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Sample records for multigroup neutron dose

  1. Multigroup and coupled forward-adjoint Monte Carlo calculation efficiencies for secondary neutron doses from proton beams

    International Nuclear Information System (INIS)

    Kelsey IV, Charles T.; Prinja, Anil K.

    2011-01-01

    We evaluate the Monte Carlo calculation efficiency for multigroup transport relative to continuous energy transport using the MCNPX code system to evaluate secondary neutron doses from a proton beam. We consider both fully forward simulation and application of a midway forward adjoint coupling method to the problem. Previously we developed tools for building coupled multigroup proton/neutron cross section libraries and showed consistent results for continuous energy and multigroup proton/neutron transport calculations. We observed that forward multigroup transport could be more efficient than continuous energy. Here we quantify solution efficiency differences for a secondary radiation dose problem characteristic of proton beam therapy problems. We begin by comparing figures of merit for forward multigroup and continuous energy MCNPX transport and find that multigroup is 30 times more efficient. Next we evaluate efficiency gains for coupling out-of-beam adjoint solutions with forward in-beam solutions. We use a variation of a midway forward-adjoint coupling method developed by others for neutral particle transport. Our implementation makes use of the surface source feature in MCNPX and we use spherical harmonic expansions for coupling in angle rather than solid angle binning. The adjoint out-of-beam transport for organs of concern in a phantom or patient can be coupled with numerous forward, continuous energy or multigroup, in-beam perturbations of a therapy beam line configuration. Out-of-beam dose solutions are provided without repeating out-of-beam transport. (author)

  2. Multi-group neutron transport theory

    International Nuclear Information System (INIS)

    Zelazny, R.; Kuszell, A.

    1962-01-01

    Multi-group neutron transport theory. In the paper the general theory of the application of the K. M. Case method to N-group neutron transport theory in plane geometry is given. The eigenfunctions (distributions) for the system of Boltzmann equations have been derived and the completeness theorem has been proved. By means of general solution two examples important for reactor and shielding calculations are given: the solution of a critical and albedo problem for a slab. In both cases the system of singular integral equations for expansion coefficients into a full set of eigenfunction distributions has been reduced to the system of Fredholm-type integral equations. Some results can be applied also to some spherical problems. (author) [fr

  3. Multigroup or multipoint thermal neutron data preparation. Programme SIGMA

    International Nuclear Information System (INIS)

    Matausek, M.V.; Kunc, M.

    1974-01-01

    When calculating the space energy distribution of thermal neutrons in reactor lattices, in either the multigroup or the multipoint approximation, it is convenient to divide the problem into two independent parts. Firstly, for all material regions of the given reactor lattice cell, the group or the point values of cross sections, scattering kernel and the outer source of thermal neutrons are calculated by a data preparation programme. These quantities are then used as input, by the programme which solves multigroup or multipoint transport equations, to generate the space energy neutron spectra in the cell considered and to determine the related integral quantities, namely the different reaction rates. The present report deals with the first part of the problem. An algorithm for constructing a set of thermal neutron input data, to be used with the multigroup or multipoint version of the code MULTI /1,2,3/, is presented and the new version of the programme SIGMA /4/, written in FORTRAN IV for the CDC-3600 computer, is described. For a given reactor cell material, composed of a number of different isotopes, this programme calculates the group or the point values of the scattering macroscopic absorption cross section, macroscopic scattering cross section, kernel and the outer source of thermal neutrons. Numerous options are foreseen in the programme, concerning the energy variation of cross sections and a scattering kernel, concerning the weighting spectrum in multigroup scheme or the procedure for constructing the scattering matrix in the multipoint scheme and, finally, concerning the organization of output. The details of the calculational algorithm are presented in Section 2 of the paper. Section 3 contains the description of the programme and the instructions for its use (author)

  4. The isotope density inverse problem in multigroup neutron transport

    International Nuclear Information System (INIS)

    Zazula, J.M.

    1981-01-01

    The inverse problem for stationary multigroup anisotropic neutron transport is discussed in order to search for isotope densities in multielement medium. The spatial- and angular-integrated form of neutron transport equation, in terms of the flux in a group - density of an element spatial correlation, leads to a set of integral functionals for the densities weighted by the group fluxes. Some methods of approximation to make the problem uniquently solvable are proposed. Particularly P 0 angular flux information and the spherically-symetrical geometry of an infinite medium are considered. The numerical calculation using this method related to sooner evaluated direct problem data gives promising agreement with primary densities. This approach would be the basis for further application in an elemental analysis of a medium, using an isotopic neutron source and a moving, energy-dependent neutron detector. (author)

  5. Complex of two-dimensional multigroup programs for neutron-physical computations of nuclear reactor

    International Nuclear Information System (INIS)

    Karpov, V.A.; Protsenko, A.N.

    1975-01-01

    Briefly stated mathematical aspects of the two-dimensional multigroup method of neutron-physical computation of nuclear reactor. Problems of algorithmization and BESM-6 computer realisation of multigroup diffuse approximations in hexagonal and rectangular calculated lattices are analysed. The results of computation of fast critical assembly having complicated composition of the core are given. The estimation of computation accuracy of criticality, neutron fields distribution and efficiency of absorbing rods by means of computer programs developed is done. (author)

  6. Application of direct discrete method (DDM) to multigroup neutron transport problems

    International Nuclear Information System (INIS)

    Vosoughi, Naser; Salehi, Ali Akbar; Shahriari, Majid

    2003-01-01

    The Direct Discrete Method (DDM), which produced excellent results for one-group neutron transport problems, has been developed for multigroup energy. A multigroup neutron transport discrete equation has been produced for a cylindrical shape fuel element with and without associated coolant regions with two boundary conditions. The calculations are illustrated for two-group energy by graphs showing the fast and thermal fluxes. The validity of the results are tested against the results obtained by the ANISN code. (author)

  7. Neutron-photon multigroup cross sections for neutron energies up to 400 MeV: HILO86R

    International Nuclear Information System (INIS)

    Kotegawa, Hiroshi; Nakane, Yoshihiro; Hasegawa, Akira; Tanaka, Shun-ichi

    1993-02-01

    A macroscopic multigroup cross section library of 66 neutron and 22 photon groups for neutron energies up to 400 MeV: HILO86R is prepared for 10 typical shielding materials; water, concrete, iron, air, graphite, polyethylene, heavy concrete, lead, aluminum and soil. The library is a revision of the DLC-119/HILO86, in which only the cross sections below 19.6 MeV have been exchanged with a group cross section processed from the JENDL-3 microscopic cross section library. In the HILO86R library, self shielding factors are used to produce effective cross sections for neutrons less than 19.6 MeV considering rather coarse energy meshes. Energy spectra and dose attenuation in water, concrete and iron have been compared among the HILO, HILO86 and HILO86R libraries for different energy neutron sources. Significant discrepancy has been observed in the energy spectra less than a couple of MeV energy in iron among the libraries, resulting large difference in the dose attenuation. The difference was attributed to the effect of self-shielding factor, namely to the difference between infinite dilution and effective cross sections. Even for 400 MeV neutron source the influence of the self-shielding factor is significant, nevertheless only the cross sections below 19.6 MeV are exchanged. (author)

  8. The problem of resonance self-shielding effect in neutron multigroup calculations

    International Nuclear Information System (INIS)

    Wang Qingming; Huang Jinghua

    1991-01-01

    It is not allowed to neglect the resonance self-shielding effect in hybrid blanket and fast reactor neutron designs. The authors discussed the importance as well as the method of considering the resonance self-shielding effect in hybrid blanket and fast reactor neutron multigroup calculations

  9. Proposal to extend CSEWG neutron and photon multigroup structures for wider applications

    International Nuclear Information System (INIS)

    LaBauve, R.J.; Wilson, W.B.

    1976-02-01

    The 239-group neutron multigroup structure recommended by the Codes and Formats Subcommittee of the cross section evaluation working group (CSEWG) for use in LMFBR design is not well suited for application in certain other areas, particularly thermal reactor design. This report describes a proposal for a neutron group structure consisting of 347 groups, which is an extension of the CSEWG group structure into the thermal range, and also includes more detail in other energy ranges important in LWR, HTGR, GCFR, and CTR design. Similarly, a proposed extension of the CSEWG 94-group photon multigroup structure to 103 groups is described. A subset of the neutron multigroup structure, consisting of 154 groups and for use in power reactor studies, is also presented

  10. Proposal to extend CSEWG neutron and photon multigroup structures for wider applications. [Tables

    Energy Technology Data Exchange (ETDEWEB)

    LaBauve, R.J.; Wilson, W.B.

    1976-02-01

    The 239-group neutron multigroup structure recommended by the Codes and Formats Subcommittee of the cross section evaluation working group (CSEWG) for use in LMFBR design is not well suited for application in certain other areas, particularly thermal reactor design. This report describes a proposal for a neutron group structure consisting of 347 groups, which is an extension of the CSEWG group structure into the thermal range, and also includes more detail in other energy ranges important in LWR, HTGR, GCFR, and CTR design. Similarly, a proposed extension of the CSEWG 94-group photon multigroup structure to 103 groups is described. A subset of the neutron multigroup structure, consisting of 154 groups and for use in power reactor studies, is also presented.

  11. An accurate solution of point reactor neutron kinetics equations of multi-group of delayed neutrons

    International Nuclear Information System (INIS)

    Yamoah, S.; Akaho, E.H.K.; Nyarko, B.J.B.

    2013-01-01

    Highlights: ► Analytical solution is proposed to solve the point reactor kinetics equations (PRKE). ► The method is based on formulating a coefficient matrix of the PRKE. ► The method was applied to solve the PRKE for six groups of delayed neutrons. ► Results shows good agreement with other traditional methods in literature. ► The method is accurate and efficient for solving the point reactor kinetics equations. - Abstract: The understanding of the time-dependent behaviour of the neutron population in a nuclear reactor in response to either a planned or unplanned change in the reactor conditions is of great importance to the safe and reliable operation of the reactor. In this study, an accurate analytical solution of point reactor kinetics equations with multi-group of delayed neutrons for specified reactivity changes is proposed to calculate the change in neutron density. The method is based on formulating a coefficient matrix of the homogenous differential equations of the point reactor kinetics equations and calculating the eigenvalues and the corresponding eigenvectors of the coefficient matrix. A small time interval is chosen within which reactivity relatively stays constant. The analytical method was applied to solve the point reactor kinetics equations with six-groups delayed neutrons for a representative thermal reactor. The problems of step, ramp and temperature feedback reactivities are computed and the results compared with other traditional methods. The comparison shows that the method presented in this study is accurate and efficient for solving the point reactor kinetics equations of multi-group of delayed neutrons

  12. Multigroup neutron transport equation in the diffusion and P{sub 1} approximation

    Energy Technology Data Exchange (ETDEWEB)

    Obradovic, D [Boris Kidric Institute of nuclear sciences Vinca, Belgrade (Yugoslavia)

    1970-07-01

    Investigations of the properties of the multigroup transport operator, width and without delayed neutrons in the diffusion and P{sub 1} approximation, is performed using Keldis's theory of operator families as well as a technique . recently used for investigations into the properties of the general linearized Boltzmann operator. It is shown that in the case without delayed neutrons, multigroup transport operator in the diffusion and P{sub 1} approximation possesses a complete set of generalized eigenvectors. A formal solution to the initial value problem is also given. (author)

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

    International Nuclear Information System (INIS)

    Smith, L.A.; Gehin, J.C.; Worley, B.A.; Renier, J.P.

    1994-01-01

    The FOEHN critical experiments were analyzed to validate the use of multigroup cross sections in the design of the Advanced Neutron Source. Eleven critical configurations were evaluated using the KENO, DORT, and VENTURE neutronics codes. Eigenvalue and power density profiles were computed and show very good agreement with measured values

  14. MCFT: a program for calculating fast and thermal neutron multigroup constants

    International Nuclear Information System (INIS)

    Yang Shunhai; Sang Xinzeng

    1993-01-01

    MCFT is a program for calculating the fast and thermal neutron multigroup constants, which is redesigned from some codes for generation of thermal neutron multigroup constants and for fast neutron multigroup constants adapted on CYBER 825 computer. It uses indifferently as basic input with the evaluated nuclear data contained in the ENDF/B (US), KEDAK (Germany) and UK (United Kingdom) libraries. The code includes a section devoted to the generation of resonant Doppler broadened cross section in the framework of single-or multi-level Breit-Wigner formalism. The program can compute the thermal neutron scattering law S (α, β, T) as the input data in tabular, free gas or diffusion motion form. It can treat up to 200 energy groups and Legendre moments up to P 5 . The output consists of various reaction multigroup constants in all neutron energy range desired in the nuclear reactor design and calculation. Three options in input file can be used by the user. The output format is arbitrary and defined by user with a minimum of program modification. The program includes about 15,000 cards and 184 subroutines. FORTRAN 5 computer language is used. The operation system is under NOS 2 on computer CYBER 825

  15. The Multigroup Neutron Diffusion Equations/1 Space Dimension

    Energy Technology Data Exchange (ETDEWEB)

    Linde, Sven

    1960-06-15

    A description is given of a program for the Ferranti Mercury computer which solves the one-dimensional multigroup diffusion equations in plane, cylindrical or spherical geometry, and also approximates automatically a two-dimensional solution by separating the space variables. In section A the method of calculation is outlined and the preparation of data for two group problems is described. The spatial separation of two-dimensional equations is considered in section B. Section C covers the multigroup equations. These parts are self contained and include all information required for the use of the program. Details of the numerical methods are given in section D. Three sample problems are solved in section E. Punching and operating instructions are given in an appendix.

  16. The Multigroup Neutron Diffusion Equations/1 Space Dimension

    International Nuclear Information System (INIS)

    Linde, Sven

    1960-06-01

    A description is given of a program for the Ferranti Mercury computer which solves the one-dimensional multigroup diffusion equations in plane, cylindrical or spherical geometry, and also approximates automatically a two-dimensional solution by separating the space variables. In section A the method of calculation is outlined and the preparation of data for two group problems is described. The spatial separation of two-dimensional equations is considered in section B. Section C covers the multigroup equations. These parts are self contained and include all information required for the use of the program. Details of the numerical methods are given in section D. Three sample problems are solved in section E. Punching and operating instructions are given in an appendix

  17. A boundary integral equation for boundary element applications in multigroup neutron diffusion theory

    International Nuclear Information System (INIS)

    Ozgener, B.

    1998-01-01

    A boundary integral equation (BIE) is developed for the application of the boundary element method to the multigroup neutron diffusion equations. The developed BIE contains no explicit scattering term; the scattering effects are taken into account by redefining the unknowns. Boundary elements of the linear and constant variety are utilised for validation of the developed boundary integral formulation

  18. MC2-2: a code to calculate fast neutron spectra and multigroup cross sections

    International Nuclear Information System (INIS)

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

    1976-06-01

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

  19. On efficiently computing multigroup multi-layer neutron reflection and transmission conditions

    International Nuclear Information System (INIS)

    Abreu, Marcos P. de

    2007-01-01

    In this article, we present an algorithm for efficient computation of multigroup discrete ordinates neutron reflection and transmission conditions, which replace a multi-layered boundary region in neutron multiplication eigenvalue computations with no spatial truncation error. In contrast to the independent layer-by-layer algorithm considered thus far in our computations, the algorithm here is based on an inductive approach developed by the present author for deriving neutron reflection and transmission conditions for a nonactive boundary region with an arbitrary number of arbitrarily thick layers. With this new algorithm, we were able to increase significantly the computational efficiency of our spectral diamond-spectral Green's function method for solving multigroup neutron multiplication eigenvalue problems with multi-layered boundary regions. We provide comparative results for a two-group reactor core model to illustrate the increased efficiency of our spectral method, and we conclude this article with a number of general remarks. (author)

  20. AMPX: a modular code system for generating coupled multigroup neutron-gamma libraries from ENDF/B

    Energy Technology Data Exchange (ETDEWEB)

    Greene, N.M.; Lucius, J.L.; Petrie, L.M.; Ford, W.E. III; White, J.E.; Wright, R.Q.

    1976-03-01

    AMPX is a modular system for producing coupled multigroup neutron-gamma cross section sets. Basic neutron and gamma cross-section data for AMPX are obtained from ENDF/B libraries. Most commonly used operations required to generate and collapse multigroup cross-section sets are provided in the system. AMPX is flexibly dimensioned; neutron group structures, and gamma group structures, and expansion orders to represent anisotropic processes are all arbitrary and limited only by available computer core and budget. The basic processes provided will (1) generate multigroup neutron cross sections; (2) generate multigroup gamma cross sections; (3) generate gamma yields for gamma-producing neutron interactions; (4) combine neutron cross sections, gamma cross sections, and gamma yields into final ''coupled sets''; (5) perform one-dimensional discrete ordinates transport or diffusion theory calculations for neutrons and gammas and, on option, collapse the cross sections to a broad-group structure, using the one-dimensional results as weighting functions; (6) plot cross sections, on option, to facilitate the ''evaluation'' of a particular multigroup set of data; (7) update and maintain multigroup cross section libraries in such a manner as to make it not only easy to combine new data with previously processed data but also to do it in a single pass on the computer; and (8) output multigroup cross sections in convenient formats for other codes. (auth)

  1. Neutron-photon multigroup cross sections for neutron energies less than or equal to400 MeV. Revision 1

    International Nuclear Information System (INIS)

    Alsmiller, R.G. Jr.; Barnes, J.M.; Drischler, J.D.

    1986-01-01

    For a variety of applications, e.g., accelerator shielding design, neutrons in radiotherapy, radiation damage studies, etc., it is necessary to carry out transport calculations involving medium-energy (greater than or equal to20 MeV) neutrons. A previous paper described neutron-photon multigroup cross sections in the ANISN format for neutrons from thermal to 400 MeV. In the present paper the cross-section data presented previously have been revised to make them agree with available experimental data. 7 refs., 1 fig

  2. PHISICS multi-group transport neutronic capabilities for RELAP5

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-07-01

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

  3. Multi-group transport methods for high-resolution neutron activation analysis

    International Nuclear Information System (INIS)

    Burns, K. A.; Smith, L. E.; Gesh, C. J.; Shaver, M. W.

    2009-01-01

    The accurate and efficient simulation of coupled neutron-photon problems is necessary for several important radiation detection applications. Examples include the detection of nuclear threats concealed in cargo containers and prompt gamma neutron activation analysis for nondestructive determination of elemental composition of unknown samples. In these applications, high-resolution gamma-ray spectrometers are used to preserve as much information as possible about the emitted photon flux, which consists of both continuum and characteristic gamma rays with discrete energies. Monte Carlo transport is the most commonly used modeling tool for this type of problem, but computational times for many problems can be prohibitive. This work explores the use of multi-group deterministic methods for the simulation of neutron activation problems. Central to this work is the development of a method for generating multi-group neutron-photon cross-sections in a way that separates the discrete and continuum photon emissions so that the key signatures in neutron activation analysis (i.e., the characteristic line energies) are preserved. The mechanics of the cross-section preparation method are described and contrasted with standard neutron-gamma cross-section sets. These custom cross-sections are then applied to several benchmark problems. Multi-group results for neutron and photon flux are compared to MCNP results. Finally, calculated responses of high-resolution spectrometers are compared. Preliminary findings show promising results when compared to MCNP. A detailed discussion of the potential benefits and shortcomings of the multi-group-based approach, in terms of accuracy, and computational efficiency, is provided. (authors)

  4. Discrete formulation for two-dimensional multigroup neutron diffusion equations

    Energy Technology Data Exchange (ETDEWEB)

    Vosoughi, Naser E-mail: vosoughi@mehr.sharif.edu; Salehi, Ali A.; Shahriari, Majid

    2003-02-01

    The objective of this paper is to introduce a new numerical method for neutronic calculation in a reactor core. This method can produce the final finite form of the neutron diffusion equation by classifying the neutronic variables and using two kinds of cell complexes without starting from the conventional differential form of the neutron diffusion equation. The method with linear interpolation produces the same convergence as the linear continuous finite element method. The quadratic interpolation is proven; the convergence order depends on the shape of the dual cell. The maximum convergence order is achieved by choosing the dual cell based on two Gauss' points. The accuracy of the method was examined with a well-known IAEA two-dimensional benchmark problem. The numerical results demonstrate the effectiveness of the new method.

  5. Discrete formulation for two-dimensional multigroup neutron diffusion equations

    International Nuclear Information System (INIS)

    Vosoughi, Naser; Salehi, Ali A.; Shahriari, Majid

    2003-01-01

    The objective of this paper is to introduce a new numerical method for neutronic calculation in a reactor core. This method can produce the final finite form of the neutron diffusion equation by classifying the neutronic variables and using two kinds of cell complexes without starting from the conventional differential form of the neutron diffusion equation. The method with linear interpolation produces the same convergence as the linear continuous finite element method. The quadratic interpolation is proven; the convergence order depends on the shape of the dual cell. The maximum convergence order is achieved by choosing the dual cell based on two Gauss' points. The accuracy of the method was examined with a well-known IAEA two-dimensional benchmark problem. The numerical results demonstrate the effectiveness of the new method

  6. Solution for the multigroup neutron space kinetics equations by the modified Picard algorithm

    Energy Technology Data Exchange (ETDEWEB)

    Tavares, Matheus G.; Petersen, Claudio Z., E-mail: matheus.gulartetavares@gmail.com [Universidade Federal de Pelotas (UFPEL), Capao do Leao, RS (Brazil). Departamento de Matematica e Estatistica; Schramm, Marcelo, E-mail: schrammmarcelo@gmail.com [Universidade Federal de Pelotas (UFPEL), RS (Brazil). Centro de Engenharias; Zanette, Rodrigo, E-mail: rodrigozanette@hotmail.com [Universidade Federal do Rio Grande do Sul (UFRGS), Porto Alegre, RS (Brazil). Instituto de Matematica e Estatistica

    2017-07-01

    In this work, we used a modified Picards method to solve the Multigroup Neutron Space Kinetics Equations (MNSKE) in Cartesian geometry. The method consists in assuming an initial guess for the neutron flux and using it to calculate a fictitious source term in the MNSKE. A new source term is calculated applying its solution, and so on, iteratively, until a stop criterion is satisfied. For the solution of the fast and thermal neutron fluxes equations, the Laplace Transform technique is used in time variable resulting in a rst order linear differential matrix equation, which are solved by classical methods in the literature. After each iteration, the scalar neutron flux and the delayed neutron precursors are reconstructed by polynomial interpolation. We obtain the fluxes and precursors through Numerical Inverse Laplace Transform using the Stehfest method. We present numerical simulations and comparisons with available results in literature. (author)

  7. Solution for the multigroup neutron space kinetics equations by the modified Picard algorithm

    International Nuclear Information System (INIS)

    Tavares, Matheus G.; Petersen, Claudio Z.; Schramm, Marcelo; Zanette, Rodrigo

    2017-01-01

    In this work, we used a modified Picards method to solve the Multigroup Neutron Space Kinetics Equations (MNSKE) in Cartesian geometry. The method consists in assuming an initial guess for the neutron flux and using it to calculate a fictitious source term in the MNSKE. A new source term is calculated applying its solution, and so on, iteratively, until a stop criterion is satisfied. For the solution of the fast and thermal neutron fluxes equations, the Laplace Transform technique is used in time variable resulting in a rst order linear differential matrix equation, which are solved by classical methods in the literature. After each iteration, the scalar neutron flux and the delayed neutron precursors are reconstructed by polynomial interpolation. We obtain the fluxes and precursors through Numerical Inverse Laplace Transform using the Stehfest method. We present numerical simulations and comparisons with available results in literature. (author)

  8. A multi-group neutron noise simulator for fast reactors

    International Nuclear Information System (INIS)

    Tran, Hoai Nam; Zylbersztejn, Florian; Demazière, Christophe; Jammes, Christian; Filliatre, Philippe

    2013-01-01

    Highlights: • The development of a neutron noise simulator for fast reactors. • The noise equation is solved fully in a frequency-domain. • A good agreement with ERANOS on the static calculations. • Noise calculations induced by a localized perturbation of absorption cross section. - Abstract: A neutron noise simulator has been developed for fast reactors based on diffusion theory with multi-energy groups and several groups of delayed neutron precursors. The tool is expected to be applicable for core monitoring of fast reactors and also for other reactor types with hexagonal fuel assemblies. The noise sources are modeled through small stationary fluctuations of macroscopic cross sections, and the induced first order noise is solved fully in the frequency domain. Numerical algorithms are implemented for solving both the static and noise equations using finite differences for spatial discretization, where a hexagonal assembly is radially divided into finer triangular meshes. A coarse mesh finite difference (CMFD) acceleration has been used for accelerating the convergence of both the static and noise calculations. Numerical calculations have been performed for the ESFR core with 33 energy groups and 8 groups of delayed neutron precursors using the cross section data generated by the ERANOS code. The results of the static state have been compared with those obtained using ERANOS. The results show an adequate agreement between the two calculations. Noise calculations for the ESFR core have also been performed and demonstrated with an assumption of the perturbation of the absorption cross section located at the central fuel ring

  9. HAMMER, 1-D Multigroup Neutron Transport Infinite System Cell Calculation for Few-Group Diffusion Calculation

    International Nuclear Information System (INIS)

    Honeck, H.C.

    1984-01-01

    1 - Description of problem or function: HAMMER performs infinite lattice, one-dimensional cell multigroup calculations, followed (optionally) by one-dimensional, few-group, multi-region reactor calculations with neutron balance edits. 2 - Method of solution: Infinite lattice parameters are calculated by means of multigroup transport theory, composite reactor parameters by few-group diffusion theory. 3 - Restrictions on the complexity of the problem: - Cell calculations - maxima of: 30 thermal groups; 54 epithermal groups; 20 space points; 20 regions; 18 isotopes; 10 mixtures; 3 thermal up-scattering mixtures; 200 resonances per group; no overlap or interference; single level only. - Reactor calculations - maxima of : 40 regions; 40 mixtures; 250 space points; 4 groups

  10. Assessment and comparison of different multigroup neutron cross section libraries for dosimetry purposes

    International Nuclear Information System (INIS)

    Erradi, L.; Karouani, K.

    1994-01-01

    Many multigroup neutron cross section libraries have been processed from basic evaluated nuclear data for use in neutron dosimetry, reactor shielding calculation and in the development of fusion reactors. Most of these libraries have been tested only for fission spectra and were not validated for fusion spectra. Fifteen of these libraries such as DOSCROS84, IRDF85 and ENDFB5 have been used along with the neutron spectra unfolding code SAND II to evaluate about fifteen threshold detector saturated activities. The comparison between these computed activities and the measured ones of a set of foils placed in different places along the axis of a paraffin cylinder and irradiated by 14 MeV neutrons generated by a D-T source, hence giving rise to complex spectra, leads to different types of discrepancies. The analysis of these discrepancies allows to select from these libraries the ones that can be recommended. 1 fig., 4 refs. (author)

  11. Numerical analysis for multi-group neutron-diffusion equation using Radial Point Interpolation Method (RPIM)

    International Nuclear Information System (INIS)

    Kim, Kyung-O; Jeong, Hae Sun; Jo, Daeseong

    2017-01-01

    Highlights: • Employing the Radial Point Interpolation Method (RPIM) in numerical analysis of multi-group neutron-diffusion equation. • Establishing mathematical formation of modified multi-group neutron-diffusion equation by RPIM. • Performing the numerical analysis for 2D critical problem. - Abstract: A mesh-free method is introduced to overcome the drawbacks (e.g., mesh generation and connectivity definition between the meshes) of mesh-based (nodal) methods such as the finite-element method and finite-difference method. In particular, the Point Interpolation Method (PIM) using a radial basis function is employed in the numerical analysis for the multi-group neutron-diffusion equation. The benchmark calculations are performed for the 2D homogeneous and heterogeneous problems, and the Multiquadrics (MQ) and Gaussian (EXP) functions are employed to analyze the effect of the radial basis function on the numerical solution. Additionally, the effect of the dimensionless shape parameter in those functions on the calculation accuracy is evaluated. According to the results, the radial PIM (RPIM) can provide a highly accurate solution for the multiplication eigenvalue and the neutron flux distribution, and the numerical solution with the MQ radial basis function exhibits the stable accuracy with respect to the reference solutions compared with the other solution. The dimensionless shape parameter directly affects the calculation accuracy and computing time. Values between 1.87 and 3.0 for the benchmark problems considered in this study lead to the most accurate solution. The difference between the analytical and numerical results for the neutron flux is significantly increased in the edge of the problem geometry, even though the maximum difference is lower than 4%. This phenomenon seems to arise from the derivative boundary condition at (x,0) and (0,y) positions, and it may be necessary to introduce additional strategy (e.g., the method using fictitious points and

  12. Neutron-photon multigroup cross sections for neutron energies less than or equal to400 MeV. Revision 1

    International Nuclear Information System (INIS)

    Alsmiller, R.G. Jr.; Barnes, J.M.; Drischler, J.D.

    1986-02-01

    Multigroup cross sections (66 neutron groups and 22 photon groups) are described for neutron energies from thermal to 400 MeV. The elements considered are hydrogen, 10 B, 11 B, carbon, nitrogen, oxygen, sodium, magnesium, aluminum, silicon, sulfur, potassium, calcium, chromium, iron, nickel, tungsten, and lead. The cross section data presented are a revision of similar data presented previously. In the case of iron, transport calculations using the earlier and the revised cross sections are presented and compared, and significant differences are found. The revised cross sections are available from the Radiation Shielding information Center of the Oak Ridge National Laboratory. 32 refs., 5 figs., 3 tabs

  13. Solution of the multilayer multigroup neutron diffusion equation in cartesian geometry by fictitious borders power method

    Energy Technology Data Exchange (ETDEWEB)

    Zanette, Rodrigo; Petersen, Caudio Zen [Univ. Federal de Pelotas, Capao do Leao (Brazil). Programa de Pos Graduacao em Modelagem Matematica; Schramm, Marcello [Univ. Federal de Pelotas (Brazil). Centro de Engenharias; Zabadal, Jorge Rodolfo [Univ. Federal do Rio Grande do Sul, Tramandai (Brazil)

    2017-05-15

    In this paper a solution for the one-dimensional steady state Multilayer Multigroup Neutron Diffusion Equation in cartesian geometry by Fictitious Borders Power Method and a perturbative analysis of this solution is presented. For each new iteration of the power method, the neutron flux is reconstructed by polynomial interpolation, so that it always remains in a standard form. However when the domain is long, an almost singular matrix arises in the interpolation process. To eliminate this singularity the domain segmented in R regions, called fictitious regions. The last step is to solve the neutron diffusion equation for each fictitious region in analytical form locally. The results are compared with results present in the literature. In order to analyze the sensitivity of the solution, a perturbation in the nuclear parameters is inserted to determine how a perturbation interferes in numerical results of the solution.

  14. Analytical synthetic methods of solution of neutron transport equation with diffusion theory approaches energy multigroup

    International Nuclear Information System (INIS)

    Moraes, Pedro Gabriel B.; Leite, Michel C.A.; Barros, Ricardo C.

    2013-01-01

    In this work we developed a software to model and generate results in tables and graphs of one-dimensional neutron transport problems in multi-group formulation of energy. The numerical method we use to solve the problem of neutron diffusion is analytic, thus eliminating the truncation errors that appear in classical numerical methods, e.g., the method of finite differences. This numerical analytical method increases the computational efficiency, since they are not refined spatial discretization necessary because for any spatial discretization grids used, the numerical result generated for the same point of the domain remains unchanged unless the rounding errors of computational finite arithmetic. We chose to develop a computational application in MatLab platform for numerical computation and program interface is simple and easy with knobs. We consider important to model this neutron transport problem with a fixed source in the context of shielding calculations of radiation that protects the biosphere, and could be sensitive to ionizing radiation

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

    International Nuclear Information System (INIS)

    Smith, L.A.; Gallmeier, F.X.; Gehin, J.C.

    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 ∼ 13%, while the average differences are < 8%

  16. Analysis of coupled neutron-gamma radiations, applied to shieldings in multigroup albedo method

    International Nuclear Information System (INIS)

    Dunley, Leonardo Souza

    2002-01-01

    The principal mathematical tools frequently available for calculations in Nuclear Engineering, including coupled neutron-gamma radiations shielding problems, involve the full Transport Theory or the Monte Carlo techniques. The Multigroup Albedo Method applied to shieldings is characterized by following the radiations through distinct layers of materials, allowing the determination of the neutron and gamma fractions reflected from, transmitted through and absorbed in the irradiated media when a neutronic stream hits the first layer of material, independently of flux calculations. Then, the method is a complementary tool of great didactic value due to its clarity and simplicity in solving neutron and/or gamma shielding problems. The outstanding results achieved in previous works motivated the elaboration and the development of this study that is presented in this dissertation. The radiation balance resulting from the incidence of a neutronic stream into a shielding composed by 'm' non-multiplying slab layers for neutrons was determined by the Albedo method, considering 'n' energy groups for neutrons and 'g' energy groups for gammas. It was taken into account there is no upscattering of neutrons and gammas. However, it was considered that neutrons from any energy groups are able to produce gammas of all energy groups. The ANISN code, for an angular quadrature order S 2 , was used as a standard for comparison of the results obtained by the Albedo method. So, it was necessary to choose an identical system configuration, both for ANISN and Albedo methods. This configuration was six neutron energy groups and eight gamma energy groups, using three slab layers (iron aluminum - manganese). The excellent results expressed in comparative tables show great agreement between the values determined by the deterministic code adopted as standard and, the values determined by the computational program created using the Albedo method and the algorithm developed for coupled neutron

  17. Integral tests of coupled multigroup neutron and gamma cross sections with fission and fusion sources

    International Nuclear Information System (INIS)

    Schriewer, J.; Hehn, G.; Mattes, M.; Pfister, G.; Keinert, J.

    1978-01-01

    Calculations were made for different benchmark experiments in order to test the coupled multigroup neutron and gamma library EURLIB-3 with 100 neutron groups and 20 gamma groups. In cooperation with EURATOM, Ispra, we produced this shielding library recently from ENDF/B-IV data for application in fission and fusion technology. Integral checks were performed for natural lithium, carbon, oxygen, and iron. Since iron is the most important structural material in nuclear technology, we started with calculations of iron benchmark experiments. Most of them are integral experiments of INR, Karlsruhe, but comparisons were also done with benchmark experiments from USA and Japan. For the experiments with fission sources we got satisfying results. All details of the resonances cannot be checked with flux measurements and multigroup cross sections used. But some averaged resonance behaviour of the measured and calculated fluxes can be compared and checked within the error limits given. We get greater differences in the calculations of benchmark experiments with 14 MeV neutron sources. For iron the group cross sections of EURLIB-3 produce an underestimation of the neutron flux in a broad energy region below the source energy. The conclusion is that the energy degradation by inelastic scattering is too strong. For fusion application the anisotropy of the inelastic scatter process must be taken into account, which isn't done by the processing codes at present. If this effect isn't enough, additional corrections have to be applied to the inelastic cross sections of iron in ENDF/B-IV. (author)

  18. MULTI - A multigroup or multipoint P{sub 3} programme for calculating thermal neutron spectra in a reactor cell

    Energy Technology Data Exchange (ETDEWEB)

    Matausek, M V [Boris Kidric Institute of Nuclear Sciences Vinca, Beograd (Yugoslavia)

    1968-06-15

    Programme MULTI calculates the space energy distribution of thermal neutrons in a multizone, cylindrical, infinitely long reactor lattice by using the multigroup or multipoint P{sub 3} approximation. This report presents a short description of the algorithm and the programme and gives the instructions for its exploitation. (author)

  19. Algorithm development and verification of UASCM for multi-dimension and multi-group neutron kinetics model

    International Nuclear Information System (INIS)

    Si, S.

    2012-01-01

    The Universal Algorithm of Stiffness Confinement Method (UASCM) for neutron kinetics model of multi-dimensional and multi-group transport equations or diffusion equations has been developed. The numerical experiments based on transport theory code MGSNM and diffusion theory code MGNEM have demonstrated that the algorithm has sufficient accuracy and stability. (authors)

  20. Some reciprocity-like relations in multi-group neutron diffusion and transport theory over bare homogeneous regions

    International Nuclear Information System (INIS)

    Modak, R.S.; Sahni, D.C.

    1996-01-01

    Some simple reciprocity-like relations that exist in multi-group neutron diffusion and transport theory over bare homogeneous regions are presented. These relations do not involve the adjoint solutions and are directly related to numerical schemes based on an explicit evaluation of the fission matrix. (author)

  1. Linear triangle finite element formulation for multigroup neutron transport analysis with anisotropic scattering

    Energy Technology Data Exchange (ETDEWEB)

    Lillie, R.A.; Robinson, J.C.

    1976-05-01

    The discrete ordinates method is the most powerful and generally used deterministic method to obtain approximate solutions of the Boltzmann transport equation. A finite element formulation, utilizing a canonical form of the transport equation, is here developed to obtain both integral and pointwise solutions to neutron transport problems. The formulation is based on the use of linear triangles. A general treatment of anisotropic scattering is included by employing discrete ordinates-like approximations. In addition, multigroup source outer iteration techniques are employed to perform group-dependent calculations. The ability of the formulation to reduce substantially ray effects and its ability to perform streaming calculations are demonstrated by analyzing a series of test problems. The anisotropic scattering and multigroup treatments used in the development of the formulation are verified by a number of one-dimensional comparisons. These comparisons also demonstrate the relative accuracy of the formulation in predicting integral parameters. The applicability of the formulation to nonorthogonal planar geometries is demonstrated by analyzing a hexagonal-type lattice. A small, high-leakage reactor model is analyzed to investigate the effects of varying both the spatial mesh and order of angular quadrature. This analysis reveals that these effects are more pronounced in the present formulation than in other conventional formulations. However, the insignificance of these effects is demonstrated by analyzing a realistic reactor configuration. In addition, this final analysis illustrates the importance of incorporating anisotropic scattering into the finite element formulation. 8 tables, 29 figures.

  2. Linear triangle finite element formulation for multigroup neutron transport analysis with anisotropic scattering

    International Nuclear Information System (INIS)

    Lillie, R.A.; Robinson, J.C.

    1976-05-01

    The discrete ordinates method is the most powerful and generally used deterministic method to obtain approximate solutions of the Boltzmann transport equation. A finite element formulation, utilizing a canonical form of the transport equation, is here developed to obtain both integral and pointwise solutions to neutron transport problems. The formulation is based on the use of linear triangles. A general treatment of anisotropic scattering is included by employing discrete ordinates-like approximations. In addition, multigroup source outer iteration techniques are employed to perform group-dependent calculations. The ability of the formulation to reduce substantially ray effects and its ability to perform streaming calculations are demonstrated by analyzing a series of test problems. The anisotropic scattering and multigroup treatments used in the development of the formulation are verified by a number of one-dimensional comparisons. These comparisons also demonstrate the relative accuracy of the formulation in predicting integral parameters. The applicability of the formulation to nonorthogonal planar geometries is demonstrated by analyzing a hexagonal-type lattice. A small, high-leakage reactor model is analyzed to investigate the effects of varying both the spatial mesh and order of angular quadrature. This analysis reveals that these effects are more pronounced in the present formulation than in other conventional formulations. However, the insignificance of these effects is demonstrated by analyzing a realistic reactor configuration. In addition, this final analysis illustrates the importance of incorporating anisotropic scattering into the finite element formulation. 8 tables, 29 figures

  3. Spectrum of the multigroup neutron transport operator for bounded spatial domains

    International Nuclear Information System (INIS)

    Larsen, E.W.

    1979-01-01

    The spectrum of the multigroup neutron transport operator A is studied for bounded spatial regions D which consist of a finite number of material subregions. Our main results provide simple conditions on the material cross sections which guarantee that (1) A possesses eigenvalues in the finite plane; (2) A possesses a ''leading'' eigenvalue lambda 0 which is real, not less than the real part of any other eigenvalue, and to which there corresponds at least one nonnegative eigenfunction psi/sub lambda/0; and (3) A possesses a ''dominant'' eigenvalue lambda 0 which is real, simple, greater than the real part of any other eigenvalue, and whose eigenfunction psi/sub lambda/0 satisfies psi/sub lambda/0> or =0 and ∫psi/sub lambda/0d 2 Ω>0. We give examples to illustrate the results and to show that a leading eigenvalue need not be simple, nor its eigenfunction(s) positive

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

    International Nuclear Information System (INIS)

    Fletcher, K.

    1981-01-01

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

  5. Solution of the multigroup neutron diffusion Eigenvalue problem in slab geometry by modified power method

    Energy Technology Data Exchange (ETDEWEB)

    Zanette, Rodrigo [Universidade Federal do Rio Grande do Sul (UFRGS), Porto Alegre, RS (Brazil). Programa de Pós-Graduação em Matemática Aplicada; Petersen, Claudio Z.; Tavares, Matheus G., E-mail: rodrigozanette@hotmail.com, E-mail: claudiopetersen@yahoo.com.br, E-mail: matheus.gulartetavares@gmail.com [Universidade Federal de Pelotas (UFPEL), RS (Brazil). Programa de Pós-Graduação em Modelagem Matemática

    2017-07-01

    We describe in this work the application of the modified power method for solve the multigroup neutron diffusion eigenvalue problem in slab geometry considering two-dimensions for nuclear reactor global calculations. It is well known that criticality calculations can often be best approached by solving eigenvalue problems. The criticality in nuclear reactors physics plays a relevant role since establishes the ratio between the numbers of neutrons generated in successive fission reactions. In order to solve the eigenvalue problem, a modified power method is used to obtain the dominant eigenvalue (effective multiplication factor (K{sub eff})) and its corresponding eigenfunction (scalar neutron flux), which is non-negative in every domain, that is, physically relevant. The innovation of this work is solving the neutron diffusion equation in analytical form for each new iteration of the power method. For solve this problem we propose to apply the Finite Fourier Sine Transform on one of the spatial variables obtaining a transformed problem which is resolved by well-established methods for ordinary differential equations. The inverse Fourier transform is used to reconstruct the solution for the original problem. It is known that the power method is an iterative source method in which is updated by the neutron flux expression of previous iteration. Thus, for each new iteration, the neutron flux expression becomes larger and more complex due to analytical solution what makes propose that it be reconstructed through an polynomial interpolation. The methodology is implemented to solve a homogeneous problem and the results are compared with works presents in the literature. (author)

  6. MC2-2, Calculation of Fast Neutron Spectra and Multigroup Cross-Sections from ENDF/B Data

    International Nuclear Information System (INIS)

    2001-01-01

    1 - Description of program or function: MC 2 -2 solves the neutron slowing-down equations using basic neutron data derived from ENDF/B data files to determine fundamental mode spectra for use in generating multigroup neutron cross sections. The current edition includes the ability to treat all ENDF/B-V and -VI data representations. It accommodates high-order P scattering representations and provides numerous capabilities such as isotope mixing, delayed neutron processing, free-format input, and flexibility in output data selection. This edition supersedes previous releases of the MC22 program and the earlier MC2 program. Improved physics algorithms and increased computational efficiency are incorporated. Input data files required by MC2-2 may be generated from ENDF/B data by the code ETOE-2. The hyper-fine-group integral transport theory module of MC2-2, RABANL, is an improved version of the RABBLE/RABID codes. Many of the MC2-2 modules are used in the SDX code. 2 - Methods: The extended transport P1, B1, consistent P1, and consistent B1 fundamental mode ultra-fine-group equations are solved using continuous slowing-down theory and multigroup methods. Fast and accurate resonance integral methods are used in the narrow resonance resolved and unresolved resonance treatments. A fundamental mode homogeneous unit cell calculation is performed using either a multigroup or a continuous slowing-down treatment. Multigroup neutron homogeneous cross sections are generated in an ISOTXS format for an arbitrary group structure. A hyper-fine-group integral transport slowing down calculation (RABANL) is available as an option. RABANL performs a homogeneous or heterogeneous (pin or slab) unit cell calculation over the resonance region (resolved and unresolved) and generates multigroup neutron cross sections in an ISOTXS format. Neutron cross sections are generated by RABANL for the homogeneous unit cell and for each heterogeneous region in the pin or slab unit cell calculation

  7. ANSL-V: ENDF/B-V based multigroup cross-section libraries for Advanced Neutron Source (ANS) reactor studies

    International Nuclear Information System (INIS)

    Ford, W.E. III; Arwood, J.W.; Greene, N.M.; Petrie, L.M.; Primm, R.T. III; Waddell, M.W.; Webster, C.C.; Westfall, R.M.; Wright, R.Q.

    1987-01-01

    Multigroup P3 neutron, P0-P3 secondary gamma ray production (SGRP), and P6 gamma ray interaction (GRI) cross section libraries have been generated to support design work on the Advanced Neutron Source (ANS) reactor. The libraries, designated ANSL-V (Advanced Neutron Source Cross-Section Libraries), are data bases in a format suitable for subsequent generation of problem dependent cross sections. The ANSL-V libraries are available on magnetic tape from the Radiation Shielding Information Center at Oak Ridge National Laboratory

  8. EPRI-LATTICE: a multigroup neutron transport code for light water reactor lattice physics calculations

    International Nuclear Information System (INIS)

    Jones, D.B.

    1986-01-01

    EPRI-LATTICE is a multigroup neutron transport computer code for the analysis of light water reactor fuel assemblies. It can solve the two-dimensional neutron transport problem by two distinct methods: (a) the method of collision probabilities and (b) the method of discrete ordinates. The code was developed by S. Levy Inc. as an account of work sponsored by the Electric Power Research Institute (EPRI). The collision probabilities calculation in EPRI-LATTICE (L-CP) is based on the same methodology that exists in the lattice codes CPM-2 and EPRI-CPM. Certain extensions have been made to the data representations of the CPM programs to improve the overall accuracy of the calculation. The important extensions include unique representations of scattering matrices and fission fractions (chi) for each composition in the problem. A new capability specifically developed for the EPRI-LATTICE code is a discrete ordinates methodology. The discrete ordinates calculation in EPRI-LATTICE (L-SN) is based on the discrete S/sub n/ methodology that exists in the TWODANT program. In contrast to TWODANT, which utilizes synthetic diffusion acceleration and supports multiple geometries, only the transport equations are solved by L-SN and only the data representations for the two-dimensional geometry are treated

  9. Application of the Laplace transform method for the albedo boundary conditions in multigroup neutron diffusion eigenvalue problems in slab geometry

    International Nuclear Information System (INIS)

    Petersen, Claudio Zen; Vilhena, Marco T.; Barros, Ricardo C.

    2009-01-01

    In this paper the application of the Laplace transform method is described in order to determine the energy-dependent albedo matrix that is used in the boundary conditions multigroup neutron diffusion eigenvalue problems in slab geometry for nuclear reactor global calculations. In slab geometry, the diffusion albedo substitutes without approximation the baffle-reflector system around the active domain. Numerical results to typical test problems are shown to illustrate the accuracy and the efficiency of the Chebysheff acceleration scheme. (orig.)

  10. Benchmarking of multigroup neutron cross sections libraries on neutron transmission through WWER-440 vessel

    International Nuclear Information System (INIS)

    Ilieva, K.; Belousov, S.; Apostolov, T.

    1998-01-01

    The verification of calculated neutron fluence onto the WWER-440/230 pressure vessel is very topical task in particular referring that some of this type of reactors have been operated the major part of its design lifetime. Since the induced activity from the neutron irradiation onto the elements is a simple response of neutron flux the neutron fluence verification usually is done using the measured activity of radionuclides produced during reactor operation. Calculational and experimental results of 54 Mn induced activity of scraps from inner wall of Unit 1 reactor pressure vessel after 18th cycle and detectors irradiated behind the vessel during the 18th cycle of Unit 1 at Kozloduy NPP as well as neutron flux attenuation through the WWER-440/230 pressure vessel are presented. Neutron cross sections libraries generated on the base of ENDF/B-IV and ENDF/B-VI have been used in the calculations. The comparative analysis of evaluated activities and attenuation coefficient demonstrates the better reliability of the neutron fluence calculations by the libraries based on ENDF/B-VI than by ones on ENDF/B-IV. The extreme rarity of data for the activity of scraps from the WWER-440 reactor vessel and its combination with the data for the detectors irradiated behind the vessel makes them especially attractive for verification of calculational methods of neutron fluence onto the WWER-440 vessel with dummy cassettes loading. (author)

  11. APPLE, Plot of 1-D Multigroup Neutron Flux and Gamma Flux and Reaction Rates from ANISN

    International Nuclear Information System (INIS)

    Kawasaki, Hiromitsu; Seki, Yasushi

    1983-01-01

    A - Description of problem or function: The APPLE-2 code has the following functions: (1) It plots multi-group energy spectra of neutron and/or gamma ray fluxes calculated by ANISN, DOT-3.5, and MORSE. (2) It gives an overview plot of multi-group neutron fluxes calculated by ANISN and DOT-3.5. The scalar neutron flux phi(r,E) is plotted with the spatial parameter r linear along the Y-axis, logE along the X-axis and log phi(r,E) in the Z direction. (3) It calculates the spatial distribution and region volume integrated values of reaction rates using the scalar flux calculated with ANISN and DOT-3.5. (4) Reaction rate distribution along the R or Z direction may be plotted. (5) An overview plot of reaction rates or scalar fluxes summed over specified groups may be plotted. R(ri,zi) or phi(ri,zi) is plotted with spatial parameters r and z along the X- and Y-axes in an orthogonal coordinate system. (6) Angular flux calculated by ANISN is rearranged and a shell source at any specified spatial mesh point may be punched out in FIDO format. The shell source obtained may be employed in solving deep penetration problems with ANISN, when the entire reactor system is divided into two or more parts and the neutron fluxes in two adjoining parts are connected by using the shell source. B - Method of solution: (a) The input data specification is made as simple as possible by making use of the input data required in the radiation transport code. For example, geometry related data in ANISN and DOT are transmitted to APPLE-2 along with scalar flux data so as to reduce duplicity and errors in reproducing these data. (b) Most the input data follow the free form FIDO format developed at Oak Ridge National Laboratory and used in the ANISN code. Furthermore, the mixture specifying method used in ANISN is also employed by APPLE-2. (c) Libraries for some standard response functions required in fusion reactor design have been prepared and are made available to users of the 42-group neutron

  12. Three-dimensional h-adaptivity for the multigroup neutron diffusion equations

    KAUST Repository

    Wang, Yaqi

    2009-04-01

    Adaptive mesh refinement (AMR) has been shown to allow solving partial differential equations to significantly higher accuracy at reduced numerical cost. This paper presents a state-of-the-art AMR algorithm applied to the multigroup neutron diffusion equation for reactor applications. In order to follow the physics closely, energy group-dependent meshes are employed. We present a novel algorithm for assembling the terms coupling shape functions from different meshes and show how it can be made efficient by deriving all meshes from a common coarse mesh by hierarchic refinement. Our methods are formulated using conforming finite elements of any order, for any number of energy groups. The spatial error distribution is assessed with a generalization of an error estimator originally derived for the Poisson equation. Our implementation of this algorithm is based on the widely used Open Source adaptive finite element library deal.II and is made available as part of this library\\'s extensively documented tutorial. We illustrate our methods with results for 2-D and 3-D reactor simulations using 2 and 7 energy groups, and using conforming finite elements of polynomial degree up to 6. © 2008 Elsevier Ltd. All rights reserved.

  13. MENDF71x. Multigroup Neutron Cross Section Data Tables Based upon ENDF/B-VII.1

    International Nuclear Information System (INIS)

    Conlin, Jeremy Lloyd; Parsons, Donald Kent; Gardiner, Steven J.; Gray, Mark Girard; Lee, Mary Beth; White, Morgan Curtis

    2015-01-01

    A new multi-group neutron cross section library has been released along with the release of NDI version 2.0.20. The library is named MENDF71x and is based upon the evaluations released in ENDF/B-VII.1 which was made publicly available in December 2011. ENDF/B-VII.1 consists of 423 evaluations of which ten are excited states evaluations and 413 are ground state evaluations. MENDF71x was created by processing the 423 evaluations into 618-group, downscatter only NDI data tables. The ENDF/B evaluation files were processed using NJOY version 99.393 with the exception of 35 Cl and 233 U. Those two isotopes had unique properties that required that we process the evaluation using NJOY version 2012. The MENDF71x library was only processed to room temperature, i.e., 293.6 K. In the future, we plan on producing a multi-temperature library based on ENDF/B-VII.1 and compatible with MENDF71x.

  14. TRIDENT: a two-dimensional, multigroup, triangular mesh discrete ordinates, explicit neutron transport code

    International Nuclear Information System (INIS)

    Seed, T.J.; Miller, W.F. Jr.; Brinkley, F.W. Jr.

    1977-03-01

    TRIDENT solves the two-dimensional-multigroup-transport equations in rectangular (x-y) and cylindrical (r-z) geometries using a regular triangular mesh. Regular and adjoint, inhomogeneous and homogeneous (k/sub eff/ and eigenvalue searches) problems subject to vacuum, reflective, white, or source boundary conditions are solved. General anisotropic scattering is allowed and anisotropic-distributed sources are permitted. The discrete-ordinates approximation is used for the neutron directional variables. An option is included to append a fictitious source to the discrete-ordinates equations that is defined such that spherical-harmonics solutions (in x-y geometry) or spherical-harmonics-like solutions (in r-z geometry) are obtained. A spatial-finite-element method is used in which the angular flux is expressed as a linear polynomial in each triangle that is discontinous at triangle boundaries. Unusual Features of the program: Provision is made for creation of standard interface output files for S/sub N/ constants, angle-integrated (scalar) fluxes, and angular fluxes. Standard interface input files for S/sub N/ constants, inhomogeneous sources, cross sections, and the scalar flux may be read. Flexible edit options as well as a dump and restart capability are provided

  15. TASK, 1-D Multigroup Diffusion or Transport Theory Reactor Kinetics with Delayed Neutron

    International Nuclear Information System (INIS)

    Buhl, A.R.; Hermann, O.W.; Hinton, R.J.; Dodds, H.L. Jr.; Robinson, J.C.; Lillie, R.A.

    1975-01-01

    1 - Description of problem or function: TASK solves the one-dimensional multigroup form of the reactor kinetics equations, using either transport or diffusion theory and allowing an arbitrary number of delayed neutron groups. The program can also be used to solve standard static problems efficiently such as eigenvalue problems, distributed source problems, and boundary source problems. Convergence problems associated with sources in highly multiplicative media are circumvented, and such problems are readily calculable. 2 - Method of solution: TASK employs a combination scattering and transfer matrix method to eliminate certain difficulties that arise in classical finite difference approximations. As such, within-group (inner) iterations are eliminated and solution convergence is independent of spatial mesh size. The time variable is removed by Laplace transformation. (A later version will permit direct time solutions.) The code can be run either in an outer iteration mode or in closed (non-iterative) form. The running mode is dictated by the number of groups times the number of angles, consistent with available storage. 3 - Restrictions on the complexity of the problem: The principal restrictions are available storage and computation time. Since the code is flexibly-dimensioned and has an outer iteration option there are no internal restrictions on group structure, quadrature, and number of ordinates. The flexible-dimensioning scheme allows optional use of core storage. The generalized cylindrical geometry option is not complete in Version I of the code. The feedback options and omega-mode search options are not included in Version I

  16. ESELEM 4: a code for calculating fine neutron spectrum and multi-group cross sections in plate lattice

    International Nuclear Information System (INIS)

    Nakagawa, Masayuki; Katsuragi, Satoru; Narita, Hideo.

    1976-07-01

    The multi-group treatment has been used in the design study of fast reactors and analysis of experiments at fast critical assemblies. The accuracy of the multi-group cross sections therefore affects strongly the results of these analyses. The ESELEM 4 code has been developed to produce multi-group cross sections with an advanced method from the nuclear data libraries used in the JAERI Fast set. ESELEM 4 solves integral transport equation by the collision probability method in plate lattice geometry to obtain the fine neutron spectrum. A typical fine group mesh width is 0.008 in lethargy unit. The multi-group cross sections are calculated by weighting the point data with the fine structure neutron flux. Some devices are applied to reduce computation time and computer core storage required for the calculation. The slowing down sources are calculated with the use of a recurrence formula derived for elastic and inelastic scattering. The broad group treatment is adopted above 2 MeV for dealing with both light any heavy elements. Also the resonance cross sections of heavy elements are represented in a broad group structure, for which we use the values of the JAERI Fast set. The library data are prepared by the PRESM code from ENDF/A type nuclear data files. The cross section data can be compactly stored in the fast computer core memory for saving the core storage and data processing time. The programme uses the variable dimensions to increase its flexibility. The users' guide for ESELEM 4 and PRESM is also presented in this report. (auth.)

  17. Advanced Neutron Source Cross Section Libraries (ANSL-V): ENDF/B-V based multigroup cross-section libraries for advanced neutron source (ANS) reactor studies

    International Nuclear Information System (INIS)

    Ford, W.E. III; Arwood, J.W.; Greene, N.M.; Moses, D.L.; Petrie, L.M.; Primm, R.T. III; Slater, C.O.; Westfall, R.M.; Wright, R.Q.

    1990-09-01

    Pseudo-problem-independent, multigroup cross-section libraries were generated to support Advanced Neutron Source (ANS) Reactor design studies. The ANS is a proposed reactor which would be fueled with highly enriched uranium and cooled with heavy water. The libraries, designated ANSL-V (Advanced Neutron Source Cross Section Libraries based on ENDF/B-V), are data bases in AMPX master format for subsequent generation of problem-dependent cross-sections for use with codes such as KENO, ANISN, XSDRNPM, VENTURE, DOT, DORT, TORT, and MORSE. Included in ANSL-V are 99-group and 39-group neutron, 39-neutron-group 44-gamma-ray-group secondary gamma-ray production (SGRP), 44-group gamma-ray interaction (GRI), and coupled, 39-neutron group 44-gamma-ray group (CNG) cross-section libraries. The neutron and SGRP libraries were generated primarily from ENDF/B-V data; the GRI library was generated from DLC-99/HUGO data, which is recognized as the ENDF/B-V photon interaction data. Modules from the AMPX and NJOY systems were used to process the multigroup data. Validity of selected data from the fine- and broad-group neutron libraries was satisfactorily tested in performance parameter calculations

  18. VENTURE: a code block for solving multigroup neutronics problems applying the finite-difference diffusion-theory approximation to neutron transport, version II

    International Nuclear Information System (INIS)

    Vondy, D.R.; Fowler, T.B.; Cunningham, G.W.

    1977-11-01

    The report documents the computer code block VENTURE designed to solve multigroup neutronics problems with application of the finite-difference diffusion-theory approximation to neutron transport (or alternatively simple P 1 ) in up to three-dimensional geometry. It uses and generates interface data files adopted in the cooperative effort sponsored by the Reactor Physics Branch of the Division of Reactor Research and Development of the Energy Research and Development Administration. Several different data handling procedures have been incorporated to provide considerable flexibility; it is possible to solve a wide variety of problems on a variety of computer configurations relatively efficiently

  19. VENTURE: a code block for solving multigroup neutronics problems applying the finite-difference diffusion-theory approximation to neutron transport, version II. [LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    Vondy, D.R.; Fowler, T.B.; Cunningham, G.W.

    1977-11-01

    The report documents the computer code block VENTURE designed to solve multigroup neutronics problems with application of the finite-difference diffusion-theory approximation to neutron transport (or alternatively simple P/sub 1/) in up to three-dimensional geometry. It uses and generates interface data files adopted in the cooperative effort sponsored by the Reactor Physics Branch of the Division of Reactor Research and Development of the Energy Research and Development Administration. Several different data handling procedures have been incorporated to provide considerable flexibility; it is possible to solve a wide variety of problems on a variety of computer configurations relatively efficiently.

  20. VENTURE: a code block for solving multigroup neutronics problems applying the finite-difference diffusion-theory approximation to neutron transport

    International Nuclear Information System (INIS)

    Vondy, D.R.; Fowler, T.B.; Cunningham, G.W.

    1975-10-01

    The computer code block VENTURE, designed to solve multigroup neutronics problems with application of the finite-difference diffusion-theory approximation to neutron transport (or alternatively simple P 1 ) in up to three-dimensional geometry is described. A variety of types of problems may be solved: the usual eigenvalue problem, a direct criticality search on the buckling, on a reciprocal velocity absorber (prompt mode), or on nuclide concentrations, or an indirect criticality search on nuclide concentrations, or on dimensions. First-order perturbation analysis capability is available at the macroscopic cross section level

  1. FENDL multigroup libraries

    International Nuclear Information System (INIS)

    Ganesan, S.; Muir, D.W.

    1992-01-01

    Selected neutron reaction nuclear data libraries and photon-atomic interaction cross section libraries for elements of interest to the IAEA's program on Fusion Evaluated Nuclear Data Library (FENDL) have been processed into MATXSR format using the NJOY system on the VAX4000 computer of the IAEA. This document lists the resulting multigroup data libraries. All the multigroup data generated are available cost-free upon request from the IAEA Nuclear Data Section. (author). 9 refs

  2. VARI-QUIR-3, 2-D Multigroup Steady-State Neutron Diffusion in X-Y R-Z or R-Theta Geometry

    International Nuclear Information System (INIS)

    Collier, George

    1984-01-01

    1 - Nature of physical problem solved: The steady-state, multigroup, two-dimensional neutron diffusion equations are solved in x-y, r-z, and r-theta geometry. 2 - Method of solution: A Gauss-Seidel type of solution with inner and outer iterations is used. The source is held constant during the inner iterations

  3. MVP/GMVP II, MC Codes for Neutron and Photon Transport Calc. based on Continuous Energy and Multigroup Methods

    International Nuclear Information System (INIS)

    2005-01-01

    A - Description of program or function: (1) Problems to be solved: MVP/GMVP can solve eigenvalue and fixed-source problems. The multigroup code GMVP can solve forward and adjoint problems for neutron, photon and neutron-photon coupled transport. The continuous-energy code MVP can solve only the forward problems. Both codes can also perform time-dependent calculations. (2) Geometry description: MVP/GMVP employs combinatorial geometry to describe the calculation geometry. It describes spatial regions by the combination of the 3-dimensional objects (BODIes). Currently, the following objects (BODIes) can be used. - BODIes with linear surfaces: half space, parallelepiped, right parallelepiped, wedge, right hexagonal prism; - BODIes with quadratic surface and linear surfaces: cylinder, sphere, truncated right cone, truncated elliptic cone, ellipsoid by rotation, general ellipsoid; - Arbitrary quadratic surface and torus. The rectangular and hexagonal lattice geometry can be used to describe the repeated geometry. Furthermore, the statistical geometry model is available to treat coated fuel particles or pebbles for high temperature reactors. (3) Particle sources: The various forms of energy-, angle-, space- and time-dependent distribution functions can be specified. (4) Cross sections: The ANISN-type PL cross sections or the double-differential cross sections can be used in the multigroup code GMVP. On the other hand, the specific cross section libraries are used in the continuous-energy code MVP. The libraries are generated from the evaluated nuclear data (JENDL-3.3, ENDF/B-VI, JEF-3.0 etc.) by using the LICEM code. The neutron cross sections in the unresolved resonance region are described by the probability table method. The neutron cross sections at arbitrary temperatures are available for MVP by just specifying the temperatures in the input data. (5) Boundary conditions: Vacuum, perfect reflective, isotropic reflective (white), periodic boundary conditions can be

  4. REX1-87, Multigroup Neutron Cross-Sections from ENDF/B

    International Nuclear Information System (INIS)

    Gopalakrishnan, V.; Ganesan, S.

