hongjuan; Li Yun

    2004-12-01

    Latent heat thermal energy storage systems can be used to recover the rejected heat from air conditioning systems, which can be used to generate low-temperature hot water. It decreases not only the consumption of primary energy for heating domestic hot water but also the calefaction to the surroundings due to the rejection of heat from air conditioning systems. A recovery system using phase change materials (PCMs) to store the rejected (sensible and condensation) heat from air conditioning system has been developed and studied, making up the shortage of other sensible heat storage system. Also, PCMs compliant for heat recovery of air conditioning system should be developed. Technical grade paraffin wax has been discussed in this paper in order to develop a paraffin wax based PCM for the recovery of rejected heat from air conditioning systems. The thermal properties of technical grade paraffin wax and the mixtures of paraffin wax with lauric acid and with liquid paraffin (paraffin oil) are investigated and discussed, including volume expansion during the phase change process, the freezing point and the heat of fusion.

  1. Sensitivity studies of beam directionality, beam size, and neutron spectrum for a fission converter-based epithermal neutron beam for boron neutron capture therapy.

    Science.gov (United States)

    Sakamoto, S; Kiger, W S; Harling, O K

    1999-09-01

    Sensitivity studies of epithermal neutron beam performance in boron neutron capture therapy are presented for realistic neutron beams with varying filter/moderator and collimator/delimiter designs to examine the relative importance of neutron beam spectrum, directionality, and size. Figures of merit for in-air and in-phantom beam performance are calculated via the Monte Carlo technique for different well-optimized designs of a fission converter-based epithermal neutron beam with head phantoms as the irradiation target. It is shown that increasing J/phi, a measure of beam directionality, does not always lead to corresponding monotonic improvements in beam performance. Due to the relatively low significance, for most configurations, of its effect on in-phantom performance and the large intensity losses required to produce beams with very high J/phi, beam directionality should not be considered an important figure of merit in epithermal neutron beam design except in terms of its consequences on patient positioning and collateral dose. Hardening the epithermal beam spectrum, while maintaining the specific fast neutron dose well below the inherent hydrogen capture dose, improves beam penetration and advantage depth and, as a desirable by-product, significantly increases beam intensity. Beam figures of merit are shown to be strongly dependent on beam size relative to target size. Beam designs with J/phi approximately 0.65-0.7, specific fast neutron doses of 2-2.6x10(-13) Gy cm2/n and beam sizes equal to or larger than the size of the head target produced the deepest useful penetration, highest therapeutic ratios, and highest intensities.

  2. Analytical calculations and Monte-Carlo simulations of a high-resolution backscattering spectrometer for the long wavelength target station at the Spallation neutron source

    Energy Technology Data Exchange (ETDEWEB)

    Bordallo, H.N. E-mail: bordallo@hmi.de; Herwig, K.W.; Zsigmond, G

    2002-09-21

    Using the Monte-Carlo simulation programs McStas and VITESS, we present the design principles of the proposed high-resolution inverse geometry spectrometer on the Spallation neutron source (SNS) - long wavelength target station (LWTS). LWTS will enable the combination of large energy and momentum transfer ranges with energy resolution. Indeed the resolution of this spectrometer lie between that routinely achieved by spin echo techniques and the design goal of the high-power target station (HPTS) backscattering spectrometer. This niche of energy resolution is interesting for the study of slow motions of large objects and we are led to the domain of large molecules - polymers and biological molecules.

  3. Analytical calculations and Monte-Carlo simulations of a high-resolution backscattering spectrometer for the long wavelength target station at the Spallation neutron source

    CERN Document Server

    Bordallo, H N; Zsigmond, G

    2002-01-01

    Using the Monte-Carlo simulation programs McStas and VITESS, we present the design principles of the proposed high-resolution inverse geometry spectrometer on the Spallation neutron source (SNS) - long wavelength target station (LWTS). LWTS will enable the combination of large energy and momentum transfer ranges with energy resolution. Indeed the resolution of this spectrometer lie between that routinely achieved by spin echo techniques and the design goal of the high-power target station (HPTS) backscattering spectrometer. This niche of energy resolution is interesting for the study of slow motions of large objects and we are led to the domain of large molecules - polymers and biological molecules.

  4. 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)

  5. SU-E-T-552: Monte Carlo Calculation of Correction Factors for a Free-Air Ionization Chamber in Support of a National Air-Kerma Standard for Electronic Brachytherapy

    Energy Technology Data Exchange (ETDEWEB)

    Mille, M; Bergstrom, P [National Institute of Standards and Technology, Gaithersburg, MD (United States)

    2015-06-15

    Purpose: To use Monte Carlo radiation transport methods to calculate correction factors for a free-air ionization chamber in support of a national air-kerma standard for low-energy, miniature x-ray sources used for electronic brachytherapy (eBx). Methods: The NIST is establishing a calibration service for well-type ionization chambers used to characterize the strength of eBx sources prior to clinical use. The calibration approach involves establishing the well-chamber’s response to an eBx source whose air-kerma rate at a 50 cm distance is determined through a primary measurement performed using the Lamperti free-air ionization chamber. However, the free-air chamber measurements of charge or current can only be related to the reference air-kerma standard after applying several corrections, some of which are best determined via Monte Carlo simulation. To this end, a detailed geometric model of the Lamperti chamber was developed in the EGSnrc code based on the engineering drawings of the instrument. The egs-fac user code in EGSnrc was then used to calculate energy-dependent correction factors which account for missing or undesired ionization arising from effects such as: (1) attenuation and scatter of the x-rays in air; (2) primary electrons escaping the charge collection region; (3) lack of charged particle equilibrium; (4) atomic fluorescence and bremsstrahlung radiation. Results: Energy-dependent correction factors were calculated assuming a monoenergetic point source with the photon energy ranging from 2 keV to 60 keV in 2 keV increments. Sufficient photon histories were simulated so that the Monte Carlo statistical uncertainty of the correction factors was less than 0.01%. The correction factors for a specific eBx source will be determined by integrating these tabulated results over its measured x-ray spectrum. Conclusion: The correction factors calculated in this work are important for establishing a national standard for eBx which will help ensure that dose

  6. 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)

  7. ITER中子通量监测器的优化计算%Optical Calculations of Neutron Flux Monitor for ITER

    Institute of Scientific and Technical Information of China (English)

    李初; 王强; 兰礼; 刘虓瀚; 曾军; 刘艺琴; 罗小兵

    2012-01-01

    中子通量监测器(NFM)可实现ITER实时的中子通量测定,转化得到聚变功率,功率密度,等离子体温度等.获得NFM探测效率对能量的相对平坦响应对准确诊断十分必要.论文针对特定的NFM裂变室结构,运用MCNP—4C对裂变室包裹层慢化剂/屏蔽材料种类及厚度进行了优化计算.这些工作对探测器裂变室结构的优化设计实验标定及定型具有重要意义.%The Neutron Flux Monitor(NFM) can provide the real - time flux of ITER( International Thermonuclear Experimental Reactor ) , and get the fusion power and temperature of the plasma after transformation. A relative flat energy response curve of neutron detection efficiency is essential for accurate diagnosis of NFM in ITER. The paper makes an optimal computation on thickness of different moderator/ shielding material with the MCNP - 4C as to the specific structure of NFM fission chamber. It is significant for the optimal design and the experimental calibration of the NFM

  8. 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

  9. Neutron Repulsion

    OpenAIRE

    Manuel, Oliver K.

    2011-01-01

    Earth is connected gravitationally, magnetically and electrically to its heat source - a neutron star that is obscured from view by waste products in the photosphere. Neutron repulsion is like the hot filament in an incandescent light bulb. Excited neutrons are emitted from the solar core and decay into hydrogen that glows in the photosphere like a frosted light bulb. Neutron repulsion was recognized in nuclear rest mass data in 2000 as the overlooked source of energy, the keystone of an arch...

  10. 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...

  11. 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...

  12. 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)

  13. Temperature of neutron stars

    Science.gov (United States)

    Tsuruta, Sachiko

    2016-07-01

    We start with a brief introduction to the historical background in the early pioneering days when the first neutron star thermal evolution calculations predicted the presence of neutron stars hot enough to be observable. We then report on the first detection of neutron star temperatures by ROSAT X-ray satellite, which vindicated the earlier prediction of hot neutron stars. We proceed to present subsequent developments, both in theory and observation, up to today. We then discuss the current status and the future prospect, which will offer useful insight to the understanding of basic properties of ultra-high density matter beyond the nuclear density, such as the possible presence of such exotic particles as pion condensates.

  14. 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

  15. Analytical calculations and Monte-Carlo simulations of a high-resolution backscattering spectrometer for the long wavelength target station at the Spallation neutron source

    Science.gov (United States)

    Bordallo, H. N.; Herwig, K. W.; Zsigmond, G.

    2002-09-01

    Using the Monte-Carlo simulation programs McStas and VITESS, we present the design principles of the proposed high-resolution inverse geometry spectrometer on the Spallation neutron source (SNS)—long wavelength target station (LWTS). LWTS will enable the combination of large energy and momentum transfer ranges with energy resolution. Indeed the resolution of this spectrometer lie between that routinely achieved by spin echo techniques and the design goal of the high-power target station (HPTS) backscattering spectrometer. This niche of energy resolution is interesting for the study of slow motions of large objects and we are led to the domain of large molecules—polymers and biological molecules.

  16. 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.

  17. 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.

  18. 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

    2016-07-01

    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.

  19. 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)

  20. 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).

  1. Solar Neutrons and the Earth's Radiation Belts.

    Science.gov (United States)

    Lingenfelter, R E; Flamm, E J

    1964-04-17

    The intensity and spectrum of solar neutrons in the vicinity of the earth are calculated on the assumption that the low-energy protons recently detected in balloon and satellite flights are products of solar neutron decay. The solar-neutron flux thus obtained exceeds the global average cosmic-ray neutron leakage above 10 Mev, indicating that it may be an important source of both the inner and outer radiation belts. Neutron measurements in the atmosphere are reviewed and several features of the data are found to be consistent with the estimated solar neutron spectrum.

  2. Analytical and numerical calculation of magnetic field distribution in the slotted air-gap of tangential surface permanent-magnet motors

    Directory of Open Access Journals (Sweden)

    Boughrara Kamel

    2009-01-01

    Full Text Available This paper deals with the analytical and numerical analysis of the flux density distribution in the slotted air gap of permanent magnet motors with surface mounted tangentially magnetized permanent magnets. Two methods for magnetostatic field calculations are developed. The first one is an analytical method in which the effect of stator slots is taken into account by modulating the magnetic field distribution by the complex relative air gap permeance. The second one is a numerical method using 2-D finite element analysis with consideration of Dirichlet and anti-periodicity (periodicity boundary conditions and Lagrange Multipliers for simulation of movement. The results obtained by the analytical method are compared to the results of finite-element analysis.

