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. Comparison of experiment and VCS calculations for transmission of air-moderated neutron and gamma radiation through a shielded structure

    International Nuclear Information System (INIS)

    Comparison between calculated and measured shielding ratios is made for a polyethylene lined positioned steel box positioned 400 metres from a source of neutron and gamma radiation. The source was suspended outdoors at an altitude of 14 metres above the ground plane. VCS, a compilation of radiation transport codes including MORSE and DOT, was used to calculate the spectral data inside the lined box. The comparison shows fair-to-good agreement between experiment calculations for total kerma shielding ratios. (author)

  4. Air Force neutron dosimetry program

    International Nuclear Information System (INIS)

    Approximately 1000 Air Force personnel are monitored for neutron radiation resulting from various sources at more than thirty worldwide locations. Neutron radiation spanning several orders of magnitude in energy is encountered. The Air Force currently uses albedo thermoluminescent neutron dosimeters for personnel monitoring. The energy dependence of the albedo neutron dosimeter is a current problem and the development of site specific correction factors is ongoing. A summary of data on the energy dependence is presented as well as efforts to develop algorithms for the dosimeter. An overview of current Air Force neutron dosimetry users and needs is also presented

  5. 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.

  6. FURNACE calculations for JET neutron diagnostics

    International Nuclear Information System (INIS)

    Neutron transport calculations have been performed for the JET-torus, using the two-dimensional toroidal geometry transport code system FURNACE, to predict the response of the time integrated neutron yield monitors on the variation of the plasma conditions. Calculations have been performed for the full aperture D-shaped and circular plasmas, for DD-operation and for DT-operation. For the neutron source distribution a simple model was used based on plasma-plasma interaction. For the torus rotation symmetry around the main torus axis was assumed. Curves have been produced that give the radial plasma shift as function of the ratio of the foil activations measured. It is shown that these curves are sufficiently accurate for application in the DT-phase. For application in the DD-phase, however, the flux of neutrons backscattered from the massive torus needs to be calculated more accurately. (Auth.)

  7. A method for tokamak neutronics calculations

    International Nuclear Information System (INIS)

    This paper presents a new method for neutron transport calculation in tokamak fusion reactors. The computational procedure is based on the solution of the even-parity transport equation in a toroidal geometry. The angular neutron distribution is treated by even-parity spherical harmonic expansion, while the spatial dependence is approximated by using R-function finite elements that are defined for regions of arbitrary geometric shape. In order to test the method, calculation of a simplified tokamak model is carried out. The results are compared with the results from the literature and for the same order of accuracy a reduction of the number of spatial unknowns is shown. (author)

  8. Uncertainty analysis of neutron transport calculation

    International Nuclear Information System (INIS)

    A cross section sensitivity-uncertainty analysis code, SUSD was developed. The code calculates sensitivity coefficients for one and two-dimensional transport problems based on the first order perturbation theory. Variance and standard deviation of detector responses or design parameters can be obtained using cross section covariance matrix. The code is able to perform sensitivity-uncertainty analysis for secondary neutron angular distribution(SAD) and secondary neutron energy distribution(SED). Covariances of 6Li and 7Li neutron cross sections in JENDL-3PR1 were evaluated including SAD and SED. Covariances of Fe and Be were also evaluated. The uncertainty of tritium breeding ratio, fast neutron leakage flux and neutron heating was analysed on four types of blanket concepts for a commercial tokamak fusion reactor. The uncertainty of tritium breeding ratio was less than 6 percent. Contribution from SAD/SED uncertainties are significant for some parameters. Formulas to estimate the errors of numerical solution of the transport equation were derived based on the perturbation theory. This method enables us to deterministically estimate the numerical errors due to iterative solution, spacial discretization and Legendre polynomial expansion of transfer cross-sections. The calculational errors of the tritium breeding ratio and the fast neutron leakage flux of the fusion blankets were analysed. (author)

  9. Evaluated neutron data for thermal reactor calculations

    International Nuclear Information System (INIS)

    The paper describes a library of evaluated neutron data designed for thermal reactor calculations and other low energy neutron physics applications. The name of the library is KORT (Evaluated Thermal Reactor Constants). The following information is given in KORT: a general characterization of the nucleus (mass, energy of capture and fission reactions, parameters of radioactive decay); partial cross-sections for neutrons of thermal energy, and the number of secondary fission neutrons (estimated errors in the measurements of these quantities are indicated); coefficients defining the deviation of capture and fission cross-sections from the 1/v law in a Maxwellian spectrum; resonance capture and fission integrals and the estimated errors in these quantities (for nuclei with Z>=90); detailed energy dependence of the cross-sections in the 10-4-5 eV region at T=300 K

  10. 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 ...

  11. Neutronic parameters calculations of a CANDU reactor

    International Nuclear Information System (INIS)

    Neutronic calculations that reproduce in a simplified way some aspects of a CANDU reactor design were performed. Starting from some prefixed reactor parameters, cylindrical and uniform iron adjuster rods were designed. An appropriate refueling scheme was established, defininig in a 2 zones model their dimensions and exit burnups. The calculations have been done using the codes WIMS-D4 (cell), SNOD (reactivity device simulations) and PUMA (reactor). Comparing with similar calculations done with codes and models usually employed for CANDU design, it is concluded that the models and methods used are appropriate. (Author)

  12. Equivalent-spherical-shield neutron dose calculations

    International Nuclear Information System (INIS)

    Neutron doses through 162-cm-thick spherical shields were calculated to be 1090 and 448 mrem/h for regular and magnetite concrete, respectively. These results bracket the measured data, for reinforced regular concrete, of /approximately/600 mrem/h. The calculated fraction of the high-energy (>20 MeV) dose component also bracketed the experimental data. The measured and calculated doses were for a graphite beam stop bombarded with 100 nA of 800-MeV protons. 6 refs., 2 figs., 1 tab

  13. RA-0 reactor. New neutronic calculations

    International Nuclear Information System (INIS)

    An updating of the neutronic calculations performed at the RA-0 reactor, located at the Natural, Physical and Exact Sciences Faculty of Cordoba National University, are herein described. The techniques used for the calculation of a reactor like the RA-0 allows prediction in detail of the flux behaviour in the core's interior and in the reflector, which will be helpful for experiments design. In particular, the use of WIMSD4 code to make calculations on the reactor implies a novelty in the possible applications of this code to solve the problems that arise in practice. (Author)

  14. Comparison of calculated integral values using measured and calculated neutron spectra for fusion neutronics analyses

    International Nuclear Information System (INIS)

    The kerma heat production density, tritum production density, and dose in a lithium-fluoride pile with a deuterium-tritum neutron source were calculated with a data processing code, UFO, from the pulse height distribution of a miniature NE213 neutron spectrometer, and compared with the values calculated with a Monte Carlo code, MORSE-CV. Both the UFO and MORSE-CV values agreed with the statistical error (less than 6%) of the MORSE-CV calculations, except for the outer-most point in the pile. The MORSE-CV values were slightly smaller than the UFO values for almost all cases, and this tendency increased with increasing distance from the neutron source

  15. Reflector modelization for neutronic diffusion calculations

    International Nuclear Information System (INIS)

    For neutron diffusion calculations in nuclear reactors, it is always difficult to modelize the reflector. There exist different ways to describe the neutrons density in non fissile areas like the reflector, each of them presenting some advantages and difficulties. The first part of this work gives a new reflector problem formulation, replacing the complete diffusion calculation of the reflector by boundary conditions using non-local operators, the Poincare-Steklov ones. They can be used for the eigenvectors and eigenvalues diffusion problem stated on reactive core only. This theoretical treatment of non fissile areas leads, in second part, to a new interpretation of response matrix methods and Green functions methods. These two methods are in fact the main numerical techniques used to treat reflector as boundary conditions, and an other point of view is given by the Poincare-Steklov operators. Then some simple physical cases are studied, giving explicit expressions of the Poincare-Steklov operators, and allowing numerical estimates of the reflector behaviour in a whole core-reflector PWR calculation. Finally, numerical results of Green functions for boundary perturbations illustrate the physical non-locality of the boundary operators. (author). 16 refs., 2 annexes

  16. Description of the CAREM Reactor Neutronic Calculation Codes

    International Nuclear Information System (INIS)

    In this work is described the neutronic calculation line used to design the CAREM reactor.A description of the codes used and the interfaces between the different programs are presented.Both, the normal calculation line and the alternative or verification calculation line are included.The calculation line used to obtain the kinetics parameters (effective delayed-neutron fraction and prompt-neutron lifetime) is also included

  17. 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.

  18. 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.

  19. The neutron 'thunder' accompanying the extensive air shower

    OpenAIRE

    Erlykin, A. D.

    2007-01-01

    Simulations show that neutrons are the most abundant component among extensive air shower hadrons. However, multiple neutrons which appear with long delays in neutron monitors nearby the EAS core ('neutron thunder') are mostly not the neutrons of the shower, but have a secondary origin. The bulk of them is produced by high energy EAS hadrons hitting the monitors. The delays are due to the termalization and diffusion of neutrons in the moderator and reflector of the monitor accompanied by the ...

  20. 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...

  1. CONDOR: neutronic code for fuel elements calculation with rods

    International Nuclear Information System (INIS)

    CONDOR neutronic code is used for the calculation of fuel elements formed by fuel rods. The method employed to obtain the neutronic flux is that of collision probabilities in a multigroup scheme on two-dimensional geometry. This code utilizes new calculation algorithms and normalization of such collision probabilities. Burn-up calculations can be made before the alternative of applying variational methods for response flux calculations or those corresponding to collision normalization. (Author)

  2. 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

  3. Non interactive calculation of effective neutron multiplication factor by using two-group neutron albedo theory

    International Nuclear Information System (INIS)

    The effective neutron multiplication factor, Keff, explicitly appears under the neutron albedo theory. An albedo scheme can be used to determine Keff value without an iterative strategy. The albedo theory is illustrated by the endeavor of calculating Keff by using two-group neutron albedo method for spherical reflected cores. (author). 4 refs, 7 tabs

  4. Neutron activitation analysis of an air-dust sample using a high-flux 14 Mev neutron generator

    International Nuclear Information System (INIS)

    The 14 MeV neutron activation analysis technique is illustrated for multielement analysis of a Milanese air-dust sample. The neutron generator and electronic system, the efficiency and flux calibration, the γ-ray background, the sample preparation and the peak analysis used are described. After careful corrections of all possible interferences and error calculations, the results of 24 elemental concentrations are compared with those of other analytical techniques in the scope of an interlaboratory test. (orig.)

  5. 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.

  6. 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)

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

    OpenAIRE

    Klos, P.; Lynn, J. E.; 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 effectiv...

  8. Calculation methods for neutron radiography spatial resolution

    International Nuclear Information System (INIS)

    Spatial resolution is an important parameter for neutron radiography facility. In this paper, different methods to define the spatial resolution,such as point spread function (PSF), line spread function (LSF), edge spread function (ESF) and modulation transfer function (MTF), are analyzed and compared. MTF turns out to be the best, as it is derived from the linear system theory in a given frequency domain, and gives the maximum amount of useful information on system signal modulation. (authors)

  9. Calculation of corrections in the neutron activation analysis of oxygen in powdered and granulated materials

    International Nuclear Information System (INIS)

    Presented is a formula for the correction calculation at the analysis of oxygen in materials by the neutron activation method. A nomogram is plotted for the calculation of corrections taking into account the oxygen of capsule material and of air being in the internal volume of the capsule due to its incomplete filling. The accuracy of corrections according to nomogram is 2-3x10-4 mass %

  10. SRAC2006: A comprehensive neutronics calculation code system

    International Nuclear Information System (INIS)

    The SRAC is a code system applicable to neutronics analysis of a variety of reactor types. Since the publication of the second version of the users manual (JAERI-1302) in 1986 for the SRAC system, a number of additions and modifications to the functions and the library data have been made to establish a comprehensive neutronics code system. The current system includes major neutron data libraries (JENDL-3.3, JENDL-3.2, ENDF/B-VII, ENDF/B-VI.8, JEFF-3.1, JEF-2.2, etc.), and integrates five elementary codes for neutron transport and diffusion calculation; PIJ based on the collision probability method applicable to 16 kind of lattice models, SN transport codes ANISN(1D) and TWOTRN(2D), diffusion codes TUD(1D) and CITATION(multi-D). The system also includes an auxiliary code COREBN for multi-dimensional core burn-up calculation. (author)

  11. INDL/F-83. Evaluated neutron reaction data library for INTOR fusion neutronics calculations (1983 version)

    International Nuclear Information System (INIS)

    The INDL/F-83 data library is a computerized library of evaluated neutron reaction data which has been assembled from a variety of other evaluated data files and is intended for use in fusion neutronics calculations of the International Tokamak Reactor (INTOR) Project. These data are available on magnetic tape from the IAEA Nuclear Data Section. (author)

  12. Standard curves and formulae for neutron kinetics calculations

    International Nuclear Information System (INIS)

    The response of the neutron kinetic equations to a wide range of step and ramp additions of reactivity has been evaluated on the PACE 231R analogue computer for two fuels, U235 and Pu239, with a full range of neutron lifetimes. The results are presented in the form of standard curves which may be readily used to assess the 'zero-energy' performance of a reactor at the early stages of a reactor concept. Appendices contain the derivation of several useful expressions associated with neutron kinetics calculations and demonstrate the use of the curves to estimate reactor behaviour during shut-down following trip action. (author)

  13. Evaluation and calculation of neutron transactinide cross-sections

    International Nuclear Information System (INIS)

    This paper reviews the state of the art of nuclear theory and its application to the evaluation and calculation of neutron reaction cross sections of transactinium isotopes. In particular, the paper describes the current evaluation of the total files of neutron reaction data for 240Pu and 241Pu in the energy range between 10-5 eV and 15 MeV based on a thorough analysis of available experimental data and on the use of modern theoretical concepts, and the work in progress on the evaluation of the total neutron reaction data file for 242Pu and 241Am. (author)

  14. 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.

  15. Design calculation of a horizontal thermal neutronic beam for neutron radiography at the Syrian MNSR

    International Nuclear Information System (INIS)

    The computer code MCNP4C and the ENDF/B-V cross-section library were used to design calculation of a horizontal thermal beam for neutron radiography (NR) at Syrian MNSR and to evaluate the safety of the reactor after installation of the NR facility (NRF). Thermal, epithermal and fast neutron energy ranges were selected as 10.0 keV, respectively. To produce a good neutron beam in terms of intensity and quality, bismuth (Bi) and silicon (Si) were used as photon and neutron filters, respectively. The ratio of L/D of the NRF ranges between 90 and 125. The thermal neutron flux at the beam exit plane can be varied from 1.836 × 105 to 3.057 × 105 n/cm2 s. If such thermal neutron beam would be built into the Syrian MNSR, many scientific applications of the NR would be available. (author)

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

    International Nuclear Information System (INIS)

    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. - Highlights: • Original light water reactor spectra can be modified by material insertions. • Calculations of resulted neutron spectra have been done. • Comparison of the calcualted data to possible further utilization and research has been done

  17. Thermal and neutronic calculation for fast breeder reactor FBR

    International Nuclear Information System (INIS)

    This research included studying of thermal and neutronic calculation for fast breeder nuclear reactor, to putting the optimum design for this reactor. So a Soviet type (BN-350) was chosen, which has its core composed of two enrichment zones, and with blanket that contains depleted uranium. A group of thermal calculation programs was made by using personal computer, to obtain core and blanket reactor dimensions and volume fractions of reaction input material and number and dimensions of fuel rods which were used for neutron calculations. Several core and blanket enrichments were used to study neutron flux behaviour for two reactors different conditions. First when control rods exist in the core reactor and second when the rods are out of the core. Breeding ratio was also studied for different core and blanket enrichment. 30 tabs.; 24 figs.; 34 refs.; 3 apps

  18. Neutronic calculations for a fast assembly by using two-group neutron albedo theory

    International Nuclear Information System (INIS)

    Under Two-Group Neutron Albedo Theory, the effective neutron multiplication factor, Keff, explicitly appears and therefore it is possible to obtain an explicit form of variation of Keff. A generalization of the two-group albedo theory can be used if a more detailed energy spectrum treatment is required. The two-group neutron albedo theory is well illustrated by the endeavor of calculating the key parameters for a fast assembly. The results obtained from diffusion approach and albedo method calculations have had excellent concordance. (author)

  19. Fusion--fission neutronics calculations for the laser solenoid

    International Nuclear Information System (INIS)

    Neutron transport calculations are presented for several laser solenoid blanket configurations containing fast-fission lattices of uranium and thorium. The presence of a small-bore pulsed magnet and a small first-wall radius results in unique neutronics characteristics relative to other fusion concepts. Parametric calculations were completed to determine the effects of increasing the pulsed magnet thickness and of varying other key blanket parameters. Attractive fissile breeding rates could be achieved for blankets with a wide range of energy multiplication under the constraints of a tritium breeding ratio of about unity and a pulsed magnet thickness of about 3 cm

  20. 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.

  1. Accuracy preserving surrogate for neutron transport calculations

    International Nuclear Information System (INIS)

    Recent advances in reduced order modeling and exact-to-precision generalized perturbation theory are combined in a novel algorithm that constructs a surrogate model for the Boltzmann equation, commonly used in assembly calculations to functionalize the few-group cross-sections in terms of the various assembly types, depletion characteristics, and thermal-hydraulics conditions. First, the algorithm employs reduced order modeling to determine the dominant input parameters, aggregated in the so-called active subspace, using a random sample of first-order derivatives calculated using an adjoint model. Next, exact-to-precision generalized perturbation theory identifies an active subspace for the state solution (i.e., angular flux) and constructs a surrogate model that is parameterized over the active subspace of the input parameters. This approach is shown to significantly reduce computational time needed for the analysis of a large number of model variations, while meeting the user-defined accuracy requirements. Numerical experiments are employed to demonstrate the mechanics and application of the proposed approach to assembly calculations commonly used in reactor physics analysis. (author)

  2. Monte Carlo calculations of the neutron coincidence gate utilisation factor for passive neutron coincidence counting

    CERN Document Server

    Bourva, L C A

    1999-01-01

    The general purpose neutron-photon-electron Monte Carlo N-Particle code, MCNP sup T sup M , has been used to simulate the neutronic characteristics of the on-site laboratory passive neutron coincidence counter to be installed, under Euratom Safeguards Directorate supervision, at the Sellafield reprocessing plant in Cumbria, UK. This detector is part of a series of nondestructive assay instruments to be installed for the accurate determination of the plutonium content of nuclear materials. The present work focuses on one aspect of this task, namely, the accurate calculation of the coincidence gate utilisation factor. This parameter is an important term in the interpretative model used to analyse the passive neutron coincidence count data acquired using pulse train deconvolution electronics based on the shift register technique. It accounts for the limited proportion of neutrons detected within the time interval for which the electronics gate is open. The Monte Carlo code MCF, presented in this work, represents...

  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. New methods for neutron response calculations with MCNP

    International Nuclear Information System (INIS)

    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

  5. 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)

  6. Calculation of neutron flux in the presence of a source

    International Nuclear Information System (INIS)

    Neutron sources are introduced into the reactors to initiate the chain reaction. For safety reasons, we have to know the distribution and evolution of the flux throughout the startup phase. The flux is calculated iteratively but convergence of the process can slow down arbitrarily as we approach criticality. A calculation method is presented, with a convergence speed which does not depend on the negative reactivity when it is small. (author). 7 refs

  7. Parallel processing of neutron transport in fuel assembly calculation

    International Nuclear Information System (INIS)

    Group constants, which are used for reactor analyses by nodal method, are generated by fuel assembly calculations based on the neutron transport theory, since one or a quarter of the fuel assembly corresponds to a unit mesh in the current nodal calculation. The group constant calculation for a fuel assembly is performed through spectrum calculations, a two-dimensional fuel assembly calculation, and depletion calculations. The purpose of this study is to develop a parallel algorithm to be used in a parallel processor for the fuel assembly calculation and the depletion calculations of the group constant generation. A serial program, which solves the neutron integral transport equation using the transmission probability method and the linear depletion equation, was prepared and verified by a benchmark calculation. Small changes from the serial program was enough to parallelize the depletion calculation which has inherent parallel characteristics. In the fuel assembly calculation, however, efficient parallelization is not simple and easy because of the many coupling parameters in the calculation and data communications among CPU's. In this study, the group distribution method is introduced for the parallel processing of the fuel assembly calculation to minimize the data communications. The parallel processing was performed on Quadputer with 4 CPU's operating in NURAD Lab. at KAIST. Efficiencies of 54.3 % and 78.0 % were obtained in the fuel assembly calculation and depletion calculation, respectively, which lead to the overall speedup of about 2.5. As a result, it is concluded that the computing time consumed for the group constant generation can be easily reduced by parallel processing on the parallel computer with small size CPU's

  8. Two level calculation of assembly neutronic data libraries

    International Nuclear Information System (INIS)

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

  9. Coupled neutron and photon cross sections for transport calculations

    International Nuclear Information System (INIS)

    A compact set of multigroup cross sections and transfer tables for use in neutron and photon transport calculations was prepared from ENDF/B-IV using the NJOY processing system. The library includes prompt and steady-state coupled sets for neutrons and photons in FIDO format, prompt and steady-state fission spectra (chi vectors) for the fissionable isotopes, and a table of useful response functions including heating and gas production. These multigroup constants should be useful for a wide variety of problems where self-shielding is not important. 15 references

  10. Calculation of 14 MeV neutron transmission

    International Nuclear Information System (INIS)

    The possibility of using the 28 group constant system (28-GCS) for calculating the transport of neutrons with initial energy of 14 MeV in thermonuclear reactor blankets is studied. A blanket project suggested by the Oak Ridge National Laboratory is used as a test version to estimate applicability of the 28-GCS. Niobium is used in a blanket as a structural material. A mixture of lithium nuclides is used for tritium production. The results of blanket test calculation and the calculational results obtained using the 28-GCS from the UKNDL library are compared. The numerical 28-group calculation of blonket is carried out by means of the ROZ-6 and ROZ-9 codes but not by the Monte-Carlo method as compared with the test calculation. Time of the blanket calculation on the BESM-6 computer by means of the ROZ-9 code in 2P5 approximation using the 28-GCS amounts to 10 min. It is noted that to create effective codes for the numerical blanket calculation different calculational grids are necessary for different energy grups. The calculations carried out have shown the possibility of using the 28-group library of cross sections for the numerical solution of the neutron transport equation in estimating analysis of blankets

  11. Exact-to-precision generalized perturbation for neutron transport calculation

    International Nuclear Information System (INIS)

    This manuscript extends the exact-to-precision generalized perturbation theory (EPGPT), introduced previously, to neutron transport calculation whereby previous developments focused on neutron diffusion calculation only. The EPGPT collectively denotes new developments in generalized perturbation theory (GPT) that place premium on computational efficiency and defendable accuracy in order to render GPT a standard analysis tool in routine design and safety reactor calculations. EPGPT constructs a surrogate model with quantifiable accuracy which can replace the original neutron transport model for subsequent engineering analysis, e.g. functionalization of the homogenized few-group cross sections in terms of various core conditions, sensitivity analysis and uncertainty quantification. This is achieved by reducing the effective dimensionality of the state variable (i.e. neutron angular flux) by projection onto an active subspace. Confining the state variations to the active subspace allows one to construct a small number of what is referred to as the 'active' responses which are solely dependent on the physics model rather than on the responses of interest, the number of input parameters, or the number of points in the state phase space. (authors)

  12. Neutron cross section calculations for fission-product nuclei

    International Nuclear Information System (INIS)

    To satisfy nuclear data requirements for fission-product nuclei, Hauser-Feshbach statistical calculations with preequilibrium corrections for neutron-induced reactions on isotopes of Se, Kr, Sr, Zr, Mo, Sn, Xe, and Ba between 0.001 and 20 MeV. Spherical neutron optical parameters were determined by simultaneous fits to resonance data and total cross sections. Isospin coefficients appearing in the optical potentials were determined through analysis of the behavior of s- and p-wave strengths as a function of mass for a given Z. Gamma-ray strength functions, determined through fits to stable-isotope capture data, were used in the calculation of capture cross sections and gamma-ray competition to particle emission. The resulting (n,γ), (n,n'), (n,2n), and (n,3n) cross sections, the secondary neutron emission spectra, and angular distributions calculated for 19 fission products will be averaged to provide a resulting ENDF-type fission-product neutronics file. 11 references

  13. Neutron absorbed dose determination by calculations of recoil energy.

    Science.gov (United States)

    Wrobel, F; Benabdesselam, M; Iacconi, P; Lapraz, D

    2004-01-01

    The aim of this work is to calculate the absorbed dose to matter due to neutrons in the 5-150 MeV energy range. Materials involved in the calculations are Al2O3, CaSO4 and CaS, which may be used as dosemeters and have already been studied for their luminescent properties. The absorbed dose is assumed to be mainly due to the energy deposited by the recoils. Elastic reactions are treated with the ECIS code while for the non-elastic ones, a Monte Carlo code has been developed and allowed to follow the nucleus decay and to determine its characteristics (nature and energy). Finally, the calculations show that the absorbed dose is mainly due to non-elastic process and that above 20 MeV this dose decreases slightly with the neutron energy. PMID:15353750

  14. 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.

  15. Miniature neutron source reactor burnup calculations using IRBURN code system

    International Nuclear Information System (INIS)

    Highlights: ► Fuel consumption of Iranian MNSR during 15 years of operation has been investigated. ► Calculations have been performed by the IRBURN code. Precision and accuracy of the implemented model has been validated. ► Our study shows the consumption rate of MNSR is about 1%. - Abstract: Fuel consumption of Iranian miniature neutron source reactor (MNSR) during 15 years of operation has been investigated. Reactor core neutronic parameters such as flux and power distributions, control rod worth and effective multiplication factor at BOL and after 15 years of irradiation has been calculated. The Monte Carlo-based depletion code system IRBURN has been used for studying the reactor core neutronic parameters as well as the isotopic inventory of the fuel during burnup. The precision and accuracy of the implemented model has been verified via validation the results for neutronic parameters in the MNSR final safety analysis report. The results show that keff decreases from 1.0034 to 0.9897 and the total U-235 consumption in the core is about 13.669 g after 15 years of operational time. Finally, our studying shows the consumption rate of MNSR is about 1%.

  16. Comparison between measured and calculated neutron spectra in FCA assemblies

    International Nuclear Information System (INIS)

    The neutron spectra measured in FCA Assembly VI-2, VI-1 and V-2 are discussed, and are compared with the results by calculation. The data were obtained by measurements of proton-recoil counter and double scintillator methods. Calculations were made with cell-program SP-2000 and fine-group cross section library AGRI/2, and the spectra with 1950 groups and broadened 64 and 26 group were derived. The measured spectra in the energy range of 5 keV to 6 MeV were effectively compared with the calculational results, by using C/E values. There are large differences between the measured and the calculated spectra near the 430 keV oxygen and 29 keV iron resonances. The experimental and the calculated central fission rate ratios were also compared. (author)

  17. Calculation of 239Pu neutron inelastic cross sections

    International Nuclear Information System (INIS)

    We have calculated cross sections for neutron-induced reactions on 239Pu between 0.001 and 5 MeV, with particular emphasis on inelastic scattering. Coupled-channel and Hauser-Feshbach statistical models were used. Within the coupled-channel calculations we employed neutron optical parameters derived from simultaneous fits to total, elastic, inelastic, and resonance data. The resulting transmission coefficients were used in Hauser-Feshbach statistical calculations having a fission channel based on a double-humped barrier representation. Barrier parameters and transition state enhancements needed to reproduce well the (n,f) cross sections between 0.001 and 5 MeV were in general agreement with those from other published analyses. Calculated compound-nucleus and direct-reaction components for inelastic scattering were combined incoherently, and the resultant cross sections agreed well with the Bruyeres-le-Chatel measurements for scattering from levels occupying the ground state rotational band. Our results are in substantial disagreement with ENDF/B-V values for these levels. We are presently performing DWBA calculations to determine direct-reaction components for states occupying higher-lying vibrational bands

  18. The neutron 'thunder' accompanying the extensive air shower

    CERN Document Server

    Erlykin, A D

    2007-01-01

    Simulations show that neutrons are the most abundant component among extensive air shower hadrons. However, multiple neutrons which appear with long delays in neutron monitors nearby the EAS core ('neutron thunder') are mostly not the neutrons of the shower, but have a secondary origin. The bulk of them is produced by high energy EAS hadrons hitting the monitors. The delays are due to the termalization and diffusion of neutrons in the moderator and reflector of the monitor accompanied by the production of secondary gamma-quanta. This conclusion raises the important problem of the interaction of EAS with the ground, the stuff of the detectors and their environment since they have often hydrogen containing materials like polyethilene in neutron monitors. Such interaction can give an additional contribution to the signal in the EAS detectors. It can be particularly important for the signals from scintillator or water tank detectors at km-long distances from the EAS core where neutrons of the shower become the do...

  19. The neutron 'thunder' accompanying the extensive air shower

    Science.gov (United States)

    Erlykin, A. D.

    2007-03-01

    Simulations show that neutrons are the most abundant component among extensive air shower (EAS) hadrons. However, multiple neutrons which appear with long delays in neutron monitors nearby the EAS core (neutron thunder) are mostly not the neutrons of the shower, but have a secondary origin. The bulk of them is produced by high energy EAS hadrons hitting the monitors. The delays are due to the thermalization and diffusion of neutrons in the moderator and reflector of the monitor accompanied by the production of secondary gamma quanta. This conclusion raises the important problem of the interaction of EAS with the ground, the stuff of the detectors and their environment since they have often hydrogen-containing materials like polyethilene in neutron monitors. Such interaction can give an additional contribution to the signal in the EAS detectors. It can be particularly important for the signals from scintillator or water tank detectors at kilometre-long distances from the EAS core, where neutrons of the shower become the dominant component after a few microseconds behind the EAS front.

  20. Study on calculation methods for the effective delayed neutron fraction

    International Nuclear Information System (INIS)

    The effective delayed neutron fraction βeff is one of the important neutronic parameters from a view point of a reactor kinetics. Several Monte-Carlo-based methods to estimate βeff have been proposed to date. In order to quantify the accuracy of these methods, we study calculation methods for βeff by analyzing various fast neutron systems including the bare spherical systems (Godiva, Jezebel, Skidoo, Jezebel-240), the reflective spherical systems (Popsy, Topsy, Flattop-23), MASURCA-R2 and MASURCA-ZONA2, and FCA XIX-1, XIX-2 and XIX-3. These analyses are performed by using SLAROM-UF and CBG for the deterministic method and MVP-II for the Monte Carlo method. We calculate βeff with various definitions such as the fundamental value β0, the standard definition, Nauchi's definition and Meulekamp's definition, and compare these results with each other. Through the present study, we find the following: The largest difference among the standard definition of βeff , Nauchi's βeff and Meulekamp's βeff is approximately 10%. The fundamental value β0 is quite larger than the others in several cases. For all the cases, Meulekamp's βeff is always higher than Nauchi's βeff. This is because Nauchi's βeff considers the average neutron multiplicity value per fission which is large in the high energy range (1MeV-10MeV), while the definition of Meulekamp's βeff does not include this parameter. Furthermore, we evaluate the multi-generation effect on βeff values and demonstrate that this effect should be considered to obtain the standard definition values of βeff. (author)

  1. TRAWA, LWR Dynamic by Coupled Neutron Diffusion and Thermohydraulics Calculation

    International Nuclear Information System (INIS)

    1 - Description of problem or function: The purpose of the program is to study reactor dynamics in thermal water-cooled reactors. It treats the core as one or a few axially one-dimensional subregions. The two group neutron diffusion equations are solved simultaneously with the heat conduction equations and the two-phase hydraulic equations for one or more channels. Neither thermal nor hydraulic mixing appear between channels. Doppler, coolant density, coolant temperature, and soluble poison density feedbacks due to the thermo- hydraulics of the channels are described by using polynomial expansions for the group constants. The hydraulic circuit outside the reactor core consists of by-pass channels and risers with two- phase flow and of pump lines with incompressible flow. Various transients can be calculated by applying external disturbances. They can affect e.g. on movements of control rods, core inlet hydraulic conditions, system pressure or coefficients of neutronic shape function expansion between subregions. 2 - Method of solution: Nontrivial implicit methods are employed in the discretization of the equations to allow for sparse spatial mesh and flexible choice of time steps. The same spatial and temporal discretization is used for neutronics and thermohydraulics. 3 - Restrictions on the complexity of the problem: The dimensions of the program variable tables can easily be extended. Now the main dimensions are: 52 axial mesh points in core; 3 subregions; 10 axial regions with different fuel compositions; 7 radial mesh points in fuel rod; 6 delayed neutron groups; 6 coupled legs in pressure balance calculation; No flow reversals are allowed

  2. 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 (< 1 okta) the neutron detector recorded a sharp (count rate = 27 neutrons/min) and brief (< 1 min) increase in the count rate. In the days before and after this event, the neutron count rate has oscillated between 0 and 3 neutrons/min. Since the occurrence of this event is not related with spurious signals, malfunctioning equipment, oscillations in the mains voltage, etc. we are led to believe that the sharp increase was caused by a physical source such as a an extensive air shower that occurred over the detector.