    1988-01-01

    1 - Description of program or function: The program calculates self- shielding factors for reactor applications from a pre-processed (linearized) evaluated nuclear data file in the ENDF/B format. 2 - Method of solution: Bondarenko definition of multigroup self- shielding factors invoking narrow resonance treatment is used. 3 - Restrictions on the complexity of the problem: a) Maximum no. of energy group is 620. b) Only the built-in forms of the weighting functions can be chosen. c) The program is strictly limited to resolved resonance region from physical considerations

  5. A multi-region boundary element method for multigroup neutron diffusion calculations

    International Nuclear Information System (INIS)

    Ozgener, H.A.; Ozgener, B.

    2001-01-01

    For the analysis of a two-dimensional nuclear system consisting of a number of homogeneous regions (termed cells), first the cell matrices which depend solely on the material composition and geometrical dimension of the cell (hence on the cell type) are constructed using a boundary element formulation based on the multigroup boundary integral equation. For a particular nuclear system, the cell matrices are utilized in the assembly of the global system matrix in block-banded form using the newly introduced concept of virtual side. For criticality calculations, the classical fission source iteration is employed and linear system solutions are by the block Gaussian-elimination algorithm. The numerical applications show the validity of the proposed formulation both through comparison with analytical solutions and assessment of benchmark problem results against alternative methods

  6. Investigation of the response of a neutron moisture meter using a multigroup, two-dimensional diffusion theory code

    International Nuclear Information System (INIS)

    Ritchie, A.I.M.; Wilson, D.J.

    1984-12-01

    A multigroup diffusion code has been used to predict the count rate from a neutron moisture meter for a range of values of soil water content ω, thermal neutron absorption cross section Ssub(a) (defined as Σsub(a)/rho) of the soil matrix and soil matrix density rho. Two dimensions adequately approximated the geometry of the source, detector and soil surrounding the detector. Seven energy groups, the data for which were condensed from 128 group data set over the neutron energy spectrum appropriate to the soil-water mixture under study, proved adequate to describe neutron slowing-down and diffusion. The soil-water mixture was an SiO 2 →water mixture, with the absorption cross section of SiO 2 increased to cover the range of Σsub(a) required. The response to changes in matrix density is, in general, linear but the response to changes in water content is not linear over the range of parameter values investigated. Tabular results are presented which allow interpolation of the response for a particular ω, Ssub(a) and rho. It is shown that R(ω, Ssub(a), rho) rho M(Ssub(a)) + C(ω) is a crude representation of the response over a very limited range of variation of ω, and Ssub(a). As the response is a slowly varying function of rho, Ssub(a) and ω, a polynomial fit will provide a better estimate of the response for values of rho, Ssub(a) and ω not tabulated

  7. A new modelling of the multigroup scattering cross section in deterministic codes for neutron transport

    International Nuclear Information System (INIS)

    Calloo, A.A.

    2012-01-01

    In reactor physics, calculation schemes with deterministic codes are validated with respect to a reference Monte Carlo code. The remaining biases are attributed to the approximations and models induced by the multigroup theory (self-shielding models and expansion of the scattering law using Legendre polynomials) to represent physical phenomena (resonant absorption and scattering anisotropy respectively). This work focuses on the relevance of a polynomial expansion to model the scattering law. Since the outset of reactor physics, the latter has been expanded on a truncated Legendre polynomial basis. However, the transfer cross sections are highly anisotropic, with non-zero values for a very small range of the cosine of the scattering angle. Besides, the finer the energy mesh and the lighter the scattering nucleus, the more exacerbated is the peaked shape of this cross section. As such, the Legendre expansion is less suited to represent the scattering law. Furthermore, this model induces negative values which are non-physical. In this work, various scattering laws are briefly described and the limitations of the existing model are pointed out. Hence, piecewise-constant functions have been used to represent the multigroup scattering cross section. This representation requires a different model for the diffusion source. The discrete ordinates method which is widely employed to solve the transport equation has been adapted. Thus, the finite volume method for angular discretization has been developed and implemented in Paris environment which hosts the S n solver, Snatch. The angular finite volume method has been compared to the collocation method with Legendre moments to ensure its proper performance. Moreover, unlike the latter, this method is adapted for both the Legendre moments and the piecewise-constant functions representations of the scattering cross section. This hybrid-source method has been validated for different cases: fuel cell in infinite lattice

  8. TEMPEST-2, Thermalization Program for Neutron Spectra and Multigroup Cross-Sections

    International Nuclear Information System (INIS)

    Gowins, G.

    1984-01-01

    Description of problem or function: TEMPEST2 is a neutron thermalization program based upon the Wigner-Wilkins approximation for light moderators and the Wilkins approximation for heavy moderators. A Maxwellian distribution may also be used. The model used may be selected as a function of energy. The second-order differential equations are integrated directly rather than transformed to the Riccati equation. The program provides microscopic and macroscopic cross-section averages over the thermal neutron spectrum

  9. Creation and validation of a neutron-gamma coupled multigroup cross section library

    International Nuclear Information System (INIS)

    Devan, K.; Gopalakrishnan, V.; Lee, S.M.

    1995-01-01

    The task of creating our own neutron-gamma coupled library was taken up. By using 1985 version of NJOY code system, a coupled set called IGC-DE4-S1 in ANISN format for 25 nuclides has been arrived at based on ENDF/B-IV neutron library and DLC-99 gamma library, with Legendre order of up to 5. The flow chart for the creation of coupled set is given. 5 refs, 1 fig., 3 tabs

  10. Status report on multigroup cross section generation code development for high-fidelity deterministic neutronics simulation system

    International Nuclear Information System (INIS)

    Yang, W.S.; Lee, C.H.

    2008-01-01

    Under the fast reactor simulation program launched in April 2007, development of an advanced multigroup cross section generation code was initiated in July 2007, in conjunction with the development of the high-fidelity deterministic neutron transport code UNIC. The general objectives are to simplify the existing multi-step schemes and to improve the resolved and unresolved resonance treatments. Based on the review results of current methods and the fact that they have been applied successfully to fast critical experiment analyses and fast reactor designs for last three decades, the methodologies of the ETOE-2/MC 2 -2/SDX code system were selected as the starting set of methodologies for multigroup cross section generation for fast reactor analysis. As the first step for coupling with the UNIC code and use in a parallel computing environment, the MC 2 -2 code was updated by modernizing the memory structure and replacing old data management package subroutines and functions with FORTRAN 90 based routines. Various modifications were also made in the ETOE-2 and MC 2 -2 codes to process the ENDF/B-VII.0 data properly. Using the updated ETOE-2/MC 2 -2 code system, the ENDF/B-VII.0 data was successfully processed for major heavy and intermediate nuclides employed in sodium-cooled fast reactors. Initial verification tests of the MC 2 -2 libraries generated from ENDF/B-VII.0 data were performed by inter-comparison of twenty-one group infinite dilute total cross sections obtained from MC 2 -2, VIM, and NJOY. For almost all nuclides considered, MC 2 -2 cross sections agreed very well with those from VIM and NJOY. Preliminary validation tests of the ENDF/B-VII.0 libraries of MC 2 -2 were also performed using a set of sixteen fast critical benchmark problems. The deterministic results based on MC 2 -2/TWODANT calculations were in good agreement with MCNP solutions within ∼0.25% Δρ, except a few small LANL fast assemblies. Relative to the MCNP solution, the MC 2 -2/TWODANT

  11. Status report on multigroup cross section generation code development for high-fidelity deterministic neutronics simulation system.

    Energy Technology Data Exchange (ETDEWEB)

    Yang, W. S.; Lee, C. H. (Nuclear Engineering Division)

    2008-05-16

    Under the fast reactor simulation program launched in April 2007, development of an advanced multigroup cross section generation code was initiated in July 2007, in conjunction with the development of the high-fidelity deterministic neutron transport code UNIC. The general objectives are to simplify the existing multi-step schemes and to improve the resolved and unresolved resonance treatments. Based on the review results of current methods and the fact that they have been applied successfully to fast critical experiment analyses and fast reactor designs for last three decades, the methodologies of the ETOE-2/MC{sup 2}-2/SDX code system were selected as the starting set of methodologies for multigroup cross section generation for fast reactor analysis. As the first step for coupling with the UNIC code and use in a parallel computing environment, the MC{sup 2}-2 code was updated by modernizing the memory structure and replacing old data management package subroutines and functions with FORTRAN 90 based routines. Various modifications were also made in the ETOE-2 and MC{sup 2}-2 codes to process the ENDF/B-VII.0 data properly. Using the updated ETOE-2/MC{sup 2}-2 code system, the ENDF/B-VII.0 data was successfully processed for major heavy and intermediate nuclides employed in sodium-cooled fast reactors. Initial verification tests of the MC{sup 2}-2 libraries generated from ENDF/B-VII.0 data were performed by inter-comparison of twenty-one group infinite dilute total cross sections obtained from MC{sup 2}-2, VIM, and NJOY. For almost all nuclides considered, MC{sup 2}-2 cross sections agreed very well with those from VIM and NJOY. Preliminary validation tests of the ENDF/B-VII.0 libraries of MC{sup 2}-2 were also performed using a set of sixteen fast critical benchmark problems. The deterministic results based on MC{sup 2}-2/TWODANT calculations were in good agreement with MCNP solutions within {approx}0.25% {Delta}{rho}, except a few small LANL fast assemblies

  12. NEUTRON AND PHOTON DOSE MAPPING OF A DD NEUTRON GENERATOR.

    Science.gov (United States)

    Metwally, Walid A; Taqatqa, Osama A; Ballaith, Mohammed M; Chen, Allan X; Piestrup, Melvin A

    2017-11-01

    Neutron generators are an excellent tool that can be effectively utilized in educational institutions for applications such as neutron activation analysis, neutron radiography, and profiling and irradiation effects. For safety purposes, it is imperative that appropriate measures are taken in order to minimize the radiation dose from such devices to the operators, students and the public. This work presents the simulation and measurement results for the neutron and photon dose rates in the vicinity of the neutron generator installed at the University of Sharjah. A very good agreement is found between the simulated and measured dose rates. All of the public dose constraints were found to be met. The occupational dose constraint was also met after imposing a 200 cm no entry zone around the generator room. © The Author 2017. Published by Oxford University Press. All rights reserved. For Permissions, please email: journals.permissions@oup.com.

  13. Spectral nodal methodology for multigroup slab-geometry discrete ordinates neutron transport problems with linearly anisotropic scattering

    Energy Technology Data Exchange (ETDEWEB)

    Oliva, Amaury M.; Filho, Hermes A.; Silva, Davi M.; Garcia, Carlos R., E-mail: aoliva@iprj.uerj.br, E-mail: halves@iprj.uerj.br, E-mail: davijmsilva@yahoo.com.br, E-mail: cgh@instec.cu [Universidade do Estado do Rio de Janeiro (UERJ), Nova Friburgo, RJ (Brazil). Instituto Politecnico. Departamento de Modelagem Computacional; Instituto Superior de Tecnologias y Ciencias Aplicadas (InSTEC), La Habana (Cuba)

    2017-07-01

    In this paper, we propose a numerical methodology for the development of a method of the spectral nodal class that will generate numerical solutions free from spatial truncation errors. This method, denominated Spectral Deterministic Method (SDM), is tested as an initial study of the solutions (spectral analysis) of neutron transport equations in the discrete ordinates (S{sub N}) formulation, in one-dimensional slab geometry, multigroup approximation, with linearly anisotropic scattering, considering homogeneous and heterogeneous domains with fixed source. The unknowns in the methodology are the cell-edge, and cell average angular fluxes, the numerical values calculated for these quantities coincide with the analytic solution of the equations. These numerical results are shown and compared with the traditional ne- mesh method Diamond Difference (DD) and the coarse-mesh method spectral Green's function (SGF) to illustrate the method's accuracy and stability. The solution algorithms problems are implemented in a computer simulator made in C++ language, the same that was used to generate the results of the reference work. (author)

  14. Analytical solution of the multigroup neutron diffusion kinetic equation in one-dimensional cartesian geometry by the integral transform technique

    International Nuclear Information System (INIS)

    Ceolin, Celina

    2010-01-01

    The objective of this work is to obtain an analytical solution of the neutron diffusion kinetic equation in one-dimensional cartesian geometry, to monoenergetic and multigroup problems. These equations are of the type stiff, due to large differences in the orders of magnitude of the time scales of the physical phenomena involved, which make them difficult to solve. The basic idea of the proposed method is applying the spectral expansion in the scalar flux and in the precursor concentration, taking moments and solving the resulting matrix problem by the Laplace transform technique. Bearing in mind that the equation for the precursor concentration is a first order linear differential equation in the time variable, to enable the application of the spectral method we introduce a fictitious diffusion term multiplied by a positive value which tends to zero. This procedure opened the possibility to find an analytical solution to the problem studied. We report numerical simulations and analysis of the results obtained with the precision controlled by the truncation order of the series. (author)

  15. TIMEX, 1-D Time-Dependent Multigroup Transport Theory with Delayed Neutron, Planar Cylindrical and Spherical Geometry

    International Nuclear Information System (INIS)

    Hill, T. R.; Reed, W. H.

    1980-01-01

    1 - Description of problem or function: TIMEX solves the time- dependent, one-dimensional multigroup transport equation with delayed neutrons in plane, cylindrical, spherical, and two-angle plane geometries. Both regular and adjoint, inhomogeneous and homogeneous problems subject to vacuum, reflective, periodic, white, albedo or inhomogeneous boundary flux conditions are solved. General anisotropic scattering is allowed and anisotropic inhomogeneous sources are permitted. 2 - Method of solution: The discrete ordinates approximation for the angular variable is used with the diamond (central) difference approximation for the angular extrapolation in curved geometries. A linear discontinuous finite element representation for the angular flux in each spatial mesh cell is used. Negative fluxes are eliminated by a local set-to-zero and correct algorithm. The time variable is differenced by an explicit technique that is unconditionally stable so that arbitrarily large time-steps can be taken. Two acceleration methods, exponential extrapolation and re-balance, are utilized to improve the accuracy of the time differencing scheme. 3 - Restrictions on the complexity of the problem: Variable dimensioning is used so that any combination of problem parameters leading to a container array less than MAXCOR can be accommodated. In addition, the CDC version permits the use of extended core storage less than MAXECS

  16. On the exact solution for the multi-group kinetic neutron diffusion equation in a rectangle

    International Nuclear Information System (INIS)

    Petersen, C.Z.; Vilhena, M.T.M.B. de; Bodmann, B.E.J.

    2011-01-01

    In this work we consider the two-group bi-dimensional kinetic neutron diffusion equation. The solution procedure formalism is general with respect to the number of energy groups, neutron precursor families and regions with different chemical compositions. The fast and thermal flux and the delayed neutron precursor yields are expanded in a truncated double series in terms of eigenfunctions that, upon insertion into the kinetic equation and upon taking moments, results in a first order linear differential matrix equation with source terms. We split the matrix appearing in the transformed problem into a sum of a diagonal matrix plus the matrix containing the remaining terms and recast the transformed problem into a form that can be solved in the spirit of Adomian's recursive decomposition formalism. Convergence of the solution is guaranteed by the Cardinal Interpolation Theorem. We give numerical simulations and comparisons with available results in the literature. (author)

  17. Generation and performance of a multigroup coupled neutron-gamma cross-section library for deterministic and Monte Carlo borehole logging analysis

    International Nuclear Information System (INIS)

    Kodeli, I.; Aldama, D. L.; De Leege, P. F. A.; Legrady, D.; Hoogenboom, J. E.; Cowan, P.

    2004-01-01

    As part of the IRTMBA (Improved Radiation Transport Modelling for Borehole Applications) project of the EU community's 5. framework program a special purpose multigroup cross-section library was prepared for use in deterministic and Monte Carlo oil well logging particle transport calculations. This library is expected to improve the prediction of the neutron and gamma spectra at the detector positions of the logging tool, and their use for the interpretation of the neutron logging measurements was studied. Preparation and testing of this library is described. (authors)

  18. HEXAGA-III-120, -30. Three dimensional multi-group neutron diffusion programmes for a uniform triangular mesh with arbitrary group scattering

    International Nuclear Information System (INIS)

    Woznicki, Z.I.

    1983-07-01

    This report presents the HEXAGA-III-programme solving multi-group time-independent real and/or adjoint neutron diffusion equations for three-dimensional-triangular-z-geometry. The method of solution is based on the AGA two-sweep iterative method belonging to the family of factorization techniques. An arbitrary neutron scattering model is permitted. The report written for users provides the description of the programme input and output and the use of HEXAGA-III is illustrated by a sample reactor problem. (orig.) [de

  19. A neutron dose equivalent meter at CAEP

    International Nuclear Information System (INIS)

    Tian Shihai; Lu Yan; Wang Heyi; Yuan Yonggang; Chen Xu

    2012-01-01

    The measurement of neutron dose equivalent has been a widespread need in industry and research. In this paper, aimed at improving the accuracy of neutron dose equivalent meter: a neutron dose counter is simulated with MCNP5, and the energy response curve is optimized. The results show that the energy response factor is from 0.2 to 1.8 for neutrons in the energy range of 2.53×10 -8 MeV to 10 MeV Compared with other related meters, it turns that the design of this meter is right. (authors)

  20. Procedure to Generate the MPACT Multigroup Library

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Kang Seog [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-12-17

    The CASL neutronics simulator MPACT is under development for the neutronics and T-H coupled simulation for the light water reactor. The objective of this document is focused on reviewing the current procedure to generate the MPACT multigroup library. Detailed methodologies and procedures are included in this document for further discussion to improve the MPACT multigroup library.

  1. Development of 3D multi-group neutron diffusion code for hexagonal geometry

    International Nuclear Information System (INIS)

    Sun Wei; Wang Kan; Ni Dongyang; Li Qing

    2013-01-01

    Based on the theory of new flux expansion nodal method to solve the neutron diffusion equations, the intra-nodal fluence rate distribution was expanded in a series of analytic basic functions for each group. In order to improve the accuracy of calculation result, continuities of neutron fluence rate and current were utilized across the nodal surfaces. According to the boundary conditions, the iteration method was adopted to solve the diffusion equation, where inner iteration speedup method is Gauss-Seidel method and outer is Lyusternik-Wagner. A new speedup method (one-outer-iteration and multi-inner-iteration method) was proposed according to the characteristic that the convergence speed of multiplication factor is faster than that of neutron fluence rate and the update of inner iteration matrix is slow. Based on the proposed model, the code HANDF-D was developed and tested by 3D two-group vver440 benchmark, experiment 2 of HFETR, 3D four-group thermal reactor benchmark, and 3D seven-group fast reactor benchmark. The numerical results show that HANDF-D can predict accurately the multiplication factor and nodal powers. (authors)

  2. The numerical analysis of eigenvalue problem solutions in the multigroup neutron diffusion theory

    International Nuclear Information System (INIS)

    Woznicki, Z.I.

    1994-01-01

    The main goal of this paper is to present a general iteration strategy for solving the discrete form of multidimensional neutron diffusion equations equivalent mathematically to an eigenvalue problem. Usually a solution method is based on different levels of iterations. The presented matrix formalism allows us to visualize explicitly how the used matrix splitting influences the matrix structure in an eigenvalue problem to be solved as well as the interdependence between inner and outer iteration within global iterations. Particular interactive strategies are illustrated by numerical results obtained for several reactor problems. (author). 21 refs, 32 figs, 15 tabs

  3. The numerical analysis of eigenvalue problem solutions in the multigroup neutron diffusion theory

    Energy Technology Data Exchange (ETDEWEB)

    Woznicki, Z I [Institute of Atomic Energy, Otwock-Swierk (Poland)

    1994-12-31

    The main goal of this paper is to present a general iteration strategy for solving the discrete form of multidimensional neutron diffusion equations equivalent mathematically to an eigenvalue problem. Usually a solution method is based on different levels of iterations. The presented matrix formalism allows us to visualize explicitly how the used matrix splitting influences the matrix structure in an eigenvalue problem to be solved as well as the interdependence between inner and outer iteration within global iterations. Particular interactive strategies are illustrated by numerical results obtained for several reactor problems. (author). 21 refs, 32 figs, 15 tabs.

  4. The numerical analysis of eigenvalue problem solutions in multigroup neutron diffusion theory

    International Nuclear Information System (INIS)

    Woznicki, Z.I.

    1995-01-01

    The main goal of this paper is to present a general iteration strategy for solving the discrete form of multidimensional neutron diffusion equations equivalent mathematically to an eigenvalue problem. Usually a solution method is based on different levels of iterations. The presented matrix formalism allows us to visualize explicitly how the used matrix splitting influences the matrix structure in an eigenvalue problem to be solved as well as the interdependence between inner and outer iterations within global iterations. Particular iterative strategies are illustrated by numerical results obtained for several reactor problems. (author). 21 refs, 35 figs, 16 tabs

  5. Nodal deterministic simulation for problems of neutron shielding in multigroup formulation

    International Nuclear Information System (INIS)

    Baptista, Josue Costa; Heringer, Juan Diego dos Santos; Santos, Luiz Fernando Trindade; Alves Filho, Hermes

    2013-01-01

    In this paper, we propose the use of some computational tools, with the implementation of numerical methods SGF (Spectral Green's Function), making use of a deterministic model of transport of neutral particles in the study and analysis of a known and simplified problem of nuclear engineering, known in the literature as a problem of neutron shielding, considering the model with two energy groups. These simulations are performed in MatLab platform, version 7.0, and are presented and developed with the help of a Computer Simulator providing a friendly computer application for their utilities

  6. TIMEX: a time-dependent explicit discrete ordinates program for the solution of multigroup transport equations with delayed neutrons

    International Nuclear Information System (INIS)

    Hill, T.R.; Reed, W.H.

    1976-01-01

    TIMEX solves the time-dependent, one-dimensional multigroup transport equation with delayed neutrons in plane, cylindrical, spherical, and two-angle plane geometries. Both regular and adjoint, inhomogeneous and homogeneous problems subject to vacuum, reflective, periodic, white, albedo or inhomogeneous boundary flux conditions are solved. General anisotropic scattering is allowed and anisotropic inhomogeneous sources are permitted. The discrete ordinates approximation for the angular variable is used with the diamond (central) difference approximation for the angular extrapolation in curved geometries. A linear discontinuous finite element representation for the angular flux in each spatial mesh cell is used. The time variable is differenced by an explicit technique that is unconditionally stable so that arbitrarily large time steps can be taken. Because no iteration is performed the method is exceptionally fast in terms of computing time per time step. Two acceleration methods, exponential extrapolation and rebalance, are utilized to improve the accuracy of the time differencing scheme. Variable dimensioning is used so that any combination of problem parameters leading to a container array less than MAXCOR can be accommodated. The running time for TIMEX is highly problem-dependent, but varies almost linearly with the total number of unknowns and time steps. Provision is made for creation of standard interface output files for angular fluxes and angle-integrated fluxes. Five interface units (use of interface units is optional), five output units, and two system input/output units are required. A large bulk memory is desirable, but may be replaced by disk, drum, or tape storage. 13 tables, 9 figures

  7. Neutron absorbed dose in a pacemaker CMOS

    International Nuclear Information System (INIS)

    Borja H, C. G.; Guzman G, K. A.; Valero L, C.; Banuelos F, A.; Hernandez D, V. M.; Vega C, H. R.; Paredes G, L.

    2012-01-01

    The neutron spectrum and the absorbed dose in a Complementary Metal Oxide Semiconductor (CMOS), has been estimated using Monte Carlo methods. Eventually a person with a pacemaker becomes an oncology patient that must be treated in a linear accelerator. Pacemaker has integrated circuits as CMOS that are sensitive to intense and pulsed radiation fields. Above 7 MV therapeutic beam is contaminated with photoneutrons that could damage the CMOS. Here, the neutron spectrum and the absorbed dose in a CMOS cell was calculated, also the spectra were calculated in two point-like detectors in the room. Neutron spectrum in the CMOS cell shows a small peak between 0.1 to 1 MeV and a larger peak in the thermal region, joined by epithermal neutrons, same features were observed in the point-like detectors. The absorbed dose in the CMOS was 1.522 x 10 -17 Gy per neutron emitted by the source. (Author)

  8. Neutron absorbed dose in a pacemaker CMOS

    Energy Technology Data Exchange (ETDEWEB)

    Borja H, C. G.; Guzman G, K. A.; Valero L, C.; Banuelos F, A.; Hernandez D, V. M.; Vega C, H. R. [Universidad Autonoma de Zacatecas, Unidad Academica de Estudios Nucleares, Cipres No. 10, Fracc. La Penuela, 98068 Zacatecas (Mexico); Paredes G, L., E-mail: fermineutron@yahoo.com [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2012-06-15

    The neutron spectrum and the absorbed dose in a Complementary Metal Oxide Semiconductor (CMOS), has been estimated using Monte Carlo methods. Eventually a person with a pacemaker becomes an oncology patient that must be treated in a linear accelerator. Pacemaker has integrated circuits as CMOS that are sensitive to intense and pulsed radiation fields. Above 7 MV therapeutic beam is contaminated with photoneutrons that could damage the CMOS. Here, the neutron spectrum and the absorbed dose in a CMOS cell was calculated, also the spectra were calculated in two point-like detectors in the room. Neutron spectrum in the CMOS cell shows a small peak between 0.1 to 1 MeV and a larger peak in the thermal region, joined by epithermal neutrons, same features were observed in the point-like detectors. The absorbed dose in the CMOS was 1.522 x 10{sup -17} Gy per neutron emitted by the source. (Author)

  9. An Experiment of Robust Parallel Algorithm for the Eigenvalue problem of a Multigroup Neutron Diffusion based on modified FETI-DP : Part 2

    International Nuclear Information System (INIS)

    Chang, Jonghwa

    2014-01-01

    Today, we can use a computer cluster consist of a few hundreds CPUs with reasonable budget. Such computer system enables us to do detailed modeling of reactor core. The detailed modeling will improve the safety and the economics of a nuclear reactor by eliminating un-necessary conservatism or missing consideration. To take advantage of such a cluster computer, efficient parallel algorithms must be developed. Mechanical structure analysis community has studied the domain decomposition method to solve the stress-strain equation using the finite element methods. One of the most successful domain decomposition method in terms of robustness is FETI-DP. We have modified the original FETI-DP to solve the eigenvalue problem for the multi-group diffusion problem in previous study. In this study, we report the result of recent modification to handle the three-dimensional subdomain partitioning, and the sub-domain multi-group problem. Modified FETI-DP algorithm has been successfully applied for the eigenvalue problem of multi-group neutron diffusion equation. The overall CPU time is decreasing as number of sub-domains (partitions) is increasing. However, there may be a limit in decrement due to increment of the number of primal points will increase the CPU time spent by the solution of the global equation. Even distribution of computational load (criterion a) is important to achieve fast computation. The subdomain partition can be effectively performed using suitable graph theory partition package such as MeTIS

  10. An Experiment of Robust Parallel Algorithm for the Eigenvalue problem of a Multigroup Neutron Diffusion based on modified FETI-DP : Part 2

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Jonghwa [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-10-15

    Today, we can use a computer cluster consist of a few hundreds CPUs with reasonable budget. Such computer system enables us to do detailed modeling of reactor core. The detailed modeling will improve the safety and the economics of a nuclear reactor by eliminating un-necessary conservatism or missing consideration. To take advantage of such a cluster computer, efficient parallel algorithms must be developed. Mechanical structure analysis community has studied the domain decomposition method to solve the stress-strain equation using the finite element methods. One of the most successful domain decomposition method in terms of robustness is FETI-DP. We have modified the original FETI-DP to solve the eigenvalue problem for the multi-group diffusion problem in previous study. In this study, we report the result of recent modification to handle the three-dimensional subdomain partitioning, and the sub-domain multi-group problem. Modified FETI-DP algorithm has been successfully applied for the eigenvalue problem of multi-group neutron diffusion equation. The overall CPU time is decreasing as number of sub-domains (partitions) is increasing. However, there may be a limit in decrement due to increment of the number of primal points will increase the CPU time spent by the solution of the global equation. Even distribution of computational load (criterion a) is important to achieve fast computation. The subdomain partition can be effectively performed using suitable graph theory partition package such as MeTIS.

  11. On an analytical evaluation of the flux and dominant eigenvalue problem for the steady state multi-group multi-layer neutron diffusion equation

    Energy Technology Data Exchange (ETDEWEB)

    Ceolin, Celina; Schramm, Marcelo; Bodmann, Bardo Ernst Josef; Vilhena, Marco Tullio Mena Barreto de [Universidade Federal do Rio Grande do Sul, Porto Alegre (Brazil). Programa de Pos-Graduacao em Engenharia Mecanica; Bogado Leite, Sergio de Queiroz [Comissao Nacional de Energia Nuclear, Rio de Janeiro (Brazil)

    2014-11-15

    In this work the authors solved the steady state neutron diffusion equation for a multi-layer slab assuming the multi-group energy model. The method to solve the equation system is based on an expansion in Taylor Series resulting in an analytical expression. The results obtained can be used as initial condition for neutron space kinetics problems. The neutron scalar flux was expanded in a power series, and the coefficients were found by using the ordinary differential equation and the boundary and interface conditions. The effective multiplication factor k was evaluated using the power method. We divided the domain into several slabs to guarantee the convergence with a low truncation order. We present the formalism together with some numerical simulations.

  12. HEXAGA-II-120, -60, -30 two-dimensional multi-group neutron diffusion programmes for a uniform triangular mesh with arbitrary group scattering

    International Nuclear Information System (INIS)

    Woznicki, Z.

    1979-06-01

    This report presents the AGA two-sweep iterative methods belonging to the family of factorization techniques in their practical application in the HEXAGA-II two-dimensional programme to obtain the numerical solution to the multi-group, time-independent, (real and/or adjoint) neutron diffusion equations for a fine uniform triangular mesh. An arbitrary group scattering model is permitted. The report written for the users provides the description of input and output. The use of HEXAGA-II is illustrated by two sample reactor problems. (orig.) [de

  13. FENDL/MG. Library of multigroup cross sections in GENDF and MATXS format for neutron-photon transport calculations. Version 1.1 of March 1995. Summary documentation

    International Nuclear Information System (INIS)

    Pashchenko, A.B.; Wienke, H.; Ganesan, S.

    1996-01-01

    Selected neutron reaction nuclear data evaluations and photon-atomic interaction cross section libraries for elements of interest to the IAEA's program on Fusion Evaluated Nuclear Data Library (FENDL) have been processed into GENDF and MATXS format using the NJOY system by R.E. MacFarlane, in VITAMIN-J group structure with VITAMIN-E weighting spectrum. This document summarizes the resulting multigroup data library FENDL/MG version 1.1. The data are available costfree, upon request from the IAEA Nuclear Data Section, online or on magnetic tape. (author). 7 refs, 1 tab

  14. Space synthesis: an application of synthesis method to two and three dimensional multigroup neutron diffusion equations; Synthese spatiale: une application de la methode de synthese aux equations de diffusion neutronique multigroupe a deux et trois dimensions

    Energy Technology Data Exchange (ETDEWEB)

    Nguyen-Ngoc, H [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1969-07-01

    In order to reduce computing time, two and three-dimensional multigroup neutron diffusion equations in cylindrical, rectangular (X, Y), (X, Y, Z) and hexagonal geometries are solved by the method of synthesis using an appropriate variational principle (stationary principle). The basic idea is to reduce the number of independent variables by constructing two or three-dimensional solutions from solutions of fewer variables, hence the name 'synthesis method'. Whatever the geometry, we are led to solve a system of ordinary differential equations with matrix coefficients to which one can apply well-known numerical methods: CHEBYSHEV's polynomial method, Gaussian elimination. Numerical results furnished by synthesis programs written for the IBM 7094, the IBM 360-75 and the CDC 6600 computers, are confronted with those which are given by programs employing the classical finite difference method. [French] En vue de reduire le-temps de calcul, les equations de diffusion neutronique, multigroupe, a deux et trois dimensions d'espace dans les geometries cylindrique, rectangulaire (X, Y), (X, Y, Z) et hexagonale sont resolues par la methode de synthese utilisant un principe variationnel approprie (principe stationnaire). L'idee consiste a reduire le nombre de variables independantes par construction d'une solution bi ou tridimensionnelle au moyen de solutions dependant d'un nombre inferieur de variables, d'ou le nom de la methode. Dans tous les cas de geometrie, nous sommes conduits a resoudre un systeme d'equations differentielles a coefficients matriciels auquel peuvent s'appliquer les methodes numeriques courantes; methode polynomiale de TCHEBYCHEFF et methode d'elimination de GAUSS. Les resultats numeriques obtenus par nos codes de synthese programmes sur IBM 7094, IBM 360-75 et CDC 6600, sont confrontes avec ceux que fournissent les programmes adoptant la methode classique des differences finies. (auteur)

  15. Monte Carlo simulated dose to the human body due to neutrons emitted in laser-fusion

    International Nuclear Information System (INIS)

    Gileadi, A.E.; Cohen, M.O.

    1977-01-01

    Considering a point neutron source located at a given distance from the human body, modeled by a 'standard reference man' phantom, neutron doses to the whole body, as well as to selected organs thereof, are determined, using the SAM-CE system, a Monte Carlo computer code, written in Fortran and designed to solve time, space and energy dependent neutron and gamma ray transport equations in complex three-dimensional geometrice. Collision density, energy deposition and dose are treated in the SAM-CE system as flux functionals. A special feature of SAM-CE is its use of the 'Combinatorial Geometry' technique which affords the user geometric capabilities exceeding those available with other commonly used geometric packages. All neutron and gamma ray cross section data, as well as gamma ray production data, are derived from the ENDF libraries. Both resolved and unresolved resonance parameters from ENDF neutron data files are treated automatically and extremely precise and detailed descriptions of cross section behavior is permitted. Such treatment avoids the ambiguities usually associated with multi-group codes, which use flux-averaged cross sections based on assumed flux distributions which may or may not be appropriate. The 'standard reference man', a heterogeneous phantom, uses simple geometric forms to approximate the shape and dimensions of the human body. Materials composition of each subregion representing a certain 'organ' is given. Typical values of neutron doses to the whole body and to selected 'organs' of interest are presented

  16. ZZ TEMPEST/MUFT, Thermal Neutron and Fast Neutron Multigroup Cross-Section Library for Program LEOPARD

    International Nuclear Information System (INIS)

    Kim, Jung-Do; Lee, Jong Tai

    1986-01-01

    Description of problem or function: Format: TEMPEST and MUFT; Number of groups: 246 thermal groups in TEMPEST Format and 54 fast groups in MUFT Format. From this library, the program SPOTS4 generates a 172-54 group library as input to the code LEOPARD. Nuclides: H, O, Zr, C, Fe, Ni, Al, Cr, Mn, U, Pu, Th, Pa, Xe, Sm, B and D. Origin: ENDF/B-4; Weighting spectrum: 1/E + U 235 fission spectrum. Data library of thermal and fast neutron group Cross sections to generate input to the program LEOPARD. The data is based on ENDF/B-4 and consists of two parts: (1) 246 thermal groups in TEMPEST Format. (2) 54 fast groups in MUFT Format. From this library, the program SPOTS4 generates a 172-54 group library as input to the code LEOPARD (NESC0279)

  17. Experimental evaluation of neutron dose in radiotherapy patients: Which dose?

    Energy Technology Data Exchange (ETDEWEB)

    Romero-Expósito, M., E-mail: mariateresa.romero@uab.cat; Domingo, C.; Ortega-Gelabert, O.; Gallego, S. [Grup de Recerca en Radiacions Ionizants (GRRI), Departament de Física, Universitat Autònoma de Barcelona, Bellaterra 08193 (Spain); Sánchez-Doblado, F. [Departamento de Fisiología Médica y Biofísica, Universidad de Sevilla, Sevilla 41009 (Spain); Servicio de Radiofísica, Hospital Universitario Virgen Macarena, Sevilla 41009 (Spain)

    2016-01-15

    Purpose: The evaluation of peripheral dose has become a relevant issue recently, in particular, the contribution of secondary neutrons. However, after the revision of the Recommendations of the International Commission on Radiological Protection, there has been a lack of experimental procedure for its evaluation. Specifically, the problem comes from the replacement of organ dose equivalent by the organ-equivalent dose, being the latter “immeasurable” by definition. Therefore, dose equivalent has to be still used although it needs the calculation of the radiation quality factor Q, which depends on the unrestricted linear energy transfer, for the specific neutron irradiation conditions. On the other hand, equivalent dose is computed through the radiation weighting factor w{sub R}, which can be easily calculated using the continuous function provided by the recommendations. The aim of the paper is to compare the dose equivalent evaluated following the definition, that is, using Q, with the values obtained by replacing the quality factor with w{sub R}. Methods: Dose equivalents were estimated in selected points inside a phantom. Two types of medical environments were chosen for the irradiations: a photon- and a proton-therapy facility. For the estimation of dose equivalent, a poly-allyl-diglicol-carbonate-based neutron dosimeter was used for neutron fluence measurements and, additionally, Monte Carlo simulations were performed to obtain the energy spectrum of the fluence in each point. Results: The main contribution to dose equivalent comes from neutrons with energy higher than 0.1 MeV, even when they represent the smallest contribution in fluence. For this range of energy, the radiation quality factor and the radiation weighting factor are approximately equal. Then, dose equivalents evaluated using both factors are compatible, with differences below 12%. Conclusions: Quality factor can be replaced by the radiation weighting factor in the evaluation of dose

  18. Neutron absorbed dose in a pacemaker CMOS

    Energy Technology Data Exchange (ETDEWEB)

    Borja H, C. G.; Guzman G, K. A.; Valero L, C. Y.; Banuelos F, A.; Hernandez D, V. M.; Vega C, H. R. [Universidad Autonoma de Zacatecas, Unidad Academica de Estudios Nucleares, Calle Cipres No. 10, Fracc. La Penuela, 98068 Zacatecas (Mexico); Paredes G, L., E-mail: candy_borja@hotmail.com [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2011-11-15

    The absorbed dose due to neutrons by a Complementary Metal Oxide Semiconductor (CMOS) has been estimated using Monte Carlo methods. Eventually a person with a pacemaker becomes a patient that must be treated by radiotherapy with a linear accelerator; the pacemaker has integrated circuits as CMOS that are sensitive to intense and pulsed radiation fields. When the Linac is working in Bremsstrahlung mode an undesirable neutron field is produced due to photoneutron reactions; these neutrons could damage the CMOS putting the patient at risk during the radiotherapy treatment. In order to estimate the neutron dose in the CMOS a Monte Carlo calculation was carried out where a full radiotherapy vault room was modeled with a W-made spherical shell in whose center was located the source term of photoneutrons produced by a Linac head operating in Bremsstrahlung mode at 18 MV. In the calculations a phantom made of tissue equivalent was modeled while a beam of photoneutrons was applied on the phantom prostatic region using a field of 10 x 10 cm{sup 2}. During simulation neutrons were isotropically transported from the Linac head to the phantom chest, here a 1 {theta} x 1 cm{sup 2} cylinder made of polystyrene was modeled as the CMOS, where the neutron spectrum and the absorbed dose were estimated. Main damages to CMOS are by protons produced during neutron collisions protective cover made of H-rich materials, here the neutron spectrum that reach the CMOS was calculated showing a small peak around 0.1 MeV and a larger peak in the thermal region, both connected through epithermal neutrons. (Author)

  19. Equivalent-spherical-shield neutron dose calculations

    International Nuclear Information System (INIS)

    Russell, G.J.; Robinson, H.

    1988-01-01

    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

  20. Neutron Dose Measurement Using a Cubic Moderator

    International Nuclear Information System (INIS)

    Sheinfeld, M.; Mazor, T.; Cohen, Y.; Kadmon, Y.; Orion, I.

    2014-01-01

    The Bonner Sphere Spectrometer (BSS), introduced In July 1960 by a research group from Rice University, Texas, is a major approach to neutron spectrum estimation. The BSS, also known as multi-sphere spectrometer, consists of a set of a different diameters polyethylene spheres, carrying a small LiI(Eu) scintillator in their center. What makes this spectrometry method such widely used, is its almost isotropic response, covering an extraordinary wide range of energies, from thermal up to even hundreds of MeVs. One of the most interesting and useful consequences of the above study is the 12'' sphere characteristics, as it turned out that the response curve of its energy dependence, have a similar shape compared with the neutron's dose equivalent as a function of energy. This inexplicable and happy circumstance makes it virtually the only monitoring device capable providing realistic neutron dose estimates over such a wide energy range. However, since the detection mechanism is not strictly related to radiation dose, one can expect substantial errors when applied to widely different source conditions. Although the original design of the BSS included a small 4mmx4mmO 6LiI(Eu) scintillator, other thermal neutron detectors has been used over the years: track detectors, activation foils, BF3 filled proportional counters, etc. In this study we chose a Boron loaded scintillator, EJ-254, as the thermal neutron detector. The neutron capture reaction on the boron has a Q value of 2.78 MeV of which 2.34 MeV is shared by the alpha and lithium particles. The high manufacturing costs, the encasement issue, the installation efficiency and the fabrication complexity, led us to the idea of replacing the sphere with a cubic moderator. This article describes the considerations, as well as the Monte-Carlo simulations done in order to examine the applicability of this idea

  1. Dose prescription in boron neutron capture therapy

    International Nuclear Information System (INIS)

    Gupta, N.M.S.; Gahbauer, R.A.; Blue, T.E.; Wambersie, A.

    1994-01-01

    The purpose of this paper is to address some aspects of the many considerations that need to go into a dose prescription in boron neutron capture therapy (BNCT) for brain tumors; and to describe some methods to incorporate knowledge from animal studies and other experiments into the process of dose prescription. Previously, an algorithm to estimate the normal tissue tolerance to mixed high and low linear energy transfer radiations in BNCT was proposed. The authors have developed mathematical formulations and computational methods to represent this algorithm. Generalized models to fit the central axis dose rate components for an epithermal neutron field were also developed. These formulations and beam fitting models were programmed into spreadsheets to simulate two treatment techniques which are expected to be used in BNCT: a two-field bilateral scheme and a single-field treatment scheme. Parameters in these spreadsheets can be varied to represent the fractionation scheme used, the 10 B microdistribution in normal tissue, and the ratio of 10 B in tumor to normal tissue. Most of these factors have to be determined for a given neutron field and 10 B compound combination from large animal studies. The spreadsheets have been programmed to integrate all of the treatment-related information and calculate the location along the central axis where the normal tissue tolerance is exceeded first. This information is then used to compute the maximum treatment time allowable and the maximum tumor dose that may be delivered for a given BNCT treatment. The effect of different treatment variables on the treatment time and tumor dose has been shown to be very significant. It has also been shown that the location of D max shifts significantly, depending on some of the treatment variables-mainly the fractionation scheme used. These results further emphasize the fact that dose prescription in BNCT is very complicated and nonintuitive. 11 refs., 6 figs., 3 tabs

  2. MVP/GMVP Version 3. General purpose Monte Carlo codes for neutron and photon transport calculations based on continuous energy and multigroup methods (Translated document)

    International Nuclear Information System (INIS)

    Nagaya, Yasunobu; Okumura, Keisuke; Sakurai, Takeshi; Mori, Takamasa

    2017-03-01

    In order to realize fast and accurate Monte Carlo simulation of neutron and photon transport problems, two Monte Carlo codes MVP (continuous-energy method) and GMVP (multigroup method) have been developed at Japan Atomic Energy Agency. The codes have adopted a vectorized algorithm and have been developed for vector-type supercomputers. They also support parallel processing with a standard parallelization library MPI and thus a speed-up of Monte Carlo calculations can be achieved on general computing platforms. The first and second versions of the codes were released in 1994 and 2005, respectively. They have been extensively improved and new capabilities have been implemented. The major improvements and new capabilities are as follows: (1) perturbation calculation for effective multiplication factor, (2) exact resonant elastic scattering model, (3) calculation of reactor kinetics parameters, (4) photo-nuclear model, (5) simulation of delayed neutrons, (6) generation of group constants. This report describes the physical model, geometry description method used in the codes, new capabilities and input instructions. (author)

  3. MVP/GMVP version 3. General purpose Monte Carlo codes for neutron and photon transport calculations based on continuous energy and multigroup methods

    International Nuclear Information System (INIS)

    Nagaya, Yasunobu; Okumura, Keisuke; Sakurai, Takeshi; Mori, Takamasa

    2017-03-01

    In order to realize fast and accurate Monte Carlo simulation of neutron and photon transport problems, two Monte Carlo codes MVP (continuous-energy method) and GMVP (multigroup method) have been developed at Japan Atomic Energy Agency. The codes have adopted a vectorized algorithm and have been developed for vector-type supercomputers. They also support parallel processing with a standard parallelization library MPI and thus a speed-up of Monte Carlo calculations can be achieved on general computing platforms. The first and second versions of the codes were released in 1994 and 2005, respectively. They have been extensively improved and new capabilities have been implemented. The major improvements and new capabilities are as follows: (1) perturbation calculation for effective multiplication factor, (2) exact resonant elastic scattering model, (3) calculation of reactor kinetics parameters, (4) photo-nuclear model, (5) simulation of delayed neutrons, (6) generation of group constants. This report describes the physical model, geometry description method used in the codes, new capabilities and input instructions. (author)

  4. Program SITHA. New version of the multigroup system of the neutron cross section GR175-V1

    International Nuclear Information System (INIS)

    Daniehl', A.V.; Perov, V.Yu.; Sokol, E.A.

    1991-01-01

    Structure of the system of the group neutron constants with band description of the resonance field and program system is described. Results of the testing calculations, made by code SITHA, for number of integral experiments devoted to measurements of the neutron flux from H 2 O,C,N,O and Al spherical targets are discussed. 13 refs.; 8 figs.; 1 tab

  5. Coupled multi-group neutron photon transport for the simulation of high-resolution gamma-ray spectroscopy applications

    Energy Technology Data Exchange (ETDEWEB)

    Burns, Kimberly A. [Georgia Inst. of Technology, Atlanta, GA (United States)

    2009-08-01

    The accurate and efficient simulation of coupled neutron-photon problems is necessary for several important radiation detection applications. Examples include the detection of nuclear threats concealed in cargo containers and prompt gamma neutron activation analysis for nondestructive determination of elemental composition of unknown samples.

  6. A multidimensional multigroup diffusion model for the determination of the frequency-dependent field of view of a neutron detector

    International Nuclear Information System (INIS)

    van der Hagen, T.H.J.J.; Hoogenboom, J.E.; van Dam, H.

    1992-01-01

    This paper reports on the sensitivity of a neutron detector to parametric fluctuations in the core of a reactor which depends on the position and the frequency of the perturbation. The basic neutron diffusion model for the calculation of this so-called field of view (FOV) of the detector is extended with respect to the dimensionality of the problem and the number of energy groups involved. The physical meaning of the FOV concept is illustrated by means of some simple examples, which can be handled analytically. The possibility of calculating the FOV by a conventional neutron diffusion code is demonstrated. In that case, the calculation in n neutron energy groups leads to 2n modified neutron diffusion equations

  7. MVP/GMVP 2: general purpose Monte Carlo codes for neutron and photon transport calculations based on continuous energy and multigroup methods

    International Nuclear Information System (INIS)

    Nagaya, Yasunobu; Okumura, Keisuke; Mori, Takamasa; Nakagawa, Masayuki

    2005-06-01

    In order to realize fast and accurate Monte Carlo simulation of neutron and photon transport problems, two vectorized Monte Carlo codes MVP and GMVP have been developed at JAERI. MVP is based on the continuous energy model and GMVP is on the multigroup model. Compared with conventional scalar codes, these codes achieve higher computation speed by a factor of 10 or more on vector super-computers. Both codes have sufficient functions for production use by adopting accurate physics model, geometry description capability and variance reduction techniques. The first version of the codes was released in 1994. They have been extensively improved and new functions have been implemented. The major improvements and new functions are (1) capability to treat the scattering model expressed with File 6 of the ENDF-6 format, (2) time-dependent tallies, (3) reaction rate calculation with the pointwise response function, (4) flexible source specification, (5) continuous-energy calculation at arbitrary temperatures, (6) estimation of real variances in eigenvalue problems, (7) point detector and surface crossing estimators, (8) statistical geometry model, (9) function of reactor noise analysis (simulation of the Feynman-α experiment), (10) arbitrary shaped lattice boundary, (11) periodic boundary condition, (12) parallelization with standard libraries (MPI, PVM), (13) supporting many platforms, etc. This report describes the physical model, geometry description method used in the codes, new functions and how to use them. (author)

  8. KAFAX-F22 : development and benchmark of multi-group library for fast reactor using JEF-2.2. Neutron 80 group and Photon 24 group

    International Nuclear Information System (INIS)

    Kim, Jung Do; Gil, Choong Sup.

    1997-03-01

    The KAFAX-F22 was developed from JEF-2.2, which is a MATXS format, multigroup library of fast reactor. The KAFAX-F22 has 80 and 24 energy group structures for neutron and photon, respectively. It includes 89 nuclide data processed by NJOY94.38. The TRANSX/TWODANT system was used for benchmark calculations of fast reactor and one- and two-dimensional calculations of ONEDANT and TWODANT were carried out with 80 group, P 3 S 16 and with 25 group, P 3 S 8 , respectively. The average values of multiplication factors are 0.99652 for MOX cores, 1.00538 for uranium cores and 1.00032 for total cores. Various central reaction rate ratios also give good agreements with the experimental values considering experimental uncertainties except for VERA-11A, VERA-1B, ZPR-6-7 and ZPR-6-6A cores of which experimental values seem to involve some problems. (author). 13 refs., 18 tabs., 2 figs

  9. Intermediate and fast neutron absorbed doses in fast neutron field at the RB reactor

    International Nuclear Information System (INIS)

    Sokcic-Kostic, M.; Pesic, M.; Antic, D.

    1987-10-01

    The experimental fuel channel EFC is created as one of the fast neutron fields at the RB reactor. The intermediate and fast neutron spectra in EFC are measured by activation technique. The intermediate and fast neutron absorbed doses are computed on the basis of these experimental results. At the end the obtained doses are compared. (author)

  10. Estimate of absorbed dose received by individuals irradiated with neutrons

    International Nuclear Information System (INIS)

    Fonseca, E.S. da; Mauricio, C.L.P.

    1995-01-01

    An innovating methodology is proposed to estimate the absorbed dose received by individuals irradiated with neutrons in an accident, even in the case that the victim is not using any kind of neutron dosemeter. The method combines direct measurements of 24 Na and 32 P activated in the human body. The calculation method was developed using data taken from previously published papers and experimental measurements. Other irradiations results in different neutron spectra prove the validity of the methodology here proposed. Using a whole body counter to measure 24 Na activity, it is possible to evaluate neutron absorbed doses in the order of 140 μGy of very soft (thermal) spectra. For fast neutron fields, the lower limit for neutron dose detection increases, but the present method continues to be very useful in accidents, with higher neutron doses. (author). 5 refs., 1 fig., 4 tabs

  11. Method and apparatus for determining the dose value of neutrons

    International Nuclear Information System (INIS)

    Burgkhardt, B.; Piesch, E.

    1976-01-01

    A method is provided for determining the dose value of neutrons leaving a body as thermal and intermediate neutrons after having been scattered in the body. A first dose value of thermal and intermediate neutrons is detected on the surface of the body by means of a first detector for neutrons which is shielded against thermal and intermediate neutrons not emerging from the body. A second detector is used to measure a second dose value of the thermal and intermediate neutrons not emerging from the body. A first correction factor based on the first and second values is obtained from a calibration diagram and is applied to the first dose value to determine a first corrected first dose value. 21 Claims, 6 Drawing Figures

  12. Neutrons in active proton therapy. Parameterization of dose and dose equivalent

    Energy Technology Data Exchange (ETDEWEB)

    Schneider, Uwe; Haelg, Roger A. [Univ. of Zurich (Switzerland). Dept. of Physics; Radiotherapy Hirslanden AG, Aarau (Switzerland); Lomax, Tony [Paul Scherrer Institute, Villigen (Switzerland). Center for Proton Therapy

    2017-08-01

    One of the essential elements of an epidemiological study to decide if proton therapy may be associated with increased or decreased subsequent malignancies compared to photon therapy is an ability to estimate all doses to non-target tissues, including neutron dose. This work therefore aims to predict for patients using proton pencil beam scanning the spatially localized neutron doses and dose equivalents. The proton pencil beam of Gantry 1 at the Paul Scherrer Institute (PSI) was Monte Carlo simulated using GEANT. Based on the simulated neutron dose and neutron spectra an analytical mechanistic dose model was developed. The pencil beam algorithm used for treatment planning at PSI has been extended using the developed model in order to calculate the neutron component of the delivered dose distribution for each treated patient. The neutron dose was estimated for two patient example cases. The analytical neutron dose model represents the three-dimensional Monte Carlo simulated dose distribution up to 85 cm from the proton pencil beam with a satisfying precision. The root mean square error between Monte Carlo simulation and model is largest for 138 MeV protons and is 19% and 20% for dose and dose equivalent, respectively. The model was successfully integrated into the PSI treatment planning system. In average the neutron dose is increased by 10% or 65% when using 160 MeV or 177 MeV instead of 138 MeV. For the neutron dose equivalent the increase is 8% and 57%. The presented neutron dose calculations allow for estimates of dose that can be used in subsequent epidemiological studies or, should the need arise, to estimate the neutron dose at any point where a subsequent secondary tumour may occur. It was found that the neutron dose to the patient is heavily increased with proton energy.

  13. Thermal neutron dose calculation in synovium membrane for BNCS

    International Nuclear Information System (INIS)

    Abdalla, Khalid; Naqvi, A.A.; Maalej, N.; El-Shahat, B.

    2006-01-01

    A D(d,n) reaction based setup has been optimized for Boron Neutron Capture Synovectomy (BNCS). The polyethylene moderator and graphite reflector sizes were optimized to deliver the highest ratio of thermal to fast neutron yield. The neutron dose was calculated at various depths in a knee phantom loaded with boron to determine therapeutic ratios of synovium dose/skin dose and synovium dose/bone dose. 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. (author)

  14. Does fast-neutron radiotherapy merely reduce the radiation dose

    International Nuclear Information System (INIS)

    Ando, Koichi

    1984-01-01

    We examined whether fast-neutron radiotherapy is superior to low-LET radiotherpy by comparing the relationship between cell survival and tumor control probabilities after exposure of tumor-bearing (species) to the two modalities. Analysis based on TCD 50 assay and lung colony assay indicated that single dose of fast neutron achieved animal cures at higher survival rates than other radiation modalities including single and fractionated γ-ray doses, fractionated doses of fast neutron, and the mixed-beam scheme with a sequence of N-γ-γ-γ-N. We conclude that fast-neutron radiotherapy cured animal tumors with lower cell killing rates other radiation modalities. (author)

  15. Generating and verification of ACE-multigroup library for MCNP

    International Nuclear Information System (INIS)

    Chen Chaobin; Hu Zehua; Chen Yixue; Wu Jun; Yang Shouhai

    2012-01-01

    The Monte Carlo code MCNP can handle multigroup calculations and a sample multigroup set based on ENDF/B-V, MGXSNP, is available for MCNP for coupled neutron-photon transport. However, this library is not suit- able for all problems, and there is a need for users to be able to generate multigroup libraries tailored to their specific applications. For these purposes CSPT (cross section processing tool) is created to generate multigroup library for MCNP from deterministic multigroup cross sections (GENDF or ANISN format at present). Several ACE-multigroup libraries based on ENDF/B-VII.0 converted and verified in this work, we drawn the conclusion that the CSPT code works correctly and the libraries produced are credible. (authors)

  16. The experimental method for neutron dose-equivalent detection

    International Nuclear Information System (INIS)

    Ji Changsong

    1992-01-01

    A new method, for getting neutron dose-equivalent Cd rode absorption method is described. The method adopts Cd-rode-swarm buck absorption, which greatly improved the neutron sensitivity and simplified the adjustment method. By this method, the author has developed BH3105 model neutron dose equivalent meter, the sensitivity of this instrument reach 10 cps/μSvh -1 . γ-ray depression rate reaches 4000:1, the measurement range is 0.1 μSv/h-10 6 μSv/h. The energy response is good (from thermal neutron-14 MeV neutron), this instrument can be used to measure the dose equivalent of the neutron areas

  17. Final report [on solving the multigroup diffusion equations

    International Nuclear Information System (INIS)

    Birkhoff, G.

    1975-01-01

    Progress achieved in the development of variational methods for solving the multigroup neutron diffusion equations is described. An appraisal is made of the extent to which improved variational methods could advantageously replace difference methods currently used

  18. HEXAGA-II. A two-dimensional multi-group neutron diffusion programme for a uniform triangular mesh with arbitrary group scattering for the IBM/370-168 computer

    International Nuclear Information System (INIS)

    Woznicki, Z.

    1976-05-01

    This report presents the AGA two-sweep iterative methods belonging to the family of factorization techniques in their practical application in the HEXAGA-II two-dimensional programme to obtain the numerical solution to the multi-group, time-independent, (real and/or adjoint) neutron diffusion equations for a fine uniform triangular mesh. An arbitrary group scattering model is permitted. The report written for the users provides the description of input and output. The use of HEXAGA-II is illustrated by two sample reactor problems. (orig.) [de

  19. Development of a neutron personal dose equivalent detector

    International Nuclear Information System (INIS)

    Tsujimura, N.; Yoshida, T.; Takada, C.; Momose, T.; Nunomiya, T.; Aoyama, K.

    2007-01-01

    A new neutron-measuring instrument that is intended to measure a neutron personal dose equivalent, H p (10) was developed. This instrument is composed of two parts: (1) a conventional moderator-based neutron dose equivalent meter and (2) a neutron shield made of borated polyethylene, which covers a backward hemisphere to adjust the angular dependence. The whole design was determined on the basis of MCNP calculations so as to have response characteristics that would generally match both the energy and angular dependencies of H p (10). This new instrument will be a great help in assessing the reference values of neutron H p (10) during field testing of personal neutron dosemeters in workplaces and also in interpreting their readings. (authors)

  20. BH3105 type neutron dose equivalent meter of high sensitivity

    International Nuclear Information System (INIS)

    Ji Changsong; Zhang Enshan; Yang Jianfeng; Zhang Hong; Huang Jiling

    1995-10-01

    It is noted that to design a neutron dose meter of high sensitivity is almost impossible in the frame of traditional designing principle--'absorption net principle'. Based on a newly proposed principle of obtaining neutron dose equi-biological effect adjustment--' absorption stick principle', a brand-new neutron dose-equivalent meter with high neutron sensitivity BH3105 has been developed. Its sensitivity reaches 10 cps/(μSv·h -1 ), which is 18∼40 times higher than one of foreign products of the same kind and is 10 4 times higher than that of domestic FJ342 neutron rem-meter. BH3105 has a measurement range from 0.1μSv/h to 1 Sv/h which is 1 or 2 orders wider than that of the other's. It has the advanced properties of gamma-resistance, energy response, orientation, etc. (6 tabs., 5 figs.)

  1. Fast Neutron Dose Distribution in a Linac Radiotherapy Facility

    International Nuclear Information System (INIS)

    Al-Othmany, D.Sh.; Abdul-Majid, S.; Kadi, M.W.

    2011-01-01

    CR-39 plastic detectors were used for fast neutron dose mapping in the radiotherapy facility at King AbdulAziz University Hospital (KAUH). Detectors were calibrated using a 252 Cf neutron source and a neutron dosimeter. After exposure chemical etching was performed using 6N NaOH solution at 70 degree C. Tracks were counted using an optical microscope and the number of tracks/cm 2 was converted to a neutron dose. 15 track detectors were distributed inside and outside the therapy room and were left for 32 days. The average neutron doses were 142.3 mSv on the accelerator head, 28.5 mSv on inside walls, 1.4 mSv beyond the beam shield, and 1 mSv in the control room

  2. Range calculations using multigroup transport methods

    International Nuclear Information System (INIS)

    Hoffman, T.J.; Robinson, M.T.; Dodds, H.L. Jr.

    1979-01-01

    Several aspects of radiation damage effects in fusion reactor neutron and ion irradiation environments are amenable to treatment by transport theory methods. In this paper, multigroup transport techniques are developed for the calculation of particle range distributions. These techniques are illustrated by analysis of Au-196 atoms recoiling from (n,2n) reactions with gold. The results of these calculations agree very well with range calculations performed with the atomistic code MARLOWE. Although some detail of the atomistic model is lost in the multigroup transport calculations, the improved computational speed should prove useful in the solution of fusion material design problems

  3. Cosmic radiation dose in aircraft - a neutron track etch detector

    Energy Technology Data Exchange (ETDEWEB)

    Vukovic, B.; Radolic, V.; Miklavcic, I.; Poje, M.; Varga, M. [Department of Physics, University of Osijek, 31000 Osijek, P.O. Box 125, Gajev trg 6 (Croatia); Planinic, J. [Department of Physics, University of Osijek, 31000 Osijek, P.O. Box 125, Gajev trg 6 (Croatia)], E-mail: planinic@ffos.hr

    2007-12-15

    Cosmic radiation bombards us at high altitude by ionizing particles. The radiation environment is a complex mixture of charged particles of solar and galactic origin, as well as of secondary particles produced in interaction of the galactic cosmic particles with the nuclei of atmosphere of the Earth. The radiation field at aircraft altitude consists of different types of particles, mainly photons, electrons, positrons and neutrons, with a large energy range. The non-neutron component of cosmic radiation dose aboard ATR 42 and A 320 aircrafts (flight level of 8 and 11 km, respectively) was measured with TLD-100 (LiF:Mg,Ti) detectors and the Mini 6100 semiconductor dosimeter. The estimated occupational effective dose for the aircraft crew (A 320) working 500 h per year was 1.64 mSv. Other experiments, or dose rate measurements with the neutron dosimeter, consisting of LR-115 track detector and boron foil BN-1 or 10B converter, were performed on five intercontinental flights. Comparison of the dose rates of the non-neutron component (low LET) and the neutron one (high LET) of the radiation field at the aircraft flight level showed that the neutron component carried about 50% of the total dose. The dose rate measurements on the flights from the Middle Europe to the South and Middle America, then to Korea and Japan, showed that the flights over or near the equator region carried less dose rate; this was in accordance with the known geomagnetic latitude effect.

  4. Cosmic radiation dose in aircraft - a neutron track etch detector

    International Nuclear Information System (INIS)

    Vukovic, B.; Radolic, V.; Miklavcic, I.; Poje, M.; Varga, M.; Planinic, J.