  3. Neutron-gamma competition for $\\beta$-delayed neutron emission

    CERN Document Server

    Mumpower, Matthew; Moller, Peter

    2016-01-01

    We present a coupled Quasi-particle Random Phase Approximation and Hauser-Feshbach (QRPA+HF) model for calculating delayed particle emission. This approach uses microscopic nuclear structure information which starts with Gamow-Teller strength distributions in the daughter nucleus, and then follows the statistical decay until the initial available excitation energy is exhausted. Explicitly included at each particle emission stage is $\\gamma$-ray competition. We explore this model in the context of neutron emission of neutron-rich nuclei and find that neutron-gamma competition can lead to both increases and decreases in neutron emission probabilities, depending on the system considered. A second consequence of this formalism is a prediction of more neutrons on average being emitted after $\\beta$-decay for nuclei near the neutron dripline compared to models that do not consider the statistical decay.

  4. On the calculation of air-sea fluxes of CO2 in the presence of temperature and salinity gradients

    Science.gov (United States)

    Woolf, D. K.; Land, P. E.; Shutler, J. D.; Goddijn-Murphy, L. M.; Donlon, C. J.

    2016-02-01

    The presence of vertical temperature and salinity gradients in the upper ocean and the occurrence of variations in temperature and salinity on time scales from hours to many years complicate the calculation of the flux of carbon dioxide (CO2) across the sea surface. Temperature and salinity affect the interfacial concentration of aqueous CO2 primarily through their effect on solubility with lesser effects related to saturated vapor pressure and the relationship between fugacity and partial pressure. The effects of temperature and salinity profiles in the water column and changes in the aqueous concentration act primarily through the partitioning of the carbonate system. Climatological calculations of flux require attention to variability in the upper ocean and to the limited validity of assuming "constant chemistry" in transforming measurements to climatological values. Contrary to some recent analysis, it is shown that the effect on CO2 fluxes of a cool skin on the sea surface is large and ubiquitous. An opposing effect on calculated fluxes is related to the occurrence of warm layers near the surface; this effect can be locally large but will usually coincide with periods of low exchange. A salty skin and salinity anomalies in the upper ocean also affect CO2 flux calculations, though these haline effects are generally weaker than the thermal effects.

  5. 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.

  6. European inter-comparison of Monte Carlo codes users for the uncertainty calculation of the kerma in air beside a caesium-137 source; Intercomparaison europeenne d'utilisateurs de codes monte carlo pour le calcul d'incertitudes sur le kerma dans l'air aupres d'une source de cesium-137

    Energy Technology Data Exchange (ETDEWEB)

    De Carlan, L.; Bordy, J.M.; Gouriou, J. [CEA Saclay, LIST, Laboratoire National Henri Becquerel, Laboratoire de Metrologie de la Dose 91 - Gif-sur-Yvette (France)

    2010-07-01

    Within the frame of the CONRAD European project (Coordination Network for Radiation Dosimetry), and more precisely within a work group paying attention to uncertainty assessment in computational dosimetry and aiming at comparing different approaches, the authors report the simulation of an irradiator containing a caesium 137 source to calculate the kerma in air as well as its uncertainty due to different parameters. They present the problem geometry, recall the studied issues (kerma uncertainty, influence of capsule source, influence of the collimator, influence of the air volume surrounding the source). They indicate the codes which have been used (MNCP, Fluka, Penelope, etc.) and discuss the obtained results for the first issue

  7. 混合堆增殖钍基燃料组件中子学分析%Neutronics Calculation of Fusion-Fission Hybrid Breeding Thorium Fuel Assembly

    Institute of Scientific and Technical Information of China (English)

    马续波; 陈义学; 全国萍; 王悦; 韩静茹; 陆道纲

    2012-01-01

    A preliminary comparative study of the physical properties among 17×17 fuel assembly in PWRs for prototype between uranium assembly and hybrid breeding thorium-based assembly has been investigated respectively using the DRAGON software. The parameters such as fuel temperature coefficient, moderator temperature coefficient and that variation as a function of operation period have been investigated. Results show that the neutron properties of uranium-based assembly and hybrid breeding thorium-based assembly are similitude, but MA mass of hybrid breeding thorium-based assembly is evidently less than those of the uranium assembly.%采用压水堆17×17燃料组件模型,用燃料组件参数计算程序DRAGON分别对混合堆增殖钍燃料组件和全铀组件的中子学特性进行了研究,分析组件的燃料温度系数、慢化剂温度系数及其与燃耗的关系.计算结果表明,混合堆增殖钍燃料组件和全铀组件的中子特性相似,但钍燃料组件中的乏燃料组件中的次锕系核素(MA)的含量明显减少.

  8. 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.

  9. A method of calculating of the thermodynamic properties and the composition of the explosion products of hydrocarbons and air under partial chemical equilibrium

    Science.gov (United States)

    Shargatov, V. A.

    2016-11-01

    We examined the approximate method to calculate composition and thermodynamic parameters of hydrocarbons-air nonequilibrium explosion products based on the assumption of the existence of a partial chemical equilibrium. With excellent accuracy of calculating thermodynamic properties and species mass fraction the respective stiff system of detailed kinetics differential equations can be replaced by the one differential equation or the two differential equations and a system of algebraic equations. This method is always consistent with the detailed kinetic mechanism. The constituent equations of the method were derived and the respective computer code written. We examine the applicability of the method by solving the test problem. The proposed method simulation results are in excellent agreement with the detailed kinetics model results corresponding the stiff ordinary differential equation solver including NO time histories.

  10. Coded source neutron imaging

    Energy Technology Data Exchange (ETDEWEB)

    Bingham, Philip R [ORNL; Santos-Villalobos, Hector J [ORNL

    2011-01-01

    Coded aperture techniques have been applied to neutron radiography to address limitations in neutron flux and resolution of neutron detectors in a system labeled coded source imaging (CSI). By coding the neutron source, a magnified imaging system is designed with small spot size aperture holes (10 and 100 m) for improved resolution beyond the detector limits and with many holes in the aperture (50% open) to account for flux losses due to the small pinhole size. An introduction to neutron radiography and coded aperture imaging is presented. A system design is developed for a CSI system with a development of equations for limitations on the system based on the coded image requirements and the neutron source characteristics of size and divergence. Simulation has been applied to the design using McStas to provide qualitative measures of performance with simulations of pinhole array objects followed by a quantitative measure through simulation of a tilted edge and calculation of the modulation transfer function (MTF) from the line spread function. MTF results for both 100um and 10um aperture hole diameters show resolutions matching the hole diameters.

  11. Coded source neutron imaging

    Science.gov (United States)

    Bingham, Philip; Santos-Villalobos, Hector; Tobin, Ken

    2011-03-01

    Coded aperture techniques have been applied to neutron radiography to address limitations in neutron flux and resolution of neutron detectors in a system labeled coded source imaging (CSI). By coding the neutron source, a magnified imaging system is designed with small spot size aperture holes (10 and 100μm) for improved resolution beyond the detector limits and with many holes in the aperture (50% open) to account for flux losses due to the small pinhole size. An introduction to neutron radiography and coded aperture imaging is presented. A system design is developed for a CSI system with a development of equations for limitations on the system based on the coded image requirements and the neutron source characteristics of size and divergence. Simulation has been applied to the design using McStas to provide qualitative measures of performance with simulations of pinhole array objects followed by a quantitative measure through simulation of a tilted edge and calculation of the modulation transfer function (MTF) from the line spread function. MTF results for both 100μm and 10μm aperture hole diameters show resolutions matching the hole diameters.

  12. 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.

  13. 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.

  14. Equation for Calculating the Concentration of Solvent in Air That Discriminates between Exposure and Non-exposure Based on Biomarker Concentrations in the Urine of Workers

    Directory of Open Access Journals (Sweden)

    Kondo,Yoshiro

    2006-12-01

    Full Text Available To develop a new method for evaluating the intensity of workers’ exposures to toluene alone or toluene in mixed solvents, regression equations were calculated between the concentrations of toluene to which workers were exposed and the concentrations of hippuric acid or toluene in workers’ urine samples taken at the end of their shifts. Thereafter, the discriminant exposure concentration of the solvents in air, which was the concentration considered to discriminate exposure from non-exposure within a fi xed level of error using fi ducial ranges of individual specimens (DEC-I or using confi dence ranges of regression equation (DEC-R, was measured by a scale. The devised equations were applied to calculate DEC-I or DEC-R accurately using the formulas expressing a regression line and its fi ducial ranges or confi dence ranges. The equations can calculate not only more precise values of DEC-I or DEC-R than can be measured by a scale, but can also calculate values corresponding to any level of error. Moreover, DEC-I and DEC-R can be defi ned by the equations. The concentration capable of discriminating TLV (threshold limit value exposure from non-TLV exposure was estimated using fi ducial ranges (DTL-I and then using confi dence ranges of the regression equation (DTL-R.

  15. 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.

  16. Summary of Simplified Two Time Step Method for Calculating Combustion Rates and Nitrogen Oxide Emissions for Hydrogen/Air and Hydrogen/Oxygen

    Science.gov (United States)

    Marek, C. John; Molnar, Melissa

    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 (greater than l x 10(exp -20)) moles per cc) in the mixture which gives the chemical kinetic time as a function of the instantaneous fuel and water mole concentrations, pressure and temperature (T(sub 4)). 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 H2/O2. A similar correlation is also developed using data from NASA's Chemical Equilibrium Applications (CEA) code to determine the equilibrium temperature (T(sub 4)) as a function of overall fuel/air ratio, pressure and initial temperature (T(sub 3)). High values of the regression coefficient R squared are obtained.

  17. Air temperature variability over three glaciers in the Ortles-Cevedale (Italian Alps): effects of glacier disintegration, intercomparison of calculation methods, and impacts on mass balance modeling

    Science.gov (United States)

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

    2014-12-01

    Glacier mass balance models rely on accurate spatial calculation of input data, in particular air temperature. Lower temperatures (the so-called glacier cooling effect), and lower temperature variability (the so-called glacier damping effect) generally occur over glaciers, compared to ambient conditions. These effects, which depend on the geometric characteristics of glaciers and display a high spatial and temporal variability, have been mostly investigated on medium- to large-size glaciers so far, while observations on smaller ice bodies are scarce. Using a dataset from 8 on-glacier and 4 off-glacier weather stations, collected in summer 2010 and 2011, we analyzed the air temperature variability and wind regime over three different glaciers in the Ortles-Cevedale. The magnitude of the cooling effect and the occurrence of katabatic boundary layer (KBL) processes showed remarkable differences among the three ice bodies, suggesting the likely existence of important reinforcing mechanisms during glacier decay and disintegration. None of the methods proposed in the literature for calculating on-glacier temperature from off-glacier data fully reproduced our observations. Among them, the more physically-based procedure of Greuell and Böhm (1998) provided the best overall results where the KBL prevail, but it was not effective elsewhere (i.e. on smaller ice bodies and close to the glacier margins). The accuracy of air temperature estimations strongly impacted the results from a mass balance model which was applied to the three investigated glaciers. Most importantly, even small temperature deviations caused distortions in parameter calibration, thus compromising the model generalizability.

  18. Neutron whispering gallery

    Science.gov (United States)

    Nesvizhevsky, Valery V.; Voronin, Alexei Yu.; Cubitt, Robert; Protasov, Konstantin V.