  3. Cronos 2: a neutronic simulation software for reactor core calculations

    International Nuclear Information System (INIS)

    The CRONOS2 software is that part of the SAPHYR code system dedicated to neutronic core calculations. CRONOS2 is a powerful tool for reactor design, fuel management and safety studies. Its modular structure and great flexibility make CRONOS2 an unique simulation tool for research and development for a wide variety of reactor systems. CRONOS2 is a versatile tool that covers a large range of applications from very fast calculations used in training simulators to time and memory consuming reference calculations needed to understand complex physical phenomena. CRONOS2 has a procedure library named CPROC that allows the user to create its own application environment fitted to a specific industrial use. (authors)

  4. The conceptual calculation for the neutron beam device at Mark 1

    International Nuclear Information System (INIS)

    The thermal neutron beam device, epithermal neutron beam device and test duct experiment device are designed by using Monte Carlo method at 30 kW Mark 1(-1). The compared calculation for transverse cross section dimension, moderator, reflector and others of neutron filter device are studied in this paper. The three optimized neutron beams including thermal neutron beam, epithermal neutron beam and the beam for measuring blood boron density, whose neutron flux density per reactor power are rather high, are also introduced. The results show that the BNCT neutron beam can be designed by using 30kW -1 reactor. (author)

  5. Neutron matter with chiral EFT interactions: Perturbative and first QMC calculations

    OpenAIRE

    Tews, I.; Krüger, T.; Gezerlis, A.; Hebeler, K.; Schwenk, A.

    2013-01-01

    Neutron matter presents a unique system in chiral effective field theory (EFT), because all many-body forces among neutrons are predicted to next-to-next-to-next-to-leading order (N3LO). We discuss perturbative and first Quantum Monte Carlo (QMC) calculations of neutron matter with chiral EFT interactions and their astrophysical impact for the equation of state and neutron stars.

  6. Improvements in the model of neutron calculations for research reactors

    International Nuclear Information System (INIS)

    Within the research program in the field of neutron physics calculations being carried out in the Nuclear Engineering Division at the Centro Atomico Bariloche, the errors which due to some typical approximations appear in the final results are researched. For research MTR type reactors, two approximations, for high and low enrichment are investigated: the treatment of the geometry and the method of few-group cell cross-sections calculation, particularly in the resonance energy region. Commonly, the cell constants used for the entire reactor calculation are obtained making an homogenization of the full fuel elements, by one-dimensional calculations. An improvement is made that explicitly includes the fuel element frames in the core calculation geometry. Besides, a detailed treatment-in energy and space- is used to find the resonance few-group cross sections, and a comparison of the results with detailed and approximated calculations is made. The least number and the best mesh of energy groups needed for cell calculations is fixed too. (Author)

  7. Improvements in the model of neutron calculations for research reactors

    International Nuclear Information System (INIS)

    Within the research program in the field of neutron physics calculations being carried out in the Nuclear Engineering Division at the Centro Atomico Bariloche, the errors which due to some typical approximations appear in the final results, are being researched. For research MTR type reactors, two approximations, for high and low enrichment are investigated: the treatment of the geometry and the method of few-group cell cross-sections calculation, particularly in the resonance energy region. Commonly, the cell constants used for the entire reactor calculation are obtained making an homogenization of the full fuel elements by means of one-dimensional calculations. An improvement is made that explicitly includes the fuel element frames in the core calculation geometry. Besides, a detailed treatment-in energy and space- is used to find the resonance few-group cross sections, and a comparison of the results with detailed and approximated calculations is made. The least number and the best mesh of energy groups needed for cell calculations is fixed too. (Author)

  8. 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/.

  9. Weisskopf-Ewing calculations: neutron-induced reactions

    International Nuclear Information System (INIS)

    The cross sections of several neutron-induced reactions on 55Mn, sup(54,56)Fe, 59Co, sup(58,60)Ni and sup(63,65)Cu are calculated for energies below 20 MeV using the Weisskopf-Ewing theory and compared with experimental data. The total (n,p) and (n, α) cross sections are generally well fitted, especially when they are dominant channels. At the higher energies the (n,p) cross sections have important contributions from pre-equilibrium processes, and these are fitted using the theory of Feshbach, Kerman and Koonin. (author)

  10. ESTIMATING THE UNCERTAINTY IN REACTIVITY ACCIDENT NEUTRONIC CALCULATIONS

    Energy Technology Data Exchange (ETDEWEB)

    DIAMOND,D.J.; YANG,C.Y.; ARONSON,A.L.

    1998-10-26

    A study of the uncertainty in calculations of the rod ejection accident in a pressurized water reactor is being carried out for the US Nuclear Regulatory Commission. This paper is a progress report on that study. Results are presented for the sensitivity of core energy deposition to the key parameters: ejected rod worth, delayed neutron fraction, Doppler reactivity coefficient, and fuel specific heat. These results can be used in the future to estimate the uncertainty in local fuel enthalpy given some assumptions about the uncertainty in the key parameters. This study is also concerned with the effect of the intra-assembly representation in calculations. The issue is the error that might be present if assembly-average power is calculated, and pin peaking factors from a static calculation are then used to determine local fuel enthalpy. This is being studied with the help of a collaborative effort with Russian and French analysts who are using codes with different intra-assembly representations. The US code being used is PARCS which calculates power on an assembly-average basis. The Russian code being used is BARS which calculates power for individual fuel pins using a heterogeneous representation based on a Green's Function method.

  11. Estimating the uncertainty in reactivity accident neutronic calculations

    Energy Technology Data Exchange (ETDEWEB)

    Diamond, D.J.; Yang, C.Y.; Aronson, A.L.

    1998-12-31

    A study of the uncertainty in calculations of the rod ejection accident in a pressurized water reactor is being carried out for the US Nuclear Regulatory Commission. This paper is a progress report on that study. Results are presented for the sensitivity of core energy deposition to the key parameters: ejected rod worth, delayed neutron fraction, Doppler reactivity coefficient, and fuel specific heat. These results can be used in the future to estimate the uncertainty in local fuel enthalpy given some assumptions about the uncertainty in the key parameters. This study is also concerned with the effect of the intra-assembly representation in calculations. The issue is the error that might be present if assembly-average power is calculated, and pin peaking factors from a static calculation are then used to determine local fuel enthalpy. This is being studied with the help of a collaborative effort with Russian and French analysts who are using codes with different intra-assembly representations. The US code being used is PARCS which calculates power on an assembly-average basis. The Russian code being used is BARS which calculates power for individual fuel pins using a heterogeneous representation based on a Green`s Function method.

  12. Aerosol and air pollution study by neutron activation analysis

    International Nuclear Information System (INIS)

    Thermal neutron activation analysis technique was used in air pollution and aerosol elemental content and size distribution investigations. Air pollution samples were collected on Whatman 41 paper filters which were activated along with known quantities of standards in a flux of approximately 1013 nxcm-2xs-1. The activity of the samples was measured with a 40 cm3 Ge(Li) detector and analyzed with the computer program JANE, which identified the isotopes and found their quantities by normalization with the standard measurement results. Correlation between the various elements, in particular those belonging to dust from the desert and those considered typical urban air pollution, is investigated. (author)

  13. Calculation and analysis of the neutron radiography spatial resolution

    International Nuclear Information System (INIS)

    Background: Spatial resolution is the key parameter for neutron radiography facility. A model of the integrated system resolution is important when designing or using a system to ensure that the realistic resolution goals can be established and achieved. Purpose: For this resolution modeling analysis we focused on the effects of the geometry effects of L/D, the optical diffusion response of the scintillator and the sampling at the sensor (CCD or CMOS camera) and a formula was derived indicating their functional relationship. Methods: This resolution modeling analysis has been down by theoretic calculations. Then this integrated system resolution model was used as an empirical methodology to verify and optimize the performance of the detection system for real-time neutron radiography at China Advance Research Reactor. Results: The special resolutions at very collimation conditions have been calculation by using this method. And three of important parameters of this resolution model have been discussed to optimize the system performance. Conclusion: These resolution analysis concepts and methods will benefit both the design and the characterization of radiography systems. (authors)

  14. 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.)

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

    It is shown that the combination of 3D neutron transport calculations and the results from activation foil measurements at a limited number of locations in a materials testing irradiation experiment can provide information at any position in the experiment for detailed neutron dosimetry and damage analysis. 4 refs

  17. CANISTER HANDLING FACILITY - VENTILATION AIR CALCULATION

    International Nuclear Information System (INIS)

    The purpose of this analysis is to establish the preliminary Ventilation Confinement Zone for the Canister Handling Facility (CHF). The results of this document will be used to determine the air quantities for each VCZ that will eventually be reflected in the development of the Ventilation Flow Diagrams. The analyses contained in this document are developed by D and E/Mechanical HVAC and are intended solely for the use of the D and E/Mechanical HVAC in its work regarding Confinement Zoning Analysis for the Canister Handling Facility. Yucca Mountain Project personnel from D and E/Mechanical HVAC should be consulted before use of the analyses for purposes other than those stated herein or used by individuals other than authorized personnel in D and E/Mechanical HVAC

  18. Solution of thermal neutron diffusion equation for the two-component system by perturbation calculation

    International Nuclear Information System (INIS)

    A method of solving the diffusion equation for the th ermal neutron flux in a heterogeneous medium is presented. Perturbation calculation is successfully applied for the cylindrical concentric system after testing this method for the spherical concentric geometry analytically solved by Czubek (1981). The method permits to calculate the t hermal neutron decay constant and the space distribution of the thermal neutron flux in a heterogeneous geom etry. The condition of the constant value of the neutron flux in the inner part of the system has to be m et. This method has an application in the measurement of the thermal neutron absorption cross section, presented by Czubek (1981). (author)

  19. Neutronic calculations for JET. Performed with the FURNACE2 program. (Final report JET contract JEO/9004)

    International Nuclear Information System (INIS)

    Neutron-transport calculations with the FURNACE(2) program system, in support of the Neutron Diagnostic Group at JET, have been performed since 1980, i.e. since the construction phase of JET. FURNACE(2) is a ray-tracing/multiple-reflection transport program system for toroidal geometries, that orginally was developed for blanket neutronics studies and which then was improved and extended for application to the neutron-diagnostics at JET. (orig./WL)

  20. 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.)

  1. 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.

  2. Implementation and training methodology of subcritical reactors neutronic calculations triggered by external neutron source and applications

    International Nuclear Information System (INIS)

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

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

    Science.gov (United States)

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

    1954-01-01

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

  4. 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)

  5. 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)

  6. How Accurately Can We Calculate Neutrons Slowing Down In Water ?

    Energy Technology Data Exchange (ETDEWEB)

    Cullen, D E; Blomquist, R; Greene, M; Lent, E; MacFarlane, R; McKinley, S; Plechaty, E; Sublet, J C

    2006-03-30

    We have compared the results produced by a variety of currently available Monte Carlo neutron transport codes for the relatively simple problem of a fast source of neutrons slowing down and thermalizing in water. Initial comparisons showed rather large differences in the calculated flux; up to 80% differences. By working together we iterated to improve the results by: (1) insuring that all codes were using the same data, (2) improving the models used by the codes, and (3) correcting errors in the codes; no code is perfect. Even after a number of iterations we still found differences, demonstrating that our Monte Carlo and supporting codes are far from perfect; in particularly we found that the often overlooked nuclear data processing codes can be the weakest link in our systems of codes. The results presented here represent the today's state-of-the-art, in the sense that all of the Monte Carlo codes are modern, widely available and used codes. They all use the most up-to-date nuclear data, and the results are very recent, weeks or at most a few months old; these are the results that current users of these codes should expect to obtain from them. As such, the accuracy and limitations of the codes presented here should serve as guidelines to code users in interpreting their results for similar problems. We avoid crystal ball gazing, in the sense that we limit the scope of this report to what is available to code users today, and we avoid predicting future improvements that may or may not actual come to pass. An exception that we make is in presenting results for an improved thermal scattering model currently being testing using advanced versions of NJOY and MCNP that are not currently available to users, but are planned for release in the not too distant future. The other exception is to show comparisons between experimentally measured water cross sections and preliminary ENDF/B-VII thermal scattering law, S({alpha},{beta}) data; although these data are strictly

  7. 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.

  8. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-07-01

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

  10. Calculation of neutron spectra on typical irradiation location of the CFBR-II reactor

    International Nuclear Information System (INIS)

    Neutron energy spectra were simulated by the MCNP code. The neutron energy spectra and corresponding average energy of off-coupling box, irradiation channel and outer surface of the off-coupling cover were calculated. The results indicate that about 90% neutrons are in the energy range of 0.05-3 MeV. The average neutron energy of off-coupling box and irradiation channel present 'S' shape along distance, and space asymmetry must be considered. The average neutron energy above off-coupling cover's 45 degree woof fluctuates slightly and it is an appropriate irradiation area. (authors)

  11. Comparison of integral values for measured and calculated fast neutron spectra in lithium fluoride piles

    International Nuclear Information System (INIS)

    The tritium production density, kerma heat production density, dose and certain integral values of scalar neutron spectra in bare and graphite-reflected lithium-fluoride piles irradiated with D-T neutrons were evaluated from the pulse height distribution of a miniature NE213 neutron spectrometer with UFO data processing code, and compared with the values calculated with MORSE-CV Monte Carlo code. (author). 8 refs.; 1 fig.; 2 tabs

  12. Improvement of neutron dose calculation algorithm using panasonic UD-809P type albedo TLD

    International Nuclear Information System (INIS)

    Panasonic UD-809P type albedo TLD mounted on a water phantom were used to measure neutron personal dose equivalent in a Korean nuclear power plant. From the measured TL readings, personal dose equivalents from thermal, epithermal and fast neutrons were evaluated by using a method adopted in a neutron dose calculation algorithm for Panasonic UD-809P type albedo TLD, which was recommended in a Panasonic TLD System User's Manual. The results showed that personal dose equivalent for fast neutrons could not be adequately evaluated in a field with high thermal neutron fraction. This seems to be related to the incomplete incidence of albedo thermal neutrons to the TLD. In order to calculate the personal dose equivalent from fast neutrons in the field condition to be encountered in a nuclear power plant, new method for the neutron dose calculation algorithm were suggested. For a known energy spectrum, it is very easy and simple to use this method for the evaluation of neutron personal dose equivalent

  13. CAREM 25: actual status of the core neutronic design. Calculation line

    International Nuclear Information System (INIS)

    This work follows the one titled 'Criteria for the CAREM 25 reactor core design. Neutronic aspects' presented at this congress, gives in detail the typical values regarding the core defined at this point. Besides, the neutronic calculation line used for the CAREM 25 reactor design is presented. (Author)

  14. 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

  15. 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.

  16. A general dimensional neutron diffusion calculation code: ADC

    International Nuclear Information System (INIS)

    A FORTRAN computer program ADC is developed for the FACOM 230-75 computer to be capable of solving eigenvalue problems of neutron diffusion equation in one, two and three spatial dimensions. The available coordinate systems are orthogonal (X), (X,Y), (X,Y,Z) and cylindrical (R,Z), (R,THETA), (R,THETA,Z). The outer boundary condition for the neutron flux can be chosen to be symmetric, zero flux or log-derivative condition. The present program can be used also for obtaining the adjoint flux. (author)

  17. Systems for neutronic, thermohydraulic and shielding calculation in personal computers

    International Nuclear Information System (INIS)

    The MTR-PC (Materials Testing Reactors-Personal Computers) system has been developed by the Nuclear Engineering Division of INVAP S.E. with the aim of providing working conditions integrated with personal computers for design and neutronic, thermohydraulic and shielding analysis for reactors employing plate type fuel. (Author)

  18. 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.

  19. 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 (shower that occurred over the detector.

  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. The adaptation of methods in multilayer optics for the calculation of specular neutron reflection

    International Nuclear Information System (INIS)

    The adaptation of standard methods in multilayer optics to the calculation of specular neutron reflection is described. Their application is illustrated with examples which include a glass optical flat and a deuterated Langmuir-Blodgett film. (author)

  2. Calculation of production cross sections of γ-rays from thermal-neutron captures

    Institute of Scientific and Technical Information of China (English)

    ZHOU Chun-Mei; WU Zhen-Dong

    2004-01-01

    The calculation methods of production cross sections of γ-rays for thermal-neutron captures are briefly presented. The check of intensity balance is made. The examples are given to illustrate its application.

  3. Calculation of the neutron importance and weighted neutron generation time using MCNIC method in accelerator driven subcritical reactors

    International Nuclear Information System (INIS)

    Highlights: • All reactor kinetic parameters are importance weighted quantities. • MCNIC method has been developed for calculating neutron importance in ADSRs. • Mean generation time has been calculated in spallation driven systems. -- Abstract: The difference between non-weighted neutron generation time (Λ) and the weighted one (Λ†) can be quite significant depending on the type of the system. In the present work, we will focus on developing MCNIC method for calculation of the neutron importance (Φ†) and importance weighted neutron generation time (Λ†) in accelerator driven systems (ADS). Two hypothetic bare and graphite reflected spallation source driven system have been considered as illustrative examples for this means. The results of this method have been compared with those obtained by MCNPX code. According to the results, the relative difference between Λ and Λ† is within 36% and 24,840% in bare and reflected illustrative examples respectively. The difference is quite significant in reflected systems and increases with reflector thickness. In Conclusion, this method may be used for better estimation of kinetic parameters rather than the MCNPX code because of using neutron importance function

  4. MCNP calculations of neutron emission anisotropy caused by the GIT-12 hardware

    Directory of Open Access Journals (Sweden)

    Šíla Ondřej

    2015-06-01

    Full Text Available The MCNP6 and MCNPX calculations for the GIT-12 device in Tomsk were performed to determine the influence of the gas-puff hardware on the neutron emission anisotropy and the neutron scattering rate. A monoenergetic 2.45 MeV neutron source and F1 and F6 tallies were declared in the simulation input. A comparison between MCNP results and the measured data was made. Differences between MCNPX and MCNP6 output data were investigated. In the experiment, two nTOF scintillation detectors with the Bicron BC-408 scintillator were used to measure the neutron waveform. Four bubble BD-PND detectors were used to estimate the amount of neutrons in different places around the neutron source.

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

    International Nuclear Information System (INIS)

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

  6. Calculation of fine neutron spectrum in irradiation holes in fuel region of JRR-3M

    International Nuclear Information System (INIS)

    The authors have a plan to evaluate TRU neutron cross sections based on the activation experiments by using JRR-3M. Fine neutron spectrum expressed by 107 energy group structure at irradiation holes in fuel region of JRR-3M core, which was utilized to analyze experimental data, was calculated by 2 step calculation. The first step is the whole core calculation taking account of burnup history and control rod pattern, and the second step is the irradiation hole calculation without any homogenization of irradiation hole components by taking into account of the neutron spectrum of surrounding region. Fine neutron spectra calculated by 2 step calculation were compared with the experimental results on reaction rate, both agreed within several percents relatively. In the comparison of absolute values, however, the maximum difference was up to 30 percents in the vicinity of control rods. This originates from the neutron transport effect around control rods. An improvement for the treatment of neutron transport effect is needed to get higher accuracy. (author)

  7. 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.

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

    International Nuclear Information System (INIS)

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

  9. Calculation of neutron and gamma ray energy spectra for fusion reactor shield design: comparison with experiment

    International Nuclear Information System (INIS)

    Integral experiments that measure the transport of approx. 14 MeV D-T neutrons through laminated slabs of proposed fusion reactor shield materials have been carried out. Measured and calculated neutron and gamma ray energy spectra are compared as a function of the thickness and composition of stainless steel type 304, borated polyethylene, and Hevimet (a tungsten alloy), and as a function of detector position behind these materials. The measured data were obtained using a NE-213 liquid scintillator using pulse-shape discrimination methods to resolve neutron and gamma ray pulse height data and spectral unfolding methods to convert these data to energy spectra. The calculated data were obtained using two-dimensional discrete ordinates radiation transport methods in a complex calculational network that takes into account the energy-angle dependence of the D-T neutrons and the nonphysical anomalies of the S/sub n/ method

  10. 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.

  11. 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

  12. 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.

  13. 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.

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

    DEFF Research Database (Denmark)

    Tarnow, Viggo

    1997-01-01

    The acoustic attenuation of acoustic fiber materials is mainly determined by the dynamic resistivity to an oscillating air flow. The dynamic resistance is calculated for a model with geometry close to the geometry of real fibre material. The model constists of parallel cylinders placed randomly.......The second procedure is an extension to oscillating air flow of the Brinkman self-consistent procedure for dc flow. The procedures are valid for volume concentrations of cylinders less than 0.1. The calculations show that for the density of fibers of interest for acoustic fibre materials the simple self......-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....

  15. Coupled hydro-neutronic calculations for fast burst reactor accidents

    International Nuclear Information System (INIS)

    Methods are described for determining the fully coupled neutronic/hydrodynamic response of fast burst reactors (FBR) under disruptive accident conditions. Two code systems, PAD (1 -D Lagrangian) and NIKE-PAGOSA (3-D Eulerian) were used to accomplish this. This is in contrast to the typical methodology that computes these responses by either single point kinetics or in a decoupled manner. This methodology is enabled by the use of modem supercomputers (CM-200). Two examples of this capability are presented: an unreflected metal fast burst assembly, and a reflected fast burst assembly typical of the Skua or SPR-III class of fast burst reactor

  16. 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.

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

    International Nuclear Information System (INIS)

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

  18. 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.

  19. Calculation of the decay power of fission products considering neutron capture transformation

    International Nuclear Information System (INIS)

    The decay power of fission products has been calculated taking into consideration the neutron capture transformation of each nuclide and its beta decay. The nuclear data library contains 1114 nuclides of which 144 are stable. Neutron capture transformation is considered for 59 nuclides, 31 of which are stable. The atom number of each nuclide is calculated analytically with code DCHAIN. The effect of neutron capture transformation in the decay power of fission products was examined by varying the neutron spectrum, neutron flux, fissioning nuclide, and irradiation and cooling time. From the results obtained the following were revealed: The effect of neutron capture increases with neutron flux and irradiation time, and it becomes salient beyond 105 sec in cooling time. It is small for less than the 104 sec which is important in the design of ECCS (emergency core cooling system) of a light-water reactor. In this region the decay power changes are small, less than 0.2%, by the neutron capture for the thermal fission of 235U irradiated for one year to thermal neutron flux 3 x 1013 n/cm2/sec. The effect of neutron capture has peaks around cooling time 106 sec and 108 sec; it is negligible beyond 109 sec. The changes in decay power are 2.4%, 10.5% and 0.2% at cooling time 106 sec, 108 sec and 109 sec, respectively, in the above irradiation. Around 106 sec, the change in decay power is mainly from the contributions of 134Cs (17%), sup(148m)Pm(60%) and 148Pm(14%). Around 108 sec 134Cs(98%) alone contributes to the change in decay power. (author)

  20. Ab initio calculations versus polarized neutron diffraction for the spin density of free radicals

    International Nuclear Information System (INIS)

    The determination of the magnetization distribution using polarized neutron diffraction has played a key role during the last twenty years in the field of molecular magnetism. This distribution can also be obtained by first principle ab initio calculations. Such calculations always rely on approximations and the question that arises is to know whether the obtained results are reliable enough to represent accurately the properties of these molecules. The comparison between polarized neutron experimental results and ab initio calculations has turned to provide stringent tests for these methods. In the resent article a comparison between experimental and theoretical results is made and is illustrated by examples based on magnetic free radicals. (author)

  1. Dose calculation from a D-D-reaction-based BSA for boron neutron capture synovectomy

    International Nuclear Information System (INIS)

    Monte Carlo simulations were carried out to calculate dose in a knee phantom from a D-D-reaction-based Beam Shaping Assembly (BSA) for Boron Neutron Capture Synovectomy (BNCS). The BSA consists of a D(d,n)-reaction-based neutron source enclosed inside a polyethylene moderator and graphite reflector. The polyethylene moderator and graphite reflector sizes were optimized to deliver the highest ratio of thermal to fast neutron yield at the knee phantom. Then neutron dose was calculated at various depths in a knee phantom loaded with boron and therapeutic ratios of synovium dose/skin dose and synovium dose/bone dose were determined. Normalized to same boron loading in synovium, the values of the therapeutic ratios obtained in the present study are 12-30 times higher than the published values.

  2. Dose calculation from a D-D-reaction-based BSA for boron neutron capture synovectomy

    Energy Technology Data Exchange (ETDEWEB)

    Abdalla, Khalid [Department of Physics, Hail University, Hail (Saudi Arabia)], E-mail: khalidafnan@uoh.edu.sa; Naqvi, A.A. [Department of Physics, King Fahd University of Petroleum and Minerals and Center for Applied Physical Sciences, Box No. 1815, Dhahran 31261 (Saudi Arabia)], E-mail: aanaqvi@kfupm.edu.sa; Maalej, N.; Elshahat, B. [Department of Physics, King Fahd University of Petroleum and Minerals and Center for Applied Physical Sciences, Box No. 1815, Dhahran 31261 (Saudi Arabia)

    2010-04-15

    Monte Carlo simulations were carried out to calculate dose in a knee phantom from a D-D-reaction-based Beam Shaping Assembly (BSA) for Boron Neutron Capture Synovectomy (BNCS). The BSA consists of a D(d,n)-reaction-based neutron source enclosed inside a polyethylene moderator and graphite reflector. The polyethylene moderator and graphite reflector sizes were optimized to deliver the highest ratio of thermal to fast neutron yield at the knee phantom. Then neutron dose was calculated at various depths in a knee phantom loaded with boron and therapeutic ratios of synovium dose/skin dose and synovium dose/bone dose were determined. Normalized to same boron loading in synovium, the values of the therapeutic ratios obtained in the present study are 12-30 times higher than the published values.

  3. 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)

  4. Demonstration of core neutronic calculation for research and training reactors via SCALE4.4

    International Nuclear Information System (INIS)

    In this work, full core modelling is performed to improve neutronic analyses capability for nuclear research reactors using SCALE4.4 system. KENOV.a module of SCALE4.4 system is utilized for full core neutronic analysis. The ORIGEN-S module is also coupled with the KENOV.a module to perform burnup dependent core analyses. Results of control rod worths for 1st cycle of Cekmece TR-2 research reactor are presented. In particular, coupling of KENOV.a and ORIGEN-S modules of SCALE4.4 is discussed. The preliminary results of 2-D burnup dependent neutronic calculations are also given. (author)

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

    International Nuclear Information System (INIS)

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

  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)

    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.

  7. 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.

  8. 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.

  9. 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.

  10. Calculated neutron KERMA factors based on the LLNL ENDL data file. Volume 27

    International Nuclear Information System (INIS)

    Neutron KERMA factors calculated from the LLNL ENDL data file are tabulated for 15 composite materials and for the isotopes or elements in the ENDL file from Z = 1 to Z = 29. The incident neutron energies range from 1.882 x 10-5 to 20. MeV for the composite materials and from 1.30 x 10-9 to 20. MeV for the isotopes and elements

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

    International Nuclear Information System (INIS)

    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/cm2 resulting in about 9.0 dpa in type 316 stainless steel

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

    International Nuclear Information System (INIS)

    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/cm2 resulting in about 9.0 dpa in type 316 stainless steel

  13. Neutronic calculations in heavy water moderated multiplying media using GGC-3 library nuclear data

    International Nuclear Information System (INIS)

    Differences in obtaining transference matrices between GGC-3 code and the system to produce multigroup cross sections using GGC-3 library, recently implemented at the Neutrons and Reactors Division, have been analized. Neutronic calculations in multiplicative systems containing heavy water have been made using both methods. From the obtained results, it is concluded that the new method is more appropriate to deal with systems including moderators other than light water. (author)

  14. Monte Carlo perturbation theory in neutron transport calculations

    International Nuclear Information System (INIS)

    The need to obtain sensitivities in complicated geometrical configurations has resulted in the development of Monte Carlo sensitivity estimation. A new method has been developed to calculate energy-dependent sensitivities of any number of responses in a single Monte Carlo calculation with a very small time penalty. This estimation typically increases the tracking time per source particle by about 30%. The method of estimation is explained. Sensitivities obtained are compared with those calculated by discrete ordinates methods. Further theoretical developments, such as second-order perturbation theory and application to k/sub eff/ calculations, are discussed. The application of the method to uncertainty analysis and to the analysis of benchmark experiments is illustrated. 5 figures

  15. 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.

  16. Development of an effective delayed neutron fraction calculation code, BETA-K

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Taek Kyum; Song, Hoon; Kim, Young Il; Kim, Young In; Kim, Young Jin [Korea Atomic Energy Research Institute, Taejon (Korea)

    1998-08-01

    BETA-K, an effective delayed neutron fraction calculation code consistent with Nodal Expansion Method (NEM), has been developed. By using relevant output files of DIF3D code, it can calculate the effective delayed neutron fraction({beta}{sub eff}), neutron lifetime(l{sub eff}), fission spectrum ({chi}-bar) and fission yield data({nu}) for each fissionable isotope, composition of fuels and over the whole core. BETA-K code has been validated by comparing the calculated values to the measured ones of effective delayed neutron fraction in two critical experiments, BFS73-1 and BFS55-1. BFS73-1 is a metal uranium core and BFS55-1 is a metal plutonium core. The C/E values, 1.007 and 0.992 for BFS73-1 and BFS55-1 respectively, agreed well with the experimental values within the experiment errors. BETA-K code predicts 0.00709 and 0.356 {mu}sec as the effective delayed neutron fraction and neutron life time for the uranium metallic fueled equilibrium core of 150MWe KALIMER. (author). 9 refs., 6 figs., 12 tabs.

  17. Combining neutron and X-ray imaging to study air and water behaviour in the soil macropores

    Science.gov (United States)

    Snehota, Michal; Sobotkova, Martina; Jelinkova, Vladimira; Kaestner, Anders

    2016-04-01

    Infiltration of water and gas trapping in soil macropores were investigated on intact sample of coarse sandy loam soil (Cambisol series) taken from the B horizon by combined X-ray and neutron tomography imaging. The soil under study is known for the occurrence of the preferential flow, in which a majority of the water flux is conducted through small, highly conductive, fraction of the soil volume. Experiment performed in the NEUTRA beamline of Paul Scherrer Institut consisted of two infiltration episodes during which a layer of heavy and light water mixture was maintained on the sample surface created a ponding boundary condition. The initial state of the sample was recorded by one X-ray and two neutron scans prior to the first infiltration. Another 20 neutron tomograms were acquired during the following 25 hours of the experiment. Fine co-registration of the reconstructed X-ray and neutron tomograms was performed. Then, bi-variate histograms helped to identify the thresholds that were subsequently used for segmentation of the macropores from the X-ray tomograms. The segmented regions served as a binary mask for calculating the water volume using the neutron tomograms. Volume of water and subsequently the average water content in the macropore system were calculated. Results then quantitatively show the extent of the water content reduction in the macropores during the second infiltration that was caused by enhanced air trapping in the wet soil.