    2007-01-01

    Cosmic radiation bombards us at high altitude by ionizing particles. The radiation environment is a complex mixture of charged particles of solar and galactic origin, as well as of secondary particles produced in interaction of the galactic cosmic particles with the nuclei of atmosphere of the Earth. The radiation field at aircraft altitude consists of different types of particles, mainly photons, electrons, positrons and neutrons, with a large energy range. The non-neutron component of cosmic radiation dose aboard ATR 42 and A 320 aircrafts (flight level of 8 and 11 km, respectively) was measured with TLD-100 (LiF:Mg,Ti) detectors and the Mini 6100 semiconductor dosimeter. The estimated occupational effective dose for the aircraft crew (A 320) working 500 h per year was 1.64 mSv. Other experiments, or dose rate measurements with the neutron dosimeter, consisting of LR-115 track detector and boron foil BN-1 or 10B converter, were performed on five intercontinental flights. Comparison of the dose rates of the non-neutron component (low LET) and the neutron one (high LET) of the radiation field at the aircraft flight level showed that the neutron component carried about 50% of the total dose. The dose rate measurements on the flights from the Middle Europe to the South and Middle America, then to Korea and Japan, showed that the flights over or near the equator region carried less dose rate; this was in accordance with the known geomagnetic latitude effect

  5. Fast neutron radiation inactivation of Bacillus subtilis: Absorbed dose determination

    International Nuclear Information System (INIS)

    Song Lingli; Zheng Chun; Ai Zihui; Li Junjie; Dai Shaofeng

    2011-01-01

    In this paper, fast neutron inactivation effects of Bacillus subtilis were investigated with fission fast neutrons from CFBR-II reactor of INPC (Institute of Nuclear Physics and Chemistry) and mono-energetic neutrons from the Van de Graaff accelerator at Peking University. The method for determining the absorbed dose in the Bacillus subtilis suspension contained in test tubes is introduced. The absorbed dose, on account of its dependence on the volume and the form of confined state, was determined by combined experiments and Monte Carlo method. Using the calculation results of absorbed dose, the fast neutron inactivation effects on Bacillus subtilis were studied. The survival rates and absorbed dose curve was constructed. (authors)

  6. Neutron dose and energy spectra measurements at Savannah River Plant

    International Nuclear Information System (INIS)

    Brackenbush, L.W.; Soldat, K.L.; Haggard, D.L.; Faust, L.G.; Tomeraasen, P.L.

    1987-08-01

    Because some workers have a high potential for significant neutron exposure, the Savannah River Plant (SRP) contracted with Pacific Northwest Laboratory (PNL) to verify the accuracy of neutron dosimetry at the plant. Energy spectrum and neutron dose measurements were made at the SRP calibrations laboratory and at several other locations. The energy spectra measurements were made using multisphere or Bonner sphere spectrometers, 3 He spectrometers, and NE-213 liquid scintillator spectrometers. Neutron dose equivalent determinations were made using these instruments and others specifically designed to determine dose equivalent, such as the tissue equivalent proportional counter (TEPC). Survey instruments, such as the Eberline PNR-4, and the thermoluminescent dosimeter (TLD)-albedo and track etch dosimeters (TEDs) were also used. The TEPC, subjectively judged to provide the most accurate estimation of true dose equivalent, was used as the reference for comparison with other devices. 29 refs., 43 figs., 13 tabs

  7. Alanine and TLD coupled detectors for fast neutron dose measurements in neutron capture therapy (NCT)

    Energy Technology Data Exchange (ETDEWEB)

    Cecilia, A.; Baccaro, S.; Cemmi, A. [ENEA-FIS-ION, Casaccia RC, Via Anguillarese 301, 00060 Santa Maria di Galeria, Rome (Italy); Colli, V.; Gambarini, G. [Dept. of Physics of the Univ., INFN, Via Celoria 16, 20133 Milan (Italy); Rosi, G. [ENEA-FIS-ION, Casaccia RC, Via Anguillarese 301, 00060 Santa Maria di Galeria, Rome (Italy); Scolari, L. [Dept. of Physics of the Univ., INFN, Via Celoria 16, 20133 Milan (Italy)

    2004-07-01

    A method was investigated to measure gamma and fast neutron doses in phantoms exposed to an epithermal neutron beam designed for neutron capture therapy (NCT). The gamma dose component was measured by TLD-300 [CaF{sub 2}:Tm] and the fast neutron dose, mainly due to elastic scattering with hydrogen nuclei, was measured by alanine dosemeters [CH{sub 3}CH(NH{sub 2})COOH]. The gamma and fast neutron doses deposited in alanine dosemeters are very near to those released in tissue, because of the alanine tissue equivalence. Couples of TLD-300 and alanine dosemeters were irradiated in phantoms positioned in the epithermal column of the Tapiro reactor (ENEA-Casaccia RC). The dosemeter response depends on the linear energy transfer (LET) of radiation, hence the precision and reliability of the fast neutron dose values obtained with the proposed method have been investigated. Results showed that the combination of alanine and TLD detectors is a promising method to separate gamma dose and fast neutron dose in NCT. (authors)

  8. Thermal neutron equivalent doses assessment around KFUPM neutron source storage area using NTDs

    Energy Technology Data Exchange (ETDEWEB)

    Abu-Jarad, F.; Fazal-ur-Rehman; Al-Haddad, M.N.; Al-Jarrallah, M.I.; Nassar, R

    2002-07-01

    Area passive neutron dosemeters based on nuclear track detectors (NTDs) have been used for 13 days to assess accumulated low doses of thermal neutrons around neutron source storage area of the King Fahd University of Petroleum and Minerals (KFUPM). Moreover, the aim of this study is to check the effectiveness of shielding of the storage area. NTDs were mounted with the boron converter on their surface as one compressed unit. The converter is a lithium tetraborate (Li{sub 2}B{sub 4}O{sub 7}) layer for thermal neutron detection via {sup 10}B(N,{alpha}){sup 7}Li and {sup 6}Li(n,{alpha}){sup 3}H nuclear reactions. The area passive dosemeters were installed on 26 different locations around the source storage area and adjacent rooms. The calibration factor for NTD-based area passive neutron dosemeters was found to be 8.3 alpha tracks.cm{sup -2}.{mu}Sv{sup -1} using active snoopy neutron dosemeters in the KFUPM neutron irradiation facility. The results show the variation of accumulated dose with locations around the storage area. The range of dose rates varied from as low as 40 nSv.h{sup -1} up to 11 {mu}Sv.h{sup -1}. The study indicates that the area passive neutron dosemeter was able to detect accumulated doses as low as 40 nSv.h{sup -1}, which could not be detected with the available active neutron dosemeters. The results of the study also indicate that an additional shielding is required to bring the dose rates down to background level. The present investigation suggests extending this study to find the contribution of doses from fast neutrons around the neutron source storage area using NTDs through proton recoil. The significance of this passive technique is that it is highly sensitive and does not require any electronics or power supplies, as is the case in active systems. (author)

  9. CHARTB multigroup transport package

    International Nuclear Information System (INIS)

    Baker, L.

    1979-03-01

    The physics and numerical implementation of the radiation transport routine used in the CHARTB MHD code are discussed. It is a one-dimensional (Cartesian, cylindrical, and spherical symmetry), multigroup,, diffusion approximation. Tests and applications will be discussed as well

  10. Low doses of neutrons induce changes in gene expression

    International Nuclear Information System (INIS)

    Woloschak, G.E.; Chang-Liu, C.M.; Panozzo, J.; Libertin, C.R.

    1993-01-01

    Studies were designed to identify genes induced following low-dose neutron but not following γ-ray exposure in fibroblasts. Our past work had shown differences in the expression of β-protein kinase C and c-fos genes, both being induced following γ-ray but not neutron exposure. We have identified two genes that are induced following neutron, but not γ-ray, exposure: Rp-8 (a gene induced by apoptosis) and the long terminal repeat (LTR) of the human immunodeficiency (HIV). Rp-8 mRNA induction was demonstrated in Syrian hamster embryo fibroblasts and was found to be induced in cells exposed to neutrons administered at low (0.5 cGy/min) and at high dose rate (12 cGy/min). The induction of transcription from the LTR of HIV was demonstrated in HeLa cells bearing a transfected construct of the chloramphenicol acetyl transferase (CAT) gene driven by the HIV-LTR promoter. Measures of CAT activity and CAT transcripts following irradiation demonstrated an unresponsiveness to γ rays over a broad range of doses. Twofold induction of the HIV-LTR was detected following neutron exposure (48 cGy) administered at low (0.5 cGy/min) but not high (12 cGy/min) dose rates. Ultraviolet-mediated HIV-LTR induction was inhibited by low-dose-rate neutron exposure

  11. Estimated neutron dose to embryo and foetus during commercial flight

    International Nuclear Information System (INIS)

    Chen, J.; Lewis, B. J.; Bennett, L. G. I.; Green, A. R.; Tracy, B. L.

    2005-01-01

    A study has been carried out to assess the radiation exposure from cosmic-ray neutrons to the embryo and foetus of pregnant aircrew and air travellers in consideration of the radiation exposure from cosmic-ray neutrons to the embryo and foetus. A Monte Carlo analysis was performed to determine the equivalent dose from neutrons to the brain and body of an embryo at 8 weeks and to the foetus at the 3, 6 and 9 month periods. Neutron fluence-to-absorbed dose conversion coefficients for the foetal brain and for the entire foetal body (isotropic irradiation geometry) have been determined at the four developmental stages. The equivalent dose rate to the foetus during commercial flights has been further evaluated considering the fluence-to-absorbed dose conversion coefficients, a neutron spectrum measured at an altitude of 11.3 km and an ICRP-92 radiation-weighting factor for neutrons. This study indicates that the foetus can exceed the annual dose limit of 1 mSv for the general public after, for example, 15 round trips on commercial trans-Atlantic flights. (authors)

  12. Rapid Measurement of Neutron Dose Rate for Transport Index

    International Nuclear Information System (INIS)

    Morris, R.L.

    2000-01-01

    A newly available neutron dose equivalent remmeter with improved sensitivity and energy response has been put into service at Rocky Flats Environmental Technology Site (RFETS). This instrument is being used to expedite measurement of the Transport Index and as an ALARA tool to identify locations where slightly elevated neutron dose equivalent rates exist. The meter is capable of measuring dose rates as low as 0.2 μSv per hour (20 μrem per hour). Tests of the angular response and energy response of the instrument are reported. Calculations of the theoretical instrument response made using MCNPtrademark are reported for materials typical of those being shipped

  13. Fast neutron flux and intracranial dose distribution at a neutron irradiation facility

    International Nuclear Information System (INIS)

    Matsumoto, Tetsuo; Aizawa, Otohiko; Nozaki, Tetsuya

    1981-01-01

    A head phantom filled with water was used to measure the fast neutron flux using 115 In(n, n')sup(115m)In and 103 Rh(n, n')sup(103m)Rh reactions. γ-ray from sup(115m)In and x-ray from sup(103m)Rh were detected by a Ge(Li) and a Na(Tl)I counter, respectively. TLD was used to investigate the γ-dose rate distribution inside the phantom. Flux of fast neutron inside the phantom was about 1 x 10 6 n/cm 2 sec, which was 3 order smaller than that of thermal neutron. The fast neutron flux decreased to 1/10 at 15 cm depth, and γ-dose rate was about 200 R/h at 100 kW inside the phantom. Total dose at the surface was 350 rad/h, to which, fast neutrons contributed more than γ-rays. The rate of fast neutron dose was about 10% of thermal neutron's in Kerma dose unit (rad), however, the rate was highly dependent on RBE value. (Nakanishi, T.)

  14. Study on the dose distribution of the mixed field with thermal and epi-thermal neutrons for neutron capture therapy

    International Nuclear Information System (INIS)

    Kobayashi, Tooru; Sakurai, Yoshinori; Kanda, Keiji

    1994-01-01

    Simulation calculations using DOT 3.5 were carried out in order to confirm the characteristics of depth-dependent dose distribution in water phantom dependent on incident neutron energy. The epithermal neutrons mixed to thermal neutron field is effective improving the thermal neutron depth-dose distribution for neutron capture therapy. A feasibility study on the neutron energy spectrum shifter was performed using ANISN-JR for the KUR Heavy Water Facility. The design of the neutron spectrum shifter is feasible, without reducing the performance as a thermal neutron irradiation field. (author)

  15. Transportable, Low-Dose Active Fast-Neutron Imaging

    Energy Technology Data Exchange (ETDEWEB)

    Mihalczo, John T. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Wright, Michael C. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); McConchie, Seth M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Archer, Daniel E. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Palles, Blake A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2017-08-01

    This document contains a description of the method of transportable, low-dose active fast-neutron imaging as developed by ORNL. The discussion begins with the technique and instrumentation and continues with the image reconstruction and analysis. The analysis discussion includes an example of how a gap smaller than the neutron production spot size and detector size can be detected and characterized depending upon the measurement time.

  16. Experimental method research on neutron equal dose-equivalent detection

    International Nuclear Information System (INIS)

    Ji Changsong

    1995-10-01

    The design principles of neutron dose-equivalent meter for neutron biological equi-effect detection are studied. Two traditional principles 'absorption net principle' and 'multi-detector principle' are discussed, and on the basis of which a new theoretical principle for neutron biological equi-effect detection--'absorption stick principle' has been put forward to place high hope on both increasing neutron sensitivity of this type of meters and overcoming the shortages of the two traditional methods. In accordance with this new principle a brand-new model of neutron dose-equivalent meter BH3105 has been developed. Its neutron sensitivity reaches 10 cps/(μSv·h -1 ), 18∼40 times higher than that of all the same kinds of meters 0.23∼0.56 cps/(μSv·h -1 ), available today at home and abroad and the specifications of the newly developed meter reach or surpass the levels of the same kind of meters. Therefore the new theoretical principle of neutron biological equi-effect detection--'absorption stick principle' is proved to be scientific, advanced and useful by experiments. (3 refs., 3 figs., 2 tabs.)

  17. Fast neutron dose equivalent rates in heavy ion target areas

    International Nuclear Information System (INIS)

    Fulmer, C.B.; Butler, H.M.; Ohnesorge, W.F.; Mosko, S.W.

    1978-01-01

    At heavy ion accelerators, personnel access to areas near the target is sometimes important for successful performance of experiments. Radiation levels determine the amount of time that can be spent in these areas without exceeding maximum permissible exposures. Inasmuch as the fast neutrons contribute the major part of the Rem dose rates in these areas, knowledge of the fast neutron levels is important for planning permissive entry to target areas. Fast neutron dose rates were measured near thick medium mass targets bombarded with beams of C, N, O, and Ne ions. beam energies ranged from 3 to 16 MeV/amu. Dose rates (mrem/h) 1 meter from the target 90 degrees from the beam direction range from approx. 0.05 at MeV/amu to approx. 50 at 16 MeV/amu. These data should be helpful in planning permissive entry to heavy ion target areas

  18. Fast neutron dose equivalent rates in heavy ion target areas

    Energy Technology Data Exchange (ETDEWEB)

    Fulmer, C.B.; Butler, H.M.; Ohnesorge, W.F.; Mosko, S.W.

    1978-01-01

    At heavy ion accelerators, personnel access to areas near the target is sometimes important for successful performance of experiments. Radiation levels determine the amount of time that can be spent in these areas without exceeding maximum permissible exposures. Inasmuch as the fast neutrons contribute the major part of the Rem dose rates in these areas, knowledge of the fast neutron levels is important for planning permissive entry to target areas. Fast neutron dose rates were measured near thick medium mass targets bombarded with beams of C, N, O, and Ne ions. beam energies ranged from 3 to 16 MeV/amu. Dose rates (mrem/h) 1 meter from the target 90 degrees from the beam direction range from approx. 0.05 at MeV/amu to approx. 50 at 16 MeV/amu. These data should be helpful in planning permissive entry to heavy ion target areas.

  19. Estimation dose of secondary neutrons in proton therapy

    International Nuclear Information System (INIS)

    Urban, T.

    2014-01-01

    Most of proton therapy centers for cancer treatment are still based on the passive scattering, in some of them there is system of the active scanning installed as well. The aim of this study is to compare secondary neutron doses in and around target volumes in proton therapy for both treatment techniques and for different energies and profile of incident proton beam. The proton induced neutrons have been simulated in the very simple geometry of tissue equivalent phantom (imitate the patient) and scattering and scanning nozzle, respectively. In simulations of the scattering nozzle, different types of scattering filters and brass collimators have been used as well. 3D map of neutron doses in and around the chosen/potential target volume in the phantom/patient have been evaluated and compared in the context of the dose deposited in the target volume. Finally, the simulation results have been compared with published data. (author)

  20. Wide-range neutron dose determination with CR-39

    International Nuclear Information System (INIS)

    Arneja, A.R.; Waker, A.J.

    1995-01-01

    Optical density measurements of CR-30 irradiated with 252 Cf neutrons and chemically etched with 6.5 N KOH solution have been used to determine neutron absorbed doses between 0.1 and 10 Gy. Optimum etching conditions will depend upon the absorbed dose. Since it is not always possible to know the range of absorbed dose on a CR-39 dosemeter collected from personnel and area monitor stations in a criticality accident situation, a three-step two-hour chemical etch at 60 o C has been found to be appropriate. If after a total of six hours of chemical etching the optical density is found to be below 0.04 for 500 nm light (transmission > 90%) then further treatment in the form of electrochemical etching can be carried out to determine the lower absorbed dose. In this manner, absorbed doses below 0.1 Gy can be determined by counting tracks over a unit area. (author)

  1. Personnel neutron dose assessment upgrade: Volume 2, Field neutron spectrometer for health physics applications

    International Nuclear Information System (INIS)

    Brackenbush, L.W.; Reece, W.D.; Miller, S.D.

    1988-07-01

    Both the (ICRP) and the (NCPR) have recommended an increase in neutron quality factors and the adoption of effective dose equivalent methods. The series of reports entitled Personnel Neutron Dose Assessment Upgrade (PNL-6620) addresses these changes. Volume 1 in this series of reports (Personnel Neutron Dosimetry Assessment) provided guidance on the characteristics, use, and calibration of personnel neutron dosimeters in order to meet the new recommendations. This report, Volume 2: Field Neutron Spectrometer for Health Physics Applications describes the development of a portable field spectrometer which can be set up for use in a few minutes by a single person. The field spectrometer described herein represents a significant advance in improving the accuracy of neutron dose assessment. It permits an immediate analysis of the energy spectral distribution associated with the radiation from which neutron quality factor can be determined. It is now possible to depart from the use of maximum Q by determining and realistically applying a lower Q based on spectral data. The field spectrometer is made up of two modules: a detector module with built-in electronics and an analysis module with a IBM PC/reg sign/-compatible computer to control the data acquisition and analysis of data in the field. The unit is simple enough to allow the operator to perform spectral measurements with minimal training. The instrument is intended for use in steady-state radiation fields with neutrons energies covering the fission spectrum range. The prototype field spectrometer has been field tested in plutonium processing facilities, and has been proven to operate satisfactorily. The prototype field spectrometer uses a 3 He proportional counter to measure the neutron energy spectrum between 50 keV and 5 MeV and a tissue equivalent proportional counter (TEPC) to measure absorbed neutron dose

  2. Age-dependent conversion coefficients for organ doses and effective doses for external neutron irradiation

    International Nuclear Information System (INIS)

    Nishizaki, Chihiro; Endo, Akira; Takahashi, Fumiaki

    2006-06-01

    To utilize dose assessment of the public for external neutron irradiation, conversion coefficients of absorbed doses of organs and effective doses were calculated using the numerical simulation technique for six different ages (adult, 15, 10, 5 and 1 years and newborn), which represent the member of the public. Calculations were performed using six age-specific anthropomorphic phantoms and a Monte Carlo radiation transport code for two irradiation geometries, anterior-posterior and rotational geometries, for 20 incident energies from thermal to 20 MeV. Effective doses defined by the 1990 Recommendation of ICRP were calculated from the absorbed doses in 21 organs. The calculated results were tabulated in the form of absorbed doses and effective doses per unit neutron fluence. The calculated conversion coefficients are used for dose assessment of the public around nuclear facilities and accelerator facilities. (author)

  3. Low-dose neutron dose response of zebrafish embryos obtained from the Neutron exposure Accelerator System for Biological Effect Experiments (NASBEE) facility

    International Nuclear Information System (INIS)

    Ng, C.Y.P.; Kong, E.Y.; Konishi, T.; Kobayashi, A.; Suya, N.; Cheng, S.H.; Yu, K.N.

    2015-01-01

    The dose response of embryos of the zebrafish, Danio rerio, irradiated at 5 h post fertilization (hpf) by 2-MeV neutrons with ≤100 mGy was determined. The neutron irradiations were made at the Neutron exposure Accelerator System for Biological Effect Experiments (NASBEE) facility in the National Institute of Radiological Sciences (NIRS), Chiba, Japan. A total of 10 neutron doses ranging from 0.6 to 100 mGy were employed (with a gamma-ray contribution of 14% to the total dose), and the biological effects were studied through quantification of apoptosis at 25 hpf. The responses for neutron doses of 10, 20, 25, and 50 mGy approximately fitted on a straight line, while those for neutron doses of 0.6, 1 and 2.5 mGy exhibited neutron hormetic effects. As such, hormetic responses were generically developed by different kinds of ionizing radiations with different linear energy transfer (LET) values. The responses for neutron doses of 70 and 100 mGy were significantly below the lower 95% confidence band of the best-fit line, which strongly suggested the presence of gamma-ray hormesis. - Highlights: • Neutron dose response was determined for embryos of the zebrafish, Danio rerio. • Neutron doses of 0.6, 1 and 2.5 mGy led to neutron hormetic effects. • Neutron doses of 70 and 100 mGy accompanied by gamma rays led to gamma-ray hormesis

  4. AMPX-77: A modular code system for generating coupled multigroup neutron-gamma cross-section libraries from ENDF/B-IV and/or ENDF/B-V

    Energy Technology Data Exchange (ETDEWEB)

    Greene, N.M.; Ford, W.E. III; Petrie, L.M.; Arwood, J.W.

    1992-10-01

    AMPX-77 is a modular system of computer programs that pertain to nuclear analyses, with a primary emphasis on tasks associated with the production and use of multigroup cross sections. AH basic cross-section data are to be input in the formats used by the Evaluated Nuclear Data Files (ENDF/B), and output can be obtained in a variety of formats, including its own internal and very general formats, along with a variety of other useful formats used by major transport, diffusion theory, and Monte Carlo codes. Processing is provided for both neutron and gamma-my data. The present release contains codes all written in the FORTRAN-77 dialect of FORTRAN and wig process ENDF/B-V and earlier evaluations, though major modules are being upgraded in order to process ENDF/B-VI and will be released when a complete collection of usable routines is available.

  5. AMPX-77: A modular code system for generating coupled multigroup neutron-gamma cross-section libraries from ENDF/B-IV and/or ENDF/B-V

    International Nuclear Information System (INIS)

    Greene, N.M.; Ford, W.E. III; Petrie, L.M.; Arwood, J.W.

    1992-10-01

    AMPX-77 is a modular system of computer programs that pertain to nuclear analyses, with a primary emphasis on tasks associated with the production and use of multigroup cross sections. AH basic cross-section data are to be input in the formats used by the Evaluated Nuclear Data Files (ENDF/B), and output can be obtained in a variety of formats, including its own internal and very general formats, along with a variety of other useful formats used by major transport, diffusion theory, and Monte Carlo codes. Processing is provided for both neutron and gamma-my data. The present release contains codes all written in the FORTRAN-77 dialect of FORTRAN and wig process ENDF/B-V and earlier evaluations, though major modules are being upgraded in order to process ENDF/B-VI and will be released when a complete collection of usable routines is available

  6. Seed irradiation with continuously increasing doses of thermal neutrons

    International Nuclear Information System (INIS)

    Uhlik, J.; Pfeifer, M.; Pittermann, P.

    1977-01-01

    In the 'Raman' pea cv. the biological activity of thermal neutrons was investigated after irradiation of a 780 mm column of seeds for 3000 and 4167 seconds with a flux of 5.607 x 10 9 n.cm -2 per second. For different fractions of the seed column the average density of the neutron flux was calculated. It was proved that for the described method of seed irradiation it was sufficient to determine only the dose approaching the lethal dose. If a sufficiently high column of seeds is used part of the column of seeds will be irradiated with the optimum range of doses. The advantages of the suggested method of irradiation are not only smaller time and technological requirements resulting from the need for the determination of only the critical lethal dose of radiation by means of inhibition tests performed with seedlings, but also a simpler irradiation procedure. The suggested method of irradiation is at least nine times cheaper. (author)

  7. Occupational dose due to neutrons in medical linear accelerators

    International Nuclear Information System (INIS)

    Larcher, Ana M.; Bonet Duran, Stella M.; Lerner, Ana M.

    2000-01-01

    This paper describes a semi-empirical method to calculate the occupational dose due to neutrons and capture gamma rays in medical linear accelerators. It compares theoretical dose values with measurements performed in several 15 MeV medical accelerators installed in the country. Good agreement has been found between calculations made using the model and dose measurements, except for those accelerator rooms in which the maze length was shorter than the postulated tenth value distance. For those cases the model seems to overestimate neutron dose. The results demonstrate that the semi-empirical model is a good tool for quick and conservative shielding calculations for radiation protection purposes. Nevertheless, it is necessary to continue with the measurements in order to perform a more accurate validation of the model. (author)

  8. Verification of an effective dose equivalent model for neutrons

    International Nuclear Information System (INIS)

    Tanner, J.E.; Piper, R.K.; Leonowich, J.A.; Faust, L.G.

    1991-10-01

    Since the effective dose equivalent, based on the weighted sum of organ dose equivalents, is not a directly measurable quantity, it must be estimated with the assistance of computer modeling techniques and a knowledge of the radiation field. Although extreme accuracy is not necessary for radiation protection purposes, a few well-chosen measurements are required to confirm the theoretical models. Neutron measurements were performed in a RANDO phantom using thermoluminescent dosemeters, track etch dosemeters, and a 1/2-in. (1.27-cm) tissue equivalent proportional counter in order to estimate neutron doses and dose equivalents within the phantom at specific locations. The phantom was exposed to bare and D 2 O-moderated 252 Cf neutrons at the Pacific Northwest Laboratory's Low Scatter Facility. The Monte Carlo code MCNP with the MIRD-V mathematical phantom was used to model the human body and calculate organ doses and dose equivalents. The experimental methods are described and the results of the measurements are compared to the calculations. 8 refs., 3 figs., 3 tabs

  9. Measurements and calculations of neutron spectra and neutron dose distribution in human phantoms

    International Nuclear Information System (INIS)

    Palfalvi, J.

    1984-11-01

    The measurement and calculation of the radiation field around and in a phantom, with regard to the neutron component and the contaminating gamma radiation, are essential for radiation protection and radiotherapy purposes. The final report includes the development of the simple detector system, automized detector measuring facilities and a computerized evaluating system. The results of the depth dose and neutron spectra experiments and calculations in a human phantom are given

  10. Dose Calibration of the ISS-RAD Fast Neutron Detector

    Science.gov (United States)

    Zeitlin, C.

    2015-01-01

    The ISS-RAD instrument has been fabricated by Southwest Research Institute and delivered to NASA for flight to the ISS in late 2015 or early 2016. ISS-RAD is essentially two instruments that share a common interface to ISS. The two instruments are the Charged Particle Detector (CPD), which is very similar to the MSL-RAD detector on Mars, and the Fast Neutron Detector (FND), which is a boron-loaded plastic scintillator with readout optimized for the 0.5 to 10 MeV energy range. As the FND is completely new, it has been necessary to develop methodology to allow it to be used to measure the neutron dose and dose equivalent. This talk will focus on the methods developed and their implementation using calibration data obtained in quasi-monoenergetic (QMN) neutron fields at the PTB facility in Braunschweig, Germany. The QMN data allow us to determine an approximate response function, from which we estimate dose and dose equivalent contributions per detected neutron as a function of the pulse height. We refer to these as the "pSv per count" curves for dose equivalent and the "pGy per count" curves for dose. The FND is required to provide a dose equivalent measurement with an accuracy of ?10% of the known value in a calibrated AmBe field. Four variants of the analysis method were developed, corresponding to two different approximations of the pSv per count curve, and two different implementations, one for real-time analysis onboard ISS and one for ground analysis. We will show that the preferred method, when applied in either real-time or ground analysis, yields good accuracy for the AmBe field. We find that the real-time algorithm is more susceptible to chance-coincidence background than is the algorithm used in ground analysis, so that the best estimates will come from the latter.

  11. Neutron spectrometry and determination of neutron ambient dose equivalents in different LINAC radiotherapy rooms

    International Nuclear Information System (INIS)

    Domingo, C.; Garcia-Fuste, M.J.; Morales, E.; Amgarou, K.; Terron, J.A.; Rosello, J.; Brualla, L.; Nunez, L.; Colmenares, R.; Gomez, F.; Hartmann, G.H.; Sanchez-Doblado, F.; Fernandez, F.

    2010-01-01

    A project has been set up to study the effect on a radiotherapy patient of the neutrons produced around the LINAC accelerator head by photonuclear reactions induced by photons above ∼8 MeV. These neutrons may reach directly the patient, or they may interact with the surrounding materials until they become thermalised, scattering all over the treatment room and affecting the patient as well, contributing to peripheral dose. Spectrometry was performed with a calibrated and validated set of Bonner spheres at a point located at 50 cm from the isocenter, as well as at the place where a digital device for measuring neutrons, based on the upset of SRAM memories induced by thermal neutrons, is located inside the treatment room. Exposures have taken place in six LINAC accelerators with different energies (from 15 to 23 MV) with the aim of relating the spectrometer measurements with the readings of the digital device under various exposure and room geometry conditions. The final purpose of the project is to be able to relate, under any given treatment condition and room geometry, the readings of this digital device to patient neutron effective dose and peripheral dose in organs of interest. This would allow inferring the probability of developing second malignancies as a consequence of the treatment. Results indicate that unit neutron fluence spectra at 50 cm from the isocenter do not depend on accelerator characteristics, while spectra at the place of the digital device are strongly influenced by the treatment room geometry.

  12. An Experiment of Robust Parallel Algorithm for the Eigenvalue problem of a Multigroup Neutron Diffusion based on modified FETI-DP

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Jonghwa [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    Parallelization of Monte Carlo simulation is widely adpoted. There are also several parallel algorithms developed for the SN transport theory using the parallel wave sweeping algorithm and for the CPM using parallel ray tracing. For practical purpose of reactor physics application, the thermal feedback and burnup effects on the multigroup cross section should be considered. In this respect, the domain decomposition method(DDM) is suitable for distributing the expensive cross section calculation work. Parallel transport code and diffusion code based on the Raviart-Thomas mixed finite element method was developed. However most of the developed methods rely on the heuristic convergence of flux and current at the domain interfaces. Convergence was not attained in some cases. Mechanical stress computation community has also work on the DDM to solve the stress-strain equation using the finite element methods. The most successful domain decomposition method in terms of robustness is FETI-DP. We have modified the original FETI-DP to solve the eigenvalue problem for the multigroup diffusion problem in this study.

  13. Neutron organ dose and the influence of adipose tissue

    Science.gov (United States)

    Simpkins, Robert Wayne

    Neutron fluence to dose conversion coefficients have been assessed considering the influences of human adipose tissue. Monte Carlo code MCNP4C was used to simulate broad parallel beam monoenergetic neutrons ranging in energy from thermal to 10 MeV. Simulated Irradiations were conducted for standard irradiation geometries. The targets were on gender specific mathematical anthropomorphic phantoms modified to approximate human adipose tissue distributions. Dosimetric analysis compared adipose tissue influence against reference anthropomorphic phantom characteristics. Adipose Male and Post-Menopausal Female Phantoms were derived introducing interstitial adipose tissue to account for 22 and 27 kg additional body mass, respectively, each demonstrating a Body Mass Index (BMI) of 30. An Adipose Female Phantom was derived introducing specific subcutaneous adipose tissue accounting for 15 kg of additional body mass demonstrating a BMI of 26. Neutron dose was shielded in the superficial tissues; giving rise to secondary photons which dominated the effective dose for Incident energies less than 100 keV. Adipose tissue impact on the effective dose was a 25% reduction at the anterior-posterior incidence ranging to a 10% increase at the lateral incidences. Organ dose impacts were more distinctive; symmetrically situated organs demonstrated a 15% reduction at the anterior-posterior Incidence ranging to a 2% increase at the lateral incidences. Abdominal or asymmetrically situated organs demonstrated a 50% reduction at the anterior-posterior incidence ranging to a 25% increase at the lateral incidences.

  14. Scaling neutron absorbed dose distributions from one medium to another

    International Nuclear Information System (INIS)

    Awschalom, M.; Rosenberg, I.; Ten Haken, R.K.

    1982-11-01

    Central axis depth dose (CADD) and off-axis absorbed dose ratio (OAR) measurements were made in water, muscle and whole skeletal bone TE-solutions, mineral oil and glycerin with a clinical neutron therapy beam. These measurements show that, for a given neutron beam quality and field size, there is a universal CADD distribution at infinity if the depth in the phantom is expressed in terms of appropriate scaling lengths. These are essentially the kerma-weighted neutron mean free paths in the media. The method used in ICRU No. 26 to scale the CADD by the ratio of the densities is shown to give incorrect results. the OAR's measured in different media at depths proportional to the respective mean free paths were also found to be independent of the media to a good approximation. It is recommended that relative CADD and OAR measurements be performed in water because of its universality and convenience. A table of calculated scaling lengths is given for various neutron energy spectra and for various tissues and materials of practical importance in neutron dosimetry

  15. Neutron activation analysis for calibration of phosphorus implantation dose

    International Nuclear Information System (INIS)

    Paul, Rick L.; Simons, David S.

    2001-01-01

    A feasibility study was undertaken to determine if radiochemical neutron activation analysis (RNAA) can be used to certify the retained dose of phosphorus implanted in silicon, with the goal of producing a phosphorus SRM. Six pieces of silicon, implanted with a nominal phosphorus dose of 8.5x10 14 atoms·cm -2 were irradiated at a neutron flux of 1.05x10 14 cm -2 ·s -1 . The samples were mixed with carrier, dissolved in acid, the phosphorus isolated by chemical separation, and 32 P measured using a beta proportional counter. A mean phosphorus concentration of (8.35±0.20)x10 14 atoms·cm -2 (uncertainty=1 standard deviation) was determined for the six samples, in agreement with the nominal implanted dose

  16. A 'hybrid' neutron area survey instrument for the determination of neutron dose quantities in the workplace

    International Nuclear Information System (INIS)

    Tanner, R.J.; Jenkins, R.; Lowe, T.; Silvie, J.; Joyce, M.J.; Winsby, A.; Molinos, C.

    2005-01-01

    Full text: Neutron survey instruments are used routinely to determine the dose rates in areas where persons may be occupationally exposed. With a few exceptions, these instruments generally use a proportional counter with a high thermal neutron response located in a moderating sphere of CH 2 . The moderating sphere in such designs contains a thermal neutron absorber to reduce the over-response to thermal and intermediate energy neutrons. However, the commercially available examples of such instruments tend to have strongly energy dependent ambient dose equivalent response characteristics. In particular, they often over-respond in the energy range between 1 eV and 10 keV. A prototype of a novel design has been produced that uses seven detectors located in a moderating sphere of CH 2 , six near the surface to detect thermal and epithermal neutrons, and one in the centre to detect fast neutrons. This has been characterized using a combination of MCNP modelling and measurements to produce an instrument that has improved energy dependence of response characteristics. Additionally, the use of seven detectors offers direction and field hardness information. The design and calibration of the instrument are described and its response in workplaces calculated. (author)

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

    International Nuclear Information System (INIS)

    Tsujimura, N.; Yoshida, T.; Takada, C.

    2011-01-01

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

  18. Neutron and photon dose assessment in Indus accelerator complex

    International Nuclear Information System (INIS)

    Verma, Dimple; Haridas Nair, G.; Bandopadhyay, Tapas; Tripathy, R.M.; Pal, Rupali; Bakshi, A.K.; Palani Selvam, T.; Datta, D.

    2016-02-01

    Indus Accelerator Complex (IAC) consists of 20 MeV Microtron, 450/550 MeV Booster, 450 MeV Indus-1 and 2.5 GeV Indus-2 storage rings. The radiation environment in Indus Accelerator Complex comprises of bremsstrahlung photons, electrons, positrons, photo neutrons and muons, out of which, bremsstrahlung photons are the major constituent of the prompt radiation. Major problem faced for on-line detection of neutrons is their severely pulsed nature. In the present study, measurement of neutron and photon dose rates in Indus Accelerator Complex was carried out using passive dosimeters such as CR-39 solid state nuclear track detector (SSNTD) and CaSO 4 :Dy Teflon disc, 6 LiF:Mg,Ti (TLD 600) and 7 LiF:Mg,Ti (TLD 700) based thermo luminescent (TL) detectors. The report describes the details of the measurement and discusses the results. (author)

  19. Verification of an effective dose equivalent model for neutrons

    International Nuclear Information System (INIS)

    Tanner, J.E.; Piper, R.K.; Leonowich, J.A.; Faust, L.G.

    1992-01-01

    Since the effective dose equivalent, based on the weighted sum of organ dose equivalents, is not a directly measurable quantity, it must be estimated with the assistance of computer modelling techniques and a knowledge of the incident radiation field. Although extreme accuracy is not necessary for radiation protection purposes, a few well chosen measurements are required to confirm the theoretical models. Neutron doses and dose equivalents were measured in a RANDO phantom at specific locations using thermoluminescence dosemeters, etched track dosemeters, and a 1.27 cm (1/2 in) tissue-equivalent proportional counter. The phantom was exposed to a bare and a D 2 O-moderated 252 Cf neutron source at the Pacific Northwest Laboratory's Low Scatter Facility. The Monte Carlo code MCNP with the MIRD-V mathematical phantom was used to model the human body and to calculate the organ doses and dose equivalents. The experimental methods are described and the results of the measurements are compared with the calculations. (author)

  20. Personnel neutron dose assessment upgrade: Volume 1, Personnel neutron dosimetry assessment: [Final report

    International Nuclear Information System (INIS)

    Hadlock, D.E.; Brackenbush, L.W.; Griffith, R.V.; Hankins, D.E.; Parkhurst, M.A.; Stroud, C.M.; Faust, L.G.; Vallario, E.J.

    1988-07-01

    This report provides guidance on the characteristics, use, and calibration criteria for personnel neutron dosimeters. The report is applicable for neutrons with energies ranging from thermal to less than 20 MeV. Background for general neutron dosimetry requirements is provided, as is relevant federal regulations and other standards. The characteristics of personnel neutron dosimeters are discussed, with particular attention paid to passive neutron dosimetry systems. Two of the systems discussed are used at DOE and DOE-contractor facilities (nuclear track emulsion and thermoluminescent-albedo) and another (the combination TLD/TED) was recently developed. Topics discussed in the field applications of these dosimeters include their theory of operation, their processing, readout, and interpretation, and their advantages and disadvantages for field use. The procedures required for occupational neutron dosimetry are discussed, including radiation monitoring and the wearing of dosimeters, their exchange periods, dose equivalent evaluations, and the documenting of neutron exposures. The coverage of dosimeter testing, maintenance, and calibration includes guidance on the selection of calibration sources, the effects of irradiation geometries, lower limits of detectability, fading, frequency of calibration, spectrometry, and quality control. 49 refs., 6 figs., 8 tabs

  1. Scaling neutron absorbed dose distributions from one medium to another

    International Nuclear Information System (INIS)

    Awschalom, M.; Rosenberg, I.; Ten Haken, R.K.

    1983-01-01

    Central axis depth dose (CADD) and off-axis absorbed dose ratio (OAR) measurements were made in water, muscle and whole skeletal bone tissue-equivalent (TE) solutions, mineral oil, and glycerin with a clinical neutron therapy beam. These measurements show that, for a given neutron beam quality and field size, there is a universal CADD distribution at infinity if the depth in the phantom is expressed in terms of appropriate scaling lengths. These are essentially the kerma-weighted neutron mean free paths in the media. The method used in ICRU Report No. 26 to scale the CADD by the ratio of the densities is shown to give incorrect results. The OARs measured in different media at depths proportional to the respective mean free paths were also found to be independent of the media to a good approximation. Therefore, neutron beam CADDs and OARs may be measured in either TE solution (USA practice) or water (European practice), and having determined the respective scaling lengths, all measurements may be scaled from one medium to any other. It is recommended that for general treatment planning purposes, scaling be made to TE muscle with a density of 1.04 g cm -3 , since this value represents muscle and other soft tissues better than TE solution of density 1.07 g cm -3 . For such a transformation, relative measurements made in water are found to require very small corrections. Hence, it is further recommended that relative CADD and OAR measurements be performed in water because of its universality and convenience. Finally, a table of calculated scaling lengths is given for various neutron energy spectra and for various tissues and materials of practical importance in neutron dosimetry

  2. Dose modification factors in boron neutron capture therapy

    Energy Technology Data Exchange (ETDEWEB)

    Allen, B.J. (Australian Nuclear Science and Technology Organization (ANSTO), Menai (Australia))

    1993-01-01

    The effective treatment depth and therapeutic ratio in boron neutron capture therapy (BNCT) depend on a number of macroscopic dose factors such as boron concentrations in the tumor, normal tissue and blood. However, the role of various microscopic dose modification factors can be of critical importance in the evaluation of normal tissue tolerance levels. An understanding of these factors is valuable in designing BNCT experiments and the selection of appropriate boron compounds. These factors are defined in this paper and applied to the case of brain tumors with particular attention to capillary endothelial cells and oligodendrocytes. (orig.).

  3. Skin Dose Equivalent Measurement from Neutron-Deficient Isotopes

    International Nuclear Information System (INIS)

    Hsu, Hsiao-Hua; Costigan, Steve A.; Romero, Leonard L.; Whicker, Jeffrey J.

    1997-12-01

    Neutron-deficient-isotopes decay via positron emission and/or electron capture often followed by x-ray, gamma-ray, and 0.511 MeV photons from positron annihilation. For cases of significant area and/or personnel contamination with these isotopes, determination of skin dose equivalent (SDE) is required by 10CFR835. For assessment of SDE, we evaluated the MICROSPEC-2(TM) system manufactured by Bubble Technology Industries of Canada which uses three different probes for dose measurement. We used two probes: (1) the X-probe which measures lower energy (4 - 120 keV) photon energy distributions and determines deep dose equivalent, SDE and dose equivalent to eyes, and (2) the B-probe which measures electron (positron) energy distributions, and determines skin dose equivalent. Also, the measured photon and beta spectra can be used to identify radioactive isotopes in the contaminated area. Measurements with several neutron-deficient sources showed that this system provided reasonably accurate SDE rate measurements when compared with calculated benchmark SDE rates with an average percent difference of 40%. Variations were expected because of differences between the assumed geometries used by MlCROSPEC-2 and the calculations when compared to the measurement conditions

  4. Multi-Group Covariance Data Generation from Continuous-Energy Monte Carlo Transport Calculations

    International Nuclear Information System (INIS)

    Lee, Dong Hyuk; Shim, Hyung Jin

    2015-01-01

    The sensitivity and uncertainty (S/U) methodology in deterministic tools has been utilized for quantifying uncertainties of nuclear design parameters induced by those of nuclear data. The S/U analyses which are based on multi-group cross sections can be conducted by an simple error propagation formula with the sensitivities of nuclear design parameters to multi-group cross sections and the covariance of multi-group cross section. The multi-group covariance data required for S/U analysis have been produced by nuclear data processing codes such as ERRORJ or PUFF from the covariance data in evaluated nuclear data files. However in the existing nuclear data processing codes, an asymptotic neutron flux energy spectrum, not the exact one, has been applied to the multi-group covariance generation since the flux spectrum is unknown before the neutron transport calculation. It can cause an inconsistency between the sensitivity profiles and the covariance data of multi-group cross section especially in resolved resonance energy region, because the sensitivities we usually use are resonance self-shielded while the multi-group cross sections produced from an asymptotic flux spectrum are infinitely-diluted. In order to calculate the multi-group covariance estimation in the ongoing MC simulation, mathematical derivations for converting the double integration equation into a single one by utilizing sampling method have been introduced along with the procedure of multi-group covariance tally

  5. Long distance elementary measurement of the radiation dose ratio produced by neutron activation

    International Nuclear Information System (INIS)

    Zhou Changgeng; Lou Benchao; Wu Chunlei; Hu Yonghong; Li Yan

    2009-04-01

    The working principle and the structure and performances of a long distance controllable individual radiation dose ratio instrument are described. The radiation dose ratio produced by neutron activation is elementarily measured by using this instrument in the neutron generator hall with high neutron yield. When neutron yield arrives to 2 x 10 11 s -1 , the radiation dose ratio produced by neutron activation is 99.9 μSv/h in 1 h after the generator being stopped. The radiation dose ratio is reduced to 24.4 μSv/h in 39 h after the generator being stopped. When neutron yield is 3.2 x 10 10 s -1 , the radiation dose ratio produced by neutron activation is 21.9 μSv/h in 36 min, after the generator being stopped. The measurement results may provide reference for physical experimenters and neutron generator operators. (authors)

  6. Evaluation of mixed energy neutron doses using TLD NG-67 type

    International Nuclear Information System (INIS)

    Akhadi, Mukhlis; Thoyib Thamrin, M; Usmiyati Dewi, K.

    2000-01-01

    A research has been carried out to develop dose evaluation method of mixed neutron source with its neutron doses can be classified to two groups, I.e neutron doses with energy ≥ 0.5 eV and thermal neutron doses with energy less than 0.5 e V consist of epithermal and fast neutron, but in this research they were classified as fast neutron. Development of this dose evaluation method was carried out by sensitivity (S) intercomparison of TLD-600 to fast neutron, mixed energy neutron of nuclear rectors, and thermal neutron. From the experiment it was obtained that the value of Sfast : Sreactor : Sthermal = 0.005 : 0.010 : 1. Calibration factor (CF) of TLD is defined as 1/S. from the sensitivity data it can be obtained that the value of Cffast : Cfreactor : Cfthermal = 200 :100 : 1. The value of Cfreactor can be applied for mixed energy neutron doses evaluation of TLD-600. Key word : dosemeter, neutron dose, calibration factor, fast neutron, thermal neutron, nuclear reactor

  7. Development of Real-Time Measurement of Effective Dose for High Dose Rate Neutron Fields

    International Nuclear Information System (INIS)

    Braby, L. A.; Reece, W. D.; Hsu, W. H.

    2003-01-01

    Studies of the effects of low doses of ionizing radiation require sources of radiation which are well characterized in terms of the dose and the quality of the radiation. One of the best measures of the quality of neutron irradiation is the dose mean lineal energy. At very low dose rates this can be determined by measuring individual energy deposition events, and calculating the dose mean of the event size. However, at the dose rates that are normally required for biology experiments, the individual events can not be separated by radiation detectors. However, the total energy deposited in a specified time interval can be measured. This total energy has a random variation which depends on the size of the individual events, so the dose mean lineal energy can be calculated from the variance of repeated measurements of the energy deposited in a fixed time. We have developed a specialized charge integration circuit for the measurement of the charge produced in a small ion chamber in typical neutron irradiation experiments. We have also developed 4.3 mm diameter ion chambers with both tissue equivalent and carbon walls for the purpose of measuring dose mean lineal energy due to all radiations and due to all radiations except neutrons, respectively. By adjusting the gas pressure in the ion chamber, it can be made to simulate tissue volumes from a few nanometers to a few millimeters in diameter. The charge is integrated for 0.1 seconds, and the resulting pulse height is recorded by a multi channel analyzer. The system has been used in a variety of photon and neutron radiation fields, and measured values of dose and dose mean lineal energy are consistent with values extrapolated from measurements made by other techniques at much lower dose rates. It is expected that this technique will prove to be much more reliable than extrapolations from measurements made at low dose rates because these low dose rate exposures generally do not accurately reproduce the attenuation and

  8. Dose inhomogeneities for photons and neutrons near interfaces

    International Nuclear Information System (INIS)

    Broerse, J. J.; Zoetelief, J.

    2004-01-01

    Perturbations of charged particle equilibrium (CPE) at interfaces of materials of different atomic composition can lead to considerable differences in the energy deposition by photons and neutrons. Specific examples of these interface perturbations are encountered during irradiation of body cavities and soft tissue adjacent to bone or metallic implants and irradiation of cells in monolayer on the bottom of culture dishes. Another example is the build-up of CPE at air-tissue interfaces, referred to in radiotherapy as the skin sparing effect. For photon irradiation excess production of secondary electrons in high-Z materials, such as glass, bone or gold, will induce appreciably higher doses and decreased cell survival compared to the equilibrium situation. The energy dissipation of fast neutrons in biological materials occurs through recoil protons, heavy recoil nuclei and products of nuclear reactions. Owing to the large contribution from recoil protons to the neutron kerma, the hydrogen content of the biological material mainly determines the energy deposition. For neutron irradiation of cells in monolayer, CPE can be established or deliberately avoided by mounting tissue-equivalent plastic or carbon discs in front of the cells, respectively. This approach makes it possible to distinguish the biological effects of the low- and high-LET radiation components. (authors)

  9. The development of BH3105E type neutron dose-equivalent meter

    International Nuclear Information System (INIS)

    Ji Changsong; Wang Tingting; Zhang Shuheng; Tan Baozeng

    2011-01-01

    A new BH3105E Type Neutron Dose-equivalent Meter has been developed. The 'multi-stick' ab- sorption method is used for thermal -14 MeV neutron equal dose-equivalent detection, what gives a high neutron sensitivity of 5 cps/μSv · h-1. RS-232 interface is accepted for signal communication (authors)

  10. On the convergence of multigroup discrete-ordinates approximations

    International Nuclear Information System (INIS)

    Victory, H.D. Jr.; Allen, E.J.; Ganguly, K.

    1987-01-01

    Our analysis is divided into two distinct parts which we label for convenience as Part A and Part B. In Part A, we demonstrate that the multigroup discrete-ordinates approximations are well-defined and converge to the exact transport solution in any subcritical setting. For the most part, we focus on transport in two-dimensional Cartesian geometry. A Nystroem technique is used to extend the discrete ordinates multigroup approximates to all values of the angular and energy variables. Such an extension enables us to employ collectively compact operator theory to deduce stability and convergence of the approximates. In Part B, we perform a thorough convergence analysis for the multigroup discrete-ordinates method for an anisotropically-scattering subcritical medium in slab geometry. The diamond-difference and step-characteristic spatial approximation methods are each studied. The multigroup neutron fluxes are shown to converge in a Banach space setting under realistic smoothness conditions on the solution. This is the first thorough convergence analysis for the fully-discretized multigroup neutron transport equations

  11. On the calculation of multi-group fission spectrum vectors

    International Nuclear Information System (INIS)

    Mueller, E.Z.

    1984-05-01

    In this report, the problem of calculating fission spectrum vectors in a consistent manner is formulated. The practical implications of using fission spectrum vectors in multi-group transport calculations are also addressed. The significance of the weighting spectra used for the calculation of fission spectrum vectors is illustrated for the case of a simple neutronic assembly

  12. FINELM: a multigroup finite element diffusion code. Part II

    International Nuclear Information System (INIS)

    Davierwalla, D.M.

    1981-05-01

    The author presents the axisymmetric case in cylindrical coordinates for the finite element multigroup neutron diffusion code, FINELM. The numerical acceleration schemes incorporated viz. the Lebedev extrapolations and the coarse mesh rebalancing, space collapsing, are discussed. A few benchmark computations are presented as validation of the code. (Auth.)

  13. Calculation of multigroup reaction rates for the Ghana Research ...

    African Journals Online (AJOL)

    The discrete ordinate spatial model, which pro-vides solution to the differential form of the transport equation by the Carlson-SN (N=4) approach was adopted to solve the Ludwig-Boltzmann multigroup neutron transport equation for this analysis. The results show that for any fissile resonance absorber, the reaction rates ...

  14. Development of Real-Time Measurement of Effective Dose for High Dose Rate Neutron Fields

    CERN Document Server

    Braby, L A; Reece, W D

    2003-01-01

    Studies of the effects of low doses of ionizing radiation require sources of radiation which are well characterized in terms of the dose and the quality of the radiation. One of the best measures of the quality of neutron irradiation is the dose mean lineal energy. At very low dose rates this can be determined by measuring individual energy deposition events, and calculating the dose mean of the event size. However, at the dose rates that are normally required for biology experiments, the individual events can not be separated by radiation detectors. However, the total energy deposited in a specified time interval can be measured. This total energy has a random variation which depends on the size of the individual events, so the dose mean lineal energy can be calculated from the variance of repeated measurements of the energy deposited in a fixed time. We have developed a specialized charge integration circuit for the measurement of the charge produced in a small ion chamber in typical neutron irradiation exp...

  15. Effects of secondary interactions on the dose calculation in treatments with Boron Neutron Capture Therapy (BNCT)

    International Nuclear Information System (INIS)

    Monteiro, E.

    2004-01-01

    The aimed of this work consists of evaluating the influence of the secondary contributions of dose (thermal neutrons dose, epithermal neutrons dose, fast neutrons dose and photon dose) in treatment planning with BNCT. MCNP4B Code was used to calculate RBE-Gy doses through the irradiation of the modified Snyder head head phantom.A reduction of the therapeutical gain of monoenergetic neutron beans was observed in non invasive treatments, provoked for the predominance of the fast neutron dose component in the skin, showing that the secondary contributions of dose can contribute more in the direction to raise the dose in the fabric healthy that in the tumor, thus reducing the treatment efficiency. (author)

  16. Influence of Neutron Spectra Unfolding Method on Fast Neutron Dose Determination

    International Nuclear Information System (INIS)

    Marinkovic, P.

    1991-01-01

    Full text: Accuracy of knowing the fast neutron spectra has great influence on equivalent dose determination. In usual fast neutron spectrum measurements with scintillation detectors based on proton recoil, the main difficulty is confidence of unfolding method. In former ones variance of obtained result is usually great and negative values are possible too, which does means that we don't now exactly is obtained neutron spectrum real one. The new unfolding method based on Shanon's information theory, which gives non-negative spectrum and relative low variance, is obtained and appropriate numerical code for application in fast neutron spectrometry based on proton recoil is realized. In this method principle of maximum entropy and maximum likelihood are used together. Unknown group density distribution functions, which are considered as desired normalized mean neutron group flux, are constl u cted using only constrain of knowing mean value. Obtained distributions are consistent to available information (counts in NCA from proton recoil), while being maximally noncommittal with respect to all other unknown circumstances. For maximum likelihood principle, distribution functions around mean value of counts in the channels of MCA are taken to be Gauss function shape. Optimal non-negative solution is searched by means of Lagrange parameter method. Nonlinear system of equations, is solved using gradient and Newton iterative algorithm. Error covariance matrix is obtained too. (author)

  17. Results of neutron dose measurements at the Rossendorf research reactors taking the actual neutron spectra into account

    International Nuclear Information System (INIS)

    Rimpler, A.; Kneschke, H.

    1985-01-01

    Based on a systematic evaluation of area dose studies at the beginning of the seventies, no individual routine neutron monitoring has been performed at the Rossendorf research reactors. To check this decision, a limited number of persons has been monitored with solid-state nuclear track detectors for several years. The dosemeters were calibrated on the basis of neutron spectra determined at the working places by means of the Bonner sphere method. Intermediate neutrons with a 1/E/sup α/ Fermi distribution were dominating. The fraction of fast neutrons was practically negligible. The obtained spectra, radiation, field quantities and results of individual dose measurements are presented. The dosemeter most appropriate for such neutron fields would be a 12-inch Bonner sphere rem counter. As the mean annual neutron exposure of research workers at the reactor amounted to only 2% of the maximum permissible dose, individual routine monitoring will, also in the future, not be neccessary. (author)

  18. The transport of neutrons and gamma-rays in the air

    International Nuclear Information System (INIS)

    Adamski, J.

    1980-01-01

    The transport of neutrons and gamma rays in the infinite homogeneous air has been investigated. For the calculations has been used the Multigroup One Dimensional Discrete Ordinates Transport Code ANISN-W. The calculations have been performed for three types of neutron sources. The neutrons and gamma ray doses in the air have been analyzed, and comparison to the other authors' results has been given. (author)

  19. Multi-group diffusion perturbation calculation code. PERKY (2002)

    Energy Technology Data Exchange (ETDEWEB)

    Iijima, Susumu; Okajima, Shigeaki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2002-12-01

    Perturbation calculation code based on the diffusion theory ''PERKY'' is designed for nuclear characteristic analyses of fast reactor. The code calculates reactivity worth on the multi-group diffusion perturbation theory in two or three dimensional core model and kinetics parameters such as effective delayed neutron fraction, prompt neutron lifetime and absolute reactivity scale factor ({rho}{sub 0} {delta}k/k) for FCA experiments. (author)

  20. Review of multigroup nuclear cross-section processing

    Energy Technology Data Exchange (ETDEWEB)

    Trubey, D.K.; Hendrickson, H.R. (comps.)

    1978-10-01

    These proceedings consist of 18 papers given at a seminar--workshop on ''Multigroup Nuclear Cross-Section Processing'' held at Oak Ridge, Tennessee, March 14--16, 1978. The papers describe various computer code systems and computing algorithms for producing multigroup neutron and gamma-ray cross sections from evaluated data, and experience with several reference data libraries. Separate abstracts were prepared for 13 of the papers. The remaining five have already been cited in ERA, and may be located by referring to the entry CONF-780334-- in the Report Number Index. (RWR)

  1. Cation disorder in high-dose, neutron-irradiated spinel

    International Nuclear Information System (INIS)

    Sickafus, K.E.; Larson, A.C.; Yu, N.; Nastasi, M.; Hollenberg, G.W.; Garner, F.A.; Bradt, R.C.

    1994-08-01

    The objective of this effort is to determine whether MgAl 2 O 4 spinel is a suitable ceramic for fusion applications. Here, the crystal structures of MgAl 2 O 4 spinel single crystals irradiated to high neutron fluences [>5·10 26 n/m 2 (E n > 0.1 MeV)] were examined by neutron diffraction. Crystal structure refinement of the highest dose sample indicated that the average scattering strength of the tetrahedral crystal sites decreased by ∼ 20% while increasing by ∼ 8% on octahedral sites. Since the neutron scattering length for Mg is considerably larger than for Al, this results is consistent with site exchange between Mg 2+ ions on tetrahedral sites and Al 3+ ions on octahedral sites. Least-squares refinements also indicated that, in all irradiated samples, at least 35% of Mg 2+ and Al 3+ ions in the crystal experienced disordering replacements. This retained dpa on the cation sublattices is the largest retained damage ever measured in an irradiated spinel material

  2. Dose Evaluation of Neutron within Containment Building of a CE type Nuclear Power Plant

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Tae Wook; Han, Jae Mun; Kim, Kyung Doek; Yun, Cheol Whan; Suh, Jang Soo; Kim, Young Jae [Nuclear Environment Technology Institute, Daejeon (Korea, Republic of)

    2005-03-15

    From measured results of the neutron fields at some principal places within the containment building in a CE type nuclear power plant in operation, the radiation exposure of a worker to the neutron at there was evaluated and the equivalent dose reflecting new recommendation (ICRP 60) was compared with that doing the old one (ICRP 26). The measured neutron field was also compared with calibration neutron field. From the analysis, the following conclusion was obtained: the average neutron radiation weighting factor according to new recommendation is 2.41 to 2.71 times higher than the old one. The average neutron radiation weighting factor at the measured place was similar to that at calibration neutron field. The average neutron energy at measured place was between 42 and 158 keV and higher than that of calibration field of 500 keV. So, the measured equivalent dose in nuclear power plant could be overestimated compared to the real equivalent dose.

  3. Assessment of doses due to secondary neutrons received by patient treated by proton therapy

    International Nuclear Information System (INIS)

    Sayah, R.; Martinetti, F.; Donadille, L.; Clairand, I.; Delacroix, S.; De Oliveira, A.; Herault, J.

    2010-01-01

    Proton therapy is a specific technique of radiotherapy which aims at destroying cancerous cells by irradiating them with a proton beam. Nuclear reactions in the device and in the patient himself induce secondary radiations involving mainly neutrons which contribute to an additional dose for the patient. The author reports a study aimed at the assessment of these doses due to secondary neutrons in the case of ophthalmological and intra-cranial treatments. He presents a Monte Carlo simulation of the room and of the apparatus, reports the experimental validation of the model (dose deposited by protons in a water phantom, ambient dose equivalent due to neutrons in the treatment room, absorbed dose due to secondary particles in an anthropomorphic phantom), and the assessment with a mathematical phantom of doses dues to secondary neutrons received by organs during an ophthalmological treatment. He finally evokes current works of calculation of doses due to secondary neutrons in the case of intra-cranial treatments

  4. The Suppression of Energy Discretization Errors in Multigroup Transport Calculations

    International Nuclear Information System (INIS)

    Larsen, Edward

    2013-01-01

    The Objective of this project is to develop, implement, and test new deterministric methods to solve, as efficiently as possible, multigroup neutron transport problems having an extremely large number of groups. Our approach was to (i) use the standard CMFD method to 'coarsen' the space-angle grid, yielding a multigroup diffusion equation, and (ii) use a new multigrid-in-space-and-energy technique to efficiently solve the multigroup diffusion problem. The overall strategy of (i) how to coarsen the spatial an energy grids, and (ii) how to navigate through the various grids, has the goal of minimizing the overall computational effort. This approach yields not only the fine-grid solution, but also coarse-group flux-weighted cross sections that can be used for other related problems.

  5. The neutron dose equivalent around high energy medical electron linear accelerators

    Directory of Open Access Journals (Sweden)

    Poje Marina

    2014-01-01

    Full Text Available The measurement of neutron dose equivalent was made in four dual energy linear accelerator rooms. Two of the rooms were reconstructed after decommissioning of 60Co units, so the main limitation was the space. The measurements were performed by a nuclear track etched detectors LR-115 associated with the converter (radiator that consist of 10B and with the active neutron detector Thermo BIOREM FHT 742. The detectors were set at several locations to evaluate the neutron ambient dose equivalent and/or neutron dose rate to which medical personnel could be exposed. Also, the neutron dose dependence on collimator aperture was analyzed. The obtained neutron dose rates outside the accelerator rooms were several times smaller than the neutron dose rates inside the accelerator rooms. Nevertheless, the measured neutron dose equivalent was not negligible from the aspect of the personal dosimetry with almost 2 mSv a year per person in the areas occupied by staff (conservative estimation. In rooms with 15 MV accelerators, the neutron exposure to the personnel was significantly lower than in the rooms having 18 MV accelerators installed. It was even more pronounced in the room reconstructed after the 60Co decommissioning. This study confirms that shielding from the neutron radiation should be considered when building vaults for high energy linear accelerators, especially when the space constraints exist.

  6. Characteristics of neutron irradiation facility and dose estimation method for neutron capture therapy at Kyoto University research reactor institute

    International Nuclear Information System (INIS)

    Kobayashi, T.; Sakurai, Y.; Kanda, K.

    2001-01-01

    The neutron irradiation characteristics of the Heavy Water Neutron Irradiation Facility (HWNIF) at the Kyoto University Research Reactor Institute (KIJRRI) for boron neutron capture therapy (BNCT), is described. The present method of dose measurement and its evaluation at the KURRI, is explained. Especially, the special feature and noticeable matters were expounded for the BNCT with craniotomy, which has been applied at present only in Japan. (author)

  7. Out‐of‐field doses and neutron dose equivalents for electron beams from modern Varian and Elekta linear accelerators

    Science.gov (United States)

    Cardenas, Carlos E.; Nitsch, Paige L.; Kudchadker, Rajat J.; Howell, Rebecca M.