    2010-02-01

    The `whispering gallery' effect has been known since ancient times for sound waves in air, later in water and more recently for a broad range of electromagnetic waves: radio, optics, Roentgen and so on. It consists of wave localization near a curved reflecting surface and is expected for waves of various natures, for instance, for atoms and neutrons. For matter waves, it would include a new feature: a massive particle would be settled in quantum states, with parameters depending on its mass. Here, we present for the first time the quantum whispering-gallery effect for cold neutrons. This phenomenon provides an example of an exactly solvable problem analogous to the `quantum bouncer'; it is complementary to the recently discovered gravitationally bound quantum states of neutrons . These two phenomena provide a direct demonstration of the weak equivalence principle for a massive particle in a pure quantum state. Deeply bound whispering-gallery states are long-living and weakly sensitive to surface potential; highly excited states are short-living and very sensitive to the wall potential shape. Therefore, they are a promising tool for studying fundamental neutron-matter interactions, quantum neutron optics and surface physics effects.

  19. 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

  20. 航空发动机空气系统验算标定%Calibration of Calculation for Aero-Engine Air System

    Institute of Scientific and Technical Information of China (English)

    呼艳丽; 徐连强; 赵维维

    2014-01-01

    利用流量特性试验得到的相关流路元件的流阻计算模型,和旋转状态下阶梯齿风阻温升计算方法,通过调节封严篦齿间隙等参数,对航空发动机空气系统的压力、温度进行验算标定,并根据验算结果分析发现后续试验中存在和需要注意的问题。验算标定结果表明:通过对发动机试验工况的验算标定,可发现试验中存在的问题,较准确地模拟出后续试验中的问题和试验风险,确保发动机的工作安全,并为空气系统的进一步改进和优化提供依据。%A calibration method for aero-engine air system was presented, with the air system throttle ele-ments flow loss coefficient experiment data and the rotating step labyrinth seal windage temperature rise cal-culation method. The pressure and temperature of air system were calibrated by adjusting the labyrinth seal clearance. Some possible and noteworthy problems in the following tests were found out according to the cal-ibrated results. The results show that through the calibration of experimental conditions, the problems could be revealed and the risk of the following tests can be simulated perfectly to ensure the safety of engine work-ing which could be referential for the further improvements of engine's air system.

  1. 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.

  2. Dose measurements around spallation neutron sources.

    Science.gov (United States)

    Fragopoulou, M; Stoulos, S; Manolopoulou, M; Krivopustov, M; Zamani, M

    2008-01-01

    Neutron dose measurements and calculations around spallation sources appear to be of great importance in shielding research. Two spallation sources were irradiated by high-energy proton beams delivered by the Nuclotron accelerator (JINR), Dubna. Neutrons produced by the spallation sources were measured by using solid-state nuclear track detectors. In addition, neutron dose was calculated after polyethylene and concrete, using a phenomenological model based on empirical relations applied in high-energy physics. The study provides an analytical and experimental neutron benchmark analysis using the transmission factor and a comparison between the experimental results and calculations.

  3. 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.

  4. 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

    The hexagonal high-temperature polymorph of LiBH4 is stabilized by solid solution with LiI to exhibit superionic Li+ ionic conductivity at room temperature. Herein, the mechanisms for the Li+ diffusion are investigated for the first time by density functional theory (DFT) calculations coupled...... defect sites, giving rise to high defect mobility. QENS results at 380 K show long-range diffusion of Li+, with jump lengths of one unit cell and jump rates in agreement with those obtained from DFT, and the application of the bias potential increases the diffusion constant by a factor of 2. At 300 K...

  5. Carbon neutron star atmospheres

    CERN Document Server

    Suleimanov, V F; Pavlov, G G; Werner, K

    2013-01-01

    The accuracy of measuring the basic parameters of neutron stars is limited in particular by uncertainties in chemical composition of their atmospheres. For example, atmospheres of thermally - emitting neutron stars in supernova remnants might have exotic chemical compositions, and for one of them, the neutron star in CasA, a pure carbon atmosphere has recently been suggested by Ho & Heinke (2009). To test such a composition for other similar sources, a publicly available detailed grid of carbon model atmosphere spectra is needed. We have computed such a grid using the standard LTE approximation and assuming that the magnetic field does not exceed 10^8 G. The opacities and pressure ionization effects are calculated using the Opacity Project approach. We describe the properties of our models and investigate the impact of the adopted assumptions and approximations on the emergent spectra.

  6. Uniformly rotating neutron stars

    CERN Document Server

    Boshkayev, Kuantay

    2016-01-01

    In this chapter we review the recent results on the equilibrium configurations of static and uniformly rotating neutron stars within the Hartle formalism. We start from the Einstein-Maxwell-Thomas-Fermi equations formulated and extended by Belvedere et al. (2012, 2014). We demonstrate how to conduct numerical integration of these equations for different central densities ${\\it \\rho}_c$ and angular velocities $\\Omega$ and compute the static $M^{stat}$ and rotating $M^{rot}$ masses, polar $R_p$ and equatorial $R_{\\rm eq}$ radii, eccentricity $\\epsilon$, moment of inertia $I$, angular momentum $J$, as well as the quadrupole moment $Q$ of the rotating configurations. In order to fulfill the stability criteria of rotating neutron stars we take into considerations the Keplerian mass-shedding limit and the axisymmetric secular instability. Furthermore, we construct the novel mass-radius relations, calculate the maximum mass and minimum rotation periods (maximum frequencies) of neutron stars. Eventually, we compare a...

  7. Fabrication of Free Air Well Type Ionization Chamber and Calculational Assessment and Measurement of Its Operational Characteristics

    Directory of Open Access Journals (Sweden)

    Koroush Arbabi

    2007-12-01

    Full Text Available Introduction: Well type ionization chamber is a measuring device which is used to determine the activity of brachytherapy sources. The chamber has a cylindrical volume in which a cylindrical tube is mounted in the middle of the chamber. For the measurements, the brachytherapy sources are transferred to the middle of the tube. Materials and Methods: For designing the well type chamber, the measurement principals of well type chambers were considered and MCNP-4C code as a calculation tool was used. The designed chamber was simulated and the response of the chamber was evaluated. In this investigation, the chamber operational parameters such as operating voltage, leakage current, reproducibility, reference measuring point, recombination and polarization factors as well as response stability for 137Cs, 57Co and 241Am sources were studied. Results: The chamber leakage currents at the operating voltage in comparison to the chamber response for the measurement of the above mentioned sources were negligible. The responses of the fabricated chamber for these sources are reproducible and its reference measurement position for these sources was obtained at 6 cm from the bottom of the chamber. The recombination factor for the well type chamber was negligible and the polarization factor is close to 1. Therefore, these two factors were not considered in the measurements. The reproducibility of the measurements in different intervals shows the stability of the chamber response for each source. Also the results of the chamber current measurement in term of source strength were compared to the response of the simulated chamber for different source positions and energy ranges of the used sources. Discussion and Conclusion: The results show that the measurement of the reference positions for each source in the simulated and fabricated chamber is quite in a good agreement. Regarding the reliable operational properties of the fabricated chamber, this chamber can be

  8. Neutron dose rate for {sup 252} Cf AT source in medical applications

    Energy Technology Data Exchange (ETDEWEB)

    Paredes, L.; Balcazar, M. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico); Azorin, J. [UAM-I, 09340 Mexico D.F. (Mexico); Francois, J.L. [FI-UNAM, 04510 Mexico D.F. (Mexico)

    2006-07-01

    The AAPM TG-43 modified protocol was used for the calculation of the neutron dose rate of {sup 252}Cf sources for two tissue substitute materials, five normal tissues and six tumours. The {sup 252}Cf AT source model was simulated using the Monte Carlo MCNPX code in spherical geometry for the following factors: a) neutron air kerma strength conversion factor, b) dose rate constant, c) radial dose function, d) geometry factor, e) anisotropy function and f) neutron dose rate. The calculated dose rate in water at 1 cm and 90 degrees from the source long axis, using the Watt fission spectrum, was D{sub n}(r{sub 0}, {theta}{sub 0})= 1.9160 cGy/h-{mu}g. When this value is compared with Rivard et al. calculation using MCNP4B code, 1.8730 cGy/h-{mu}g, a difference of 2.30% is obtained. The results for the reference neutron dose rate in other media show how small variations in the elemental composition between the tissues and malignant tumours, produce variations in the neutron dose rate up to 12.25%. (Author)

  9. Development of fast neutron pinhole camera using nuclear emulsion for neutron emission profile measurement in KSTAR

    Science.gov (United States)

    Izumi, Y.; Tomita, H.; Nakayama, Y.; Hayashi, S.; Morishima, K.; Isobe, M.; Cheon, M. S.; Ogawa, K.; Nishitani, T.; Naka, T.; Nakano, T.; Nakamura, M.; Iguchi, T.

    2016-11-01

    We have developed a compact fast neutron camera based on a stack of nuclear emulsion plates and a pinhole collimator. The camera was installed at J-port of Korea superconducting tokamak advanced research at National Fusion Research Institute, Republic of Korea. Fast neutron images agreed better with calculated ones based on Monte Carlo neutron simulation using the uniform distribution of Deuterium-Deuterium (DD) neutron source in a torus of 40 cm radius.

  10. 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)

  11. Development of pulsed neutron uranium logging instrument

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Xin-guang, E-mail: wangxg@upc.edu.cn [School of Geosciences, China University of Petroleum, Qingdao 266580 (China); Engineering Research Center of Nuclear Technology Application (East China Institute of Technology), Ministry of Education, Nanchang 330013 (China); Liu, Dan [China Institute of Atomic Energy, Beijing 102413 (China); Zhang, Feng [School of Geosciences, China University of Petroleum, Qingdao 266580 (China)

    2015-03-15

    This article introduces a development of pulsed neutron uranium logging instrument. By analyzing the temporal distribution of epithermal neutrons generated from the thermal fission of {sup 235}U, we propose a new method with a uranium-bearing index to calculate the uranium content in the formation. An instrument employing a D-T neutron generator and two epithermal neutron detectors has been developed. The logging response is studied using Monte Carlo simulation and experiments in calibration wells. The simulation and experimental results show that the uranium-bearing index is linearly correlated with the uranium content, and the porosity and thermal neutron lifetime of the formation can be acquired simultaneously.

  12. Sensitivity of a new-developed neutron detector

    Institute of Scientific and Technical Information of China (English)

    PENG Tai-Ping; ZHU Xue-Bin; YANG Hong-Qiong; YANG Jian-Lun; YANG Gao-Zhao; LI Lin-Bo; SONG Xian-Cai

    2005-01-01

    We develop a kind of neutron detector, which consists of a polyethylene thin film and two PIN semiconductors connected face-to-face. The detector is insensitive to γ-rays. Its sensitivity to neutron has been calculated with MCNP program and calibrated by experiments, and the results indicate that the neutron sensitivity of the compensation detector will vary with polyethylene converter. The compensation PIN detector can be employed to measure pulse neutron in neutron and gamma mixture radiation field.