  18. Program POD. A computer code to calculate cross sections for neutron-induced nuclear reactions

    International Nuclear Information System (INIS)

    A computer code, POD, was developed for neutron-induced nuclear data evaluations. This program is based on four theoretical models, (1) the optical model to calculate shape-elastic scattering and reaction cross sections, (2) the distorted wave Born approximation to calculate neutron inelastic scattering cross sections, (3) the preequilibrium model, and (4) the multi-step statistical model. With this program, cross sections can be calculated for reactions (n, γ), (n, n'), (n, p), (n, α), (n, d), (n, t), (n, 3He), (n, 2n), (n, np), (n, nα), (n, nd), and (n, 3n) in the neutron energy range above the resonance region to 20 MeV. The computational methods and input parameters are explained in this report, with sample inputs and outputs. (author)

  19. Calculation of neutron die-away times in a large-vehicle portal monitor

    International Nuclear Information System (INIS)

    Monte Carlo methods have been used to calculate neutron die-away times in a large-vehicle portal monitor. These calculations were performed to investigate the adequacy of using neutron die-away time measurements to detect the clandestine movement of shielded nuclear materials. The geometry consisted of a large tunnel lined with He3 proportional counters. The time behavior of the (n,p) capture reaction in these counters was calculated when the tunnel contained a number of different tractor-trailer load configurations. Neutron die-away times obtained from weighted least squares fits to these data were compared. The change in neutron die-away time due to the replacement of cargo in a fully loaded truck with a spherical shell containing 240 kg of borated polyethylene was calculated to be less than 3%. This result together with the overall behavior of neutron die-away time versus mass inside the tunnel strongly suggested that measurements of this type will not provide a reliable means of detecting shielded nuclear materials in a large vehicle. 5 figures, 4 tables

  20. CHC program for calculation of the adjoint neutron cross sections on the basis of evaluated neutron data of the ENDF/B

    International Nuclear Information System (INIS)

    The features and the algorithm of the program to calculate adjoint neutron cross sections on the basis of the continuous energy neutron cross sections as well as energy and angular distributions are described. The calculated adjoint cross sections are intended for Monte Carlo investigation of the nonuniform adjoint Boltzmann equation. 16 refs

  1. Calculation of the Inelastic Scattering of Neutrons from Polyethylene and Water

    International Nuclear Information System (INIS)

    A model for the calculation of the scattering of thermal neutrons from chemical system was proposed by Nelkin. This model considered the actual dynamics of the scattering system as composed of a set of oscillatory motions, each describable by a Hamiltonian which commuted with each of the others. It was then possible to express the differential scattering cross-section in closed form. This model has been used to calculate the scattering of neutrons by water. Some care must be taken in performing the numerical integration over angle and energy. The scattering model has been extended to the calculation of neutron scattering from polyethylene CnH2n. Analogous levels of polyethylene can be noted at 0.089 eV, 0.182 eV, 0.354 eV, and 0.533 eV. The differential and total cross-sections have been calculated for the scattering and the latter has been seen to be in reasonable agreement with experiment at room temperature. Scattering kernels have been calculated for a number of temperatures and where possible the results have been compared with experiment. In addition, neutron flux spectra and diffusion lengths have been calculated using the equations of reactor physics. Comparison of these Results with experimental data indicates that such integral measurements are indicative of at least the gross features of the scattering system and should be analysed in conduction with the detailed differential cross-section results. (author)

  2. Comparisons of Measured and Calculated Neutron Fluxes in Laminated iron and Heavy Water

    International Nuclear Information System (INIS)

    Measurements of neutron fluxes have been performed in configurations depicting the regions extending radially and axially outwards from the core of a PHWR reactor in order to test the accuracy of the available methods in shield design on thin alternating laminae of Fe and D2O. A 'dry' experimental set-up was constructed, i.e. the D2O was contained in flat tanks made of Al. The first set of measurements was performed through solid Fe and D2O layers, and only the results of these experiments are described in this report. The set-up allowed measurements also in a mock-up of a reactor top penetrated by D2O or air-filled channels (to be reported later). The results are compared to fluxes calculated by the British 18-group removal-diffusion method and by the NRN method developed at AE. The results show that the values predicted may be expected to be within a factor of 2 from the true values in most cases. The predicted relative flux distributions follow the observed ones with a very good accuracy in spite of the apparent misuse of diffusion theory for the thin regions in question. Finally, it is shown that the predicted change in the fast spectrum while penetrating these set-ups should be confirmable with certain threshold detectors

  3. Calculation of the angular distribution of delay times in neutron scattering on 58Ni nuclei

    International Nuclear Information System (INIS)

    Angular distributions of average delay times and time variances are calculated for resonance-neutron scattering on 58Ni nuclei at neutron energies in the range E = 600−700 keV. The effect of the energy spectrum and polarization of the beam on the scattering-process time is discussed. The angular dependence of the time law is also considered for the decay of an intermediate compound nuclear system. It is shown that the results of stationary and nonstationary calculations are in good agreement.

  4. Calculation of the angular distribution of delay times in neutron scattering on {sup 58}Ni nuclei

    Energy Technology Data Exchange (ETDEWEB)

    Prokopets, G. A., E-mail: gaprok@uos.net.ua [National University of Kyiv-Mohyla Academy (Ukraine)

    2011-05-15

    Angular distributions of average delay times and time variances are calculated for resonance-neutron scattering on {sup 58}Ni nuclei at neutron energies in the range E = 600-700 keV. The effect of the energy spectrum and polarization of the beam on the scattering-process time is discussed. The angular dependence of the time law is also considered for the decay of an intermediate compound nuclear system. It is shown that the results of stationary and nonstationary calculations are in good agreement.

  5. Role of the Tapiro Fast Research Reactor in Neutron Capture Therapy in Italy Calculations and Measurements

    International Nuclear Information System (INIS)

    Thermal-neutron research reactors are currently the most common source of neutron beams for both research and clinical trials of neutron capture therapy (NCT). Neutron spectra suitable for NCT are typically produced either by beam filtering or spectrum shifting techniques. However, fast-neutron reactors are also being considered for NCT application as it is recognized that they may allow for improved beam quality. TAPIRO is a low power, high flux, highly enriched (93.5% 235U) fast reactor. The power is 5 kW and the maximum neutron flux in the core is 3x1012 cm-2.s-1. Both a thermal and an epithermal column have been designed and constructed, aimed at dosimetry and animal experiments. The configurations of the columns have been designed by means of Monte Carlo calculations. The columns have been characterized by means of measurements performed with activation techniques and thermoluminescence and gel dosimeters. Experimental results have shown good consistency with calculations. Moreover, they have confirmed the good quality of the beams obtainable with such a reactor. An epithermal column for clinical trials of patients with brain gliomas has been designed and is under construction. The treatment planning figures-of-merit in an anthropomorphic phantom look very satisfactory. (author)

  6. Calculation of Neutron Resonance Spacing with Microscopic Theory

    Institute of Scientific and Technical Information of China (English)

    A.N. Behkami

    2003-01-01

    Nuclear level spacings calculated with a microscopic theory are compared with spacings determined fromneutron resonance experiment. The gross features of the experimental data due to nuclear shells are reproduced with themicroscopic theory. The experimental data for nuclei with statistically deformed nuclei have also been tested with leveldensity formula including low energy rotational levels. The experimental data for the actinide nuclei and the lanthanidenuclei are found to be consistent with the theory which includes collective rotational levels.

  7. Neutron reflectivity measurement of polymer monolayer and brush at the air/water interface

    International Nuclear Information System (INIS)

    We have been studied on amphiphilic polymer monolayer structure at the air/water interface by X-ray and neutron reflectometry. By complemently use of X-ray and neutron reflectometry, we have found (1) the existence of carpet layer in ionic polymer brush in monolayer system and (2) characteristic structural change in polymer/subphase interface. Furthermore, interesting experiment on small ion distribution was carried out by NR with contrast variation method. With our experimental examples, characteristic points in the neutron reflectivity measurement at the air/water interface and further possibility in this research area are discussed. (author)

  8. 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.

  9. Monte Carlo simulation in the reaction rate's calculation with neutron-activation method

    International Nuclear Information System (INIS)

    With MCNP/4B code, the influence of cut-off energy, flux tallies, nuclear databases and perturbation on the reaction rate's calculation with neutron-activation method are analysed. When the effective reaction threshold is chosen as the cut-off energy, calculation time is considerably reduced and yet the results are not changed. Comparing calculations with cell tallies (F4) with those performed with detector tallies (F5), the counting efficiency of cell tallies is higher and the results are slightly higher, but still credible. With different nuclear databases, calculated results can be different. The perturbation among the detectors doesn't effect on the calculated results. (authors)

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

    International Nuclear Information System (INIS)

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

  11. Scission neutrons for U, Pu, Cm, and Cf isotopes: Relative multiplicities calculated in the sudden limit

    Science.gov (United States)

    Capote, R.; Carjan, N.; Chiba, S.

    2016-02-01

    The multiplicities of scission neutrons νs c are calculated for series of U, Pu, Cm, and Cf isotopes assuming a sudden transition between two different nuclear configurations (αi→αf ): one just before the neck rupture and one immediately after the disappearance of the neck. This calculation requires only the knowledge of the corresponding two sets of neutron eigenstates. The nuclear shapes around the scission point are described in terms of Cassinian ovals with only two parameters: α (that positions the shape with respect to the zero-neck shape) and α1 (that defines the mass asymmetry). Based on these shapes, a neutron mean field of the Woods-Saxon type is constructed using two prescriptions to calculate the distance to the nuclear surface. The accent in the present work is put on the dependence of νs c on the neutron number Nf of the fissioning nucleus and on the mass asymmetry AL/AH of the primary fission fragments. The relative dependence of these multiplicities, averaged over the mass yields, , are finally compared with existing experimental data on prompt fission neutrons .

  12. 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.

  13. 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.)

  14. 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.

  15. Monte Carlo calculation of neutron generation time in critical reactor and subcritical reactor with an external source

    International Nuclear Information System (INIS)

    The neutron generation time Λ plays an important role in the reactor kinetics. However, it is not straightforward nor standard in most continuous energy Monte Carlo codes which are able to calculate the prompt neutron lifetime lp directly. The difference between Λ and lp are sometimes very apparent. As very few delayed neutrons are produced in the reactor, they have little influence on Λ. Thus on the assumption that no delayed neutrons are produced in the system, the prompt kinetics equations for critical system and subcritical system with an external source are proposed. And then the equations are applied to calculating Λ with pulsed neutron technique using Monte Carlo. Only one fission neutron source is simulated with Monte Carlo in critical system while two neutron sources, including a fission source and an external source, are simulated for subcritical system. Calculations are performed on both critical benchmarks and subcritical system with an external source and the results are consistent with the reference values. (author)

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

    OpenAIRE

    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...

  17. Calculation of effective delayed neutron fraction with modified library of Monte Carlo code

    International Nuclear Information System (INIS)

    Highlights: ► We propose a new Monte Carlo method to calculate the effective delayed neutron fraction by changing the library. ► We study the stability of our method. When the particles and cycles are sufficiently great, the stability is very good. ► The final result is determined to make the deviation least. ► We verify our method on several benchmarks, and the results are very good. - Abstract: A new Monte Carlo method is proposed to calculate the effective delayed neutron fraction βeff. Based on perturbation theory, βeff is calculated with modified library of Monte Carlo code. To verify the proposed method, calculations are performed on several benchmarks. The error of the method is analyzed and the way to reduce error is proposed. The results are in good agreement with the reference data

  18. 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)

  19. 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

  20. 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

  1. The neutron 'thunder' accompanying large extensive air showers

    OpenAIRE

    Erlykin, A. D.

    2006-01-01

    The bulk of neutrons which appear with long delays in neutron monitors nearby the EAS core (~'neutron thunder'~) are produced by high energy EAS hadrons hitting the monitors. This conclusion raises an important problem of the interaction of EAS with the ground, the stuff of the detectors and their environment. Such interaction can give an additional contribution to the signal in the EAS detectors at {\\em km}-long distances from the large EAS core after a few $\\mu s$ behind the EAS front.

  2. 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.

  3. Calculation of the moderator temperature coefficient of reactivity for miniature neutron source reactors

    International Nuclear Information System (INIS)

    This paper presents results of the evaluated group constants for fuel and other important materials of the Miniature Neutron Source Reactor (Mnr) and the moderator temperature coefficient of reactivity through global reactor calculation. In this study the group constants were calculated with the WIMSD code and the global reactor calculation is accomplished by the CITATION code. This work also presents a method for evaluation of the moderator temperature coefficient of reactivity at different temperatures and it's average value in a range of temperature directly through the values of moderator temperature for MNSRs. This method provides simple analytical representation convenient for reactor kinetics calculation and reactor safety assessment. (author)

  4. Calculations to support JET neutron yield calibration: Modelling of the JET remote handling system

    Energy Technology Data Exchange (ETDEWEB)

    Snoj, Luka, E-mail: luka.snoj@ijs.si [JET-EFDA, Culham Science Centre, OX14 3DB Abingdon (United Kingdom); EURATOM-MHEST Association, Reactor Physics Division, Jožef Stefan Institute, Jamova Cesta 39, SI-1000 Ljubljana (Slovenia); Lengar, Igor; Čufar, Aljaž [JET-EFDA, Culham Science Centre, OX14 3DB Abingdon (United Kingdom); EURATOM-MHEST Association, Reactor Physics Division, Jožef Stefan Institute, Jamova Cesta 39, SI-1000 Ljubljana (Slovenia); Syme, Brian; Popovichev, Sergey [JET-EFDA, Culham Science Centre, OX14 3DB Abingdon (United Kingdom); EURATOM-CCFE Fusion Association, Culham Science Centre, Abingdon OX14 3DB, OXON (United Kingdom); Conroy, Sean [JET-EFDA, Culham Science Centre, OX14 3DB Abingdon (United Kingdom); EURATOM-VR Association, Department of Physics and Astronomy, Uppsala University, Box 516, SE-75120 Uppsala (Sweden); Meredith, Lewis [JET-EFDA, Culham Science Centre, OX14 3DB Abingdon (United Kingdom); EURATOM-CCFE Fusion Association, Culham Science Centre, Abingdon OX14 3DB, OXON (United Kingdom)

    2013-08-15

    Highlights: ► We model JET remote handling system in MCNP. ► We examine the effect of JET remote handling system on neutron monitor response. ► The integral effect of JET RH system on neutron monitors is less than 5%. -- Abstract: After the coated CFC wall to ITER-Like Wall (Beryllium/Tungsten/Carbon) transition in 2010–2011, confirmation of the neutron yield calibration will be ensured by direct measurements using a calibrated {sup 252}Cf neutron source deployed by the in-vessel remote handling boom and Mascot manipulator inside the JET vacuum vessel. Neutronic calculations are required to calculate the effects of the JET remote handling (RH) system on the neutron monitors. We developed a simplified geometrical computational model of the JET remote handling system in MCNP. In parallel we developed a script that translates the RH movement data to transformations of individual geometrical parts of the RH model in MCNP. After that a benchmarking of the model was performed to verify and validate the accordance of the target positions of source and RH system with the ones from our model. In the last phase we placed the JET RH system in the simplified MCNP model of the JET tokamak and studied its effect on neutron monitor response for some example source positions and boom configurations. As the correction factors due to presence of the JET RH system can potentially be significant in cases when the boom is blocking a port close to the detector under investigation, we have chosen boom configurations so that this is avoided in the vast majority of the source locations. Examples are given.

  5. Calculation of diffusion coefficients in air-metal thermal plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Cressault, Y; Gleizes, A [Universite de Toulouse, UPS, INPT, LAPLACE (Laboratoire Plasma et Conversion d' Energie), 118 route de Narbonne, F-31062 Toulouse Cedex 9 (France)

    2010-11-03

    This paper presents the combined diffusion coefficients of metal vapours (silver, copper and iron) in air thermal plasmas for temperatures ranging from 300 to 30 000 K. The theory used to calculate these coefficients is remembered and validated by comparison with the literature values in several cases such as Ar-He, Ar-Cu and N{sub 2}-O{sub 2} mixtures. The results are discussed showing the influences of the metal concentration, of the vapour nature and of the pressure. The results show rather similar behaviour for the three metals. The maximum values of the combined ordinary diffusion coefficient in the evolution with temperature are obtained for temperature around 10 000 K but this peak is shifted to the highest temperatures when the metal proportion increases. Another result shows that the diffusion coefficient decreases when pressure increases.

  6. 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)

  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. Measurement and calculation of 238U fission reaction rates induced by neutrons reflected by carbon material

    International Nuclear Information System (INIS)

    To check the data of carbon material reflecting neutrons, the distribution of 238U fission reaction rates induced by D-T fusion neutrons reflected by carbon material was measured by using the small depleted uranium fission chamber and the capturing detector. For comparison, 238U fission rates without carbon material was measured too. The combined standard uncertainty of 238U fission reaction rate is 5.1%-6.4%. The measured results are consistent with the calculated ones with MCNP/4A code and ENDF/B-IV library data in the range of the error

  9. Neutron dosimetry and damage calculations for the JP-17, 18 and 19 experiments in HFIR

    Energy Technology Data Exchange (ETDEWEB)

    Greenwood, L.R.; Baldwin, C.A.

    1996-04-01

    Neutron fluence measurements and radiation damage calculations are reported for the joint US-Japanese experiments JP-17, 18, and 19 in the target of the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL). These experiments were irradiated at 85 MW for two cycles resulting in 43.55 EFPD for JP-17 and 42.06 EFPD for JP-18 and 19. The maximum fast neutron fluence > 0.1 MeV was about 3.7E + 21 n/cm{sup 2} for all three irradiations, resulting in about 3 dpa in 316 stainless steel.

  10. Neutron dosimetry and damage calculation for the JP-10, 11, 13, and 16 experiments in HFIR

    Energy Technology Data Exchange (ETDEWEB)

    Greenwood, L.R.; Ratner, R.T.

    1996-04-01

    Neutron fluence measurements and radiation damage calculations are reported for the joint U.S./Japanese experiments JP-10, 11, 13, and 16 in the target of the High Flux Isotope Reactor (HFIR) at Oak Ridge National Lab (ORNL). These experiments were irradiated at 85 MW for 238.5 EFPD. The maximum fast neutron fluence >0.1 MeV was about 2.1E + 22 n/cm{sup 2} for all of the experiments resulting in about 17.3 dpa in 316 stainless steel.

  11. 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.

  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-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.

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

    International Nuclear Information System (INIS)

    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 x 1022 n/cm2 (1.9 x 1022 n/cm2 above 0.1 MeV), resulting in about 12 dpa and 28 appm helium in type 316 stainless steel

  15. Results of coupled channels calculations for the neutrons cross sections of a set of actinide nuclei

    International Nuclear Information System (INIS)

    This report gathers recents results of neutrons interactions with the following actinide nuclei: 230Th, 232Th, 234U, 238U, 242Pu, 246Cm and 252Cf from the use of the coupled channels optical model. Tabulations of the following quantities are given in Annexe: total, direct elastic and inelastic scattering (integrated and differential), and compound nucleus formation cross sections; ground state generalized transmission coefficients needed to calculate the cross sections of partial compound nucleus processes. This work was carried out within the framework of the IAEA-NDS Coordinated Research Programme on the Intercomparison of Actinide Neutron Cross Section Evaluations

  16. Single event upsets calculated from new ENDF/B-VI proton and neutron data up to 150 MeV

    Energy Technology Data Exchange (ETDEWEB)

    Chadwick, M.B. [Los Alamos National Lab., NM (United States). Theoretical Div.; Normand, E. [Boeing Military Aircraft and Missile Systems, Seattle, WA (United States)

    1999-06-01

    Single-event upsets (SEU) in microelectronics are calculated from newly-developed silicon nuclear reaction recoil data that extend up to 150 MeV, for incident protons and neutrons. Calculated SEU cross sections are compared with measured data.

  17. 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.

  18. Using of discrete ordinate method in the spallation target neutronics and shielding calculations

    International Nuclear Information System (INIS)

    A discrete ordinate algorithm for coupled charged/neutral particle transport calculations in 2D pencil beam problems is developed. It is based on the use of the second order of accuracy adaptive WDD (AWDD) scheme for approximation both the continuous slowing down (CSD) and streaming terms of the charged particle transport equation in z geometry, and a suitable algorithm for treatment of the extended uncollided flux from an initially monodirectional beam of charged particles with given radial distribution. The developed algorithm is implemented in the 2D transport code KASKAD-S-1.5 and is applied to the high-energy coupled proton-pion-neutron-photon transport calculations. The multigroup cross-section library SADCO-2 for nucleon-meson cascade calculations coupled with standard neutron and gamma-ray cross-section libraries below 20 MeV is used. Some numerical examples are given.(author)

  19. MAMONT program for neutron field calculation by the Monte Carlo method

    International Nuclear Information System (INIS)

    The MAMONT program (MAthematical MOdelling of Neutron Trajectories) designed for three-dimensional calculation of neutron transport by analogue and nonanalogue Monte Carlo methods in the range of energies from 15 MeV to the thermal ones is described. The program is written in FORTRAN and is realized at the BESM-6 computer. Group constants of the library modulus are compiled of the ENDL-83, ENDF/B-4 and JENDL-2 files. The possibility of calculation for the layer spherical, cylindrical and rectangular configurations is envisaged. Accumulation and averaging of slowing-down kinetics functionals (averaged logarithmic losses of energy, time of slowing- down, free paths, the number of collisions, age), diffusion parameters, leakage spectra and fluxes as well as formation of separate isotopes over zones are realized in the process of calculation. 16 tabs

  20. Lattice EFT calculation of thermal properties of low-density neutron matter

    International Nuclear Information System (INIS)

    Thermal properties of low-density neutron matter are investigated by lattice calculation with nuclear effective field theory without pions up to the next-to-leading order. The 1S0 pairing gap is extracted near zero temperature at low densities. We find that the pairing gap is smaller than the BCS approximation with the conventional NN potentials, but not as small as those by various many-body calculations beyond BCS approximation. Our result is consistent with the recent Green's Function Monte Carlo calculation within the statistical errors. The critical temperature of the normal-to-superfluid phase transition and the pairing temperature scale are also extracted at low densities, and the phase diagram is given. We find that the physics of low-density neutron matter is clearly identified as being BCS-BEC crossover.

  1. 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.

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

    Science.gov (United States)

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

    1986-02-01

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

  3. Measurement and calculation of the neutron flux distribution in the RP-10 reactor

    International Nuclear Information System (INIS)

    In this work implementing experimental methods are implemented for easy reproduction for measuring the spatial distribution or thermal neutron flux in the RP-10 reactor core. Using two measuring methods: the passive and the active ones. In the passive method was used the activation technique using foils such as gold, manganese, and indium. These were irradiated in the reactor core and treated through the Westcott's formalism. In the active method was used the Self Powered Neutron Detectors (SPNs) for which was necessary to condition the detectors response for the data acquisition. The knowledge of the spatial distribution of RP-10 reactor neutrons flux will contribute in the understanding of other interesting parameters of reactor physics such as power density, reactivity, buckling, etc.. Wish knowledge is important for reactor operation. Fuel burnup calculations as well as others related to safety. (author)

  4. 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.

  5. 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.

  6. 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.

  7. Calculational study on neutron kinetics and thermodynamics of a gaseous core fission reactor. Doctoral thesis

    Energy Technology Data Exchange (ETDEWEB)

    Kuijper, J.C.

    1992-01-01

    The aim of the authors' work was to investigate the static and dynamic properties of a GCFR with oscillating (moving) fuel gas. A simplified schematic diagram of such a GCFR, similar to the concept of Kistemaker (Kis78a), is shown. It consists of a graphite cylinder of, say, 2 m diameter and 10 m length, filled with a mixture of uranium and carbon fluorides (UCF) at high temperature in ionized state, in chemical and thermodynamical equilibrium with the graphite cylinder wall (Kis78a, Kis86, Kle87). The cylindrical gas space is divided into an active 'core' region, surrounded by an effective (thick) neutron reflector, and a so-called 'expander' region, surrounded by a much less effective (thinner or with neutron poison) neutron reflector. In operation, part of the fuel gas oscillates back and forth between core and expander region. The investigation requires the study of neutron statics, neutron kinetics, reactor gas thermodynamics and gas dynamics, resulting in a combined calculational model, containing these aspects. In order to achieve this the authors followed a step-by-step approach.

  8. 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...

  9. 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

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

    International Nuclear Information System (INIS)

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

  11. 232Th and 238U neutron emission cross section calculations and analysis of experimental data

    International Nuclear Information System (INIS)

    In this study, pre-equilibrium neutron-emission spectra produced by (n,xn) reactions on nuclei 232Th and 238U have been calculated. Angle-integrated cross sections in neutron induced reactions on targets 232Th and 238U have been calculated at the bombarding energies up to 18 MeV. We have investigated multiple pre-equilibrium matrix element constant from internal transition for 232Th (n,xn) neutron emission spectra. In the calculations, the geometry dependent hybrid model and the cascade exciton model including the effects of pre-equilibrium have been used. In addition, we have described how multiple pre-equilibrium emissions can be included in the Feshbach-Kerman-Koonin (FKK) fully quantum-mechanical theory. By analyzing (n,xn) reaction on 232Th and 238U, with the incident energy from 2 Me V to 18 Me V, the importance of multiple pre-equilibrium emission can be seen cleady. All calculated results have been compared with experimental data. The obtained results have been discussed and compared with the available experimental data and found agreement with each other

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

    International Nuclear Information System (INIS)

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

  13. Benchmark test of TRIPOLI-4 code through simple model calculation and analysis of fusion neutronics experiments at JAEA/FNS

    Energy Technology Data Exchange (ETDEWEB)

    Ohta, Masayuki, E-mail: ohta.masayuki@jaea.go.jp [Japan Atomic Energy Agency, Tokai-mura, Naka-gun, Ibaraki-ken 319-1195 (Japan); Takakura, Kosuke; Ochiai, Kentaro; Sato, Satoshi; Konno, Chikara [Japan Atomic Energy Agency, Tokai-mura, Naka-gun, Ibaraki-ken 319-1195 (Japan)

    2013-10-15

    In order to examine a basic performance of the TRIPOLI code, two types of analyses were carried out with TRIPOLI-4.4 and MCNP5-1.40; one is a simple model calculation and the other is an analysis of iron fusion neutronics experiments with DT neutrons at the Fusion Neutronics Source (FNS) facility in Japan Atomic Energy Agency (JAEA). In the simple model calculation, we adopted a sphere of 0.5 m in radius with a 20 MeV neutron source in the center and calculated leakage neutron spectra from the sphere. We also analyzed in situ and Time-of-Flight (TOF) experiments for iron at JAEA/FNS. For the in situ experiment, neutron spectra and reaction rates for dosimetry reactions were calculated for several points inside the assembly. For the TOF experiment, angular neutron leakage spectra from the assembly were calculated. Results with TRIPOLI were comparable to those with MCNP in most calculations, but a difference between TRIPOLI and MCNP calculation results, probably caused by inadequate treatment of inelastic scattering data in TRIPOLI, appears in some calculations.

  14. The neutron 'thunder' accompanying the extensive air shower

    Energy Technology Data Exchange (ETDEWEB)

    Erlykin, A D [P. N. Lebedev Physical Institute, Moscow (Russian Federation)

    2007-03-15

    Simulations show that neutrons are the most abundant component among extensive air shower (EAS) hadrons. However, multiple neutrons which appear with long delays in neutron monitors nearby the EAS core (neutron thunder) are mostly not the neutrons of the shower, but have a secondary origin. The bulk of them is produced by high energy EAS hadrons hitting the monitors. The delays are due to the thermalization and diffusion of neutrons in the moderator and reflector of the monitor accompanied by the production of secondary gamma quanta. This conclusion raises the important problem of the interaction of EAS with the ground, the stuff of the detectors and their environment since they have often hydrogen-containing materials like polyethilene in neutron monitors. Such interaction can give an additional contribution to the signal in the EAS detectors. It can be particularly important for the signals from scintillator or water tank detectors at kilometre-long distances from the EAS core, where neutrons of the shower become the dominant component after a few microseconds behind the EAS front.

  15. Relativistic collision rate calculations for electron-air interactions

    International Nuclear Information System (INIS)

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

  16. dpa calculations for neutron irradiated amorphous Fe40Ni40B20 ribbons

    International Nuclear Information System (INIS)

    Calculations of the displacement per atom have been performed for an amorphous metal Fe40Ni40B20, which has been exposed to incore reactor irradiation. Up to 99% of the radiation induced displacements result from the nuclear reaction 10B(n,α)7Li + 2.79 MeV, which is initiated by thermal neutrons. The number of primary knock-on atoms generated by α- and Li-particles and the size of cascades have also been calculated. A fluence of thermal neutrons of 1 x 1019 nsub(th)/cm2 is found to produce a damage in amorphous Fe40Ni40B20 which corresponds to 0.65 dpa. (orig.)

  17. Power and neutron flux calculation for the PUSPATI TRIGA Reactor using MCNP

    International Nuclear Information System (INIS)

    The Malaysian 1 MW TRIGA MARK II research reactor at Malaysian Nuclear Agency is designed to effectively implement the various fields of basic nuclear research, manpower training, and production of radioisotopes for their use in agriculture, industry, and medicine. This study deals with the calculation of neutron flux and power distribution in PUSPATI TRIGA REACTOR (RTP) 14th core configuration. The 3-D continuous energy Monte Carlo code MCNP was used to develop a versatile and accurate full model of the TRIGA core and fuels. The model represents in detailed all components of the core with literally no physical approximation. Continuous energy cross-section data from the more recent nuclear data as well as S (α, β) thermal neutron scattering functions distributed with the MCNP code were used. Results of calculations are analyzed and discussed. (author)

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

    Institute of Scientific and Technical Information of China (English)

    李先庭; 李冬宁; 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.

  19. 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.

  20. Methods for the calculation of neutron nuclear data for structural materials of fast and fusion reactors

    International Nuclear Information System (INIS)

    This report contains the texts of the invited presentations (20) delivered at the third Research Co-ordination Meeting of the Co-ordinated Research Programme on Methods for the Calculation of Neutron Nuclear Data for Structural Materials of Fast and Fusion Reactors. The meeting was held at the IAEA Headquarters, Vienna, Austria, from 20 to 22 June 1990. A separate abstract was prepared for each of these presentations. Refs, figs and tabs

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

    International Nuclear Information System (INIS)

    Highlights: ► We have extended the KAERI library generation system to include gamma cross section generation capability. ► A gamma transport/diffusion calculation module has been implemented in KARMA 1.2. ► The computational results for benchmark problems show that the gamma library and gamma simulation in KARMA are reasonable. - Abstract: KAERI has developed a lattice transport calculation code KARMA (Kernel Analyzer by Ray-tracing Method for fuel Assembly) and its library generation system. Recently, the library generation system has been extended to include a gamma cross section generation capability and a gamma transport/diffusion calculation module has been implemented in KARMA 1.2. The method of characteristics for the neutron transport calculation to estimate eigenvalue has been utilized to predict gamma flux distribution and energy deposition. In addition, the coarse mesh finite difference method with diffusion approximation has also been utilized to estimate gamma flux distribution and energy depositions for each coarse mesh with homogenized pins as a computationally efficient alternative. This paper describes the procedure to generate neutron induced gamma production and gamma cross section data, and the methods to predict gamma flux distribution, gamma energy deposition and gamma smeared pin power distribution. The computational results for benchmark problems show that the gamma library and gamma simulation in KARMA are reasonable. And it is noted that gamma smeared power distributions predicted by coarse mesh diffusion calculation are very accurate compared to the results of transport calculation

  2. Calculation And Design Of A New Configuration For Radiation Shielding At Neutron Beam No.3 For Fundamental And Applied Researches

    International Nuclear Information System (INIS)

    The tangential horizontal channel of No. 3 of the Dalat Research Reactor has been opened and used during the 1990s. The utilizations of the thermal neutron beam at this channel were the Neutron Radiography and the Prompt Gamma Neutron Activation Analysis method (PGNAA). At present, the neutron beam used for nuclear structure data researches based on the Summing of Amplitude Coincident Pulses system (SACP). Beside, several related research equipments have been set up and operated for the research purposes. A renovation of the neutron channel, therefore, will play an important role in safe and effective utilizations of the neutron beam in fields of nuclear physic training and researches. A new configuration for radiation shielding has been simulated by MCNP code. The calculated results of dose rates for neutron and gamma at working positions are in range of dose rate limit. (author)

  3. 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.