    2016-01-01

    Out‐of‐field doses from radiotherapy can cause harmful side effects or eventually lead to secondary cancers. Scattered doses outside the applicator field, neutron source strength values, and neutron dose equivalents have not been broadly investigated for high‐energy electron beams. To better understand the extent of these exposures, we measured out‐of‐field dose characteristics of electron applicators for high‐energy electron beams on two Varian 21iXs, a Varian TrueBeam, and an Elekta Versa HD operating at various energy levels. Out‐of‐field dose profiles and percent depth‐dose curves were measured in a Wellhofer water phantom using a Farmer ion chamber. Neutron dose was assessed using a combination of moderator buckets and gold activation foils placed on the treatment couch at various locations in the patient plane on both the Varian 21iX and Elekta Versa HD linear accelerators. Our findings showed that out‐of‐field electron doses were highest for the highest electron energies. These doses typically decreased with increasing distance from the field edge but showed substantial increases over some distance ranges. The Elekta linear accelerator had higher electron out‐of‐field doses than the Varian units examined, and the Elekta dose profiles exhibited a second dose peak about 20 to 30 cm from central‐axis, which was found to be higher than typical out‐of‐field doses from photon beams. Electron doses decreased sharply with depth before becoming nearly constant; the dose was found to decrease to a depth of approximately E(MeV)/4 in cm. With respect to neutron dosimetry, Q values and neutron dose equivalents increased with electron beam energy. Neutron contamination from electron beams was found to be much lower than that from photon beams. Even though the neutron dose equivalent for electron beams represented a small portion of neutron doses observed under photon beams, neutron doses from electron beams may need to be considered for

  8. Out-of-field doses and neutron dose equivalents for electron beams from modern Varian and Elekta linear accelerators.

    Science.gov (United States)

    Cardenas, Carlos E; Nitsch, Paige L; Kudchadker, Rajat J; Howell, Rebecca M; Kry, Stephen F

    2016-07-08

    Out-of-field doses from radiotherapy can cause harmful side effects or eventually lead to secondary cancers. Scattered doses outside the applicator field, neutron source strength values, and neutron dose equivalents have not been broadly investigated for high-energy electron beams. To better understand the extent of these exposures, we measured out-of-field dose characteristics of electron applicators for high-energy electron beams on two Varian 21iXs, a Varian TrueBeam, and an Elekta Versa HD operating at various energy levels. Out-of-field dose profiles and percent depth-dose curves were measured in a Wellhofer water phantom using a Farmer ion chamber. Neutron dose was assessed using a combination of moderator buckets and gold activation foils placed on the treatment couch at various locations in the patient plane on both the Varian 21iX and Elekta Versa HD linear accelerators. Our findings showed that out-of-field electron doses were highest for the highest electron energies. These doses typically decreased with increasing distance from the field edge but showed substantial increases over some distance ranges. The Elekta linear accelerator had higher electron out-of-field doses than the Varian units examined, and the Elekta dose profiles exhibited a second dose peak about 20 to 30 cm from central-axis, which was found to be higher than typical out-of-field doses from photon beams. Electron doses decreased sharply with depth before becoming nearly constant; the dose was found to decrease to a depth of approximately E(MeV)/4 in cm. With respect to neutron dosimetry, Q values and neutron dose equivalents increased with electron beam energy. Neutron contamination from electron beams was found to be much lower than that from photon beams. Even though the neutron dose equivalent for electron beams represented a small portion of neutron doses observed under photon beams, neutron doses from electron beams may need to be considered for special cases.

  9. NDS multigroup cross section libraries

    International Nuclear Information System (INIS)

    DayDay, N.

    1981-12-01

    A summary description and documentation of the multigroup cross section libraries which exist at the IAEA Nuclear Data Section are given in this report. The libraries listed are available either on tape or in printed form. (author)

  10. Time-Dependent Neutron and Photon Dose-Field Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Wooten, Hasani Omar [Georgia Inst. of Technology, Atlanta, GA (United States)

    2005-08-01

    A unique tool is developed that allows the user to model physical representations of complicated glovebox facilities in two dimensions and determine neutral-particle flux and ambient dose-equivalent fields throughout that geometry. The Pandemonium code, originally designed to determine flux and dose-rates only, is improved to include realistic glovebox geometries, time-dependent source and detector positions, time-dependent shielding thickness calculations, time-integrated doses, a representative criticality accident scenario based on time-dependent reactor kinetics, and more rigorous photon treatment. A primary benefit of this work has been an extensive analysis and improvement of the photon model that is not limited to the application described in this thesis. The photon model has been extended in energy range to 10 MeV to include photons from fission and new photon buildup factors have been included that account for the effects of photon buildup at slant-path thicknesses as a function of angle, where the mean free path thickness has been preserved. The overall system of codes is user-friendly and it is directly applicable to facilities such as the plutonium facility at Los Alamos National Laboratory, where high-intensity neutron and photon emitters are regularly used. The codes may be used to determine a priori doses for given work scenarios in an effort to supply dose information to process models which will in turn assist decision makers on ensuring as low as reasonably achievable (ALARA) compliance. In addition, coupling the computational results of these tools with the process model visualization tools will help to increase worker safety and radiological safety awareness.

  11. GANTRAS - a system of codes for the solution of the multigroup transport equation with a rigorous treatment of anisotropic neutron scattering

    International Nuclear Information System (INIS)

    Schwenk-Ferrero, A.

    1986-11-01

    GANTRAS is a system of codes for neutron transport calculations in which the anisotropy of elastic and inelastic (including (n,n'x)-reactions) scattering is fully taken into account. This is achieved by employing a rigorous method, so-called I * -method, to represent the scattering term of the transport equation and with the use of double-differential cross-sections for the description of the emission of secondary neutrons. The I * -method was incorporated into the conventional transport code ONETRAN. The ONETRAN subroutines were modified for the new purpose. An implementation of the updated version ANTRA1 was accomplished for plane and spherical geometry. ANTRA1 was included in GANTRAS and linked to another modules which prepare angle-dependent transfer matrices. The GANTRAS code consists of three modules: 1. The CROMIX code which calculates the macroscopic transfer matrices for mixtures on the base of microscopic nuclide-dependent data. 2. The ATP code which generates discretized angular transfer probabilities (i.e. discretizes the I * -function). 3. The ANTRA1 code to perform S N transport calculations in one-dimensional plane and spherical geometries. This structure of GANTRAS allows to accommodate the system to various transport problems. (orig.) [de

  12. Alterations in water and electrolyte absorption in the rat colon following neutron irradiation: influence of neutron component and irradiation dose.

    Science.gov (United States)

    Dublineau, I; Ksas, B; Joubert, C; Aigueperse, J; Gourmelon, P; Griffiths, N M

    2002-12-01

    To study the absorptive function of rat colon following whole-body exposure to neutron irradiation, either to the same total dose with varying proportion of neutrons or to the same neutron proportion with an increasing irradiation dose. Different proportions of neutron irradiation were produced from the reactor SILENE using a fissile solution of uranium nitrate (8, 47 and 87% neutron). Water and electrolyte fluxes were measured in the rat in vivo under anaesthesia by insertion into the descending colon of an agarose gel cylinder simulating the faeces. Functional studies were completed by histological analyses. In the first set of experiments, rats received 3.8 Gy with various neutron percentages and were studied from 1 to 14 days after exposure. In the second set of experiments, rats were exposed to increasing doses of irradiation (1-4Gy) with a high neutron percentage (87%n) and were studied at 4 days after exposure. The absorptive capacity of rat colon was diminished by irradiation at 3-5 days, with a nadir at 4 days. The results demonstrate that an increase in the neutron proportion is associated with an amplification of the effects. Furthermore, a delay in the re-establishment of normal absorption was observed with the high neutron proportion (87%n). A dose-dependent reduction of water absorption by rat colon was also observed following neutron irradiation (87%n), with a 50% reduction at 3 Gy. Comparison of this dose-effect curve with the curve obtained following gamma (60)Co-irradiation indicates an RBE of 2.2 for absorptive colonic function in rat calculated at 4 days after exposure.

  13. High dose neutron irradiation damage in beryllium as blanket material

    Energy Technology Data Exchange (ETDEWEB)

    Chakin, V.P. E-mail: fae@niiar.ru; Kazakov, V.A.; Teykovtsev, A.A.; Pimenov, V.V.; Shimansky, G.A.; Ostrovsky, Z.E.; Suslov, D.N.; Latypov, R.N.; Belozerov, S.V.; Kupriyanov, I.B. E-mail: vniinm.400@g23.relkom.ru

    2001-11-01

    The paper presents the investigation results of beryllium products that operated in the SM and BOR-60 reactors up to neutron doses of 2.8x10{sup 22} and 8.0x10{sup 22} cm{sup -2} (E>1 MeV), respectively. The calculated and experimental data are given on helium and tritium accumulation, swelling, micro-hardness and thermal conductivity. The microstructural investigation results of irradiated beryllium are also presented. It is shown that the rate of helium and tritium accumulation in beryllium in the SM and BOR-60 reactors is high enough, which is of interest from the viewpoint of modeling the working conditions of the DEMO fusion reactor. Swelling of beryllium at irradiation temperature of 70-150 deg. C and neutron fluence of 2.8x10{sup 22} cm{sup -2} (E>1 MeV) makes up 0.8-1.5%, at 400 deg. C and fluence of 8x10{sup 22} cm{sup -2} (E>1 MeV)-3.2-5.0%. Irradiation hardening and decrease of thermal conductivity strongly depend on the irradiation temperature and are more significant at reduced temperatures. All results presented in the paper were analyzed with due account of the supposed working parameters of the DEMO fusion reactor blanket.

  14. High dose neutron irradiation damage in beryllium as blanket material

    International Nuclear Information System (INIS)

    Chakin, V.P.; Kazakov, V.A.; Teykovtsev, A.A.; Pimenov, V.V.; Shimansky, G.A.; Ostrovsky, Z.E.; Suslov, D.N.; Latypov, R.N.; Belozerov, S.V.; Kupriyanov, I.B.

    2001-01-01

    The paper presents the investigation results of beryllium products that operated in the SM and BOR-60 reactors up to neutron doses of 2.8x10 22 and 8.0x10 22 cm -2 (E>1 MeV), respectively. The calculated and experimental data are given on helium and tritium accumulation, swelling, micro-hardness and thermal conductivity. The microstructural investigation results of irradiated beryllium are also presented. It is shown that the rate of helium and tritium accumulation in beryllium in the SM and BOR-60 reactors is high enough, which is of interest from the viewpoint of modeling the working conditions of the DEMO fusion reactor. Swelling of beryllium at irradiation temperature of 70-150 deg. C and neutron fluence of 2.8x10 22 cm -2 (E>1 MeV) makes up 0.8-1.5%, at 400 deg. C and fluence of 8x10 22 cm -2 (E>1 MeV)-3.2-5.0%. Irradiation hardening and decrease of thermal conductivity strongly depend on the irradiation temperature and are more significant at reduced temperatures. All results presented in the paper were analyzed with due account of the supposed working parameters of the DEMO fusion reactor blanket

  15. High Fidelity Ion Beam Simulation of High Dose Neutron Irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Was, Gary; Wirth, Brian; Motta, Athur; Morgan, Dane; Kaoumi, Djamel; Hosemann, Peter; Odette, Robert

    2018-04-30

    Project Objective: The objective of this proposal is to demonstrate the capability to predict the evolution of microstructure and properties of structural materials in-reactor and at high doses, using ion irradiation as a surrogate for reactor irradiations. “Properties” includes both physical properties (irradiated microstructure) and the mechanical properties of the material. Demonstration of the capability to predict properties has two components. One is ion irradiation of a set of alloys to yield an irradiated microstructure and corresponding mechanical behavior that are substantially the same as results from neutron exposure in the appropriate reactor environment. Second is the capability to predict the irradiated microstructure and corresponding mechanical behavior on the basis of improved models, validated against both ion and reactor irradiations and verified against ion irradiations. Taken together, achievement of these objectives will yield an enhanced capability for simulating the behavior of materials in reactor irradiations

  16. Neutron and gamma dose and spectra measurements on the Little Boy replica

    International Nuclear Information System (INIS)

    Hoots, S.; Wadsworth, D.

    1984-01-01

    The radiation-measurement team of the Weapons Engineering Division at Lawrence Livermore National Laboratory (LLNL) measured neutron and gamma dose and spectra on the Little Boy replica at Los Alamos National Laboratory (LANL) in April 1983. This assembly is a replica of the gun-type atomic bomb exploded over Hiroshima in 1945. These measurements support the National Academy of Sciences Program to reassess the radiation doses due to atomic bomb explosions in Japan. Specifically, the following types of information were important: neutron spectra as a function of geometry, gamma to neutron dose ratios out to 1.5 km, and neutron attenuation in the atmosphere. We measured neutron and gamma dose/fission from close-in to a kilometer out, and neutron and gamma spectra at 90 and 30 0 close-in. This paper describes these measurements and the results. 12 references, 13 figures, 5 tables

  17. Preliminary characterization of the passive neutron dose equivalent monitor with TLDs

    Energy Technology Data Exchange (ETDEWEB)

    Tsujimura, Norio; Kanai, Katsuta; Momose, Takumaro; Hayashi, Naomi [Japan Nuclear Cycle Development Inst., Tokai Works, Tokai, Ibaraki (Japan); Chen Erhu [Beijing Institute of Nuclear Engineering, Beijing (China)

    2001-02-01

    The passive neutron dose equivalent monitor with TLDs is composed of a cubic polyethylene moderator and TLDs at the center of moderator. This monitor was originally designed for measurements of neutron doses over long-term period of time around the nuclear facilities. In this study, the energy response of this monitor was calculated by Monte Carlo methods and experimentally obtained under {sup 241}Am-Be, {sup 252}Cf and moderated {sup 252}Cf neutron irradiation. Additionally, the responses of two types of conventional neutron dose equivalent meters (rem counters) were also investigated as comparison. The authors concluded that this passive neutron monitor with TLDs had a good energy response similar to conventional rem counters and could evaluate neutron doses within 10% of accuracy to the moderated fission spectra. (author)

  18. Recursive solutions for multi-group neutron kinetics diffusion equations in homogeneous three-dimensional rectangular domains with time dependent perturbations

    Energy Technology Data Exchange (ETDEWEB)

    Petersen, Claudio Z. [Universidade Federal de Pelotas, Capao do Leao (Brazil). Programa de Pos Graduacao em Modelagem Matematica; Bodmann, Bardo E.J.; Vilhena, Marco T. [Universidade Federal do Rio Grande do Sul, Porto Alegre, RS (Brazil). Programa de Pos-graduacao em Engenharia Mecanica; Barros, Ricardo C. [Universidade do Estado do Rio de Janeiro, Nova Friburgo, RJ (Brazil). Inst. Politecnico

    2014-12-15

    In the present work we solve in analytical representation the three dimensional neutron kinetic diffusion problem in rectangular Cartesian geometry for homogeneous and bounded domains for any number of energy groups and precursor concentrations. The solution in analytical representation is constructed using a hierarchical procedure, i.e. the original problem is reduced to a problem previously solved by the authors making use of a combination of the spectral method and a recursive decomposition approach. Time dependent absorption cross sections of the thermal energy group are considered with step, ramp and Chebyshev polynomial variations. For these three cases, we present numerical results and discuss convergence properties and compare our results to those available in the literature.

  19. Neutron dose equivalent next to the target shield of a neutron therapy facility using an LET counter

    International Nuclear Information System (INIS)

    Stinchcomb, T.G.; Kuchnir, F.T.

    1981-01-01

    The use of a spherical tissue-equivalent proportional counter for measurements of the lineal energy (y) and derivations of the linear energy transfer (LET) for fast neutrons has the advantage of giving distributions of dose and dose equivalent as functions of either LET or y. A measurement next to the target shielding of the neutron therapy facility at the University of Chicago Hospitals and Clinics (UCHC) is described, and the data processing is outlined. The distributions are presented and compared to those from measurements in the neutron beam. The average quality factors are presented

  20. Measurement of stray neutron doses inside the treatment room from a proton pencil beam scanning system

    Czech Academy of Sciences Publication Activity Database

    Mojzeszek, N.; Farah, J.; Klodowska, M.; Ploc, Ondřej; Stolarczyk, L.; Waligorski, M. P. R.; Olko, P.

    2017-01-01

    Roč. 34, č. 2 (2017), s. 80-84 ISSN 1120-1797 Institutional support: RVO:61389005 Keywords : secondary neutrons * proton therapy * pencil beam scanning systtems * out-of-field doses * stray neutron doses * TEPC Subject RIV: FP - Other Medical Disciplines OBOR OECD: Radiology, nuclear medicine and medical imaging Impact factor: 1.990, year: 2016

  1. Genetic effects induced by neutrons in Drosophila melanogaster I. Determination of absorbed dose

    International Nuclear Information System (INIS)

    Delfin, A.; Paredes, L.C.; Zambrano, F.; Guzman-Rincon, J.; Urena-Nunez, F.

    2001-01-01

    A method to obtain the absorbed dose in Drosophila melanogaster irradiated in the thermal column facility of the Triga Mark III Reactor has been developed. The method is based on the measurements of neutron activation of gold foils produced by neutron capture to obtain the neutron fluxes. These fluxes, combined with the calculations of kinetic energy released per unit mass, enables one to obtain the absorbed doses in Drosophila melanogaster

  2. Neutron spectrum and dose-equivalent in shuttle flights during solar maximum

    Energy Technology Data Exchange (ETDEWEB)

    Keith, J E; Badhwar, G D; Lindstrom, D J [National Aeronautics and Space Administration, Houston, TX (United States). Lyndon B. Johnson Space Center

    1992-01-01

    This paper presents unambiguous measurements of the spectrum of neutrons found in spacecraft during spaceflight. The neutron spectrum was measured from thermal energies to about 10 MeV using a completely passive system of metal foils as neutron detectors. These foils were exposed to the neutron flux bare, covered by thermal neutron absorbers (Gd) and inside moderators (Bonner spheres). This set of detectors was flown on three U.S. Space Shuttle flights, STS-28, STS-36 and STS-31, during the solar maximum. We show that the measurements of the radioactivity of these foils lead to a differential neutron energy spectrum in all three flights that can be represented by a power law, J(E){approx equal}E{sup -0.765} neutrons cm{sup -2} day {sup -1} MeV{sup -1}. We also show that the measurements are even better represented by a linear combination of the terrestrial neutron albedo and a spectrum of neutrons locally produced in a aluminium by protons, computed by a previous author. We use both approximations to the neutron spectrum to produce a worst case and most probable case for the neutron spectra and the resulting dose-equivalents, computed using ICRP-51 neutron fluence-dose conversion tables. We compare these to the skin dose-equivalents due to charged particles during the same flights. (author).

  3. 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)

  4. Neutron dose to patients treated with high-energy medical accelerators

    International Nuclear Information System (INIS)

    McGinley, P.H.

    2001-01-01

    The neutron dose equivalent received by patients treated with high energy x-ray beams was measured in this research. A total of 13 different medical accelerators were evaluated in terms of the neutron dose equivalent in the patient plane and at the beam center. The neutron dose equivalent at the beam center was found to ranged from 0.02 to 9.4 mSv per Sv of x-ray dose and values from 0.029 to 2.58 mSv per Sv of x-ray were measured in the patient plane. It was concluded that the neutron levels meet the International Electrotechnical Commission standard for the patient plane. It was also concluded that when intensity modulated radiation treatment is conducted the neutron dose equivalent received by the patient will increase by a factor of 2 to 10. (author)

  5. Assessment of fast and thermal neutron ambient dose equivalents around the KFUPM neutron source storage area using nuclear track detectors

    Energy Technology Data Exchange (ETDEWEB)

    Fazal-ur-Rehman [Physics Department, King Fahd University of Petroleum and Minerals, Dhahran 31261 (Saudi Arabia)]. E-mail: fazalr@kfupm.edu.sa; Al-Jarallah, M.I. [Physics Department, King Fahd University of Petroleum and Minerals, Dhahran 31261 (Saudi Arabia); Abu-Jarad, F. [Radiation Protection Unit, Environmental Protection Department, Saudi Aramco, P. O. Box 13027, Dhahran 31311 (Saudi Arabia); Qureshi, M.A. [Center for Applied Physical Sciences, King Fahd University of Petroleum and Minerals, Dhahran 31261 (Saudi Arabia)

    2005-11-15

    A set of five {sup 241}Am-Be neutron sources are utilized in research and teaching at King Fahd University of Petroleum and Minerals (KFUPM). Three of these sources have an activity of 16Ci each and the other two are of 5Ci each. A well-shielded storage area was designed for these sources. The aim of the study is to check the effectiveness of shielding of the KFUPM neutron source storage area. Poly allyl diglycol carbonate (PADC) Nuclear track detectors (NTDs) based fast and thermal neutron area passive dosimeters have been utilized side by side for 33 days to assess accumulated low ambient dose equivalents of fast and thermal neutrons at 30 different locations around the source storage area and adjacent rooms. Fast neutron measurements have been carried out using bare NTDs, which register fast neutrons through recoils of protons, in the detector material. NTDs were mounted with lithium tetra borate (Li{sub 2}B{sub 4}O{sub 7}) converters on their surfaces for thermal neutron detection via B10(n,{alpha})Li6 and Li6(n,{alpha})H3 nuclear reactions. The calibration factors of NTD both for fast and thermal neutron area passive dosimeters were determined using thermoluminescent dosimeters (TLD) with and without a polyethylene moderator. The calibration factors for fast and thermal neutron area passive dosimeters were found to be 1.33 proton tracks cm{sup -2}{mu}Sv{sup -1} and 31.5 alpha tracks cm{sup -2}{mu}Sv{sup -1}, respectively. The results show variations of accumulated dose with the locations around the storage area. The fast neutron dose equivalents rates varied from as low as 182nSvh{sup -1} up to 10.4{mu}Svh{sup -1} whereas those for thermal neutron ranged from as low as 7nSvh{sup -1} up to 9.3{mu}Svh{sup -1}. The study indicates that the area passive neutron dosimeter was able to detect dose rates as low as 7 and 182nSvh{sup -1} from accumulated dose for thermal and fast neutrons, respectively, which were not possible to detect with the available active neutron

  6. Impact of the Revised 10 CFR 835 on the Neutron Dose Rates at LLNL

    International Nuclear Information System (INIS)

    Radev, R.

    2009-01-01

    In June 2007, 10 CFR 835 (1) was revised to include new radiation weighting factors for neutrons, updated dosimetric models, and dose terms consistent with the newer ICRP recommendations. A significant aspect of the revised 10 CFR 835 is the adoption of the recommendations outlined in ICRP-60 (2). The recommended new quantities demand a review of much of the basic data used in protection against exposure to sources of ionizing radiation. The International Commission on Radiation Units and Measurements has defined a number of quantities for use in personnel and area monitoring (3,4,5) including the ambient dose equivalent H*(d) to be used for area monitoring and instrument calibrations. These quantities are used in ICRP-60 and ICRP-74. This report deals only with the changes in the ambient dose equivalent and ambient dose rate equivalent for neutrons as a result of the implementation of the revised 10 CFR 835. In the report, the terms neutron dose and neutron dose rate will be used for convenience for ambient neutron dose and ambient neutron dose rate unless otherwise stated. This report provides a qualitative and quantitative estimate of how much the neutron dose rates at LLNL will change with the implementation of the revised 10 CFR 835. Neutron spectra and dose rates from selected locations at the LLNL were measured with a high resolution spectroscopic neutron dose rate system (ROSPEC) as well as with a standard neutron rem meter (a.k.a., a remball). The spectra obtained at these locations compare well with the spectra from the Radiation Calibration Laboratory's (RCL) bare californium source that is currently used to calibrate neutron dose rate instruments. The measurements obtained from the high resolution neutron spectrometer and dose meter ROSPEC and the NRD dose meter compare within the range of ±25%. When the new radiation weighting factors are adopted with the implementation of the revised 10 CFR 835, the measured dose rates will increase by up to 22%. The

  7. Implementation of an Analytical Model for Leakage Neutron Equivalent Dose in a Proton Radiotherapy Planning System

    Energy Technology Data Exchange (ETDEWEB)

    Eley, John [Department of Radiation Physics, The University of Texas MD Anderson Cancer Center, 1515 Holcombe Blvd., Houston, TX 77030 (United States); Graduate School of Biomedical Sciences, The University of Texas, 6767 Bertner Ave., Houston, TX 77030 (United States); Newhauser, Wayne, E-mail: newhauser@lsu.edu [Department of Physics and Astronomy, Louisiana State University and Agricultural and Mechanical College, 202 Nicholson Hall, Tower Drive, Baton Rouge, LA 70803 (United States); Mary Bird Perkins Cancer Center, 4950 Essen Lane, Baton Rouge, LA 70809 (United States); Homann, Kenneth; Howell, Rebecca [Department of Radiation Physics, The University of Texas MD Anderson Cancer Center, 1515 Holcombe Blvd., Houston, TX 77030 (United States); Graduate School of Biomedical Sciences, The University of Texas, 6767 Bertner Ave., Houston, TX 77030 (United States); Schneider, Christopher [Department of Physics and Astronomy, Louisiana State University and Agricultural and Mechanical College, 202 Nicholson Hall, Tower Drive, Baton Rouge, LA 70803 (United States); Mary Bird Perkins Cancer Center, 4950 Essen Lane, Baton Rouge, LA 70809 (United States); Durante, Marco; Bert, Christoph [GSI Helmholtzzentrum für Schwerionenforschung, Planckstr. 1, Darmstadt 64291 (Germany)

    2015-03-11

    Equivalent dose from neutrons produced during proton radiotherapy increases the predicted risk of radiogenic late effects. However, out-of-field neutron dose is not taken into account by commercial proton radiotherapy treatment planning systems. The purpose of this study was to demonstrate the feasibility of implementing an analytical model to calculate leakage neutron equivalent dose in a treatment planning system. Passive scattering proton treatment plans were created for a water phantom and for a patient. For both the phantom and patient, the neutron equivalent doses were small but non-negligible and extended far beyond the therapeutic field. The time required for neutron equivalent dose calculation was 1.6 times longer than that required for proton dose calculation, with a total calculation time of less than 1 h on one processor for both treatment plans. Our results demonstrate that it is feasible to predict neutron equivalent dose distributions using an analytical dose algorithm for individual patients with irregular surfaces and internal tissue heterogeneities. Eventually, personalized estimates of neutron equivalent dose to organs far from the treatment field may guide clinicians to create treatment plans that reduce the risk of late effects.

  8. SU-E-T-566: Neutron Dose Cloud Map for Compact ProteusONE Proton Therapy

    Energy Technology Data Exchange (ETDEWEB)

    Syh, J; Patel, B; Syh, J; Rosen, L; Wu, H [Willis-Knighton Medical Center, Shreveport, LA (United States)

    2015-06-15

    Purpose: To establish the base line of neutron cloud during patient treatment in our new compact Proteus One proton pencil beam scanning (PBS) system with various beam delivery gantry angles, with or without range shifter (RS) at different body sites. Pencil beam scanning is an emerging treatment technique, for the concerns of neutron exposure, this study is to evaluate the neutron dose equivalent per given delivered dose under various treatment conditions at our proton therapy center. Methods: A wide energy neutron dose equivalent detector (SWENDI-II, Thermo Scientific, MA) was used for neutron dose measurements. It was conducted in the proton therapy vault during beam was on. The measurement location was specifically marked in order to obtain the equivalent dose of neutron activities (H). The distances of 100, 150 and 200 cm at various locations are from the patient isocenter. The neutron dose was measured of proton energy layers, # of spots, maximal energy range, modulation width, field radius, gantry angle, snout position and delivered dose in CGE. The neutron dose cloud is reproducible and is useful for the future reference. Results: When distance increased the neutron equivalent dose (H) reading did not decrease rapidly with changes of proton energy range, modulation width or spot layers. For cranial cases, the average mSv/CGE was about 0.02 versus 0.032 for pelvis cases. RS will induce higher H to be 0.10 mSv/CGE in average. Conclusion: From this study, neutron per dose ratio (mSv/CGE) slightly depends upon various treatment parameters for pencil beams. For similar treatment conditions, our measurement demonstrates this value for pencil beam scanning beam has lowest than uniform scanning or passive scattering beam with a factor of 5. This factor will be monitored continuously for other upcoming treatment parameters in our facility.

  9. ANSL-V: ENDF/B-V based multigroup cross-section libraries for Advanced Neutron Source (ANS) reactor studies. Supplement 1

    Energy Technology Data Exchange (ETDEWEB)

    Wright, R.Q.; Renier, J.P.; Bucholz, J.A.

    1995-08-01

    The original ANSL-V cross-section libraries (ORNL-6618) were developed over a period of several years for the physics analysis of the ANS reactor, with little thought toward including the materials commonly needed for shielding applications. Materials commonly used for shielding applications include calcium barium, sulfur, phosphorous, and bismuth. These materials, as well as {sup 6}Li, {sup 7}Li, and the naturally occurring isotopes of hafnium, have been added to the ANSL-V libraries. The gamma-ray production and gamma-ray interaction cross sections were completely regenerated for the ANSL-V 99n/44g library which did not exist previously. The MALOCS module was used to collapse the 99n/44g coupled library to the 39n/44g broad- group library. COMET was used to renormalize the two-dimensional (2- D) neutron matrix sums to agree with the one-dimensional (1-D) averaged values. The FRESH module was used to adjust the thermal scattering matrices on the 99n/44g and 39n/44g ANSL-V libraries. PERFUME was used to correct the original XLACS Legendre polynomial fits to produce acceptable distributions. The final ANSL-V 99n/44g and 39n/44g cross-section libraries were both checked by running RADE. The AIM module was used to convert the master cross-section libraries from binary coded decimal to binary format (or vice versa).

  10. A Study on the Neutron Dose Distribution in Case of 10 MV X-rays Radiotherapy

    International Nuclear Information System (INIS)

    Park, Cheol Soo; Shin, Seong Soo; Lim, Cheong Hwan; Jung, Hong Ryang

    2008-01-01

    This study is to measure the radiation dose of neutrons generated by the particle accelerator during X-ray (photon) treatment with a neutron detection method by using CR-39, and to research how the generation of neutrons may incur problems associated with radiation doses for patient treatment when using high energy photons for cancer treatment as a clinical application. The findings are summarized as follows : The results showed that average 0.35 mSv was measured with exposure of 1 Gy photon in case of fast neutron, 0.65 mSv with exposure of 2 Gy photon, 1.82 mSv exposure of 5 Gy, 0.26 mSv with exposure of 1 Gy photon in case of thermal neutron, 0.56 mSv with exposure of 2 Gy photon, and 1.23 mSv with exposure of 5 Gy of photon. By measuring the occurrence of neutron by using Wedge Filter, it has been confirmed that the occurrence of neutrons increased when using Wedge Filter. The results also showed that more neutrons were detected over the existing experiments when using an SRS Cone requiring high doses of radiation. Total 2.85 mSv neutrons were found on the average with exposure of 5 Gy photon in case of fast neutron and 1.37 mSv neutrons were found on the average with exposure of 5 Gy photon in case of thermal neutron. During the general treatment, about 1.6 times more neutrons over 5 Gy photon were found in case of fast neutron and about 1.12 time more neutrons over 5 Gy photon were found in case of thermal neutron.

  11. Calculation of fast neutron dose in plastic-coated optical fibers

    International Nuclear Information System (INIS)

    Siebert, B.R.L.; Henschel, H.

    1998-01-01

    The dose of fast neutrons in optical fibers with hydrogen-containing coating materials is considerably increased by energetic recoil protons. Their contribution to the dose in a SiO 2 fiber core is calculated by the Monte Carlo method for different fiber geometries and a fiber optic cable. With 14 MeV neutrons the dose in a single fiber is increased by about 21%, whereas in fiber bundles the dose increase can reach about 170%. Maximum dose enhancement in fiber bundles (about 610%) occurs at neutron energies around 5.5 MeV. The dose increase caused by 14 MeV neutrons in the fiber of a typical laboratory cable is about 124%

  12. Neutron fluence-to-dose conversion coefficients for embryo and fetus

    International Nuclear Information System (INIS)

    Chen, J.; Meyerhof, D.; Vlahovich, S.

    2004-01-01

    A problem of concern in radiation protection is the exposure of pregnant women to ionising radiation, because of the high radiosensitivity of the embryo and fetus. External neutron exposure is of concern when pregnant women travel by aeroplane. Dose assessments for neutrons frequently rely on fluence-to-dose conversion coefficients. While neutron fluence-to-dose conversion coefficients for adults are recommended in International Commission on Radiological Protection publications and International Commission on Radiological Units and Measurements reports, conversion coefficients for embryos and fetuses are not given in the publications. This study undertakes Monte Carlo calculations to determine the mean absorbed doses to the embryo and fetus when the mother is exposed to neutron fields. A new set of mathematical models for the embryo and fetus has been developed at Health Canada and is used together with mathematical phantoms of a pregnant female developed at Oak Ridge National Laboratory. Monoenergetic neutrons from 1 eV to 10 MeV are considered in this study. The irradiation geometries include antero-posterior (AP), postero-anterior (PA), lateral (LAT), rotational (ROT) and isotropic (ISO) geometries. At each of these standard irradiation geometries, absorbed doses to the fetal brain and body are calculated; for the embryo at 8 weeks and the fetus at 3, 6 or 9 months. Neutron fluence-to-absorbed dose conversion coefficients are derived for the four age groups. Neutron fluence-to-equivalent dose conversion coefficients are given for the AP irradiations which yield the highest radiation dose to the fetal body in the neutron energy range considered here. The results indicate that for neutrons <10 MeV more protection should be given to pregnant women in the first trimester due to the higher absorbed dose per unit neutron fluence to the fetus. (authors)

  13. Neutron fluence-to-dose conversion coefficients for embryo and fetus.

    Science.gov (United States)

    Chen, Jing; Meyerhof, Dorothy; Vlahovich, Slavica

    2004-01-01

    A problem of concern in radiation protection is the exposure of pregnant women to ionising radiation, because of the high radiosensitivity of the embryo and fetus. External neutron exposure is of concern when pregnant women travel by aeroplane. Dose assessments for neutrons frequently rely on fluence-to-dose conversion coefficients. While neutron fluence-to-dose conversion coefficients for adults are recommended in International Commission on Radiological Protection publications and International Commission on Radiological Units and Measurements reports, conversion coefficients for embryos and fetuses are not given in the publications. This study undertakes Monte Carlo calculations to determine the mean absorbed doses to the embryo and fetus when the mother is exposed to neutron fields. A new set of mathematical models for the embryo and fetus has been developed at Health Canada and is used together with mathematical phantoms of a pregnant female developed at Oak Ridge National Laboratory. Monoenergetic neutrons from 1 eV to 10 MeV are considered in this study. The irradiation geometries include antero-posterior (AP), postero-anterior (PA), lateral (LAT), rotational (ROT) and isotropic (ISO) geometries. At each of these standard irradiation geometries, absorbed doses to the fetal brain and body are calculated; for the embryo at 8 weeks and the fetus at 3, 6 or 9 months. Neutron fluence-to-absorbed dose conversion coefficients are derived for the four age groups. Neutron fluence-to-equivalent dose conversion coefficients are given for the AP irradiations which yield the highest radiation dose to the fetal body in the neutron energy range considered here. The results indicate that for neutrons <10 MeV more protection should be given to pregnant women in the first trimester due to the higher absorbed dose per unit neutron fluence to the fetus.

  14. Development of slow neutron dose meter by track counting

    International Nuclear Information System (INIS)

    Nozaki, Tetsuya; Takeuchi, Hiroshi; Hasnel, S.; Honda, Teruyuki; Harasawa, Susumu.

    1993-01-01

    Prototypes of a plate type and two types of semispherical dosemeter were manufactured and their performances were tested. Polyallyl diglycol carbonate (PADC) was employed as neutron detector and covered with natural LiF and 6 Li-enriched LiF ceramics. They were irradiated in the TRIGA II Reactor of the Rikkyo University, and the sensitivity characteristics for incident angles and neutron energies were analyzed. It is concluded that it may be possible to manufacture small sized neutron dosemeter of flat response to wide energy range from thermal neutrons to epithermal neutrons with sufficient sensitivity for personal monitoring, using LiF ceramic as neutron filters and neutron converters. However the following issues are to be solved: the optimization of thickness of the LiF filter, the effects of albedo of the human body, and the applicability to the intermediate neutrons. (A.Y.)

  15. Multigroup computation of the temperature-dependent Resonance Scattering Model (RSM) and its implementation

    Energy Technology Data Exchange (ETDEWEB)

    Ghrayeb, S. Z. [Dept. of Mechanical and Nuclear Engineering, Pennsylvania State Univ., 230 Reber Building, Univ. Park, PA 16802 (United States); Ouisloumen, M. [Westinghouse Electric Company, 1000 Westinghouse Drive, Cranberry Township, PA 16066 (United States); Ougouag, A. M. [Idaho National Laboratory, MS-3860, PO Box 1625, Idaho Falls, ID 83415 (United States); Ivanov, K. N.

    2012-07-01

    A multi-group formulation for the exact neutron elastic scattering kernel is developed. This formulation is intended for implementation into a lattice physics code. The correct accounting for the crystal lattice effects influences the estimated values for the probability of neutron absorption and scattering, which in turn affect the estimation of core reactivity and burnup characteristics. A computer program has been written to test the formulation for various nuclides. Results of the multi-group code have been verified against the correct analytic scattering kernel. In both cases neutrons were started at various energies and temperatures and the corresponding scattering kernels were tallied. (authors)

  16. Distributions of neutron and gamma doses in phantom under a mixed field

    International Nuclear Information System (INIS)

    Beraud-Sudreau, E.

    1982-06-01

    A calculation program, based on Monte Carlo method, allowed to estimate the absorbed doses relatives to the reactor primary radiation, in a water cubic phantom and in cylindrical phantoms modelized from tissue compositions. This calculation is a theoretical approach of gamma and neutron dose gradient study in an animal phantom. PIN junction dosimetric characteristics have been studied experimentally. Air and water phantom radiation doses measured by PIN junction and lithium 7 fluoride, in reactor field have been compared to doses given by dosimetry classical techniques as tissue equivalent plastic and aluminium ionization chambers. Dosimeter responses have been employed to evaluate neutron and gamma doses in plastinaut (tissue equivalent plastic) and animal (piglet). Dose repartition in the piglet bone medulla has been also determined. This work has been completed by comparisons with Doerschell, Dousset and Brown results and by neutron dose calculations; the dose distribution related to lineic energy transfer in Auxier phantom has been also calculated [fr

  17. Peripheral photon and neutron doses from prostate cancer external beam irradiation.

    Science.gov (United States)

    Bezak, Eva; Takam, Rundgham; Marcu, Loredana G

    2015-12-01

    Peripheral photon and neutron doses from external beam radiotherapy (EBRT) are associated with increased risk of carcinogenesis in the out-of-field organs; thus, dose estimations of secondary radiation are imperative. Peripheral photon and neutron doses from EBRT of prostate carcinoma were measured in Rando phantom. (6)LiF:Mg,Cu,P and (7)LiF:Mg,Cu,P glass-rod thermoluminescence dosemeters (TLDs) were inserted in slices of a Rando phantom followed by exposure to 80 Gy with 18-MV photon four-field 3D-CRT technique. The TLDs were calibrated using 6- and 18-MV X-ray beam. Neutron dose equivalents measured with CR-39 etch-track detectors were used to derive readout-to-neutron dose conversion factor for (6)LiF:Mg,Cu,P TLDs. Average neutron dose equivalents per 1 Gy of isocentre dose were 3.8±0.9 mSv Gy(-1) for thyroid and 7.0±5.4 mSv Gy(-1) for colon. For photons, the average dose equivalents per 1 Gy of isocentre dose were 0.2±0.1 mSv Gy(-1) for thyroid and 8.1±9.7 mSv Gy(-1) for colon. Paired (6)LiF:Mg,Cu,P and (7)LiF:Mg,Cu,P TLDs can be used to measure photon and neutron doses simultaneously. Organs in close proximity to target received larger doses from photons than those from neutrons whereas distally located organs received higher neutron versus photon dose. © The Author 2015. Published by Oxford University Press. All rights reserved. For Permissions, please email: journals.permissions@oup.com.

  18. NEUTRON GENERATOR FACILITY AT SFU: GEANT4 DOSE RATE PREDICTION AND VERIFICATION.

    Science.gov (United States)

    Williams, J; Chester, A; Domingo, T; Rizwan, U; Starosta, K; Voss, P

    2016-11-01

    Detailed dose rate maps for a neutron generator facility at Simon Fraser University were produced via the GEANT4 Monte Carlo framework. Predicted neutron dose rates throughout the facility were compared with radiation survey measurements made during the facility commissioning process. When accounting for thermal neutrons, the prediction and measurement agree within a factor of 2 or better in most survey locations, and within 10 % inside the vault housing the neutron generator. © The Author 2015. Published by Oxford University Press. All rights reserved. For Permissions, please email: journals.permissions@oup.com.

  19. Optimal calculational schemes for solving multigroup photon transport problem

    International Nuclear Information System (INIS)

    Dubinin, A.A.; Kurachenko, Yu.A.

    1987-01-01

    A scheme of complex algorithm for solving multigroup equation of radiation transport is suggested. The algorithm is based on using the method of successive collisions, the method of forward scattering and the spherical harmonics method, and is realized in the FORAP program (FORTRAN, BESM-6 computer). As an example the results of calculating reactor photon transport in water are presented. The considered algorithm being modified may be used for solving neutron transport problems

  20. Photon and neutron doses of the personnel using moisture and density measurement devices

    Energy Technology Data Exchange (ETDEWEB)

    Carinou, E.; Papadomarkaki, E.; Tritakis, P.; Hourdakis, C.I.; Kamenopoulou, V. [Greek Atomic Energy Commission, Agia Paraskevi, Attiki, 60092 (Greece)

    2006-07-01

    The objective of this study is to present the evolution of the photon doses received by the workers who use mobile devices for measuring the moisture and the density in various materials and to estimate the neutron doses. The workers employed in more than 30 construction companies in Greece were 76 in 2004. The devices used for that purpose incorporate a {sup 137}Cs source for density measurements and an {sup 241}Am-Be source for moisture measurements of soil, asphalt or concrete. Photon and neutron measurements were performed occasionally during the on site inspections. The results of the measurements showed that the photon and neutron dose rates were not negligible. The workers were monitored for photon radiation using film badges (Kodak Type 2, Holder NRPB type) till the year 2000 and then TLD badges issued by the Greek Atomic Energy Commission (GAEC), on a monthly basis. Since the neutron dose rates measured by a rem-meter were not so high, no neutron dosemeters were issued for them. Their personal dose equivalent data for photons are kept in the National Dose Registry Information System (N.D.R.I.S.) in G.A.E.C. and were used for statistical analysis for the period from 1997 till 2004. As far as the neutrons are concerned, a Monte Carlo code was used to simulate the measuring devices and the working positions in order to calculate the neutron individual doses. (authors)

  1. Photon and neutron doses of the personnel using moisture and density measurement devices

    International Nuclear Information System (INIS)

    Carinou, E.; Papadomarkaki, E.; Tritakis, P.; Hourdakis, C.I.; Kamenopoulou, V.

    2006-01-01

    The objective of this study is to present the evolution of the photon doses received by the workers who use mobile devices for measuring the moisture and the density in various materials and to estimate the neutron doses. The workers employed in more than 30 construction companies in Greece were 76 in 2004. The devices used for that purpose incorporate a 137 Cs source for density measurements and an 241 Am-Be source for moisture measurements of soil, asphalt or concrete. Photon and neutron measurements were performed occasionally during the on site inspections. The results of the measurements showed that the photon and neutron dose rates were not negligible. The workers were monitored for photon radiation using film badges (Kodak Type 2, Holder NRPB type) till the year 2000 and then TLD badges issued by the Greek Atomic Energy Commission (GAEC), on a monthly basis. Since the neutron dose rates measured by a rem-meter were not so high, no neutron dosemeters were issued for them. Their personal dose equivalent data for photons are kept in the National Dose Registry Information System (N.D.R.I.S.) in G.A.E.C. and were used for statistical analysis for the period from 1997 till 2004. As far as the neutrons are concerned, a Monte Carlo code was used to simulate the measuring devices and the working positions in order to calculate the neutron individual doses. (authors)

  2. Accuracy of neutron dose evaluation in the area monitoring for LHD experiments

    CERN Document Server

    Yamanishi, H; Uda, T; Tanahashi, S; Saitou, M; Handa, H

    2000-01-01

    The error in the evaluation of neutron dose during calculation of the neutron field around the large helical device (LHD) in D-D operation is discussed. The expected neutron dose at each monitoring point was derived from the dose conversion factor and neutron fluence data, which was calculated with the radiation transport code DOT-3.5. In contrast, the detected dose at the neutron counter was obtained from the fluence data and the detector response given by calculation with MCNP-4b. The neutron counter used in these calculations consisted of a helium-3 proportional counter with a cylindrical polyethylene moderator. According to the results of the calculations, the ratio of the detected dose to the expected dose was found to lie in the range 1.0-3.0 on the outdoor monitoring points. Since the response of a single neutron counter may lead to inconsistencies in the dose conversion factor, we attempted to minimize these inconsistencies by using a pair of counters with moderators of different thickness. The ratio ...

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

    Energy Technology Data Exchange (ETDEWEB)

    Verbeke, Jerome M.

    1999-12-14

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

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

    International Nuclear Information System (INIS)

    Verbeke, Jerome M.

    1999-01-01

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

  5. Tensile property changes of metals and irradiated to low doses with fission, fusion and spallation neutrons

    International Nuclear Information System (INIS)

    Heinisch, H.L.; Hamilton, M.L.; Sommer, W.F.; Ferguson, P.D.

    1992-01-01

    The objective of this work is to investigate the effects of the neutron energy spectrum in low dose irradiations on the microstructures and mechanical properties of metals. Radiation effects due to low doses of spallation neutrons are compared directly to those produced by fission and fusion neutrons. Yield stress changes of pure Cu, alumina-dispersion-strengthened Cu and AISI 316 stainless steel irradiated at 36-55 C in the Los Alamos Spallation Radiation Effects Facility (LASREF) are compared with earlier results of irradiations at 90 C using 14 MeV D-T fusion neutrons at the Rotating Target Neutron Source and fission reactor neutrons in the Omega West Reactor. At doses up to 0.04 displacements per atom (dpa), the yield stress changes due to the three quite different neutron spectra correlate well on the basis of dpa in the stainless steel and the Cu alloy. However, in pure Cu, the measured yield stress changes due to spallation neutrons were anomalously small and should be verified by additional irradiations. With the exception of pure Cu, the low dose, low temperature experiments reveal no fundamental differences in radiation hardening by fission, fusion or spallation neutrons when compared on the basis of dpa

  6. Study on method of dose estimation for the Dual-moderated neutron survey meter

    International Nuclear Information System (INIS)

    Zhou, Bo; Li, Taosheng; Xu, Yuhai; Gong, Cunkui; Yan, Qiang; Li, Lei

    2013-01-01

    In order to study neutron dose measurement in high energy radiation field, a Dual-moderated survey meter in the range from 1 keV to 300 MeV mean energies spectra has been developed. Measurement results of some survey meters depend on the neutron spectra characteristics in different neutron radiation fields, so the characteristics of the responses to various neutron spectra should be studied in order to get more reasonable dose. In this paper the responses of the survey meter were calculated under different neutron spectra data from IAEA of Technical Reports Series No. 318 and other references. Finally one dose estimation method was determined. The range of the reading per H*(10) for the method estimated is about 0.7–1.6 for the neutron mean energy range from 50 keV to 300 MeV. -- Highlights: • We studied a novel high energy neutron survey meter. • Response characteristics of the survey meter were calculated by using a series of neutron spectra. • One significant advantage of the survey meter is that it can provide mean energy of radiation field. • Dose estimate deviation can be corrected. • The range of corrected reading per H*(10) is about 0.7–1.6 for the neutron fluence mean energy range from 0.05 MeV to 300 MeV

  7. Nuclear data and multigroup methods in fast reactor calculations

    International Nuclear Information System (INIS)

    Gur, Y.

    1975-03-01

    The work deals with fast reactor multigroup calculations, and the efficient treatment of basic nuclear data, which serves as raw material for the calculations. Its purpose is twofold: to build a computer code system that handles a large, detailed library of basic neutron cross section data, (such as ENDF/B-III) and yields a compact set of multigroup cross sections for reactor calculations; to use the code system for comparative analysis of different libraries, in order to discover basic uncertainties that still exist in the measurement of neutron cross sections, and to determine their influence upon uncertainties in nuclear calculations. A program named NANICK which was written in two versions is presented. The first handles the American basic data library, ENDF/B-III, while the second handles the German basic data library, KEDAK. The mathematical algorithm is identical in both versions, and only the file management is different. This program calculates infinitely diluted multigroup cross sections and scattering matrices. It is complemented by the program NASIF that calculates shielding factors from resonance parameters. Different versions of NASIF were written to handle ENDF/B-III or KEDAK. New methods for evaluating in reactor calculations the long term behavior of the neutron flux as well as its fine structure are described and an efficient calculation of the shielding factors from resonance parameters is offered. (B.G.)

  8. Development of neutron dosimeter using CR-39 for measurement of ambient dose equivalent

    International Nuclear Information System (INIS)

    Maki, Daisuke; Shinozaki, Wakako; Ohguchi, Hiroyuki; Yamamoto, Takayoshi; Nakamura, Takayoshi

    2010-01-01

    A CR-39 has good advantages such as cumulative type dosimeter, small fading effect and gamma-ray insensitive. Therefore, we developed the wide energy-range environmental neutron dosimeter using eight CR-39s for area monitoring in this study. This dosimeter is made of octagonal columnar polyethylene block which height is 60 mm and bottom side is 25 mm. The dosimeter contains two types of CR-39s for fast neutron detection and slow neutron detection. Four CR-39s for fast neutron detection are used for detection of recoil protons produced by H (n, p) reactions. Four CR-39s for slow neutron detection are used with boron nitride converter to detect alpha-rays produced by 10 B (n, α) 7 Li reactions. Ambient dose equivalent is obtained by adding the number of etch-pits observed in four CR-39s for fast neutron detection to the number of etch-pits observed in four CR-39s for slow neutron detection with appropriate constants respectively. Dosimeters were irradiated with some energetic neutrons and evaluated results of ambient dose equivalent were compared with results from neutron transport calculations. Energy response of dosimeter shows good agreement with neutron fluence to ambient dose equivalent conversion coefficients. Directional dependence of dosimeter is at the same level as the rem-counter. (author)

  9. Phantom experiment of depth-dose distributions for gadolinium neutron capture therapy

    International Nuclear Information System (INIS)

    Matsumoto, T.; Kato, K.; Sakuma, Y.; Tsuruno, A.; Matsubayashi, M.

    1993-01-01

    Depth-dose distributions in a tumor simulated phantom were measured for thermal neutron flux, capture gamma-ray and internal conversion electron dose rates for gadolinium neutron capture therapy. The results show that (i) a significant dose enhancement can be achieved in the tumor by capture gamma-rays and internal conversion electrons but the dose is mainly due to capture gamma-rays from the Gd(n, γ) reactions, therefore, is not selective at the cellular level, (ii) the dose distribution was a function of strongly interrelated parameters such as gadolinium concentrations, tumor site and neutron beam size (collimator aperture size), and (iii) the Gd-NCT by thermal neutrons appears to be a potential for treatment of superficial tumor. (author)

  10. Investigation of dose distribution in mixed neutron-gamma field of boron neutron capture therapy using N isopropylacrylamide gel

    Energy Technology Data Exchange (ETDEWEB)

    Bavarmegin, Elham; Sadremomtaz, Alireza [Nuclear Science and Technology Research Institute (NSTRI), Tehran (Iran, Islamic Republic of); Khalafi, Hossein; Kasesaz, Yaser [Dept. of Physics, University of Guilan, Rasht (Iran, Islamic Republic of); Khajeali, Azim [Medical Education Research Center, Tabriz (Iran, Islamic Republic of)

    2017-02-15

    Gel dosimeters have unique advantages in comparison with other dosimeters. Until now, these gels have been used in different radiotherapy techniques as a reliable dosimetric tool. Because dose distribution measurement is an important factor for appropriate treatment planning in different radiotherapy techniques, in this study, we evaluated the ability of the N-isopropylacrylamide (NIPAM) polymer gel to record the dose distribution resulting from the mixed neutron-gamma field of boron neutron capture therapy (BNCT). In this regard, a head phantom containing NIPAM gel was irradiated using the Tehran Research Reactor BNCT beam line, and then by a magnetic resonance scanner. Eventually, the R2 maps were obtained in different slices of the phantom by analyzing T2-weighted images. The results show that NIPAM gel has a suitable potential for recording three-dimensional dose distribution in mixed neutron-gamma field dosimetry.

  11. Dose planning with comparison to in vivo dosimetry for epithermal neutron irradiation of the dog brain

    International Nuclear Information System (INIS)

    Seppaelae, Tiina; Auterinen, Iiro; Aschan, Carita; Seren, Tom; Benczik, Judit; Snellman, Marjatta; Huiskamp, Rene; Ramadan, Usama Abo; Kankaanranta, Leena; Joensuu, Heikki; Savolainen, Sauli

    2002-01-01

    Boron neutron capture therapy (BNCT) is an experimental type of radiotherapy, presently being used to treat glioblastoma and melanoma. To improve patient safety and to determine the radiobiological characteristics of the epithermal neutron beam of Finnish BNCT facility (FiR 1) dose-response studies were carried on the brain of dogs before starting the clinical trials. A dose planning procedure was developed and uncertainties of the epithermal neutron-induced doses were estimated. The accuracy of the method of computing physical doses was assessed by comparing with in vivo dosimetry. Individual radiation dose plans were computed using magnetic resonance images of the heads of 15 Beagle dogs and the computational model of the FiR 1 epithermal neutron beam. For in vivo dosimetry, the thermal neutron fluences were measured using Mn activation foils and the gamma-ray doses with MCP-7s type thermoluminescent detectors placed both on the skin surface of the head and in the oral cavity. The degree of uncertainty of the reference doses at the thermal neutron maximum was estimated using a dose-planning program. The estimated uncertainty (±1 standard deviation) in the total physical reference dose was ±8.9%. The calculated and the measured dose values agreed within the uncertainties at the point of beam entry. The conclusion is that the dose delivery to the tissue can be verified in a practical and reliable fashion by placing an activation dosimeter and a TL detector at the beam entry point on the skin surface with homogeneous tissues below. However, the point doses cannot be calculated correctly in the inhomogeneous area near air cavities of the head model with this type of dose-planning program. This calls for attention in dose planning in human clinical trials in the corresponding areas

  12. Dose levels due to neutrons in the vicinity of high energy medical accelerators

    International Nuclear Information System (INIS)

    McGinley, P.H.; Wood, M.; Sohrabi, M.; Mills, M.; Rodriguez, R.

    1976-01-01

    High energy photons are generated for use in radiation therapy by the decelleration of electrons in metal targets. Fast neutrons are also generated as a result of (γ, n) and (e, e'n) interactions in the target, beam compensator filter, and collimator material. In this work the adsorbed dose to neutrons was measured at the center of a 10 x 10 cm photon beam and 5 cm outside of the beam edge for a number of treatment units. Dose levels due to slow and fast neutrons were also established outside of the treatment rooms and a Bonner sphere neutron spectrometer system was employed to determine the neutron energy spectrum due to stray neutron radiation at each accelerator. For the linac it was found that the neutron dose at the beam center was 0.0039% of the photon dose and values of 0.049% and 0.053% were observed for the Allis Chalmers betatron and the Brown Boveri Betatron. Dose equivalent rates in the range of 0.3 to 22.5 mrem/hr were measured for points outside the treatment rooms when the accelerators were operated at a photon dose rate of 100 rad/min at the treatment position

  13. Intercomparison of personnel dosimetry for thermal neutron dose equivalent in neutron and gamma-ray mixed fields

    International Nuclear Information System (INIS)

    Ogawa, Yoshihiro

    1985-01-01

    In order to consider the problems concerned with personnel dosimetry using film badges and TLDs, an intercomparison of personnel dosimetry, especially dose equivalent responses of personnel dosimeters to thermal neutron, was carried out in five different neutron and gamma-ray mixed fields at KUR and UTR-KINKI from the practical point of view. For the estimation of thermal neutron dose equivalent, it may be concluded that each personnel dosimeter has good performances in the precision, that is, the standard deviations in the measured values by individual dosimeter were within 24 %, and the dose equivalent responses to thermal neutron were almost independent on cadmium ratio and gamma-ray contamination. However, the relative thermal neutron dose equivalent of individual dosimeter normalized to the ICRP recommended value varied considerably and a difference of about 4 times was observed among the dosimeters. From the results obtained, it is suggested that the standardization of calibration factors and procedures is required from the practical point of radiation protection and safety. (author)

  14. Neutron doses to personnel from a 24 MeV betatron

    International Nuclear Information System (INIS)

    Beckham, W.A; Entwistle, R.F.

    1987-01-01

    Neutrons are produced by bombardment of most materials by high-energy photons. Because the x-ray shielding around high-energy x-ray generators may not have been designed with neutrons in mind there may be unexpected contributions to the radiation doses of staff working in the immediate vicinity. Neutron fluxes in the working area close to an Allis-Chalmers 24 MeV betatron have been measured using a lithium-6-loaded scintillator and the dose rates calculated. Hazard of staff has been found to be low; typical dose-equivalent rates in occupied areas range from 0.0042 to 0.012 mrem/hour. The flux of fast neutrons in the treatment room was found to be essentially zero. Measurements of neutron flux may be routinely performed using the scintillation detector (NE 912) described, and could usefully form part of the acceptance protocol for any new accelerator

  15. Absorbed dose conversion coefficients for embryo and foetus in neutron fields

    International Nuclear Information System (INIS)

    Chen, J.

    2007-01-01

    The Monte Carlo code MCNPX has been used to determine mean absorbed doses to the embryo and foetus when the mother is exposed to neutron fields. There are situations, such as on-board aircraft, where high-energy neutrons are often peaked in top down (TOP) direction. In addition to previous publications for standard irradiation geometries, this study provides absorbed dose conversion coefficients for the embryo of 8 weeks and the foetus of 3, 6 or 9 months at TOP irradiation geometry. The conversion coefficients are compared with the coefficients in isotropic irradiation (ISO). With increasing neutron energies, the conversion coefficients in TOP irradiation become dominant. A set of conversion coefficients is constructed from the higher value in either ISO or TOP irradiation at a given neutron energy. In cases where the irradiation geometry is not adequately known, this set of conversion coefficients can be used in a conservative dose assessment for embryo and foetus in neutron fields. (authors)

  16. Monitor units are not predictive of neutron dose for high-energy IMRT

    Directory of Open Access Journals (Sweden)

    Hälg Roger A

    2012-08-01

    Full Text Available Abstract Background Due to the substantial increase in beam-on time of high energy intensity-modulated radiotherapy (>10 MV techniques to deliver the same target dose compared to conventional treatment techniques, an increased dose of scatter radiation, including neutrons, is delivered to the patient. As a consequence, an increase in second malignancies may be expected in the future with the application of intensity-modulated radiotherapy. It is commonly assumed that the neutron dose equivalent scales with the number of monitor units. Methods Measurements of neutron dose equivalent were performed for an open and an intensity-modulated field at four positions: inside and outside of the treatment field at 0.2 cm and 15 cm depth, respectively. Results It was shown that the neutron dose equivalent, which a patient receives during an intensity-modulated radiotherapy treatment, does not scale with the ratio of applied monitor units relative to an open field irradiation. Outside the treatment volume at larger depth 35% less neutron dose equivalent is delivered than expected. Conclusions The predicted increase of second cancer induction rates from intensity-modulated treatment techniques can be overestimated when the neutron dose is simply scaled with monitor units.

  17. Axial distribution of absorbed doses in fast neutron field at the RB reactor

    International Nuclear Information System (INIS)

    Sokcic-Kostic, M.; Pesic, M.; Antic, D.; Ninkovic, M.

    1988-11-01

    The coupled fast thermal system CFTS at the RB reactor is created for obtaining fast neutron fields. The axial distribution of fast neutron flux density in its second configuration (CFTS-2) is measured. The axial distribution of absorbed doses is computed on the basis of mentioned experimental results. At the end these experimental and computed results are given. (Author)

  18. Use of prompt gamma emissions from polyethylene to estimate neutron ambient dose equivalent

    Energy Technology Data Exchange (ETDEWEB)

    Priyada, P.; Sarkar, P.K., E-mail: pradip.sarkar@manipal.edu

    2015-06-11

    The possibility of using measured prompt gamma emissions from polyethylene to estimate neutron ambient dose equivalent is explored theoretically. Monte Carlo simulations have been carried out using the FLUKA code to calculate the response of a high density polyethylene cylinder to emit prompt gammas from interaction of neutrons with the nuclei of hydrogen and carbon present in polyethylene. The neutron energy dependent responses of hydrogen and carbon nuclei are combined appropriately to match the energy dependent neutron fluence to ambient dose equivalent conversion coefficients. The proposed method is tested initially with simulated spectra and then validated using experimental measurements with an Am–Be neutron source. Experimental measurements and theoretical simulations have established the feasibility of estimating neutron ambient dose equivalent using measured neutron induced prompt gammas emitted from polyethylene with an overestimation of neutron dose at very low energies. - Highlights: • A new method for estimating H{sup ⁎}(10) using prompt gamma emissions from HDPE. • Linear combination of 2.2 MeV and 4.4 MeV gamma intensities approximates DCC (ICRP). • Feasibility of the method was established theoretically and experimentally. • The response of the present technique is very similar to that of the rem meters.

  19. Nominal effective radiation doses delivered during clinical trials of boron neutron capture therapy

    International Nuclear Information System (INIS)

    Capala, J.; Diaz, A.Z.; Chanana, A.D.