  13. Neutron flux calculation for central channel in first cycle of SPRR-300%300#研究堆首炉中央孔道中子通量密度计算

    Institute of Scientific and Technical Information of China (English)

    杨万奎; 曾和荣; 冷军; 刘耀光

    2012-01-01

    The physical model of the 300 # swimming pool research reactor(SPRR-300) based on the Monte Carlo code MCNP has been verified. Sophisticated modeling is conducted. An effective multiplication factor value of 1. 002 29 is obtained, existing a relative error of 0. 229% compared with the critical value. Meanwhile, a problem comes out that the interrupt and con-tinue-run with parallel version of MCNP doesn't work. The problem is solved through trail and error process. A reasonable application of flux tally average over a cell and flux tally at a point is suggested, namely the former is prior to the latter to tally in big volume. Comparison between calculation results and experimental data shows that the thermal neutron flux has a deviation of 4. 6% at a power level of 3 MW. That is to say, the calculated value and the experimental value agree well with each other, and the neutron flux result is dependable.%基于MCNP程序对300#研究堆首炉堆芯进行精细建模,通过并行计算方式得到了实验临界棒位下堆芯的有效增殖因数为1.002 29,与临界值之间的相对误差为0.229%,验证了物理模型的正确性.探讨并解决了并行计算的中断与接续问题,提出了体通量计数与点探测器计数应用中的合理化建议,即对大体积空间计数时尽量使用体通量计数.计算值与实验值对比结果表明:两者在3 MW功率水平下热中子通量密度相差4.6%,符合得较好.

  14. 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)

  15. Forming images with thermal neutrons

    Science.gov (United States)

    Vanier, Peter E.; Forman, Leon

    2003-01-01

    Thermal neutrons passing through air have scattering lengths of about 20 meters. At further distances, the majority of neutrons emanating from a moderated source will scatter multiple times in the air before being detected, and will not retain information about the location of the source, except that their density will fall off somewhat faster than 1/r2. However, there remains a significant fraction of the neutrons that will travel 20 meters or more without scattering and can be used to create an image of the source. A few years ago, a proof-of-principle "camera" was demonstrated that could produce images of a scene containing sources of thermalized neutrons and could locate a source comparable in strength with an improvised nuclear device at ranges over 60 meters. The instrument makes use of a coded aperture with a uniformly redundant array of openings, analogous to those used in x-ray and gamma cameras. The detector is a position-sensitive He-3 proportional chamber, originally used for neutron diffraction. A neutron camera has many features in common with those designed for non-focusable photons, as well as some important differences. Potential applications include detecting nuclear smuggling, locating non-metallic land mines, assaying nuclear waste, and surveying for health physics purposes.

  16. Performance characteristics of the MIT fission converter based epithermal neutron beam.

    Science.gov (United States)

    Riley, K J; Binns, P J; Harling, O K

    2003-04-07

    A pre-clinical characterization of the first fission converter based epithermal neutron beam (FCB) designed for boron neutron capture therapy (BNCT) has been performed. Calculated design parameters describing the physical performance of the aluminium and Teflon filtered beam were confirmed from neutron fluence and absorbed dose rate measurements performed with activation foils and paired ionization chambers. The facility currently provides an epithermal neutron flux of 4.6 x 10(9) n cm(-2) s(-1) in-air at the patient position that makes it the most intense BNCT source in the world. This epithermal neutron flux is accompanied by very low specific photon and fast neutron absorbed doses of 3.5 +/- 0.5 and 1.4 +/- 0.2 x 10(-13) Gy cm2, respectively. A therapeutic dose rate of 1.7 RBE Gy min(-1) is achievable at the advantage depth of 97 mm when boronated phenylalanine (BPA) is used as the delivery agent, giving an average therapeutic ratio of 5.7. In clinical trials of normal tissue tolerance when using the FCB, the effective prescribed dose is due principally to neutron interactions with the nonselectively absorbed BPA present in brain. If an advanced compound is considered, the dose to brain would instead be predominately from the photon kerma induced by thermal neutron capture in hydrogen and advantage parameters of 0.88 Gy min(-1), 121 mm and 10.8 would be realized for the therapeutic dose rate, advantage depth and therapeutic ratio, respectively. This study confirms the success of a new approach to producing a high intensity, high purity epithermal neutron source that attains near optimal physical performance and which is well suited to exploit the next generation of boron delivery agents.

  17. Neutron star structure from QCD

    Energy Technology Data Exchange (ETDEWEB)

    Fraga, Eduardo S. [Universidade Federal do Rio de Janeiro, Instituto de Fisica, Rio de Janeiro, RJ (Brazil); Kurkela, Aleksi [PH-TH, Case C01600, CERN, Theory Division, Geneva (Switzerland); University of Stavanger, Faculty of Science Technology, Stavanger (Norway); Vuorinen, Aleksi [University of Helsinki, Helsinki Institute of Physics and Department of Physics (Finland)

    2016-03-15

    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. (orig.)

  18. Neutron star structure from QCD

    Science.gov (United States)

    Fraga, Eduardo S.; Kurkela, Aleksi; Vuorinen, Aleksi

    2016-03-01

    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.

  19. 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.

  20. 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.

  1. Radioprotection calculations for MEGAPIE.

    Science.gov (United States)

    Zanini, L

    2005-01-01

    The MEGAwatt PIlot Experiment (MEGAPIE) liquid lead-bismuth spallation neutron source will commence operation in 2006 at the SINQ facility of the Paul Scherrer Institut. Such an innovative system presents radioprotection concerns peculiar to a liquid spallation target. Several radioprotection issues have been addressed and studied by means of the Monte Carlo transport code, FLUKA. The dose rates in the room above the target, where personnel access may be needed at times, from the activated lead-bismuth and from the volatile species produced were calculated. Results indicate that the dose rate level is of the order of 40 mSv h(-1) 2 h after shutdown, but it can be reduced below the mSv h(-1) level with slight modifications to the shielding. Neutron spectra and dose rates from neutron transport, of interest for possible damage to radiation sensitive components, have also been calculated.

  2. Neutron metrology in the HFR

    Energy Technology Data Exchange (ETDEWEB)

    Voorbraak, W.P.; Freudenreich, W.E.; Paardekooper, A.; Stecher-Rasmussen, F.; Verhagen, H.W.

    1991-11-01

    Results are presented of the ECN measurements at the filtered HFR beam HB11. The neutron measurements took place in the free beam at full power. Several gamma measurements were performed at full power under different conditions. The neutron spectrum was obtained by adjusting a calculated spectrum with experimental results from activation foils. The gamma data were obtained with thermoluminescent dosimeters. (author). 5 refs.; 4 figs.; 4 tabs.

  3. Properties of Rotating Neutron Star

    Directory of Open Access Journals (Sweden)

    Shailesh K. Singh

    2015-08-01

    Full Text Available Using the nuclear equation of states for a large variety of relativistic and non-relativistic force parameters, we calculate the static and rotating masses and radii of neutron stars. From these equation of states, we evaluate the properties of rotating neutron stars, such as rotational frequencies, moment of inertia, quadrupole deformation parameter, rotational ellipticity and gravitational wave strain amplitude. The estimated gravitational wave strain amplitude of the star is found to be~sim 10-23.

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

    Energy Technology Data Exchange (ETDEWEB)

    Maglieri, Robert, E-mail: robert.maglieri@mail.mcgill.ca; Evans, Michael; Seuntjens, Jan; Kildea, John [Medical Physics Unit, McGill University, Montreal, Quebec H4A 3J1 (Canada); Licea, Angel [Canadian Nuclear Safety Commission, Ottawa, Ontario K1P 5S9 (Canada)

    2015-11-15

    Purpose: 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. Methods: 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. Results: 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. Conclusions: The NNS may

  5. 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

  6. The measurement of neutron and neutron induced photon spectra in fusion reactor related assemblies

    CERN Document Server

    Unholzer, S; Klein, H; Seidel, K

    2002-01-01

    The spectral neutron and photon fluence (or flux) measured outside and inside of assemblies related to fusion reactor constructions are basic quantities of fusion neutronics. The comparison of measured spectra with the results of MCNP neutron and photon transport calculations allows a crucial test of evaluated nuclear data as generally used in fusion applications to be carried out. The experiments concern mixed neutron/photon fields with about the same intensity of the two components. An NE-213 scintillation spectrometer, well described by response matrices for both neutrons and photons, is used as proton-recoil and Compton spectrometer. The experiments described here in more detail address the background problematic of two applications, an iron benchmark experiment with an ns-pulsed neutron source and a deep penetration mock-up experiment for the investigation of the ITER in-board shield system. The measured spectral neutron and photon fluences are compared with spectra calculated with the MCNP code on the b...

  7. 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.

  8. Neutron area monitor with TLD 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., E-mail: ing_karen_guzman@yahoo.com.mx [Universidad Autonoma de Zacatecas, Unidad Academica de Estudios Nucleares, Calle Cipres No. 10. Fracc. La Penuela, 98068 Zacatecas (Mexico)

    2011-11-15

    The response of a passive neutron area monitor with pairs of thermoluminescent dosimeters has been calculated using the Monte Carlo code MCNP5. The response was calculated for one TLD 600 located at the center of a polyethylene cylinder, as moderator. When neutrons collide with the moderator lose their energy reaching the TLD with thermal energies where the ambient dose equivalent is calculated. The response was calculated for 47 monoenergetic neutron sources ranging from 1E(-9) to 20 MeV. Response was calculated using two irradiation geometries, one with an upper source and another with a lateral source. For both irradiation schemes the response was calculated with the TLDs in two positions, one parallel to the source and another perpendicular to the source. The advantage of this passive neutron monitor area is that can be used in locations with intense, pulsed and mixed radiation fields. (Author)

  9. Modeling of ground albedo neutrons to investigate seasonal cosmic ray-induced neutron variations measured at high-altitude stations

    Science.gov (United States)

    Hubert, G.; Pazianotto, M. T.; Federico, C. A.

    2016-12-01

    This paper investigates seasonal cosmic ray-induced neutron variations measured over a long-term period (from 2011 to 2016) in both the high-altitude stations located in medium geomagnetic latitude and Antarctica (Pic-du-Midi and Concordia, respectively). To reinforce analysis, modeling based on ground albedo neutrons simulations of extensive air showers and the solar modulation potential was performed. Because the local environment is well known and stable over time in Antarctica, data were used to validate the modeling approach. A modeled scene representative to the Pic-du-Midi was simulated with GEANT4 for various hydrogen properties (composition, density, and wet level) and snow thickness. The orders of magnitudes of calculated thermal fluence rates are consistent with measurements obtained during summers and winters. These variations are dominant in the thermal domain (i.e., En 20 MeV) is weakly impacted. The role of hydrogen content on ground albedo neutron generation was investigated with GEANT4 simulations. These investigations focused to mountain environment; nevertheless, they demonstrate the complexity of the local influences on neutron fluence rates.