  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 neutron and gamma-ray emission spectra produced by p + 27Al reactions

    International Nuclear Information System (INIS)

    Preliminary calculations of neutron and gamma-ray spectra induced by proton reactions on aluminum have been made to provide data required for shielding design for a proposed proton linear accelerator. The nuclear models used in this study were the preequilibrium and Hauser-Feshbach models as embodied in the GNASH program. This nuclear model code has been used in the past to successfully investigate higher energy (E less than or equal to 50 MeV) neutron and proton interactions with nuclei in the structural materials region. Because this study was of an exploratory nature, we did not attempt to optimize input parameters but instead relied upon global sets, especially for optical parameters. In particular, for neutrons we chose the Wilmore-Hodgson parameter set after confirmation of its suitability through comparison to n+27Al total cross-section data between 0.5 and 60 MeV. Agreement with the data on the level of 5-10% occurred. Comparisons were also made to measured nonelastic data for incident energies between 10 and 60 MeV. Again, there was generally good agreement although there was some tendency to overpredict such data for incident neutron energies below several MeV. For protons we found the Becchetti-Greenlees parameter set reproduced nonelastic data recently measured by McGill et al

  6. 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.

  7. Specific yield functions of neutron monitors and accuracy of spectrum calculation of solar cosmic rays (SCR)

    International Nuclear Information System (INIS)

    The specific yield functions m(R) for the neutron component of cosmic rays have been calculated on the basis of latitude surveys made on board the scientific ship Akademik Kurchatov' in late 1971 - early 1972. Calculations have been performed using the primary cosmic ray spectrum reconstructed with regard for possible anomalous modulation effects of the Sun. Significant discrepancies (up to 2 orders of magnitude) have been found in m(R) values obtained by different authors. They seem to be due to non-adequacy of the primary spectra to the latitude curves used in calculations. The accuracy of SCR absolute spectrum in the range of R > = 1 GV determined with the help of m(R) does not exceed a factor of 2. However it may be improved by comparing the calculated spectra with direct measurements in R < 1 GV range

  8. 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.

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

    International Nuclear Information System (INIS)

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

  10. 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...

  11. Neutronic modelling of the reflector for the calculation of pressurized water reactors: application to EPR

    International Nuclear Information System (INIS)

    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 calculation

  12. Calculation of neutron fluence-to-dose conversion factors for extremities

    International Nuclear Information System (INIS)

    The Pacific Northwest Laboratory is developing a standard for the performance testing of personnel extremity dosimeters for the US Department of Energy. Part of this effort requires the calculation of neutron fluence-to-dose conversion factors for finger and wrist extremities. This study focuses on conversion factors for two types of extremity models: namely the polymethyl methacrylate (PMMA) phantom (as specified in the draft standard for performance testing of extremity dosimeters) and more realistic extremity models composed of tissue-and-bone. Calculations for each type of model are based on both bare and D2O-moderated 252Cf sources. The results are then tabulated and compared with whole-body conversion factors. More appropriate energy-averaged quality factors for the extremity models have also been computed from the neutron fluence in 50 equally spaced energy bins with energies from 2.53 x 10-8 to 15 MeV. Tabulated results show that conversion factors for both types of extremity phantom are 15 to 30% lower than the corresponcung whole-body phantom conversion factors for 252Cf neutron sources. This difference in extremity and whole-body conversion factors is attributable to the proportionally smaller amount of back-scattering that occurs in the extremity phantoms compared with whole-body phantoms

  13. Neutrons and Gamma-Ray Dose Calculations in Subcritical Reactor Facility Using MCNP

    Directory of Open Access Journals (Sweden)

    Ned Xoubi

    2016-06-01

    Full Text Available In nuclear experimental, training and teaching laboratories such as a subcritical reactor facility, huge measures of external radiation doses could be caused by neutron and gamma radiation. It becomes imperative to place the health and safety of staff and students in the reactor facility under proper scrutiny. The protection of these individuals against ionization radiation is facilitated by expected dose mapping and shielding calculations. A three-dimensional (3D Monte Carlo model was developed to calculate the dose rate from neutrons and gamma, using the ANSI/ANS-6.1.1 and the ICRP-74 flux-to-dose conversion factors. Estimation for the dose was conducted across 39 areas located throughout the reactor hall of the facility and its training platform. It was found that the range of the dose rate magnitude is between 7.50 E−01 μSv/h and 1.96 E−04 μSv/h in normal operation mode. During reactor start-up/shut-down mode, it was observed that a large area of the facility can experience exposure to a significant radiation field. This field ranges from 2.99 E+03 μSv/h to 3.12 E+01 μSv/h. There exists no noticeable disparity between results using the ICRP-74 or ANSI/ANS-6.1.1 flux-to-dose rate conversion factors. It was found that the dose rate due to gamma rays is higher than that of neutrons.

  14. 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.

  15. 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%.

  16. Calculation of the ex-core neutron noise induced by fuel vibrations in PWRs

    International Nuclear Information System (INIS)

    Calculation of the neutron noise induced by fuel assembly vibrations in two pressurized water reactor (PWR) cores has been performed to investigate the effect of cycle burnup on the properties of the ex-core detector noise. Pendular vibrations of individual fuel assemblies were assumed to occur at different locations in the core. The auto power spectra density (APSD) of the ex-core detector noise was evaluated with the assumption of stochastic vibrations along a random two-dimensional trajectory. The results show that no general monotonic variation of APSD was found. The increase of APSD occurs predominantly for peripheral assemblies. Assuming simultaneous vibrations of a number of fuel assemblies uniformly distributed over the core with the more realistic perturbation model, the effect of the peripheral assemblies will dominate and the increase of the amplitude of the ex-core neutron noise with burnup can be confirmed. (author)

  17. Direct Numerical Calculation of X-Ray and Neutron Imaging Using Apertures

    Science.gov (United States)

    Christensen, C. R.; Murphy, T. J.

    2001-10-01

    The ICF program makes extensive use of X-ray imaging utilizing apertures. Pinhole, penumbral, and ring aperture imaging have all been demonstrated. For neutrons, pinhole and penumbral aperture imaging are being developed for applications to National Ignition Facility and Laser Megajoule. Previous analysis techniques have used approximations using Fourier transforms to reconstruct a source from the measured image. We proceed in a more straightforward manner: integration of the probability distribution over the areas of the square pixels followed by matrix inversion. Penetration of and scattering within the aperture substrate are explicitly calculated. Consideration of noise and matrix conditioning allow optimal choices for system geometry. Noise reduction is perfomed using constrained singular value decomposition. Using simulated ICF implosions, a noise-reduction algorithm will be demonstrated. Reconstructions will be shown for simulated and real data at different neutron yields. Work performed at LANL under DOE contract No. W-7405-Eng-36

  18. Sensitivity analysis of neutronics calculations in the preliminary design of JAERI experimental fusion reactor

    International Nuclear Information System (INIS)

    Sensitivity of principal neutronics characteristic quantities for the neutron cross sections of JAERI Experimental Fusion Reactor (JXFR) has been studied by means of sensitivity analysis method based on linear perturbation theory. The same study was made previously. After publication of the previous results, however, the SWANLAKE code used to calculate sensitivities was found to include error derived during its conversion process. The study was thus repeated with corrected SWANLAKE. The quantities studied are calculational results for the first preliminary design of JXFR such as the (n, p) reaction rates of 58Ni and 54Fe in the outer part of superconducting toroidal field coil (TFC), the copper atomic displacement rate in the inner part of TFC and the tritium production rate in the outer blanket. Though the calculational results do not contradict essentially the results in the former study, the newly calculated sensivitities were found to be more or less different from the previous ones. Therefore, the results and discussion of analysis given in this report are revised, with the values corrected. The errors of the (n, p) reaction rates and the copper displacement rate due to the uncertainties of cross sections were estimated to be about 50 - 70% and 25 - 65%, respectively, taking into account the direct sensitivity of (n, p) reaction cross sections in the former. (author)

  19. 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)

  20. Study of suitability of Fricke-gel-layer dosimeters for in-air measurements to characterise epithermal/thermal neutron beams for NCT.

    Science.gov (United States)

    Gambarini, G; Artuso, E; Giove, D; Felisi, M; Volpe, L; Barcaglioni, L; Agosteo, S; Garlati, L; Pola, A; Klupak, V; Viererbl, L; Vins, M; Marek, M

    2015-12-01

    The reliability of Fricke gel dosimeters in form of layers for measurements aimed at the characterization of epithermal neutron beams has been studied. By means of dosimeters of different isotopic composition (standard, containing (10)B or prepared with heavy water) placed against the collimator exit, the spatial distribution of gamma and fast neutron doses and of thermal neutron fluence are attained. In order to investigate the accuracy of the results obtained with in-air measurements, suitable MC simulations have been developed and experimental measurements have been performed utilizing Fricke gel dosimeters, thermoluminescence detectors and activation foils. The studies were related to the epithermal beam designed for BNCT irradiations at the research reactor LVR-15 (Řež). The results of calculation and measurements have revealed good consistency of gamma dose and fast neutron 2D distributions obtained with gel dosimeters in form of layers. In contrast, noticeable modification of thermal neutron fluence is caused by the neutron moderation produced by the dosimeter material. Fricke gel dosimeters in thin cylinders, with diameter not greater than 3mm, have proved to give good results for thermal neutron profiling. For greater accuracy of all results, a better knowledge of the dependence of gel dosimeter sensitivity on radiation LET is needed. PMID:26249744

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

    Energy Technology Data Exchange (ETDEWEB)

    Traub, Richard J.

    2010-03-26

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

  2. 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.

  3. 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)

  4. 3D Monte-Carlo transport calculations of whole slab reactor cores: validation of deterministic neutronic calculation routes

    International Nuclear Information System (INIS)

    This paper presents how Monte-Carlo calculations (French TRIPOLI4 poly-kinetic code with an appropriate pre-processing and post-processing software called OVNI) are used in the case of 3-dimensional heterogeneous benchmarks (slab reactor cores) to reduce model biases and enable a thorough and detailed analysis of the performances of deterministic methods and their associated data libraries with respect to key neutron parameters (reactivity, local power). Outstanding examples of application of these tools are presented regarding the new numerical methods implemented in the French lattice code APOLLO2 (advanced self-shielding models, new IDT characteristics method implemented within the discrete-ordinates flux solver model) and the JEFF3.1 nuclear data library (checked against JEF2.2 previous file). In particular we have pointed out, by performing multigroup/point-wise TRIPOLI4 (assembly and core) calculations, the efficiency (in terms of accuracy and computation time) of the new IDT method developed in APOLLO2. In addition, by performing 3-dimensional TRIPOLI4 calculations of the whole slab core (few millions of elementary volumes), the high quality of the new JEFF3.1 nuclear data files and revised evaluations (U235, U238, Hf) for reactivity prediction of slab cores critical experiments has been stressed. As a feedback of the whole validation process, improvements in terms of nuclear data (mainly Hf capture cross-sections) and numerical methods (advanced quadrature formulas accounting validation results, validation of new self-shielding models, parallelization) are suggested to improve even more the APOLLO2-CRONOS2 standard calculation route. (author)

  5. 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.

  6. INDRA: a program system for calculating the neutronics and photonics characteristics of a fusion reactor blanket

    International Nuclear Information System (INIS)

    INDRA is a program system for calculating the neutronics and photonics characteristics of fusion reactor blankets. It incorporates a total of 19 different codes and 5 large data libraries. 10 of the codes are available from the code distribution organizations. Some of them, however, have been slightly modified in order to permit a convenient transfer of information from one program module to the next. The remaining 9 programs have been prepared by the authors to complete the system with respect to flexibility and to facilitate the handling of the results. (orig./WBU)

  7. Scattering cross sections of liquid deuterium for ultracold neutrons: Experimental results and a calculation model

    CERN Document Server

    Döge, Stefan; Müller, Stefan; Morkel, Christoph; Gutsmiedl, Erwin; Geltenbort, Peter; Lauer, Thorsten; Fierlinger, Peter; Petry, Winfried; Böni, Peter

    2015-01-01

    We present scattering cross sections $\\sigma_\\text{scatt}$ of ultracold neutrons (UCN) in liquid deuterium at T = 20.6 K, as recently measured by means of a transmission experiment. The indispensable thorough raw data treatment procedure is explained. A calculation model for coherent and incoherent scattering in liquid deuterium in the hydrodynamic limit based on appropriate physical concepts is provided and shown to ?t the data well. The applicability of the incoherent approximation for UCN scattering in liquid deuterium was tested and found to deliver acceptable results.

  8. Relativistic Hartree-Bogoliubov Calculation of Specific Heat of the Inner Crust of Neutron Stars

    OpenAIRE

    Nakano, Takuya; Matsuzaki, Masayuki

    2001-01-01

    We calculate the specific heat of the inner crust of neutron stars within a local-density approximation to an improved relativistic Hartree-Bogoliubov theory. Non-uniformness of the system enhances the specific heat in particular at low temperatures. The degree of enhancement is similar to that in the spherical phase of Elgar{\\o}y et al. We examine a schematic interpolation between the results of Broglia et al. adopting the Gogny force and ours based on the Lagrangian of the relativistic mean...

  9. Relativistic Hartree-Bogoliubov calculation of specific heat of the inner crust of neutron stars

    International Nuclear Information System (INIS)

    We calculate the specific heat of the inner crust of neutron stars within a local-density approximation to an improved relativistic Hartree-Bogoliubov theory. Non-uniformness of the system enhances the specific heat in particular at low temperatures. The degree of enhancement is similar to that in the spherical phase of Elgaroey et al. We examine a schematic interpolation between the results of Broglia et al. adopting the Gogny force and ours based on the Lagrangian of the relativistic mean field model. (author)

  10. Calculation of gamma-rays and fast neutrons fluxes with the program Mercure-4

    International Nuclear Information System (INIS)

    The program MERCURE-4 evaluates gamma ray or fast neutron attenuation, through laminated or bulky three-dimensionnal shields. The method used is that of line of sight point attenuation kernel, the scattered rays being taken into account by means of build-up factors for γ and removal cross sections for fast neutrons. The integration of the point kernel over the range of sources distributed in space and energy, is performed by the Monte-Carlo method, with an automatic adjustment of the importance functions. Since it is operationnal the program MERCURE-4 has been intensively used for many various problems, for example: - the calculation of gamma heating in reactor cores, control rods and shielding screens, as well as in experimental devices and irradiation loops; - the evaluation of fast neutron fluxes and corresponding damage in structural materials of reactors (vessel steels...); - the estimation of gamma dose rates on nuclear instrumentation in the reactors, around the reactor circuits and around spent fuel shipping casks

  11. 40 CFR 86.166-12 - Method for calculating emissions due to air conditioning leakage.

    Science.gov (United States)

    2010-07-01

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

  12. GEANT4 calculations of neutron dose in radiation protection using a homogeneous phantom and a Chinese hybrid male phantom.

    Science.gov (United States)

    Geng, Changran; Tang, Xiaobin; Guan, Fada; Johns, Jesse; Vasudevan, Latha; Gong, Chunhui; Shu, Diyun; Chen, Da

    2016-03-01

    The purpose of this study is to verify the feasibility of applying GEANT4 (version 10.01) in neutron dose calculations in radiation protection by comparing the calculation results with MCNP5. The depth dose distributions are investigated in a homogeneous phantom, and the fluence-to-dose conversion coefficients are calculated for different organs in the Chinese hybrid male phantom for neutrons with energy ranging from 1 × 10(-9) to 10 MeV. By comparing the simulation results between GEANT4 and MCNP5, it is shown that using the high-precision (HP) neutron physics list, GEANT4 produces the closest simulation results to MCNP5. However, differences could be observed when the neutron energy is lower than 1 × 10(-6) MeV. Activating the thermal scattering with an S matrix correction in GEANT4 with HP and MCNP5 in thermal energy range can reduce the difference between these two codes. PMID:26156875

  13. GEANT4 calculations of neutron dose in radiation protection using a homogeneous phantom and a Chinese hybrid male phantom

    International Nuclear Information System (INIS)

    The purpose of this study is to verify the feasibility of applying GEANT4 (version 10.01) in neutron dose calculations in radiation protection by comparing the calculation results with MCNP5. The depth dose distributions are investigated in a homogeneous phantom, and the fluence-to-dose conversion coefficients are calculated for different organs in the Chinese hybrid male phantom for neutrons with energy ranging from 1 x 10-9 to 10 MeV. By comparing the simulation results between GEANT4 and MCNP5, it is shown that using the high-precision (HP) neutron physics list, GEANT4 produces the closest simulation results to MCNP5. However, differences could be observed when the neutron energy is lower than 1 x 10-6 MeV. Activating the thermal scattering with an S matrix correction in GEANT4 with HP and MCNP5 in thermal energy range can reduce the difference between these two codes. (authors)

  14. A simple method to calculate the neutron flow through full ducts; une simple methode pour calculer le flux des neutrons a travers les canaux pleins

    Energy Technology Data Exchange (ETDEWEB)

    Faik Ouahab, Z.; Jehouani, A.; Ghassoun, J.; Senhou, N. [EPRA, Departement de Physique, Faculte de Sciences Semlalia, BP. 2390, Universite Cadi Ayyad, Marrakech (Morocco); Groetz, J.E. [Universite de Franche-Comte, Laboratoire de Chimie Physique et Rayonnements Alain Chambaudet, UMR CEA E4, 16, route de Gray, 25030 Besancon Cedex (France)

    2010-07-01

    Summary of a study of assessment of the probability for neutrons to be guided in a full duct with a square cross section and doubly bent. Two software have been developed, based on the Monte Carlo simulation, to compute the neutron transmission probability at the end of the duct. Results are in good agreement with that obtained with the MCNP-5 code. The neutron flow and probability at the duct end have been determined for different materials and different duct dimensions

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

    International Nuclear Information System (INIS)

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

  16. 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)

  17. 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.

  18. A development of NRESPG Monte Carlo code for the calculation of neutron response function for gas counters

    Energy Technology Data Exchange (ETDEWEB)

    Takeda, N. [Electrotechnical Laboratory, 1-1-4 Umezono, Tsukuba-shi, Ibaraki 305-8568 (Japan); Kudo, K. [Electrotechnical Laboratory, 1-1-4 Umezono, Tsukuba-shi, Ibaraki 305-8568 (Japan); Toyokawa, H. [Electrotechnical Laboratory, 1-1-4 Umezono, Tsukuba-shi, Ibaraki 305-8568 (Japan); Torii, T. [Japan Power Reactor and Nuclear Fuel Development Corporation, Tsuruga Office, Fukui 919-12 (Japan); Hashimoto, M. [Japan Power Reactor and Nuclear Fuel Development Corporation, O-arai Engineering Center, Ibaraki 311-13 (Japan); Sugita, T. [Science System Laboratory, Ibaraki 309-17 (Japan); Dietze, G. [Physikalisch-Technische Bundesanstalt, 38023 Braunschweig (Germany); Yang, X. [China Institute of Atomic Energy (China)

    1999-02-11

    A Monte Carlo code Neutron RESPonse function for Gas counters (NRESPG) has been developed for the calculation of neutron response functions and efficiencies for neutron energies up to 20 MeV, which can be applied for {sup 3}He, H{sub 2}, or BF{sub 3} gas proportional counters with or without moderator. This code can simulate the neutron behavior in a two-dimensional detector configuration and treat the thermal motion of a moderator atom which becomes important as the neutron energy becomes sufficiently low. Further, a more precise measured data was taken to simulate the position-dependent gas multiplication in the sensitive and insensitive gas region of a proportional counter. The NRESPG code has been applied for the calculation of response functions of {sup 3}He cylindrical proportional counters to determine neutron energy and neutron fluence in a monoenergetic calibration field. Thus, a remarkable discrepancy in the lower portion of the full-energy peak produced by the {sup 3}He(n,p)T reaction can be removed which results in a good agreement between simulations and experiments. The code has been also used for the simulation of the response of a McTaggart-type long counter consisting of a central cylindrical BF{sub 3} counter surrounded by a polyethylene moderator. The results of the NRESPG simulations were compared with those obtained from MCNP calculations.

  19. A development of NRESPG Monte Carlo code for the calculation of neutron response function for gas counters

    International Nuclear Information System (INIS)

    A Monte Carlo code Neutron RESPonse function for Gas counters (NRESPG) has been developed for the calculation of neutron response functions and efficiencies for neutron energies up to 20 MeV, which can be applied for 3He, H2, or BF3 gas proportional counters with or without moderator. This code can simulate the neutron behavior in a two-dimensional detector configuration and treat the thermal motion of a moderator atom which becomes important as the neutron energy becomes sufficiently low. Further, a more precise measured data was taken to simulate the position-dependent gas multiplication in the sensitive and insensitive gas region of a proportional counter. The NRESPG code has been applied for the calculation of response functions of 3He cylindrical proportional counters to determine neutron energy and neutron fluence in a monoenergetic calibration field. Thus, a remarkable discrepancy in the lower portion of the full-energy peak produced by the 3He(n,p)T reaction can be removed which results in a good agreement between simulations and experiments. The code has been also used for the simulation of the response of a McTaggart-type long counter consisting of a central cylindrical BF3 counter surrounded by a polyethylene moderator. The results of the NRESPG simulations were compared with those obtained from MCNP calculations

  20. 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.

  1. Calculation of conversion coefficients Hp(3)/K air using the PENELOPE Monte Carlo code and comparison with MCNP calculation results

    International Nuclear Information System (INIS)

    The authors report calculations performed using the MNCP and PENELOPE codes to determine the Hp(3)/K air conversion coefficient which allows the Hp(3) dose equivalent to be determined from the measured value of the kerma in the air. They report the definition of the phantom, a 20 cm diameter and 20 cm high cylinder which is considered as representative of a head. Calculations are performed for an energy range corresponding to interventional radiology or cardiology (20 keV-110 keV). Results obtained with both codes are compared

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

    CERN Document Server

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

    2002-01-01

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

  3. Different approach to calculating average angular distributions of elastically scattered neutrons in the resonance region

    International Nuclear Information System (INIS)

    A relatively simple formalism for calculating the average neutron elastic angular distribution dσel/dΩ in the resonance region below several hundred keV is presented. The expression for dσel/dΩ depends mainly on the R-matrix parameters S0, R', S1, and R1∞. Comparisons between calculated and experimental angular distributions are presented for 103Rh, 139La, 232Th, and 238U. A fit to 238U data at 75 keV led to a value of the p-wave strength function of S1=1.81±0.35x10-4. Except for measuring a complete set of individual l=1 resonances, determining the p-wave strength function by fitting low-energy angular distributions is probably more reliable than, or competitive with, other techniques which are available. An analysis of elastic angular distributions as a function of neutron energy is also well suited to a search for intermediate structure in the s- or p-wave strength function. copyright 1997 The American Physical Society

  4. WETAIR: A computer code for calculating thermodynamic and transport properties of air-water mixtures

    Science.gov (United States)

    Fessler, T. E.

    1979-01-01

    A computer program subroutine, WETAIR, was developed to calculate the thermodynamic and transport properties of air water mixtures. It determines the thermodynamic state from assigned values of temperature and density, pressure and density, temperature and pressure, pressure and entropy, or pressure and enthalpy. The WETAIR calculates the properties of dry air and water (steam) by interpolating to obtain values from property tables. Then it uses simple mixing laws to calculate the properties of air water mixtures. Properties of mixtures with water contents below 40 percent (by mass) can be calculated at temperatures from 273.2 to 1497 K and pressures to 450 MN/sq m. Dry air properties can be calculated at temperatures as low as 150 K. Water properties can be calculated at temperatures to 1747 K and pressures to 100 MN/sq m. The WETAIR is available in both SFTRAN and FORTRAN.

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

    International Nuclear Information System (INIS)

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

  6. Calculation of neutron and gamma-ray energy spectra for fusion reactor shield design: comparison with Experiment II

    International Nuclear Information System (INIS)

    Measured and calculated neutron and gamma-ray energy spectra resulting from the transport of approx. 14 MeV neutrons through a 0.30-m-thick lithium hydride slab and through a 0.05-m-thick lead slab followed by 0.30 m of lithium hydride are compared. Also reported are comparisons of the measured and calculated neutron energy spectra behind an 0.80-m-thick assembly comprised of stainless steel type 304 and borated polyethylene. The spatial dependence of the gamma-ray energy deposition rate measured using thermoluminescent detectors is compared with calculated data. The calculated data obtained using two-dimensional radiation transport methods and ENDF/B-IV cross section data are in good agreement for all of the experimental configurations

  7. Neutron activation analysis application to the study of air pollution bio monitors

    International Nuclear Information System (INIS)

    Full text: This work has been done within the IAEA Research Contract Arg 9929, Research Co-ordinated Programme on Validation and application of plants as bio monitors of trace-element atmospheric pollution, analysed using nuclear and related techniques. Knowledge on air pollution levels and identification of polluted areas and potential emission sources are of increasing concern all over the world. Chemical characterisation of atmospheric aerosol, especially its heavy metal contents, is therefore of great importance and neutron activation analysis is a powerful technique for its determination. The advantages of using bio monitors instead of direct sampling lies not only on its lower cost but also on the possibility of using them to measure and/or evaluate deposition over large areas. The general objective of this project is the use of lichen to evaluate pollution levels in an area of Cordoba province (Argentina) and to establish baseline levels and temporal trends and draw distribution maps of pollutants. Based on lichen distribution maps, two species were selected: Raumalina ecklonii and Usnea amblyoclada. Different tests were done to adjust sample preparation methodologies previous to irradiation. The tests included grinding and drying assays to investigate their influence on the following determination using NAA. Sample grinding with and without the addition of liquid nitrogen was tried and oven-dry and freeze-dry were tried on samples of the two selected species. Elemental determination was done using instrumental Neutron Activation Analysis. Samples were irradiated for 5 hours at the RA-3 reactor of the Ezeiza Atomic Center (thermal flux 3.1013cm-2-2.s-1-1, 4.5 M w), and measured twice with different decay times 86 and 30 days) for the determination of medium and long-lived nuclides. The measurements were done using GeHP detectors (30 % efficiency, resolution 1.9 keV for 6060Co 1332.5 keV peak) coupled to a Canberra Series 85 multichannel analyser

  8. 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 $\\AA$.

  9. Calculating the neutron yield from thick targets irradiated by electrons with the energy upto 500 MeV

    International Nuclear Information System (INIS)

    The mathematical simulation technique used for calculating the photoneutron yield from thick targets made of different Materials is suggested. Cascade-evaporative nucleus model being a part of the IMITATOR program complex is used for calculations. Three groups of materials are investigated: light-oxygen and aluminium, medium- iron and nickel, heavy,tungsten and lead. Maximum thickness of targets consistuting 10 radiation length is determined on the basis of the experiment and represents the thickness at which in the investigated energy range secondary neutron flux ''saturation'' arises. The dependences of total neutron yield on electron beam energy and target material are obtained. The values of fast neutrons yield from thick targets, their spatial distribution and dependences on the energy of primary electrons and target thickness are determined. Anomalies of photoneutrons yield near magic and double magic nuclei are pointed out. A considerable drop of total yield of fast neutrons with increase of atomic number of target material is noted

  10. Validation of Three-Dimensional Synthesis RPV Neutron Fluence Calculations Using VVER-1000 Ex-Vessel Reference Dosimetry Results

    International Nuclear Information System (INIS)

    According to Russian federal norms and the safety guide of the nuclear regulatory body of Russia, the maximum fast neutron fluence above 0.5 MeV at critical positions of the reactor pressure vessel (RPV) of VVER-type reactors is used for prediction of the RPV lifetime. For the computation of neutron fluences in the RPV near the reactor core midplane level, the three-dimensional (3-D) synthesis method based on two- and one-dimensional SN calculations may be acceptable but needs validation. The present validation analysis was carried out on the basis of neutron transport calculations for a VVER-1000 model by means of the well-known codes DORT (R, Θ- and R, Z geometry) and ANISN (R geometry) using the multigroup library BUGLE-96. The 3-D spatial neutron source distribution, including pin-to-pin power variations and the complex baffle construction, were modeled in detail

  11. Neutron Thermal Cross Sections, Westcott Factors, Resonance Integrals, Maxwellian Averaged Cross Sections and Astrophysical Reaction Rates Calculated from Major Evaluated Data Libraries

    OpenAIRE

    Pritychenko, B.; Mughabghab, S.F.

    2012-01-01

    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-...

  12. 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.

  13. 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

  14. Results of the MUSE Benchmark - 3D distribution of reaction yields and delayed neutrons in criticality calculations

    International Nuclear Information System (INIS)

    The MUSE project, carried out within the European fifth Framework Program, focuses on the coupling of a sub-critical reactor core with an external neutron source. In the first stage of the project a benchmark has been defined in order to define a reference calculational route, which is able to accurately predict the neutronics behavior in an accelerator driven system. Benchmark calculations will be carried out by several members of the project and the results will be compared, also with experimental results. The contribution of NRG to the project consists of the benchmark calculations and additional work that focuses on the calculation of 3D distributions of reaction yields. This paper discusses the non-conventional methods used to perform the benchmark calculations, including the 3D reaction yield distributions. The 3D distributions calculated for the sub-critical core will be Shown and discussed. With the ORANGE-extension to MCNP it is possible to tally 3D distributions, without adding extra cells and surfaces to the geometry and without a significant slowing down of the calculation. These are major advantages when compared to the conventional way of tallying in the MCNP-code. The distributions show details that can be understood in terms of the expected neutron behavior in the different parts of the geometry. For instance, the results show that: 1) a large number of fast neutrons is found in the fuel regions, 2) the reflector region shows an increased number of slower neutrons and 3) the reaction yield in the shielding region declines steeply. The extension therefore seems a useful tool in generating a better understanding of the behavior of neutrons throughout large and complex geometries like accelerator driven systems, but we also expect to use the extension in a variety of different fields. (authors)

  15. 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 ...

  16. 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.

  17. Recent advances in neutron tomography

    International Nuclear Information System (INIS)

    Neutron imaging has been shown to be an excellent imaging tool for many nondestructive evaluation applications. Significantly improved contrast over X-ray images is possible for materials commonly found in engineering assemblies. The major limitations have been the neutron source and detection. A low cost, position sensitive neutron tomography detector system has been designed and built based on an electro-optical detector system using a LiF-ZnS scintillator screen and a cooled charge coupled device. This detector system can be used for neutron radiography as well as two and three-dimensional neutron tomography. Calculated performance of the system predicted near-quantum efficiency for position sensitive neutron detection. Experimental data was recently taken using this system at McClellan Air Force Base, Air Logistics Center, Sacramento, CA. With increased availability of low cost neutron sources and advanced image processing, neutron tomography will become an increasingly important nondestructive imaging method

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

    International Nuclear Information System (INIS)

    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 241Am-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. (author)

  19. Analysis of calculated neutron flux response at detectors of G.A. Siwabessy multipurpose reactor (RSG-GAS Reactor)

    International Nuclear Information System (INIS)

    Multi Purpose Reactor G.A. Siwabessy (RSG-GAS) reactor core possesses 4 fission-chamber detectors to measure intermediate power level of RSG-GAS reactor. Another detector, also fission-chamber detector, is intended to measure power level of RSG-GAS reactor. To investigate influence of space to the neutron flux values for each detector measuring intermediate and power levels has been carried out. The calculation was carried out using combination of WIMS/D4 and CITATION-3D code and focused on calculation of neutron flux at different detector location of RSG-GAS typical working core various scenarios. For different scenarios, all calculation results showed that each detector, located at different location in the RSG-GAS reactor core, causes different neutron flux occurred in the reactor core due to spatial time effect

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

    International Nuclear Information System (INIS)

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

  1. Monte Carlo methodologies for neutron streaming in diffusion calculations - Application to directional diffusion coefficients and leakage models in XS generation

    OpenAIRE

    Dorval, Eric

    2016-01-01

    Neutron transport calculations by Monte Carlo methods are finding increased application in nuclear reactor simulations. In particular, a versatile approach entails the use of a 2-step pro-cedure, with Monte Carlo as a few-group cross section data generator at lattice level, followed by deterministic multi-group diffusion calculations at core level. In this thesis, the Serpent 2 Monte Carlo reactor physics burnup calculation code is used in order to test a set of diffusion coefficient model...