    1997-01-01

    Boron neutron capture therapy (BNCT) is a binary system that, in theory, should selectively deliver lethal, high linear energy transfer (LET) radiation to tumor cells dispersed within normal tissues. It is based on the nuclear reaction 10-B(n, α)7-Li, which occurs when the stable nucleus of boron-10 captures a thermal neutron. Due to the relatively high cross-section of the 10-B nucleus for thermal neutron capture and short ranges of the products of this reaction, tumor cells in the volume exposed to thermal neutrons and containing sufficiently high concentration of 10-B would receive a much higher radiation dose than the normal cells contained within the exposed volume. Nevertheless, radiation dose deposited in normal tissue by gamma and fast neutron contamination of the neutron beam, as well as neutron capture in nitrogen, 14-N(n,p)14-C, hydrogen, 1-H(n,γ)2-H, and in boron present in blood and normal cells, limits the dose that can be delivered to tumor cells. It is, therefore, imperative for the success of the BNCT the dosed delivered to normal tissues be accurately determined in order to optimize the irradiation geometry and to limit the volume of normal tissue exposed to thermal neutrons. These are the major objectives of BNCT treatment planning

  20. Radiation dose distribution monitoring at neutron radiography facility area, Nuclear Energy Unit, Malaysia

    International Nuclear Information System (INIS)

    Abdul Razak Daud

    1995-01-01

    One experiment was carried out to get the distribution of radiation doses at the neutron radiography facilities, Nuclear Energy Unit, Malaysia. The analysis was done to evaluate the safety level of the area. The analysis was used in neutron radiography work

  1. Dose conversion coefficients for high-energy photons, electrons, neutrons and protons

    International Nuclear Information System (INIS)

    Sakamoto, Yukio

    2005-01-01

    Dose conversion coefficients for photons, electrons and neutrons based on new ICRP recommendations were cited in the ICRP Publication 74, but the energy ranges of these data were limited and there are no data for high energy radiations produced in accelerator facilities. For the purpose of designing the high intensity proton accelerator facilities at JAERI, the dose evaluation code system of high energy radiations based on the HERMES code was developed and the dose conversion coefficients of effective dose were evaluated for photons, neutrons and protons up to 10 GeV, and electrons up to 100 GeV. The dose conversion coefficients of effective dose equivalent were also evaluated using quality factors to consider the consistency between radiation weighting factors and Q-L relationship. The effective dose conversion coefficients obtained in this work were in good agreement with those recently evaluated by using FLUKA code for photons and electrons with all energies, and neutrons and protons below 500 MeV. There were some discrepancy between two data owing to the difference of cross sections in the nuclear reaction models. The dose conversion coefficients of effective dose equivalents for high energy radiations based on Q-L relation in ICRP Publication 60 were evaluated only in this work. The previous comparison between effective dose and effective dose equivalent made it clear that the radiation weighting factors for high energy neutrons and protons were overestimated and the modification was required. (author)

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

    International Nuclear Information System (INIS)

    Moin-Vasiri, M.

    1990-06-01

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

  3. An analytical multigroup benchmark for (n, γ) and (n, n', γ) verification of diffusion theory algorithms

    International Nuclear Information System (INIS)

    Ganapol, B.D.

    2011-01-01

    Highlights: → Coupled neutron and gamma transport is considered in the multigroup diffusion approximation. → The model accommodates fission, up- and down-scattering and common neutron-gamma interactions. → The exact solution to the diffusion equation in a heterogeneous media of any number of regions is found. → The solution is shown to parallel the one-group case in a homogeneous medium. → The discussion concludes with a heterogeneous, 2 fuel-plate 93.2% enriched reactor fuel benchmark demonstration. - Abstract: The angular flux for the 'rod model' describing coupled neutron/gamma (n, γ) diffusion has a particularly straightforward analytical representation when viewed from the perspective of a one-group homogeneous medium. Cast in the form of matrix functions of a diagonalizable matrix, the solution to the multigroup equations in heterogeneous media is greatly simplified. We shall show exactly how the one-group homogeneous medium solution leads to the multigroup solution.

  4. Device for measuring the dose rate of pulsed neutrons

    International Nuclear Information System (INIS)

    Klett, A.

    2009-01-01

    The author presents a new apparatus, developed in collaboration by Berthold Technologies and the German company DESY, allowing neutron pulsed fields to be measured. It is based on the activation by high energy neutrons of carbon 12 present in the sensor materials, and on the decay of short life radionuclides produced by this activation. The detection principle and system are briefly presented

  5. Dose field research of analysis room for in-hospital neutron irradiator

    International Nuclear Information System (INIS)

    Zhang Zizhu; Song Mingzhe; Li Wei; Chen Jun; Yang Yong; Li Yiguo

    2012-01-01

    Neutron equivalent dose rate and y ray dose rate inside the analysis room of the in-hospital neutron irradiator (IHNI) and outdoor were measured. The results show that γ ray dose rate inside the analysis room exceeds calculation value many times and γ/ ray dose rate outdoor is higher than supervision region dose limit of 7.5 μSv/h. According to the measurement results and the Monte Carlo simulation, the following shielding plan was adopted. Lead shielding with thickness of 16 cm was installed on the wall, which faces the neutron beam, to shield γ ray, and lithium polyethylene plate with thickness of l cm was installed on all the wall (not including ceiling and floor) to shield scattering neutron. After shielding transformation, the highest γ ray dose rate point inside the analysis room decreased 277 times, the neutron equivalent dose rate decreased 5.8 times, and the outdoor γ/ray dose rate decreased nearly 90 times. (authors)

  6. Multigroup neutron data base for nuclear geophysics

    International Nuclear Information System (INIS)

    Dworak, D.; Loskiewicz, J.

    1989-01-01

    The average group constants for the total, elastic, inelastic and capture cross sections as well as the average cosine of the scattering angle for elastic scattering and the average logarithmic energy decrement for elastic scattering have been obtained at two temperatures (300 and 400 deg K), using the ENDF/B-4 data and the IAEA-NDS pre-processing codes. The extended Abagyan group structure and the weighting spectrum of type 1/E were applied in course of the calculations. Self-shielding effect was not taken into account. All cross sections were Doppler broadened for both, 300 and 400 deg K temperatures. Under above assumptions, the average group constants were obtained for exactly 22 ENDF materials, which are of special importance for nuclear geophysics applications. 10 refs., 15 figs., 44 tabs. (author)

  7. Experimental Determination of the Neutron Radiation-Dose Distribution in the Human Phantom

    Energy Technology Data Exchange (ETDEWEB)

    Stipcic, Neda [Institute Rudjer Bogkovic, Zagreb, Yugoslavia (Serbia)

    1967-01-15

    The quality of the radiation delivering the radiation dose to the human phantom is quite different from that of the incident neutron beam. This paper describes the experimental investigation of the variation of neutron dose related to the variation of neutron fluence with depth in the human phantom. The distribution of neutron radiation was determined in the human phantom - a cube of paraffin wax 25 cm x 25 cm x 50 cm with a density of 0.92 cm{sup -3}. Po-Be and Ra-Be point sources were used as neutron sources. Neutron fluences were measured using different types of detector: scintillation detector, BF{sub 3} counter, and nuclear-track emulsions. Since the fluence measurements with these three types of detectors were carried out under the same experimental conditions, it was possible to separate and analyse each part of the radiation dose in the paraffin. From the investigations, the distribution of the total radiation dose was obtained as a function of the paraffin depth. The maximum value of this dose distribution is constant with respect to the distance between the source and the paraffin phantom. From the results obtained, some conclusions may be drawn concerning the amount of absorbed radiation dose in the human phantom. (author)

  8. Kalpakkam multigroup cross section set for fast reactor applications - status and performance

    International Nuclear Information System (INIS)

    Ramanadhan, M.M.; Gopalakrishnan, M.M.

    1986-01-01

    This report documents the status of the presently created set of multigroup constants at Kalpakkam. The list of nuclides processed and the details of multigroup structure are given. Also included are the particulars of dilutions and temperatures for each nuclide in the multigroup cross section set for which self shielding factors have been calculated. Using this new multigroup cross section set, measured integral quantities such as K-eff, central reaction rate ratios, central reactivity worths etc. were calculated for a few fast critical benchmark assemblies and the calculated values of neutronic parameters obtained were compared with those obtained using the available Cadarache cross section library and those published in literature for ENDF/B-IV based set and Japanese evaluated nuclear data library (JENDL). The details of analyses are documented along with the conclusions. (author). 17 refs., 12 tabs

  9. Dose conversion coefficients for high-energy photons, electrons, neutrons and protons

    CERN Document Server

    Sakamoto, Y; Sato, O; Tanaka, S I; Tsuda, S; Yamaguchi, Y; Yoshizawa, N

    2003-01-01

    In the International Commission on Radiological Protection (ICRP) 1990 Recommendations, radiation weighting factors were introduced in the place of quality factors, the tissue weighting factors were revised, and effective doses and equivalent doses of each tissues and organs were defined as the protection quantities. Dose conversion coefficients for photons, electrons and neutrons based on new ICRP recommendations were cited in the ICRP Publication 74, but the energy ranges of theses data were limited and there are no data for high energy radiations produced in accelerator facilities. For the purpose of designing the high intensity proton accelerator facilities at JAERI, the dose evaluation code system of high energy radiations based on the HERMES code was developed and the dose conversion coefficients of effective dose were evaluated for photons, neutrons and protons up to 10 GeV, and electrons up to 100 GeV. The dose conversion coefficients of effective dose equivalent were also evaluated using quality fact...

  10. Characterization of thermal neutron fields for calibration of neutron monitors in accordance with great equivalent dose environment H⁎(10)

    International Nuclear Information System (INIS)

    Silva, Larissa P. S. da; Silva, Felipe S.; Fonseca, Evaldo S.; Patrao, Karla C.S.; Pereira, Walsan W.

    2017-01-01

    The Laboratório Brasileiro de Nêutrons do Instituto de Radioproteção e Dosimetria (IRD/CNEN) has developed and built a thermal neutron flux facility to provide neutron fluence for dosimeters (Astuto, 2014). This fluency is obtained by four 16 Ci sources 241 AmBe (α, n) positioned around the channel positioned in the center of the Thermal Flow Unit (UFT). The UFT was built with blocks of paraffin with graphite addition and graphite blocks of high purity to obtain a central field with a homogeneous thermal neutron fluence for calibration purposes with the following measurements: 1.2 x 1.2 x 1.2 m 3 . The objective of this work is to characterize several points, in the thermal energy range, in terms of the equivalent ambient dose quantity H⁎(10) for calibration and irradiation of monitors neutrons

  11. Experimental possibilities and fast neutron dose map of the fast neutron fields at the RB reactor facility

    International Nuclear Information System (INIS)

    Sokcic-Kostic, M.; Pesic, M.; Antic, D.; Ninkovic, M.

    1993-01-01

    The RB is an unshielded, zero power nuclear facility with natural and enriched uranium fuel (2% and 80%) and D 2 O as moderator. It is possible to create different configurations of non-reflected and partially reflected critical systems and to make experiments in the fields of thermal neutrons. The fields of fast neutrons with 'softened' fission spectrum are made by modifying the system: modified experimental fuel channel EFC, coupled fast-thermal system in two configurations CFTS-1 and CFTS-2, coupled fast-thermal core HERBE. The intermediate and fast neutron absorbed doses in fast neutron fields are given. In first configuration of RB reactor it was almost impossible to perform dosimetric and other experiments. By creating these fields, with in our circumstances available fuel elements, the possibilities for different experiments are greatly improved. Now we can irradiate food samples, soil samples, electronic devices, study material properties, perform various dosimetry experiments, etc. (1 tab.)

  12. Validation of dose planning calculations for boron neutron capture therapy using cylindrical and anthropomorphic phantoms

    Energy Technology Data Exchange (ETDEWEB)

    Koivunoro, Hanna; Seppaelae, Tiina; Uusi-Simola, Jouni; Merimaa, Katja; Savolainen, Sauli [Department of Physics, POB 64, FI-00014 University of Helsinki (Finland); Kotiluoto, Petri; Seren, Tom; Auterinen, Iiro [VTT Technical Research Centre of Finland, Espoo, POB 1000, FI-02044 VTT (Finland); Kortesniemi, Mika, E-mail: hanna.koivunoro@helsinki.f [HUS Helsinki Medical Imaging Center, University of Helsinki, POB 340, FI-00029 HUS (Finland)

    2010-06-21

    In this paper, the accuracy of dose planning calculations for boron neutron capture therapy (BNCT) of brain and head and neck cancer was studied at the FiR 1 epithermal neutron beam. A cylindrical water phantom and an anthropomorphic head phantom were applied with two beam aperture-to-surface distances (ASD). The calculations using the simulation environment for radiation application (SERA) treatment planning system were compared to neutron activation measurements with Au and Mn foils, photon dose measurements with an ionization chamber and the reference simulations with the MCNP5 code. Photon dose calculations using SERA differ from the ionization chamber measurements by 2-13% (disagreement increased along the depth in the phantom), but are in agreement with the MCNP5 calculations within 2%. The {sup 55}Mn(n,{gamma}) and {sup 197}Au(n,{gamma}) reaction rates calculated using SERA agree within 10% and 8%, respectively, with the measurements and within 5% with the MCNP5 calculations at depths >0.5 cm from the phantom surface. The {sup 55}Mn(n,{gamma}) reaction rate represents the nitrogen and boron depth dose within 1%. Discrepancy in the SERA fast neutron dose calculation (of up to 37%) is corrected if the biased fast neutron dose calculation option is not applied. Reduced voxel cell size ({<=}0.5 cm) improves the SERA calculation accuracy on the phantom surface. Despite the slight overestimation of the epithermal neutrons and underestimation of the thermal neutrons in the beam model, neutron calculation accuracy with the SERA system is sufficient for reliable BNCT treatment planning with the two studied treatment distances. The discrepancy between measured and calculated photon dose remains unsatisfactorily high for depths >6 cm from the phantom surface. Increasing discrepancy along the phantom depth is expected to be caused by the inaccurately determined effective point of the ionization chamber.

  13. Calculation of neutron fluence to dose equivalent conversion coefficients using GEANT4

    International Nuclear Information System (INIS)

    Ribeiro, Rosane M.; Santos, Denison de S.; Queiroz Filho, Pedro P. de; Mauricio, CLaudia L.P.; Silva, Livia K. da; Pessanha, Paula R.

    2014-01-01

    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 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 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 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)

  14. Measurement of neutron and gamma absorbed doses in phantoms exposed to mixed fields

    International Nuclear Information System (INIS)

    Beraud-Sudreau, E.; Lemaire, G.; Maas, J.

    1985-01-01

    In order to study the dosimetric characteristics of PIN junctions, the absorbed doses measured by junctions and FLi7 in air and water phantoms were compared with the doses measured by classical neutron dosimetry in mixed fields. The validity of the experimental responses of PIN junctions being thus checked and established, neutron and gamma dose distributions in tissue equivalent plastic phantoms (plastinaut) and mammals (piglets) were evaluated as well as the absorbed dose distributions in the pig bone-marrow producing areas. By using correlatively a Monte-Carlo calculation method and applying some simplifying assumptions, the absorbed doses were derived from the spectrum of SILENE's neutrons at various depths inside a cubic water phantom and the results were compared with some from the literature [fr

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2006-07-01

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

  16. Secondary neutron doses received by patients of different ages during intracranial proton therapy treatments

    International Nuclear Information System (INIS)

    Sayah, R.

    2012-01-01

    Proton therapy is an advanced radiation therapy technique that allows delivering high doses to the tumor while saving the healthy surrounding tissues due to the protons' ballistic properties. However, secondary particles, especially neutrons, are created during protons' nuclear reactions in the beam-line and the treatment room components, as well as inside the patient. Those secondary neutrons lead to unwanted dose deposition to the healthy tissues located at distance from the target, which may increase the secondary cancer risks to the patients, especially the pediatric ones. The aim of this work was to calculate the neutron secondary doses received by patients of different ages treated at the Institut Curie-centre de Protontherapie d'Orsay (ICPO) for intracranial tumors, using a 178 MeV proton beam. The treatments are undertaken at the new ICPO room equipped with an IBA gantry. The treatment room and the beam-line components, as well as the proton source were modeled using the Monte Carlo code MCNPX. The obtained model was then validated by a series of comparisons between model calculations and experimental measurements. The comparisons concerned: a) depth and lateral proton dose distributions in a water phantom, b) neutron spectrometry at one position in the treatment room, c) ambient dose equivalents at different positions in the treatment room and d) secondary absorbed doses inside a physical anthropomorphic phantom. A general good agreement was found between calculations and measurements, thus our model was considered as validated. The University of Florida hybrid voxelized phantoms of different ages were introduced into the MCNPX validated model, and secondary neutron doses were calculated to many of these phantoms' organs. The calculated doses were found to decrease as the organ's distance to the treatment field increases and as the patient's age increases. The secondary doses received by a one year-old patient may be two times higher than the doses

  17. In-wire measurement of the neutron dose rate on patients with 238Pu pacemakers implanted

    International Nuclear Information System (INIS)

    Piesch, E.; Burgkhardt, B.; Kollmeier, W.

    1975-01-01

    In-vivo measurements of the neutron dose on Medtronic pacemakers have been performed by using a proportional counter and a scintillation counter. The paper discusses the technique of free air and phantom calibration and the method of in-vivo measurement of the neutron fluence and the estimation of the dose equivalent. The neutron dose equivalent rate measured on seven patients with 238 Pu pacemakers implanted were found to be (5.6+-0.1) mRem/h at the surface of the pacemaker in 1.25 cm distance from the center of the source corresponding to a neutron emission rate of 940 ns -1 . The results are in good agreement with results of other methods reported by different authors. (Auth.)

  18. Application of Whole Body Counter to Neutron Dose Assessment in Criticality Accidents

    Energy Technology Data Exchange (ETDEWEB)

    Kurihara, O.; Tsujimura, N.; Takasaki, K.; Momose, T.; Maruo, Y. [Japan Nuclear Cycle Development Institute, Tokai (Japan)

    2001-09-15

    Neutron dose assessment in criticality accidents using Whole Body Counter (WBC) was proved to be an effective method as rapid neutron dose estimation at the JCO criticality accident in Tokai-mura. The 1.36MeV gamma-ray of {sup 24}Na in a body can be detected easily by a germanium detector. The Minimum Detectable Activity (MDA) of {sup 24}Na is approximately 50Bq for 10minute measurement by the germanium-type whole body counter at JNC Tokai Works. Neutron energy spectra at the typical shielding conditions in criticality accidents were calculated and the conversion factor, whole body activity-to-organ mass weighted neutron absorbed dose, corresponding to each condition were determined. The conversion factor for uncollied fission spectrum is 7.7 [(Bq{sup 24}Na/g{sup 23}Na)/mGy].

  19. The alanine detector in BNCT dosimetry: dose response in thermal and epithermal neutron fields.

    Science.gov (United States)

    Schmitz, T; Bassler, N; Blaickner, M; Ziegner, M; Hsiao, M C; Liu, Y H; Koivunoro, H; Auterinen, I; Serén, T; Kotiluoto, P; Palmans, H; Sharpe, P; Langguth, P; Hampel, G

    2015-01-01

    The response of alanine solid state dosimeters to ionizing radiation strongly depends on particle type and energy. Due to nuclear interactions, neutron fields usually also consist of secondary particles such as photons and protons of diverse energies. Various experiments have been carried out in three different neutron beams to explore the alanine dose response behavior and to validate model predictions. Additionally, application in medical neutron fields for boron neutron capture therapy is discussed. Alanine detectors have been irradiated in the thermal neutron field of the research reactor TRIGA Mainz, Germany, in five experimental conditions, generating different secondary particle spectra. Further irradiations have been made in the epithermal neutron beams at the research reactors FiR 1 in Helsinki, Finland, and Tsing Hua open pool reactor in HsinChu, Taiwan ROC. Readout has been performed with electron spin resonance spectrometry with reference to an absorbed dose standard in a (60)Co gamma ray beam. Absorbed doses and dose components have been calculated using the Monte Carlo codes fluka and mcnp. The relative effectiveness (RE), linking absorbed dose and detector response, has been calculated using the Hansen & Olsen alanine response model. The measured dose response of the alanine detector in the different experiments has been evaluated and compared to model predictions. Therefore, a relative effectiveness has been calculated for each dose component, accounting for its dependence on particle type and energy. Agreement within 5% between model and measurement has been achieved for most irradiated detectors. Significant differences have been observed in response behavior between thermal and epithermal neutron fields, especially regarding dose composition and depth dose curves. The calculated dose components could be verified with the experimental results in the different primary and secondary particle fields. The alanine detector can be used without

  20. Neutron dose measurements with the GSI ball at high energy accelerators

    International Nuclear Information System (INIS)

    Fehrenbacher, G.; Gutermuth, F.; Radon, T.; Kozlova, E.

    2005-01-01

    Full text: At high energy particle accelerators the production of neutron radiation dominates radiation protection. For the radiation survey at accelerators there is a need for reliable detection systems (passive radiation monitors), which can measure the dose for a wide range of neutron energies independently on the beam pulse structure of the produced radiation. In this work a passive neutron dosemeter for the measurement of the ambient dose equivalent is presented. The dosemeter is suitable for measurements of the emerging neutron radiation at accelerators for the whole energy range up to about 10 GeV. The dosemeter consists of a polyethylene sphere, TL elements (pairs of TLD600/700) and an additional lead layer (PE/Pb) in neutron fields at high energy accelerators is investigated in this work. Results of dose measurements which were performed in realistic neutron fields at the high energy accelerator SPS at CERN (CERF facility) and in Cave A at the heavy ion synchrotron SIS at GSI are presented. The results of these measurements are compared with the expected dose values from the neutron spectra determined for the measurement positions at CERF and in Cave A (FLUKA) and with the dosemeter response derived by the calculated response functions (FLUKA) folded with the neutron spectra. The comparisons show that the additional lead layer in the PE/Pb-sphere improves significantly the response of the dosemeter. The response of the PE/Pb-sphere is 40 to 50 % higher at CERF and Cave A in comparison to the bare PE-sphere. At CERF the dose values of the PE/Pb-sphere is about 25 % lower than the expected dose value, whilst for Cave A, a rather good agreement was found (2 % deviation). (author)

  1. Method for measuring and evaluation dose equivalent rate from fast neutrons in mixed gamma-neutron fields around particles accelerators

    International Nuclear Information System (INIS)

    Cruceru, I.; Sandu, M.; Cruceru, M.

    1994-01-01

    A method for measuring and evaluation of doses and dose equivalent rate in mixed gamma- neutron fields is discussed in this paper. The method is basedon a double detector system consist of an ionization chamber with components made from a plastic scintillator, coupled to on photomultiplier. Generally the radiation fields around accelerators are complex, often consisting of many different ionizing radiations extending over a broad range of energies. This method solve two major difficulties: determination of response functions of radiation detectors; interpretation of measurement and determination of accuracy. The discrimination gamma-fast neutrons is assured directly without a pulse shape discrimination circuit. The method is applied to mixed fields in which particle energies are situated in the energy range under 20 MeV and an izotropic emision (Φ=10 4 -10 11 n.s -1 ). The dose equivalent rates explored is 0.01mSV--0.1SV

  2. Estimate of neutron secondary doses received by patients in proton therapy: cases of ophthalmologic treatments

    International Nuclear Information System (INIS)

    Martinetti, F.

    2009-12-01

    This research thesis aims at assessing doses due to secondary neutrons and received by the organs of a patient which are located outside of the treatment field. The study focused on ophthalmological treatments performed at the Orsay proton therapy centre. A 75 eV beam line model has first been developed with the MCNPX Monte Carlo code. Several experimental validations of this model have been performed: proton dose distribution in a water phantom, ambient equivalent dose due to secondary neutrons and neutron spectra in the treatment room, and doses deposited by secondary neutrons in an anthropomorphous phantom. Simulations and measurements are in correct agreement. Then, a numeric assessment of secondary doses received by the patient's organs has been performed by using a MIRD-type mathematical phantom. These doses have been computed for several organs: the non-treated eye, the brain, the thyroid, and other parts of the body situated either in the front part of the body (the one directly exposed to neutrons generated in the treatment line) or deeper and further from the treatment field

  3. Monte Carlo simulation of secondary neutron dose for scanning proton therapy using FLUKA.

    Directory of Open Access Journals (Sweden)

    Chaeyeong Lee

    Full Text Available Proton therapy is a rapidly progressing field for cancer treatment. Globally, many proton therapy facilities are being commissioned or under construction. Secondary neutrons are an important issue during the commissioning process of a proton therapy facility. The purpose of this study is to model and validate scanning nozzles of proton therapy at Samsung Medical Center (SMC by Monte Carlo simulation for beam commissioning. After the commissioning, a secondary neutron ambient dose from proton scanning nozzle (Gantry 1 was simulated and measured. This simulation was performed to evaluate beam properties such as percent depth dose curve, Bragg peak, and distal fall-off, so that they could be verified with measured data. Using the validated beam nozzle, the secondary neutron ambient dose was simulated and then compared with the measured ambient dose from Gantry 1. We calculated secondary neutron dose at several different points. We demonstrated the validity modeling a proton scanning nozzle system to evaluate various parameters using FLUKA. The measured secondary neutron ambient dose showed a similar tendency with the simulation result. This work will increase the knowledge necessary for the development of radiation safety technology in medical particle accelerators.

  4. Monte Carlo calculations of lung dose in ORNL phantom for boron neutron capture therapy

    International Nuclear Information System (INIS)

    Krstic, D.; Markovic, V.M.; Jovanovic, Z.; Milenkovic, B.; Nikezic, D.; Atanackovic, J.

    2014-01-01

    Monte Carlo simulations were performed to evaluate dose for possible treatment of cancers by boron neutron capture therapy (BNCT). The computational model of male Oak Ridge National Laboratory (ORNL) phantom was used to simulate tumours in the lung. Calculations have been performed by means of the MCNP5/X code. In this simulation, two opposite neutron beams were considered, in order to obtain uniform neutron flux distribution inside the lung. The obtained results indicate that the lung cancer could be treated by BNCT under the assumptions of calculations. The difference in evaluated dose in cancer and normal lung tissue suggests that BNCT could be applied for the treatment of cancers. The difference in exposure of cancer and healthy tissue can be observed, so the healthy tissue can be spared from damage. An absorbed dose ratio of metastatic tissue-to-the healthy tissue was ∼5. Absorbed dose to all other organs was low when compared with the lung dose. Absorbed dose depth distribution shows that BNC therapy can be very useful in the treatments for tumour. The ratio of the tumour absorbed dose and irradiated healthy tissue absorbed dose was also ∼5. It was seen that an elliptical neutron field was better irradiation choice. (authors)

  5. Antiproton Radiotherapy Peripheral Dose from Secondary Neutrons produced in the Annihilation of Antiprotons in the Target

    CERN Document Server

    Fahimian, Benjamin P; Keyes, Roy; Bassler, Niels; Iwamoto, Keisuke S; Zankl, Maria; Holzscheiter, Michael H

    2009-01-01

    The AD-4/ACE collaboration studies the biological effects of antiprotons with respect to a possible use of antiprotons in cancer therapy. In vitro experiments performed by the collaboration have shown an enhanced biological effectiveness for antiprotons relative to protons. One concern is the normal tissue dose resulting from secondary neutrons produced in the annihilation of antiprotons on the nucleons of the target atoms. Here we present the first organ specific Monte Carlo calculations of normal tissue equivalent neutron dose in antiproton therapy through the use of a segmented CT-based human phantom. The MCNPX Monte Carlo code was employed to quantify the peripheral dose for a cylindrical spread out Bragg peak representing a treatment volume of 1 cm diameter and 1 cm length in the frontal lobe of a segmented whole-body phantom of a 38 year old male. The secondary neutron organ dose was tallied as a function of energy and organ.

  6. Neutron production and dose rate in the IFMIF/EVEDA LIPAc injector beam commissioning

    Energy Technology Data Exchange (ETDEWEB)

    Kondo, Keitaro, E-mail: kondo.keitaro@jaea.go.jp [Rokkasho Fusion Institute, Japan Atomic Energy Agency, Rokkasho-mura, Kamikita-gun, Aomori (Japan); Narita, Takahiro; Usami, Hiroki; Takahashi, Hiroki; Ochiai, Kentaro; Shinto, Katsuhiro; Kasugai, Atsushi [Rokkasho Fusion Institute, Japan Atomic Energy Agency, Rokkasho-mura, Kamikita-gun, Aomori (Japan); Okumura, Yoshikazu [IFMIF/EVEDA Project Team, Rokkasho-mura, Kamikita-gun, Aomori (Japan)

    2016-11-01

    Highlights: • A dedicated neutron production yield monitoring system for LIPAc has been developed. • The biological dose rate during operation of the LIPAc injector was analyzed. • The neutron streaming effect due to penetrations in the shielding wall was investigated. - Abstract: The construction of the Linear IFMIF Prototype Accelerator (LIPAc) is in progress in Rokkasho, Japan, and the deuteron beam commissioning of the injector began in July 2015. Due to the huge beam current of 125 mA, a large amount of d-D neutrons are produced in the commissioning. The neutron streaming effect through pipe penetrations and underground pits may dominate the radiation dose at the outside of the accelerator vault during the injector operation. In the present study the effective dose rate expected during the injector commissioning was analyzed by a Monte Carlo calculation and compared with the measured value. For the comparison it is necessary to know the total neutron production yield in the accelerator vault, thus a dedicated neutron production yield monitoring system was developed. The yield obtained was smaller than that previously reported in a literature by a factor of a few and seems to depend on some beam conditions. From the comparison it was proved that the calculation always provides a conservative estimate and the dose rates in places where occupational works can always access and the controlled area boundary are expected to be far less than the legal criteria throughout the injector commissioning.

  7. A new online detector for estimation of peripheral neutron equivalent dose in organ

    Energy Technology Data Exchange (ETDEWEB)

    Irazola, L., E-mail: leticia@us.es; Sanchez-Doblado, F. [Departamento de Fisiología Médica y Biofísica, Universidad de Sevilla, Sevilla 41009, Spain and Servicio de Radiofísica, Hospital Universitario Virgen Macarena, Sevilla 41007 (Spain); Lorenzoli, M.; Pola, A. [Departimento di Ingegneria Nuclear, Politecnico di Milano, Milano 20133 (Italy); Bedogni, R. [Laboratori Nazionali di Frascati, Istituto Nazionale di Fisica Nucleare (INFN), Frascati Roma 00044 (Italy); Terrón, J. A. [Servicio de Radiofísica, Hospital Universitario Virgen Macarena, Sevilla 41007 (Spain); Sanchez-Nieto, B. [Instituto de Física, Pontificia Universidad Católica de Chile, Santiago 4880 (Chile); Expósito, M. R. [Departamento de Física, Universitat Autònoma de Barcelona, Bellaterra 08193 (Spain); Lagares, J. I.; Sansaloni, F. [Centro de Investigaciones Energéticas y Medioambientales y Tecnológicas (CIEMAT), Madrid 28040 (Spain)

    2014-11-01

    Purpose: Peripheral dose in radiotherapy treatments represents a potential source of secondary neoplasic processes. As in the last few years, there has been a fast-growing concern on neutron collateral effects, this work focuses on this component. A previous established methodology to estimate peripheral neutron equivalent doses relied on passive (TLD, CR39) neutron detectors exposed in-phantom, in parallel to an active [static random access memory (SRAMnd)] thermal neutron detector exposed ex-phantom. A newly miniaturized, quick, and reliable active thermal neutron detector (TNRD, Thermal Neutron Rate Detector) was validated for both procedures. This first miniaturized active system eliminates the long postprocessing, required for passive detectors, giving thermal neutron fluences in real time. Methods: To validate TNRD for the established methodology, intrinsic characteristics, characterization of 4 facilities [to correlate monitor value (MU) with risk], and a cohort of 200 real patients (for second cancer risk estimates) were evaluated and compared with the well-established SRAMnd device. Finally, TNRD was compared to TLD pairs for 3 generic radiotherapy treatments through 16 strategic points inside an anthropomorphic phantom. Results: The performed tests indicate similar linear dependence with dose for both detectors, TNRD and SRAMnd, while a slightly better reproducibility has been obtained for TNRD (1.7% vs 2.2%). Risk estimates when delivering 1000 MU are in good agreement between both detectors (mean deviation of TNRD measurements with respect to the ones of SRAMnd is 0.07 cases per 1000, with differences always smaller than 0.08 cases per 1000). As far as the in-phantom measurements are concerned, a mean deviation smaller than 1.7% was obtained. Conclusions: The results obtained indicate that direct evaluation of equivalent dose estimation in organs, both in phantom and patients, is perfectly feasible with this new detector. This will open the door to an

  8. Dose-effect relationship of apoptosis induced by fission-neutron in murine thymocytes

    International Nuclear Information System (INIS)

    Yuan Bin; Li Liang; Xue Wencheng; Sun Jianmin; Wang Baoqin

    2000-01-01

    Objective: To investigate the effectiveness of high LET fission-neutron to induce apoptosis in murine thymocytes and to compare it with that of low LET 60 Co γ-ray. Methods: Apoptosis induction was studied qualitatively by light and transmission electron microscopy and DNA gel electrophoresis,also quantitatively by flow cytometry(FCM) and diphenylamine (DPA)methods. Results: DNA ladders of murine thymocytes were detectable, the typical apoptosis of thymocytes could be observed morphologically by means of light and electron microscopy at 6 h after fission-neutron irradiation with doses ranging from 0.5 to 5.0 Gy, meanwhile the percentages of apoptosis increased with increasing doses. After exposure to γ-rays with doses ranging from 1.0 to 30 Gy, the experimental results were similar to those from neutron radiation. The incidence of apoptosis peaked at about 20 Gy, the percentages did not increase further when doses increased. Conclusion: Apoptosis of murine thymocytes can be induced when mice are exposed to either fission-neutron (0.5-5.0 Gy) or to γ-ray (1-30 Gy). Although the relationship between apoptosis and radiation doses is similar, the percentage of apoptosis induced by neutron irradiation is higher than that induced by γ-irradiation. The RBE values of fission-neutron for inducing apoptosis murine thymocytes are 2.09 (by FCM method) and 2.37 (by DPA method), respectively. These results also suggest that fission-neutron-induced murine immune tissue is more severe than that induced by γ-rays at several hours post-irradiation and this might be the basis for heavy damage to immune tissues induced by fission-neutron-irradiation in later period

  9. A new online detector for estimation of peripheral neutron equivalent dose in organ

    International Nuclear Information System (INIS)

    Irazola, L.; Sanchez-Doblado, F.; Lorenzoli, M.; Pola, A.; Bedogni, R.; Terrón, J. A.; Sanchez-Nieto, B.; Expósito, M. R.; Lagares, J. I.; Sansaloni, F.

    2014-01-01

    Purpose: Peripheral dose in radiotherapy treatments represents a potential source of secondary neoplasic processes. As in the last few years, there has been a fast-growing concern on neutron collateral effects, this work focuses on this component. A previous established methodology to estimate peripheral neutron equivalent doses relied on passive (TLD, CR39) neutron detectors exposed in-phantom, in parallel to an active [static random access memory (SRAMnd)] thermal neutron detector exposed ex-phantom. A newly miniaturized, quick, and reliable active thermal neutron detector (TNRD, Thermal Neutron Rate Detector) was validated for both procedures. This first miniaturized active system eliminates the long postprocessing, required for passive detectors, giving thermal neutron fluences in real time. Methods: To validate TNRD for the established methodology, intrinsic characteristics, characterization of 4 facilities [to correlate monitor value (MU) with risk], and a cohort of 200 real patients (for second cancer risk estimates) were evaluated and compared with the well-established SRAMnd device. Finally, TNRD was compared to TLD pairs for 3 generic radiotherapy treatments through 16 strategic points inside an anthropomorphic phantom. Results: The performed tests indicate similar linear dependence with dose for both detectors, TNRD and SRAMnd, while a slightly better reproducibility has been obtained for TNRD (1.7% vs 2.2%). Risk estimates when delivering 1000 MU are in good agreement between both detectors (mean deviation of TNRD measurements with respect to the ones of SRAMnd is 0.07 cases per 1000, with differences always smaller than 0.08 cases per 1000). As far as the in-phantom measurements are concerned, a mean deviation smaller than 1.7% was obtained. Conclusions: The results obtained indicate that direct evaluation of equivalent dose estimation in organs, both in phantom and patients, is perfectly feasible with this new detector. This will open the door to an

  10. Elaboration of a nodal method to solve the steady state multigroup diffusion equation. Study and use of the multigroup diffusion code DAHRA

    International Nuclear Information System (INIS)

    Halilou, A.; Lounici, A.

    1981-01-01

    The subject is divided in two parts: In the first part a nodal method has been worked out to solve the steady state multigroup diffusion equation. This method belongs to the same set of nodal methods currently used to calculate the exact fission powers and neutron fluxes in a very short computing time. It has been tested on a two dimensional idealized reactors. The effective multiplication factor and the fission powers for each fuel element have been calculated. The second part consists in studying and mastering the multigroup diffusion code DAHRA - a reduced version of DIANE - a two dimensional code using finite difference method

  11. Research of the application of multi-group libraries based on ENDF/B-VII library in the reactor design

    International Nuclear Information System (INIS)

    Mi Aijun; Li Junjie

    2010-01-01

    In this paper the multi-group libraries were constructed by processing ENDF/B-VII neutron incident files into multi-group structure, and the application of the multi-group libraries in the pressurized-water reactor(PWR) design was studied. The construction of the multi-group library is realized by using the NJOY nuclear data processing system. The code can process the neutron cross section files form ENDF format to MATXS format which was required in SN code. Two dimension transport theory code of discrete ordinates DORT was used to verify the multi-group libraries and the method of the construction by comparing calculations for some representative benchmarks. We made the PWR shielding calculation by using the multi-group libraries and studied the influence of the parameters involved during the construction of the libraries such as group structure, temperatures and weight functions on the shielding design of the PWR. This work is the preparation for the construction of the multi-group library which will be used in PWR shielding design in engineering. (authors)

  12. Development and testing of multigroup library with correction of self-shielding effects in fusion-fission hybrid reactor

    International Nuclear Information System (INIS)

    Zou Jun; He Zhaozhong; Zeng Qin; Qiu Yuefeng; Wang Minghuang

    2010-01-01

    A multigroup library HENDL2.1/SS (Hybrid Evaluated Nuclear Data Library/Self-Shielding) based on ENDF/B-VII.0 evaluate data has been generated using Bondarenko and flux calculator method for the correction of self-shielding effect of neutronics analyses. To validate the reliability of the multigroup library HENDL2.1/SS, transport calculations for fusion-fission hybrid system FDS-I were performed in this paper. It was verified that the calculations with the HENDL2.1/SS gave almost the same results with MCNP calculations and were better than calculations with the HENDL2.0/MG which is another multigroup library without self-shielding correction. The test results also showed that neglecting resonance self-shielding caused underestimation of the K eff , neutron fluxes and waste transmutation ratios in the multigroup calculations of FDS-I.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-05-01

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

  14. Dose-equivalent response CR-39 track detector for personnel neutron dosimetry

    International Nuclear Information System (INIS)

    Oda, K.; Ito, M.; Yoneda, H.; Miyake, H.; Yamamoto, J.; Tsuruta, T.

    1991-01-01

    A dose-equivalent response detector based on CR-39 has been designed to be applied for personnel neutron dosimetry. The intrinsic detection efficiency of bare CR-39 was first evaluated from irradiation experiments with monoenergetic neutrons and theoretical calculations. In the second step, the radiator effect was investigated for the purpose of sensitization to fast neutrons. A two-layer radiator consisting of deuterized dotriacontane (C 32 D 66 ) and polyethylene (CH 2 ) was designed. Finally, we made the CR-39 detector sensitive to thermal neutrons by doping with orthocarbone (B 10 H 12 C 2 ), and also estimated the contribution of albedo neutrons. It was found that the new detector - boron-doped CR-39 with the two-layer radiator - would have a flat response with an error of about 70% in a wide energy region, ranging from thermal to 15 MeV. (orig.)

  15. ACDOS2: a code for neutron-induced activities and dose rates

    International Nuclear Information System (INIS)

    Ruby, L.; Keney, G.S.; Lagache, J.C.

    1981-10-01

    In order to anticipate problems from the radioactivation of neutral beam sources as a result of testing, a code has been developed which calculates both the radioactivities produced and the dose rates resulting therefrom. The code ACDOS2 requires neutron source strength and spectral distribution as input, or alternately, the source strength can be calculated internally from an input of neutral beam source parameters. A variety of simple geometries can be specified, and up to 12 times of interest following the shutdown of the neutron source. Radiation attenuating and daughter radioactivities are treated accurately. ACDOS2 is also of use for neutron-induced radioactivation problems involving accelerators, fusion reactors, or fission reactors

  16. American National Standard: neutron and gamma-ray flux-to-dose rate factors

    International Nuclear Information System (INIS)

    Anon.

    1977-01-01

    This Standard presents data recommended for computing biological dose rates due to neutron and gamma-ray radiation fields. Neutron flux-to-dose-rate conversion factors for energies from 2.5 x 10 -8 to 20 MeV are given; the energy range for the gamma-ray conversion factors is 0.01 to 15 MeV. Specifically, this Standard is intended for use by shield designers to calculate wholebody dose rates to radiation workers and the general public. Establishing dose-rate limits is outside the scope of this Standard. Use of this Standard in cases where the dose equivalents are far in excess of occupational exposure guidelines is not recommended

  17. Dose equivalent response of personal neutron dosemeters as a function of angle

    International Nuclear Information System (INIS)

    Tanner, J.E.; McDonald, J.C.; Stewart, R.D.; Wernli, C.

    1997-01-01

    The measured and calculated dose equivalent response as a function of angle has been examined for an albedo-type thermoluminescence dosemeter (TLD) that was exposed to unmoderated and D 2 O-moderated 252 Cf neutron sources while mounted on a 40 x 40 15 cm 3 polymethylmethacrylate phantom. The dosemeter used in this study is similar to many neutron personal dosemeters currently in use. The detailed construction of the dosemeter was modelled, and the dose equivalent response was calculated, using the MCNP code. Good agreement was found between the measured and calculated values of the relative dose equivalent angular response for the TLD albedo dosemeter. The relative dose equivalent angular response was also compared with the values of directional and personal dose equivalent as a function of angle published by Siebert and Schuhmacher. (author)

  18. Optimization of a neutron ambient dose equivalent rate meter

    International Nuclear Information System (INIS)

    Burgkhardt, B.; Fieg, G.; Piesch, E.; Klett, A.; Maushart, R.

    1994-01-01

    Co-operating in a technology transfer project, the Karlsruhe Nuclear Research Center and EG and G Berthold have developed a neutron equivalent doserate probe with high sensitivity and an energy dependent detection efficiency in accordance with the ICRP60 requirements. The special features of this probe were realized, on the one hand, by optimizing the moderator and absorber geometry through simulation calculations with the neutron transport code MCNP, and, on the other hand, by using a newly designed 3 He-methane proportional counter tube. The measurements were not yet completed when this paper went to press, however, it is to be expected that the response sensitivity will be more than 3 counts/nSv. (orig.) [de

  19. Estimation of low-level neutron dose-equivalent rate by using extrapolation method for a curie level Am–Be neutron source

    International Nuclear Information System (INIS)

    Li, Gang; Xu, Jiayun; Zhang, Jie

    2015-01-01

    Neutron radiation protection is an important research area because of the strong radiation biological effect of neutron field. The radiation dose of neutron is closely related to the neutron energy, and the connected relationship is a complex function of energy. For the low-level neutron radiation field (e.g. the Am–Be source), the commonly used commercial neutron dosimeter cannot always reflect the low-level dose rate, which is restricted by its own sensitivity limit and measuring range. In this paper, the intensity distribution of neutron field caused by a curie level Am–Be neutron source was investigated by measuring the count rates obtained through a 3 He proportional counter at different locations around the source. The results indicate that the count rates outside of the source room are negligible compared with the count rates measured in the source room. In the source room, 3 He proportional counter and neutron dosimeter were used to measure the count rates and dose rates respectively at different distances to the source. The results indicate that both the count rates and dose rates decrease exponentially with the increasing distance, and the dose rates measured by a commercial dosimeter are in good agreement with the results calculated by the Geant4 simulation within the inherent errors recommended by ICRP and IEC. Further studies presented in this paper indicate that the low-level neutron dose equivalent rates in the source room increase exponentially with the increasing low-energy neutron count rates when the source is lifted from the shield with different radiation intensities. Based on this relationship as well as the count rates measured at larger distance to the source, the dose rates can be calculated approximately by the extrapolation method. This principle can be used to estimate the low level neutron dose values in the source room which cannot be measured directly by a commercial dosimeter. - Highlights: • The scope of the affected area for

  20. Sequential measurements of spectrum and dose for cosmic-ray neutrons on the ground

    International Nuclear Information System (INIS)

    Hirabayashi, N.; Nunomiya, T.; Suzuki, H.; Nakamura, T.

    2002-01-01

    The earth is continually bathed in high-energy particles that come from outside the solar system, known as galactic cosmic rays. When these particles penetrate the magnetic fields of the solar system and the Earth and reach the Earth's atmosphere, they collide with atomic nuclei in air and secondary cosmic rays of every kind. On the other hand, levels of accumulation of the semiconductor increase recently, and the soft error that the cosmic-ray neutrons cause has been regarded as questionable. There have been long-term measurements of cosmic-ray neutron fluence at several places in the world, but no systematic study on cosmic-ray neutron spectrum measurements. This study aimed to measure the cosmic-ray neutron spectrum and dose on the ground during the solar maximum period of 2000 to 2002. Measurements have been continuing in a cabin of Tohoku University Kawauchi campus, by using five multi-moderator spectrometers (Bonner sphere), 12.7 cm diam by 12.7 cm long NE213 scintillator, and rem counter. The Bonner sphere uses a 5.08 cm diam spherical 3 He gas proportional counter and the rem counter uses a 12.7 cm diam 3 He gas counter. The neutron spectra were obtained by unfolding from the count rates measured with the Bonner sphere using the SAND code and the pulse height spectra measured with the NE213 scintillator using the FORIST code . The cosmic- ray neutron spectrum and ambient dose rates have been measured sequentially from April 2001. Furthermore, the correlation between ambient dose rate and the atmospheric pressure was investigated with a barometer. We are also very much interested in the variation of neutron spectrum following big solar flares. From the sequential measurements, we found that the cosmic-ray neutron spectrum has two peaks at around 1 MeV and at around 100 MeV, and the higher energy peak increases with a big solar flare

  1. Dose distributions in thorax inhomogeneity for fast neutron beam from NIRS cyclotron

    International Nuclear Information System (INIS)

    Kutsutani-Nakamura, Yuzuru; Furukawa, Shigeo; Iinuma, T.A.; Kawashima, Katsuhiro; Hoshino, Kazuo; Hiraoka, Takeshi; Maruyama, Takashi; Sakashita, Kunio; Tsunemoto, Hiroshi

    1990-01-01

    The power law tissue-air ratio (TAR) method developed by Batho appears to be practical use for inhomogeneity corrections to the dose calculated in a layered media for photon beam therapy. The validity was examined in applying the modified power law TAR and the isodose shift methods to the dose calculation in thorax tissue inhomogeneity containing the boundary region for fast neutron beam. The neutron beam is produced by bombarding a thick beryllium target with 30 MeV deuterons. Lung phantom was made of granulated tissue equivalent plastic, which resulted in density of 0.30 and 0.60 g/cm 3 . Depth dose distributions for neutron beam were measured in thorax phantom by an air-filled cylindrical ionization chamber with TE plastic wall. The power law TAR method considering TAR of zero depth at boundary was compared with the measured data and a good result was obtained that the calculated dose was within ±3 % against the measured. But the isodose shift method is not so good for dose calculation in thorax tissue inhomogeneity using fast neutron beam. (author)

  2. 6LiF sandwich type detectors for low dose individual monitoring in mixed neutron-photon fields

    International Nuclear Information System (INIS)

    Olko, P.; Budzanowski, M.; Bilski, P.; Burgkhardt, B.; Piesch, E.

    1994-01-01

    ICRP Publication 60 recommends the reduction of the annual dose limit for occupational exposure from 50 to 20 mSv and a doubling of the quality factor for medium energy neutrons. If occupational doses are evaluated every month (which is obligatory e.g. in Germany and in Poland), the individual neutron dosemeter will have to measure neutron doses in the range of 100 μSv. No commercially available, automatic individual dosimetry monitoring system exists that fulfils this requirement. Some of the parameters which influence the evaluation of the neutron dose from readings of TL dosemeters have been studied in order to decrease the variance of the measured neutron signal. In mixed neutron-photon fields, clear separation of the neutron component from the total reading depends also on the uncertainty of the gamma dose measurements. While the thermal albedo neutrons are absorbed mostly at the surface of the 6 LiF detector, the reduction of the detector thickness results in a decrease of its photon sensitivity, while its neutron sensitivity is almost principally maintained. As a consequence, the uncertainty of gamma dose contributes with lower weight to the variance of the evaluated neutron signal. First tests of an optimised 200 μm thick sandwich detector and 0.9 mm thick standard LiF chips were made at low neutron and photon dose ranges using different readers, in order to determine the uncertainty versus dose for different neutron-photon combinations. The conditions under which the new sandwich type detectors may improve albedo neutron dosimetry are demonstrated. (Author)

  3. Life shortening, tumor induction, and tissue dose for fission-neutron and gamma-ray irradiations

    International Nuclear Information System (INIS)

    Grahn, D.; Duggal, K.; Lombard, L.S.

    1985-01-01

    The primary focus of this program is to obtain information on the late effects of whole body exposure to low doses of a high linear-energy-transfer (LET) and a low-LET ionizing radiation in experimental animals to provide guidance for the prediction of radiation hazards to man. The information obtained takes the form of dose-response curves for life shortening and for the induction of numerous specific types of tumors. The animals are irradiated with fission neutrons from the Janus reactor and with 60 Co gamma rays, delivered as single, weekly, or duration-of-life exposures covering the range of doses and dose rates. 6 refs

  4. Determination of dose components in mixed gamma neutron fields by use of high pressure ionization chambers

    International Nuclear Information System (INIS)

    Golnik, N.; Pliszczynski, T.; Wysocka, A.; Zielczynski, M.

    1985-01-01

    The two ionization chamber method for determination of dose components in mixed γ-neutron field has been improved by increasing gas pressure in the chambers up to some milions pascals. Advantages of high pressure gas filling are the followings: 1) significant reduction of the ratio of neutron-to gamma sensitivity for the hydrogen-free chamber, 2) possibility of sensitivity correction for both chambers by application of appropriate voltage, 3) high sensitivity for small detectors. High-pressure, pen-like ionization chambers have been examined in fields of different neutron sources: a TE-chamber, filled with 0.2 MPa of quasi-TE-gas and a conductive PTFE chamber, filled with 3.1 MPa of CO 2 . The ratio of neutron-to-gamma sensitivity for the PTFE chamber, operated at electrical field strength below 100 V/cm, has not exceeded 0.01 for neutrons with energy below 8 MeV. Formula is presented for calculation of this ratio for any high-pressure, CO 2 -filled ionization chamber. Contribution of gamma component to total tissue dose in the field of typical neutron sources has been found to be 3 to 70%

  5. Evaluation of energy responses for neutron dose-equivalent meters made in Japan

    International Nuclear Information System (INIS)

    Saegusa, J.; Yoshizawa, M.; Tanimura, Y.; Yoshida, M.; Yamano, T.; Nakaoka, H.

    2004-01-01

    Energy responses of three types of Japanese neutron dose-equivalent (DE) meters were evaluated by Monte Carlo simulations and measurements. The energy responses were evaluated for thermal neutrons, monoenergetic neutrons with energies up to 15.2 MeV, and also for neutrons from such radionuclide sources as 252 Cf and 241 Am-Be. The calculated results were corroborated with the measured ones. The angular dependence of the response and the DE response were also evaluated. As a result, reliable energy responses were obtained by careful simulations of the proportional counter, moderator and absorber of the DE meters. Furthermore, the relationship between pressure of counting gas and response of the DE meter was discussed. By using the obtained responses, relations between predicted readings of the DE meters and true DE values were studied for various workplace spectra

  6. Comparison of Out-Of-Field Neutron Equivalent Doses in Scanning Carbon and Proton Therapies for Cranial Fields

    DEFF Research Database (Denmark)

    Athar, B.; Henker, K.; Jäkel, O.

    2010-01-01

    Purpose: The purpose of this analysis is to compare the secondary neutron lateral doses from scanning carbon and proton beam therapies. Method and Materials: We simulated secondary neutron doses for out-of-field organs in an 11-year old male patient. Scanned carbon and proton beams were simulated...

  7. Limiting values for the RBE of fission neutrons at low doses for life shortening in mice

    International Nuclear Information System (INIS)

    Storer, J.B.; Mitchell, T.J.

    1984-01-01

    The authors have analyzed recently published data on the effects of low doses of fission neutrons on the mean survival times of mice. The analysis for single-dose exposures was confined to doses of 20 rad or less, while for fractionated exposures only total doses of 80 rad or less were considered. They fitted the data to the frequently used power function model: life shortening = βD/sup γ/, where D is the radiation dose. They show that, at low doses per fraction, either the effects are not additive or the dose-effect curve for single exposures cannot show a greater negative curvature than about the 0.9 power of dose. Analysis of the data for γ rays showed that an exponent of 1.0 gave an acceptable fit. They conclude that at neutron doses of 20 rad or less the RBE for life shortening is constant and ranges from 13 to 22 depending on mouse strain and sex

  8. High dose effect of gamma and neutrons on the N-JFET electronic components

    International Nuclear Information System (INIS)

    Assaf, Jamal-Eddin

    2006-11-01

    Two types of N-JFET components have been irradiated by high doses of thermal neutrons and gamma rays up to 2000x10 12 n/cm 2 and 1000 kGy, respectively. The static tests show a decrease of the g m and I d s parameters. The behaviour of electronic noise on the output was the principal dynamic test after irradiation. The result of this test gives an increase of the noise with radiation dose increasing. The noise was described as the Equivalent Noise of Charge (ENC) at the output of the measurements set-up. The quantities and the qualities of the noise depend on the N-JEET type and the type of radiation (neutrons or gamma). Other tests were carried out like the relaxation or recovery phenomena after radiation, and the superposed effects of gamma and neutrons.(author)

  9. Neutron and gamma-ray dose-rates from the Little Boy replica

    International Nuclear Information System (INIS)

    Plassmann, E.A.; Pederson, R.A.

    1984-01-01

    We report dose-rate information obtained at many locations in the near vicinity of, and at distances out to 0.64 km from, the Little Boy replica while it was operated as a critical assembly. The measurements were made with modified conventional dosimetry instruments that used an Anderson-Braun detector for neutrons and a Geiger-Mueller tube for gamma rays with suitable electronic modules to count particle-induced pulses. Thermoluminescent dosimetry methods provide corroborative data. Our analysis gives estimates of both neutron and gamma-ray relaxation lengths in air for comparison with earlier calculations. We also show the neutron-to-gamma-ray dose ratio as a function of distance from the replica. Current experiments and further data analysis will refine these results. 7 references, 8 figures

  10. Approximate albedo boundary conditions for energy multigroup X,Y-geometry discrete ordinates nuclear global calculations

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Davi J.M.; Nunes, Carlos E.A.; Alves Filho, Hermes; Barros, Ricardo C., E-mail: davijmsilva@yahoo.com.br, E-mail: ceanunes@yahoo.com.br, E-mail: rcbarros@pq.cnpq.br [Secretaria Municipal de Educacao de Itaborai, RJ (Brazil); Universidade Estacio de Sa (UNESA), Rio de Janeiro, RJ (Brazil); Universidade do Estado do Rio de Janeiro (UERJ), Novra Friburgo, RJ (Brazil). Instituto Politecnico. Departamento de Modelagem Computacional

    2017-11-01

    Discussed here is the accuracy of approximate albedo boundary conditions for energy multigroup discrete ordinates (S{sub N}) eigenvalue problems in two-dimensional rectangular geometry for criticality calculations in neutron fission reacting systems, such as nuclear reactors. The multigroup (S{sub N}) albedo matrix substitutes approximately the non-multiplying media around the core, e.g., baffle and reflector, as we neglect the transverse leakage terms within these non-multiplying regions. Numerical results to a typical model problem are given to illustrate the accuracy versus the computer running time. (author)

  11. Multigroup cross section collapsing optimization of a He-3 detector assembly model using deterministic transport techniques

    International Nuclear Information System (INIS)

    Huang, Mi; Yi, Ce; Manalo, Kevin L.; Sjoden, Glenn E.

    2011-01-01

    Multigroup optimization is performed on a neutron detector assembly to examine the validity of transport response in forward and adjoint modes. For SN transport simulations, we discuss the multigroup collapse of an 80 group library to 40, 30, and 16 groups, constructed from using the 3-D parallel PENTRAN and macroscopic cross section collapsing with YGROUP contribution weighting. The difference in using P_1 and P_3 Legendre order in scattering cross sections is investigated; also, associated forward and adjoint transport responses are calculated. We conclude that for the block analyzed, a 30 group cross section optimizes both computation time and accuracy relative to the 80 group transport calculations. (author)

  12. Multi-level methods for solving multigroup transport eigenvalue problems in 1D slab geometry

    International Nuclear Information System (INIS)

    Anistratov, D. Y.; Gol'din, V. Y.

    2009-01-01

    A methodology for solving eigenvalue problems for the multigroup neutron transport equation in 1D slab geometry is presented. In this paper we formulate and compare different variants of nonlinear multi-level iteration methods. They are defined by means of multigroup and effective one-group low-order quasi diffusion (LOQD) equations. We analyze the effects of utilization of the effective one-group LOQD problem for estimating the eigenvalue. We present numerical results to demonstrate the performance of the iteration algorithms in different types of reactor-physics problems. (authors)

  13. Multi-level nonlinear diffusion acceleration method for multigroup transport k-Eigenvalue problems

    International Nuclear Information System (INIS)

    Anistratov, Dmitriy Y.

    2011-01-01

    The nonlinear diffusion acceleration (NDA) method is an efficient and flexible transport iterative scheme for solving reactor-physics problems. This paper presents a fast iterative algorithm for solving multigroup neutron transport eigenvalue problems in 1D slab geometry. The proposed method is defined by a multi-level system of equations that includes multigroup and effective one-group low-order NDA equations. The Eigenvalue is evaluated in the exact projected solution space of smallest dimensionality, namely, by solving the effective one- group eigenvalue transport problem. Numerical results that illustrate performance of the new algorithm are demonstrated. (author)

  14. Effects of split fast neutron doses on the liver cells of albino Swiss mice

    International Nuclear Information System (INIS)

    Abdelmeguid, N.; Ramadan, A.A.; El-Khatib, A.M.

    1990-01-01

    The effect of neutron doses from a compact D-T neutron generator on the liver cells of adult male and female albino Swiss mice was investigated. Fast neutrons (14.5 MeV) were delivered to the whole body in a single dose or in two, four, six or eight equal doses separated by 3-day intervals. The lowest dose, 100 rem, was given over an exposure time of 6 hours and was then steadily raised to 912 rem over an exposure time of 48 hours. During exposure the neutron flux was controlled by the activation foil technique. The animals were killed for testing after each irradiation. Histological examination of the hepatocytes with a light microscope showed marked degenerative changes only after the longer irradiation periods (24, 36 and 48 h). Electron microscopy showed cytological (cytoplasmic and nuclear) changes in the hepatocytes after only 12 hours' irradiation. Densitometric scans of electron micrographs of control and 12 h-irradiated livers indicated that the control hepatocyte interphase nucleus contains approximately 72% heterochromatin, while the irradiated nucleus contains only 64% heterochromatin. (author). 13 figs., 1 tab., 18 refs

  15. ACDOS2: an improved neutron-induced dose rate code

    International Nuclear Information System (INIS)

    Lagache, J.C.

    1981-06-01

    To calculate the expected dose rate from fusion reactors as a function of geometry, composition, and time after shutdown a computer code, ACDOS2, was written, which utilizes up-to-date libraries of cross-sections and radioisotope decay data. ACDOS2 is in ANSI FORTRAN IV, in order to make it readily adaptable elsewhere

  16. ACDOS2: an improved neutron-induced dose rate code

    Energy Technology Data Exchange (ETDEWEB)

    Lagache, J.C.

    1981-06-01

    To calculate the expected dose rate from fusion reactors as a function of geometry, composition, and time after shutdown a computer code, ACDOS2, was written, which utilizes up-to-date libraries of cross-sections and radioisotope decay data. ACDOS2 is in ANSI FORTRAN IV, in order to make it readily adaptable elsewhere.

  17. Evaluating secondary neutron doses of a refined shielded design for a medical cyclotron using the TLD approach

    International Nuclear Information System (INIS)

    Lin, Jye-Bin; Tseng, Hsien-Chun; Liu, Wen-Shan; Lin, Ding-Bang; Hsieh, Teng-San; Chen, Chien-Yi

    2013-01-01

    An increasing number of cyclotrons at medical centers in Taiwan have been installed to generate radiopharmaceutical products. An operating cyclotron generates immense amounts of secondary neutrons from reactions such the 18 O(p, n) 18 F, used in the production of FDG. This intense radiation can be hazardous to public health, particularly to medical personnel. To increase the yield of 18 F-FDG from 4200 GBq in 2005 to 48,600 GBq in 2011, Chung Shan Medical University Hospital (CSMUH) has prolonged irradiation time without changing the target or target current to meet requirements regarding the production 18 F. The CSMUH has redesigned the CTI Radioisotope Delivery System shield. The lack of data for a possible secondary neutron doses has increased due to newly designed cyclotron rooms. This work aims to evaluate secondary neutron doses at a CTI cyclotron center using a thermoluminescent dosimeter (TLD-600). Two-dimensional neutron doses were mapped and indicated that neutron doses were high as neutrons leaked through self-shielded blocks and through the L-shaped concrete shield in vault rooms. These neutron doses varied markedly among locations close to the H 2 18 O target. The Monte Carlo simulation and minimum detectable dose are also discussed and demonstrated the reliability of using the TLD-600 approach. Findings can be adopted by medical centers to identify radioactive hot spots and develop radiation protection. - Highlights: • Neutron doses were verified using TLD approach. • Neutron doses were increased at cyclotron centers. • Revised L-shaped shield suppresses effectively the neutrons. • Neutron dose can be attenuated to 1.13×10 6 %

  18. Neutron dose rate analysis on HTGR-10 reactor using Monte Carlo code

    Science.gov (United States)

    Suwoto; Adrial, H.; Hamzah, A.; Zuhair; Bakhri, S.; Sunaryo, G. R.