  10. Prediction analysis of dose equivalent responses of neutron dosemeters used at a MOX fuel facility.

    Science.gov (United States)

    Tsujimura, N; Yoshida, T; Takada, C

    2011-07-01

    To predict how accurately neutron dosemeters can measure the neutron dose equivalent (rate) in MOX fuel fabrication facility work environments, the dose equivalent responses of neutron dosemeters were calculated by the spectral folding method. The dosemeters selected included two types of personal dosemeter, namely a thermoluminescent albedo neutron dosemeter and an electronic neutron dosemeter, three moderator-based neutron survey meters, and one special instrument called an H(p)(10) monitor. The calculations revealed the energy dependences of the responses expected within the entire range of neutron spectral variations observed in neutron fields at workplaces.

  11. 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.

  12. Neutron drops radii probed by the neutron skin thickness of nuclei

    CERN Document Server

    Zhao, P W

    2016-01-01

    Multi-neutron systems are crucial to understand the physics of neutron-rich nuclei and neutron stars. Neutron drops, neutrons confined in an external field, are investigated systematically in both non-relativistic and relativistic density functional theories and with ab initio calculations. We demonstrate a strong linear correlation, which is universal in the realm of mean-field models, between the root-mean-square (rms) radii of neutron drops and the neutron skin thickness of Pb-208 and Ca-48; i.e., the difference between the neutron and proton rms radii of a nucleus. Due to its high quality, this correlation can be used to deduce the radii of neutron drops from the measured neutron skin thickness in a model-independent way, and the radii obtained for neutron drops can provide a useful constraint for realistic three neutron forces. This correlation, together with high- precision measurements of the neutron skin thicknesses of Pb-208 and Ca-48, will have an enduring impact on the understanding of multi-neutro...

  13. A fundamental study on hyper-thermal neutrons for neutron capture therapy.

    Science.gov (United States)

    Sakurai, Y; Kobayashi, T; Kanda, K

    1994-12-01

    The utilization of hyper-thermal neutrons, which have an energy spectrum with a Maxwellian distribution at a higher temperature than room temperature (300 K), was studied in order to improve the thermal neutron flux distribution at depth in a living body for neutron capture therapy. Simulation calculations were carried out using a Monte Carlo code 'MCNP-V3' in order to investigate the characteristics of hyper-thermal neutrons, i.e. (i) depth dependence of the neutron energy spectrum, and (ii) depth distribution of the reaction rate in a water phantom for materials with 1/v neutron absorption. It is confirmed that hyper-thermal neutron irradiation can improve the thermal neutron flux distribution in the deeper areas in a living body compared with thermal neutron irradiation. When hyper-thermal neutrons with a 3000 K Maxwellian distribution are incident on a body, the reaction rates of 1/v materials such as 14N, 10B etc are about twice that observed for incident thermal neutrons at 300 K, at a depth of 5 cm. The limit of the treatable depth for tumours having 30 ppm 10B is expected to be about 1.5 cm greater by utilizing hyper-thermal neutrons at 3000 K compared with the incidence of thermal neutrons at 300 K.

  14. Neutron Repulsion

    CERN Document Server

    Manuel, Oliver K

    2011-01-01

    Earth is connected gravitationally, magnetically and electrically to its heat source - a neutron star that is obscured from view by waste products in the photosphere. Neutron repulsion is like the hot filament in an incandescent light bulb. Excited neutrons are emitted from the solar core and decay into hydrogen that glows in the photosphere like a frosted light bulb. Neutron repulsion was recognized in nuclear rest mass data in 2000 as the overlooked source of energy, the keystone of an arch that locked together these puzzling space-age observations: 1.) Excess 136Xe accompanied primordial helium in the stellar debris that formed the solar system (Fig. 1); 2.) The Sun formed on the supernova core (Fig. 2); 3.) Waste products from the core pass through an iron-rich mantle, selectively carrying lighter elements and lighter isotopes of each element into the photosphere (Figs. 3-4); and 4.) Neutron repulsion powers the Sun and sustains life (Figs. 5-7). Together these findings offer a framework for understanding...

  15. Neutron-Mirror Neutron Oscillations in a Residual Gas Environment

    Science.gov (United States)

    Varriano, Louis; Kamyshkov, Yuri

    2017-01-01

    A precise measurement of the neutron lifetime is important for calculating the rate at which nucleosynthesis occurred after the Big Bang. The history of neutron lifetime measurements has demonstrated impressive continuous improvement in experimental technique and in accuracy. However, two most precise recent measurements performed by different techniques differ by about 3 standard deviations. This difference of 9.2 seconds can possibly be resolved by future experiments, but it may also be evidence of a mirror matter effect present in these experiments. Both mirror matter, a candidate for dark matter, and ordinary matter can have similar properties and self-interactions but will interact only gravitationally with each other, in accordance with observational evidence of dark matter. Three separate experiments have been performed in the last decade to detect the possibility of neutron-mirror neutron oscillations. This work provides a formalism for understanding the interaction of the residual gas in an experiment with ultra-cold neutrons. This residual gas effect was previously considered negligible but can have a significant impact on the probability of neutron-mirror neutron transition.

  16. 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.

  17. Improvement of Delayed Neutron Counting System

    Institute of Scientific and Technical Information of China (English)

    YUAN; Guo-jun; XIAO; Cai-jin; YANG; Wei; ZHANG; Gui-ying; JIN; Xiang-chun; WANG; Ping-sheng; NI; Bang-fa

    2012-01-01

    <正>A new delayed neutron counting system, which is good at qualitative and quantitative analysis of fissionable nuclide mixture, will be established at China Advanced Research Reactor (CARR). We use 3 He proportional counters to count the delayed neutrons after the samples irradiated by reactor neutrons, including U3O8-stantard, uranium ore and enriched uranium. Then, the counting efficiency and limit of this system were calculated.

  18. Calculation of gas temperature at the outlet of the combustion chamber and in the air-gas channel of a gas-turbine unit by data of acceptance tests in accordance with ISO

    Science.gov (United States)

    Kostyuk, A. G.; Karpunin, A. P.

    2016-01-01

    This article describes a high accuracy method enabling performance of the calculation of real values of the initial temperature of a gas turbine unit (GTU), i.e., the gas temperature at the outlet of the combustion chamber, in a situation where manufacturers do not disclose this information. The features of the definition of the initial temperature of the GTU according to ISO standards were analyzed. It is noted that the true temperatures for high-temperature GTUs is significantly higher than values determined according to ISO standards. A computational procedure for the determination of gas temperatures in the air-gas channel of the gas turbine and cooling air consumptions over blade rims is proposed. As starting equations, the heat balance equation and the flow mixing equation for the combustion chamber are assumed. Results of acceptance GTU tests according to ISO standards and statistical dependencies of required cooling air consumptions on the gas temperature and the blade metal are also used for calculations. An example of the calculation is given for one of the units. Using a developed computer program, the temperatures in the air-gas channel of certain GTUs are calculated, taking into account their design features. These calculations are performed on the previously published procedure for the detailed calculation of the cooled gas turbine subject to additional losses arising because of the presence of the cooling system. The accuracy of calculations by the computer program is confirmed by conducting verification calculations for the GTU of the Mitsubishi Comp. and comparing results with published data of the company. Calculation data for temperatures were compared with the experimental data and the characteristics of the GTU, and the error of the proposed method is estimated.

  19. The Realization of Air Quality Index Calculation Method based on VBA in Excel%空气质量指数计算方法在Excel中VBA的实现

    Institute of Scientific and Technical Information of China (English)

    曾明智; 韩波; 莫海萍

    2014-01-01

    Air quality index calculation method is the basis of the release of important urban air quality information, which is widely used in air quality assessment in urban air automatic monitor-ing. According to the calculation method of air quality index (AQI) from the Technical Regulation on Ambient Air Quality Index (on trial, HJ 633-2012), the macro-program based on VBA in Ex-cellwas built to achieve the easy release of important daily urban air quality information, which can relieve the users from their repeated filling and inserting function.%空气质量指数计算方法是目前全国各城市空气质量重要信息发布的基础,广泛应用于城市空气自动监测的空气质量评价。依据《HJ 633-2012环境空气质量指数(AQI)技术规定(试行)》发布环境空气质量指数(AQI)计算方法,利用VBA在Excel下编写宏程序,可以轻松实现空气质量日报信息的发布,使用户从手工重复性的填充和插入函数操作方式中彻底解放出来。

  20. Monte-Carlo simulations of elastically backscattered neutrons from hidden explosives using three different neutron sources

    Energy Technology Data Exchange (ETDEWEB)

    ElAgib, I. [College of Science, King Saud University, P.O. Box 2455 (Saudi Arabia)], E-mail: elagib@ksu.edu.sa; Elsheikh, N. [College of Applied and Industrial Science, University of Juba, Khartoum, P.O. Box 321 (Sudan); AlSewaidan, H. [College of Science, King Saud University, P.O. Box 2455 (Saudi Arabia); Habbani, F. [Faculty of Science, Physics Department, University of Khartoum, Khartoum, P.O. Box 321 (Sudan)

    2009-01-15

    Calculations of elastically backscattered (EBS) neutrons from hidden explosives buried in soil were performed using Monte-Carlo N-particle transport code MCNP5. Three different neutron sources were used in the study. The study re-examines the performance of the neutron backscattering methods in providing identification of hidden explosives through their chemical composition. The EBS neutron energy spectra of fast and slow neutrons of the major constituent elements in soil and an explosive material in form of TNT have shown definite structures that can be used for the identification of a buried landmine.

  1. 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.

  2. Neutronic design of the ITER radial neutron camera

    Energy Technology Data Exchange (ETDEWEB)

    Petrizzi, L. [Associazione EURATOM-ENEA sulla Fusione, ENEA Centro Ricerche, C.P. 65, 00044 Frascati, Rome (Italy)], E-mail: petrizzi@frascati.enea.it; Barnsley, R. [EFDA CSU-Garching (Germany); Bertalot, L.; Esposito, B. [Associazione EURATOM-ENEA sulla Fusione, ENEA Centro Ricerche, C.P. 65, 00044 Frascati, Rome (Italy); Haskell, H. [ITER International Team, Garching (Germany); Mainardi, E.; Marocco, D.; Podda, S. [Associazione EURATOM-ENEA sulla Fusione, ENEA Centro Ricerche, C.P. 65, 00044 Frascati, Rome (Italy); Walker, C. [ITER International Team, Garching (Germany); Villari, S. [Associazione EURATOM-ENEA sulla Fusione, ENEA Centro Ricerche, C.P. 65, 00044 Frascati, Rome (Italy)

    2007-10-15

    This paper summarizes the work, performed in the frame of various EFDA contracts during 2004-2005, on the design review and upgrade of the ITER radial neutron camera (RNC). The RNC, which should provide information on the spatial distribution and energy spectrum of the neutron emission, consists of an ex-vessel system (fan-like collimator with 12 x 3 lines of sights) and an in-vessel system with further 9 lines for a full coverage of the plasma. A Monte Carlo code (MCNP) has been used for the neutronic calculations. The basic ITER model has been developed from the CATIA drawings to include the RNC with all details relevant for the neutronic analysis. In the model the collimator diameters have been set to 2 and 4 cm, respectively, for the ex-vessel and in-vessel systems. A detailed space dependent fusion neutron source (DD and DT phases in various plasma scenarios) has been used with a consistent ion temperature radial profile. A special variance reduction treatment has been developed so that neutrons reach the far regions in the high collimated neutron beam and score with a satisfying statistical error. Neutron and photon fluxes and spectra have been calculated. Approximately, one neutron out of 10{sup 11} emitted in all the plasma reaches a single ex-vessel detector. Therefore, for an emission rate of 1.8 x 10{sup 20} n/s (corresponding to 500 MW fusion power) the flux on the detectors is in the range (1-5) x 10{sup 8} n/(cm{sup 2} s) depending on the poloidal orientation. The fraction of scattered neutrons (>1 MeV) is lower than few % of the total. A measurement simulation software tool (MSST) performing asymmetric Abel inversion of simulated measured neutron signals has also been developed for line of sight and design optimization. Combining information from MCNP calculations and MSST, it has been possible to evaluate the performance of the RNC, check whether the present design of the RNC meets the measurement requirements and optimize the RNC design.