  2. Neutronic calculation to the TRIGA Ipr-R1 reactor using the WIMSD4 and CITATION codes

    International Nuclear Information System (INIS)

    The WIMSD4 and CITATION codes are used to calculate neutronic parameters of a TRIGA reactor. The results are compared with experimental values. Five configurations are analysed and the excess reactivity worth, the fuel temperature reactivity coefficient, the control reactivity worth, safety and regulation rod of the TRIGA IPR-R1 reactor are calculated. The idea is to obtain the systematic error for k∞ for this methodology comparing the calculated and the experimental results

  3. Characterisation of air particulate matter in Klang Valley by neutron activation analysis technique

    International Nuclear Information System (INIS)

    Air particulate matter is known to affect human health, impairs visibility and can cause climate change. Study on air particulate matter in term of particle size and chemical contents is very important to indicate the quality of air in a sampling area. Information on concentration of important constituents in air particles can be used to identify some of emission sources which contribute to the pollution problem. The data collected may also be, used as a basis to design a strategy in order to overcome the air pollution problem in the area. The study involved sampling of air dust at two stations, one in Bangi and the other in Kuala Lumpur using Gent Stack Sampler units. Each sampler capable of collecting air particle sizes smaller than 2.5 micron (PM 2.5) and between 2.5 - O micron on two different filters simultaneously. The filters were measured for their mass, elemental carbon and elemental concentrations using analytical equipment or techniques including reflectometer and Neutron Activation Analysis. The results of analysis on samples collected in 1997-1998 are discussed. (author)

  4. 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.

  5. 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...

  6. Monte Carlo calculation of 60Co γ-ray's albedo-dose rate from the air

    International Nuclear Information System (INIS)

    The Monte Carlo calculation of 60Co γ-ray's albedo-dose rate from the air is reported. A formula is presented with which the relations of the albedo-doserate with some parameters are simulated and fitted

  7. Time-resolved Fast Neutron Radiography of Air-water Two-phase Flows

    Science.gov (United States)

    Zboray, Robert; Dangendorf, Volker; Mor, Ilan; Tittelmeier, Kai; Bromberger, Benjamin; Prasser, Horst-Michael

    Neutron imaging, in general, is a useful technique for visualizing low-Z materials (such as water or plastics) obscured by high-Z materials. However, when significant amounts of both materials are present and full-bodied samples have to be examined, cold and thermal neutrons rapidly reach their applicability limit as the samples become opaque. In such cases one can benefit from the high penetrating power of fast neutrons. In this work we demonstrate the feasibility of time-resolved, fast neutron radiography of generic air-water two-phase flows in a 1.5 cm thick flow channel with Aluminum walls and rectangular cross section. The experiments have been carried out at the high-intensity, white-beam facility of the Physikalisch-Technische Bundesanstalt, Germany. Exposure times down to 3.33 ms have been achieved at reasonable image quality and acceptable motion artifacts. Different two-phase flow regimes such as bubbly slug and churn flows have been examined. Two-phase flow parameters like the volumetric gas fraction, bubble size and bubble velocities have been measured.

  8. Air pollution assessment in two Moroccan cities using instrumental neutron activation analysis on bio-accumulators

    International Nuclear Information System (INIS)

    Full text: Biomonitoring is an appropriate tool for the air pollution assessment studies. In this work, lichens and barks have been used as bio-accumulators in several sites in two Moroccan cities (Rabat and Mohammadia). The specific ability of absorbing and accumulating heavy metals and toxic element from the air, their longevity and resistance to the environmental stresses, make those bioindicators suitable for this kind of studies. The Instrumental Neutron Activation Analysis (INAA) is universally accepted as one of the most reliable analytical tools for trace and ultra-trace elements determination. Its use in trace elements atmospheric pollution related studies has been and is still extensive as can be demonstrated by several specific works and detailed reviews. In this work, a preliminary investigation employing lichens, barks and instrumental neutron activation analysis (INAA) was carried out to evaluate the trace elements distribution in six different areas of Rabat and Mohammadia cities characterised by the presence of many industries and heavy traffic. Samples were irradiated with thermal neutrons in a nuclear reactor and the induced activity was counted using high-resolution Germanium-Lithium detectors. More than 30 elements were determined using two modes : short irradiation (1 minute) and long irradiation (17 hours). Accuracy and quality control were assessed using the reference standard material IAEA-336. This was less than 1% for major and about 5 to 10% for traces.

  9. 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)

  10. 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.

  11. Calculating the effective delayed neutron fraction in the Molten Salt Fast Reactor: Analytical, deterministic and Monte Carlo approaches

    International Nuclear Information System (INIS)

    Highlights: • Calculation of effective delayed neutron fraction in circulating-fuel reactors. • Extension of the Monte Carlo SERPENT-2 code for delayed neutron precursor tracking. • Forward and adjoint multi-group diffusion eigenvalue problems in OpenFOAM. • Analytical approach for βeff calculation in simple geometries and flow conditions. • Good agreement among the three proposed approaches in the MSFR test-case. - Abstract: This paper deals with the calculation of the effective delayed neutron fraction (βeff) in circulating-fuel nuclear reactors. The Molten Salt Fast Reactor is adopted as test case for the comparison of the analytical, deterministic and Monte Carlo methods presented. The Monte Carlo code SERPENT-2 has been extended to allow for delayed neutron precursors drift, according to the fuel velocity field. The forward and adjoint eigenvalue multi-group diffusion problems are implemented and solved adopting the multi-physics tool-kit OpenFOAM, by taking into account the convective and turbulent diffusive terms in the precursors balance. These two approaches show good agreement in the whole range of the MSFR operating conditions. An analytical formula for the circulating-to-static conditions βeff correction factor is also derived under simple hypotheses, which explicitly takes into account the spatial dependence of the neutron importance. Its accuracy is assessed against Monte Carlo and deterministic results. The effects of in-core recirculation vortex and turbulent diffusion are finally analysed and discussed

  12. 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)

  13. Neutronics calculation of a reactor cold neutron source%反应堆冷中子源中子物理学计算

    Institute of Scientific and Technical Information of China (English)

    胡春明; 余朝举; 童剑飞

    2011-01-01

    用MCNP软件计算反应堆冷中子源,慢化剂室内平均中子注量率为6.69× 1013/cm-2.s-1,波长为0.4 nm和0.6 nm的冷中子增益因子~16和32.冷源慢化剂中正仲氢比例对输出的冷中子能谱有较大影响,而在3K范围内慢化剂温度变化对冷中子能谱的影响很小.计算结果表明,冷中子源性能达到基本设计要求.%The construction of a reactor cold neutron source (CNS) will be completed in the near future. To evaluate performance of the CNS, a neutronics calculation using MCNP4C code has been carried out. The results show that the average neutron flux in the moderator is 6.69×1013/cm2·s, and the cold neutron gain factors corresponding to 4-(A) and 6-(A) wavelengths are 16 and 32, respectively. The results also indicate that different ratios of ortho-H2/para-H2 have an obvious impact on cold neutron spectrum in the moderator, but within 3 K of the moderator temperature changes, the spectrum varies slightly.

  14. FLUENT-based neutronics and thermal-hydraulics coupling calculation for a liquid-fuel molten salt reactor

    International Nuclear Information System (INIS)

    Molten Salt Reactor (MSR) is the only one using liquid fuel in the six candidate reactors of the Generation IV advanced nuclear power systems with expected remarkable advantages in safety, economics, sustainability, and proliferation resistance. The strong coupling between neutronics and thermal-hydraulics due to fuel movement in the liquid-fuel MSRs induces many new challenges in reactor analyses from the perspective of both theoretical models and solution methods. In this study, the multi-group diffusion theory was adopted to deduce the neutronics model for the liquid-fuel MSRs, in which the salt flow effects on the delayed neutron precursor distributions in space were considered particularly. Since the liquid-fuel salt is a Newton fluid, the single-phase thermal hydraulics model for liquid-fuel MSRs was generally established based on the fundamental laws of the mass, momentum and energy conservation equations as used in the computational fluid dynamic (CFD) method. Since the control equations of the neutronic model can be written in the same form of those solved in the CFD softwares, a neutronics and thermal-hydraulics coupling scheme was proposed and a program was developed based on the FLUENT software by using its user-defined functions and subroutines (UDF and UDS). This program was applied to perform the steady state calculation of the molten salt fast reactor (MSFR), and the main results such as the space distributions of the neutron fluxes, delayed neutron precursors, temperatures, velocities were obtained. The results show that the liquid fuel flow influences the delayed neutron precursors significantly, while slightly affects the neutron fluxes. The flow in the MSFR core generates a vortex near the fertile tank leading the maximal temperature to about 1200 K at the centre of the vortex, which will be optimized in the future core design. (author)

  15. 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.

  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. Improvement of the neutron flux calculations in thick shield by conditional Monte Carlo and deterministic methods

    Energy Technology Data Exchange (ETDEWEB)

    Ghassoun, Jillali; Jehoauni, Abdellatif [Nuclear physics and Techniques Lab., Faculty of Science, Semlalia, Marrakech (Morocco)

    2000-01-01

    In practice, the estimation of the flux obtained by Fredholm integral equation needs a truncation of the Neuman series. The order N of the truncation must be large in order to get a good estimation. But a large N induces a very large computation time. So the conditional Monte Carlo method is used to reduce time without affecting the estimation quality. In a previous works, in order to have rapid convergence of calculations it was considered only weakly diffusing media so that has permitted to truncate the Neuman series after an order of 20 terms. But in the most practical shields, such as water, graphite and beryllium the scattering probability is high and if we truncate the series at 20 terms we get bad estimation of flux, so it becomes useful to use high orders in order to have good estimation. We suggest two simple techniques based on the conditional Monte Carlo. We have proposed a simple density of sampling the steps for the random walk. Also a modified stretching factor density depending on a biasing parameter which affects the sample vector by stretching or shrinking the original random walk in order to have a chain that ends at a given point of interest. Also we obtained a simple empirical formula which gives the neutron flux for a medium characterized by only their scattering probabilities. The results are compared to the exact analytic solution, we have got a good agreement of results with a good acceleration of convergence calculations. (author)

  18. Improvement of the neutron flux calculations in thick shield by conditional Monte Carlo and deterministic methods

    International Nuclear Information System (INIS)

    In practice, the estimation of the flux obtained by Fredholm integral equation needs a truncation of the Neuman series. The order N of the truncation must be large in order to get a good estimation. But a large N induces a very large computation time. So the conditional Monte Carlo method is used to reduce time without affecting the estimation quality. In a previous works, in order to have rapid convergence of calculations it was considered only weakly diffusing media so that has permitted to truncate the Neuman series after an order of 20 terms. But in the most practical shields, such as water, graphite and beryllium the scattering probability is high and if we truncate the series at 20 terms we get bad estimation of flux, so it becomes useful to use high orders in order to have good estimation. We suggest two simple techniques based on the conditional Monte Carlo. We have proposed a simple density of sampling the steps for the random walk. Also a modified stretching factor density depending on a biasing parameter which affects the sample vector by stretching or shrinking the original random walk in order to have a chain that ends at a given point of interest. Also we obtained a simple empirical formula which gives the neutron flux for a medium characterized by only their scattering probabilities. The results are compared to the exact analytic solution, we have got a good agreement of results with a good acceleration of convergence calculations. (author)

  19. 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.

  20. Sensitivity of the scalar and the angular neutron flux to the anisotropic scattering in Monte Carlo calculations

    International Nuclear Information System (INIS)

    The anisotropic scattering influences both the transport and slowing processes of neutrons. Since Practical shields are usually anisotropic scatters, several parameterized, anisotropic scattering kernels were used to present a general class of anisotropies. To study the anisotropic sensitivity of the flux in thick shield medium, the feasibility of track- length distribution biasing for calculations of scalar and angular neutron flux and their sensitivity to anisotropic scattering was investigated. To represent more realistic angular distribution in a parameterized functional form, an exponential angular density for sampling the scattering angle is proposed, also an empirical formula for the choice of optimal parameter for track length biasing depending on the anisotropic scattering is proposed. The calculations are performed for a particle transport model having an exact solution. The results show that this distribution covers also isotropic and the anisotropic scattering case. The anisotropic effect has a great influence on the behavior of neutron distribution particularly in thick shield. (author)

  1. Neutronic calculations of hexagonal lattice nuclear reactors: Modelling of the CAREM-25 reactor

    International Nuclear Information System (INIS)

    This work was carried out in the frame of the Cnea CAREM-25 project (Central Argentina de Elementos Modulares).This project involves the development and construction of an argentinian design nuclear reactor for producing electricity. It's a PWR type (light water moderated and enriched U02 fueled) integrated reactor in an hexagonal lattice.The total power of this prototype is 100 MW thermal. In this frame, the main objective of this work is to consolidate and validate a neutronic line of calculus which can be applied to the CAREM-25 core.At a first analysis at cell level, the different fuel elements were modeled with the Dragon code, obtaining homogenised and condensed cross sections.Then a core level analysis with the Puma code was performed at full power condition and room temperature. A comparison of the obtained results is needed.For this reason, a Monte Carlo analysis (at room temperature) was performed.Also a validation of the Dragon code was carried out on the base of experimental data of WWER type lattices (similars to CAREM).The confidence on the results is then granted and their uncertainties were quantified.The Dragon-Puma line of calculus is then established and the main objective of this work is achieved. A full neutronic analysis should be followed by thermohydraulics calculations in an iterative procedure, and it would be the objective of future works.Finally, a burnup analysis was performed, at cell and core level.The design condition for extraction burnup and fuel cycle duration were verified.

  2. ZZ ALBEDO-DATA, Data for the Calculation of Albedos from Concrete, Iron, Lead and Water for Photons and Neutrons

    International Nuclear Information System (INIS)

    1 - Description: The use of albedo techniques is central to many radiation streaming codes and has been widely used as an alternative to much more expensive transport calculations. Key to the albedo technique is the availability of either a large set of albedo data or, preferably, an empirical formula that approximates the albedo over the range of source energies and incident and exit radiation directions involved in a particular problem. Previously proposed neutron and photon albedo approximating formulas have been based on limited energy-angular ranges, a single reflecting material, old cross section data, and, most important, obsolete fluence-to-dose response functions. This library contains differential neutron dose albedo functions, based on modern cross section and response function data. Newly evaluated parameters are tabulated for several empirical differential dose albedo formulas. The albedos considered are (1) two approximations for photon albedo, (2) a new approximation for the neutron albedo, and (3) the secondary-photon albedo for incident neutrons. Albedo data is provided for four Materials: concrete, iron, lead, and water. Unlike previous compilations of albedo data, modern dosimetric units have been employed. Data are presented for (1) the ambient dose equivalent H*(10 mm) and (2) the effective dose equivalent for anteroposterior (AP) illumination of the ICRP anthropomorphic phantom. 2 - Methods: Monte Carlo code, MCNP, was used to calculate the albedo reflected from thick slabs of various materials. In particular, a homogeneous cylindrical slab surrounded by a vacuum was used. The incident neutrons were modeled by a point mono-directional source positioned just inside the center of the circular scoring (reflecting) surface. This was done to facilitate scoring because all particles crossing the surface must be outgoing (reflected) particles. Slab thickness and radius were sufficiently large (1000 cm) so that negligible numbers of neutrons were

  3. Improvement on the calculation of D2O moderated critical systems with new thermal neutron scattering libraries

    International Nuclear Information System (INIS)

    Highlights: • We analize the performance of neutron scattering libraries for D and O in D2O for nuclear criticality calculations. • We calculated 65 ICSBEP benchmark cases from 8 heavy water moderated thermal systems using MCNP5. • A significant improvement is found when our library is combined with the ROSFOND-2010 evaluation for deuterium. • In 48 of the 65 benchmark cases we obtained a C/E ratio closer to 1.0. • The percentage of benchmark cases calculated within 1-sigma increases from 42% to 82%, compared to ENDF/B-VII calculations. - Abstract: To improve the evaluations in thermal sublibraries, we developed a set of thermal neutron scattering cross sections (scattering kernels) for the deuterium and oxygen bound in heavy water in the ENDF-6 format. These new libraries are based on molecular dynamics simulations and recent experimental data, and result in an improvement of the calculation of single neutron scattering quantities. In this work, we show how the use of this new set of cross sections also improves the calculation of thermal critical systems moderated and/or reflected with heavy water. The use of the new thermal scattering library for heavy water, combined with the ROSFOND-2010 evaluation of the deuterium cross sections, results in an improvement of the C/E ratio in 48 out of 65 benchmark cases calculated with the Monte Carlo code MCNP5, in comparison with the existing library based on the ENDF/B-VII evaluation

  4. 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.

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

    International Nuclear Information System (INIS)

    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/(cm2 s) (neutrons above 500 keV are 53% of total) whereas total fluency 2.53E18 cm−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 58Co and 54Mn (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 51Cr, 0.93–1.21 for 54Mn, 0.77–0.98 for 57Co, 0.91–1.21 for 58Co, 1.17–1.27 for 59Fe, and 1.75–2.44 for 60Co

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

    International Nuclear Information System (INIS)

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

  7. 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).

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

    NARCIS (Netherlands)

    Jacimovic, R; Maucec, M; Trkov, A

    2003-01-01

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

  9. Status review of methods for the calculation of fast neutron nuclear data for structural materials of fast and fusion reactors

    International Nuclear Information System (INIS)

    The report contains the texts of the 9 invited papers delivered during the Second Research Co-ordination Meeting on ''Methods for the Calculation of Fast Neutron Nuclear Data for Structural Materials and Fast and Fusion Reactors'' held in Vienna during 15-17 February 1988. A separate abstract was prepared for each of these 9 papers. Refs, figs and tabs

  10. Comparison of calculated and measured spectral response and intrinsic efficiency for a boron-loaded plastic neutron detector

    International Nuclear Information System (INIS)

    Boron-loaded scintillators offer the potential for neutron spectrometers with a simplified, peak-shaped response. The Monte Carlo code, MCNP, has been used to calculate the detector characteristics of a scintillator made of a boron-loaded plastic, BC454, for neutrons between 1 and 7 MeV. Comparisons with measurements are made of spectral response for neutron energies between 4 and 6 MeV and of intrinsic efficiencies for neutrons up to 7 MeV. In order to compare the calculated spectra with measured data, enhancements to MCNP were introduced to generate tallies of light output spectra for recoil events terminating in a final capture by 10B. The comparison of measured and calculated spectra shows agreement in response shape, full width at half maximum, and recoil energy deposition. Intrinsic efficiencies measured to 7 MeV are also in agreement with the MCNP calculations. These results validate the code predictions and affirm the value of MCNP as a useful tool for development of sensor concepts based on boron-loaded plastics. (orig.)

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

    CERN Document Server

    Routti, J T

    1975-01-01

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

  12. VSOP, Neutron Spectra, 2-D Flux Synthesis, Fuel Management, Thermohydraulics Calculation

    International Nuclear Information System (INIS)

    1 - Description of problem or function: VSOP (Very Superior Old Programs) is a system of codes linked together for the simulation of reactor life histories. It comprises neutron cross section libraries and processing routines, repeated neutron spectrum evaluation, 2-D and 3-D diffusion calculation, depletion and shut-down features, in- core and out-of-pile fuel management, fuel cycle cost analysis, and thermal hydraulics (steady state and transient). Various techniques have been employed to accelerate the iterative processes and to optimize the internal data transfer. The code system has been used extensively for comparison studies of thermal reactors, their fuel cycles, thermal transients, and safety assessment. Besides its use in research and development work for the Gas Cooled High Temperature Reactor, the system has been applied successfully to Light Water and Heavy Water Reactors, MAGNOX, and RBMK. 2 - Method of solution: The nuclear data for 184 isotopes are contained in two libraries. Fast and epithermal data in a 68 group GAM-I structure have been prepared mainly from ENDF/B-V and JEF-1. Resonance cross section data are given as input. Thermal data in a 30 group THERMOS structure have been collapsed from a 96 group THERMALIZATION (GATHER) library by a relevant neutron energy spectrum generated by the THERMALIZATION code. Graphite scattering matrices are based on the Young phonon spectrum in graphite. The neutron spectrum is calculated by a combination of the GAM and THERMOS codes. They can simultaneously be employed for many core regions differing in temperature, burnup, and fuel element lay-out. The thermal cell code THERMOS has been extended to treat the grain structure of the coated particles inside the fuel elements, and the epithermal GAM code uses modified cross sections for the resonance absorbers prepared from double heterogeneous ZUT-DGL calculations. The diffusion module of the code is CITATION with 2 - 8 energy groups. It provides the neutron

  13. Neutron flux calculation and fluence in the encircling of the core and vessel of a reactor BWR

    International Nuclear Information System (INIS)

    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-θ 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-θ, 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, θ 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/cm2s, at a height H 4 (239.07 cm) and angle 32.236o in the core shroud and 4.00 E + 09 n/cm2s at a height H 4 and angle 35.27o in the inner wall of the reactor vessel, positions that are consistent to within ±10% over the ones reported in the literature. (Author)

  14. A simple method to calculate the neutron flow through full ducts

    International Nuclear Information System (INIS)

    Summary of a study of assessment of the probability for neutrons to be guided in a full duct with a square cross section and doubly bent. Two software have been developed, based on the Monte Carlo simulation, to compute the neutron transmission probability at the end of the duct. Results are in good agreement with that obtained with the MCNP-5 code. The neutron flow and probability at the duct end have been determined for different materials and different duct dimensions

  15. Neutron spectrum calculation and safety analysis for supercritical water-cooled reactor

    International Nuclear Information System (INIS)

    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. (authors)

  16. Neutronic performance calculations with alternative fluids in a hybrid reactor by using the Monte Carlo method

    Energy Technology Data Exchange (ETDEWEB)

    Guenay, Mehtap [Malatya Univ. (Turkey). Physics Department

    2015-03-15

    In this study, salt-heavy metal mixtures consisting of 93-85% Li{sub 20}Sn{sub 80} + 5% SFG-PuO{sub 2} and 2-10% UO{sub 2}, 93-85% Li{sub 20}Sn{sub 80} + 5% SFG-PuO{sub 2} and 2-10% NpO{sub 2}, and 93-85% Li{sub 20}Sn{sub 80} + 5% SFG-PuO{sub 2} and 2-10% UCO were used as fluids. The fluids were used in the liquid first wall, blanket, and shield zones of a fusion-fission hybrid reactor system. A beryllium (Be) zone with a width of 3 cm was used for neutron multiplicity between the liquid first wall and the blanket. 9Cr2WVTa ferritic steel with the width of 4 cm was used as the structural material. The contributions of each isotope in the fluids to the nuclear parameters, such as tritium breeding ratio (TBR), energy multiplication factor (M), and heat deposition rate, of the fusion-fission hybrid reactor were calculated in the liquid first wall, blanket, and shield zones. Three-dimensional analyses were performed using the Monte Carlo code MCNPX-2.7.0 and nuclear data library ENDF/B-VII.0.

  17. 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.

  18. Review of Methods for Calculating Pressure Profiles of Explosive Air Blast and its Sample Application

    OpenAIRE

    Chock, Jeffrey Mun Kong

    1999-01-01

    Blast profiles and two primary methods of determining them were reviewed for use in the creation of a computer program for calculating blast pressures which serves as a design tool to aid engineers or analysts in the study of structures subjected to explosive air blast. These methods were integrated into a computer program, BLAST.F, to generate air blast pressure profiles by one of these two differing methods. These two methods were compared after the creation of the program and can conserv...

  19. Computer calculation of neutron cross sections with Hauser-Feshbach code STAPRE incorporating the hybrid pre-compound emission model

    International Nuclear Information System (INIS)

    Computer codes incorporating advanced nuclear models (optical, statistical and pre-equilibrium decay nuclear reaction models) were used to calculate neutron cross sections needed for fusion reactor technology. The elastic and inelastic scattering (n,2n), (n,p), (n,n'p), (n,d) and (n,γ) cross sections for stable molybdenum isotopes Mosup(92,94,95,96,97,98,100) and incident neutron energy from about 100 keV or a threshold to 20 MeV were calculated using the consistent set of input parameters. The hydrogen production cross section which determined the radiation damage in structural materials of fusion reactors can be simply deduced from the presented results. The more elaborated microscopic models of nuclear level density are required for high accuracy calculations

  20. Investigations of neutron spectra and dose distributions - with calculations and measurements - eleptical phantom for light-water moderated reactor spectrum

    International Nuclear Information System (INIS)

    Calculations and measurements for the dose distribution in a water-filled elliptical phantom when irradiated with neutrons of different unshielded light water moderated reactors are presented. The calculations were performed by a Monte Carlo code, for the measurements activation, TL and solid state nuclear track detectors were used. It was observed that the neutron spectra do not vary significantly inside the phantom and that not only the total absorbed dose but the kerma value at a depth of 2 cm can be higher than that on the front, in our cases by a factor of about 1.2. The measurements and calculations resulted in a kerma attenuation from the front to the back of the phantom of a factor of about 5. (author)

  1. Benchmarking of the WIMSD/CITATION deterministic code system for the neutronic calculations of TRIGA Mark-III research reactors

    International Nuclear Information System (INIS)

    Highlights: • Benchmark study performed for the neutronic calculations of TRIGA research reactors. • WIMSD-5B/CITATION is the utilized code system along with the WIMSD-IAEA-69 library. • The studied condensed spectra are five and seven energy groups spectra. • Analyzed: lattice parameters, reactivities, CR worth, flux and power distribution. • The lattice and neutronic parameters showed the accuracy of both condensed spectra. - Abstract: The objective of this paper is to assess the suitability and accuracy of the deterministic diffusion method for the neutronic calculations of the TRIGA Mark-III research reactors using the WIMSD/CITATION code system in proposed condensed energy spectra of five and seven energy groups with one and three thermal groups respectively. The utilized cell transport calculations code and core diffusion calculations code are the WIMSD-5B and the CITVAP v3.1 codes respectively, along with the WIMSD-IAEA-69 nuclear data library. Firstly, the assessment goes through analyzing the integral parameters – keff, ρ238, δ235, δ238, and C* – of the TRX and BAPL benchmark lattices and comparison with experimental and previous reference results using other ENDLs at the full energy spectra which show good agreement with the references at both spectra. Secondly, evaluation of the 3D nuclear characteristics of three different cores of the TRR-1/M1 TRIGA Mark-III Thai research reactor at the condensed energy spectra. The results include the excess reactivities of the cores and the worth of selected control rods which were compared with reference Monte Carlo results and experimental values. The results show good agreement with the references at both energy spectra and the better accuracy are attainable in the five energy groups spectrum. The results also include neutron flux distributions which are evaluated for future comparisons with other calculational techniques even they are comparable to reactors and fuels of the same type. The

  2. 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).

  3. Thermoluminescence dosimetry of a thermal neutron field and comparison with Monte Carlo calculations.

    Science.gov (United States)

    Fernandes, A C; Santos, J P; Kling, A; Marques, J G; Gonçalves, I C; Carvalho, A Ferro; Santos, L; Cardoso, J; Osvay, M

    2004-01-01

    The characteristics of thermoluminescence dosemeters (TLDs) regarding the determination of photon and neutron absorbed doses were investigated in a thermal neutron beam. Harshaw TLD-100 (LiF:Mg,Ti) and TLD-700 (7LiF:Mg,Ti) were compared with similar materials from Solid Dosimetric Detector and Method Laboratory (People's Republic of China). Harshaw TLD-700H (7LiF:Mg,Cu,P) and aluminium oxide (Al2O3:Mg,Y) from Hungary were also considered for photon dose measurement. The neutron sensitivity of the investigated materials was measured and found to be consistent with values reported by other authors. A comparison was made between the TL dose measurements and results obtained via conventional methods. An agreement within 20% was obtained, which demonstrates the ability of TLD for measuring neutron and photon doses in a mixed field, using careful calibration procedures and determining the neutron sensitivity for the usage conditions. PMID:15367765

  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. A formalism for the calculation of the effect of sodium voiding on neutron leakages in fast reactor

    International Nuclear Information System (INIS)

    A simple formalism for the calculation of the effect of sodium voiding on neutron leakages in a fast reactor is presented. The diffusion coefficients in a plane or 2-D lattice are calculated following a method which is very analogous to the method proposed earlier by the author for the treatment of thermal reactors. The two situations, sodium present, sodium voided, are calculated with the same approximations. It is known that it is impossible, in the situation where the sodium is voided, to calculate buckling-independent diffusion coefficients, for they diverge these coefficients are hence calculated, in both situations, at the lowest order of the expansion in terms of the buckling, which introduces a logarithmic term. The calculation is performed in the actual geometry of the lattice, without cylindricalizing the cell

  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. Analysis of the Pecore experimental programme carried out on Masurca and neutronics calculations for the PEC reactor

    International Nuclear Information System (INIS)

    A number of critical configurations installed in the Masurca experimental reactor of the French Commissariat a l'energie atomique (CEA) in Cadarache were used to simulate, by means of certain approximations, the neutron properties of the core of the PEC reactor. As the amount of plutonium available was insufficient, the Pecore experiment was carried out by using two different fuel zones: one with mixed U and Pu oxide, the other with enriched U oxide. However, an effort was made to keep the two zones similar as far as the neutron energy spectrum was concerned. The first interpretation of Pecore was carried out by the CEA with the assistance of staff of the Italian National Nuclear Energy Committee (CNEN), on the basis of French nuclear data and calculation methods. The paper describes the interpretation made in Italy on the basis of the codes and cross-sections used for the PEC neutron calculations, so that use could be made in this project of the discrepancies between calculation and experiment evaluated for Pecore. The characteristics compared were: critical mass, negative reactivity introduced by the control rods, ratios of reaction rates, efficiency of the nickel-base reflector and the axial and radial power curves. By way of a general conclusion, it can be stated that the interpretations of the CEA and the CNEN are in good agreement; the differences which they show are nearly always within the margin of experimental error. In addition to the main results of the Pecore experiment, the paper deals with certain aspects of: the PEC neutron calculation, from the definition of fuel enrichment to the calculation of the power level reached in each element, and from the updating of reactivity coefficients to the formulation of refuelling strategy. For some sectors, the results of the Pecore experiment had to be taken into account. (author)

  8. 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.)

  9. 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.

  10. 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.