    2018-02-01

    The HTGR-10 reactor is cylinder-shaped core fuelled with kernel TRISO coated fuel particles in the spherical pebble with helium cooling system. The outlet helium gas coolant temperature outputted from the reactor core is designed to 700 °C. One advantage HTGR type reactor is capable of co-generation, as an addition to generating electricity, the reactor was designed to produce heat at high temperature can be used for other processes. The spherical fuel pebble contains 8335 TRISO UO2 kernel coated particles with enrichment of 10% and 17% are dispersed in a graphite matrix. The main purpose of this study was to analysis the distribution of neutron dose rates generated from HTGR-10 reactors. The calculation and analysis result of neutron dose rate in the HTGR-10 reactor core was performed using Monte Carlo MCNP5v1.6 code. The problems of double heterogeneity in kernel fuel coated particles TRISO and spherical fuel pebble in the HTGR-10 core are modelled well with MCNP5v1.6 code. The neutron flux to dose conversion factors taken from the International Commission on Radiological Protection (ICRP-74) was used to determine the dose rate that passes through the active core, reflectors, core barrel, reactor pressure vessel (RPV) and a biological shield. The calculated results of neutron dose rate with MCNP5v1.6 code using a conversion factor of ICRP-74 (2009) for radiation workers in the radial direction on the outside of the RPV (radial position = 220 cm from the center of the patio HTGR-10) provides the respective value of 9.22E-4 μSv/h and 9.58E-4 μSv/h for enrichment 10% and 17%, respectively. The calculated values of neutron dose rates are compliant with BAPETEN Chairman’s Regulation Number 4 Year 2013 on Radiation Protection and Safety in Nuclear Energy Utilization which sets the limit value for the average effective dose for radiation workers 20 mSv/year or 10μSv/h. Thus the protection and safety for radiation workers to be safe from the radiation source has

  19. Multi-Group Library Generation with Explicit Resonance Interference Using Continuous Energy Monte Carlo Calculation

    Energy Technology Data Exchange (ETDEWEB)

    Park, Ho Jin; Cho, Jin Young [KAERI, Daejeon (Korea, Republic of); Kim, Kang Seog [Oak Ridge National Laboratory, Oak Ridge (United States); Hong, Ser Gi [Kyung Hee University, Yongin (Korea, Republic of)

    2016-05-15

    In this study, multi-group cross section libraries for the DeCART code were generated using a new procedure. The new procedure includes generating the RI tables based on the MC calculations, correcting the effective fission product yield calculations, and considering most of the fission products as resonant nuclides. KAERI (Korea Atomic Energy Research Institute) has developed the transport lattice code KARMA (Kernel Analyzer by Ray-tracing Method for fuel Assembly) and DeCART (Deterministic Core Analysis based on Ray Tracing) for a multi-group neutron transport analysis of light water reactors (LWRs). These codes adopt the method of characteristics (MOC) to solve the multi-group transport equation and resonance fixed source problem, the subgroup and the direct iteration method with resonance integral tables for resonance treatment. With the development of the DeCART and KARMA code, KAERI has established its own library generation system for a multi-group transport calculation. In the KAERI library generation system, the multi-group average cross section and resonance integral (RI) table are generated and edited using PENDF (point-wise ENDF) and GENDF (group-wise ENDF) produced by the NJOY code. The new method does not need additional processing because the MC method can handle any geometry information and material composition. In this study, the new method is applied to the dominant resonance nuclide such as U{sup 235} and U{sup 238} and the conventional method is applied to the minor resonance nuclides. To examine the newly generated multi-group cross section libraries, various benchmark calculations such as pin-cell, FA, and core depletion problem are performed and the results are compared with the reference solutions. Overall, the results by the new method agree well with the reference solution. The new procedure based on the MC method were verified and provided the multi-group library that can be used in the SMR nuclear design analysis.

  20. Calculation of neutron and gamma-ray flux-to-dose-rate conversion factors

    International Nuclear Information System (INIS)

    Kwon, S.G.; Lee, S.Y.; Yook, C.C.

    1981-01-01

    This paper presents flux-to-dose-rate conversion factors for neutrons and gamma rays based on the American National Standard Institute (ANSI) N666. These data are used to calculate the dose rate distribution of neutron and gamma ray in radiation fields. Neutron flux-to-dose-rate conversion factors for energies from 2.5 x 10 -8 to 20 MeV are presented; the corresponding energy range for gamma rays is 0.01 to 15 MeV. Flux-to-dose-rate conversion factors were calculated, under the assumption that radiation energy distribution has nonlinearity in the phantom, have different meaning from those values obtained by monoenergetic radiation. Especially, these values were determined with the cross section library. The flux-to-dose-rate conversion factors obtained in this work were in a good agreement to the values presented by ANSI. Those data will be useful for the radiation shielding analysis and the radiation dosimetry in the case of continuous energy distributions. (author)

  1. Neutron/gamma dose separation by the multiple-ion-chamber technique

    International Nuclear Information System (INIS)

    Goetsch, S.J.

    1983-01-01

    Many mixed n/γ dosimetry systems rely on two dosimeters, one composed of a tissue-equivalent material and the other made from a non-hydrogenous material. The paired chamber technique works well in fields of neutron radiation nearly identical in spectral composition to that in which the dosimeters were calibrated. However, this technique is drastically compromised in phantom due to the degradation of the neutron spectrum. The three-dosimeter technique allows for the fall-off in neutron sensitivity of the two non-hydrogenous dosimeters. Precise and physically meaningful results were obtained with this technique with a D-T source in air and in phantom and with simultaneous D-T neutron and 60 Co gamma ray irradiation in air. The MORSE-CG coupled n/γ three-dimensional Monte Carlo code was employed to calculate neutron and gamma doses in a water phantom. Gamma doses calculated in phantom with this code were generally lower than corresponding ion chamber measurements. This can be explained by the departure of irradiation conditions from ideal narrow-beam geometry. 97 references

  2. The effect of low-dose neutron irradiation on extracellular matrix

    International Nuclear Information System (INIS)

    Chen Tiehe; Lu Yongjie; Chai Mingsheng; Peng Wulin; Yang Yifang; Pan Yan; Chen Jinguo

    2003-01-01

    Projective: To study the effect of neutron irradiation on extracellular matrix. Methods: 120 male wistar rats were divided into four groups at random, and then exposed to neutron of 252 Cf-source at the doses of 0, 0.29, 0.62 and 1.20 Gy, respectively. After the exposure of 3 days, 1 month and 2 months, the rats were sacrificed and lung tissue specimens stored at -30 degree C. Hyaluronan, laminin, type III procollagen and type IV collagen in the lung tissue were detected by the method of radioimmunoassay. Results: The differences of the levels of hyaluronan in lung tissue among the groups were unsignificant. The levels of laminin in 0.29, 0.62 and 1.20 Gy groups after the 3-day exposure were remarkably different to those of the control group, and unable to recover completely even 2 months after the exposure. The levels of type IV collagen in higher three irradiated groups were all higher, but not significantly. The levels of type III procollagen in the early stage after exposure were higher, and later they lowered. Conclusion: The levels of some components of extracellular matrix in the lung tissue of rat can be changed by low-dose of neutron irradiation, but their variational modes and degrees depend on the dose of neutron irradiation and the length of period after exposure

  3. A Laplace transform method for energy multigroup hybrid discrete ordinates

    International Nuclear Information System (INIS)

    Segatto, C.F.; Vilhena, M.T.; Barros, R.C.

    2010-01-01

    In typical lattice cells where a highly absorbing, small fuel element is embedded in the moderator, a large weakly absorbing medium, high-order transport methods become unnecessary. In this work we describe a hybrid discrete ordinates (S N) method for energy multigroup slab lattice calculations. This hybrid S N method combines the convenience of a low-order S N method in the moderator with a high-order S N method in the fuel. The idea is based on the fact that in weakly absorbing media whose physical size is several neutron mean free paths in extent, even the S 2 method (P 1 approximation), leads to an accurate result. We use special fuel-moderator interface conditions and the Laplace transform (LTS N ) analytical numerical method to calculate the two-energy group neutron flux distributions and the thermal disadvantage factor. We present numerical results for a range of typical model problems.

  4. Some thoughts on tolerance, dose, and fractionation in boron neutron capture therapy

    International Nuclear Information System (INIS)

    Gahbauer, R.; Goodman, J.; Blue, T.

    1988-01-01

    Unique to boron neutron capture therapy, the tolerance very strongly depends on the boron concentration in normal brain, skin and blood. If one first considers the ideal situation of a 2 KeV beam and a compound clearing from normal tissues and blood, the tolerance dose to epithermal beams relates to the maximum tolerated capture gamma dose and capture high LET dose, H (n,gamma)D and N(n,p) 14 C. The authors can relate this gamma and high LET dose to known clinical experience. Assuming gamma and high LET dose ratios as given by Fairchild and Bond, one may first choose a clearly safe high LET whole brain dose and calculate the unavoidably resulting gamma dose. To a first approximation 500 cGy of high LET dose results in 3,000 cGy gamma dose. One can speculate that this approximates the tolerance of whole brain to the 2 KeV beam with no contributing boron dose if the radiation is fractionated. It would clearly be beyond tolerance in a single fraction where most therapists would be uncomfortable to deliver even one third of the above doses

  5. Temperature and neutron dose rate measurements at a spent fuel shipping cask

    International Nuclear Information System (INIS)

    Krause, F.

    1982-01-01

    Apart from some other requirements, spent fuel shipping casks have to ensure sufficient heat removal and radiation shielding. Results of temperature and neutron dose rate measurements at a spent fuel shipping cask are presented for different loading and heat removal by air. The measurements show that in shipping higher burnup fuel assemblies neutron radiation has to be taken into account when estimating the shielding of the shipping cask. On the other hand, unallowable high temperatures have been observed neither at the fuel assemblies nor at the shipping cask for a maximum heat output of Q <= 12 kW. (author)

  6. Spectral correction factors for conventional neutron dose meters used in high-energy neutron environments improved and extended results based on a complete survey of all neutron spectra in IAEA-TRS-403

    International Nuclear Information System (INIS)

    Oparaji, U.; Tsai, Y. H.; Liu, Y. C.; Lee, K. W.; Patelli, E.; Sheu, R. J.

    2017-01-01

    This paper presents improved and extended results of our previous study on corrections for conventional neutron dose meters used in environments with high-energy neutrons (E n > 10 MeV). Conventional moderated-type neutron dose meters tend to underestimate the dose contribution of high-energy neutrons because of the opposite trends of dose conversion coefficients and detection efficiencies as the neutron energy increases. A practical correction scheme was proposed based on analysis of hundreds of neutron spectra in the IAEA-TRS-403 report. By comparing 252 Cf-calibrated dose responses with reference values derived from fluence-to-dose conversion coefficients, this study provides recommendations for neutron field characterization and the corresponding dose correction factors. Further sensitivity studies confirm the appropriateness of the proposed scheme and indicate that (1) the spectral correction factors are nearly independent of the selection of three commonly used calibration sources: 252 Cf, 241 Am-Be and 239 Pu-Be; (2) the derived correction factors for Bonner spheres of various sizes (6''-9'') are similar in trend and (3) practical high-energy neutron indexes based on measurements can be established to facilitate the application of these correction factors in workplaces. (authors)

  7. The LAW Library -- A multigroup cross-section library for use in radioactive waste analysis calculations

    International Nuclear Information System (INIS)

    Greene, N.M.; Arwood, J.W.; Wright, R.Q.; Parks, C.V.

    1994-08-01

    The 238-group LAW Library is a new multigroup neutron cross-section library based on ENDF/B-V data, with five sets of data taken from ENDF/B-VI ( 14 N 7 , 15 N 7 , 16 O 8 , 154Eu 63 , and 155 Eu 63 ). These five nuclides are included because the new evaluations are thought to be superior to those in Version 5. The LAW Library contains data for over 300 materials and will be distributed by the Radiation Shielding Information Center, located at Oak Ridge National Laboratory. It was generated for use in neutronics calculations required in radioactive waste analyses, although it has equal utility in any study requiring multigroup neutron cross sections

  8. A sensitivity study on neutron flux variation due to 10B concentration in dose calculation for BNCT

    International Nuclear Information System (INIS)

    Jung, Sang Hoon

    2006-02-01

    The effects of inclusion of 10 B concentration on neutron flux and dose in dose calculation were studied. In order to provide the quantitative effects of inclusion of 10 B concentrations on depressions of neutron and photon flux and dose, the fluxes and doses with voxel head phantoms for various 10 B concentrations homogeneously distributed were calculated by using MCNPX simulations. A lithium target system and beam shaping assembly, which have been developed at the Hanyang University, were used as epithermal neutron beam. The calculation results show that the neutron flux at the center of the head phantom decreases by approximately 5.4% per 10 ppm of 10 B concentration in comparison with the neutron flux in the case of boron-free. It was also observed that the tissue dose at the center of the head phantom is depressed by approximately 4.7% per 10 ppm of the 10 B concentration and the tumor dose by approximately 5.3% per 10 ppm. According to depth of tumors, it was observed that the depressions of the doses in the tumors are ranged in 3.7 ∼ 9.2%. The dose calculations in the case of boron-free show that it is overestimated in comparison with the dose calculations in the cases of the inclusion of 10 B concentrations for the normal tissue and the tumors. Therefore, in dose calculation for BNCT, the depressions of neutron flux and dose should be considered. The results in this study are available to setting up the depression ratios which can be used for converting neutron and gamma fluxes and doses in phantom with boron free into the fluxes and doses in phantom with inclusion of 10 B concentrations in treatment. It is expected that the depression ratios is practicable to dose evaluation for BNCT

  9. The effect of the neutron spectra unfolding method on the fast neutron dose determination

    International Nuclear Information System (INIS)

    Marinkovic, P.; Zavaljevski, N.

    1992-01-01

    Based on Shanon's information theory, a new unfolding method which gives non-negative spectrum values and a relatively low variance, is proposed, and a numerical code suitable for application in fast neutron spectroscopy based on proton recoil is developed. The principles of maximum entropy and maximum likelihood are jointly applied. According to the principle of maximum likelihood, the distribution functions around the mean value of the counts in the MCA channels are assumed to be Gaussians. The Lagrange parameter method is applied in the search for an optimal non-negative solution. The nonlinear system of equations is solved using the gradient and Newton iterative algorithms. (orig.)

  10. The effect of the neutron spectra unfolding method on the fast neutron dose determination

    International Nuclear Information System (INIS)

    Marinkovic, P.; Avdic, S.; Pesic, M.; Zavaljevski, N

    1992-09-01

    Based on Shanon's information theory, a new unfolding method which gives non-negative spectrum values and a relatively low variance, is proposed, and a numerical code suitable for application in fast neutron spectroscopy based on proton recoil is developed. The principles of maximum entropy and maximum likelihood are jointly applied. According to the principle of maximum likelihood, the distribution functions around the mean value of the counts in the MCA channels are assumed to be Gaussians. The Lagrange parameter method is applied in the search for an optimal non-negative solution. The nonlinear system of equations is solved using the gradient and Newton iterative algorithms. (author)

  11. Boron neutron capture irradiation of the rat spinal cord: effects of variable doses of borocaptate sodium

    International Nuclear Information System (INIS)

    Morris, Gerard M.; Coderre, Jeffrey A.; Hopewell, John W.; Micca, Peggy L.; Fisher, Craig

    1996-01-01

    The Fischer 344 rat spinal cord model has been used to evaluate the response of the central nervous system to boron neutron capture irradiation with variable doses of the neutron capture agent, borocaptate sodium (BSH). Three doses of BSH, 190, 140 and 80 mg/kg body weight, administered by i.p. injection, were used to establish the time course of 10 B accumulation in and removal from the blood. After administration of the two lower doses of BSH, blood 10 B levels peaked at 0.5 h after injection, with no significant (P > 0.1) change at 1 h after injection. Beyond this time point, levels of 10 B in the blood began to decrease after a dose of 80 mg/kg BSH, but remained constant until 3 h after administration after the two higher doses of BSH. Myelopathy developed after latent intervals of 20.4 ± 0.1, 20.8 ± 1.4, 15.0 ± 0.8, 15.4 ± 0.4 and 15.6 ± 0.4 weeks, following irradiation with thermal neutrons in combination with BSH at doses of 20, 40, 80, 140 and 190 mg/kg body weight, respectively. The radiation-induced lesion in the spinal cord was white matter necrosis. ED 50 values for myelopathy were calculated from probit-fitted dose-effect curves. Expressed as total physical absorbed doses, these values were 20.7 ± 1.9, 24.9 ± 1.2, 27.2 ± 0.9, 28.4 ± 0.6 and 32.4 ± 1.9 Gy after irradiation with thermal neutrons in the presence of 20, 40, 80, 140 and 190 mg/kg body weight of BSH, respectively. The compound biological effectiveness (CBE) factor values, estimated from this data, were in the range 0.49-0.55. There was no significant (P >0.1) variation in the CBE factor for BSH as a function of increasing 10 B concentration in the blood. It was concluded that there was no significant synergistic interaction between the low and high linear energy transfer (LET) components of the boron neutron capture (BNC) radiation field

  12. Effective dose evaluation for BNCT treatment in the epithermal neutron beam at THOR

    Energy Technology Data Exchange (ETDEWEB)

    Wang, J.N. [Department of Engineering and System Science, National Tsing Hua University, No. 101, Section 2, Kuang-Fu Rd., Hsinchu 30013, Taiwan (China)] [Division of Health Physics, Institute of Nuclear Energy Research, No. 1000, Wenhua Rd., Jiaan Village, Longtan Township, Taoyuan County 32546, Taiwan (China); Huang, C.K. [Institute of Nuclear Engineering and Science, National Tsing Hua University, No. 101, Section 2, Kuang-Fu Rd., Hsinchu 30013, Taiwan (China); Tsai, W.C. [Department of Engineering and System Science, National Tsing Hua University, No. 101, Section 2, Kuang-Fu Rd., Hsinchu 30013, Taiwan (China); Liu, Y.H. [Nuclear Science and Technol. Develop. Center, National Tsing Hua University, No. 101, Section 2, Kuang-Fu Rd., Hsinchu 30013, Taiwan (China); Jiang, S.H., E-mail: shjiang@mx.nthu.edu.tw [Department of Engineering and System Science, National Tsing Hua University, No. 101, Section 2, Kuang-Fu Rd., Hsinchu 30013, Taiwan (China)] [Institute of Nuclear Engineering and Science, National Tsing Hua University, No. 101, Section 2, Kuang-Fu Rd., Hsinchu 30013, Taiwan (China)

    2011-12-15

    This paper aims to evaluate the effective dose as well as equivalent doses of several organs of an adult hermaphrodite mathematical phantom according to the definition of ICRP Publication 60 for BNCT treatments of brain tumors in the epithermal neutron beam at THOR. The MCNP5 Monte Carlo code was used for the calculation of the average absorbed dose of each organ. The effective doses for a typical brain tumor treatment with a tumor treatment dose of 20 Gy-eq were evaluated to be 0.59 and 0.35 Sv for the LLAT and TOP irradiation geometries, respectively. In addition to the stochastic effect, it was found that it is also likely to produce deterministic effects, such as cataracts and depression of haematopoiesis.

  13. Workplace characterization using spectrometry for reliable neutron dose assessment in mixed neutron photon fields

    International Nuclear Information System (INIS)

    Bolognese-Milsztajn, T.; Asselineau, B.; Itie, C.; Lahaye, T.; Muller, H.

    2002-01-01

    Radiation protection for practices should be based on justification of exposure, optimization of protection and dose limitations. The 96/29 European Union directive clearly requires that operational protection of exposed workers shall be based on prior evaluation to identify the nature and magnitude of the radiological risk to exposed workers and implementation of optimization of radiation protection in all working conditions. Operational protection of exposed workers requires the implementation of measures at workplaces. The ICRP60 publication (1991) defines and recommends the use of the effective dose E for radiological protections purposes. This quantity can not be directly measured; the operational quantities ambient dose equivalent: H and personal dose equivalent: Hp as defined in the ICRU 51 report (1993) are recommended to estimate the effective dose E. These operational quantities are related, through calculated conversion coefficients h Φ , to the measurable function fluence Φ by: H = ∫ h Φ (E) Φ E dE (1a); Hp = ∫ hp Φ (E, Ω) Φ E , Ω dE dΩ (1b). Φ E (Φ E ,Ω) is the distribution of the fluence with respect to energy (energy E and direction Ω); h Φ or hp Φ are the fluence to operational quantity conversion coefficients as defined by the ICRU 57 report (1998)

  14. The radiobiology of boron neutron capture therapy: Are ''photon-equivalent'' doses really photon-equivalent?

    International Nuclear Information System (INIS)

    Coderre, J.A.; Diaz, A.Z.; Ma, R.

    2001-01-01

    Boron neutron capture therapy (BNCT) produces a mixture of radiation dose components. The high-linear energy transfer (LET) particles are more damaging in tissue than equal doses of low-LET radiation. Each of the high-LET components can multiplied by an experimentally determined factor to adjust for the increased biological effectiveness and the resulting sum expressed in photon-equivalent units (Gy-Eq). BNCT doses in photon-equivalent units are based on a number of assumptions. It may be possible to test the validity of these assumptions and the accuracy of the calculated BNCT doses by 1) comparing the effects of BNCT in other animal or biological models where the effects of photon radiation are known, or 2) if there are endpoints reached in the BNCT dose escalation clinical trials that can be related to the known response to photons of the tissue in question. The calculated Gy-Eq BNCT doses delivered to dogs and to humans with BPA and the epithermal neutron beam of the Brookhaven Medical Research Reactor were compared to expected responses to photon irradiation. The data indicate that Gy-Eq doses in brain may be underestimated. Doses to skin are consistent with the expected response to photons. Gy-Eq doses to tumor are significantly overestimated. A model system of cells in culture irradiated at various depths in a lucite phantom using the epithermal beam is under development. Preliminary data indicate that this approach can be used to detect differences in the relative biological effectiveness of the beam. The rat 9L gliosarcoma cell survival data was converted to photon-equivalent doses using the same factors assumed in the clinical studies. The results superimposed on the survival curve derived from irradiation with Cs-137 photons indicating the potential utility of this model system. (author)

  15. Improved Dose Targeting for a Clinical Epithermal Neutron Capture Beam Using Optional 6Li Filtration

    International Nuclear Information System (INIS)

    Binns, Peter J.; Riley, Kent J.; Ostrovsky, Yakov; Gao Wei; Albritton, J. Raymond; Kiger, W.S.; Harling, Otto K.

    2007-01-01

    Purpose: The aim of this study was to construct a 6 Li filter and to improve penetration of thermal neutrons produced by the fission converter-based epithermal neutron beam (FCB) for brain irradiation during boron neutron capture therapy (BNCT). Methods and Materials: Design of the 6 Li filter was evaluated using Monte Carlo simulations of the existing beam line and radiation transport through an ellipsoidal water phantom. Changes in beam performance were determined using three figures of merit: (1) advantage depth (AD), the depth at which the total biologically weighted dose to tumor equals the maximum weighted dose to normal tissue; (2) advantage ratio (AR), the ratio of the integral tumor dose to that of normal tissue averaged from the surface to the AD; and (3) advantage depth dose rate (ADDR), the therapeutic dose rate at the AD. Dosimetry performed with the new filter installed provided calibration data for treatment planning. Past treatment plans were recalculated to illustrate the clinical potential of the filter. Results: The 8-mm-thick Li filter is more effective for smaller field sizes, increasing the AD from 9.3 to 9.9 cm, leaving the AR unchanged at 5.7 but decreasing the ADDR from 114 to 55 cGy min -1 for the 12 cm diameter aperture. Using the filter increases the minimum deliverable dose to deep seated tumors by up to 9% for the same maximum dose to normal tissue. Conclusions: Optional 6 Li filtration provides an incremental improvement in clinical beam performance of the FCB that could help to establish a therapeutic window in the future treatment of deep-seated tumors

  16. Effects of high neutron doses and duration of the chemical etching on the optical properties of CR-39

    International Nuclear Information System (INIS)

    Sahoo, G.S.; Tripathy, S.P.; Paul, S.; Sharma, S.C.; Joshi, D.S.; Gupta, A.K.; Bandyopadhyay, T.

    2015-01-01

    Effects of the duration of chemical etching on the transmittance, absorbance and optical band gap width of the CR-39 (Polyallyl diglycol carbonate) detectors irradiated to high neutron doses (12.7, 22.1, 36.0 and 43.5 Sv) were studied. The neutrons were produced by bombardment of a thick Be target with 12 MeV protons of different fluences. The unirradiated and neutron-irradiated CR-39 detectors were subjected to a stepwise chemical etching at 1 h intervals. After each step, the transmission spectra of the detectors were recorded in the range from 200 to 900 nm, and the absorbances and optical band gap widths were determined. The effect of the etching on the light transmittance of unirradiated detectors was insignificant, whereas it was very significant in the case of the irradiated detectors. The dependence of the optical absorbance on the neutron dose is linear at short etching periods, but exponential at longer ones. The optical band gap narrows with increasing etching time. It is more significant for the irradiated dosimeters than for the unirradiated ones. The rate of the narrowing of the optical band gap with increasing neutron dose increases with increasing duration of the etching. - Highlights: • The variation of optical properties of CR-39 at very high neutron dose is analyzed. Etching process is found to play a crucial role for change in optical properties of neutron-irradiated CR-39. • The optical absorbance varies linearly at lower dose, at very high dose absorbance saturation occurs. The dose at which saturation absorbance is observed shifts towards lower neutron dose with increase in etching time. • The rate of decrease in optical band gap with respect to neutron dose is found to be more at higher etching durations

  17. A Monte Carlo Study on the Effect of Various Neutron Capturers on Dose Distribution in Brachytherapy with 252Cf Source

    Directory of Open Access Journals (Sweden)

    Firoozabadi M. M.

    2017-03-01

    Full Text Available Background: In neutron interaction with matter and reduction of neutron energy due to multiple scatterings to the thermal energy range, increasing the probability of thermal neutron capture by neutron captures makes dose enhancement in the tumors loaded with these materials. Objective: The purpose of this study is to evaluate dose distribution in the presence of 10B, 157Gd and 33S neutron capturers and to determine the effect of these materials on dose enhancement rate for 252Cf brachytherapy source. Methods: Neutron-ray flux and energy spectra, neutron and gamma dose rates and dose enhancement factor (DEF are determined in the absence and presence of 10B, 157Gd and 33S using Monte Carlo simulation. Results: The difference in the thermal neutron flux rate in the presence of 10B and 157Gd is significant, while the flux changes in the fast and epithermal energy ranges are insensible. The dose enhancement factor has increased with increasing distance from the source and reached its maximum amount equal to 258.3 and 476.1 cGy/h/µg for 157Gd and 10B, respectively at about 8 cm distance from the source center. DEF for 33S is equal to one. Conclusion: Results show that the magnitude of dose augmentation in tumors containing 10B and 157Gd in brachytherapy with 252Cf source will depend not only on the capture product dose level, but also on the tumor distance from the source. 33S makes dose enhancement under specific conditions that these conditions depend on the neutron energy spectra of source, the 33S concentration in tumor and tumor distance from the source.

  18. A punched-card library of neutron cross-sections and its use in the mechanized preparation of group cross-sections for use in Monte Carlo, Carlson S{sub n} and other multi-group neutronics calculations on high-speed computers

    Energy Technology Data Exchange (ETDEWEB)

    Parker, K [Atomic Weapons Research Establishment, Aldermaston (United Kingdom)

    1962-03-15

    The AWRE punched-card library of neutron cross-sections is described together with associated IBM-7090 programmes which process this data to give group-averaged cross-sections for use in Monte Carlo, Carlson S{sub n} and other multi-group neutronics calculations. The methods developed to deal with both isotropic and anisotropic elastic scattering are described. These include the multi-group transport approximation and the full treatment of anisotropic scattering using the Legendre polynomial moments of the scattering transfer matrix. The principles of group-constant formation are considered and illustrated by describing systems of group constants suitable for fast-reactor calculations. Practical problems such as the empirical adjustment of group constants to reproduce integral results and the collapsing of a many-group set of constants to give a few-group set are discussed. (author) [French] L'auteur decrit le fichier de cartes perforees sur lesquelles on enregistre a l'Atomic Weapons Research Establishment (AWRE) les sections efficaces neutroniques ainsi que les programmes IBM-7090 associes qui sont employes pour le traitement de ces informations, en vue d'obtenir des sections efficaces moyennes par groupe pouvant servir aux calculs de neutroniques a plusieurs groupes, effectues a l'aide des methodes de Monte-Carlo, S{sub n} de Carlson et autres methodes. L'auteur expose ensuite les methodes mises au point roda etudier la diffusion elastique, tant isotrope qu'anisotrope. Elles comprennent l'approximation de transport a plusieurs groupes, ainsi que le traitement complet de la diffusion anisotrope par les moments polynomiaux de Legendre de la matrice de transfert de la diffusion. L'auteur examine les principes de la formation des constantes de groupes; a titre d'illustration, il decrit les systemes de constantes de groupes qui se pretent aux calculs de reacteurs a neutrons rapides. Il expose quelques problemes pratiques, tels que l'ajustement empirique des

  19. The Primary Origin of Dose Rate Effects on Microstructural Evolution of Austenitic Alloys During Neutron Irradiation

    International Nuclear Information System (INIS)

    Okita, Taira; Sato, Toshihiko; Sekimura, Naoto; Garner, Francis A.; Greenwood, Lawrence R.

    2002-01-01

    The effect of dose rate on neutron-induced microstructural evolution was experimentally estimated. Solution-annealed austenitic model alloys were irradiated at approximately 400 degrees C with fast neutrons at seven different dose rates that vary more than two orders difference in magnitude, and two different doses were achieved at each dose rate. Both cavity nucleation and growth were found to be enhanced at lower dose rate. The net vacancy flux is calculated from the growth rate of cavities that had already nucleated during the first cycle of irradiation and grown during the second cycle. The net vacancy flux was found to be proportional to (dpa/sec) exp (1/2) up to 28.8 dpa and 8.4 x 10 exp (-7) dpa/sec. This implies that mutual recombination dominates point defect annihilation, in this experiment even though point defect sinks such as cavities and dislocations were well developed. Thus, mutual recombination is thought to be the primary origin of the effect of dose rate on microstructural evolution

  20. Swelling of spinel after low-dose neutron irradiation

    International Nuclear Information System (INIS)

    Coghlan, W.A.; Clinard, F.W. Jr.; Itoh, N.; Greenwood, L.R.

    1986-01-01

    Swelling was determined in samples of single-crystal MgAl 2 O 4 spinel, irradiated to doses as high as 8 x 10 22 n/m 2 (E > 0.1 MeV) at approx. =50 0 C in the Omega West Reactor. Swelling effectively saturated at approx. =2 x 10 22 n/m 2 which corresponds to a damage level of only approx. =2 x 10 -3 dpa. In addition subsequent measurements after irradiation have revealed that the samples continued swelling for several weeks. These results imply that irradiation defects begin to interact by recombination and aggregation at low damage levels in this material at 50 0 C and perhaps continue to cluster at room temperature after irradiation. Rate equations have been employed to determine defect concentrations at saturation. Results to date show that the observed swelling is consistent with the number of surviving defects if swelling per Frenkel defect pair is taken to be one atomic volume

  1. Analysis of neutron dose rates on RGTT200K core using MCNP5

    International Nuclear Information System (INIS)

    Suwoto; Zuhair

    2016-01-01

    The conceptual design of RGTT200K (High Temperature Gas-cooled Reactor of 200 MWth Cogeneration) is the non-annular cylindrical reactor core with TRISO kernel coated fuel particles in the form of balls called pebble and cooled by helium gas. The RGTT200K reactor core design adopts high temperature gas cooled reactor (HTGR) technology with inherent passive safety. The RGTT200K spherical fuel called pebble fuel containing thousand of TRISO-coated fuel particles of uranium oxide (UO 2 ) 10 % enriched. TRISO coating comprises four layers, namely: porous carbon buffer layer, inner pyrolytic carbon layer (IPyC, Inner Pyrolytic Carbon), silicon carbide layer (SiC) and a layer of pyrolytic carbon outer portion (OPyC, Outer Pyrolytic Carbon). Modeling and analysis of preliminary calculation of neutron dose rate on normal operating temperature (T kernel =1200K) and accident temperature (T kernel =1800K) of the RGTT200K core were performed using Monte Carlo MCNP5v1.2 code. The continuous energy nuclear data cross-sections was taken from ENDF/B-VII, JENDL-4 and JEFF-3.1 nuclear data files . Double heterogeneity model in TRISO-coated fuel particles kernel and the pebble of RGTT200K core. By utilizing EGS99304 code, the 640 amount of energy group structures (SAND-II neutron group structures) is used in the neutron fluxes and spectrum calculation in RGTT200K reactor. The RGTT200K reactor core is divided into 25 zones (5 zones in radial and 10 zones in axial directions), while the modeling of radiation and biological shielding reactor RGTT200K are used to determine of preliminary neutron dose rate emitted by the neutron source with tally cards are available in the MCNP5v1.2 code. The calculation result analyses of the neutron dose rate distributions are determined using a conversion factor of flux-to-dose taken from International Commission on Radiological Protection, ICRP. The preliminary calculations result show that the neutrons dose rate using ICRP-74 conversion factor for

  2. Effect of low level doses of fast neutrons on the activity of the snake venom

    International Nuclear Information System (INIS)

    Hanafy, Magda S.; Amin, Aida M.

    1998-01-01

    In this work, the effect of low level doses of fast neutrons from 252 Cf on snake venom (Cerastes cerastes) was studied through measurements of biophysical and haematological changes. The absorption spectrum (200-700 nm) of haemoglobin (Hb) collected from the blood of rats after 3 and 24 hours post injection with irradiated and non-irradiated snake venom with neutron fluences of 3x10 6 , 2.8x10 7 and 3X10 8 n/cm 2 was measured. The results indicated that injection of animals with either non- irradiated or irradiated venom ( with different neutron fluences) resulted into the decrease of the absorption band intensities of Hb. These changes in the properties of the characteristic band showed to be a marker for irradiated venom and is dose dependent. It was concluded that neutron irradiation of the venom leads to the decrease of its toxicity and, consequently, to the increase of the chance of repair mechanism in livings. Obvious changes of most haematological erythrocytic values of Hb, packed cell volume (PCV), red blood counts (RBC), mean corpuscular volume (MCV), mean corpuscular haemoglobin (MCHb) and mean corpuscular haemoglobin concentration (MCHC) were observed in the blood of the rats injected with non-irradiated venom (as a first group) and those injected with the irradiated venom (as a second group). The microcytic haemolytic anemia was more acute in the first group than in the second one which showed lesser extent. It is concluded from this study that low level doses of fast neutrons could postpone and lower acute haematological action induced by the venom. (authors)

  3. Estimation of dose distribution and neutron spectra in JCO critical accident by shielding calculations

    International Nuclear Information System (INIS)

    Sakamoto, Yukio

    2001-01-01

    The information about neutrons at the surrounding of JCO site in the critical accident is limited to survey results by neutron Rem counter in the period of accident and activation data very near the test facility measured after the shut down of accident. This caused the big uncertainty in the dose estimation by detailed shielding calculation codes. On the other hand, environmental activity data measured by radiochemical researchers included the information about fast neutrons inside of JCO site and thermal neutrons up to 1 km from test facility. It is important to grasp the actual circumstance and examine the executed evaluation of the critical accident as scientifically as possible. Therefore, it is meaningful for different field researchers to corporate and exchange the information. In the Technical Divisions of Radiation Science and Technology in Atomic Energy Society of Japan, the information about neutron spectra are released from their home page and three groups of JAERI/CRC, Sumitomo Atomic Energy Industry and Nuclear Power Engineering Corp. (NUPEC)/Mitsubishi Research Institute Inc. (MRI), tried the shielding calculation by Monte Carlo Code MCNP-4B. The procedures and main results of shielding calculations were reviewed in this report. The main difference of shielding calculation by three groups was density and water content of autoclaved light-weight concrete (ALC) as the wall and ceiling. From the result by NUPEC/MRI, it was estimated that the water content in ALC was from 0.05 g/cm 3 to 0.10 g/cm 3 . The behavior of dose equivalent attenuation obtained by shielding calculation was very similar with the measured data from 250 m to 1,700 m obtained by survey meter, TLD and monitoring post. For more exact dose estimation, more detail examination of density and water content of ALC will be needed. (author)

  4. Dose profile modeling of Idaho National Laboratory's active neutron interrogation laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Chichester, D.L. [Idaho National Laboratory, 2525 N. Fremont Avenue, Idaho Falls, ID 83415 (United States)], E-mail: david.chichester@inl.gov; Seabury, E.H.; Zabriskie, J.M.; Wharton, J.; Caffrey, A.J. [Idaho National Laboratory, 2525 N. Fremont Avenue, Idaho Falls, ID 83415 (United States)

    2009-06-15

    A new laboratory has been commissioned at Idaho National Laboratory for performing active neutron interrogation research and development. The facility is designed to provide radiation shielding for deuterium-tritium (DT) fusion (14.1 MeV) neutron generators (2x10{sup 8} n/s), deuterium-deuterium (DD) fusion (2.5 MeV) neutron generators (1x10{sup 7} n/s), and {sup 252}Cf spontaneous fission neutron sources (6.96x10{sup 7} n/s, 30 {mu}g). Shielding at the laboratory is comprised of modular concrete shield blocks 0.76 m thick with tongue-in-groove features to prevent radiation streaming, arranged into one small and one large test vault. The larger vault is designed to allow operation of the DT generator and has walls 3.8 m tall, an entrance maze, and a fully integrated electrical interlock system; the smaller test vault is designed for {sup 252}Cf and DD neutron sources and has walls 1.9 m tall and a simple entrance maze. Both analytical calculations and numerical simulations were used in the design process for the building to assess the performance of the shielding walls and to ensure external dose rates are within required facility limits. Dose rate contour plots have been generated for the facility to visualize the effectiveness of the shield walls and entrance mazes and to illustrate the spatial profile of the radiation dose field above the facility and the effects of skyshine around the vaults.

  5. Incidence of leukemia among atomic bomb survivors in relation to neutron and gamma dose, Hiroshima and Nagasaki, 1950-71

    International Nuclear Information System (INIS)

    Ishimaru, Toranosuke; Otake, Masanori; Ichimaru, Michito.

    1978-03-01

    The incidence of leukemia during 1950-71 in the fixed mortality sample of atomic bomb survivors in Hiroshima and Nagasaki has been analyzed as a function of individual gamma and neutron kerma and marrow dose. Two dose response models were tested for each of acute leukemia, chronic granulocytic leukemia, and all types of leukemia, respectively. Each model postulates that leukemia incidence depends upon the sum of the separate risks imposed by the gamma ray and neutron doses; in Model I both are assumed to be directly proportional to the respective doses, while Model II assumes that while the risk from neutrons is directly proportional to the dose, the risk from gamma rays is proportional to dose-squared. Weighted regression analyses were performed for each model. When the two models were fitted to the data for all types of leukemia, the estimated regression coefficients corresponding to the neutron and gamma ray doses both differed significantly from zero, for each model. However, when analysis was restricted to acute leukemia, both the neutron and gamma ray coefficients were significant only for Model II, and with respect to chronic granulocytic leukemia, only the coefficient of the neutron dose was significant, using either Model I or Model II. It appeared that the responses of the two leukemia types differed by type of radiation. If the chronic granulocytic and acute leukemias are considered together, the Model II appears to fit the data slightly better than Model I, but neither models is rejected by the data. (author)

  6. Ambient neutron dose equivalent outside concrete vault rooms for 15 and 18 MV radiotherapy accelerators

    International Nuclear Information System (INIS)

    Martinez-ovalle, S. A.; Barquero, R.; Gomez-ros, J. M.; Lallena, A. M.

    2012-01-01

    In this work, the ambient dose equivalent, H*(10), due to neutrons outside three bunkers that house a 15- and a 18-MV Varian Clinac 2100C/D and a 15-MV Elekta Inor clinical linacs, has been calculated. The Monte Carlo code MCNPX (v. 2.5) has been used to simulate the neutron production and transport. The complete geometries including linacs and full installations have been built up according to the specifications of the manufacturers and the planes provided by the corresponding medical physical services of the hospitals where the three linacs operate. Two of these installations, those lodging the Varian linacs, have an entrance door to the bunker while the other one does not, although it has a maze with two bends. Various treatment orientations were simulated in order to establish plausible annual equivalent doses. Specifically anterior-posterior, posterior-anterior, left lateral, right lateral orientations and an additional one with the gantry rotated 30 deg. have been studied. Significant dose rates have been found only behind the walls and the door of the bunker, near the entrance and the console, with a maximum of 12 μSv h -1 . Dose rates per year have been calculated assuming a conservative workload for the three facilities. The higher dose rates in the corresponding control areas were 799 μSv y -1 , in the case of the facility which operates the 15-MV Clinac, 159 μSv y -1 , for that with the 15-MV Elekta, and 21 μSv y -1 for the facility housing the 18-MV Varian. A comparison with measurements performed in similar installations has been carried out and a reasonable agreement has been found. The results obtained indicate that the neutron contamination does not increase the doses above the legal limits and does not produce a significant enhancement of the dose equivalent calculated. When doses are below the detection limits provided by the measuring devices available today, MCNPX simulation provides an useful method to evaluate neutron dose equivalents

  7. Angular dependence of dose equivalent response of an albedo neutron dosimeter

    International Nuclear Information System (INIS)

    Torres, B.A.; Boswell, E.; Schwartz, R.B.

    1994-01-01

    The ANSI provides procedures for testing the performance of dosimetry services. Although neutron dose equivalent angular response studies are not now mandated, future standards may well require that such studies be performed. Current studies with an albedo dosimeter will yield information regarding the angular dependence of dose equivalent response for this type of personnel dosimeter. Preliminary data for bare 252 Cf fluences show a marked decrease in dosimeter reading with increasing angle. The response decreased by an approximate factor of four. For the horizontal orientation, the same response was noted from both positive and negative angles. However, for the vertical orientation, the response was unexplainably assymetric. We are also examining the response of the personnel badge in moderated 252 Cf fluences. Responses from the moderated and unmoderated 252 Cf fields and theoretical calculations of the neutron angular response will be compared. This information will assist in building a data base for future comparisons of neutron angular responses with other neutron albedo dosimeters and phantoms

  8. MC2-3: Multigroup Cross Section Generation Code for Fast Reactor Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Lee, C. H. [Argonne National Lab. (ANL), Argonne, IL (United States); Yang, W. S. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2013-11-08

    The MC2-3 code is a Multigroup Cross section generation Code for fast reactor analysis, developed by improving the resonance self-shielding and spectrum calculation methods of MC2-2 and integrating the one-dimensional cell calculation capabilities of SDX. The code solves the consistent P1 multigroup transport equation using basic neutron data from ENDF/B data files to determine the fundamental mode spectra for use in generating multigroup neutron cross sections. A homogeneous medium or a heterogeneous slab or cylindrical unit cell problem is solved in ultrafine (~2000) or hyperfine (~400,000) group levels. In the resolved resonance range, pointwise cross sections are reconstructed with Doppler broadening at specified isotopic temperatures. The pointwise cross sections are directly used in the hyperfine group calculation whereas for the ultrafine group calculation, self-shielded cross sections are prepared by numerical integration of the pointwise cross sections based upon the narrow resonance approximation. For both the hyperfine and ultrafine group calculations, unresolved resonances are self-shielded using the analytic resonance integral method. The ultrafine group calculation can also be performed for two-dimensional whole-core problems to generate region-dependent broad-group cross sections. Multigroup cross sections are written in the ISOTXS format for a user-specified group structure. The code is executable on UNIX, Linux, and PC Windows systems, and its library includes all isotopes of the ENDF/BVII. 0 data.

  9. Application of laboratory sourceless object counting for the estimation of the neutron dose

    International Nuclear Information System (INIS)

    Cheng Jie; Ning Jing; Zhang Xiaomin; Qu Decheng; Xie Xiangdong; Nan Hongjie

    2011-01-01

    Objective: To estimate the neutron dose using 24 Na energy spectrum analysis method. Methods: Genius-2000 GeomComposer software package was used to calibrate the efficiency of the detector. Results: The detection efficiency of the detector toward γ photon with an energy of 1.368 MeV was quickly found to be 4.05271×10 -3 while the error of the software was 4.0% . The estimated dose value of the neutron irradiation samples was between 1.94 Gy and 2.82 Gy, with an arithmetic mean value of 2.38 Gy. The uncertainty of the dosimetry was about 20.07% . Conclusion: The application of efficiency calibration without a radioactive source of the energy spectrum analysis of the 24 Na contained in human blood with accelerate the estimation process. (authors)

  10. Stimulation growth effect of Eriocheir sinensis treated with low-dose neutron

    International Nuclear Information System (INIS)

    Luo Keyong; Liu Chunquan; Xu Lixin; Peng Zhangji

    2006-01-01

    This paper was dealt with the relationship between biochemical indexes and different growth stages of Eriocheir sinensis megalopa treated with Low-dose Neutron at 55.24 to 73.66 mGy. It showed that some biochemical component indexes were increased, such as-SH group in protain (between 23.40% to 69.59%), albumen (between 4.99% to 22.6%) and Hyp compared with CK. However, free radical level (between 7.67% to 32.68%) and AKP were decreased. The carapace color was turned into darker than that of CK; Antibacterial immunity of younger crab during the growing stage was increased, the body size of treated Eriocheir sinensis megalopa became uniform and early sexual maturity was inhibited in a certain degree with a low dose neutron treatment. (authors)

  11. Measurement of two-dimensional thermal neutron flux in a water phantom and evaluation of dose distribution characteristics

    International Nuclear Information System (INIS)

    Yamamoto, Kazuyoshi; Kumada, Hiroaki; Kishi, Toshiaki; Torii, Yoshiya; Horiguchi, Yoji

    2001-03-01

    To evaluate nitrogen dose, boron dose and gamma-ray dose occurred by neutron capture reaction of the hydrogen at the medical irradiation, two-dimensional distribution of the thermal neutron flux is very important because these doses are proportional to the thermal neutron distribution. This report describes the measurement of the two-dimensional thermal neutron distribution in a head water phantom by neutron beams of the JRR-4 and evaluation of the dose distribution characteristic. Thermal neutron flux in the phantom was measured by gold wire placed in the spokewise of every 30 degrees in order to avoid the interaction. Distribution of the thermal neutron flux was also calculated using two-dimensional Lagrange's interpolation program (radius, angle direction) developed this time. As a result of the analysis, it was confirmed to become distorted distribution which has annular peak at outside of the void, though improved dose profile of the deep direction was confirmed in the case which the radiation field in the phantom contains void. (author)

  12. Spectra and neutron dose of an 18 MV Linac using two geometric models of the head

    International Nuclear Information System (INIS)

    Barrera, M. T.; Pino, F.; Barros, H.; Sajo-Bohus, L.; Davila, J.; Salcedo, E.; Vega C, H. R.; Benites R, J. L.

    2015-10-01

    Full text: Using the Monte Carlo method, by MCNP5 code, simulations were performed with different source terms and 2 geometric models of the head to obtain spectra in energy, flow and doses of photo-neutrons at different positions on the stretcher and in the radiotherapy room. The simplest model was a spherical shell of tungsten; the second was the complete model of a heterogeneous head of an accelerator Varian ix. In both models Tosi function was used as a source term. In addition, for the second model Sheikh-Bagheri distribution was used for photons and photo-neutrons were generated. Also in both models the radiotherapy room of Gurve group of the Teaching Medical Center La Trinidad was included, which is equipped with an accelerator Varian Clinic 2100. In this Center passive detectors PADC (Cr-39) were irradiated with neutron converters, with 18 MeV photons radiation. The measured neutron flow was compared with that obtained with Monte Carlo calculations. The Monte Carlo flows are similar to those measured at the isocenter. The simplest model underestimates the neutron flow compared with the calculated flows with the heterogeneous model of the head. (Author)

  13. SHINE-III. Simple code for skyshine dose calculation up to 3 GeV neutrons

    Energy Technology Data Exchange (ETDEWEB)

    Tsukiyama, Toshihisa; Tayama, Ryuichi; Handa, Hiroyuki [Hitachi Engineering Co. Ltd., Ibaraki (Japan)] [and others

    2000-03-01

    Skyshine dose at site boundary is considered as one of the most fundamental issues to get approval of constructing nuclear installations. Skyshine conical beam response functions (CBRF) for high energy neutrons up to 3 GeV are obtained using NMTC-JAERI and MCNP code. This CBRF is fitted to the four parameters equation. Simple code named SHINE-III using this equation with updated data is developed. (author)

  14. A study of the responses of neutron dose equivalent survey meters with computer codes

    International Nuclear Information System (INIS)

    Sartori, D.E.; Beer, G.P. de

    1983-01-01

    The ANISN and DOT discrete-ordinates radiation transport codes for one and two dimensions have been proved as effective and simple techniques to study the response of dose equivalent neutron detectors. Comparisons between results of an experimental calibration of the Harwell 95/0075 survey meter and calculated results rendered satisfactory agreement, considering the different techniques and sources of error involved. Possible improvements in the methods and designs and causes of error are discussed. (author)

  15. Effects of low-dose gamma and neutron radiation on genotoxicity and cytotoxicity of reticulocytes in a mouse model

    International Nuclear Information System (INIS)

    Phan, N.; McFarlane, N.M.; Lemon, J.; Boreham, D.R.

    2008-01-01

    Using a successful new automation of micronucleated reticulocyte (MN-RET) scoring, the effects of low-dose (< 1.0 Gy) gamma and neutron radiation on genotoxicity and cytotoxicity of reticulocytes (RET) in a mouse model were investigated. Gamma and neutron irradiation induced significant (p<0.001) increases in the levels of %MN-RET and decreases in the levels of %RET (p<0.001) as the dose level increased. Increasing dose levels showed that gamma radiation induced significantly (p<0.05) more %MN-RET and more %RET than neutron radiation. The results suggest that neutron irradiation may be more cytotoxic (less %RET) than gamma irradiation; however, gamma irradiation may be producing cells with more chromosomal aberrations (more %MN-RET) than neutron irradiation. (author)

  16. Effects of low-dose gamma and neutron radiation on genotoxicity and cytotoxicity of reticulocytes in a mouse model

    Energy Technology Data Exchange (ETDEWEB)

    Phan, N.; McFarlane, N.M.; Lemon, J.; Boreham, D.R. [McMaster Univ., Medical Physics and Applied Radiation Sciences Unit, Hamilton, Ontario (Canada)

    2008-07-01

    Using a successful new automation of micronucleated reticulocyte (MN-RET) scoring, the effects of low-dose (< 1.0 Gy) gamma and neutron radiation on genotoxicity and cytotoxicity of reticulocytes (RET) in a mouse model were investigated. Gamma and neutron irradiation induced significant (p<0.001) increases in the levels of %MN-RET and decreases in the levels of %RET (p<0.001) as the dose level increased. Increasing dose levels showed that gamma radiation induced significantly (p<0.05) more %MN-RET and more %RET than neutron radiation. The results suggest that neutron irradiation may be more cytotoxic (less %RET) than gamma irradiation; however, gamma irradiation may be producing cells with more chromosomal aberrations (more %MN-RET) than neutron irradiation. (author)

  17. Development of a phoswich detector for neutron dose rate measurements in the Earth's atmosphere

    International Nuclear Information System (INIS)

    Doensdorf, Esther Miriam

    2014-01-01

    The Earth is constantly exposed to a stream of energetic particles from outer space. Through the interaction of this radiation with the Earth's magnetosphere and atmosphere a complex radiation field is formed which varies with the location inside the Earth's atmosphere. This radiation field consists of charged and uncharged particles leading to the constant exposure of human beings to radiation. As this ionizing radiation can be harmful for humans, it is necessary to perform dose rate measurements in different altitudes in the Earth's atmosphere. Due to their higher biological effectiveness the exposure to neutrons is more harmful than the exposure to γ-rays and charged particles, which is why the determination of neutron dose rates is the focus of this work. In this work the prototype of a Phoswich detector called PING (Phoswich Instrument for Neutrons and Gammas) is developed to determine dose rates caused by neutrons in the Earth's atmosphere and to distinguish these from γ-rays. The instrument is composed of two different scintillators optically coupled to each other and read out by one common photomultiplier tube. The scintillator package consists of an inner plastic scintillator made of the material BC-412 and a surrounding anti-coincidence made of sodium doped caesium iodide (CsI(Na)). In this work the instrument is calibrated, tested and flown and a procedure for a pulse shape analysis for this instrument is developed. With this analysis it is possible to distinguish pulses from the plastic scintillator and pulses from the CsI(Na). The pulses from the plastic scintillator are mainly due to the interaction of neutrons but there is an energy-dependent contribution of γ-rays to these events. Measurements performed on board an airplane show that the dose rates measured with the developed detector are in the same order of magnitude as results of other instruments. During measurements on board stratospheric balloons the altitude dependence of count rates and

  18. Neutron fluence-to-dose equivalent conversion factors: a comparison of data sets and interpolation methods

    International Nuclear Information System (INIS)

    Sims, C.S.; Killough, G.G.

    1983-01-01

    Various segments of the health physics community advocate the use of different sets of neutron fluence-to-dose equivalent conversion factors as a function of energy and different methods of interpolation between discrete points in those data sets. The major data sets and interpolation methods are used to calculate the spectrum average fluence-to-dose equivalent conversion factors for five spectra associated with the various shielded conditions of the Health Physics Research Reactor. The results obtained by use of the different data sets and interpolation methods are compared and discussed. (author)

  19. Correlation of clinical outcome to the estimated radiation dose from Boron Neutron Capture Therapy (BNCT)

    Energy Technology Data Exchange (ETDEWEB)

    Chadha, M. [Beth Israel Medical Center, NY (United States). Dept. of Radiation Oncology; Coderre, J.A.; Chanana, A.D. [Brookhaven National Lab., Upton, NY (United States)] [and others

    1996-12-31

    A phase I/II trial delivering a single fraction of BNCT using p-Boronophenylalanine-Fructose and epithermal neutrons at the the Brookhaven Medical Research Reactor was initiated in September 1994. The primary endpiont of the study was to evaluate the feasibility and safety of a given BNCT dose. The clinical outcome of the disease was a secondary endpoint of the study. The objective of this paper is to evaluate the correlation of the clinical outcome of patients to the estimated radiation dose from BNCT.

  20. Correlation of clinical outcome to the estimated radiation dose from Boron Neutron Capture Therapy (BNCT)

    International Nuclear Information System (INIS)

    Chadha, M.

    1996-01-01

    A phase I/II trial delivering a single fraction of BNCT using p-Boronophenylalanine-Fructose and epithermal neutrons at the the Brookhaven Medical Research Reactor was initiated in September 1994. The primary endpiont of the study was to evaluate the feasibility and safety of a given BNCT dose. The clinical outcome of the disease was a secondary endpoint of the study. The objective of this paper is to evaluate the correlation of the clinical outcome of patients to the estimated radiation dose from BNCT

  1. The application of computer and automatic technology in dose measurement of neutron radiation

    International Nuclear Information System (INIS)

    Zhou Yu; Li Chenglin; Luo Yisheng; Guo Yong; Chen Di; Xiaojiang

    1999-01-01

    Generally the dose measurement of neutron radiation requires three electrometers, two bias, three workers in the same time. To improve the accuracy and efficiency of measurement, a Model 6517A electrometer that accommodate Model 6521 scanner cards and a portable computer are used to make up of a automatic measurement system. Corresponding software is developed and used to control it. Because of the application of computer and automatic technology, this system can not only measure dose rate automatically, but also make data's calculating, saving, querying, printing and comparing ease

  2. NUMERICAL MULTIGROUP TRANSIENT ANALYSIS OF SLAB NUCLEAR REACTOR WITH THERMAL FEEDBACK

    Directory of Open Access Journals (Sweden)

    Filip Osuský

    2016-12-01

    Full Text Available The paper describes a new numerical code for multigroup transient analyses with thermal feedback. The code is developed at Institute of Nuclear and Physical Engineering. It is necessary to carefully investigate transient states of fast neutron reactors, due to recriticality issues after accident scenarios. The code solves numerical diffusion equation for 1D problem with possible neutron source incorporation. Crank-Nicholson numerical method is used for the transient states. The investigated cases are describing behavior of PWR fuel assembly inside of spent fuel pool and with the incorporated neutron source for better illustration of thermal feedback.

  3. Quantitative radiation dose-response relationships for normal tissues in man - I. Gustatory tissues response during photon and neutron radiotherapy

    International Nuclear Information System (INIS)

    Mossman, K.L.

    1982-01-01

    Quantitative radiation dose-response curves for normal gustatory tissue in man were studied. Taste function, expressed as taste loss, was evaluated in 84 patients who were given either photon or neutron radiotherapy for tumors in the head and neck region. Patients were treated to average tumor doses of 6600 cGy (photon) or 2200 cGy intervals for photon patients and 320-cGy intervals for neutron patients during radiotherapy. The dose-response curves for photons and neutrons were analyzed by fitting a four-parameter logistic equation to the data. Photon and neutron curves differed principally in their relative position along the dose axis. Comparison of the dose-response curves were made by determination of RBE. At 320 cGy, the lowest neutron dose at which taste measurements were made, RBE = 5.7. If this RBE is correct, then the therapeutic gain factor may be equal to or less than 1, indicating no biological advantage in using neutrons over photons for this normal tissue. These studies suggest measurements of taste function and evaluation of dose-response relationships may also be useful in quantitatively evaluating the efficacy of chemical modifiers of radiation response such as hypoxic cell radiosensitizers and radioprotectors

  4. Neutron flux and gamma dose measurement in the BNCT irradiation facility at the TRIGA reactor of the University of Pavia

    Science.gov (United States)

    Bortolussi, S.; Protti, N.; Ferrari, M.; Postuma, I.; Fatemi, S.; Prata, M.; Ballarini, F.; Carante, M. P.; Farias, R.; González, S. J.; Marrale, M.; Gallo, S.; Bartolotta, A.; Iacoviello, G.; Nigg, D.; Altieri, S.

    2018-01-01

    University of Pavia is equipped with a TRIGA Mark II research nuclear reactor, operating at a maximum steady state power of 250 kW. It has been used for many years to support Boron Neutron Capture Therapy (BNCT) research. An irradiation facility was constructed inside the thermal column of the reactor to produce a sufficient thermal neutron flux with low epithermal and fast neutron components, and low gamma dose. In this irradiation position, the liver of two patients affected by hepatic metastases from colon carcinoma were irradiated after borated drug administration. The facility is currently used for cell cultures and small animal irradiation. Measurements campaigns have been carried out, aimed at characterizing the neutron spectrum and the gamma dose component. The neutron spectrum has been measured by means of multifoil neutron activation spectrometry and a least squares unfolding algorithm; gamma dose was measured using alanine dosimeters. Results show that in a reference position the thermal neutron flux is (1.20 ± 0.03) ×1010 cm-2 s-1 when the reactor is working at the maximum power of 250 kW, with the epithermal and fast components, respectively, 2 and 3 orders of magnitude lower than the thermal component. The ratio of the gamma dose with respect to the thermal neutron fluence is 1.2 ×10-13 Gy/(n/cm2).

  5. Comparison of doses delivered in clinical trials of neutron capture therapy in the USA

    International Nuclear Information System (INIS)

    Albritton, J.R.; Binns, P.J.; Riley, K.J.; Coderre, J.A.; Harling, O.K.; Kiger, W.S. III

    2006-01-01

    A combined 81 brain tumor patients have been treated in dose escalation trials of Neutron Capture Therapy (NCT) at Harvard-MIT and Brookhaven National Laboratory (BNL). Pooling the clinical outcomes from these trials will permit evaluation with more statistical rigor. However, differences in physical and computational dosimetry between the institutions make direct comparison of the clinical dosimetry difficult. This paper describes work performed to normalize the BNL clinical dosimetry to that of Harvard-MIT for combined dose response analysis. This normalization involved analysis of MIT measurements and calculations using the BNL treatment planning system (TPS), BNCT - Rtpe, for two different phantoms. The BNL TPS was calibrated to dose measurements made by MIT at the BMRR in the BNL calibration phantom, a Lucite cube, and then validated by MIT dose measurements at the BMMR in an ellipsoidal water phantom. Treatment plans for all BNL patients were recomputed using the newly determined TPS calibration, yielding reductions in reported mean brain doses of 19% on average in the initial 15 patients and 31% in the latter 38 patients. These reductions in reported doses have clinically significant implications for those relying on reported BNL doses as a basis for initial dose selection in clinical studies. (author)

  6. Application of semiconductor MOSFET and pin diode dosimeters to epithermal neutron beam dose distribution measurements in phantoms

    International Nuclear Information System (INIS)

    Carolan, M.G.; Wallace, S.A.; Allen, B.J.; Rosenfeld, A.B.; Mathur, J.N.

    1996-01-01

    For any clinical application of Boron Neutron Capture Therapy (BNCT) fast and accurate dose calculations will be required for treatment planning. Such calculations are also necessary for the planning and interpretation of results from pre-clinical and clinical trials where the speed of calculation is not so critical. A dose calculation system based on the MCNP Monte Carlo Neutron transport code has been developed by Wallace. This system takes image data from CT scans and constructs a voxel based geometrical model for input into MCNP. To validate the calculations, a number of phantoms were constructed and exposed in the HB11 epithermal neutron beam at the HFR of the CEC Joint Research Centre in Petten. The doses recorded by arrays of PIN diode neutron dosimeters and MOSFET gamma dosimeters in these phantoms were compared with the calculated results from the MCNP dose planning system. Initial results have been reported elsewhere. Poster 197. (author)

  7. Development of an anthropomorfic simulator for simulation and measurements of neutron dose and flux the facility for BNCT studies

    International Nuclear Information System (INIS)

    Muniz, Rafael Oliveira Rondon

    2010-01-01

    IPEN facility for researches in BNCT (Boron Neutron Capture Therapy) uses IEA-R1 reactor's irradiation channel number 3, where there is a mixed radiation field - neutrons and gamma. The researches in progress require the radiation fields, in the position of the irradiation of sample, to have in its composition maximized thermal neutrons component and minimized, fast and epithermal neutron flux and gamma radiation. This work was developed with the objective of evaluating whether the present radiation field in the facility is suitable for BNCT researches. In order to achieve this objective, a methodology for the dosimetry of thermal neutrons and gamma radiation in mixed fields of high doses, which was not available in IPEN, was implemented in the Center of Nuclear Engineering of IPEN, by using thermoluminescent dosimeters - TLDs 400, 600 and 700. For the measurements of thermal and epithermal neutron flux, activation detectors of gold were used applying the cadmium ratio technique. A cylindrical phantom composed by acrylic discs was developed and tested in the facility and the DOT 3.5. computational code was used in order to obtain theoretical values of neutron flux and the dose along phantom. In the position corresponding to about half the length of the cylinder of the phantom, the following values were obtained: thermal neutron flux (2,52 ± 0,06).10 8 n/cm 2 s, epithermal neutron flux (6,17 ± 0,26).10 7 .10 6 n/cm 2 s, absorbed dose due to thermal neutrons (4,2 ± 1,8)Gy and (10,1 ± 1,3)Gy due to gamma radiation. The obtained values show that the fluxes of thermal and epithermal neutrons flux are appropriate for studies in BNCT, however, the dose due to gamma radiation is high, indicating that the facility should be improved. (author)

  8. Organ and Effective Dose Coefficients for Cranial and Caudal Irradiation Geometries: Neutrons

    Science.gov (United States)

    Veinot, K. G.; Eckerman, K. F.; Hertel, N. E.; Hiller, M. M.

    2017-09-01

    With the introduction of new recommendations by ICRP Publication 103, the methodology for determining the protection quantity, effective dose, has been modified. The modifications include changes to the defined organs and tissues, the associated tissue weighting factors, radiation weighting factors, and the introduction of reference sex-specific computational phantoms (ICRP Publication 110). Computations of equivalent doses in organs and tissues are now performed in both the male and female phantoms and the sex-averaged values used to determine the effective dose. Dose coefficients based on the ICRP 103 recommendations were reported in ICRP Publication 116, the revision of ICRP Publication 74 and ICRU Publication 57. The coefficients were determined for the following irradiation geometries: anterior-posterior (AP), posterior-anterior (PA), right and left lateral (RLAT and LLAT), rotational (ROT), and isotropic (ISO). In this work, the methodology of ICRP Publication 116 was used to compute dose coefficients for neutron irradiation of the body with parallel beams directed upward from below the feet (caudal) and directed downward from above the head (cranial). These geometries may be encountered in the workplace from personnel standing on contaminated surfaces or volumes and from overhead sources. Calculations of organ and tissue absorbed doses for caudal and cranial exposures to neutrons ranging in energy from 10-9 MeV to 10 GeV have been performed using the MCNP6 radiation transport code and the adult reference voxel phantoms of ICRP Publication 110. At lower energies the effective dose per particle fluence for cranial and caudal exposures is less than AP orientations while above about 30 MeV the cranial and caudal values are greater.