  3. Revision to the humidity correction equation in the calculation formulae of the air refractive index based on a phase step interferometer with three frequency-stabilized lasers

    Science.gov (United States)

    Chen, Qianghua; Zhang, Mengce; Liu, Shuaijie; He, Yongxi; Luo, Huifu; Luo, Jun; Lv, Weiwei

    2016-12-01

    At present the formulae proposed by G Boensch and E Potulski in 1998 (Boensch and Potulski 1998 Metrologia 35 133-9) are mostly used to calculate the air refractive index. However, the humidity correction equation in the formulae is derived by using the light source of a Cd lamp whose light frequency stability is poor and at a narrow temperature range, around 20 °C. So it is no longer suitable in present optical precision measurements. To solve this problem, we propose a refractive index measurement system based on phase step interferometer with three frequency stabilized lasers (532 nm, 633 nm, 780 nm), corrected coefficients of the humidity are measured and a corresponding revised humidity correction equation is acquired. Meanwhile, the application temperature range is extended from 14.6 °C to 25.0 °C. The experiment comparison results at the temperature of 22.2-23.2 °C show the accuracy by the presented equation is better than that of Boensch and Potulski.

  4. NIST Calibration of a Neutron Spectrometer ROSPEC.

    Science.gov (United States)

    Heimbach, Craig

    2006-01-01

    A neutron spectrometer was acquired for use in the measurement of National Institute of Standards and Technology neutron fields. The spectrometer included options for the measurement of low and high energy neutrons, for a total measurement range from 0.01 eV up to 17 MeV. The spectrometer was evaluated in calibration fields and was used to determine the neutron spectrum of an Americium-Beryllium neutron source. The calibration fields used included bare and moderated (252)Cf, monoenergetic neutron fields of 2.5 MeV and 14 MeV, and a thermal-neutron beam. Using the calibration values determined in this exercise, the spectrometer gives a good approximation of the neutron spectrum, and excellent values for neutron fluence, for all NIST calibration fields. The spectrometer also measured an Americium-Beryllium neutron field in a NIST exposure facility and determined the field quite well. The spectrometer measured scattering effects in neutron spectra which previously could be determined only by calculation or integral measurements.

  5. 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.

  6. Neutron and photon spectra in LINACs.

    Science.gov (United States)

    Vega-Carrillo, H R; Martínez-Ovalle, S A; Lallena, A M; Mercado, G A; Benites-Rengifo, J L

    2012-12-01

    A Monte Carlo calculation, using the MCNPX code, was carried out in order to estimate the photon and neutron spectra in two locations of two linacs operating at 15 and 18 MV. Detailed models of both linac heads were used in the calculations. Spectra were estimated below the flattening filter and at the isocenter. Neutron spectra show two components due to evaporation and knock-on neutrons. Lethargy spectra under the filter were compared to the spectra calculated from the function quoted by Tosi et al. that describes reasonably well neutron spectra beyond 1 MeV, though tends to underestimate the energy region between 10(-6) and 1 MeV. Neutron and the Bremsstrahlung spectra show the same features regardless of the linac voltage.

  7. Optimization of the beam shaping assembly in the D-D neutron generators-based BNCT using the response matrix method.

    Science.gov (United States)

    Kasesaz, Y; Khalafi, H; Rahmani, F

    2013-12-01

    Optimization of the Beam Shaping Assembly (BSA) has been performed using the MCNP4C Monte Carlo code to shape the 2.45 MeV neutrons that are produced in the D-D neutron generator. Optimal design of the BSA has been chosen by considering in-air figures of merit (FOM) which consists of 70 cm Fluental as a moderator, 30 cm Pb as a reflector, 2mm (6)Li as a thermal neutron filter and 2mm Pb as a gamma filter. The neutron beam can be evaluated by in-phantom parameters, from which therapeutic gain can be derived. Direct evaluation of both set of FOMs (in-air and in-phantom) is very time consuming. In this paper a Response Matrix (RM) method has been suggested to reduce the computing time. This method is based on considering the neutron spectrum at the beam exit and calculating contribution of various dose components in phantom to calculate the Response Matrix. Results show good agreement between direct calculation and the RM method.

  8. “核-光转换”中子探测系统物理结构的模拟研究%Simulation Calculation on Nuclear- Optical Converters for Neutron Detection

    Institute of Scientific and Technical Information of China (English)

    张建华; 李波均; 彭太平

    2012-01-01

    “核-光转换”中子探测器是以惰性气体为介质将裂变碎片能量转换为光辐射的裂变室,拥有电离探测器所没有的优点:不需要供电电源;信号传输方式采用光导或光纤,而不是绝缘电缆;对伽马辐射极不灵敏;输出信号较大,可以避免在探测器附近使用前置放大器.根据“核-光转换”中子测量系统的特点,采用Geant4模拟了铀裂变靶厚度、惰性气体成分、腔体材料等对到达惰性气体的裂变碎片和可见光的影响,给出了NOC结构设计的最佳参数和中子能量响应.%Nucear - optical converters ( NOC) are fission chambers based upon fission fragement energy conversion to optical radiation in gas luminescent media. NOCs offer a number of potential advantages over ionization detectors. The detectors require no power supply. Signals are transmitted via light - pipe or fiber optics rather than insulated electrical cable. The detectors are less sensitive to gamma radiation. NOC can produce large signals , obviating the need for pre - amplifiers near the detector. According to the characteristics of NOC, under the different conditions, like as the thickness of U target, type of inert gases, material of container , fission fragments and light intensities was simulated by Geant4; We obtained the optimal parameter to design the NOCs, and calculated the energy response of neutron.

  9. Proton Fraction in Neutron Stars

    Institute of Scientific and Technical Information of China (English)

    张丰收; 陈列文

    2001-01-01

    The proton fraction in β-stable neutron stars is investigated within the framework of the Skyrme-Hartree-Fock theory using the extended Skyrme effective interaction for the first time. The calculated results show that the proton fraction disappears at high density, which implies that the pure neutron matter may exist in the interior of neutron stars. The incompressibility of the nuclear equation-of-state is shown to be more important to determine the proton fraction. Meanwhile, it is indicated that the addition of muons in neutron stars will change the proton fraction. It is also found that the higher-order terms of the nuclear symmetry energy have obvious effects on the proton fraction and the parabolic law of the nuclear symmetry energy is not enough to determine the proton fraction.

  10. Radiation shielding for neutron guides

    Science.gov (United States)

    Ersez, T.; Braoudakis, G.; Osborn, J. C.

    2006-11-01

    Models of the neutron guide shielding for the out of bunker guides on the thermal and cold neutron beam lines of the OPAL Reactor (ANSTO) were constructed using the Monte Carlo code MCNP 4B. The neutrons that were not reflected inside the guides but were absorbed by the supermirror (SM) layers were noted to be a significant source of gammas. Gammas also arise from neutrons absorbed by the B, Si, Na and K contained in the glass. The proposed shielding design has produced compact shielding assemblies. These arrangements are consistent with safety requirements, floor load limits, and cost constraints. To verify the design a prototype was assembled consisting of 120 mm thick Pb(96%)Sb(4%) walls resting on a concrete block. There was good agreement between experimental measurements and calculated dose rates for bulk shield regions.

  11. Calculation of neutron cross sections for sup 9 sup 0 Zr, sup 2 sup 0 sup 8 Pb and sup 2 sup 0 sup 9 Bi in the energy range of 0.5-25 MeV by using the optical model potentials

    CERN Document Server

    Miah, M M H; Faruque, M R I

    2003-01-01

    Neutron total cross sections and differential elastic scattering cross sections for the nuclides sup 9 sup 0 Zr, sup 2 sup 0 sup 8 Pb and sup 2 sup 0 sup 9 Bi were calculated using different global spherical optical potential (SOP) parameter sets at neutron energies from 0.5-25 MeV. Calculated cross sections for the corresponding nuclides were compared with their experimental data obtained by the EXFOR file to select the best fit parameter sets. It is found that the parameter sets of Ferer Rapaport for sup 9 sup 0 Zr and Bechetti and Greenless for sup 2 sup 0 sup 8 Pb and sup 2 sup 0 sup 9 Bi are the best fitted set to obtain the experimental data of total cross sections and angular distributions of these nuclides. (author)

  12. 高海拔城市污水处理厂表曝机供气量计算%Calculation on Air Supply Quantity of Surface Aerator for Sewage Treatment Plant of High-altitude City

    Institute of Scientific and Technical Information of China (English)

    刘奕

    2015-01-01

    高海拔地区气温和气压较低、水中饱和溶解氧浓度低、空气中氧含量低,在污水处理厂供气量计算上存在一定的特殊性。本文结合国外某高海拔城市污水处理厂氧化沟工艺的设计实践,对高海拔地区表曝机供气量的计算方法进行了总结,提出了计算要点和需要注意的问题。%Due to low air temperature and air pressure, low concentration of saturated dissolved oxygen in water and low oxygen content in air, calculation on air supply quantity of sewage treatment plant has a certain particularity in high-altitude area. In combination with design practice of oxidation ditch process in sewage treatment plant of a high-altitude city, the paper summarizes calculation method of air supply quantity of surface aerator in high-altitude area and puts forward calculation points and some problems needs to be paid.

  13. 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.

  14. Rapidly rotating neutron star progenitors

    Science.gov (United States)

    Postnov, K. A.; Kuranov, A. G.; Kolesnikov, D. A.; Popov, S. B.; Porayko, N. K.