  11. High-speed algorithm for calculating the neutron field in a reactor when working in dialog mode with a computer

    International Nuclear Information System (INIS)

    The large-scale construction of atomic power stations results in a need for trainers to instruct power-station personnel. The present work considers one problem of developing training computer software, associated with the development of a high-speed algorithm for calculating the neutron field after control-rod (CR) shift by the operator. The case considered here is that in which training units are developed on the basis of small computers of SM-2 type, which fall significantly short of the BESM-6 and EC-type computers used for the design calculations, in terms of speed and memory capacity. Depending on the apparatus for solving the criticality problem, in a two-dimensional single-group approximation, the physical-calculation programs require ∼ 1 min of machine time on a BESM-6 computer, which translates to ∼ 10 min on an SM-2 machine. In practice, this time is even longer, since ultimately it is necessary to determine not the effective multiplication factor K/sub ef/, but rather the local perturbations of the emergency-control (EC) system (to reach criticality) and change in the neutron field on shifting the CR and the EC rods. This long time means that it is very problematic to use physical-calculation programs to work in dialog mode with a computer. The algorithm presented below allows the neutron field following shift of the CR and EC rods to be calculated in a few seconds on a BESM-6 computer (tens of second on an SM-2 machine. This high speed may be achieved as a result of the preliminary calculation of the influence function (IF) for each CR. The IF may be calculated at high speed on a computer. Then it is stored in the external memory (EM) and, where necessary, used as the initial information

  12. Neutron response calculation on the basis of variable track etch rates along the secondary particle trajectories in CR-39

    CERN Document Server

    Hermsdorf, D; Dörschel, B; Henniger, J

    1999-01-01

    The calculation of the response of CR-39 detectors exposed to neutrons is of high importance for their dosimetric application. A computer code system has been developed for this purpose. Whereas the generation of secondary charged particles is carried out using non-analogue Monte-Carlo techniques with variance reduction the simulation of the track formation process is treated without any free parameter starting from the etch rate ratio V(REL) only. Results are given for the contribution of recoil protons to the response as a function of the neutron energy and angle of incidence. Furthermore, the influence of an external radiator has been studied. The comparison of the calculated values with experimental data confirm the reliability of the track etch model applied.

  13. 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.)

  14. Neutronics calculations in support of the OOR-MFE-4A and -4B spectral tailoring experiments

    International Nuclear Information System (INIS)

    The objective of this work was to provide the neutronic design for materials irradiation experiments in the Oak Ridge Research Reactor (ORR). Spectral tailoring to control the fast and thermal fluxes is required to provide the desired displacement and helium production rates in alloys containing nickel. The calculated fluences from the ongoing three-dimensional neutronics calculations are being scaled to agree with experimental data. As of March 27, 1984, this treatment yields 158.5 at. ppm He (not including 2.0 at. ppm He from 10B) and 10.36 dpa for type 316 stainless steel in ORR-MFE-4A and 99.4 at. ppm He and 7.10 dpa in ORR-MFE-4B

  15. Neutronics calculations in support of the ORR-MFE-4A and -4B spectral tailoring experiments

    International Nuclear Information System (INIS)

    The objective of this work is to provide the neutronic design for materials irradiation experiments in the Oak Ridge Research Reactor (ORR). Spectral tailoring to control the fast and thermal fluxes is required to provide the desired displacement and helium production rates in alloys containing nickel. The calculated fluences from the ongoing three-dimensional neutronics calculations are being scaled to agree with experimental data. As of march 31, 1985, this treatment yields 201.9 at. ppm He (not including 2.0 at. ppm He from 10B) and 13.1 dpa for type 316 stainless steel in ORR-MFE-4A and 155.9 at. ppm He and 10.60 dpa in ORR-MFE-4B

  16. Neutronics calculations in support of the ORR-MFE-4A and -4B spectral tailoring experiments

    International Nuclear Information System (INIS)

    The objective of this work is to provide the neutronic design for materials irradiation experiments in the Oak Ridge Research Reactor (ORR). Spectral tailoring to control the fast and thermal fluxes is required to provide the desired displacement and helium production rates in alloys containing nickel. The calculated fluences from the ongoing three-dimensional neutronics calculations are being scaled to agree with experimental data. As of September 30, 1984, this treatment yields 176.2 at. ppm He (not including 2.0 at. ppm He from 10B) and 11.90 dpa for type 316 stainless steel in ORR-MFE-4A and 133.0 at. ppm He and 9.13 dpa in ORR-MFE-4B. 2 references, 2 figures, 1 table

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

    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 ν, α, 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

  19. 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)

  20. MCANG - a group data library of cumulative angular distributions of elastically scattered neutrons for Monte Carlo calculations

    International Nuclear Information System (INIS)

    A code has been written for producing group data suited to take into account the center of mass anisotropy of elastic neutron scattering in Monte Carlo calculations. The format of the generated data library is described. Up to now variants of the library based on KEDAK2 and KEDAK3/ENDL78 have been produced. One of the libraries is listed in the appendix. (author)

  1. 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.

  2. 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.

  3. 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.

  4. Use of neutron activation analysis for the control of air pollution of Algiers

    International Nuclear Information System (INIS)

    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)

  5. 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.

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

    DEFF Research Database (Denmark)

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

    2015-01-01

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

  7. The Calculation of Collision Risk on Air-Routes Based on Variable Nominal Separation

    Institute of Scientific and Technical Information of China (English)

    QU Yu-ling; HAN Song-chen

    2010-01-01

    In this paper, a new method to calculate collision risk of air-routes, based on variable nominal separation, is proposed. The collision risk model of air-routes, based on the time variable and initial time interval variable, is given. Because the distance and the collision probability vary with time when the nominal relative speed between aircraft is not zero for a fixed initial time interval, the distance, the variable nominal separation, and the collision probability at any time can be expressed as functions of time and initial time interval. By the probabilistic theory, a model for calculating collision risk is acquired based on initial time interval distribution, flow rates, and the proportion of aircraft type. From the results of calculations, the collision risk can be characterized by the model when the nominal separation changes with time. As well the roles of parameters can be shown more readily.

  8. NEUTRON ACTIVATION ANALYSIS FOR SIMULTANEOUS DETERMINATION OF TRACE ELEMENTS IN AMBIENT AIR COLLECTED ON GLASS-FIBER FILTERS

    Science.gov (United States)

    Arsenic with 25 other elements are simultaneously determined in ambient air samples collected on glass-fiber filter composites at 250 United States sites. The instrumental neutron activation analysis (NAA) technique combined with the power of a dedicated mini-computer resulted in...

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

    International Nuclear Information System (INIS)

    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. - Highlights: • Localized ozone production in air may be an indicator of radioactive material. • Radiolytic ozone work is dominated by high radiation fields in the saturation regime. • For low level measurements we used a reactor as a mixed photon/neutron source. • Monte Carlo simulations were performed to understand the dose profile to air. • Different contributions to ozone production are discussed for neutrons and photons

  10. Air pollution biomonitoring in Argentina, application of neutron activation analysis to the study of biomonitors

    International Nuclear Information System (INIS)

    The assessment of baseline levels of atmospheric pollutants and the identification of polluted areas is a complex problem, as pollutant contents at a certain geographical location is usually a combination of contributions from various diverse sources, including long-range transport. Elemental chemical characterization of atmospheric pollutants is thus of great importance and Neutron Activation Analysis has proved to be a powerful technique for multielemental determination of trace elements in biomonitors and aerosols. The general objective of this project is to study the use of biomonitors, specially lichens, for evaluating pollutant levels over a wide geographic area of Argentina and for establishing baseline values and assessing time trends. Two lichen species (Usnea sp. and Ramalina ecklonii (Spreng.) Mey. and Flot) have been identified as suitable monitors of air pollution, with potential regional application at the central area of the country (province of Cordoba) and pilot studies have been initiated to test the practicability of sampling and sample collection. An area of approximately 40,000 km2 will be covered by a sampling network, using in situ growing lichens. The distribution maps for the two selected species are already drawn and sampling of local soils will also be conducted. Current efforts at the Neutron Activation Analysis laboratory are put on assessing, for the selected lichen species, the influence of sample preparation methods on trace element concentrations. The use of other analytical techniques will allow the evaluation of the bioindicator chemical response and its relationship to different atmospheric quality levels. Source identification and apportionment will be done by statistical fingerprinting of the elemental concentrations, as sources of pollution are characterized by being composed of different mixtures of elements in different proportions. In this way and as a long-term objective, regional maps will be drawn showing the

  11. 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)

  12. Neutronics calculations in support of the ORR-MFE-4A and -4B spectral tailoring experiments

    International Nuclear Information System (INIS)

    The calculated fluences from the ongoing three-dimensional neutronics calculations are being scaled to agree with experimental data. As of September 30, 1983, this treatment yields 121.6 at. ppM He (not including 2.0 at. ppM He from 10B) and 8.25 dpa for type 316 stainless steel in ORR-MFE-4A and 69.3 at. ppM He and 5.19 dpa in ORR-MFE-4B

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

    International Nuclear Information System (INIS)

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

  14. Theoretical methods for neutronics calculations of core-blanket and core-reflector systems in fast reactors

    International Nuclear Information System (INIS)

    The present work is a contribution to the neutronics calculational methods of fast neutron reactors. The first step is devoted to the analysis of the validity of the few-groups (of the order of 25) multigroup scheme, and of the transport-correction approximation for the treatment of the scattering anisotropy. This analysis includes both the reactor core, where the usual approximations are found to be satisfactory, and the reflector, where it turns out that the rapid variations of the neutron flux and of it's spectrum necessitate the improvement of the multigroup cross-sections' generation. Therefore, a zero-dimensional simple and accurate model for the average spectrum in the reflector is developed by the space-energy synthesis method. Finally using the Rayleigh-Ritz method, a model is developed in which the flux is spatially represented by an analytical function. This model is applied to the analysis of the sensitivity of reflector neutronics parameters to the variations of the cross sections

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-07-01

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

  16. 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.

  17. US/JAERI fusion neutronics calculational benchmarks for nuclear data and codes intercomparison

    International Nuclear Information System (INIS)

    The US Department of Energy (DOE) and the Japan Atomic Energy Research Institute (JAERI) have been involved in a collaborative research program on fusion neutronics. The program focuses on performing post- and pre-analyses for the integral experiments conducted at the fusion neutronics source (FNS) facility at JAERI. The main general objectives of the program are: (a) to provide experimental data needed to determine the accuracy, guide the development and establish the validity of computational methods and the nuclear data base; (b) to provide the data base required to evaluate the overall uncertainty (both analytical and experimental) in estimating key parameters of importance in fusion blanket design (e.g., tritium production rate, nuclear heating rate, dose rate, etc.); (c) to intercompare various measuring techniques and to increase the reliability of measurements by developing more advanced detectors; (d) to provide experimental data to assist in the selection of materials and configuration of candidate blanket concepts from the neutronics viewpoint

  18. 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.

  19. Coupling Serpent and OpenFOAM for neutronics - CFD multi-physics calculations

    OpenAIRE

    Tuominen, Riku

    2015-01-01

    The main goal of this work was to couple the Monte Carlo neutronics code Serpent 2 with a CFD solver from the OpenFOAM toolbox. The coupling was implemented with the already available multi-physics interface of Serpent. The interface allows the passing of high fidelity density and temperature distributions from an external solver to Serpent and also the passing of fission power distribution from Serpent to the external solver. The coupled CFD-neutronics problem was solved by iteration. At eac...

  20. BCG: a computer code for calculating neutron spectra and criticality in cells of fast reactors

    International Nuclear Information System (INIS)

    The BCG code for determining the space and energy neutron flux distribution and criticality of fast reactor cylindrical cells is discussed. The code solves the unidimensional neutron transport equation together with interface current relations at each energy point in an unionized energy grid prepared for the cell and at an arbitrary number of spatial zones. While the spatial resolution is user specified, the energy dependence of the flux distribution is resolved according to the degree of variation in the reconstruced total microscopic cross sections of the atomic species in the cell. Results for a simplified fuel cell illustrate the high resolution and accuracy that can be obtained with the code. (author)

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

    International Nuclear Information System (INIS)

    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

  2. Neutronics and dose calculation for prospective spent nuclear fuel cask for Ghana Research Reactor - 1 facility

    International Nuclear Information System (INIS)

    Ghana Research Reactor-1 core is to be converted from highly enrich Uranium (HEU) fuel to low enriched Uranium (LEU) fuel in the near future: a storage cask will be needed to store the HEU fuel. Notwithstanding the core conversion process, It is also important for the facilitv to have a storage cask ready when the fuel is finally spent to temporarily store the fuel until permanent storage is provided. Winfrith Improved Multigroup Scheme-Argonne National Laboratory (WIMS-ANL). Reactor Burnup System (REBUS). Oak Ridge Isotope Generation (ORIGEN2) and Monte Carlo ''N'' Particle (MCNP5) codes have been used to design the cask. WIMS-ANL was used in generating cross sections for the REBUS code which was used in the burnup calculations. The REBUS code was used to estimate the core life time. An estimated core life of approximatcly 750 full-power-equivaicnt-days was obtained for reactor operation of 2hours a day. 4 days a week and 48 weeks in a year. The ORIGIN2 code recorded U-235 burnup weight percent of 2.90% whilst the result from the REBUS3 code was 2.86%. The amount of Pu-239 at the end of the irradiation period was 145 mg which is very low relative to other low power reactors. Isotopic inventory obtained from the ORIGIN2 and REBUS3 runs were used in setting up the MCNP5 input deck for the MCNP5 calculation of the criticality and dose rate. Six cask design options were investigated. The materials for the casks designs were selected based on their attenuation coefficient properties and their high removal cross section properties. The various materials were arranged in no specific order in multilayered casks. The reason for investigating six casks was to look at various arrangements of the cask layers that will optimize effective shielding. The spent nuclear fuel at discharge was used as the radioactivity source during the MCNP simulation. The multilayer cask shield comprise of serpentine concrete of density 5.14 g/cm3 and thickness 21.94cm which

  3. Study of room-return neutrons

    International Nuclear Information System (INIS)

    The coefficient that relates the neutron source strength and inner room surface area with the thermal neutron fluence rates has been calculated for neutrons whose energy goes from 1eV to 20MeV. This coefficient was calculated using Monte Carlo methods for 150, 200 300, 535.24, 832.10 and 1010cm-radius spherical cavity with and without air. In the calculations monoenergetic neutron sources were located at the center of cavity and, along the spherical cavity radius, neutron spectra were determined at several source-to-detector distances. From the neutron spectra the thermal neutron (E=<0.414eV) contribution was calculated and several coefficients were estimated. A weighted average of all coefficients was estimated being 5.6+/-0.1, this is the c-value for the case of rooms without air, and 4.8+/-0.4, for rooms with air. In the aim to compare the results with monoenergetic neutrons, calculations were also performed using two isotopic neutron sources, Cf252 and AmBe241 in vacuum

  4. Constraints on Skyrme equations of state from properties of doubly magic nuclei and ab-initio calculations of low-density neutron matter

    OpenAIRE

    Brown, B.Alex; Schwenk, A.

    2014-01-01

    We use properties of doubly-magic nuclei and ab-initio calculations of low-density neutron matter to constrain Skyrme equations of state for neutron-rich conditions. All of these properties are consistent with a Skyrme functional form and a neutron-matter equation of state that depends on three parameters. With a reasonable range for the neutron-matter effective mass, the values of the two other Skyrme parameters are well constrained. This leads to predictions for other quantities. The neutro...

  5. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-07-01

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

  7. The effect of three-body cluster energy on LOCV calculation for hot nuclear and neutron matter

    International Nuclear Information System (INIS)

    The two-body correlation functions, obtained in a lowest-order constrained variational calculation for hot nuclear and neutron matter, with the Reid potential and the explicit inclusion of Δ(1234), are state averaged and used to calculate the three-body cluster energy. The three-body cluster energy is found to vary between about 1 and 2 MeV through and beyond twice the nuclear-matter saturation density for temperatures between 5 and 20 MeV. However, the inclusion of a three-body cluster reduces the nuclear-matter flashing and critical temperatures. A critical temperature of 15.8 MeV and a critical exponent of 0.35 is found. The results of entropy calculations are in good agreement with experimental prediction and other theoretical results. Finally it is shown that by allowing an explicit Δ(1234) degree of freedom through the Reid potential up to and including the three-body clusters, the lowest-constrained variational calculation yields other nuclear- and neutron-matter properties close to the available semi-empirical and experimental data at zero and finite temperatures. (author)

  8. Neutronic calculations in support of the ORR-MFE-4 spectral tailoring experiments

    International Nuclear Information System (INIS)

    The objective of this work is to provide the neutronic design for materials irradiation experiments in the Oak Ridge Research Reactor (ORR). Spectral tailoring to control the fast and thermal fluxes is required to provide the desired displacement and helium production rates in alloys containing nickel

  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. Activities of research-reactor-technology project in FNCA from FY2005 to FY2007. Sharing neutronics calculation technique for core management and utilization of research reactors

    International Nuclear Information System (INIS)

    RRT project (Research-Reactor-Technology Project) was carried out with the theme of 'sharing neutronics calculation technique for core management and utilization of research reactors' in the framework of FNCA (Forum for Nuclear Cooperation in Asia) from FY2005 to FY2007. The objective of the project was to improve and equalize the level of neutronics calculation technique for the reactor core management among participating countries to assure the safe and stable operation of research reactors and the promotion of the effective utilization. Neutronics calculation codes, namely SRAC code system and MVP code, were adopted as common codes. Participating countries succeeded in applying the common codes to analyzing the core of each domestic research reactor. Some participating countries succeeded in applying the common codes to analyzing for utilization of own research reactors. Activities of RRT project have improved and equalized the level of neutronics calculation technique among participating countries. (author)

  11. 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.

  12. Effects of frequency on fatigue behavior of type 316 low-carbon, nitrogen-added stainless steel in air and mercury for the spallation neutron source

    Science.gov (United States)

    Tian, H.; Liaw, P. K.; Fielden, D. E.; Brooks, C. R.; Brotherton, M. D.; Jiang, L.; Yang, B.; Wang, H.; Strizak, J. P.; Mansur, L. K.

    2006-01-01

    The high-cycle fatigue behavior of type 316 low-carbon, nitrogen-added (LN) stainless steel (SS), the prime-candidate target-container material for the spallation neutron source (SNS), was investigated in air and mercury. Test frequencies ranged from 0.2 to 10 Hz with an R ratio of -1, and 10 to 700 Hz with an R ratio of 0.1. During tension-compression fatigue studies, a significant increase in the specimen temperature was observed at 10 Hz in air, which decreased the fatigue life of the 316 LN SS relative to that at 0.2 Hz. Companion tests in air were carried out, while cooling the specimen with nitrogen gas at 10 Hz in air. In these experiments, fatigue lives were comparable at 10 Hz in air with nitrogen cooling and at 0.2 Hz in air. During tension-tension fatigue studies, a higher specimen temperature was observed at 700 than at 10 Hz. After cooling the specimen, comparable fatigue lives were found at 10 and at 700 Hz. The frequency effect on the fatigue life in mercury was found to be much less than that in air, due to the fact that mercury acts as an effective coolant during the fatigue experiment. Striation spacing on the fracture surface at different test frequencies was closely examined, relative to calculated Δ K values, during fatigue of the 316 LN SS. Specimen self-heating has to be considered in understanding fatigue characteristics of 316 LN SS in air and mercury.

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

    International Nuclear Information System (INIS)

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

  15. The Neutronic And Power Distribution Calculations For Triga 2 MW Reactor Using WIMS-D/4 And Citation Codes

    International Nuclear Information System (INIS)

    . The neutronic calculation has been carried out for TRIGA 2 MW reactor. These included criticality flux and power distributions. Computer code Citation which solves 7-groups, 3-dimensional hexagonal geometry has been used. The multi groups-cross-section is generated by the WIMS-D/4 code.This 7-group-39x39x38-mesh-points problem takes about 90 minutes on the Pentium-133 MHz PC. The calculation of the initial core of TRIGA 2 MW reactor shows that the excess reactivity of the core is 7,8% and the thermal fluxes in the irradiation positions are between 1.0-2.9*1013n cm-2s-1. The results are about 10% deviate from those calculated by General Atomics. In the initial core, the highest power is produced in the C-9 position. The fuel element in this position produces 30.7 k W thermal power

  16. 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.

  17. Neutronic simulation calculations to assess the proliferation resistance of nuclear technologies

    International Nuclear Information System (INIS)

    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.)

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

    International Nuclear Information System (INIS)

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

  19. Quantum Monte Carlo calculations of neutron matter with chiral three-body forces

    Science.gov (United States)

    Tews, I.; Gandolfi, S.; Gezerlis, A.; Schwenk, A.

    2016-02-01

    Chiral effective field theory (EFT) enables a systematic description of low-energy hadronic interactions with controlled theoretical uncertainties. For strongly interacting systems, quantum Monte Carlo (QMC) methods provide some of the most accurate solutions, but they require as input local potentials. We have recently constructed local chiral nucleon-nucleon (NN) interactions up to next-to-next-to-leading order (N2LO ). Chiral EFT naturally predicts consistent many-body forces. In this paper, we consider the leading chiral three-nucleon (3N) interactions in local form. These are included in auxiliary field diffusion Monte Carlo (AFDMC) simulations. We present results for the equation of state of neutron matter and for the energies and radii of neutron drops. In particular, we study the regulator dependence at the Hartree-Fock level and in AFDMC and find that present local regulators lead to less repulsion from 3N forces compared to the usual nonlocal regulators.

  20. Application of a calculational model for thermal neutrons through biological shields

    International Nuclear Information System (INIS)

    In this work a computational program, based on the Boltzmann transport integrodifferential equation, is applied. The scattering kernel is represented by the synthetic scattering model. The behaviour of thermal neutron in hydrogenous materials, which can be used as biological shields, are studied. These materials are water, polyethylene, Oak-Ridge concrete, ordinary concrete and manganese concrete. The data obtained are presented in tables. The results are analysed and compared with similar experimental values. Safety evaluation and environmental impact are discussed. 2 tabs

  1. Ab initio calculations as a quantitative tool in the inelastic neutron scattering study of a single-molecule magnet analogue.

    Science.gov (United States)

    Vonci, Michele; Giansiracusa, Marcus J; Gable, Robert W; Van den Heuvel, Willem; Latham, Kay; Moubaraki, Boujemaa; Murray, Keith S; Yu, Dehong; Mole, Richard A; Soncini, Alessandro; Boskovic, Colette

    2016-02-01

    Ab initio calculations carried out on the Tb analogue of the single-molecule magnet family Na9[Ln(W5O18)2] (Ln = Nd, Gd, Ho and Er) have allowed interpretation of the inelastic neutron scattering spectra. The combined experimental and theoretical approach sheds new light on the sensitivity of the electronic structure of the Tb(III) ground and excited states to small structural distortions from axial symmetry, thus revealing the subtle relationship between molecular geometry and magnetic properties of the two isostructural species that comprise the sample. PMID:26690503

  2. 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.

  3. Neutronic characterization and decay heat calculations in the in-vessel fuel storage facilities for MYRRHA/FASTEF

    International Nuclear Information System (INIS)

    Highlights: ► Monte Carlo design of reactor facilities. ► Neutron coupling assessment between critical core and fresh fuel in the storage vessels. ► Power contribution by induced fission from neutrons leaving the core, spontaneous fission and (α, n) sources. ► Power decay heat estimation for different reactor fuel cycles scenarios. ► Material damage assessment in the storage vessels. - Abstract: The main objective of the Central Design Team (CDT) project is to establish an engineering design of a Fast Spectrum Transmutation Experimental Facility (FASTEF) that is the pilot plant of an experimental-scale of both an Accelerator Driven System (ADS) and a Lead Fast Reactor (LFR), based on the MYRRHA reactor concept, planned to be built during the next decade. The MYRRHA reactor concept is devoted to be a multi-purpose irradiation facility aimed at demonstrating the efficient transmutation of long-lived and high radiotoxicity minor actinides, fission products and the associated technology. An important issue regarding the reactor design of the MYRRHA/FASTEF experiment is the In-Vessel Fuel Storage Facilities (IVFSFs), both for fresh and spent fuel, as it might have an impact on the criticality of the overall system that must be quantified. In this work, the neutronic analysis of the in-vessel fuel storage facility and its coupling with the critical core was performed, using the state of the art Monte Carlo program MCNPX 2.6.0 and ORIGEN 2.2 computer code system for calculating the buildup and decay heat of spent fuel. Several parameters were analyzed, like the criticality behavior (namely the Keff), the neutron fluxes and their variations, the fission power production and the radiation damage (the displacements per atom). Finally, also the heat power generated by the fission products decay in the spent fuel was assessed.

  4. Status of the McCad geometry conversion tool and related visualization capabilities for 3D fusion neutronics calculations

    Energy Technology Data Exchange (ETDEWEB)

    Große, D.; Fischer, U., E-mail: ulrich.fischer@kit.edu; Kondo, K.; Leichtle, D.; Pereslavtsev, P.; Serikov, A.

    2013-10-15

    Highlights: • McCad – software tool developed at KIT for the automatic conversion of CAD models into the geometry representation of Monte Carlo particle transport codes. • Open source software running under the Linux operating system and utilizing Open Cascade CAD kernel with the Qt4 libraries for the graphical user interface (GUI). • Converted geometry models can be output in the syntax of MCNP and TRIPOLI of the Monte Carlo codes. • Related visualization capabilities, based on coupling of McCad with the ParaView software, allow to overlay mesh tally distributions to the CAD geometry. • McCad applied to solve fusion neutronics problems of ITER and the IFMIF neutron source. -- Abstract: The McCad geometry conversion tool has been developed at KIT to enable the automatic conversion of CAD models into the semi-algebraic geometry representation as utilized in Monte Carlo particle transport simulations. McCad is entirely based on open source software, it is running under the Linux operating system and utilizes the Open Cascade CAD kernel with the Qt4 libraries for the graphical user interface (GUI). The converted geometry models can be output in the syntax of the Monte Carlo codes MCNP and TRIPOLI. Related visualization capabilities are based on the coupling of McCad with the ParaView software and allow to overlay mesh tally distributions to the CAD geometry. This enables perspective 3D representations or animations on the CAD geometry. The paper presents the current status of the McCad approach and its implementation, and discusses its capabilities, limitations as well as future development needs. The use of McCad for fusion neutronics applications is illustrated on the examples of the MCNP model generation for ITER Test Blanket Modules (TBM) and the test cell facility of the IFMIF neutron source including Monte Carlo shielding calculations using the converted models.

  5. Water-extended polyester neutron shield for a 252Cf neutron source

    International Nuclear Information System (INIS)

    A Monte Carlo study to determine the shielding features to neutrons of water-extended polyester was carried out. During calculations, 252Cf and shielding were modelled and the neutron spectra as well as the H*(10) were calculated in four sites. The calculation was extended to include a water shielding, the source in vacuum and in air. Besides neutron shielding characteristics, the Kerma in air due to gammas emitted by 252Cf and due to capture γ rays in the shielding were included. (authors)

  6. 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.

  7. 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...

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

    Science.gov (United States)

    Minato, Futoshi

    2016-06-01

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

  9. 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

  10. Improved calculation of the prompt fission neutron spectrum from the spontaneous fission of /sup 252/Cf: Preliminary results

    International Nuclear Information System (INIS)

    An improved calculation is presented for the prompt fission neutron spectrum N(E) from the spontaneous fission of /sup 252/Cf. In this calculation the fission-spectrum model of Madland and Nix is used, but with several improvements leading to a physically more accurate representation of the spectrum. Specifically, the contributions to N(E) from the entire fission-fragment mass and charge distributions will be calculated instead of calculating on the basis of a seven- point approximation to the peaks of these distributions as has been done in the past. Therefore, values of the energy release in fission, fission-fragment kinetic energy, and compound nucleus cross section for the inverse process will be considered on a point-by-point basis over the fragment yield distributions instead of considering averages of these quantities over the peaks of the distributions. Preliminary results will be presented and compared with a measurement, an earlier calculation, and a recent evaluation of the spectrum. 14 refs., 4 figs

  11. The all particle method: Coupled neutron, photon, electron, charged particle Monte Carlo calculations

    International Nuclear Information System (INIS)

    At the present time a Monte Carlo transport computer code is being designed and implemented at Lawrence Livermore National Laboratory to include the transport of: neutrons, photons, electrons and light charged particles as well as the coupling between all species of particles, e.g., photon induced electron emission. Since this code is being designed to handle all particles this approach is called the ''All Particle Method''. The code is designed as a test bed code to include as many different methods as possible (e.g., electron single or multiple scattering) and will be data driven to minimize the number of methods and models ''hard wired'' into the code. This approach will allow changes in the Livermore nuclear and atomic data bases, used to described the interaction and production of particles, to be used to directly control the execution of the program. In addition this approach will allow the code to be used at various levels of complexity to balance computer running time against the accuracy requirements of specific applications. This paper describes the current design philosophy and status of the code. Since the treatment of neutrons and photons used by the All Particle Method code is more or less conventional, emphasis in this paper is placed on the treatment of electron, and to a lesser degree charged particle, transport. An example is presented in order to illustrate an application in which the ability to accurately transport electrons is important. 21 refs., 1 fig

  12. 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.

  13. 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.

  14. 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.

  15. Constraints on Skyrme Equations of State from Properties of Doubly Magic Nuclei and Ab-Initio Calculations of Low-Density Neutron Matter

    CERN Document Server

    Brown, B Alex

    2013-01-01

    We use properties of doubly-magic nuclei and ab-initio calculations of low-density neutron matter to constrain Skyrme equations of state for neutron-rich conditions. All of these properties are consistent with a Skyrme functional form and a neutron-matter equation of state that depends on three parameters. With a reasonable range for the neutron-matter effective mass, the values of the two other Skyrme parameters are well constrained. This leads to predictions for other quantities. The neutron skins for $^{208}$Pb and $^{48}$Ca are predicted to be 0.182(10) fm and 0.173(5) fm, respectively. Other results including the dipole polarizability are discussed.

  16. Program MCU for Monte-Carlo calculations of neutron-physical characteristics of nuclear reactors

    International Nuclear Information System (INIS)

    A description of the MCU data modification is presented. The calculation results by the MCU-2 and MCU-3 codes are compared for the critical assemblies of a different reactor types. The full list of the critical assemblies calculation results obtained by all MCU code versions is given. 32 refs.; 32 tabs

  17. High energy neutron cross-sections and kerma values of biomedical interest calculated with a nuclear model applicable to light nuclei

    International Nuclear Information System (INIS)

    A model has been developed for calculating fast neutron cross sections (E > 14 MeV) for light nuclei of biomedical interest. The model explicitly includes experimental nuclear structure information. Some calculations for 12C, 14N, and 16O are presented

  18. Computer Programs for Calculating the Isentropic Flow Properties for Mixtures of R-134a and Air

    Science.gov (United States)

    Kvaternik, Raymond G.

    2000-01-01

    Three computer programs for calculating the isentropic flow properties of R-134a/air mixtures which were developed in support of the heavy gas conversion of the Langley Transonic Dynamics Tunnel (TDT) from dichlorodifluoromethane (R-12) to 1,1,1,2 tetrafluoroethane (R-134a) are described. The first program calculates the Mach number and the corresponding flow properties when the total temperature, total pressure, static pressure, and mole fraction of R-134a in the mixture are given. The second program calculates tables of isentropic flow properties for a specified set of free-stream Mach numbers given the total pressure, total temperature, and mole fraction of R-134a. Real-gas effects are accounted for in these programs by treating the gases comprising the mixture as both thermally and calorically imperfect. The third program is a specialized version of the first program in which the gases are thermally perfect. It was written to provide a simpler computational alternative to the first program in those cases where real-gas effects are not important. The theory and computational procedures underlying the programs are summarized, the equations used to compute the flow quantities of interest are given, and sample calculated results that encompass the operating conditions of the TDT are shown.