  9. An assessment of the secondary neutron dose in the passive scattering proton beam facility of the national cancer center

    Energy Technology Data Exchange (ETDEWEB)

    Han, Sang Eun [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of); Cho, Gyuseong [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of); Lee, Se Byeong [Proton Therapy Center, National Cancer Center, Goyang (Korea, Republic of)

    2017-06-15

    The purpose of this study is to assess the additional neutron effective dose during passive scattering proton therapy. Monte Carlo code (Monte Carlo N-Particle 6) simulation was conducted based on a precise modeling of the National Cancer Center's proton therapy facility. A three-dimensional neutron effective dose profile of the interior of the treatment room was acquired via a computer simulation of the 217.8-MeV proton beam. Measurements were taken with a 3He neutron detector to support the simulation results, which were lower than the simulation results by 16% on average. The secondary photon dose was about 0.8% of the neutron dose. The dominant neutron source was deduced based on flux calculation. The secondary neutron effective dose per proton absorbed dose ranged from 4.942 ± 0.031 mSv/Gy at the end of the field to 0.324 ± 0.006 mSv/Gy at 150 cm in axial distance.

  10. Three-dimensional neutron dose distribution in the environment around a 1-GeV electron synchrotron facility at INS

    International Nuclear Information System (INIS)

    Uwamino, Y.; Nakamura, T.

    1987-01-01

    The three-dimensional (surface and altitude) skyshine neutron-dose-equivalent distribution around the 1-GeV electron synchrotron (ES) of the Institute for Nuclear Study, University of Tokyo, was measured with a high-sensitivity dose-equivalent counter. The neutron spectrum in the environment was also measured with a multimoderator spectrometer incorporating a 3 He counter. The dose-equivalent distribution and the leakage neutron spectrum at the surface of the ES building were measured with a Studsvik 2202D counter and the multimoderator spectrometer, including an indium activation detector. Skyshine neutron transport calculations, beginning with the photoneutron spectrum and yielding the dose-equivalent distribution in the environment, were performed with the DOT3.5 code and two Monte Carlo codes, MMCR-2 and MMCR-3, using the DLC-87/HILO group cross sections. The calculated neutron spectra at the top surface of the concrete ceiling and at a point 111 m from the ES agreed well with the measured results, and the calculated three-dimensional dose-equivalent distribution also agreed. The dose value increased linearly with altitude, and the slope was estimated for neutron-producing facilities. (author)

  11. Determination of neutron dose from criticality accidents with bioassays for sodium-24 in blood and phosphorus-32 in hair

    International Nuclear Information System (INIS)

    Feng, Y.; Miller, L.F.; Brown, K.S.; Casson, W.H.; Mei, G.T.; Thein, M.

    1993-06-01

    A comprehensive review of accident neutron dosimetry using blood and hair analysis was performed and is summarized in this report. Experiments and calculations were conducted at Oak Ridge National Laboratory (ORNL) and the University of Tennessee (UT) to develop measurement techniques for the activity of 24 Na in blood and 32 P in hair for nuclear accident dosimetry. An operating procedure was established for the measurement of 24 Na in blood using an HPGe detector system. The sensitivity of the measurement for a 20-mL sample is 0.01-0.02 Gy of total neutron dose for hard spectra and below 0.005 Gy for soft spectra based on a 30- to 60-min counting time. The operating procedures for direct counting of hair samples are established using a liquid scintillation detector. Approximately 0.06-0.1 Gy of total neutron dose can be measured from a 1-g hair sample using this procedure. Detailed procedures for chemical dissolution and ashing of hair samples are also developed. A method is proposed to use blood and hair analysis for assessing neutron dose based on a collection of 98 neutron spectra. Ninety-eight blood activity-to-dose conversion factors were calculated. The calculated results for an uncollided fission spectrum compare favorably with previously published data for fission neutrons. This nuclear accident dosimetry system makes it possible to estimate an individual's neutron dose within a few hours after an accident if the accident spectrum can be approximated from one of 98 tabulated neutron spectrum descriptions. If the information on accident and spectrum description is not available, the activity ratio of 32 P in hair and 24 Na in blood can provide information related to the neutron spectrum for dose assessment

  12. Cassandre : a two-dimensional multigroup diffusion code for reactor transient analysis

    International Nuclear Information System (INIS)

    Arien, B.; Daniels, J.

    1986-12-01

    CASSANDRE is a two-dimensional (x-y or r-z) finite element neutronics code with thermohydraulics feedback for reactor dynamics prior to the disassembly phase. It uses the multigroup neutron diffusion theory. Its main characteristics are the use of a generalized quasistatic model, the use of a flexible multigroup point-kinetics algorithm allowing for spectral matching and the use of a finite element description. The code was conceived in order to be coupled with any thermohydraulics module, although thermohydraulics feedback is only considered in r-z geometry. In steady state criticality search is possible either by control rod insertion or by homogeneous poisoning of the coolant. This report describes the main characterstics of the code structure and provides all the information needed to use the code. (Author)

  13. Boron neutron capture therapy using mixed epithermal and thermal neutron beams in patients with malignant glioma-correlation between radiation dose and radiation injury and clinical outcome

    International Nuclear Information System (INIS)

    Kageji, Teruyoshi; Nagahiro, Shinji; Matsuzaki, Kazuhito; Mizobuchi, Yoshifumi; Toi, Hiroyuki; Nakagawa, Yoshinobu; Kumada, Hiroaki

    2006-01-01

    Purpose: To clarify the correlation between the radiation dose and clinical outcome of sodium borocaptate-based intraoperative boron neutron capture therapy in patients with malignant glioma. Methods and Materials: The first protocol (P1998, n = 8) prescribed a maximal gross tumor volume (GTV) dose of 15 Gy. In 2001, a dose-escalated protocol was introduced (P2001, n 11), which prescribed a maximal vascular volume dose of 15 Gy or, alternatively, a clinical target volume (CTV) dose of 18 Gy. Results: The GTV and CTV doses in P2001 were 1.1-1.3 times greater than those in P1998. The maximal vascular volume dose of those with acute radiation injury was 15.8 Gy. The mean GTV and CTV dose in long-term survivors with glioblastoma was 26.4 and 16.5 Gy, respectively. A statistically significant correlation between the GTV dose and median survival time was found. In the 11 glioblastoma patients in P2001, the median survival time was 19.5 months and 1- and 2-year survival rate was 60.6% and 37.9%, respectively. Conclusion: Dose escalation contributed to the improvement in clinical outcome. To avoid radiation injury, the maximal vascular volume dose should be <12 Gy. For long-term survival in patients with glioblastoma after boron neutron capture therapy, the optimal mean dose of the GTV and CTV was 26 and 16 Gy, respectively

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-08-11

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

  15. Measurement of neutron dose equivalent outside and inside of the treatment vault of GRID therapy

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Xudong; Charlton, Michael A.; Esquivel, Carlos; Eng, Tony Y.; Li, Ying; Papanikolaou, Nikos [University of Texas Health Science Center, San Antonio, Texas 78229 (United States)

    2013-09-15

    Purpose: To evaluate the neutron and photon dose equivalent rates at the treatment vault entrance (H{sub n,D} and H{sub G}), and to study the secondary radiation to the patient in GRID therapy. The radiation activation on the grid was studied.Methods: A Varian Clinac 23EX accelerator was working at 18 MV mode with a grid manufactured by .decimal, Inc. The H{sub n,D} and H{sub G} were measured using an Andersson–Braun neutron REM meter, and a Geiger Müller counter. The radiation activation on the grid was measured after the irradiation with an ion chamber γ-ray survey meter. The secondary radiation dose equivalent to patient was evaluated by etched track detectors and OSL detectors on a RANDO{sup ®} phantom.Results: Within the measurement uncertainty, there is no significant difference between the H{sub n,D} and H{sub G} with and without a grid. However, the neutron dose equivalent to the patient with the grid is, on average, 35.3% lower than that without the grid when using the same field size and the same amount of monitor unit. The photon dose equivalent to the patient with the grid is, on average, 44.9% lower. The measured average half-life of the radiation activation in the grid is 12.0 (±0.9) min. The activation can be categorized into a fast decay component and a slow decay component with half-lives of 3.4 (±1.6) min and 15.3 (±4.0) min, respectively. There was no detectable radioactive contamination found on the surface of the grid through a wipe test.Conclusions: This work indicates that there is no significant change of the H{sub n,D} and H{sub G} in GRID therapy, compared with a conventional external beam therapy. However, the neutron and scattered photon dose equivalent to the patient decrease dramatically with the grid and can be clinical irrelevant. Meanwhile, the users of a grid should be aware of the possible high dose to the radiation worker from the radiation activation on the surface of the grid. A delay in handling the grid after the beam

  16. Contributions to the developments in apollo 2 neutron transport code: 1) scattering kernel taking into account the thermal agitation, resonance effect on energy transfers - 2) criteria for the construction of a multigroup mesh

    International Nuclear Information System (INIS)

    Ouisloumen, M.

    1989-03-01

    This research thesis reports an investigation of neutron diffusion by the fuel of a nuclear reactor while taking thermal agitation of diffusing nuclei into account. The author adopted the free gas model to describe the atom vibrations in the crystal network. He reports the elaboration of a diffusion law which gives the neutron energy spectrum after collision. Some particular theoretical cases are then studied in order to explain the observed physical phenomena, and to confirm the obtained results. The Monte Carlo method is also used to validate these results. The author presents the various physical criteria which must be met by an energetic grid in order to minimize errors in reaction rate calculations

  17. Qualitative dose response of the normal canine head to epithermal neutron irradiation with and without boron capture

    International Nuclear Information System (INIS)

    DeHaan, C.E.; Gavin, P.R.; Kraft, S.L.; Wheeler, F.J.; Atkinson, C.A.

    1992-01-01

    Boron Neutron Capture Therapy is being re-evaluated for the treatment of intracranial tumors. Prior to human clinical trials, determination of normal tissue tolerance is critical. Dogs were chosen as a large animal model for the following reasons. Dogs can be evaluated with advanced imaging, diagnostic and therapeutic modalities. Dogs are amenable to detailed neurologic examination and subtle behavioral changes are easily detected. Specifically, Labrador retrievers were chosen for their large body and head size. The dogs received varying doses of epithermal neutron irradiation and boron neutron capture irradiation using an epithermal neutron source. The dogs were closely monitored for up to one year post irradiation

  18. Spectral effects in low-dose fission and fusion neutron irradiated metals and alloys

    International Nuclear Information System (INIS)

    Heinisch, H.L.; Atkin, S.D.; Martinez, C.

    1986-04-01

    Flat miniature tensile specimens were irradiated to neutron fluences up to 9 x 10 22 n/m 2 in the RTNS-II and in the Omega West Reactor. Specimen temperatures were the same in both environments, with runs being made at both 90 0 C and 290 0 C. The results of tensile tests on AISI 316 stainless steel, A302B pressure vessel steel and pure copper are reported here. The radiation-induced changes in yield strength as a function of neutron dose in each spectrum are compared. The data for 316 stainless steel correlate well on the basis of displacements per atom (dpa), while those for copper and A302B do not. In copper the ratio of fission dpa to 14 MeV neutron dpa for a given yield stress change is about three to one. In A302B pressure vessel steel this ratio is more than three at lower fluences, but the yield stress data for fission and 14 MeV neutron-irradiated A302B steel appears to coalesce or intersect at the higher fluences

  19. The impact of ICRP 60 recommendations on the dose equivalent in low- and high energy neutron fields

    Energy Technology Data Exchange (ETDEWEB)

    Jakes, J; Schraube, H [GSF-Forschungszentrum Neuberg, D-85758 Oberschleissheim (Germany). Inst. fuer Strahlenschutz

    1996-12-31

    The objectives of this study was to determine the impact of the increased risk factors for neutrons after ICRP 60 on the operational dose equivalent quantities at a few neutron fields selected with the respect to cover the broad variety of neutron spectra: (1) Cadarache calibration assembly, with average neutron energy around 0.6 MeV, designed to simulate realistic neutron spectra at workplaces. This assembly is basically composed of an almost spherical {sup 238}U converter irradiated by 14.6 MeV neutrons from an accelerator target, placed at its center, and a scattering chamber consisting of a cylindrical polyethylene duct and a series of additional shieldings; (2) Neutron spectra at exposed workplaces in nuclear power plants; (3) Moderated spectra of {sup 252}Cf fission source; (4) Neutron spectra behind a shielding made of the iron (the average energy 5.,89 MeV) and concrete (the average energy 46.51 MeV), respectively; (5) Cosmic rays induced neutron spectra measured on the top of the Zugspitze (2968 m) where there is the average neutron energy around 40 MeV. From the derived neutron spectra, the mean quality factors and conversion factors h after ICRP 21 and ICRP 60, respectively, were calculated. The dose equivalent conversion factors were taken for the region below 20 MeV, and the energy region above 20 MeV. The results show that the operational quantities were affected predominately in the low energy fields, where the changes are given by a factor of 1,3 for the neutron fields given above. As has been expected, the impact of the new recommendations depends on the shape of the neutron spectra. Therefore, this factor can be much higher in the fields where the intermediate energy region is dominant, which is the case of moderated and scattered spectra at some places in the nuclear power plant and around containers with the spent fuel elements. (J.K.) 9 refs.

  20. Dose determination of Neutron contamination in radiothrapy rooms equiped with high energy linear accelerators

    International Nuclear Information System (INIS)

    Shweikani, R.; Anjak, O.

    2014-03-01

    Radiotherapy represents the most widely spread technique to control and treat cancer. To increase the treatment efficiency, high-energy linear accelerators are used. However, applying high energy photon beams leads to a non-negligible dose of neutrons contaminating therapeutic beams. A high-energy (23 MV) linear accelerator (Varian 21EX) was studied. The CR-39 nuclear track detectors (NTDs) were used to study the variation of fast neutron relative intensities around a linear accelerator high energy photon beam and to determined the its variation on the patient plane at 0, 50, 100, 150 and 200 cm from the center of the photon beam was. By increasing the distance from the center of the X-ray beam towards the periphery, the photoneutron dose equivalent decreased rapidly for the fields. Photoneutron intensity and distributions at isocenter level with the field sizes of 40*40 cm'2 at SSD=100cm around 23 MV photon beam using Nuclear Track Detectors were determined. The advantages of CR-39 NTD s over active detectors: 1- there is no pulse pileup problem. 2- no photon interference with neutron measurement. 3- no electronics are required. 4 - less prone to noise and interference. The photoneutron intensities were rapidly decreased as we move away from the isocenter of linear accelerators. As the use of simulation software MCNP match in the results we have obtained through direct measurements and the modeling results using the code MCNP (author).

  1. Comparison of neutron dose measured by Albedo TLD and etched tracks detector at PNC plutonium fuel facilities

    International Nuclear Information System (INIS)

    Tsujimura, N.; Momose, T.; Shinohara, K.; Ishiguro, H.

    1996-01-01

    Power Reactor and Nuclear Fuel Development Corporation (PNC) has fabricated Plutonium and Uranium Mixed OXide (MOX) fuel for FBR MONJU at Tokai works. In this site, PNC/Panasonic albedo TLDs/1/ are used for personnel neutron monitoring. And a part of workers wore Etched Tracks Detector (ETD) combined with TLD in order to check the accuracy of the neutron dose estimated by albedo TLD. In this paper, the neutron dose measured by TLD and ETD in the routine monitoring is compared at PNC plutonium fuel facilities. (author)

  2. Verification and validation of multi-group library MUSE1.0 created from ENDF/B-VII.0

    International Nuclear Information System (INIS)

    Chen Yixue; Wu Jun; Yang Shouhai; Zhang Bin; Lu Daogang; Chen Chaobin

    2010-01-01

    A multi-group library set named MUSE1.0 with 172-neutron group and 42-photon group is produced based on ENDF/B-VII.0 using NJOY code. Weight function of the multi-group library set is taken from the Vitanim-e library and the max legendre order of scattering matrix is six. All the nuclides have thermal scattering data created using free-gas scattering law and 10 Bondarenko background cross sections se lected to generate the self-shielded multi-group cross sections. The final libraries have GENDF-format, MATXS-format and ACE-multi-group sub-libraries and each sub-library generated under 4 temperatures(293 K,600 K,800 K and 900 K). This paper provides a summary of the procedure to produce the library set and a detail description of the validation of the multi-group library set by several critical benchmark devices and shielding benchmark devices using MCNP code. The ability to handle the thermal neutron transport and resonance self-shielding problems are investigated specially. In the end, we draw the conclusion that the multi-group libraries produced is credible and can be used in the R and D process of Supercritical Water Reactor Design. (authors)

  3. Dose distribution and clinical response of glioblastoma treated with boron neutron capture therapy

    Energy Technology Data Exchange (ETDEWEB)

    Matsuda, M. [Department of Neurosurgery, Graduate School of Comprehensive Human Science, University of Tsukuba, Tennodai 1-1-1, Tsukuba (Japan)], E-mail: mhide-m@gk9.so-net.ne.jp; Yamamoto, T. [Department of Neurosurgery, Graduate School of Comprehensive Human Science, University of Tsukuba, Tennodai 1-1-1, Tsukuba (Japan); Kumada, H. [Japan Atomic Energy Agency, Shirakatashirane 2-4, Tokai (Japan); Nakai, K.; Shirakawa, M.; Tsurubuchi, T.; Matsumura, A. [Department of Neurosurgery, Graduate School of Comprehensive Human Science, University of Tsukuba, Tennodai 1-1-1, Tsukuba (Japan)

    2009-07-15

    The dose distribution and failure pattern after treatment with the external beam boron neutron capture therapy (BNCT) protocol were retrospectively analyzed. BSH (5 g/body) and BPA (250 mg/kg) based BNCT was performed in eight patients with newly diagnosed glioblastoma. The gross tumor volume (GTV) and clinical target volume (CTV)-1 were defined as the residual gadolinium-enhancing volume. CTV-2 and CTV-3 were defined as GTV plus a margin of 2 and 3 cm, respectively. As additional photon irradiation, a total X-ray dose of 30 Gy was given to the T2 high intensity area on MRI. Five of the eight patients were alive at analysis for a mean follow-up time of 20.3 months. The post-operative median survival time of the eight patients was 27.9 months (95% CI=21.0-34.8). The minimum tumor dose of GTV, CTV-2, and CTV-3 averaged 29.8{+-}9.9, 15.1{+-}5.4, and 12.4{+-}2.9 Gy, respectively. The minimum tumor non-boron dose of GTV, CTV-2, and CTV-3 averaged 2.0{+-}0.5, 1.3{+-}0.3, and 1.1{+-}0.2 Gy, respectively. The maximum normal brain dose, skin dose, and average brain dose were 11.4{+-}1.5, 9.6{+-}1.4, and 3.1{+-}0.4 Gy, respectively. The mean minimum dose at the failure site in cases of in-field recurrence (IR) and out-field recurrence (OR) was 26.3{+-}16.7 and 14.9 GyEq, respectively. The calculated doses at the failure site were at least equal to the tumor control doses which were previously reported. We speculate that the failure pattern was related to an inadequate distribution of boron-10. Further improvement of the microdistribution of boron compounds is expected, and may improve the tumor control by BNCT.

  4. Evaluation of neutron irradiation fields for BNCT by using absorbed dose in a phantom

    International Nuclear Information System (INIS)

    Aizawa, O.

    1993-01-01

    In a previous paper, the author defined the open-quotes irradiation timeclose quotes as the time of irradiation in which the maximum open-quotes total background doseclose quotes becomes 2,500 RBE-cGy. In this paper, he has modified the definition a little as the time of irradiation in which the maximum open-quotes lμg/g B-10 doseclose quotes becomes 3,000 RBE-cGy, because he assumed that normal tissue contained 1μg/g B-10. Moreover, he has modified the dose criteria for BNCT as follows: The open-quotes eye doseclose quotes, open-quotes total body doseclose quotes and open-quotes except-head doseclose quotes should be less that 200, 100 and 50 RBE-cGy, respectively. He has added one more criterion for BNCT that the thermal neutron fluence at the tumor position should be over 2.5x10 12 n/cm 2 at the open-quotes irradiation timeclose quotes. The distance from the core side to the irradiation port in the open-quotes old configurationclose quotes of the Musashi reactor (TRIGA-II, 100kW) was 160 cm. He is now planning to design an eccentric core and to move the reactor core nearer to the irradiation port, distance between the core side and the irradiation port to be 140, 130 and 120cm. The other assumptions used in this paper are as follows: (1) The B-10 concentrations in tumor are 30 and/or 10μg/g. (2) The depth of the tumor is 5.0 cm to 5.5 cm from the surface. (3) The RBE values used are 1.0 for all gamma rays and 2.3 for B 10 (n,α) reaction products. (4) The RBE values for neutrons are the following three cases: the first case is using 1.6 for all neutrons; the second one is using 3.2 for non-thermal neutrons and 1.6 for thermal neutrons; the third case is using 4.8 for fast neutrons, 3.2 for faster epithermal and epithermal neutrons, and 1.6 for thermal neutrons

  5. A code system to generate multigroup cross-sections using basic data

    International Nuclear Information System (INIS)

    Garg, S.B.; Kumar, Ashok

    1978-01-01

    For the neutronic studies of nuclear reactors, multigroup cross-sections derived from the basic energy point data are needed. In order to carry out the design based studies, these cross-sections should also incorporate the temperature and fuel concentration effects. To meet these requirements, a code system comprising of RESRES, UNRES, FIGERO, INSCAT, FUNMO, AVER1 and BGPONE codes has been adopted. The function of each of these codes is discussed. (author)

  6. Solution of the multigroup diffusion equation for two-dimensional triangular regions by finite Fourier transformation

    International Nuclear Information System (INIS)

    Takeshi, Y.; Keisuke, K.

    1983-01-01

    The multigroup neutron diffusion equation for two-dimensional triangular geometry is solved by the finite Fourier transformation method. Using the zero-th-order equation of the integral equation derived by this method, simple algebraic expressions for the flux are derived and solved by the alternating direction implicit method. In sample calculations for a benchmark problem of a fast breeder reactor, it is shown that the present method gives good results with fewer mesh points than the usual finite difference method

  7. Survey of computer codes which produce multigroup data from ENDF/B-IV

    International Nuclear Information System (INIS)

    Greene, N.M.

    1975-01-01

    The features of three code systems that produce multigroup neutron data are contrasted. This includes the ETOE-2/MC 2 -2/SDX, MINX/SPHINX and AMPX code packages. These systems all contain a fairly extensive set of processing capabilities with the current evaluated nuclear data files--ENDF/B. They were designed with different goals and applications in mind. This paper discusses some of their differences and the implications for particular situations

  8. New Multi-group Transport Neutronics (PHISICS) Capabilities for RELAP5-3D and its Application to Phase I of the OECD/NEA MHTGR-350 MW Benchmark

    Energy Technology Data Exchange (ETDEWEB)

    Gerhard Strydom; Cristian Rabiti; Andrea Alfonsi

    2012-10-01

    PHISICS is a neutronics code system currently under development at the Idaho National Laboratory (INL). Its goal is to provide state of the art simulation capability to reactor designers. The different modules for PHISICS currently under development are a nodal and semi-structured transport core solver (INSTANT), a depletion module (MRTAU) and a cross section interpolation (MIXER) module. The INSTANT module is the most developed of the mentioned above. Basic functionalities are ready to use, but the code is still in continuous development to extend its capabilities. This paper reports on the effort of coupling the nodal kinetics code package PHISICS (INSTANT/MRTAU/MIXER) to the thermal hydraulics system code RELAP5-3D, to enable full core and system modeling. This will enable the possibility to model coupled (thermal-hydraulics and neutronics) problems with more options for 3D neutron kinetics, compared to the existing diffusion theory neutron kinetics module in RELAP5-3D (NESTLE). In the second part of the paper, an overview of the OECD/NEA MHTGR-350 MW benchmark is given. This benchmark has been approved by the OECD, and is based on the General Atomics 350 MW Modular High Temperature Gas Reactor (MHTGR) design. The benchmark includes coupled neutronics thermal hydraulics exercises that require more capabilities than RELAP5-3D with NESTLE offers. Therefore, the MHTGR benchmark makes extensive use of the new PHISICS/RELAP5-3D coupling capabilities. The paper presents the preliminary results of the three steady state exercises specified in Phase I of the benchmark using PHISICS/RELAP5-3D.

  9. Multigroup P8 - elastic scattering matrices of main reactor elements

    International Nuclear Information System (INIS)

    Garg, S.B.; Shukla, V.K.

    1979-01-01

    To study the effect of anisotropic scattering phenomenon on shielding and neutronics of nuclear reactors multigroup P8-elastic scattering matrices have been generated for H, D, He, 6 Li, 7 Li, 10 B, C, N, O, Na, Cr, Fe, Ni, 233 U, 235 U, 238 U, 239 Pu, 240 Pu, 241 Pu and 242 Pu using their angular distribution, Legendre coefficient and elastic scattering cross-section data from the basic ENDF/B library. Two computer codes HSCAT and TRANS have been developed to complete this task for BESM-6 and CDC-3600 computers. These scattering matrices can be directly used as input to the transport theory codes ANISN and DOT. (auth.)

  10. Calculation of dose distribution for 252Cf fission neutron source in tissue equivalent phantoms using Monte Carlo method

    International Nuclear Information System (INIS)

    Ji Gang; Guo Yong; Luo Yisheng; Zhang Wenzhong

    2001-01-01

    Objective: To provide useful parameters for neutron radiotherapy, the author presents results of a Monte Carlo simulation study investigating the dosimetric characteristics of linear 252 Cf fission neutron sources. Methods: A 252 Cf fission source and tissue equivalent phantom were modeled. The dose of neutron and gamma radiations were calculated using Monte Carlo Code. Results: The dose of neutron and gamma at several positions for 252 Cf in the phantom made of equivalent materials to water, blood, muscle, skin, bone and lung were calculated. Conclusion: The results by Monte Carlo methods were compared with the data by measurement and references. According to the calculation, the method using water phantom to simulate local tissues such as muscle, blood and skin is reasonable for the calculation and measurements of dose distribution for 252 Cf

  11. Self-ion emulation of high dose neutron irradiated microstructure in stainless steels

    Science.gov (United States)

    Jiao, Z.; Michalicka, J.; Was, G. S.

    2018-04-01

    Solution-annealed 304L stainless steel (SS) was irradiated to 130 dpa at 380 °C, and to 15 dpa at 500 °C and 600 °C, and cold-worked 316 SS (CW 316 SS) was irradiated to 130 dpa at 380 °C using 5 MeV Fe++/Ni++ to produce microstructures and radiation-induced segregation (RIS) for comparison with that from neutron irradiation at 320 °C to 46 dpa in the BOR60 reactor. For the 304L SS alloy, self-ion irradiation at 380 °C produced a dislocation loop microstructure that was comparable to that by neutron irradiation. No voids were observed in either the 380 °C self-ion irradiation or the neutron irradiation conditions. Irradiation at 600 °C produced the best match to radiation-induced segregation of Cr and Ni with the neutron irradiation, consistent with the prediction of a large temperature shift by Mansur's invariant relations for RIS. For the CW 316 SS alloy irradiated to 130 dpa at 380 °C, both the irradiated microstructure (dislocation loops, precipitates and voids) and RIS reasonably matched the neutron-irradiated sample. The smaller temperature shift for RIS in CW 316 SS was likely due to the high sink (dislocation) density induced by the cold work. A single self-ion irradiation condition at a dose rate ∼1000× that in reactor does not match both dislocation loops and RIS in solution-annealed 304L SS. However, a single irradiation temperature produced a reasonable match with both the dislocation/precipitate microstructure and RIS in CW 316 SS, indicating that sink density is a critical factor in determining the temperature shift for self-ion irradiations.

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

    Science.gov (United States)

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

    2014-01-01

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

  13. In vivo transcriptome modulation after low dose of high energy neutron irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Amendola, R; Fratini, E; Piscitelli, M; Sallustio, D E [ENEA, BAS BIOTEC MED, Roma (Italy); Angelone, M; Pillon, M [ENEA, FUS TEC, Frascati (Italy); Chiani, F; Licursi, V; Negri, R [Universita La Sapienza, Roma (Italy). Dip. Biologia Cellulare e dello Sviluppo

    2007-07-01

    Complete text of publication follows. Objective: This project aims to the identification of an hypothetical transcriptome modulation of mouse peripheral blood lymphocytes and skin after exposure to high energy neutron in vivo. Positive candidate genes isolated from mice in in vivo experiments will be selected and evaluated for both radioprotection issues dealing with cosmic ray exposure, and for biomedical issues mainly for low doses and non-cancer effects. Methods: High energy neutron irradiation is performed at the ENEA Frascati, neutron generator facilities (FNG), specifically dedicated to biological samples. FNG is a linear electrostatic accelerator that produces up to 1.0 x 10{sup 11} n/s 14 MeV neutrons via the D-T nuclear reaction. The dose-rate applied for this study is of 0.7 cGy/min. The functional genomic approach has been performed on six animals for each experimental points: un-irradiated; 20 cGy, 6 hours and 24 hours delayed time after exposure. Preliminarily, a pool of total RNA is evaluated on commercial micro-arrays containing large collections of mus musculus cDNAs. Statistical filtering and functional clustering of the data is carried out using dedicated software packages. Results: Candidate genes are selected on the basis of responsiveness to 20 cGy of exposure, with a defined temporal regulation. We plan to organize a systematic screen focused on genes responding to our selection criteria, in in vivo mouse experiments, and correlate their differential expression to the human counterparts. A specific cross species database will be created with all the functional information available in standardized format (MIAME: minimal information about micro-arrays experiments). Conclusions: A lack of information on in vivo experiments is still evident for low doses exposure, especially for neutron of cosmic interest. Individual susceptibility, extensive number of animals to be processed, lack of standardization methodologies are among problems to be solved

  14. The effect of a paraffin screen on the neutron dose at the maze door of a 15 MV linear accelerator.

    Science.gov (United States)

    Krmar, M; Nikolić, D; Kuzmanović, A; Kuzmanović, Z; Ganezer, K

    2013-08-01

    The purpose of this study was to explore the effects of a paraffin screen located at various positions in the maze on the neutron dose equivalent at the maze door. The neutron dose equivalent was measured at the maze door of a room containing a 15 MV linear accelerator for x-ray therapy. Measurements were performed for several positions of the paraffin screen covering only 27.5% of the cross-sectional area of the maze. The neutron dose equivalent was also measured at all screen positions. Two simple models of the neutron source were considered in which the first assumed that the source was the cross-sectional area at the inner entrance of the maze, radiating neutrons in an isotropic manner. In the second model the reduction in the neutron dose equivalent at the maze door due to the paraffin screen was considered to be a function of the mean values of the neutron fluence and energy at the screen. The results of this study indicate that the equivalent dose at the maze door was reduced by a factor of 3 through the use of a paraffin screen that was placed inside the maze. It was also determined that the contributions to the dosage from areas that were not covered by the paraffin screen as viewed from the dosimeter, were 2.5 times higher than the contributions from the covered areas. This study also concluded that the contributions of the maze walls, ceiling, and floor to the total neutron dose equivalent were an order of magnitude lower than those from the surface at the far end of the maze. This study demonstrated that a paraffin screen could be used to reduce the neutron dose equivalent at the maze door by a factor of 3. This paper also found that the reduction of the neutron dose equivalent was a linear function of the area covered by the maze screen and that the decrease in the dose at the maze door could be modeled as an exponential function of the product φ·E at the screen.

  15. Angular finite volume method for solving the multigroup transport equation with piecewise average scattering cross sections

    Energy Technology Data Exchange (ETDEWEB)

    Calloo, A.; Vidal, J.F.; Le Tellier, R.; Rimpault, G., E-mail: ansar.calloo@cea.fr, E-mail: jean-francois.vidal@cea.fr, E-mail: romain.le-tellier@cea.fr, E-mail: gerald.rimpault@cea.fr [CEA, DEN, DER/SPRC/LEPh, Saint-Paul-lez-Durance (France)

    2011-07-01

    This paper deals with the solving of the multigroup integro-differential form of the transport equation for fine energy group structure. In that case, multigroup transfer cross sections display strongly peaked shape for light scatterers and the current Legendre polynomial expansion is not well-suited to represent them. Furthermore, even if considering an exact scattering cross sections representation, the scattering source in the discrete ordinates method (also known as the Sn method) being calculated by sampling the angular flux at given directions, may be wrongly computed due to lack of angular support for the angular flux. Hence, following the work of Gerts and Matthews, an angular finite volume solver has been developed for 2D Cartesian geometries. It integrates the multigroup transport equation over discrete volume elements obtained by meshing the unit sphere with a product grid over the polar and azimuthal coordinates and by considering the integrated flux per solid angle element. The convergence of this method has been compared to the S{sub n} method for a highly anisotropic benchmark. Besides, piecewise-average scattering cross sections have been produced for non-bound Hydrogen atoms using a free gas model for thermal neutrons. LWR lattice calculations comparing Legendre representations of the Hydrogen scattering multigroup cross section at various orders and piecewise-average cross sections for this same atom are carried out (while keeping a Legendre representation for all other isotopes). (author)

  16. Angular finite volume method for solving the multigroup transport equation with piecewise average scattering cross sections

    International Nuclear Information System (INIS)

    Calloo, A.; Vidal, J.F.; Le Tellier, R.; Rimpault, G.

    2011-01-01

    This paper deals with the solving of the multigroup integro-differential form of the transport equation for fine energy group structure. In that case, multigroup transfer cross sections display strongly peaked shape for light scatterers and the current Legendre polynomial expansion is not well-suited to represent them. Furthermore, even if considering an exact scattering cross sections representation, the scattering source in the discrete ordinates method (also known as the Sn method) being calculated by sampling the angular flux at given directions, may be wrongly computed due to lack of angular support for the angular flux. Hence, following the work of Gerts and Matthews, an angular finite volume solver has been developed for 2D Cartesian geometries. It integrates the multigroup transport equation over discrete volume elements obtained by meshing the unit sphere with a product grid over the polar and azimuthal coordinates and by considering the integrated flux per solid angle element. The convergence of this method has been compared to the S_n method for a highly anisotropic benchmark. Besides, piecewise-average scattering cross sections have been produced for non-bound Hydrogen atoms using a free gas model for thermal neutrons. LWR lattice calculations comparing Legendre representations of the Hydrogen scattering multigroup cross section at various orders and piecewise-average cross sections for this same atom are carried out (while keeping a Legendre representation for all other isotopes). (author)

  17. Evaluation of the HTTR criticality and burnup calculations with continuous-energy and multigroup cross sections

    Energy Technology Data Exchange (ETDEWEB)

    Chiang, Min-Han; Wang, Jui-Yu [Institute of Nuclear Engineering and Science, National Tsing Hua University, 101, Section 2, Kung-Fu Road, Hsinchu 30013, Taiwan (China); Sheu, Rong-Jiun, E-mail: rjsheu@mx.nthu.edu.tw [Institute of Nuclear Engineering and Science, National Tsing Hua University, 101, Section 2, Kung-Fu Road, Hsinchu 30013, Taiwan (China); Department of Engineering System and Science, National Tsing Hua University, 101, Section 2, Kung-Fu Road, Hsinchu 30013, Taiwan (China); Liu, Yen-Wan Hsueh [Institute of Nuclear Engineering and Science, National Tsing Hua University, 101, Section 2, Kung-Fu Road, Hsinchu 30013, Taiwan (China); Department of Engineering System and Science, National Tsing Hua University, 101, Section 2, Kung-Fu Road, Hsinchu 30013, Taiwan (China)

    2014-05-01

    The High Temperature Engineering Test Reactor (HTTR) in Japan is a helium-cooled graphite-moderated reactor designed and operated for the future development of high-temperature gas-cooled reactors. Two detailed full-core models of HTTR have been established by using SCALE6 and MCNP5/X, respectively, to study its neutronic properties. Several benchmark problems were repeated first to validate the calculation models. Careful code-to-code comparisons were made to ensure that two calculation models are both correct and equivalent. Compared with experimental data, the two models show a consistent bias of approximately 20–30 mk overestimation in effective multiplication factor for a wide range of core states. Most of the bias could be related to the ENDF/B-VII.0 cross-section library or incomplete modeling of impurities in graphite. After that, a series of systematic analyses was performed to investigate the effects of cross sections on the HTTR criticality and burnup calculations, with special interest in the comparison between continuous-energy and multigroup results. Multigroup calculations in this study were carried out in 238-group structure and adopted the SCALE double-heterogeneity treatment for resonance self-shielding. The results show that multigroup calculations tend to underestimate the system eigenvalue by a constant amount of ∼5 mk compared to their continuous-energy counterparts. Further sensitivity studies suggest the differences between multigroup and continuous-energy results appear to be temperature independent and also insensitive to burnup effects.

  18. Evaluation of the HTTR criticality and burnup calculations with continuous-energy and multigroup cross sections

    International Nuclear Information System (INIS)

    Chiang, Min-Han; Wang, Jui-Yu; Sheu, Rong-Jiun; Liu, Yen-Wan Hsueh

    2014-01-01

    The High Temperature Engineering Test Reactor (HTTR) in Japan is a helium-cooled graphite-moderated reactor designed and operated for the future development of high-temperature gas-cooled reactors. Two detailed full-core models of HTTR have been established by using SCALE6 and MCNP5/X, respectively, to study its neutronic properties. Several benchmark problems were repeated first to validate the calculation models. Careful code-to-code comparisons were made to ensure that two calculation models are both correct and equivalent. Compared with experimental data, the two models show a consistent bias of approximately 20–30 mk overestimation in effective multiplication factor for a wide range of core states. Most of the bias could be related to the ENDF/B-VII.0 cross-section library or incomplete modeling of impurities in graphite. After that, a series of systematic analyses was performed to investigate the effects of cross sections on the HTTR criticality and burnup calculations, with special interest in the comparison between continuous-energy and multigroup results. Multigroup calculations in this study were carried out in 238-group structure and adopted the SCALE double-heterogeneity treatment for resonance self-shielding. The results show that multigroup calculations tend to underestimate the system eigenvalue by a constant amount of ∼5 mk compared to their continuous-energy counterparts. Further sensitivity studies suggest the differences between multigroup and continuous-energy results appear to be temperature independent and also insensitive to burnup effects

  19. Neutron dose rate at the SwissFEL injector test facility: first measurements

    International Nuclear Information System (INIS)

    Hohmann, E.; Frey, N.; Fuchs, A.; Harm, C.; Hoedlmoser, H.; Luescher, R.; Mayer, S.; Morath, O.; Philipp, R.; Rehmann, A.; Schietinger, T.

    2014-01-01

    At the Paul Scherrer Institute, the new SwissFEL Free Electron Laser facility is currently in the design phase. It is foreseen to accelerate electrons up to a maximum energy of 7 GeV with a pulsed time structure. An injector test facility is operated at a maximum energy of 300 MeV and serves as the principal test and demonstration plant for the SwissFEL project. Secondary radiation is created in unavoidable interactions of the primary beam with beamline components. The resulting ambient dose-equivalent rate due to neutrons was measured along the beamline with different commercially available survey instruments. The present study compares the readings of these neutron detectors (one of them is specifically designed for measurements in pulsed fields). The experiments were carried out in both, a normal and a diagnostic mode of operation of the injector. Measurements were taken at the SwissFEL injector test facility using three different types of commercially available survey instruments for normal and diagnostic mode of operation at different positions inside the accelerator vault. During normal operation, the doses indicated by the different instruments agree within the measurement uncertainty except for the beam dump region. There, due to its limited energy range and high sensitivity, the LB6411 shows significantly lower dose values than the other instruments. The photon background in the vault associated with each pulse causes the scintillator used by the LB6419 to saturate. As a result, only the channel using the delayed 12 C(n,p)12-reaction could be used during the measurements. The highest doses per pulse were measured next to the beam dump and the bunch compressor. For the optimisation of the accelerator, luminescent screens can be inserted into the beam path causing a dose distributed over several metres depending on the screen type. The dose arise to 40 % from neutrons with energies of >20 MeV. Although the charge of each pulse were reduced to decrease

  20. Influence of the neutron flux shape on the value of absorbed neutron dose; Uticaj oblika neutronskog spektra na vrednost apsorbovane doze neutrona

    Energy Technology Data Exchange (ETDEWEB)

    Miric, I; Miric, P [Institute of nuclear sciences Boris Kidric, Vinca, Beograd (Yugoslavia)

    1974-07-01

    This paper deals with the study od specific doses dependence on the type and approximation procedures of neutron spectra. Values of specific dose rates (dose per neutron cm{sub 2}) were analysed for neutron spectra from RB reactor in Vinca, Crac facility in Valduc (France) and HPRR reactor in Oak Ridge (USA). Data used in this analysis were obtained by methods used in Harwell (AERE), Oak Ridge (ORNL), Chalk River (AECL), CEN de Cadarache (CEA) and in the Boris Kidric Institute (IBK). Specific absorbed neutron doses were determined for each of the estimated spectra and presented in the form of kerma/(n.cm{sup -2}) and rad/((n.cm{sup -2}) units. The obtained results have shown the influence of the flux approximation procedure on the values of conversion factors for obtaining neutron doses from neutron flux. U okviru ovog rada radjeno je na ispitivanju zavisnosti specificnih doza od vrste i nacina aproksimacije neutronskog spektra. U radu su analizirane vrednosti specificnih doza (doza po n.cm{sup -2}) za neutronske spektre koji se dobijaju oko sledecih nuklearnih postrojenja: reaktora RB u Vinci, postrojenja CRAC u Valduc-u (Francuska), reaktora HPRR u Oak Ridge-u (SAD). Za analizu su korisceni podaci dobijeni metodama koje se koriste u nuklearnim centrima Harwell (AERE), Oak Ridge-u (ORNL), Chalk River-u (AECL), CEN de Cadarache (CEA) i Institutu Boris Kidric (IBK). Za svaki procenjeni spektar odredjene su specificne apsorbovane doze neutrona izrazene u kerma/(n.cm{sup -2}) i rad/(n.cm{sup -2}) jedinicama. Dobijeni rezultati su pokazali koliko nacin aproksimacije spektra utice na vrednost konverzionih faktora koji sluze za prelazak sa fluksa na dozu neutrona (author)

  1. Determination of the energy dose of a neutron beam using a ionization chamber. A compendium

    International Nuclear Information System (INIS)

    Schraube, H.; Alberts, W.G.; Brede, H.; Burgkhardt, B.; Piesch, E.; Doerschel, B.; Heinzelmann, M.; Hess, A.; Hoefert, M.

    2003-01-01

    This report is addressed to scientists and technicians, who are engaged in dosimetry of fast neutrons, especially for purposes of percutaneous radiation therapy. The range of mean energies of the radiation sources may be approximately between 1 MeV and 50 MeV. The report exhibits a compendium, which describes the basic methods and procedures for the determination of energy dose in tissue in a phantom or free-in-air, where applicable. Furthermore, requirements for monitor, test devices and phantom materials are described. The calculation methods are comprehensibly derived and supplemented with numerical data. A detailed analysis of experimental uncertainties is completed with practical examples. (orig.) [de

  2. Response functions for computing absorbed dose to skeletal tissues from neutron irradiation

    Science.gov (United States)

    Bahadori, Amir A.; Johnson, Perry; Jokisch, Derek W.; Eckerman, Keith F.; Bolch, Wesley E.

    2011-11-01

    Spongiosa in the adult human skeleton consists of three tissues—active marrow (AM), inactive marrow (IM) and trabecularized mineral bone (TB). AM is considered to be the target tissue for assessment of both long-term leukemia risk and acute marrow toxicity following radiation exposure. The total shallow marrow (TM50), defined as all tissues lying within the first 50 µm of the bone surfaces, is considered to be the radiation target tissue of relevance for radiogenic bone cancer induction. For irradiation by sources external to the body, kerma to homogeneous spongiosa has been used as a surrogate for absorbed dose to both of these tissues, as direct dose calculations are not possible using computational phantoms with homogenized spongiosa. Recent micro-CT imaging of a 40 year old male cadaver has allowed for the accurate modeling of the fine microscopic structure of spongiosa in many regions of the adult skeleton (Hough et al 2011 Phys. Med. Biol. 56 2309-46). This microstructure, along with associated masses and tissue compositions, was used to compute specific absorbed fraction (SAF) values for protons originating in axial and appendicular bone sites (Jokisch et al 2011 Phys. Med. Biol. 56 6857-72). These proton SAFs, bone masses, tissue compositions and proton production cross sections, were subsequently used to construct neutron dose-response functions (DRFs) for both AM and TM50 targets in each bone of the reference adult male. Kerma conditions were assumed for other resultant charged particles. For comparison, AM, TM50 and spongiosa kerma coefficients were also calculated. At low incident neutron energies, AM kerma coefficients for neutrons correlate well with values of the AM DRF, while total marrow (TM) kerma coefficients correlate well with values of the TM50 DRF. At high incident neutron energies, all kerma coefficients and DRFs tend to converge as charged-particle equilibrium is established across the bone site. In the range of 10 eV to 100 Me

  3. Attenuation of the neutron and γ ray dose in concrete channels

    International Nuclear Information System (INIS)

    Paratte, J.M.

    1983-08-01

    The calculations of the γ and neutron dose in concrete channels is described. The method is based on the Monte Carlo procedure. One series of results obtained in straight channels shows the influence of the source spectra and geometry and thus the channel form. A second series shows the attenuation produced by bends along the length of the channel; the variation of the branch length is also studied. The results are generalised and represented by a simple formula. The parameters are adjusted to the curves obtained by the Monte Carlo programme. (G.T.H.)

  4. Radiation doses from radiation sources of neutrons and photons by different computer calculation

    International Nuclear Information System (INIS)

    Siciliano, F.; Lippolis, G.; Bruno, S.G.

    1995-12-01

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

  5. Crystal growth in EPDM by chemi-crystallisation as a function of the neutron irradiation dose and flux level

    International Nuclear Information System (INIS)

    Lambri, O.A.; Salvatierra, L.M.; Sanchez, F.A.; Matteo, C.L.; Sorichetti, P.A.; Celauro, C.A.

    2005-01-01

    Neutron irradiation at room temperature were performed on EPDM (ethylene-propylene-diene monomer) in two different nuclear reactors at different fluxes. The effect of the irradiation on the chain arrangement in the polymer, as a function of the dose is discussed. Different crystal concentrations and crystal shapes, developed by chemi-crystallisation, are obtained depending on the neutron dose. In addition the radiation damage degree in the polymer depends both on the dose and the flux level. Dynamical mechanical analysis, swelling studies, X-ray diffraction, differential thermal analysis and infrared studies were employed as experimental techniques

  6. Neutron dose rate in the upper part of a PWR containment. Comparison between measurements and TRIPOLI-2 calculations

    International Nuclear Information System (INIS)

    Vergnaud, T.; Bourdet, L.; Gonnord, J.; Nimal, J.C.; Champion, G.

    1984-01-01

    Conception of a reactor building requires large openings in the primary concrete shield for a postulated loss-of-coolant accident. Through these openings neutrons escape and produce dose rates in several parts of the reactor building. Some calculations using ANISN, DOT and essentially TRIPOLI-2 codes allow to compute the neutron dose rates at several places such as reactor containment operating floor and containment annulus. Some complementary shields are provided and the instrumentations are placed in area where the dose rate is lower. Comparisons are presented between measurements and calculations

  7. Neutron dose study with bubble detectors aboard the International Space Station as part of the Matroshka-R experiment

    International Nuclear Information System (INIS)

    Machrafi, R.; Garrow, K.; Ing, H.; Smith, M. B.; Andrews, H. R.; Akatov, Yu; Arkhangelsky, V.; Chernykh, I.; Mitrikas, V.; Petrov, V.; Shurshakov, V.; Tomi, L.; Kartsev, I.; Lyagushin, V.

    2009-01-01

    As part of the Matroshka-R experiments, a spherical phantom and space bubble detectors (SBDs) were used on board the International Space Station to characterise the neutron radiation field. Seven experimental sessions with SBDs were carried out during expeditions ISS-13, ISS-14 and ISS-15. The detectors were positioned at various places throughout the Space Station, in order to determine dose variations with location and on/in the phantom in order to establish the relationship between the neutron dose measured externally to the body and the dose received internally. Experimental data on/in the phantom and at different locations are presented. (authors)

  8. Capability of NIPAM polymer gel in recording dose from the interaction of 10B and thermal neutron in BNCT

    International Nuclear Information System (INIS)

    Khajeali, Azim; Reza Farajollahi, Ali; Kasesaz, Yaser; Khodadadi, Roghayeh; Khalili, Assef; Naseri, Alireza

    2015-01-01

    The capability of N-isopropylacrylamide (NIPAM) polymer gel to record the dose resulting from boron neutron capture reaction in BNCT was determined. In this regard, three compositions of the gel with different concentrations of 10 B were prepared and exposed to gamma radiation and thermal neutrons. Unlike irradiation with gamma rays, the boron-loaded gels irradiated by neutron exhibited sensitivity enhancement compared with the gels without 10 B. It was also found that the neutron sensitivity of the gel increased by the increase of concentration of 10 B. It can be concluded that NIPAM gel might be suitable for the measurement of the absorbed dose enhancement due to 10 B and thermal neutron reaction in BNCT. - Highlights: • Three compositions of NIPAM gel with different concentration of 10 B have been exposed by gamma and thermal neutron. • The vials containing NIPAM gel have been irradiated by an automatic system capable of providing for dose uniformity. • Suitability of NIPAM polymer gel in measuring radiation doses in BNCT has been investigated.

  9. Sequential measurements of cosmic-ray neutron spectrum and dose rate at sea level in Sendai, Japan

    International Nuclear Information System (INIS)

    Nakamura, Takashi; Nunomiya, Tomoya; Abe, Shigeru; Terunuma, Kazutaka; Suzuki, Hiroyuki

    2005-01-01

    The cosmic-ray neutron energy spectrum and dose rate were measured sequentially for two years from April 2001 up to March 2003 by using three neutron detectors, a 3 He-loaded multi-moderator detector (Bonner ball), 12.7 cm diameter by 12.7 cm long NE213 organic liquid scintillator, and high-sensitivity rem (dose equivalent) counter at the Kawauchi campus of Tohoku University in Sendai, Japan of geomagnetic latitude, 29degN, and cutoff rigidity, 10.43 GV. The neutron spectrum has three major peaks, thermal energy peak, evaporation peak around 1 MeV and cascade peak around 100 MeV. The ambient neutron dose equivalent rates measured by the rem counter, and the Bonner ball keep almost constant values of 4.0 and 6.5 (nSv/h), respectively, throughout this time period, after atmospheric pressure correction, and it often decreased about 30% after a large Solar Flare, that is called as the Forbush decrease. The total neutron flux was also obtained by the Bonner ball measurements to be 7.5x10 -3 (ncm -2 ·s -1 ) in average. The altitude variation of neutron flux and dose was also investigated by comparing the measured results with other results measured at Mt. Fuji area and aboard an airplane, where the cutoff rigidities are similar. (author)

  10. High-Dose Neutron Detector Development Using 10B Coated Cells

    Energy Technology Data Exchange (ETDEWEB)

    Menlove, Howard Olsen [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Henzlova, Daniela [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2016-11-08

    During FY16 the boron-lined parallel-plate technology was optimized to fully benefit from its fast timing characteristics in order to enhance its high count rate capability. To facilitate high count rate capability, a novel fast amplifier with timing and operating properties matched to the detector characteristics was developed and implemented in the 8” boron plate detector that was purchased from PDT. Each of the 6 sealed-cells was connected to a fast amplifier with corresponding List mode readout from each amplifier. The FY16 work focused on improvements in the boron-10 coating materials and procedures at PDT to significantly improve the neutron detection efficiency. An improvement in the efficiency of a factor of 1.5 was achieved without increasing the metal backing area for the boron coating. This improvement has allowed us to operate the detector in gamma-ray backgrounds that are four orders of magnitude higher than was previously possible while maintaining a relatively high counting efficiency for neutrons. This improvement in the gamma-ray rejection is a key factor in the development of the high dose neutron detector.

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

    International Nuclear Information System (INIS)

    Stewart, R.D.; Harty, R.; McDonald, J.C.; Tanner, J.E.

    1993-04-01

    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 D 2 O-moderated 252 Cf 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 252 Cf 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

  12. Hardening of single crystals of magnesium by low neutron doses at 77 K

    International Nuclear Information System (INIS)

    Gonzalez, H.C.

    1984-01-01

    Radiation hardening in Mg single crystals at 77 K is studied with a microtensile machine operating in-situ in the CNEA reactor facility RA1. Experimental results show that the dose dependence of the yield stress is similar to that previously observed in irradiated Cu and Zn. The radiation-induced yield stress, due to the presence of radiation obstacles operating alone, increases according to a 0.5 power law. It adds algebraically to the athermal component of the initial yield stress, but is not exactly additive to the other thermally activated mechanisms. For doses higher than 4.5 x 10 16 neutrons/cm 2 , a strong instability in the deformation is observed. Post irradiation experiments in tensile tests performed with a hard machine show a continuous stress drop. This effect is attributed to the dislocation channeling phenomenon which takes place during the tensile test. (author)

  13. Synergies Between ' and Cavity Formation in HT-9 Following High Dose Neutron Irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Field, Kevin G. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Parish, Chad M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Saleh, Tarik A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Eftink, Benjamin P. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2017-06-01

    Candidate cladding materials for advanced nuclear power reactors including fast reactor designs require materials capable of withstanding high dose neutron irradiation at elevated temperatures. One candidate material, HT-9, through various research programs have demonstrated the ability to withstand significant swelling and other radiation-induced degradation mechanisms in the high dose regime (>50 displacements per atom, dpa) at elevated temperatures (>300 C). Here, high efficiency multi-dimensional scanning transmission electron microscopy (STEM) acquisition with the aid of a three-dimensional (3D) reconstruction and modeling technique is used to probe the microstructural features that contribute to the exceptional swelling resistance of HT-9. In particular, the synergies between ' and fine-scale and moderate-scale cavity formation is investigated.

  14. The Influence of Used Construction Material and Its Thickness on the Neutron Dose Rate Around the Linear Accelerator - Experimental Results

    International Nuclear Information System (INIS)

    Krpan, I; Miklavcic, I.; Poje, M.; Radolic, V.; Vukovic, B.; Zivkovic, A.; Faj, D.; Ivkovic, A.

    2013-01-01

    Since linear accelerators for medical radiotherapy do not have active radioactive sources it makes them adequate from the radioprotection point of view. However, when operating at the energy higher than 10 MeV, they can become a source of unwanted neutron radiation in the giant dipole resonance reaction between the photon beam and the accelerator head material. Neutrons created in this reaction are almost isotropic in direction with an energy range between 700 keV and 1 MeV. During the accelerator installation and different phases of the construction work around the accelerator, a neutron dose rate at several important locations was investigated. Both passive (solid state nuclear track etched detectors - CR 39 and/or LR-115 with the 10B foil) and active detectors (Thermo Biorem FHT 752) were used. A higher photon dose rate was measured around the accelerator facility. An effective photon dose reduction was achieved using steel plates. However, this was the secondary source of neutrons in the reaction between the photons and steel plates, since higher values were measured. Neutron reduction was done by additional layers of barite concrete. A very conservative assessment of the effective dose was done for the medical personnel inside the control room. At the accelerator extreme operating regime (fixed accelerator direction - gantry angle, highest energy possible used), the neutron dose rate in the control room of 12 μSv/h was measured. Knowing the number of working days and number of patients per technician (per day), an exposure to the neutron dose of 1,1 mSv per year was calculated.(author)

  15. SU-E-T-568: Neutron Dose Survey of a Compact Single Room Proton Machine

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Y; Prusator, M; Islam, M; Johnson, D; Ahmad, S [University of Oklahoma Health Sciences Center, Oklahoma City, OK (United States)

    2015-06-15

    Purpose: To ensure acceptable radiation limits are maintained for those working at and near the machine during its operation, a comprehensive radiation survey was performed prior to the clinical release of Mevion S250 compact proton machine at Stephenson Oklahoma Cancer Center. Methods: The Mevion S250 proton therapy system consists of the following: a superconducting cyclotron to accelerate the proton particles, a passive double scattering system for beam shaping, and paired orthogonal x-ray imaging systems for patient setup and verification via a 6D robotic couch. All equipment is housed within a single vault of compact design. Two beam delivery applicators are available for patient treatment, offering field sizes of as great as 14 cm and 25 cm in diameter, respectively. Typical clinical dose rates are between 1 and 2 Gy/min with a fixed beam energy of 250 MeV. The large applicator (25 cm in diameter) was used in conjunction with a custom cut brass aperture to create a 20 cm x 20 cm field size at beam isocenter. A 30 cm − 30 cm − 35 cm high density plastic phantom was placed in the beam path to mimic the conditions creating patient scatter. Measurements integrated-ambient-neutron-dose-equivalence were made with a SWENDII detector. Gantry angles of 0, 90 and 180 degrees, with a maximum dose rate of 150 MU/min (for large applicator) and beam configuration of option 1 (range 25 cm and 20 cm modulation), were selected as testing conditions. At each point of interest, the highest reading was recorded at 30 cm from the barrier surface. Results: The highest neutron dose was estimated to be 0.085 mSv/year at the console area. Conclusion: All controlled areas are under 5 mSv/year and the uncontrolled areas are under 1 mSv/year. The radiation protection provided by the proton vault is of sufficient quality.

  16. DIANE, a simulation code for the interaction of neutrons with living tissues. Application to low doses of fast neutrons on human tumoral cells

    International Nuclear Information System (INIS)

    Nenot, M.L.

    2003-07-01

    Our work deals with the irradiation of cells and living tissues by 14 MeV neutrons at very low doses (a few 10 -2 Gy). Such experiments require an accurate knowledge of the values of neutron dose rates and fluences at the level of cell cultures. We have performed measurements of fluence rates through an activation method applied to gold and copper foils. The fluence rate is deduced from the gamma rays emitted by the irradiated foils. Neutron doses and dose rates have been measured through varied methods: PIN diodes, ionization tissue equivalent chambers, and Geiger-Mueller counters. We have designed the DIANE code to simulate the impact of energetic neutrons on cells. This code can be used with isolated cells or macroscopic tissues, it takes into account the roles of the ionisation electrons produced by recoil nuclei entering the cell. This point is all the more important since recent works have highlighted the impact of very low energy electrons on DNA. (A.C.)

  17. Photo neutron dose equivalent rate in 15 MV X-ray beam from a Siemens Primus Linac

    Directory of Open Access Journals (Sweden)

    A Ghasemi

    2015-01-01

    Full Text Available Fast and thermal neutron fluence rates from a 15 MV X-ray beams of a Siemens Primus Linac were measured using bare and moderated BF 3 proportional counter inside the treatment room at different locations. Fluence rate values were converted to dose equivalent rate (DER utilizing conversion factors of American Association of Physicist in Medicine′s (AAPM report number 19. For thermal neutrons, maximum and minimum DERs were 3.46 × 10 -6 (3 m from isocenter in +Y direction, 0 × 0 field size and 8.36 × 10 -8 Sv/min (in maze, 40 × 40 field size, respectively. For fast neutrons, maximum DERs using 9" and 3" moderators were 1.6 × 10 -5 and 1.74 × 10 -5 Sv/min (2 m from isocenter in +Y direction, 0 × 0 field size, respectively. By changing the field size, the variation in thermal neutron DER was more than the fast neutron DER and the changes in fast neutron DER were not significant in the bunker except inside the radiation field. This study showed that at all points and distances, by decreasing field size of the beam, thermal and fast neutron DER increases and the number of thermal neutrons is more than fast neutrons.

  18. Evaluation of neutron doses beyond of primary shielding of rooms housing clinical linear accelerators

    International Nuclear Information System (INIS)

    Rezende, Gabriel Fonseca da Silva

    2011-01-01

    The growing need to build radiotherapy rooms in places with lack of available space leads to the necessity of unconventional solutions for the shielding projects. In most cases, adding metals to the primary barriers is the best way to shield the rooms properly. However, when photons with energies equal to or great than 10 MeV interact with nuclei of materials with high atomic number, neutrons are ejected and can result in a problem of radioprotection both inside and outside the room. Currently, the only empirical formula existing in the literature to assess the dose equivalent due to neutrons beyond the laminated barriers works only under very specific conditions, and a validation of this formula had not yet been done. In this work, the Monte Carlo code MCNPX was used to verify the validity of the above formula for cases of primary barriers containing lead or iron sheets in rooms that house linear accelerators with 10, 15 and 18 MV. Moreover, such a code was used to evaluate the coefficient of neutron production and tenth-value layer for neutrons in concrete, both parameters that directly influence the equation studied. The study results showed that over 90% of the values compared between the formula and the simulations present discrepancies above 100%, which led to conclude that the formula from the literature produces values that do not match the reality. In addition, there were inconsistencies in the parameters that make up the formula, leading to a need to review this formula in order to build a new model that will better represent the real case. (author)

  19. Dose dependence of the microstructural evolution in neutron-irradiated austenitic stainless steel

    International Nuclear Information System (INIS)

    Zinkle, S.J.; Maziasz, P.J.; Stoller, R.E.

    1993-01-01

    Microstructural data on the evolution of the dislocation loop, cavity, and precipitate populations in neutron-irradiated austenitic stainless steels are reviewed in order to estimate the displacement damage levels needed to achieve the 'steady state' condition. The microstructural data can be conveniently divided into two temperature regimes. In the low temperature regime (below about 200 degrees C) the microstructure of austenitic stainless steel is dominated by 'black spot' defect clusters and faulted interstitial dislocation loops. The dose needed to approach saturation of the loop and defect cluster densities is generally on the order of 1 displacement per atom (dpa) in this regime. In the high temperature regime (∼300 to 700 degrees C), cavities, precipitates, loops and network dislocations are all produced during irradiation; doses in excess of 10 dpa are generally required to approach a 'steady state' microstructural condition. Due to complex interactions between the various microstructural components that form during irradiation, a secondary transient regime is typically observed in commercial stainless steels during irradiation at elevated temperatures. This slowly evolving secondary transient may extend to damage levels in excess of 50 dpa in typical 300-series stainless steels, and to >100 dpa in radiation-resistant developmental steels. The detailed evolution of any given microstructural component in the high-temperature regime is sensitive to slight variations in numerous experimental variables, including heat-to-heat composition changes and neutron spectrum

  20. Photon and neutron dose discrimination using low pressure proportional counters with graphite and A150 walls

    International Nuclear Information System (INIS)

    Kylloenen, J.; Lindborg, L.