    2016-12-01

    Rotating proto-neutron stars can be important sources of gravitational waves to be searched for by present-day and future interferometric detectors. It was demonstrated by Imshennik that in extreme cases the rapid rotation of a collapsing stellar core may lead to fission and formation of a binary proto-neutron star which subsequently merges due to gravitational wave emission. In this paper, we show that such dynamically unstable collapsing stellar cores may be the product of a former merger process of two stellar cores in a common envelope. We applied population synthesis calculations to assess the expected fraction of such rapidly rotating stellar cores which may lead to fission and formation of a pair of proto-neutron stars. We have used the BSE (Binary Star Evolution) population synthesis code supplemented with a new treatment of stellar core rotation during the evolution via effective core-envelope coupling, characterized by the coupling time, τc. The validity of this approach is checked by direct MESA calculations of the evolution of a rotating 15 M⊙ star. From comparison of the calculated spin distribution of young neutron stars with the observed one, reported by Popov and Turolla, we infer the value τc ≃ 5 × 105 yr. We show that merging of stellar cores in common envelopes can lead to collapses with dynamically unstable proto-neutron stars, with their formation rate being ˜0.1-1 per cent of the total core collapses, depending on the common envelope efficiency.

  15. Rapidly rotating neutron star progenitors

    Science.gov (United States)

    Postnov, K. A.; Kuranov, A. G.; Kolesnikov, D. A.; Popov, S. B.; Porayko, N. K.

    2016-08-01

    Rotating proto-neutron stars can be important sources of gravitational waves to be searched for by present-day and future interferometric detectors. It was demonstrated by Imshennik that in extreme cases the rapid rotation of a collapsing stellar core may lead to fission and formation of a binary proto-neutron star which subsequently merges due to gravitational wave emission. In the present paper, we show that such dynamically unstable collapsing stellar cores may be the product of a former merger process of two stellar cores in a common envelope. We applied population synthesis calculations to assess the expected fraction of such rapidly rotating stellar cores which may lead to fission and formation of a pair of proto-neutron stars. We have used the BSE population synthesis code supplemented with a new treatment of stellar core rotation during the evolution via effective core-envelope coupling, characterized by the coupling time, τc. The validity of this approach is checked by direct MESA calculations of the evolution of a rotating 15 M⊙ star. From comparison of the calculated spin distribution of young neutron stars with the observed one, reported by Popov and Turolla, we infer the value τc ≃ 5 × 105 years. We show that merging of stellar cores in common envelopes can lead to collapses with dynamically unstable proto-neutron stars, with their formation rate being ˜0.1 - 1% of the total core collapses, depending on the common envelope efficiency.

  16. Neutron tomography

    Science.gov (United States)

    Crump, James C., III; Richards, Wade J.; Shields, Kevin C.

    1995-07-01

    The McClellan Nuclear Radiation Center's (MNRC) staff in conjunction with a Cooperative Research and Development Agreement (CRDA) with the U.C. Santa Barbara facility has developed a system that can be used for aircraft inspection of jet engine blades. The problem was to develop an inspection system that can detect very low concentrations of hydrogen (i.e., greater than 100 ppm) in metal matricies. Specifically in Titanium alloy jet engine blades. Entrapment and precipitation of hydrogen in metals is an undesirable phenomenon which occurs in many alloys of steel and titanium. In general, metals suffer a loss of mechanical properties after long exposures to hydrogen, especially at high temperatures and pressures, thereby becoming embrittled. Neutron radiography has been used as a nondestructive testing technique for many years. Neutrons, because of their unique interactions with materials, are especially useful in the detection of hydrogen. They have an extremely high interaction cross section for low atomic number nuclei (i.e., hydrogen). Thus hydrogen in a metal matrix can be visualized using neutrons. Traditional radiography is sensitive to the total attenuation integrated over the path of radiation through the material. Increased sensitivity and quantitative cross section resolution can be obtained using three-dimensional volumetric imaging techniques such as tomography. The solution used to solve the problem was to develop a neutron tomography system. The neutron source is the McClellan Nuclear Radiation Center's 1 MW TRIGA reactor. This paper describes the hardware used in the system as well as some of the preliminary results.

  17. Type approval of instruments and uncertainty calculations in the new EN-standards for the measurement of SO{sub 2}, NO{sub 2}, CO a nd O{sub 3} in ambient air

    Energy Technology Data Exchange (ETDEWEB)

    Sneek, E.J. [Quality of Environmental Data - QUENDA, Arnheim (Netherlands)

    2003-04-01

    The new European directives for ambient air quality are setting data quality objectives for the measurement of air polluting compounds. On request of the European Commission a working group of CEN 264 ''Air quality'' has developed standards with which these data quality objectives could be realised. Next to the measurement method itself, sampling procedure and quality assurance and quality control, these standards are dealing with type approval tests of instruments and uncertainty calculations of measurement data. Explanation is given about some of the considerations for the choices made by the working group in developing the requirements for performance characteristics, in developing the test methods to establish the values of these performance characteristics and in developing the uncertainty calculations. (orig.)

  18. Detector for advanced neutron capture experiments at LANSCE

    Energy Technology Data Exchange (ETDEWEB)

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

    2002-01-01

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

  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. Can three-neutron forces be constrained by empirical information on the neutron skin of 48Ca and 208Pb?

    CERN Document Server

    Sammarruca, Francesca

    2016-01-01

    We calculate the neutron matter equation of state from two-neutron forces up to ?fifth order of the chiral expansion and investigate the order-by-order convergence of the predictions. Based on these equations of state, we derive the binding energies and the neutron and proton density distributions in 208Pb and 48Ca, with particular attention to the neutron skins. Anticipating future experiments which will provide reliable information on the weak charge density in nuclei, we discuss the theoretical uncertainties and the possibility of constraining the size of three-neutron forces in neutron matter.

  1. Determination of neutron spectra using the programs GNSR and SPECTRIX

    CERN Document Server

    Weyrauch, M; Matzke, M

    2002-01-01

    We describe the capabilities and the application of two computer programs, which have been developed in order to facilitate common tasks in neutron spectrometry: GNSR (calculation of response matrices) and SPECTRIX (unfolding). Gas-filled Neutron Spectrometer Response calculates response functions and response matrices of various gas-filled neutron detectors. It can be configured to accommodate the appropriate gas-fillings and supports a number of different neutron beam configurations with a possibility to input calculated or measured neutron beam spectra. The program includes graphical capabilities as well as a context-sensitive help system. SPECTRIX implements several unfolding algorithms as well as support algorithms for unfolding and includes graphics capabilities and context-sensitive help. We apply both programs to a specific example: calculation of the response matrix of a sup 3 He detector and unfolding of the neutron spectrum of a thick accelerator target using the calculated response matrix.

  2. Complete Monte Carlo Simulation of Neutron Scattering Experiments

    Science.gov (United States)

    Drosg, M.

    2011-12-01

    In the far past, it was not possible to accurately correct for the finite geometry and the finite sample size of a neutron scattering set-up. The limited calculation power of the ancient computers as well as the lack of powerful Monte Carlo codes and the limitation in the data base available then prevented a complete simulation of the actual experiment. Using e.g. the Monte Carlo neutron transport code MCNPX [1], neutron scattering experiments can be simulated almost completely with a high degree of precision using a modern PC, which has a computing power that is ten thousand times that of a super computer of the early 1970s. Thus, (better) corrections can also be obtained easily for previous published data provided that these experiments are sufficiently well documented. Better knowledge of reference data (e.g. atomic mass, relativistic correction, and monitor cross sections) further contributes to data improvement. Elastic neutron scattering experiments from liquid samples of the helium isotopes performed around 1970 at LANL happen to be very well documented. Considering that the cryogenic targets are expensive and complicated, it is certainly worthwhile to improve these data by correcting them using this comparatively straightforward method. As two thirds of all differential scattering cross section data of 3He(n,n)3He are connected to the LANL data, it became necessary to correct the dependent data measured in Karlsruhe, Germany, as well. A thorough simulation of both the LANL experiments and the Karlsruhe experiment is presented, starting from the neutron production, followed by the interaction in the air, the interaction with the cryostat structure, and finally the scattering medium itself. In addition, scattering from the hydrogen reference sample was simulated. For the LANL data, the multiple scattering corrections are smaller by a factor of five at least, making this work relevant. Even more important are the corrections to the Karlsruhe data due to the

  3. From nuclear structure to neutron stars

    CERN Document Server

    Gandolfi, Stefano

    2013-01-01

    Recent progress in quantum Monte Carlo with modern nucleon-nucleon interactions have enabled the successful description of properties of light nuclei and neutron-rich 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.

  4. Tutorial on Neutron Physics in Dosimetry

    CERN Document Server

    Pomp, S

    2009-01-01

    Almost since the time of the discovery of the neutron more than 70 years ago, efforts have been made to understand the effects of neutron radiation on tissue and, eventually, to use neutrons for cancer treatment. In contrast to charged particle or photon radiations which directly lead to release of electrons, neutrons interact with the nucleus and induce emission of several different types of charged particles such as protons, alpha particles or heavier ions. Therefore, a fundamental understanding of the neutron-nucleus interaction is necessary for dose calculations and treatment planning with the needed accuracy. We will discuss the concepts of dose and kerma, neutron-nucleus interactions and have a brief look at nuclear data needs and experimental facilities and set-ups where such data are measured.

  5. 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...

  6. Strong CP violation and the neutron electric dipole form factor

    CERN Document Server

    Kuckei, J; Faessler, A; Gutsche, T; Kovalenko, S; Lyubovitskij, V E; Pumsa-ard, K; Dib, Claudio; Faessler, Amand; Gutsche, Th.; Kovalenko, Sergey

    2005-01-01

    We calculate the neutron electric dipole form factor induced by the CP violating theta-term of QCD, within a perturbative chiral quark model which includes pion and kaon clouds. On this basis we derive the neutron electric dipole moment and the electron-neutron Schiff moment. From the existing experimental upper limits on the neutron electric dipole moment we extract constraints on the theta-parameter and compare our results with other approaches.

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

    Science.gov (United States)

    Khaial, Anas M.

    Neutron radiography has the ability to differentiate between gas and liquid in two-phase flow due both to the density difference and the high neutron scattering probability of hydrogen. Previous studies have used dynamic neutron radiography -- in both real-time and high-speed -- for air-water, steam-water and gas-liquid metal two-phase flow measurements. Radiography with thermal neutrons is straightforward and efficient as thermal neutrons are easier to detect with relatively higher efficiency and can be easily extracted from nuclear reactor beam ports. The quality of images obtained using neutron radiography and the imaging speed depend on the neutron beam intensity at the imaging plane. A high quality neutron beam, with thermal neutron intensity greater than 3.0x 10 6 n/cm2-s and a collimation ratio greater than 100 at the imaging plane, is required for effective dynamic neutron radiography up to 2000 frames per second. The primary objectives of this work are: (1) to optimize a neutron radiography facility for dynamic neutron radiography applications and (2) to investigate a new technique for three-dimensional neutron radiography using information obtained from neutron scattering. In this work, neutron transport analysis and experimental validation of a dynamic neutron radiography facility is studied with consideration of real-time and high-speed neutron radiography requirements. A beam port based dynamic neutron radiography facility, for a target thermal neutron flux of 1.0x107 n/cm2-s, has been analyzed, constructed and experimentally verified at the McMaster Nuclear Reactor. The neutron source strength at the beam tube entrance is evaluated experimentally by measuring the thermal and fast neutron fluxes using copper activation flux-mapping technique. The development of different facility components, such as beam tube liner, gamma ray filter, beam shutter and biological shield, is achieved analytically using neutron attenuation and divergence theories. Monte

  8. Design and simulations of the neutron dump for the back-streaming white neutron beam at CSNS

    Science.gov (United States)

    Zhang, L. Y.; Jing, H. T.; Tang, J. Y.; Wang, X. Q.