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

    International Nuclear Information System (INIS)

    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

  20. A comparison of the moisture gauge and the neutron log in air-filled holes at NTS

    International Nuclear Information System (INIS)

    Two methods are commonly used to measure water content of geologic materials by neutron diffusion. One is used mostly in agricultural, mining and civil engineering areas and is called a moisture gauge. The other is used principally in petroleum and mineral exploration, and is called a neutron log. Both are used at NTS, the moisture gauge principally in tunnels, the neutron log in vertical drilled holes. There is little communication between the two industrial groups, and the measurement instruments have evolved with very different operational characteristics, and one important physics difference, the source to detector spacing. The moisture gauge has a very short, 0-6 cm, spacing, with little internal shielding, and count increases with water. In contrast, the neutron log has a long spacing, 30-50 cm, substantial internal shielding, and exhibits decreasing count with increasing water. Because of its short spacing the moisture gauge gives better bed resolution than the neutron log. Because its count increases with water, the moisture gauge is more strongly affected by water in the borehole, especially in dry formations. In these conditions the neutron log is the method of choice. In air-filled holes, if source size or logging time is not a constraint, the relative sensitivity of the two tools to water is determined by the relative strengths of borehole effects as fluid, holesize, or tool-wall gap. If source size is a constraint for safety reasons, the short spacing provides higher countrates for a given detector efficiency and thus better relative precision in determining the true count. If source size is limited because of detector or electronics saturation, the short spacing will be better at high water content, while the long spacing will be better at low water content. In any case the short spacing may have an advantage because it can make better contact with the hole wall and it can be more easily corrected for gap

  1. Calculation and analysis of neutron and radiation characteristics of lead coolants with isotopic tailoring for future nuclear power facilities

    Energy Technology Data Exchange (ETDEWEB)

    Blokhin, A.I.; Ivanov, A.P.; Korobeinikov, V.V.; Lunev, V.P.; Manokhin, V.N.; Khorasanov, G.L. [SSC RF A. I. Leypunsky Institute for Physics and Power Engineering, Obninsk, Kaluga Region (Russian Federation)

    2000-03-01

    A new type of safe fast reactor with lead coolant was proposed in Russia. The use of coolants with low moderating properties is one of the ways to get a hard neutron spectrum and an increase in the burning of Np-237, Am-243 and other miner actinides(MA) fissionable preferentially in the fast reactor. The stable lead isotope, Pb-208, is proposed as the one of such coolants. The neutron inelastic scattering cross-section of Pb-208 is 3.0-3.5 times less than the one of other lead isotopes. Calculation of the MA transmutation rates in the standard BN-type fast reactor with different coolants is performed by Monte-Carlo method using Code MMKFK. Six various models are simulated for the fast reactor blanket with different kinds of fuel and coolant. The fast reactor with natural-lead coolant practically does not differ from the reactor with sodium coolant relative to MA incineration. The use of Pb-208 as a coolant in the fast reactor results in increasing incineration of MA from 18 to 26% in comparison with a usual fast reactor. Calculation of induced radioactivity was performed using the FISPACT-3 inventory code, also. The results include total induced radioactivity and dose rate for initial material composition and selected long-lived radionuclides. The calculations show that the coolant consisting of lead isotope, Pb-206, or Pb-207, can be considered as the low-activation one because it does not practically contain long-lived toxic radionuclides. (M. Suetake)

  2. 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.

  3. Calculation of conversion coefficients for effective dose for neutrons using a female voxel anthropomorphic model and the MCNPX code

    International Nuclear Information System (INIS)

    This work aims to calculate the fluence to effective dose conversion coefficients, (E/Φ), for monoenergetic neutrons from 10-9 to 20 MeV, based on the radiation (wR) and tissue (wT) weighting factors values recommended by ICRP publications numbers 60 and 103. The organs and tissues absorbed doses were calculated using the radiation transport code MCNPX and a female anthropomorphic voxel-based simulator, assuming whole-body irradiation by plane-parallel beams, on the geometries of the antero-posterior (AP) and postero-anterior (PA) irradiation. Dose calculations were performed for 21 selected organs of the body, for which the International Commission on Radiological Protection and the International Commission on Radiological Units and Measurements have set tissue weighting factors for the determination of the effective dose. From comparison between the dose results calculated and the data reported for the MIRD model, it can be concluded that, the fluence to effective dose conversion coefficients obtained using the voxel simulator are underestimated by a factor of up to 5 times when compared with the one obtained by ICRP 74, using mathematical simulators. (author)

  4. 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.

  5. Fast neutron reaction data calculations with the computer code STAPRE-H

    International Nuclear Information System (INIS)

    Description of the specific features of the version STAPRE-H are given. Illustration of the model options and parameter influence on the calculated results is done to trace the accurate reproducing of large body of correlated data. (authors)

  6. Calculated neutron-activation cross sections for E/sub n/ /le/ 100 MeV for a range of accelerator materials

    International Nuclear Information System (INIS)

    Activation problems associated with particle accelerators are commonly dominated by reactions of secondary neutrons produced in reactions of beam particles with accelerator or beam stop materials. Measured values of neutron-activation cross sections above a few MeV are sparse. Calculations with the GNASH code have been made for neutrons incident on all stable nuclides of a range of elements common to accelerator materials. These elements include B, C, N, O, Ne, Mg, Al, Si, P, S, Ar, K, Ca, Cr, Mn, Fe, Co, Ni, Cu, Zn, Zr, Mo, Nd, and Sm. Calculations were made for a grid of incident neutron energies extending to 100 MeV. Cross sections leading to the direct production of as many as 87 activation products for each of 84 target nuclide were tabulated on this grid of neutron energies, each beginning with the threshold for the product nuclide's formation. Multigrouped values of these cross sections have been calculated and are being integrated into the cross-section library of the REAC-2 neutron activation code. Illustrative cross sections are presented. 20 refs., 6 figs., 1 tab

  7. Calculations of neutron flux for BNCT facility of typical working core Multipurpose Reactor (RSG-GAS) using MCNP4B Code

    International Nuclear Information System (INIS)

    Calculation of neutron flux distributions of RSG-GAS typical working core using MCNP 4b Code has been done. Prior to the calculations, modelling of fuel element of meat as well as surfaces of cladding cell and geometry should be made. The model was then included water as a containment also developed. To achieve neutron flux behavior, it was simulated 200,000 to 2,000,000 neutrons. The calculation results indicated that the neutron flux in TWC core is in the order of 1014. Meanwhile, the best flux order for the BNCT facility should be in the order of 1010. With the use of any method, such as constructing of shielding and collimator, the order of neutron flux will decrease. In the previous research in 2001, the results showed the neutron flux in the order of 1010 by installing the collimator with 45 cm thick, made of Pb and 380 cm from the core centre. The results of this research completed with the research done in 2001, 2000 and 1999 certainly support the possibility to construct the BNCT facility in RSG-GAS reactor core

  8. 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)

  9. Revised Dalton's method for calculation of thermodynamic properties of unsaturated humid air and gas mixture after combustion in humid air turbine cycle

    International Nuclear Information System (INIS)

    The article applies Revised Dalton's method for calculation of thermodynamic properties (i.e. specific enthalpy, specific entropy and specific volume) of unsaturated humid air and gas mixture after combustion in humid air turbine cycle. The research temperature range is from 280 K to 1600 K and pressure range from 0.1 MPa to 5 MPa. “Improvement Factor” and “Cutting Off Temperature” for unsaturated humid air are explored in depth. Two “Improvement Factor” formulas are proposed. The discovery of the changing trends of “Improvement Factors” reveals the fundamental behaviors of dry air and water vapor in unsaturated humid air. Another discovery is “Cutting Off Temperature”. It is a crucial temperature point, above which the interaction of dissimilar molecules may be omitted. Revised Dalton's method may also be applied to gas mixture after combustion. The thermodynamic properties of unsaturated humid air and gas mixture after combustion are calculated by the Revised Dalton's method. The average error of Revised Dalton's method is within 0.1% compared to experimental data. - Highlights: • Revised Dalton's Model is suitable to unsaturated humid air and gas mixture after combustion. • Two “Improvement Factor” formulas are proposed for unsaturated humid air. • Changing trends of “Improvement Factors” reveal fundamental behaviors of dry air and water vapor in gas mixture. • When temperature exceeds “Cutting Off Temperature”, interactions between dissimilar molecules may be omitted. • This method deviates from experimental data less than 0.1%

  10. Comparison of Calculated and Measured Neutron Fluence in Fuel/Cladding Irradiation Experiments in HFIR

    International Nuclear Information System (INIS)

    A recently-designed thermal neutron irradiation facility has been used for a first series of irradiations of PWR fuel pellets in the high flux isotope reactor (HFIR) at Oak Ridge National Laboratory. Since June 2010, irradiations of PWR fuel pellets made of UN or UO2, clad in SiC, have been ongoing in the outer small VXF sites in the beryllium reflector region of the HFIR, as seen in Fig. 1. HFIR is a versatile, 85 MW isotope production and test reactor with the capability and facilities for performing a wide variety of irradiation experiments. HFIR is a beryllium-reflected, light-water-cooled and -moderated, flux-trap type reactor that uses highly enriched (in 235U) uranium (HEU) as the fuel. The reactor core consists of a series of concentric annular regions, each about 2 ft (0.61 m) high. A 5-in. (12.70-cm)-diam hole, referred to as the flux trap, forms the center of the core. The fuel region is composed of two concentric fuel elements made up of many involute-shaped fuel plates: an inner element that contains 171 fuel plates, and an outer element that contains 369 fuel plates. The fuel plates are curved in the shape of an involute, which provides constant coolant channel width between plates. The fuel (U3O8-Al cermet) is nonuniformly distributed along the arc of the involute to minimize the radial peak-to-average power density ratio. A burnable poison (B4C) is included in the inner fuel element primarily to reduce the negative reactivity requirements of the reactor control plates. A typical HEU core loading in HFIR is 9.4 kg of 235U and 2.8 g of 10B. The thermal neutron flux in the flux trap region can exceed 2.5 x 1015 n/cm2 · s while the fast flux in this region exceeds 1 x 1015 n/cm2 · s. The inner and outer fuel elements are in turn surrounded by a concentric ring of beryllium reflector approximately 1 ft (0.30 m) thick. The beryllium reflector consists of three regions: the removable reflector, the semi-permanent reflector, and the permanent reflector

  11. 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.

  12. Hamming method for solving the delayed neutron precursor concentration for reactivity calculation

    International Nuclear Information System (INIS)

    Highlights: ► We present a new formulation to calculate the reactivity using the Hamming method. ► This method shows better accuracy than existing methods for reactivity calculation. ► The reactivity is calculated without limitation of the nuclear power form. ► The method can be implemented in reactivity meters with time step of up to 0.1 s. - Abstract: We propose a new method for numerically solving the inverse point kinetic equation for a nuclear reactor using the Hamming method, without requiring the nuclear power history and without using the Laplace transform. This new method converges with accuracy of order h5, where h is the step in the computation time. The procedure is validated for different forms of the nuclear power and with different time steps. The results indicate that this method has a better accuracy and lower computational effort compared with other conventional methods that use the nuclear power history.

  13. 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.

  14. 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

    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...... 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...... 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...

  15. Neutronic Parameters Calculations for the RP-10 and RP-0 peruvian research reactors

    International Nuclear Information System (INIS)

    Theoretical safety calculations were done with proved codes utilized by the staff of the RERTR program in the HEU to LEU core conversions. The studies were designed to evaluate the reactivity coefficients and kinetics parameters of the reactor involved in the evolution of peak power transients by reactivity insertion accidents. It was done to show the trend of these reactivity coefficients as a function of the core size and fuel depletion for RP-10 cores. It was useful to get a better understanding of the progression of the reactivity insertion transients monitoring the critical thermal-hydraulic parameters to avoid core damage. To confirm the accuracy of these studies the results were compared with experimental data of the SPERT I reactors. The microscopic cross section calculations were condensed to 15 broad groups using the WIMSD4M code for all the isotopes from eight different regions that model seven different assemblies. The Supercell, multiplate and homogenized options were used to represent the different assemblies in the reactor. For diffusion theory calculations the DIF3D code was used in planar geometry with input axial buckling to simulate axial leakage. To Benchmark the designed models used in the cross sections generation and the DIF3D designed model, the VIM Monte Carlo code was used. The RECOEFF code was used to calculate the reactivity coefficients

  16. Calculation of Spin Observables for Proton-Neutron Elastic Scattering in the Bethe-Salpeter Equation

    CERN Document Server

    Kinpara, Susumu

    2016-01-01

    Bethe-Salpeter equation is applied to $p$-$n$ elastic scattering. The spin observables are calculated by the M matrix similar to $p$-$p$ case. The parameters of the meson-exchange model are used with the cut-off for the pion exchange interaction. Change of the M matrix indicates breaking of the charge independence in the nucleon-nucleon system.

  17. 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)

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

    International Nuclear Information System (INIS)

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

  19. $\\beta$-decay properties for neutron-rich Kr-Tc isotopes from deformed pn-QRPA calculations with realistic forces

    CERN Document Server

    Fang, Dong-Liang; Suzuki, Toshio

    2012-01-01

    In this work we studied $\\beta$-decay properties for deformed neutron-rich nuclei in the region Z=36-43. We use the deformed pn-QRPA methods with the realistic CD-Bonn forces, and include both the Gamow-Teller and first-forbidden types of decays in the calculation. The obtained $\\beta$-decay half-lives and neutron-emission probabilities of deformed isotopes are compared with experiment as well as with previous calculations. The advantages and disadvantages of the method are discussed.

  20. Calculational determination of neutron flux densities in the IR-8 reactor with the aim of choosing the additional cells for material irradiation

    International Nuclear Information System (INIS)

    The calculated analysis of the neutronic characteristics of working loading of the IR-8 reactor at its physical start-up is carried out for the estimation of flux definition error in the reflector. The calculated analysis of working loading of the IR-8 reactor with ampoule rigs is carried out with the purpose of choice of the cell for irradiation of constructional materials in the reflector of the IR-8 reactor under condition of neutron flux density ∼ 2 x 1011 cm-2 s-1 (E > 0.5 MeV)

  1. Monte Carlo neutron fluence calculations, activation measurements and spectrum adjustment for the KORPUS dosimetry experiment

    International Nuclear Information System (INIS)

    KORPUS is an irradiation facility located at the lateral core surface of the 6 MW experimental reactor RBT-6 in Dimitrovgrad. In this work the KORPUS irradiation experiment has been used to demonstrate the capability of the pressure vessel dosimetry methodology developed in Rossendorf to solve these problems. At the same time the experiments were used to test recent improvements of this methodology including a new procedure for treatment of elastic scattering in the Monte Carlo code TRAMO and a new multispectrum version of the adjustment code. By means of a series of calculations the influence of model and data approximations were investigated aiming at an evaluation of the uncertainties of the calculations. Further, uncertainty investigations were carried out in connection with spectrum adjustment resulting in covariances of spectra, measured reaction rates and fluence integrals. (orig.)

  2. Neutronic calculations in support of the ORR-MFE-4 spectral tailoring experiments

    International Nuclear Information System (INIS)

    New scaling factors have been obtained to force agreement between the experimentally measured and calculated fluences. As of March 31, 1983, these factors yield 89.6 at. ppM He (not including 2.0 at. ppM He from 10B) and 6.18 dpa for type 316 stainless steel in ORR-MFE-4A and 57.7 at. ppM He and 4.48 dpa in ORR-MFE-4B

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

    International Nuclear Information System (INIS)

    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

  4. Analytical calculations of multiple scattering for high energy photons and neutrons

    International Nuclear Information System (INIS)

    Radiography of large dense objects often require the use of highly penetrating radiation. For example, a couple of centimeters of steel attenuates 50 keV x-rays by a factor of approximately 10-14 whereas this same amount of steel would attenuate a 500 keV photon beam by only a factor of about 0.25. However, this increase in penetrating power comes with a price. In the case of x-radiation there are two bills to pay: (1) For projection radiography, this increase in penetration directly causes a corresponding decrease in resolution. (2) This increase in penetration occurs in a region where the interaction of radiation and matter is changing from absorption to scattering. In the above example the fraction of scattering goes from about 0.1 at 50 keV to over 0.99 at 500 keV. These scattered photons can significantly degrade contrast. In order to overcome some of these difficulties, radiography using scattered photons has been studied by myself and numerous other authors. In all the above cases, calculation of the intensity of scattered radiation is of primary importance. In cases where scattering is probable, multiple scattering can also be probable. Calculations of multiple scattering are generally very difficult and usually require the use of extremely sophisticated Monte Carlo simulations. It is not unusual for these calculations to require several hours of CPU time on some of the worlds largest and fastest supercomputers. In this paper the author presents an alternative approach. He presents an analytical solution to the equations of double scattering, and shows how this solution can be extended to the case of higher order scattering. Finally, he gives numerical examples of these solutions and compares them to solutions obtained by Monte Carlo simulations

  5. 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.

  6. HEXANN-EVALU - a Monte Carlo program system for pressure vessel neutron irradiation calculation

    International Nuclear Information System (INIS)

    The Monte Carlo program HEXANN and the evaluation program EVALU are intended to calculate Monte Carlo estimates of reaction rates and currents in segments of concentric angular regions around a hexagonal reactor-core region. The report describes the theoretical basis, structure and activity of the programs. Input data preparation guides and a sample problem are also included. Theoretical considerations as well as numerical experimental results suggest the user a nearly optimum way of making use of the Monte Carlo efficiency increasing options included in the program

  7. Calculations of Branching Ratios for Radiative-Capture, One-Proton, and Two-Neutron Channels in the Fusion Reaction $^{209}$Bi+$^{70}$Zn

    CERN Document Server

    Ichikawa, Takatoshi; 10.1143/JPSJ.79.074201

    2010-01-01

    We discuss the possibility of the non-one-neutron emission channels in the cold fusion reaction $^{70}$Zn + $^{209}$Bi to produce the element Z=113. For this purpose, we calculate the evaporation-residue cross sections of one-proton, radiative-capture, and two-neutron emissions relative to the one-neutron emission in the reaction $^{70}$Zn + $^{209}$Bi. To estimate the upper bounds of those quantities, we vary model parameters in the calculations, such as the level-density parameter and the height of the fission barrier. We conclude that the highest possibility is for the 2n reaction channel, and its upper bounds are 2.4$%$ and at most less than 7.9% with unrealistic parameter values, under the actual experimental conditions of [J. Phys. Soc. Jpn. {\\bf 73} (2004) 2593].

  8. The neutronic calculations for some fluids, libraries and structural materials in a hybrid reactor system

    Energy Technology Data Exchange (ETDEWEB)

    Guenay, Mehtap [Inoenue Univ., Malatya (Turkey). Physics Dept.

    2015-11-15

    In the present investigation, a hybrid reactor system was designed by using 100 % Flibe, 90 % Flibe-10 % ThF{sub 4}, 90 % Flibe-10 % UF{sub 4} fluids, ENDF/B-VII, JEFF-3.1, JENDL-4.0, ROSFOND, BROND-2.2, CENDL-3.1 evaluated nuclear data libraries and Ferritic Steel, 9Cr2WVTa, V4Cr4Ti, SiC structural materials. The fluids were used in the liquid first wall, liquid second wall (blanket) and shield zones of a fusion-fission hybrid reactor system. The nuclear parameters of a fusion-fission hybrid reactor such as tritium breeding ratio (TBR), energy multiplication factor (M), heat deposition rate were computed in liquid first wall, blanket and shield zones. Three-dimensional nucleonic calculations were performed using the most recent version MCNPX-2.7.0 the Monte Carlo code.

  9. The neutronic calculations for some fluids, libraries and structural materials in a hybrid reactor system

    International Nuclear Information System (INIS)

    In the present investigation, a hybrid reactor system was designed by using 100 % Flibe, 90 % Flibe-10 % ThF4, 90 % Flibe-10 % UF4 fluids, ENDF/B-VII, JEFF-3.1, JENDL-4.0, ROSFOND, BROND-2.2, CENDL-3.1 evaluated nuclear data libraries and Ferritic Steel, 9Cr2WVTa, V4Cr4Ti, SiC structural materials. The fluids were used in the liquid first wall, liquid second wall (blanket) and shield zones of a fusion-fission hybrid reactor system. The nuclear parameters of a fusion-fission hybrid reactor such as tritium breeding ratio (TBR), energy multiplication factor (M), heat deposition rate were computed in liquid first wall, blanket and shield zones. Three-dimensional nucleonic calculations were performed using the most recent version MCNPX-2.7.0 the Monte Carlo code.

  10. Spatial homogenization methods for pin-by-pin neutron transport calculations

    Science.gov (United States)

    Kozlowski, Tomasz

    For practical reactor core applications low-order transport approximations such as SP3 have been shown to provide sufficient accuracy for both static and transient calculations with considerably less computational expense than the discrete ordinate or the full spherical harmonics methods. These methods have been applied in several core simulators where homogenization was performed at the level of the pin cell. One of the principal problems has been to recover the error introduced by pin-cell homogenization. Two basic approaches to treat pin-cell homogenization error have been proposed: Superhomogenization (SPH) factors and Pin-Cell Discontinuity Factors (PDF). These methods are based on well established Equivalence Theory and Generalized Equivalence Theory to generate appropriate group constants. These methods are able to treat all sources of error together, allowing even few-group diffusion with one mesh per cell to reproduce the reference solution. A detailed investigation and consistent comparison of both homogenization techniques showed potential of PDF approach to improve accuracy of core calculation, but also reveal its limitation. In principle, the method is applicable only for the boundary conditions at which it was created, i.e. for boundary conditions considered during the homogenization process---normally zero current. Therefore, there exists a need to improve this method, making it more general and environment independent. The goal of proposed general homogenization technique is to create a function that is able to correctly predict the appropriate correction factor with only homogeneous information available, i.e. a function based on heterogeneous solution that could approximate PDFs using homogeneous solution. It has been shown that the PDF can be well approximated by least-square polynomial fit of non-dimensional heterogeneous solution and later used for PDF prediction using homogeneous solution. This shows a promise for PDF prediction for off

  11. 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

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

    International Nuclear Information System (INIS)

    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 Univ. of Mainz (DE), 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-a-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 7Li(n,a)3H-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 and 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 speculate on

  13. 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

  14. The sensitivity of calculated critical masses of small fast systems to changes in the U236 and U238 neutron-scattering data

    International Nuclear Information System (INIS)

    The sensitivity of the calculated critical masses of a number of simple systems, to changes in the basic neutron scattering data, have been investigated. The systems considered are spheres of 29% U235 and 93.5% U235, both bare, and reflected by thick natural uranium. The calculations have been carried out using the Carlson Sn method with 4 energy groups, and the percentage changes in the calculated critical masses of the different systems, due to specified changes in the various aspects of the neutron scattering data, have been obtained. The results are presented and discussed with particular reference to the adjustment of the basic data to give good agreement with experimental critical sizes. The basic data on which these calculations have been based are those given in AWRE Report 0-28/60. (author)

  15. 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, ...

  16. Computer program /P1-GAS/ calculates the P-0 and P-1 transfer matrices for neutron moderation in a monatomic gas

    Science.gov (United States)

    Collier, G.; Gibson, G.

    1968-01-01

    FORTRAN 4 program /P1-GAS/ calculates the P-O and P-1 transfer matrices for neutron moderation in a monatomic gas. The equations used are based on the conditions that there is isotropic scattering in the center-of-mass coordinate system, the scattering cross section is constant, and the target nuclear velocities satisfy a Maxwellian distribution.

  17. Calculation of self-shielding coefficients, flux depression and cadmium factor for thermal neutron flux measurement of the IPEN/MB-01 reactor

    International Nuclear Information System (INIS)

    A calculation methodology of Flux Depression, Self-Shielding and Cadmium Factors is presented, using the ANISN code, for experiments conducted at the IPEN/MB-01 Research Reactor. The correction factors were determined considering thermal neutron flux and 0.125 e 0.250 mm diameter of 197 Au wires. (author)

  18. 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.

  19. 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.

  20. 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.

  1. 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

  2. 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.

  3. The impact of heavy metals from environmental tobacco smoke on indoor air quality as determined by Compton suppression neutron activation analysis.

    Science.gov (United States)

    Landsberger, S; Wu, D

    1995-12-01

    The method of instrumental neutron activation analysis (NAA) has been improved for air filter samples in the determination of low level heavy metals in indoor air. By using the techniques of epithermal neutron irradiation in conjunction with Compton suppression, the detection limits of cadmium, arsenic and antimony measurements have been dramatically reduced to 2 ng for Cd, 0.2 ng for As, and 0.03 ng for Sb. The determination of these heavy metals in particulate material generated from cigarette smoking in indoor environments has been conducted. Other elements, Br, Cl, Na, K, Zn were also found at elevated levels. PMID:8560226

  4. Using neutron powder diffraction and first-principles calculations to understand the working mechanisms of porous coordination polymer sorbents.

    Science.gov (United States)

    Chevreau, Hubert; Duyker, Samuel G; Peterson, Vanessa K

    2015-12-01

    Metal-organic frameworks (MOFs) are promising solid sorbents, showing gas selectivity and uptake capacities relevant to many important applications, notably in the energy sector. To improve and tailor the sorption properties of these materials for such applications, it is necessary to gain an understanding of their working mechanisms at the atomic and molecular scale. Specifically, it is important to understand how features such as framework porosity, topology, chemical functionality and flexibility underpin sorbent behaviour and performance. Such information is obtained through interrogation of structure-function relationships, with neutron powder diffraction (NPD) being a particularly powerful characterization tool. The combination of NPD with first-principles density functional theory (DFT) calculations enables a deep understanding of the sorption mechanisms, and the resulting insights can direct the future development of MOF sorbents. In this paper, experimental approaches and investigations of two example MOFs are summarized, which demonstrate the type of information and the understanding into their functional mechanisms that can be gained. Such information is critical to the strategic design of new materials with targeted gas-sorption properties. PMID:26634721

  5. 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.

  6. Verification of Monte Carlo Calculations by Means of Neutron and Gamma Fluence Spectra Measurements behind and inside of Iron-Water Configurations

    International Nuclear Information System (INIS)

    Neutron and gamma spectra were measured behind and inside of modules consisting of variable iron and water slabs that were installed in radial beams of the zero-power training and research reactors AKR of the Technical University Dresden and ZLFR of the University of Applied Sciences Zittau/Goerlitz. The applied NE-213 scintillation spectrometer did allow the measurement of gamma and neutron fluence spectra in the energy regions 0.3-10 MeV for photons and 1.0-20 MeV for neutrons. The paper describes the experiments and presents important results of the measurements. They are compared with the results of Monte Carlo transport calculations made by means of the codes MCNP and TRAMO on an absolute scale of fluences

  7. 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.

  8. Development calculational procedures for the neutron physics design of advanced pressurized water reactors (APWR) with tight triangular lattices in hexagonal fuel assemblies

    International Nuclear Information System (INIS)

    The new procedures for the calculation of infinite reactor zones build a synthesis of wellknown fast breeder (FBR) and light water reactor (LWR) methods. The data libraries are based on the 69 energy group structure of the WIMS code for thermal reactors and use the flexible storage mode of the FBR libraries. For the calculation of effective cross sections in the energy of neutron resonances, being very important in the APWR with its strongly epithermal neutron spectrum, several options are available. In most applications improved selfshielding tabulation formalisms (f-factor concept) are used. For more accurate investigations the fine flux programs ULFISP (own development) or RESABK (IKE, Stuttgart) may be selected. All cross section calculations use a modified version of the FBR code GRUCAL. Particularly the calculation of voided lattices may be improved at 69 groups with the program REMOCO or with a new 334 group library. The new calculational procedures could be qualified for a large number of LWR, APWR and FBR applications. The fuel assembly and whole core calculations are performed with available FBR methods. A new reactor core simulation program ARCOSI has been developed for the investigation of an APWR equilibrium core. The required cross sections come from fast interpolations of fuel assembly data on code-own libraries. The whole core calculations are performed with the fast nodal code HEXNODK, adopted from KWU. All calculational procedures are part of the powerful FBR code system KAPROS. (orig.)

  9. 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

  10. A method to calculate fission-fragment yields Y(Z,N) versus proton and neutron number in the Brownian shape-motion model. Application to calculations of U and Pu charge yields

    Energy Technology Data Exchange (ETDEWEB)

    Moeller, Peter [Los Alamos National Laboratory, Theoretical Division, Los Alamos, NM (United States); Ichikawa, Takatoshi [Kyoto University, Yukawa Institute for Theoretical Physics, Kyoto (Japan)

    2015-12-15

    We propose a method to calculate the two-dimensional (2D) fission-fragment yield Y(Z,N) versus both proton and neutron number, with inclusion of odd-even staggering effects in both variables. The approach is to use the Brownian shape-motion on a macroscopic-microscopic potential-energy surface which, for a particular compound system is calculated versus four shape variables: elongation (quadrupole moment Q{sub 2}), neck d, left nascent fragment spheroidal deformation ε{sub f1}, right nascent fragment deformation ε{sub f2} and two asymmetry variables, namely proton and neutron numbers in each of the two fragments. The extension of previous models 1) introduces a method to calculate this generalized potential-energy function and 2) allows the correlated transfer of nucleon pairs in one step, in addition to sequential transfer. In the previous version the potential energy was calculated as a function of Z and N of the compound system and its shape, including the asymmetry of the shape. We outline here how to generalize the model from the ''compound-system'' model to a model where the emerging fragment proton and neutron numbers also enter, over and above the compound system composition. (orig.)