    2016-10-01

    For nuclear data measurements with a white neutron source, to control the background at the detector is a key issue. The neutron dump which locates at the end of the white neutron beam line at CSNS has a very important impact to the neutron and gamma backgrounds in the endstation. A sophisticated neutron dump was designed to reduce the backgrounds to the level of about 10-8 relative to the neutron flux. In this paper, the method to suppress both neutron and gamma backgrounds near a white-spectrum neutron dump is introduced. The optimized geometry structure and materials of the dump are described, and the neutron and gamma space distributions have been calculated by using the FLUKA code for different operation settings which are defined by beam spots of Φ30 mm, Φ60 mm and 90 mm×90 mm, respectively.

  9. 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)

  10. Neutron absorbed dose in a pacemaker CMOS

    Energy Technology Data Exchange (ETDEWEB)

    Borja H, C. G.; Guzman G, K. A.; Valero L, C.; Banuelos F, A.; 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 (Mexico); Paredes G, L., E-mail: fermineutron@yahoo.com [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2012-06-15

    The neutron spectrum and the absorbed dose in a Complementary Metal Oxide Semiconductor (CMOS), has been estimated using Monte Carlo methods. Eventually a person with a pacemaker becomes an oncology patient that must be treated in a linear accelerator. Pacemaker has integrated circuits as CMOS that are sensitive to intense and pulsed radiation fields. Above 7 MV therapeutic beam is contaminated with photoneutrons that could damage the CMOS. Here, the neutron spectrum and the absorbed dose in a CMOS cell was calculated, also the spectra were calculated in two point-like detectors in the room. Neutron spectrum in the CMOS cell shows a small peak between 0.1 to 1 MeV and a larger peak in the thermal region, joined by epithermal neutrons, same features were observed in the point-like detectors. The absorbed dose in the CMOS was 1.522 x 10{sup -17} Gy per neutron emitted by the source. (Author)

  11. 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.

  12. Measurement of in-phantom neutron flux and gamma dose in Tehran research reactor boron neutron capture therapy beam line

    OpenAIRE

    Elham Bavarnegin; Alireza Sadremomtaz; Hossein Khalafi; Yaser Kasesaz

    2016-01-01

    Aim: Determination of in-phantom quality factors of Tehran research reactor (TRR) boron neutron capture therapy (BNCT) beam. Materials and Methods: 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. Results: Different dose components have been measured in a head phantom which has been designed an...

  13. 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)

  14. A system of materials composition and geometry arrangement for fast neutron beam thermalization: An MCNP study

    Science.gov (United States)

    Uhlář, Radim; Alexa, Petr; Pištora, Jaromír

    2013-03-01

    Compact deuterium-tritium neutron generators emit fast neutrons (14.2 MeV) that have to be thermalized for neutron activation analysis experiments. To maximize thermal neutron flux and minimize epithermal and fast neutron fluxes across the output surface of the neutron generator facility, Monte Carlo calculations (MCNP5; Los Alamos National Laboratory) for different moderator types and widths and collimator and reflector designs have been performed. A thin lead layer close to the neutron generator as neutron multiplier followed by polyethylene moderator and surrounded by a massive lead and nickel collimator and reflector was obtained as the optimum setup.

  15. On the crystal energy and structure of A{sub 2}Ti{sub n}O{sub 2n+1} (A=Li, Na, K) titanates by DFT calculations and neutron diffraction

    Energy Technology Data Exchange (ETDEWEB)

    Catti, Michele, E-mail: catti@mater.unimib.it [Dipartimento di Scienza dei Materiali, Università di Milano Bicocca, Via Cozzi 53, 20125 Milano (Italy); Pinus, Ilya [Dipartimento di Scienza dei Materiali, Università di Milano Bicocca, Via Cozzi 53, 20125 Milano (Italy); Scherillo, Antonella [ISIS Facility, CCLRC Rutherford Appleton Laboratory, Chilton, Didcot, Oxon, OX11 0QX (United Kingdom)

    2013-09-15

    First-principles quantum-mechanical calculations (CRYSTAL09 code, B3LYP functional) were performed on alkali titanates A{sub 2}Ti{sub n}O{sub 2n+1} with layered structure (n=3,4,6). Monoclinic structural types with unshifted (P2{sub 1}/m) and with shifted (C2/m) layers were considered. Crystal energies and full structural details were obtained for all Li, Na, and K phases. Neutron diffraction data were collected on powder samples of P2{sub 1}/m-Li{sub 2}Ti{sub 3}O{sub 7} (a=9.3146(3), b=3.7522(1), c=7.5447(3) Å, β=97.611(4)°) and C2/m-K{sub 2}Ti{sub 4}O{sub 9} (a=18.2578(8), b=3.79160(9), c=12.0242(4) Å, β=106.459(4)°) and their structures were Rietveld-refined. Computed energies show the P2{sub 1}/m arrangement as favoured over the C2/m one for n=3, and the opposite holds for n=6. In the n=4 case the P2{sub 1}/m configuration is predicted to be more stable for Li and Na, and the C2/m one for K titanates. Analysis of Li–O and K–O crystal-chemical environments from experiment and theory shows that the alkali atom bonding is stabilized/destabilized in the different phases consistently with the energy trend. - Graphical abstract: Display Omitted - Highlights: • The P2{sub 1}/m structure-type is found to be more stable for A{sub 2}Ti{sub 3}O{sub 7} layer titanates. • The C2/m structure-type is found to be more stable for A{sub 2}Ti{sub 6}O{sub 13} layer titanates. • Tetratitanates are predicted to prefer the P2{sub 1}/m (Li and Na) or C2/m (K) structure. • Li–O and K–O bond distances follow a trend consistent with computed phase energies.

  16. 超临界水冷堆中子能谱计算及安全性分析%Neutron spectrum calculation and safety analysis for supercritical water-cooled reactor

    Institute of Scientific and Technical Information of China (English)

    汤晓斌; 谢芹; 耿长冉; 陈达

    2012-01-01

    超临界水堆是国际第Ⅳ代核能系统论坛推荐的六种第Ⅳ代核电反应堆堆型之一,与现有的轻水堆相比,具有热效率高、系统结构简单、造价低等优点.建立了MCNP程序下的超临界水堆堆芯物理计算模型,解决了燃料组件几何结构过于复杂精细难以建模的技术难题;考虑了堆芯轴向冷却剂密度的不均匀分布,计算并分析各区域的中子能谱分布;对失水事故下的超临界水冷堆安全性进行了分析,研究了不同区域冷却剂丢失程度对反应性及有效增殖系数的影响,表明所设计堆型具有较高的安全性;分析处理失水事故的应对措施,验证了使用注入硼水措施处理超临界水冷堆失水事故的可行性.%The supercritical water reactor is one of the six reactors recommended by Generation IV International Forum, Compared with existing light water reactors, the supercritical water reactor has advantages of high thermal efficiency, simplified system structure and low cost. The physical model of the supercritical water reactor is established with MCNP program in this paper, which solves the problem of intricate geometry of fuel assembly. The change of coolant density along the axis is considered and the neutron spectrum distribution of different regions of the core is calculated. The safety in loss of coolant accident for the supercritical water reactor and the effect of missing coolant in different regions on the reactivity and effective multiplication factor analyzed. The results show the supercritical water reactor core has high security. The countermeasures of loss of coolant accident is studied and the effectiveness of boron water cooling is validated. The research not only provide important reference for the construction and security analysis of the supercritical water reactor, but also has great significance for the application and development of the supercritical water reactor.

  17. Neutrino Processes in Neutron Stars

    Science.gov (United States)

    Kolomeitsev, E. E.; Voskresensky, D. N.

    2010-10-01

    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 interaction effects can be

  18. Neutron transport simulation (selected topics)

    Science.gov (United States)

    Vaz, P.

    2009-10-01

    Neutron transport simulation is usually performed for criticality, power distribution, activation, scattering, dosimetry and shielding problems, among others. During the last fifteen years, innovative technological applications have been proposed (Accelerator Driven Systems, Energy Amplifiers, Spallation Neutron Sources, etc.), involving the utilization of intermediate energies (hundreds of MeV) and high-intensity (tens of mA) proton accelerators impinging in targets of high Z elements. Additionally, the use of protons, neutrons and light ions for medical applications (hadrontherapy) impose requirements on neutron dosimetry-related quantities (such as kerma factors) for biologically relevant materials, in the energy range starting at several tens of MeV. Shielding and activation related problems associated to the operation of high-energy proton accelerators, emerging space-related applications and aircrew dosimetry-related topics are also fields of intense activity requiring as accurate as possible medium- and high-energy neutron (and other hadrons) transport simulation. These applications impose specific requirements on cross-section data for structural materials, targets, actinides and biologically relevant materials. Emerging nuclear energy systems and next generation nuclear reactors also impose requirements on accurate neutron transport calculations and on cross-section data needs for structural materials, coolants and nuclear fuel materials, aiming at improved safety and detailed thermal-hydraulics and radiation damage studies. In this review paper, the state-of-the-art in the computational tools and methodologies available to perform neutron transport simulation is presented. Proton- and neutron-induced cross-section data needs and requirements are discussed. Hot topics are pinpointed, prospective views are provided and future trends identified.

  19. Neutron transport simulation (selected topics)

    Energy Technology Data Exchange (ETDEWEB)

    Vaz, P. [Instituto Tecnologico e Nuclear, Estrada Nacional 10, P-2686-953 Sacavem (Portugal)], E-mail: pedrovaz@itn.pt

    2009-10-15

    Neutron transport simulation is usually performed for criticality, power distribution, activation, scattering, dosimetry and shielding problems, among others. During the last fifteen years, innovative technological applications have been proposed (Accelerator Driven Systems, Energy Amplifiers, Spallation Neutron Sources, etc.), involving the utilization of intermediate energies (hundreds of MeV) and high-intensity (tens of mA) proton accelerators impinging in targets of high Z elements. Additionally, the use of protons, neutrons and light ions for medical applications (hadrontherapy) impose requirements on neutron dosimetry-related quantities (such as kerma factors) for biologically relevant materials, in the energy range starting at several tens of MeV. Shielding and activation related problems associated to the operation of high-energy proton accelerators, emerging space-related applications and aircrew dosimetry-related topics are also fields of intense activity requiring as accurate as possible medium- and high-energy neutron (and other hadrons) transport simulation. These applications impose specific requirements on cross-section data for structural materials, targets, actinides and biologically relevant materials. Emerging nuclear energy systems and next generation nuclear reactors also impose requirements on accurate neutron transport calculations and on cross-section data needs for structural materials, coolants and nuclear fuel materials, aiming at improved safety and detailed thermal-hydraulics and radiation damage studies. In this review paper, the state-of-the-art in the computational tools a