  11. Nodal model for calculating the variations in neutron flux density due to stochastic vibrations of control elements of hexagonal cross section

    International Nuclear Information System (INIS)

    Based on a three-dimensional modal geometry model for the WWER 440 reacotr, with nodes in the hexagonal z geometry, the equations for the interative calculation of the mean neutron flux density in a node and their variations due to stochastic control element vibration are shown. For modelling sources of noise, two different geometric and neutron-physics equations are used, according to the design of a control element as a spatial double pendulum with the absorber and fuel part. The neutron flux noise caused by vibration of the fuel parts is due to area sources. These are induced by material parameter variation due to control element displacement within the guide duct. The model of the 'thermal black body' absorbing hollow cylinder is transferred to bodies of hexagonal crossection for the absorber part. Both sources of noise are described as disturbances for the partial neutron current densities averaged over the node surfaces in the two group diffusion approximation. The transfer of the noise signals is dealt with in the prompt response approximation. The 'two group swelling nodes' are coupled to the 'one group transmission nodes' on the basis of the modified one group diffusion approximation. The algorithms shown are the basis for development of a computer program for examining the transfer functions depending on location of neutron flux density variations with stochastic control element vibrations as the source of noise. (orig./HP)

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

    International Nuclear Information System (INIS)

    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 PbF2 reflector, the epithermal neutron flux at the filter/moderator exit can reach ∼8.5 x 109 n/cm2/s. When the collimator is added, the epithermal neutron beam intensity at the beam exit is reduced to 3 x 109 n/cm2/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)

  13. 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

  14. 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)

  15. A methodology for benefit assessment of using in-core neutron detector signals in core protection calculator system (CPCS) for Korea standard nuclear power plants (KSNPP)

    International Nuclear Information System (INIS)

    Core Protection Calculator System (CPCS) is a digital computer based safety system generating trip signals based on the calculation of departure from nucleate boiling ratio (DNBR) and local power density (LPD). Currently, CPCS uses ex-core detector signals for core power calculation and it has some uncertainties. In this work, a quantitative economic benefit assessment of using in-core neutron detector signals is carried out. In-core detector signals which directly measure the inside neutron flux of core are applied to CPCS to obtain more accurate power distribution profile, DNBR and LPD to reduce the calculation uncertainties. In order to improve axial power distribution calculation, piecewise cubic spline method is applied. Simulation is also carried out to verify its applicability to power distribution calculation in this work. Simulation result shows that the improved method reduces the calculational uncertainties significantly and it allows larger operational margin. It is also assured that no power reduction is required while Core Operating Limit Supervisory System (COLSS) is out-of-service when the improved method is applied

  16. Measurement of neutron activation cross sections for major elements of water, air and soil between 30 and 70 MeV

    International Nuclear Information System (INIS)

    Neutron activation cross sections between 30 and 70 MeV were measured by the activation method using a semi-monoenergetic neutron field settled at the AVF cyclotron of the Cyclotron and Radioisotope Center (CYRIC), Tohoku University. Natural samples of N, O, Si, Na, Ca and Mg which are the major elements of water, air and soil were irradiated in this neutron field generated through the 7Li(p,n)7Be reaction by 30, 35, 40, 50, 60 and 70 MeV protons on thin Li target. Neutron yields were measured with the time-of-flight method using a calibrated NE213 organic liquid scintillator. From the induced activities measured with the HPGe detectors, we estimated the excitation functions of 15 cross sections. (author)

  17. Gamma-ray exposure from neutron-induced radionuclides in soil in Hiroshima and Nagasaki based on DS02 calculations.

    Science.gov (United States)

    Imanaka, Tetsuji; Endo, Satoru; Tanaka, Kenichi; Shizuma, Kiyoshi

    2008-07-01

    As a result of joint efforts by Japanese, US and German scientists, the Dosimetry System 2002 (DS02) was developed as a new dosimetry system, to evaluate individual radiation dose to atomic bomb survivors in Hiroshima and Nagasaki. Although the atomic bomb radiation consisted of initial radiation and residual radiation, only initial radiation was reevaluated in DS02 because, for most survivors in the life span study group, the residual dose was negligible compared to the initial dose. It was reported, however, that there were individuals who entered the city at the early stage after the explosion and experienced hemorrhage, diarrhea, etc., which were symptoms of acute radiation syndrome. In this study, external exposure due to radionuclides induced in soil by atomic bomb neutrons was reevaluated based on DS02 calculations, as a function of both the distance from the hypocenters and the elapsed time after the explosions. As a result, exposure rates of 6 and 4 Gy h(-1) were estimated at the hypocenter at 1 min after the explosion in Hiroshima and Nagasaki, respectively. These exposure rates decreased rapidly by a factor of 1,000 1 day later, and by a factor of 1 million 1 week later. Maximum cumulative exposure from the time of explosion was 1.2 and 0.6 Gy at the hypocenters in Hiroshima and Nagasaki, respectively. Induced radiation decreased also with distance from the hypocenters, by a factor of about 10 at 500 m and a factor of three to four hundreds at 1,000 m. Consequently, a significant exposure due to induced radiation is considered feasible to those who entered the area closer to a distance of 1,000 m from the hypocenters, within one week after the bombing. PMID:18368418

  18. Calculation of water/air stopping-power ratios using EGS4 with explicit treatment of electron-positron differences

    International Nuclear Information System (INIS)

    Using the EGS4 Monte Carlo simulation program, a general purpose code has been written to calculate Bragg--Gray and Spencer--Attix stopping-power ratios for use in radiation dosimetry. The stopping-power ratios can be calculated in any material in any region in a general cylindrical geometry with a large number of source geometries possible. The calculations take into account for the first time the differences between the stopping powers and the inelastic scattering of positrons and electrons. The results show that previous calculations ignoring these effects were accurate. The present results agree, typically within 0.1%, with the Spencer--Attix water-to-air stopping-power ratios for broad parallel beams of electrons given in the AAPM and IAEA protocols except at the surface where the present calculations follow the buildup of secondary electrons in more detail and see a 2% reduction in the stopping-power ratios

  19. Dosimetry intercomparisons between fast neutron radiotherapy facilities

    International Nuclear Information System (INIS)

    Neutron dosimetry intercomparisons have been made between M.D. Anderson Hospital and Tumor Institute, Naval Research Laboratory, University of Washington Hospital, and Hammersmith Hospital. The parameters that are measured during these visits are: tissue kerma in air, tissue dose at depth of dose maximum, depth dose, beam profiles, neutron/gamma ratios and photon calibrations of ionization chambers. A preliminary report of these intercomparisons will be given including a comparison of the calculation and statement of tumor doses for each institution

  20. A new calculation method adapted to the experimental conditions for determining samples γ-activities induced by 14 MeV neutrons

    Science.gov (United States)

    Rzama, A.; Erramli, H.; Misdaq, M. A.

    1994-09-01

    Induced gamma-activities of different disk shaped irradiated samples and standards with 14 MeV neutrons have been determined by using a Monte Carlo calculation method adapted to the experimental conditions. The self-absorption of the multienergetic emitted gamma rays has been taken into account in the final samples activities. The influence of the different activation parameters has been studied. Na, K, Cl and P contents in biological (red beet) samples have been determined.

  1. Use of different programs for calculating the flux density of neutrons activating sodium in the secondary circuit of a NPP with the BN-600 reactor

    International Nuclear Information System (INIS)

    Possibilities of application of the RADAR, TVK-2D and MMKFK program complexes to calculate the BN-600 type reactor shields are analyzed. TVK-2D program (ALGOL-DDR, BESM-6 computer) is designed for two-dimensional calculations of reactors in diffusion multigroup finite-difference approximation using classical and unified perturbation theory. The RADAR system (FORTRAN-4, BESM-6 computer) realizes Boltzmann equation solution by iterative synthesis method in multigroup diffusion approximation. The MMKFK complex (FORTRAN, BESM-6 computer) is used to calculate radiation transport in reactors and cells. The complex is improved: at large ratioes of neutron flux attenuation the methods of splitting and roulette are realized. Calculational results of the integral by energy and mean by zones values of neutron flux density in radial shield and sodium activity in the secondary coolant circuits are presented. Good conformity of the data obtained is pointed out. Conclusion is made about the applicability of the program systems investigated to calculate fast reactor shields at different stages of design. The RADAR system due to its quick operation will be more efficient at the initial stages, while the MMKFK system - at final ones, when high accuracy of calculation is required

  2. 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

  3. Diagnostic and Impact Estimation of Nuclear Data Inconsistencies on Energy Deposition Calculations in Coupled Neutron-Photon Monte-Carlo Simulation, with TRIPOLI-4®

    Science.gov (United States)

    Péron, A.; Malouch, F.; Zoia, A.; Diop, C. M.

    2014-06-01

    Nuclear heating evaluation by Monte-Carlo simulation requires coupled neutron-photon calculation so as to take into account the contribution of secondary photons. Nuclear data are essential for a good calculation of neutron and photon energy deposition and for secondary photon generation. However, a number of isotopes of the most common nuclear data libraries happen to be affected by energy and/or momentum conservation errors concerning the photon production or inaccurate thresholds for photon emission sections. In this paper, we perform a comprehensive survey of the three evaluations JEFF3.1.1, JEFF3.2T2 (beta version) and ENDF/B-VII.1, over 142 isotopes. The aim of this survey is, on the one hand, to check the existence of photon production data by neutron reaction and, on the other hand, to verify the consistency of these data using the kinematic limits method recently implemented in the TRIPOLI-4 Monte-Carlo code, developed by CEA (Saclay center). Then, the impact of these inconsistencies affecting energy deposition scores has been estimated for two materials using a specific nuclear heating calculation scheme in the context of the OSIRIS Material Testing Reactor (CEA/Saclay).

  4. VIWI, Neutron Speeds and Weights for Scattering Kernel Calculation. BASKER, Isotropic Scattering Kernel Calculation Using VIWI. FLAKER, Legendre Moments from Scattering Law Tables

    International Nuclear Information System (INIS)

    1 - Description of problem or function: These three programs are concerned with scattering kernel calculations using energy points. The calculations are made using the speed rather than the energy variable as the functions considered vary generally more smoothly in the former variable. VIWI calculates the speed points and associated weights for all energy intervals. BASKER calculates the isotropic scattering kernel using the points and weights calculated by VIWI. FLAKER calculates the desired Legendre moments of the scattering kernel. 2 - Method of solution: The variation of flux between energy points is accounted for using a Lagrange interpolation method. The speed points have been chosen with the suitability of Pu systems as a major goal. The calculated points and weights are written on file by VIWI, as well as the ordinary Gaussian points and weights. BASKER uses the free gas or Brown-St. John model to calculate the isotropic scattering kernel. FLAKER calculates the Legendre moments using any tabulated scattering law (given as input)

  5. 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.

  6. 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.

  7. 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.

  8. REFRACTIVE NEUTRON LENS

    OpenAIRE

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

    2013-01-01

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

  9. 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.

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

    OpenAIRE

    Kaicun Wang; Chunlüe Zhou

    2015-01-01

    Global analyses of surface mean air temperature (T m ) are key datasets for climate change studies and provide fundamental evidences for global warming. However, the causes of regional contrasts in the warming rate revealed by such datasets, i.e., enhanced warming rates over the northern high latitudes and the “warming hole” over the central U.S., are still under debate. Here we show these regional contrasts depend on the calculation methods of T m . Existing global analyses calculate T m fro...

  11. 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.

  12. 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...

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

    CERN Document Server

    Ferrari, A; Rancati, T

    2001-01-01

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

  14. Carbon and oxygen fluxes in the Barents and Norwegian Seas : production, air-sea exchange and budget calculations

    OpenAIRE

    Kivimäe, Caroline

    2007-01-01

    This thesis focus on the carbon and oxygen fluxes in the Barents and Norwegian Seas and presents four studies where the main topics are variability of biological production, air-sea exchange and budget calculations. The world ocean is the largest short term reservoir of carbon on Earth, consequently it has the potential to control the atmospheric concentrations of carbon dioxide (CO2) and has already taken up ~50 % of the antropogenically emitted CO2. It is thus important to...

  15. Calculation software for efficiency and penetration of a fibrous filter medium based on the mathematical models of air filtration

    OpenAIRE

    Kouropoulos, Giorgos

    2014-01-01

    At this article will be created a software written in visual basic for efficiency and penetration calculation in a fibrous filter medium for given values of particles diameter that are retained in the filter. Initially, will become report of mathematical models of air filtration in fibrous filters media and then will develop the code and the graphical interface of application, that are the base for software creation in the visual basic platform.

  16. Time-resolved fast-neutron radiography of air-water two-phase flows in a rectangular channel by an improved detection system

    OpenAIRE

    Zboray, Robert; Dangendorf, Volker; Mor, Ilan; Bromberger, Benjamin; Tittelmeier, Kai

    2015-01-01

    In a previous work we have demonstrated the feasibility of high-frame-rate, fast-neutron radiography of generic air-water two-phase flows in a 1.5 cm thick, rectangular flow channel. The experiments have been carried out at the high-intensity, white-beam facility of the Physikalisch-Technische Bundesanstalt, Germany, using an multi-frame, time-resolved detector developed for fast neutron resonance radiography. The results were however not fully optimal and therefore we have decided to modify ...

  17. 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.

  18. 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.

  19. Fine-mesh deterministic modeling of PWR fuel assemblies: Proof-of-principle of coupled neutronic/thermal–hydraulic calculations

    International Nuclear Information System (INIS)

    Highlights: • We implemented a fine-mesh coupled neutronic/thermal–hydraulic tool. • A CFD approach is used together with the multi-group neutron diffusion approximation. • Temperature-dependent cross-sections are generated with a Monte Carlo method. • We applied the tool to a simplified PWR fuel assembly. • Discrepancies in multiplication factor are seen against radial coarse-mesh averaging. - Abstract: This paper investigates the feasibility of developing a fine mesh coupled neutronic/thermal–hydraulic solver within the same computing platform for selected fuel assemblies in nuclear cores. As a first step in this developmental work, a Pressurized Water Reactor at steady-state conditions was considered. The system being simulated has a finite axial size, but is infinite in the radial direction. The platform used for the modeling is based on the open source C++ library OpenFOAM. The thermal–hydraulics is solved using the built-in SIMPLE algorithm for the mass and momentum fields of the fluid, complemented by an equation for the temperature field applied simultaneously to all the regions (i.e. fluid and solid structures). For the neutronics, a two-group neutron diffusion-based solver was developed, with sets of macroscopic cross-sections generated by the Monte Carlo code SERPENT. The meshing of the system was created by the open source software SALOME. Successful convergence of the neutronic and thermal–hydraulic fields was achieved, thus bringing the solution of the coupled problem to an unprecedented level of details. Most importantly, the true interdependence of the different fields is automatically guaranteed at all scales. In addition, comparisons with a coarse-mesh radial averaging of the thermal–hydraulic variables show that a coarse-mesh fuel temperature identical for all fuel pins can lead to discrepancies of up to 0.5% in pin powers, and of several tens of pcm in multiplication factor

  20. 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.

  1. Application of the Synthesis method to the calculations of neutron flow in 3D in the enveloping of a BWR reactor with the DORT code

    International Nuclear Information System (INIS)

    The surveillance program of the vessel materials of a BWR reactor requires the determination of the neutron flux in 3D in the core enveloping. To carry out these calculations of the neutron flux, the Regulatory Guide 1.190 of the NRC recommends the use of the following codes: MCNP, TORT and DORT. In the case of using the DORT code, the one which solves the transport equation in discreet coordinates and in two dimensions (xy, rθ, and rz), the regulatory guide in reference, requires to make an approach of the flow in three dimensions by means of the call Synthesis Method. It is denominated like this due to that a flow representation in 3D is achieved 'combining' or 'synthesizing' the calculated flows by DORT in rθ, rz and r. In this work the application of the Synthesis Method it is presented, according to the Regulatory Guide 1.190, to determine the 3D flows in a BWR reactor. To achieve the above mentioned it was implemented the Synthesis Method in a computer program developed in the ININ to which is denominated SYNTHESIS. This program applies the synthesis method, and is 'coupled' with the DORT code to determine by this way the neutronic fluxes in 3D on the enveloping of a BWR reactor. (Author)

  2. A Method to Calculate Fission-Fragment Yields $Y(Z,N)$ versus Proton and Neutron Number in the Brownian Shape-Motion Model. Application to calculations of U and Pu charge yields

    CERN Document Server

    Moller, P

    2015-01-01

    We propose a method to calculate the two-dimensional (2D) fission-fragment yield $Y(Z,N)$ versus both proton and neutron number, with inclusion of odd-even staggering effects in both variables. The approach is to use Brownian shape-motion on a macroscopic-microscopic potential-energy surface which, for a particular compound system is calculated versus four shape variables: elongation (quadrupole moment $Q_2$), neck $d$, left nascent fragment spheroidal deformation $\\epsilon_{\\rm f1}$, right nascent fragment deformation $\\epsilon_{\\rm f2}$ and two asymmetry variables, namely proton and neutron numbers in each of the two fragments. The extension of previous models 1) introduces a method to calculate this generalized potential-energy function and 2) allows the correlated transfer of nucleon pairs in one step, in addition to sequential transfer. In the previous version the potential energy was calculated as a function of $Z$ and $N$ of the compound system and its shape, including the asymmetry of the shape. We ou...

  3. 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.

  4. Calculation of dose equivalent index, effective dose equivalent and ambient dose equivalent for the giant resonance neutron spectra produced at an electron accelerator

    International Nuclear Information System (INIS)

    The ANISN code has been used in this study to evaluate the attenuation of neutron beams of various spectra incident normally on slabs of different kinds of concrete. Spectra of the most common sources (Am-Be and Cf-252) and those of giant resonance neutrons, produced at electron accelerators, were studied. The concretes examined had densities between 2.1 and 4.64 g.cm-3. The calculation were made in terms of the deep dose equivalent index, the effective dose equivalent and the ambient dose equivalent. Values of attenuation length in the various materials were derived from the attenuation curves. The results found should allow for useful evaluations in every day practice for health physicist

  5. Measurements of Neutron and Gamma Attenuation in Massive Laminated Shields of Concrete and a Study of the Accuracy of some Methods of Calculation

    International Nuclear Information System (INIS)

    Extensive neutron and gamma attenuation measurements have been performed in magnetite and ordinary concrete up to a depth of 2 metres in order to study the accuracy attainable by some shield calculation methods. The effect of thin, heavy layers (Pb) has also been studied. Experimental facilities and instrumentation, especially the foil detection methods used for thermal and epithermal neutrons, are described in some detail. Great weight is laid upon a thorough error analysis. The fluxes measured are compared to those calculated by an earlier version of the British 18-group removal method (RASH B3), by an improved removal method (NRN) developed at AB Atomenergi, and by numerical integration of the Boltzmann equation (NIOBE). The results show that shielding calculations with the newer methods give fluxes that are generally within a factor of 2-3 from the true values. A greater accuracy seems to be difficult to obtain in practice in spite of possible improvements in the mathematical solution of the transport problem. The greatest errors originate in the translation between the true and calculation geometries in the uncertainty of material properties in the case of concrete, and in approximations and inaccuracies of radiation sources

  6. Calculations of neutron and proton induced reaction cross sections for actinides in the energy region from 10MeV to 1GeV

    International Nuclear Information System (INIS)

    Several nuclear model codes were applied to calculations of nuclear data in the energy region from 10MeV to 1GeV. At energies up to 100MeV the nuclear theory code GNASH was used for nuclear data calculation for neutrons incident for on 238U, 233-236U, 238-242Pu, 237Np, 232Th, 241-243Am and 242-247Cm. At energies from 100MeV to 1GeV the intranuclear cascade exciton model including the fission process was applied to calculations of protons and neutrons with 233U, 235U, 238U, 232Th, 232Pa, 237Np, 238Np, 239Pu, 241Am, 242Am and 242-248Cm. Determination of parameter systematics was a major effort in the present work that was aimed at improving the predictive capability of the models used. An emphasis was placed upon a simultaneous analysis of data for a variety of reaction channels for the nuclei considered, as well as of data that are available for nearby nuclei or for other incident particles. Comparisons with experimental data available on multiple reaction cross sections, isotope yields, fission cross sections, particle multiplicities, secondary particle spectra, and double differential cross sections indicate that the calculations reproduce the trends, and often the details, of the measurements data. (author) 82 refs

  7. Biomonitoring of air pollution with heavy metals in the Republic of Macedonia by neutron activation analysis

    International Nuclear Information System (INIS)

    Atmospheric deposition of trace metals was studied over the entire territory of the Republic of Macedonia in 2002 and 2005. Samples of the terrestrial mosses Hypnum cupressiforme, Camptothecium lutescens, and Homalothecium sericeum were collected at 73 sites in 2002 and at 72 sites in 2005. Instrumental neutron activation analysis allowed determination of 41 elements in 2002 and 38 elements in 2005. Principal component factor analysis was used to identify the most polluted areas and characterize different pollution sources. The most important sources of trace metal deposition are ferrous and non-ferrous smelters, oil refineries, fertilizer production plants, and central heating station. four areas appear to be particularly exposed to metal pollution: Veles, Skopje, Tetovo, and Kavadarci - Negotino. Comparison of the results from the first and the second moss survey, showed that there is no significant difference in median values of elemental concentrations determined, besides Ni. The more then twofold increase of its median value is explained by renewed activity of ferronickel smelter in Kavadartsi. (Author)

  8. A numerical scheme to calculate temperature and salinity dependent air-water transfer velocities for any gas

    Directory of Open Access Journals (Sweden)

    M. T. Johnson

    2010-02-01

    Full Text Available The transfer velocity determines the rate of exchange of a gas across the air-water interface for a given deviation from Henry's law equilibrium between the two phases. In the thin film model of gas exchange, which is commonly used for calculating gas exchange rates from measured concentrations of trace gases in the atmosphere and ocean/freshwaters, the overall transfer is controlled by diffusion-mediated films on either side of the air-water interface. Calculating the total transfer velocity (i.e. including the influence from both molecular layers requires the Henry's law constant and the Schmidt number of the gas in question, the latter being the ratio of the viscosity of the medium and the molecular diffusivity of the gas in the medium. All of these properties are both temperature and (on the water side salinity dependent and extensive calculation is required to estimate these properties where not otherwise available. The aim of this work is to standardize the application of the thin film approach to flux calculation from measured and modelled data, to improve comparability, and to provide a numerical framework into which future parameter improvements can be integrated. A detailed numerical scheme is presented for the calculation of the gas and liquid phase transfer velocities (ka and kw respectively and the total transfer velocity, K. The scheme requires only basic physical chemistry data for any gas of interest and calculates K over the full range of temperatures, salinities and wind-speeds observed in and over the ocean. Improved relationships for the wind-speed dependence of ka and for the salinity-dependence of the gas solubility (Henry's law are derived. Comparison with alternative schemes and methods for calculating air-sea flux parameters shows good agreement in general but significant improvements under certain conditions. The scheme is provided as a downloadable

  9. Mass Properties Calculation and Fuel Analysis in the Conceptual Design of Uninhabited Air Vehicles

    OpenAIRE

    Ohanian, Osgar John

    2003-01-01

    The determination of an aircraft's mass properties is critical during its conceptual design phase. Obtaining reliable mass property information early in the design of an aircraft can prevent design mistakes that can be extremely costly further along in the development process. In this thesis, several methods are presented in order to automatically calculate the mass properties of aircraft structural components and fuel stored in tanks. The first method set forth calculates the mass prope...

  10. Integrated system for production of neutronics and photonics calculational constants. Major neutron-induced interactions (Z less than or equal to 55): graphical, experimental data

    International Nuclear Information System (INIS)

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

  11. Simulation based energy consumption calculation of an office building using solar-assisted air conditioning

    OpenAIRE

    Thomas, Sébastien; Andre, Philippe

    2008-01-01

    To minimize environmental impact and CO2 production associated with air-conditioning system operation, it is reasonable to evaluate the prospects of a clean energy source. The targets of the study are to evaluate cooling energy consumption to maintain thermal comfort in an office building and to point out solar energy to satisfy these cooling needs. Simulations were carried out with three different cooling systems in the same operating conditions to determine as accurately as possible the pot...

  12. 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.

  13. Tests of Neutron Spectrum Calculations with the Help of Foil Measurements in a D2O and in an H2O-Moderated Reactor and in Reactor Shields of Concrete an Iron

    International Nuclear Information System (INIS)

    Foil measurements covering the fast, epithermal and thermal neutron energy regions have been made in the centre of the Swedish D2O-moderated reactor R1, in the pool reactor R2-0, and in different positions in reactor shields of iron, magnetite concrete and ordinary concrete. Neutron spectra have also been calculated for most of these positions, often with the help of a numerical integration of the Boltzmann equation. The measurements and the calculated spectra are presented

  14. Cronos 2: a neutronic simulation software for reactor core calculations; Cronos 2: un logiciel de simulation neutronique des coeurs de reacteurs

    Energy Technology Data Exchange (ETDEWEB)

    Lautard, J.J.; Magnaud, C.; Moreau, F.; Baudron, A.M. [CEA Saclay, Dept. de Mecanique et de Technologie (DMT/SERMA), 91 - Gif-sur-Yvette (France)

    1999-07-01

    The CRONOS2 software is that part of the SAPHYR code system dedicated to neutronic core calculations. CRONOS2 is a powerful tool for reactor design, fuel management and safety studies. Its modular structure and great flexibility make CRONOS2 an unique simulation tool for research and development for a wide variety of reactor systems. CRONOS2 is a versatile tool that covers a large range of applications from very fast calculations used in training simulators to time and memory consuming reference calculations needed to understand complex physical phenomena. CRONOS2 has a procedure library named CPROC that allows the user to create its own application environment fitted to a specific industrial use. (authors)

  15. 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.

  16. Neutron source for Neutron Capture Synovectomy

    International Nuclear Information System (INIS)

    Monte Carlo calculations were performed to obtain a thermal neutron field from a 239PuBe neutron source inside a cylindrical heterogeneous moderators for Neutron Capture Synovectomy. Studied moderators were light water and heavy water, graphite and heavy water, lucite and polyethylene and heavy water. The neutron spectrum of polyethylene and heavy water moderator was used to determine neutron spectra inside a knee model. In this model the elemental composition of synovium and synovial liquid was assumed like blood. Kerma factors for synovium and synovial liquid were calculated to compare with water Kerma factors, in this calculations the synovium was loaded with two different concentrations of Boron

  17. 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.

  18. Use of a heating test to calculate the air renewable rate in an MV/LV substation; Exploitation d`un essai d`echauffement pour le calcul du renouvellement d`air dans un poste HTA/BT

    Energy Technology Data Exchange (ETDEWEB)

    Duquerroy, P.

    1997-12-31

    In order to determine the behaviour of the ventilation system in a MV/LV distribution substation, results of heating tests at different load levels have been used. Knowing the power generated by the transformer and the switchgear, and after calculating the power dissipated through the walls, it was possible to estimate the power evacuated by ventilation and thus the air flow through the grid. The resulting equations were introduced in the thermal model of the substation using the CLIM 2000 building heat engineering software. Simulation results are in agreement with experimental data

  19. 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.)

  20. Calculation of the neutron induced fission cross-section of 233Pa up to 20 MeV

    International Nuclear Information System (INIS)

    Since very recently, direct measurements of the 233Pa(n,f) cross-section are available in the energy range from 1.0 to 8.5 MeV. This has stimulated a new, self-consistent, neutron cross-section evaluation for the n+233Pa system, in the incident neutron energy range 0.01-20 MeV. Since higher fission chances are involved also the lighter Pa-isotopes had to be re-evaluated in a consistent manner. The results are quite different compared to earlier evaluation attempts. Since 233Pa is a key isotope in the thorium based fuel cycle the quality of its reaction cross-sections is important for the modeling of future advanced fuel and reactor concepts. The present status of the evaluated libraries is that they differ by a factor of two in the absolute fission cross-section and also in the threshold energy value

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

    Science.gov (United States)

    Wang, Kaicun; Zhou, Chunlüe

    2015-01-01

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

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

    Science.gov (United States)

    Wang, Kaicun; Zhou, Chunlüe

    2016-04-01

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

  3. Development of a surrogate model for simplified neutronic calculations involved in the design stage of a thermonuclear fusion reactor

    OpenAIRE

    Martínez Arroyo, Javier

    2012-01-01

    Several system codes have been developed since the eighties, with different objectives and appropriate architecture and level of development, aiming to explore possible operating condition ranges of a fusion power reactor. In some “system codes” technology/engineering assumptions/models (e.g. thermodynamic efficiency of coolant cycle, neutron multiplication coefficient, Tritium Breeding Ratio, radial built) are treated as input data inserted by the user and integrated in a main module esse...

  4. Development of a data base system and concentration calculation for neutron activation analysis as per the k0 method

    International Nuclear Information System (INIS)

    One of the most important nuclear analytical techniques is the neutron activation analysis used to determine which elements and their proportion are included within an analysis sample. A sample is undergone to the procedures of the technique, and the information, which is dispersed, is generated in each phase of this process. Therefore, it is necessary this information should be organized properly for its better use

  5. 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.

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

    Science.gov (United States)

    Benstead, J.; Tostevin, J. A.; Escher, J. E.; Burke, J. T.; Hughes, R. O.; Ota, S.; Casperson, R. J.; Thompson, I. J.

    2016-06-01

    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,t)90Zr reaction, and are compared against measured data at an incident proton energy of 28.56 MeV.

  7. 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)

  8. 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...

  9. SU-E-T-554: Monte Carlo Calculation of Source Terms and Attenuation Lengths for Neutrons Produced by 50–200 MeV Protons On Brass

    Energy Technology Data Exchange (ETDEWEB)

    Ramos-Mendez, J; Faddegon, B [University of California San Francisco, San Francisco, CA (United States); Paganetti, H [Massachusetts General Hospital, Boston, MA (United States)

    2015-06-15

    Purpose: We used TOPAS (TOPAS wraps and extends Geant4 for medical physicists) to compare Geant4 physics models with published data for neutron shielding calculations. Subsequently, we calculated the source terms and attenuation lengths (shielding data) of the total ambient dose equivalent (TADE) in concrete for neutrons produced by protons in brass. Methods: Stage1: The Bertini and Binary nuclear models available in Geant4 were compared with published attenuation at depth of the TADE in concrete and iron. Stage2: Shielding data of the TADE in concrete was calculated for 50– 200 MeV proton beams on brass. Stage3: Shielding data from Stage2 was extrapolated for 235 MeV proton beams. This data was used in a point-line-source analytical model to calculate the ambient dose per unit therapeutic dose at two locations inside one treatment room at the Francis H Burr Proton Therapy Center. Finally, we compared these results with experimental data and full TOPAS simulations. Results: At larger angles (∼130o) the TADE in concrete calculated with the Bertini model was about 9 times larger than that calculated with the Binary model. The attenuation length in concrete calculated with the Binary model agreed with published data within 7%±0.4% (statistical uncertainty) for the deepest regions and 5%±0.1% for shallower regions. For iron the agreement was within 3%±0.1%. The ambient dose per therapeutic dose calculated with the Binary model, relative to the experimental data, was a ratio of 0.93±0.16 and 1.23±0.24 for two locations. The analytical model overestimated the dose by four orders of magnitude. These differences are attributed to the complexity of the geometry. Conclusion: The Binary and Bertini models gave comparable results, with the Binary model giving the best agreement with published data at large angle. Shielding data we calculated using the Binary model is useful for fast shielding calculations with other analytical models. This work was supported by

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

    Directory of Open Access Journals (Sweden)

    Ozdemir Omer Faruk

    2015-01-01

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

  11. 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

  12. Measurements and calculations of neutron leakage spectra from slabs irradiated with 9Be(d, n)10B 2H(d, n)3He and Pu-Be neutrons

    International Nuclear Information System (INIS)

    The spectra of neutrons from the 9Be(d, n)10B, 2H(d, n)3He and Pu-Be sources passing through slabs of water, graphite, Al, Fe and Pb up to 20 cm in thickness were measured by a pulse height response spectrometer in the 1.5-15 MeV range. The measured leakage spectra have been compared with calculated results obtained using the three dimensional Monte-Carlo code MCNP-4A and point-wise cross sections from the ENDF/B-IV, ENDF/B-VI and JENDL-3.1 data files. A comparison of the measured and calculated data has shown that the MCNP-4A code with an appropriate library can reasonably approximate the measured leakage spectra

  13. Measurements and calculations of neutron leakage spectra from slabs irradiated with {sup 9}Be(d, n){sup 10}B {sup 2}H(d, n){sup 3}He and Pu-Be neutrons

    Energy Technology Data Exchange (ETDEWEB)

    Olah, L.; Jordanova, J.; El-Megrab, A.M.; Darsono, Perez N.; Yousif, M.Y.A.; Csikai, J. [Institute of Experimental Physics, Kossuth University, Debrecen (Hungary); Fenyvesi, A.; Majdeddin, A.D. [Institute of Nuclear Research of the Hungarian Academy of Sciences (ATOMKI), Debrecen (Hungary)

    1999-03-01

    The spectra of neutrons from the {sup 9}Be(d, n){sup 10}B, {sup 2}H(d, n){sup 3}He and Pu-Be sources passing through slabs of water, graphite, Al, Fe and Pb up to 20 cm in thickness were measured by a pulse height response spectrometer in the 1.5-15 MeV range. The measured leakage spectra have been compared with calculated results obtained using the three dimensional Monte-Carlo code MCNP-4A and point-wise cross sections from the ENDF/B-IV, ENDF/B-VI and JENDL-3.1 data files. A comparison of the measured and calculated data has shown that the MCNP-4A code with an appropriate library can reasonably approximate the measured leakage spectra.

  14. Calculation of nonequilibrium hydrogen-air reactions with implicit flux vector splitting method

    Science.gov (United States)

    Lee, Seung-Ho; Deiwert, George S.

    1989-01-01

    Two methods, fully- and loosely-coupled, are developed to incorporate nonequilibrium hydrogen-air chemistry into the fluid dynamic implicit flux vector splitting code (F3D). The new code (F3D/Chem) is validated against other existing codes for two cases: nozzle expansion, and shock-induced combustion around a blunt body. The shock-induced combustion case is compared also with an experimental data. The reaction rate constants are varied in an effort to reproduce the experimental data. The fully- and loosely-coupled methods are found to yield comparable results, but the computation time is shorter using the loosely-coupled method. The present method is found to reproduce results obtained using different existing codes. The experimental data was not reproduced with any selected rate coefficients set.

  15. 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.

  16. Integrated system for production of neutronics and photonics calculational constants. Volume XVI. Tabular and graphical presentation of 175 neutron group constants derived from the LLL evaluated neutron data library (ENDL)

    International Nuclear Information System (INIS)

    As of February 3, 1975, 175 neutron group constants had been derived from the Evaluated Nuclear Data Library (ENDL) at LLL. In this volume, tables and graphs of the constants are presented along with the conventions used in their preparation. (U.S.)

  17. 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

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

    Energy Technology Data Exchange (ETDEWEB)

    Morgan C. White

    2000-07-01

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

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

    International Nuclear Information System (INIS)

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

  20. 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)