    2005-01-01

    Full text: The determination of both the low- and high-LET components of ambient dose equivalent in mixed fields is possible with microdosimetric methods. With the multiple-event microdosimetric variance covariance method the sum of those components are directly obtained also in pulsed beams. However, if the value of each dose component is needed a more extended analysis is required. The use of a graphite walled proportional detector in combination with a tissue-equivalent proportional counter in combination with the variance covariance method was here investigated. MCNP simulations were carried out for relevant energies to investigate the photon and neutron responses of the two detectors. The combined graphite and TEPC system, the Sievert instrument, was used for measurements at IRSN, Cadarache, in the workplace calibration fields of CANEL+, SIGMA, a Cf-252 and a moderated Cf(D 2 O,Cd) radiation field. The response of the instrument in various monoenergetic neutron fields is also known from measurements at PTB. The instrument took part in the measurement campaigns in workplace fields in the nuclear industry organized within the EVIDOS contract. The results are analyzed and the method of using a graphite detector compared with alternative methods of analysis is discussed. (author)

  1. Dose Measurements of Bremsstrahlung-Produced Neutrons at the Advanced Photon Source

    International Nuclear Information System (INIS)

    Job, P.K.; Pisharody, M.; Semones, E.

    1998-01-01

    Bremsstrahlung is generated in the storage rings of the synchrotron radiation facilities by the radiative interaction of the circulating particle beam with both the residual gas molecules and storage ring components. These bremsstrahlung photons, having an energy range of zero to the maximum energy of the particle beam, interact with beamline components like beam stops and collimators generating photoneutrons of varying energies. There are three main processes by which photoneutrons may be produced by the high energy bremsstrahlung photons: giant nuclear dipole resonance and decay (10 MeV γ γ γ > 140 MeV). The giant resonance neutrons are emitted almost isotropically and have an average energy of about 2 MeV. High energy neutrons (E > 10 MeV) emitted from the quasi-deuteron decay and intranuclear cascade are peaked in the forward direction. At the Advanced Photon Source (APS), where bremsstrahlung energy can be as high as 7 GeV, production of photoneutrons in varying yields is possible from all of the above three processes. The bremsstrahlung produced along a typical 15.38-m straight path of the insertion device (ID) beamline of the APS has been measured and analyzed in previous studies. High-Z materials constituting the beamline components, such as collimators and beam stops, can produce photoneutrons upon interaction with these bremsstrahlung photons. The 1/E nature of the bremsstrahlung spectrum and the fact that the photoneutron production cross section is comparatively larger in the energy region 10 MeV γ 3 detector, as well as a very sensitive pressurized 3 He detector, is used for neutron dose measurements. The dose equivalent rates, normalized to bremsstrahlung power, beam current, and storage ring vacuum, are measured for various targets. This report details the experimental setup,

  2. A point-kernel shielding code for calculations of neutron and secondary gamma-ray 1cm dose equivalents: PKN

    International Nuclear Information System (INIS)

    Kotegawa, Hiroshi; Tanaka, Shun-ichi

    1991-09-01

    A point-kernel integral technique code, PKN, and the related data library have been developed to calculate neutron and secondary gamma-ray dose equivalents in water, concrete and iron shields for neutron sources in 3-dimensional geometry. The comparison between calculational results of the present code and those of the 1-dimensional transport code ANISN = JR, and the 2-dimensional transport code DOT4.2 showed a sufficient accuracy, and the availability of the PKN code has been confirmed. (author)

  3. Optimization of beam shaping assembly based on D-T neutron generator and dose evaluation for BNCT

    Science.gov (United States)

    Naeem, Hamza; Chen, Chaobin; Zheng, Huaqing; Song, Jing

    2017-04-01

    The feasibility of developing an epithermal neutron beam for a boron neutron capture therapy (BNCT) facility based on a high intensity D-T fusion neutron generator (HINEG) and using the Monte Carlo code SuperMC (Super Monte Carlo simulation program for nuclear and radiation process) is proposed in this study. The Monte Carlo code SuperMC is used to determine and optimize the final configuration of the beam shaping assembly (BSA). The optimal BSA design in a cylindrical geometry which consists of a natural uranium sphere (14 cm) as a neutron multiplier, AlF3 and TiF3 as moderators (20 cm each), Cd (1 mm) as a thermal neutron filter, Bi (5 cm) as a gamma shield, and Pb as a reflector and collimator to guide neutrons towards the exit window. The epithermal neutron beam flux of the proposed model is 5.73 × 109 n/cm2s, and other dosimetric parameters for the BNCT reported by IAEA-TECDOC-1223 have been verified. The phantom dose analysis shows that the designed BSA is accurate, efficient and suitable for BNCT applications. Thus, the Monte Carlo code SuperMC is concluded to be capable of simulating the BSA and the dose calculation for BNCT, and high epithermal flux can be achieved using proposed BSA.

  4. Estimated neutron-activation data for TFTR. Part II. Biological dose rate from sample-materials activation

    International Nuclear Information System (INIS)

    Ku, L.; Kolibal, J.G.

    1982-06-01

    The neutron induced material activation dose rate data are summarized for the TFTR operation. This report marks the completion of the second phase of the systematic study of the activation problem on the TFTR. The estimations of the neutron induced activation dose rates were made for spherical and slab objects, based on a point kernel method, for a wide range of materials. The dose rates as a function of cooling time for standard samples are presented for a number of typical neutron spectrum expected during TFTR DD and DT operations. The factors which account for the variations of the pulsing history, the characteristic size of the object and the distance of observation relative to the standard samples are also presented

  5. The relative biological effectiveness of fractionated doses of fast neutrons (42 MeVd→Be) for normal tissues. Pt. 3

    International Nuclear Information System (INIS)

    Rezvani, M.; Hopewell, J.W.; Robbins, M.E.C.; Hamlet, R.; Barnes, D.W.H.; Sansom, J.M.; Adams, P.J.V.

    1990-01-01

    The effect of single and fractionated doses of fast neutrons (42 MeV d→Bc ) on the early and late radiation responses of the pig lung have been assessed by the measurement of changes in lung function using a 133 Xe washout technique. The results obtained for irradiation schedules with fast neutrons have been compared with those after photon irradiation. There was no statistically significant difference between the values for the relative biological effectiveness (RBE) for the early and late radiation response of the lung. The RBE of the neutron beam increased with decreasing size of dose/fraction with an upper limit value of 4.39 ± 0.94 for infinitely small X-ray doses per fraction. (author)

  6. Achievement and qualification of multigroup cross-section library for light water reactor calculation

    International Nuclear Information System (INIS)

    Gastaldi, B.

    1986-07-01

    This study intends to improve then to check on integral experiments, the calculation of the main neutronic parameters in light water moderated lattices: Uranium 238 capture and consequently Plutonium 239 build-up, multiplication factor, temperature coefficient. The first part of this work concerns the resonant reaction rate calculation method implemented in the APOLLO code, the so-called LIVOLANT and JEANPIERRE formalism. The errors introduced by the corresponding assumptions are quantified and we propose substitution methods which avoid large biases and supply satisfactory results. The second part is dedicated to the cross-section evaluation of uranium major isotopes and to the achievement of APOLLO multigroup cross-sections. This cross-section set takes into considerations on the one hand the recent differential information and the other hand the various integral information obtained in the French Atomic Energy Commission facilities. The nuclear data file (JEF abd ENDF/B5) processing, for multigroup and self-shielded cross-sections achieving enable us to check the new THEMIS computer code. In the last part, the experimental validation of the proposed procedure (accurate formalism mutuel shielding and new multigroup library) is presented. This qualification is based on the reinterpretation of critical experiments performed in the EOLE reactor at Cadarache and spent fuel analysis. The corresponding results demonstrate that our propositions provide improvements on the computation of the PWR neutronic parameters; calculation-experiment discrepancies are now consistent with experimental uncertainty margins. 46 refs; 31 figs; 23 tabl [fr

  7. Portable instrument for measuring neutron energy spectra and neutron dose in a mixed n-γ field

    International Nuclear Information System (INIS)

    Daniels, C. J.; Silberberg, J. L.

    1980-01-01

    A portable high-speed neutron spectrometer consists of an organic scintillator, a true zero-crossing pulse shape discriminator, a 1 MHZ conversion-rate multichannel analyzer, an 8-bit microcomputer, and appropriate displays. The device can be used to measure neutron energy spectra and kerma rate in intense n- gamma radiation fields in which the neutron energy is from 5 to 15 MEV

  8. Reduction of the secondary neutron dose in passively scattered proton radiotherapy, using an optimized pre-collimator/collimator

    International Nuclear Information System (INIS)

    Brenner, David J; Elliston, Carl D; Hall, Eric J; Paganetti, Harald

    2009-01-01

    Proton radiotherapy represents a potential major advance in cancer therapy. Most current proton beams are spread out to cover the tumor using passive scattering and collimation, resulting in an extra whole-body high-energy neutron dose, primarily from proton interactions with the final collimator. There is considerable uncertainty as to the carcinogenic potential of low doses of high-energy neutrons, and thus we investigate whether this neutron dose can be significantly reduced without major modifications to passively scattered proton beam lines. Our goal is to optimize the design features of a patient-specific collimator or pre-collimator/collimator assembly. There are a number of often contradictory design features, in terms of geometry and material, involved in an optimal design. For example, plastic or hybrid plastic/metal collimators have a number of advantages. We quantify these design issues, and investigate the practical balances that can be achieved to significantly reduce the neutron dose without major alterations to the beamline design or function. Given that the majority of proton therapy treatments, at least for the next few years, will use passive scattering techniques, reducing the associated neutron-related risks by simple modifications of the collimator assembly design is a desirable goal.

  9. Extended use of alanine irradiated in experimental reactor for combined gamma- and neutron-dose assessment by ESR spectroscopy and thermal neutron fluence assessment by measurement of (14)C by LSC.

    Science.gov (United States)

    Bartoníček, B; Kučera, J; Světlík, I; Viererbl, L; Lahodová, Z; Tomášková, L; Cabalka, M

    2014-11-01

    Gamma- and neutron doses in an experimental reactor were measured using alanine/electron spin resonance (ESR) spectrometry. The absorbed dose in alanine was decomposed into contributions caused by gamma and neutron radiation using neutron kerma factors. To overcome a low sensitivity of the alanine/ESR response to thermal neutrons, a novel method has been proposed for the assessment of a thermal neutron flux using the (14)N(n,p) (14)C reaction on nitrogen present in alanine and subsequent measurement of (14)C by liquid scintillation counting (LSC). Copyright © 2014 Elsevier Ltd. All rights reserved.

  10. Parallel computation of multigroup reactivity coefficient using iterative method

    Science.gov (United States)

    Susmikanti, Mike; Dewayatna, Winter

    2013-09-01

    One of the research activities to support the commercial radioisotope production program is a safety research target irradiation FPM (Fission Product Molybdenum). FPM targets form a tube made of stainless steel in which the nuclear degrees of superimposed high-enriched uranium. FPM irradiation tube is intended to obtain fission. The fission material widely used in the form of kits in the world of nuclear medicine. Irradiation FPM tube reactor core would interfere with performance. One of the disorders comes from changes in flux or reactivity. It is necessary to study a method for calculating safety terrace ongoing configuration changes during the life of the reactor, making the code faster became an absolute necessity. Neutron safety margin for the research reactor can be reused without modification to the calculation of the reactivity of the reactor, so that is an advantage of using perturbation method. The criticality and flux in multigroup diffusion model was calculate at various irradiation positions in some uranium content. This model has a complex computation. Several parallel algorithms with iterative method have been developed for the sparse and big matrix solution. The Black-Red Gauss Seidel Iteration and the power iteration parallel method can be used to solve multigroup diffusion equation system and calculated the criticality and reactivity coeficient. This research was developed code for reactivity calculation which used one of safety analysis with parallel processing. It can be done more quickly and efficiently by utilizing the parallel processing in the multicore computer. This code was applied for the safety limits calculation of irradiated targets FPM with increment Uranium.

  11. Dose dependence of microstructural evolution and mechanical properties of neutron irradiated copper and copper alloys

    Energy Technology Data Exchange (ETDEWEB)

    Singh, B N; Edwards, D J; Horsewell, A; Toft, P

    1995-09-01

    The present investigation of the effects of neutron irradiation on microstructures and mechanical properties of copper alloys is a part of the ITER (International Thermonuclear Experimental Reactor) programme. Tensile specimens of the candidate alloys Cu-Al{sub 2}O{sub 3}, CuCrZr and CuNiBe were irradiated with fission neutrons in the DR-3 reactor at Risoe with a flux of 2.5 x 10{sup 17} n/m{sup 2}s (E > 1 MeV, i.e. a dose rate of {approx}5 x 10{sup -8} dpa/s) to fluences of 5 x 10{sup 22}, 5 x 10{sup 23} and 1 x 10{sup 24} n/m{sup 2} (E > 1 MeV, i.e. displacement doses of 0.01, 0.1 and 0.2 dpa) at 47 deg. C. The Cu-Al{sub 2}O{sub 3} (CuA125) specimens, were irradiated in the as-cold worked state. Tensile properties and Vickers hardness of both irradiated and unirradiated specimens were determined at 22 deg. C. Pre- and post-deformation microstructures of irradiated as well as unirradiated specimens were examined using a transmission electron microscope. The fractured surfaces of tensile tested specimens were investigated in a scanning electron microscope. The results show the following general trend: (a) that the CuNiBe alloy is stronger than CuCrZr as well as Cu Al{sub 2}O{sub 3}, (b) that even relatively low dose irradiations cause significant increase in the yield strength, but rather drastic decreases in the uniform elongation of CuCrZr and CuNiBe alloys and that the low dose irradiation of the cold-worked Cu-Al{sub 2}O{sub 3} alloy causes a decrease in the yield strength and an increase in the uniform elongation, at higher doses irradiation hardening occurs. The SEM examinations of the fractured surfaces demonstrate that both unirradiated and irradiated specimens fracture in a ductile manner. The lack of uniform elongation in the irradiated copper alloys may be understood in terms of difficulty in dislocation generation due to pinning of grown-in dislocation by defect clusters (loops) at or around them. (EG) 5 tabs., 18 ills., 13 refs.

  12. Dose dependence of microstructural evolution and mechanical properties of neutron irradiated copper and copper alloys

    International Nuclear Information System (INIS)

    Singh, B.N.; Edwards, D.J.; Horsewell, A.; Toft, P.

    1995-09-01

    The present investigation of the effects of neutron irradiation on microstructures and mechanical properties of copper alloys is a part of the ITER (International Thermonuclear Experimental Reactor) programme. Tensile specimens of the candidate alloys Cu-Al 2 O 3 , CuCrZr and CuNiBe were irradiated with fission neutrons in the DR-3 reactor at Risoe with a flux of 2.5 x 10 17 n/m 2 s (E > 1 MeV, i.e. a dose rate of ∼5 x 10 -8 dpa/s) to fluences of 5 x 10 22 , 5 x 10 23 and 1 x 10 24 n/m 2 (E > 1 MeV, i.e. displacement doses of 0.01, 0.1 and 0.2 dpa) at 47 deg. C. The Cu-Al 2 O 3 (CuA125) specimens, were irradiated in the as-cold worked state. Tensile properties and Vickers hardness of both irradiated and unirradiated specimens were determined at 22 deg. C. Pre- and post-deformation microstructures of irradiated as well as unirradiated specimens were examined using a transmission electron microscope. The fractured surfaces of tensile tested specimens were investigated in a scanning electron microscope. The results show the following general trend: (a) that the CuNiBe alloy is stronger than CuCrZr as well as Cu Al 2 O 3 , (b) that even relatively low dose irradiations cause significant increase in the yield strength, but rather drastic decreases in the uniform elongation of CuCrZr and CuNiBe alloys and that the low dose irradiation of the cold-worked Cu-Al 2 O 3 alloy causes a decrease in the yield strength and an increase in the uniform elongation, at higher doses irradiation hardening occurs. The SEM examinations of the fractured surfaces demonstrate that both unirradiated and irradiated specimens fracture in a ductile manner. The lack of uniform elongation in the irradiated copper alloys may be understood in terms of difficulty in dislocation generation due to pinning of grown-in dislocation by defect clusters (loops) at or around them. (EG) 5 tabs., 18 ills., 13 refs

  13. Multigroup perturbation model for kinetic analysis of nuclear reactors

    International Nuclear Information System (INIS)

    Souza, G.M.

    1989-01-01

    The scope of this work is the development of a multigroup perturbation theory for the purpose of Kinetic and dynamic analysis of nuclear reactors. The equations that describe the reactor behavior were presented in all generality and written in the shorthand notation of matrices and vectors. In the derivation of those equations indetermined operators and discretizing factors were introduced and then determined by comparision with conventional equations. Fick's Law was developed in higher orders for neutron and importance current density. The solution of the direct and adjoint fields were represented by combination of the eigenfunctions of the B and B* operators and the eigenvalue modulus equality was established mathematically. In the derivation of the reactivity expression the B operator perturbation was split in two non coupled to the flux form and level. The prompt neutrons effective mean life was derived from reactor equations and importance conservation. The establishment of the Nordheim's equation, although modified, was based on Gandini. Finally, a mathematical interpretation of the flux-trap region was avented. (author)

  14. The mechanical behavior and reliability prediction of the HTR graphite component at various temperature and neutron dose ranges

    International Nuclear Information System (INIS)

    Fang, Xiang; Yu, Suyuan; Wang, Haitao; Li, Chenfeng

    2014-01-01

    Highlights: • The mechanical behavior of graphite component in HTRs under high temperature and neutron irradiation conditions is simulated. • The computational process of mechanical analysis is introduced. • Deformation, stresses and failure probability of the graphite component are obtained and discussed. • Various temperature and neutron dose ranges are selected in order to investigate the effect of in-core conditions on the results. - Abstract: In a pebble-bed high temperature gas-cooled reactor (HTR), nuclear graphite serves as the main structural material of the side reflectors. The reactor core is made up of a large number of graphite bricks. In the normal operation case of the reactor, the maximum temperature of the helium coolant commonly reaches about 750 °C. After around 30 years’ full power operation, the peak value of in-core fast neutron cumulative dose reaches to 1 × 10 22 n cm −2 (EDN). Such high temperature and neutron irradiation strongly impact the behavior of graphite component, causing obvious deformation. The temperature and neutron dose are unevenly distributed inside a graphite brick, resulting in stress concentrations. The deformation and stress concentration can both greatly affect safety and reliability of the graphite component. In addition, most of the graphite properties (such as Young's modulus and coefficient of thermal expansion) change remarkably under high temperature and neutron irradiations. The irradiation-induced creep also plays a very important role during the whole process, and provides a significant impact on the stress accumulation. In order to simulate the behavior of graphite component under various in-core conditions, all of the above factors must be considered carefully. In this paper, the deformation, stress distribution and failure probability of a side graphite component are studied at various temperature points and neutron dose levels. 400 °C, 500 °C, 600 °C and 750 °C are selected as the

  15. Hair {sup 32}P measurement for body dose mapping in non-fatal exposures to fast neutrons

    Energy Technology Data Exchange (ETDEWEB)

    Mianji, Fereidoun A. [Nuclear Science and Technology Research Institute, Tehran (Iran, Islamic Republic of); Iran Nuclear Regulatory Authority, Tehran (Iran, Islamic Republic of); Jafari, Sheyda; Zaryouni, Saiedeh [Bu-Ali Sina University, Faculty of Science, Hamedan (Iran, Islamic Republic of); Hajizadeh, Bardia [Iran Nuclear Regulatory Authority, Tehran (Iran, Islamic Republic of)

    2015-03-15

    Dosimetry bioassay methods are the backbone of a personal dosimetry in criticality accidents. Although methods like hair dosimetry and the use of activation foils (e.g., {sup 32}S) have been employed for decades, capabilities of different techniques, effects of hair type and neutron spectrum on the dose response, sensitivity and uncertainties of different techniques, etc., need more investigations. For this reason, the use of the {sup 32}S(n,p){sup 32}P reaction and hair samples for estimating non-fatal doses from fast neutrons was studied. The experiments were carried out with the hair samples attached on a RANDO phantom in a Cf-252 neutron field, in the dose range of about 0.05-1.15 Gy. In addition, the adequate post-accident preparation for hair samples including optimum conditioning and timing were investigated. Experimental results prove the good sensitivity and merit of the method for neutron quantification in the mentioned dose range for which other bioassay methods are of poor resolution and sensitivity. A rough estimation of the dose-response curve for Iranian hair was also derived. (orig.)

  16. Measurement of dose rates and Monte Carlo analysis of neutrons in a spent-fuel shipping vessel

    International Nuclear Information System (INIS)

    Ueki, K.; Namito, Y.; Fuse, T.

    1986-01-01

    On-board experiments were carried out in a spent-fuel shipping vessel, the Pacific Swan, in which 13 casks of TN-12A and Excellox 3 were loaded in five holds, and neutron and gamma-ray dose rates were measured on the hatch covers of the holds. Before shipping those casks, dose rates were also measured on the cask surfaces, one by one, to eliminate radiation from other casks. The Monte Carlo coupling technique was employed successfully to analyze the measured neutron dose rate distributions in the spent-fuel shipping vessel. Through this study, the Monte Carlo coupling code system, MORSE-CG/CASK-VESSEL, on which the MORSE-CG code was based, was established. The agreement between the measured and the calculated neutron dose rates on the TN-12A cask surface was quite satisfactory. The calculated neutron dose rates agreed with the measured values within a factor of 1.5 on the hold 3 hatch cover and within a factor of 2 on the hold 5 hatch cover in which the concrete shield was fixed in the Pacific Swan

  17. Temperature and dose dependencies of microstructure and hardness of neutron irradiated OFHC copper

    International Nuclear Information System (INIS)

    Singh, B.N.; Horsewell, A.; Toft, P.; Edwards, D.J.

    1995-01-01

    Tensile specimens of pure oxygen free high conductivity (OFHC) copper were irradiated with fission neutrons between 320 and 723 K to fluences in the range 5x10 21 to 1.5x10 24 n/m 2 (E>1 MeV) with a flux of 2.5x10 17 n/m 2 s. Irradiated specimens were investigated by transmission electron microscopy (TEM) and quantitative determinations were made of defect clusters and cavities. The dose dependence of tensile properties of specimens irradiated at 320 K was determined at 295 K. Hardness measurements were made at 295 K on specimens irradiated at different temperatures and doses. Microstructures of tensile tested specimens were also investigated by TEM. Results show that the increase in cluster density and hardening nearly saturate at a dose of similar 0.3 dpa. Irradiations at 320 K cause a drastic decrease in the uniform elongation already at ∼ =0.1 dpa. It is suggested that the irradiation-induced increase in the initial yield stress and a drastic decrease in the ability of copper to deform plastically in a homogeneous fashion are caused by a substantial reduction in the ability of grown-in dislocations to act as efficient dislocation sources. ((orig.))

  18. Measurement of the relationship of 24 Na activity and the received neutron dose

    International Nuclear Information System (INIS)

    Gossio, S.; Carrelli, J.; Villella, A.; Soppe, E.

    2013-01-01

    In cases of criticality accidents it is required a fast dosimetric system that allows to evaluate the doses of the personnel involved. The reaction (n,y) with sodium presented in the body ( 23 Na), generates 24 Na, that emits two gamma of 1369KeV and 2754 KeV that can be measured using a whole body counter. The experienced were carried out with the irradiation of 252 Cf of a phantom with a solution of NaCl in water. After the irradiation it was measured the 24 Na activity in the whole body counter, which has a HPGe detector previously calibrated in energy and efficiency. Considering the correction by decay, the quantity of 23 Na presented in the body of an adult and elimination curve of 24 Na, it was established a coefficient of neutronic doses by unity of activity of 24 Na measured in the whole body counter. This method is useful for the retrospective estimation of the doses, as well as to carry out a radiological sorting in case of criticality

  19. Dynamics of elements in soil treated with increasing doses sewage sludge for instrumental neutron activation analysis

    International Nuclear Information System (INIS)

    Oliveira, Helder de; Mortatti, Jefferson; Vendramini, Diego; Lopes, Renato A.; Nolasco, Murilo M.; Sarries, Gabriel A.; Furlan, Adriana

    2007-01-01

    In this work the dynamics of the elements was analyzed The, Br, Ce, Co, Cr, Cs, Fe, Hf, La, In the, Sb, Sc, Sm, Ta, Th, U, Yb and Zn in a profile of a red-yellow latossolo, in the depths of 0-5, 5-10, 10-30 and 30-50 cm, and dose of the biosolid of 0, 25, 124 and 375 t ha -1 , of the station of treatment of sewer of Barueri, Sao Paulo. The experiment was carried out in areas of 3,05 m 2 in the times of 2,2; 4,0; 6,6; 14,3 and 21 months. For analysis of the elementary composition, it was used of the analysis technique by instrumental neutron activation analysis (INAA). The experiment was submitted under normal tropical conditions in a forest station in Itatinga, Sao Paulo, of the University of Sao Paulo. For better details, the factors depth, doses and times statistical analyses of the results of the elementary composition of the soil samples were made. For all the biossolid doses conditioned with polymeric and applied in the soil, the composition of 17 of the 18 elements in the soil were not altered, with exception for Cr in the studied times. The elements As, Br, Ce, Co, Fe, Hf, La, Sm, Ta, Th, U and Yb presented higher levels in the deepest layers of soil; already the elements Cr, In the, Sb and Zn presented higher concentrations in the superficial layers. (author)

  20. Genetic injury in hybrid male mice exposed to low doses of 60CO γ-rays or fission neutrons. 1

    International Nuclear Information System (INIS)

    Grahn, D.; Carnes, B.A.; Farrington, B.H.; Lee, C.H.

    1984-01-01

    Young adult male B6CF 1 mice were exposed to single whole body doses of fission neutrons or 60 Co γ rays. Postspermatogonial dominant lethal injury, incidence of reciprocal chromosome translocations induced in spermatogonia, incidence of abnormal epididymal sperm 4-6 weeks after exposure, and testis weight loss 3-6 weeks after exposure were all measured. Significant effects were seen at 1 and 2.5 rad of neutrons consistent with extrapolation from higher doses, with the exception of dominant lethal mutations, which occurred in significant excess of expectation. Dose-response functions were linear or linear-quadratic, depending upon end point, radiation quality, and dose range. For translocation frequencies, the D 2 term was negative for neutron and positive for γ-ray irradiations. RBE values varied with dose and end point. For testis weight loss and abnormal sperm over the full dose range, the RBEs were between 5 and 6. They were between 7 and 9 at lower doses (< 10 rad) for translocations. RBEs for postimplantation and total dominant lethal rates were 5-6 above 10 rad and 10-14 below 10 rad. The RBEs for preimplant losses were between 15 and 25 above 10 rad and possibly higher below 10 rad, although the data are statistically 'noisy'. (Auth.)

  1. Initial measurement of site boundary neutron dose and comparison with calculations

    International Nuclear Information System (INIS)

    Degtyarenko, P.; Dotson, D.; May, R.; Schwahn, S.; Stapleton, G.

    1996-01-01

    For most accelerators adequate side shielding can be provided at minimal cost to meet the most aggressive radiation protection regulations and, further, the likely requirement to increase shielding thickness still more at a later date can be done usually by heaping more earth or applying local shielding at minimal expense and inconvenience. This moderately happy state of affairs does not unfortunately hold true with roof shielding. The cost of roof shielding is largely predicated on the roof span and the necessary structural engineering requirements for its support. These measures can be extremely expensive and where one is dealing with the rather extensive unsupported spans typical of experimental halls devoted to experiments with high energy electron beams; it is necessary to specify the roof thickness as carefully as possible with the constant concern that adding more earth later is not likely to be possible without rebuilding the hall. Because of the nature of roof skyshine, and for most high energy accelerator facilities neutron skyshine, the effect of the radiation is likely to extend to the facility fence-line where one is concerned about the exposure of the general population. Very properly the dose limit for the general population is set at a rather low value (1 mSv y -1 ) and in order for the Jefferson Lab (JLab) to ensure strict compliance with this limit they have a design goal for the fence line of 0.1 mSv y -1 . However, because natural neutron backgrounds are low (30--40 microSv y -1 ) and the methods of detection and measurement permit rejection of background interference from photons, they can measure the JLab produced neutron radiation with good sensitivity and precision

  2. RBE/absorbed dose relationship of d(50)-Be neutrons determined for early intestinal tolerance in mice

    International Nuclear Information System (INIS)

    Gueulette, J.; Wambersie, A.

    1978-01-01

    RBE/absorbed dose relationship of d(50)-Be neutrons (ref.: 60 Co) was determined using intestinal tolerance in mice (LD50) after single and fractionated irradiation. RBE is 1.8 for a single fraction (about 1000 rad 60 Co dose); it increases when decreasing dose and reaches the plateau value of 2.8 for a 60 Co dose of about 200 rad. This RBE value is used for the clinical applications with the cyclotron 'Cyclone' at Louvain-la-Neuve [fr

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

    Science.gov (United States)

    Brovchenko, Mariya; Dechenaux, Benjamin; Burn, Kenneth W.; Console Camprini, Patrizio; Duhamel, Isabelle; Peron, Arthur

    2017-09-01

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

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

    Directory of Open Access Journals (Sweden)

    Brovchenko Mariya

    2017-01-01

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

  5. Novel Concrete Chemistry Achieved with Low Dose Gamma Radiation Curing and Resistance to Neutron Activation

    Science.gov (United States)

    Burnham, Steven Robert

    As much as 50% of ageing-related problems with concrete structures can be attributed to con-struction deficiencies at the time of placement. The most influential time affecting longevity of concrete structures is the curing phase, or commonly the initial 28 days following its placement. A novel advanced atomistic analysis of novel concrete chemistry is presented in this dissertation with the objective to improve concrete structural properties and its longevity. Based on experiments and computational models, this novel concrete chemistry is discussed in two cases: (a) concrete chemistry changes when exposed to low-dose gamma radiation in its early curing stage, thus improving its strength in a shorter period of time then curing for the conventional 28 days; (b) concrete chemistry is controlled by its atomistic components to assure strength is not reduced but that its activation due to long-term exposure to neutron flux in nuclear power plants is negligible. High dose gamma radiation is well documented as a degradation mechanism that decreases concrete's compressive strength; however, the effects of low-dose gamma radiation on the initial curing phase of concrete, having never been studied before, proved its compressive strength increases. Using a 137 Cs source, concrete samples were subjected to gamma radiation during the initial curing phase for seven, 14, and 28 days. The compressive strength after seven days is improved for gamma cured concrete by 24% and after 14 days by 76%. Concrete shows no improvement in compressive strength after 28 days of exposure to gamma radiation, showing that there is a threshold effect. Scanning Electron Microscopy is used to examine the microstructure of low-dose gamma radiation where no damage to its microstructure is found, showing no difference between gamma cured and conventionally cured concrete. Molecular dynamics modeling based on the MOPAC package is used to study how gamma radiation during the curing stage improves

  6. Quick evaluation of the neutron dose following a criticality accident by measurement of sodium 24 activity

    International Nuclear Information System (INIS)

    Tabardel, R.; Ricourt, A.; Parmentier, N.

    1984-07-01

    In order to quickly sort out the irradiated individuals following a criticality accident, the neutron dose can be evaluated quickly by measuring the sodium-24 activity induced in the human body. The report supplies the information necessary for this evaluation from the response of various detectors of current use in radiation protection. The first part describes the method of evaluation of sodium-24 activity (A) given by the reading (M) of each instrument. The second part describes the method of kerma evaluation from the measured sodium-24 activity. The third part is an experimental application of the method of kerma evaluation from the sodium-24 activity measured in a phantom irradiated in the SILENE reactor flux. The results given by radiation protection instruments are in good agreement with the calculated values for a front exposure and demonstrate the usefulness of measuring the induced sodium-24 activity by radiation protection instruments of current use [fr

  7. Dose conversion coefficients for neutron exposure to the lens of the human eye

    International Nuclear Information System (INIS)

    Manger, Ryan P.; Bellamy, Michael B.; Eckerman, Keith F.

    2011-01-01

    Dose conversion coefficients for the lens of the human eye have been calculated for neutron exposure at energies from 1 x 10 -9 to 20 MeV and several standard orientations: anterior-to-posterior, rotational and right lateral. MCNPX version 2.6.0, a Monte Carlo-based particle transport package, was used to determine the energy deposited in the lens of the eye. The human eyeball model was updated by partitioning the lens into sensitive and insensitive volumes as the anterior portion (sensitive volume) of the lens being more radiosensitive and prone to cataract formation. The updated eye model was used with the adult UF-ORNL mathematical phantom in the MCNPX transport calculations.

  8. Dose conversion coefficients for neutron exposure to the lens of the human eye

    International Nuclear Information System (INIS)

    Manger, R. P.; Bellamy, M. B.; Eckerman, K. F.

    2012-01-01

    Dose conversion coefficients for the lens of the human eye have been calculated for neutron exposure at energies from 1 x 10 -9 to 20 MeV and several standard orientations: anterior-to-posterior, rotational and right lateral. MCNPX version 2.6.0, a Monte Carlo-based particle transport package, was used to determine the energy deposited in the lens of the eye. The human eyeball model was updated by partitioning the lens into sensitive and insensitive volumes as the anterior portion (sensitive volume) of the lens being more radiosensitive and prone to cataract formation. The updated eye model was used with the adult UF-ORNL mathematical phantom in the MCNPX transport calculations. (authors)

  9. Occupational doses due to photoneutrons in medical linear accelerators rooms; Doses ocupacionais devido a neutrons em salas de aceleradores lineares de uso medico

    Energy Technology Data Exchange (ETDEWEB)

    Soares, Alessandro Facure Neves de Salles

    2006-04-15

    Medical linear accelerators, with maximum photon energies above 10 MeV, are becoming of common use in Brazil. Although desirable in the therapeutic point of view, the increase in photon energies causes the generation of undesired neutrons, which are produced through nuclear reactions between photons and the high Z target nuclei of the materials that constitute the accelerator head. In this work, MCNP simulation was undertaken to examine the neutron equivalent doses around the accelerators head and at the entrance of medical linear accelerators treatment rooms, some of them licensed in Brazil by the National Regulatory Agency (CNEN). The simulated neutron dose equivalents varied between 2 e 26 {mu} Sv/Gy{sub RX}, and the results were compared with calculations performed with the use of some semi-empirical equations found in literature. It was found that the semi-empirical equations underestimate the simulated neutron doses in the majority of the cases, if compared to the simulated values, suggesting that these equations must be revised, due to the increasing number of high energy machines in the country. (author)

  10. Depth-dose evaluation for lung and pancreas cancer treatment by BNCT using an epithermal neutron beam

    International Nuclear Information System (INIS)

    Matsumoto, Tetsuo; Fukushima, Yuji

    2000-01-01

    The depth-dose distributions were evaluated for possible treatment of both lung and pancreas cancers using an epithermal neutron beam. The MCNP calculations showed that physical dose in tumors were 6 and 7 Gy/h, respectively, for lung and pancreas, attaining an epithermal neutron flux of 5x10 8 ncm -2 s -1 . The boron concentrations were assumed at 100 ppm and 30 ppm, respectively, for lung and pancreas tumors and normal tissues contains 1/10 tumor concentrations. The dose ratios of tumor to normal tissue were 2.5 and 2.4, respectively, for lung and pancreas. The dose evaluation suggests that BNCT could be applied for both lung and pancreas cancer treatment. (author)

  11. Update of neutron dose yields as a function of energy for protons and deuterons incident on beryllium targets

    International Nuclear Information System (INIS)

    Ten Haken, R.K.; Awschalom, M.; Rosenberg, I.

    1982-11-01

    Neutron absorbed dose yields (absorbed dose rates per unit incident current on targets at a given SAD or SSD) increase with incident charged particle energy for both protons and deuterons. Analyses of neutron dose yield versus incident particle energy have been performed for both deuterons and protons. It is the purpose of this report to update those analyses by pooling all of the more recent published results and to reanalyze the trend of yield, Y, versus incident energy, E, which in the past has been described by an expression of the form Y = aE/sup b/, where a and b are empirical constants. From the reanalyzed trend it is concluded that for a given size cyclotron (E/sub p/ = 2E/sub d/), the dose yields using protons are higher than those using deuterons up to a proton energy E/sub p/ of 64 MeV

  12. Neutron relative biological effectiveness for solid cancer incidence in the Japanese A-bomb survivors: an analysis considering the degree of independent effects from γ-ray and neutron absorbed doses with hierarchical partitioning

    Energy Technology Data Exchange (ETDEWEB)

    Walsh, Linda [Federal Office for Radiation Protection, Department Radiation Protection and Health, Oberschleissheim (Germany); University of Manchester, The Faculty of Medical and Human Sciences, Manchester (United Kingdom)

    2013-03-15

    It has generally been assumed that the neutron and γ-ray absorbed doses in the data from the life span study (LSS) of the Japanese A-bomb survivors are too highly correlated for an independent separation of the all solid cancer risks due to neutrons and due to γ-rays. However, with the release of the most recent data for all solid cancer incidence and the increased statistical power over previous datasets, it is instructive to consider alternatives to the usual approaches. Simple excess relative risk (ERR) models for radiation-induced solid cancer incidence fitted to the LSS epidemiological data have been applied with neutron and γ-ray absorbed doses as separate explanatory covariables. A simple evaluation of the degree of independent effects from γ-ray and neutron absorbed doses on the all solid cancer risk with the hierarchical partitioning (HP) technique is presented here. The degree of multi-collinearity between the γ-ray and neutron absorbed doses has also been considered. The results show that, whereas the partial correlation between the neutron and γ-ray colon absorbed doses may be considered to be high at 0.74, this value is just below the level beyond which remedial action, such as adding the doses together, is usually recommended. The resulting variance inflation factor is 2.2. Applying HP indicates that just under half of the drop in deviance resulting from adding the γ-ray and neutron absorbed doses to the baseline risk model comes from the joint effects of the neutrons and γ-rays - leaving a substantial proportion of this deviance drop accounted for by individual effects of the neutrons and γ-rays. The average ERR/Gy γ-ray absorbed dose and the ERR/Gy neutron absorbed dose that have been obtained here directly for the first time, agree well with previous indirect estimates. The average relative biological effectiveness (RBE) of neutrons relative to γ-rays, calculated directly from fit parameters to the all solid cancer ERR model with both

  13. Transmutation approximations for the application of hybrid Monte Carlo/deterministic neutron transport to shutdown dose rate analysis

    International Nuclear Information System (INIS)

    Biondo, Elliott D.; Wilson, Paul P. H.

    2017-01-01

    In fusion energy systems (FES) neutrons born from burning plasma activate system components. The photon dose rate after shutdown from resulting radionuclides must be quantified. This shutdown dose rate (SDR) is calculated by coupling neutron transport, activation analysis, and photon transport. The size, complexity, and attenuating configuration of FES motivate the use of hybrid Monte Carlo (MC)/deterministic neutron transport. The Multi-Step Consistent Adjoint Driven Importance Sampling (MS-CADIS) method can be used to optimize MC neutron transport for coupled multiphysics problems, including SDR analysis, using deterministic estimates of adjoint flux distributions. When used for SDR analysis, MS-CADIS requires the formulation of an adjoint neutron source that approximates the transmutation process. In this work, transmutation approximations are used to derive a solution for this adjoint neutron source. It is shown that these approximations are reasonably met for typical FES neutron spectra and materials over a range of irradiation scenarios. When these approximations are met, the Groupwise Transmutation (GT)-CADIS method, proposed here, can be used effectively. GT-CADIS is an implementation of the MS-CADIS method for SDR analysis that uses a series of single-energy-group irradiations to calculate the adjoint neutron source. For a simple SDR problem, GT-CADIS provides speedups of 200 100 relative to global variance reduction with the Forward-Weighted (FW)-CADIS method and 9 _± 5 • _1_0_"_4 relative to analog. As a result, this work shows that GT-CADIS is broadly applicable to FES problems and will significantly reduce the computational resources necessary for SDR analysis.

  14. Effects of low dose rate fission neutron irradiation on the lymphocyte subpopulations of peripheral blood in rats

    International Nuclear Information System (INIS)

    Jiang Dingwen; Lei Chengxiang; Shen Xianrong; Ma Li; Yang Yifang; Peng Wulin; Dai Shourong

    2008-01-01

    Objective: To evaluate the effects of long-term, low dose rate fission neutron irradiation on lymphocyte subpopulations in peripheral blood of rats. Methods: Ninety-six rats were randomly divided into control group and irradiated group exposed to low dose rate fission neutron ( 252 Cf,0.35 mGy/h) for 20.5 h every day. At days 14,28,42,56 and 70 d after irradiation and 35 d after stopping irradiation, After 8 rats of each group were killed, WBC and lymphocyte subpopulations of CD4 + CD3 + , CD8 + CD3 + and CD45RA + /CD161α + in peripheral blood were estimated respectively. Results: Compared with the control group, WBC was reduced significantly at dose of 0.3, 0.4 and 0.5 Gy (P + CD3 - was evidently higher compared with control group at doses of 0.1,0.3, 0.4 and 0.5 Gy and 35 d after stopping irradiation (P + CD3 - was obviously higher compared with control group at dose of 0.2 and 0.3 Gy (P + CD3 + at dose of 0.1 Gy (P + CD3 + at doses of 0.1 and 0.2 Gy (P + CD45RA - ) was increased significantly at doses of 0.2-0.3 Gy, and peripheral blood B cells(CD161α - CD45RA + ) was reduced remarkably at doses of 0.1-0.5 Gy and 35 d after stopping irradiation compared with the control group. Conclusions: Long-term irradiation with low dose rate fission neutron could make TCR (T-cell-receptor) mutant, therefore, WBC, B cells in peripheral blood significantly reduced and NK cells increased. These changes may could not recover at 35 d after Stopping irradiation. (authors)

  15. DOSE-Analyzer. A computer program with graphical user interface to analyze absorbed dose inside a body of mouse and human upon external neutron exposure

    International Nuclear Information System (INIS)

    Satoh, Daiki; Takahashi, Fumiaki; Shigemori, Yuji; Sakamoto, Kensaku

    2010-06-01

    DOSE-Analyzer is a computer program to retrieve the dose information from a database and generate a graph through a graphical user interface (GUI). The database is constructed for absorbed dose, fluence, and energy distribution inside a body of mouse and human exposed upon external neutrons, which is calculated by our developed Monte-Carlo simulation method using voxel-based phantom and particle transport code PHITS. The input configurations of irradiation geometry, subject, and energy are set by GUI. The results are tabulated at particle types, i.e. electron, proton, deuteron, triton, and alpha particle, and target organs on a data sheet of Microsoft Office Excel TM . Simple analysis to compare the output values for two subjects is also performed on DOSE-Analyzer. This report is a user manual of DOSE-Analyzer. (author)

  16. Neutron spectrometry by means of threshold detectors - Neutron spectrometry by means of activation detectors. Studies of the method of approximation by polygonal function. Application to dose determination

    International Nuclear Information System (INIS)

    Bricka, M.

    1962-03-01

    This report addresses the problem of determination of neutron spectrum by using a set of detectors. The spectrum approximation method based on a polygonal function is more particularly studied. The author shows that the coefficients of the usual mathematical model can be simply formulated and assessed. The study of spectra approximation by a polygonal function shows that dose can be expressed by a linear function of the activity of the different detectors [fr

  17. Immediate Dose Assessment for Radiation Accident in Laboratory Containing Gamma Source and/or Neutron Source

    International Nuclear Information System (INIS)

    Ahmed, E.M.

    2012-01-01

    One of the most important safety requirements for any place containing radiation sources is an accurate and fast way to assess the dose rate in both normal and accidental case. In normal case, the source is completely protected inside its surrounded shields in case of non use. In some cases this source may stuck outside its shield. In this case the walls of the place act as a shield. Many studies were carried for obtaining the most appropriate materials that may be used as shielding depending on their efficiency and also their cost. As concrete- with different densities- is the most available constructive material, this study presented a theoretical model using MCNP-4B code, based on Monte Carlo method to estimate the dose rate distribution in a laboratory with concrete walls in case of source stuck accident. The study dealt with Cs-137 as gamma source and Am-Be-241 as neutron source. Two different densities of concrete and also different thicknesses of walls were studied. The used model was verified by comparing the results with a practical study concerning with the effect of adding carbon powder to the concrete. The results showed good agreement

  18. Comparison between steel and lead shieldings for radiotherapy rooms regarding neutron doses to patients

    Energy Technology Data Exchange (ETDEWEB)

    Silva, M.G.; Rebello, W.F.; Andrade, E.R.; Medeiros, M.P.C.; Mendes, R.M.S.; Braga, K.L.; Gomes, R.G., E-mail: eng.cavaliere@gmail.com, E-mail: ggrprojetos@gmail.com [Instituto Militar de Engenharia (IME), Rio de Janeiro, RJ (Brazil). Secao de Engenharia Nuclear; Silva, A.X., E-mail: ademir@con.ufrj.br [Coordenacao dos Programas de Pos-Graduacao em Engenharia (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil)

    2015-07-01

    The NCRP Report No. 151, Structural Shielding Design and Evaluation for Megavoltage X- and Gamma-Ray Radiotherapy Facilities, considers, in shielding calculations for radiotherapy rooms, the use of lead and/or steel to be applied on bunker walls. The NCRP Report calculations were performed foreseeing a better protection of people outside the radiotherapy room. However, contribution of lead and steel to patient dose should be taken into account for radioprotection purposes. This work presents calculations performed by MCNPX code in analyzing the Ambient Dose Equivalent due to neutron, H *(10){sub n}, within a radiotherapy room, in the patients area, considering the use of additional shielding of 1 TVL of lead or 1 TVL of steel, positioned at the inner faces of walls and ceiling of a bunker. The head of the linear accelerator Varian 2100/2300 C/D was modeled working at 18MeV, with 5 x 5 cm{sup 2}, 10 x 10 cm{sup 2}, 20 x 20 cm{sup 2}, 30 x 30 cm{sup 2} and 40 x 40 cm{sup 2} openings for jaws and MLC and operating in eight gantry's angles. This study shows that the use of lead generates an average value of H *(10){sub n} at patients area, 8.02% higher than the expected when using steel. Further studies should be performed based on experimental data for comparison with those from MCNPX simulation. (author)

  19. Comparison between steel and lead shieldings for radiotherapy rooms regarding neutron doses to patients

    International Nuclear Information System (INIS)

    Silva, M.G.; Rebello, W.F.; Andrade, E.R.; Medeiros, M.P.C.; Mendes, R.M.S.; Braga, K.L.; Gomes, R.G.

    2015-01-01

    The NCRP Report No. 151, Structural Shielding Design and Evaluation for Megavoltage X- and Gamma-Ray Radiotherapy Facilities, considers, in shielding calculations for radiotherapy rooms, the use of lead and/or steel to be applied on bunker walls. The NCRP Report calculations were performed foreseeing a better protection of people outside the radiotherapy room. However, contribution of lead and steel to patient dose should be taken into account for radioprotection purposes. This work presents calculations performed by MCNPX code in analyzing the Ambient Dose Equivalent due to neutron, H *(10) n , within a radiotherapy room, in the patients area, considering the use of additional shielding of 1 TVL of lead or 1 TVL of steel, positioned at the inner faces of walls and ceiling of a bunker. The head of the linear accelerator Varian 2100/2300 C/D was modeled working at 18MeV, with 5 x 5 cm 2 , 10 x 10 cm 2 , 20 x 20 cm 2 , 30 x 30 cm 2 and 40 x 40 cm 2 openings for jaws and MLC and operating in eight gantry's angles. This study shows that the use of lead generates an average value of H *(10) n at patients area, 8.02% higher than the expected when using steel. Further studies should be performed based on experimental data for comparison with those from MCNPX simulation. (author)

  20. Neutron-photon mixed field dosimetry by TLD-700 glow curve analysis and its implementation in dose monitoring for Boron Neutron Capture Therapy (BNCT) treatments

    Energy Technology Data Exchange (ETDEWEB)

    Boggio, E. F.; Longhino, J. M. [Centro Atomico Bariloche, Departamento de Fisica de Reactores y Radiaciones / CNEA, Av. E. Bustillo Km 9.5, R8402AGP San Carlos de Bariloche (Argentina); Andres, P. A., E-mail: efboggio@cab.cnea.gov.ar [Centro Atomico Bariloche, Division Proteccion Radiologica / CNEA, Av. E. Bustillo Km 9.5, R8402AGP San Carlos de Bariloche (Argentina)

    2015-10-15

    BNCT is a cancerous cells selective, non-conventional radiotherapy modality to treat malignant tumors such as glioblastoma, melanoma and recurrent head and neck cancer. It consists of a two-step procedure: first, the patient is injected with a tumor localizing drug containing a non-radioactive isotope (Boron-10) with high slow neutron capture cross-section. In a second step, the patient is irradiated with neutrons, which are absorbed by the Boron-10 agent with the subsequently nuclear reaction B- 10(n,a)Li-7, thereby resulting in dose at cellular level due to the high-Let particles. The neutron fields suitable for BNCT are characterized by high neutron fluxes and low gamma dose. Determination of each component is not an easy task, especially when the volume of measurement is quite small or inaccessible for a miniature ionization chamber, for example. A method of measuring the photon and slow neutron dose(mainly by N-14 and B-10) from the glow curve (GC) analysis of a single {sup 7}LiF thermoluminescence detector is evaluated. This method was suggested by the group headed by Dr. Grazia Gambarini. The dosemeters used were TLD-600 ({sup 6}LiF:Mg,Ti with 95.6% {sup 6}Li) and TLD-700 ({sup 7}LiF:Mg,Ti with 99.9% {sup 7}LiF) from Harshaw. Photon dose measurement using the GC analysis method with TLD-700 in mixed fields requires the relation of the two main peaks of a TLD-600 GC shape obtained from an exposition to the same neutron field, and a photon calibrated GC with TLD-700. The requirements for slow neutron dose measurements are similar. In order to properly apply the GC analysis method at the Ra-6 Research Reactor BNCT facility, measurements were carried out in a standard water phantom, fully characterized on the BNCT beam by conventional techniques (activation detectors and paired ionization chambers technique). Next, the method was implemented in whole body dose monitoring of a patient undergoing a BNCT treatment, using a Bo MAb (Bottle Manikin Absorption) phantom

  1. Off-axis dose equivalent due to secondary neutrons from uniform scanning proton beams during proton radiotherapy

    Science.gov (United States)

    Islam, M. R.; Collums, T. L.; Zheng, Y.; Monson, J.; Benton, E. R.

    2013-11-01

    The production of secondary neutrons is an undesirable byproduct of proton therapy and it is important to quantify the contribution from secondary neutrons to patient dose received outside the treatment volume. The purpose of this study is to investigate the off-axis dose equivalent from secondary neutrons experimentally using CR-39 plastic nuclear track detectors (PNTD) at ProCure Proton Therapy Center, Oklahoma City, OK. In this experiment, we placed several layers of CR-39 PNTD laterally outside the treatment volume inside a phantom and in air at various depths and angles with respect to the primary beam axis. Three different proton beams with max energies of 78, 162 and 226 MeV and 4 cm modulation width, a 5 cm diameter brass aperture, and a small snout located 38 cm from isocenter were used for the entire experiment. Monte Carlo simulations were also performed based on the experimental setup using a simplified snout configuration and the FLUKA Monte Carlo radiation transport code. The measured ratio of secondary neutron dose equivalent to therapeutic primary proton dose (H/D) ranged from 0.3 ± 0.08 mSv Gy-1 for 78 MeV proton beam to 37.4 ± 2.42 mSv Gy-1 for 226 MeV proton beam. Both experiment and simulation showed a similar decreasing trend in dose equivalent with distance to the central axis and the magnitude varied by a factor of about 2 in most locations. H/D was found to increase as the energy of the primary proton beam increased and higher H/D was observed at 135° compared to 45° and 90°. The overall higher H/D in air indicates the predominance of external neutrons produced in the nozzle rather than inside the body.

  2. Off-axis dose equivalent due to secondary neutrons from uniform scanning proton beams during proton radiotherapy

    International Nuclear Information System (INIS)

    Islam, M R; Collums, T L; Monson, J; Benton, E R; Zheng, Y

    2013-01-01

    The production of secondary neutrons is an undesirable byproduct of proton therapy and it is important to quantify the contribution from secondary neutrons to patient dose received outside the treatment volume. The purpose of this study is to investigate the off-axis dose equivalent from secondary neutrons experimentally using CR-39 plastic nuclear track detectors (PNTD) at ProCure Proton Therapy Center, Oklahoma City, OK. In this experiment, we placed several layers of CR-39 PNTD laterally outside the treatment volume inside a phantom and in air at various depths and angles with respect to the primary beam axis. Three different proton beams with max energies of 78, 162 and 226 MeV and 4 cm modulation width, a 5 cm diameter brass aperture, and a small snout located 38 cm from isocenter were used for the entire experiment. Monte Carlo simulations were also performed based on the experimental setup using a simplified snout configuration and the FLUKA Monte Carlo radiation transport code. The measured ratio of secondary neutron dose equivalent to therapeutic primary proton dose (H/D) ranged from 0.3 ± 0.08 mSv Gy −1  for 78 MeV proton beam to 37.4 ± 2.42 mSv Gy −1  for 226 MeV proton beam. Both experiment and simulation showed a similar decreasing trend in dose equivalent with distance to the central axis and the magnitude varied by a factor of about 2 in most locations. H/D was found to increase as the energy of the primary proton beam increased and higher H/D was observed at 135° compared to 45° and 90°. The overall higher H/D in air indicates the predominance of external neutrons produced in the nozzle rather than inside the body. (paper)

  3. Characterization of the neutron irradiation system for use in the Low-Dose-Rate Irradiation Facility at Sandia National Laboratories.

    Energy Technology Data Exchange (ETDEWEB)

    Franco, Manuel [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2014-08-01

    The objective of this work was to characterize the neutron irradiation system consisting of americium-241 beryllium (241AmBe) neutron sources placed in a polyethylene shielding for use at Sandia National Laboratories (SNL) Low Dose Rate Irradiation Facility (LDRIF). With a total activity of 0.3 TBq (9 Ci), the source consisted of three recycled 241AmBe sources of different activities that had been combined into a single source. The source in its polyethylene shielding will be used in neutron irradiation testing of components. The characterization of the source-shielding system was necessary to evaluate the radiation environment for future experiments. Characterization of the source was also necessary because the documentation for the three component sources and their relative alignment within the Special Form Capsule (SFC) was inadequate. The system consisting of the source and shielding was modeled using Monte Carlo N-Particle transport code (MCNP). The model was validated by benchmarking it against measurements using multiple techniques. To characterize the radiation fields over the full spatial geometry of the irradiation system, it was necessary to use a number of instruments of varying sensitivities. First, the computed photon radiography assisted in determining orientation of the component sources. With the capsule properly oriented inside the shielding, the neutron spectra were measured using a variety of techniques. A N-probe Microspec and a neutron Bubble Dosimeter Spectrometer (BDS) set were used to characterize the neutron spectra/field in several locations. In the third technique, neutron foil activation was used to ascertain the neutron spectra. A high purity germanium (HPGe) detector was used to characterize the photon spectrum. The experimentally measured spectra and the MCNP results compared well. Once the MCNP model was validated to an adequate level of confidence, parametric analyses was performed on the model to optimize for potential

  4. Effects of long-term, low dose rate fission neutron irradiation on the peripheral hematological cells in rats

    International Nuclear Information System (INIS)

    Jiang Dingwen; Lei Chengxiang; Shen Xianrong; Ma Li; Yang Xufang; Peng Wulin; Dai Shourong

    2008-01-01

    Objective: To evaluate the effects of long-term, low dose rate fission neutron irradiation on the peripheral hematological cells in rats. Methods: 96 rats were randomly divided into the control group and the irradiation group with low dose rate fission neutron ( 252 Cf, 0.35 mGy/h) irradiation 20.5 h every day. 8 rats of each group were killed at 14 d, 28 d, 42d, 56d, 70d after irradiation and 35d after the irradiation, and their peripheral hematological cells were tested respectively. Results: Compared with the control group, peripheral blood WBC was reduced significantly at the dose of 0.3Gy and 0.4Gy (P < 0.05), and was reduced remarkably at dose of 0.5Gy (P<0.01) and 35d after stopping irradiation(P<0.01). At dose of 0.2Gy, Peripheral blood RBC was abnormally higher comparing with the control group (P<0.01), accompanying with higher HCT and HGB, which suggests condensed blood. At the other point, RBC tend to become lower, but only at dose 0.5Gy, and the difference is significant comparing with control group(P <0.05). At dose of 0.3Gy, 0.4Gy and 0.5Gy, HCT were significantly lower comparing with control group. Comparing with control group, MCV was higher at 35d after stopping irradiation, and PLT was significantly lower in dose of 0.2Gy. Conclusion: Long-term irradiation with low dose rate fission neutron could significantly reduce peripheral blood WBC, with less effects on RBC and PLT. The reduced WBC could not recover at 35d after stopping irradiation. (authors)

  5. SU-E-T-108: Development of a Novel Clinical Neutron Dose Monitor for Proton Therapy Based On Twin TLD500 Chips in a Small PE Moderator

    International Nuclear Information System (INIS)

    Hentschel, R; Mukherjee, B

    2014-01-01

    Purpose: In proton therapy, it could be desirable to measure out-of-field fast neutron doses at critical locations near and outside the patient body. Methods: The working principle of a novel clinical neutron dose monitor is verified by MCNPX simulation. The device is based on a small PE moderator of just 5.5cm side length for easy handling covered with a thermal neutron suppression layer. In the simulation, a polystyrene phantom is bombarded with a standard proton beam. The secondary thermal neutron flux produced inside the moderator by the impinging fast neutrons from the treatment volume is estimated by pairs of α-Al2O3:C (TLD500) chips which are evaluated offline after the treatment either by TL or OSL methods. The first chip is wrapped with 0.5mm natural Gadolinium foil converting the thermal neutrons to gammas via (n,γ) reaction. The second chip is wrapped with a dummy material. The chip centers have a distance of 2cm from each other. Results: The simulation shows that the difference of gamma doses in the TLD500 chips is correlated to the mean fast neutron dose delivered to the moderator material. Different outer shielding materials have been studied. 0.5mm Cadmium shielding is preferred for cost reasons and convenience. Replacement of PE moderator material by other materials like lead or iron at any place is unfavorable. The spatial orientation of the moderator cube is uncritical. Using variance reduction techniques like splitting/Russian roulette, the TLD500 gamma dose simulation give positive differences up to distances of 0.5m from the treatment volume. Conclusion: Applicability and basic layout of a novel clinical neutron dose monitor are demonstrated. The monitor measures PE neutron doses at locations outside the patient body up to distances of 0.5m from the treatment volume. Tissue neutron doses may be calculated using neutron kerma factors

  6. Neutron spectra calculation and doses in a subcritical nuclear reactor based on thorium

    International Nuclear Information System (INIS)

    Medina C, D.; Hernandez A, P. L.; Hernandez D, V. M.; Vega C, H. R.; Sajo B, L.

    2015-10-01

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

  7. Morphological differences in the response of mouse small intestine to radiobiologically equivalent doses of X and neutron irradiation

    International Nuclear Information System (INIS)

    Carr, K.E.; Hamlet, R.; Nias, A.H.; Watt, C.

    1984-01-01

    A scale has been developed to describe the effects of radiation on small intestinal villi. The scale has been used to compare the damage done to the villi in the period 0-5 days after irradiation by X-irradiation or neutron irradiation, using 10 Gy X-rays and 5 Gy neutrons, doses which are radiobiologically equivalent when assessed by the microcolony assay method. Use of the scale indicates that the damage done to the villi by neutrons is greater than that produced by X-rays. This has implications for the interpretation of radiobiological equivalent doses (R.B.E.). Resin light microscopy and transmission electron microscopy (T.E.M.) have also been used to examine small intestinal damage after 10 Gy X-irradiation and 5 Gy neutron irradiation. Differences include variations in crypt shape, mitotic activity and the proportion of crypts which are heavily parasitised. As well as the differences in villous shape which have been reflected in the different values on the scoring system, there are also variations in the response of the constituent cells of the epithelial compartment of the villi. In general, the effect of the neutron irradiation is more severe than that of the X-rays, particularly as would be suggested by a simple quantitation of crypt regeneration

  8. Determination of the dose equivalents due to neutrons produced during therapeutic irradiations with a Varian CLINAC 2500

    International Nuclear Information System (INIS)

    Carrillo, Ricardo E.

    1991-01-01

    This experiment it was designed to quantify that so important it is the dose equivalent deposited by the neutron flow that is generated by photonuclear reactions during therapeutic irradiations with X rays of produced high-energy for an accelerator Varian CLINAC 2500. This accelerator type is routinely used in the Department of Radiotherapy of the Hospital of the University of Wisconsin, E.U. The equivalent dose was measured in diverse towns of the room of irradiations using the activation of thin sheets of gold put in the center of plastic recipients full with water. In general, the recipients were 1 m or more than the floor and at distances still bigger than the walls. The irradiations were made using photons with the highest energy that you can select with this team - 24 MeV. The due equivalent dose to neutrons taken place here by the energy photons used they were measured and reported. (author)

  9. Optimization of shadow cone length and mass for determination the amount of scattered radiation dose in the calibration laboratory of Am/Be neutron source

    International Nuclear Information System (INIS)

    Raisali, G.; Hamidi, S.; Hallajfard, E.; Shahvar, A.; Hajiloo, N.

    2007-01-01

    The shadow cone technique is one of the methods which is used for determining the contribution of scattered particles on the response of neutron detectors. This technique is used for neutron field calibration in Agriculture, Medicine and Industry Research School. In this investigation, we have designed and constructed an optimized shadow cone. According to the calculated neutron dose equivalent attenuation factors, a cone with 20 cm of iron and 30 cm of polyethylene has been found as optimum. For this cone, the neutron dose equivalent attenuation factor for 241 Am/Be neutron source, is 0.00035 for which the contribution of scattered neutrons in Agriculture, Medicine and Industry Research School neutron calibration laboratory according to the calculation and measurement results, can be evaluated with less than 0.5% of error

  10. Desorption of tritium and helium from high dose neutron irradiated beryllium

    Science.gov (United States)

    Kupriyanov, I. B.; Nikolaev, G. N.; Vlasov, V. V.; Kovalev, A. M.; Chakin, V. P.

    2007-08-01

    The effect of high dose neutron irradiation on tritium and helium desorption in beryllium is described. Beryllium samples were irradiated in the SM and BOR-60 reactors to a neutron fluences ( E > 0.1 MeV) of (5-16) × 10 22 cm -2 at 70-100 °C and 380-420 °C. A mass-spectrometry technique was used in out of pile tritium release experiments during stepped annealing in the 250-1300 °C temperature range. The total amount of helium accumulated in irradiated beryllium samples varied from 6000 to 7200 appm. The first signs of tritium and helium release were detected at temperature of 312-445 °C and 500-740 °C, respectively. It is shown that most tritium (˜82%) from sample irradiated at 70-100 °C releases in temperature range of 312-700 °C before the beginning of helium release (740 °C). In the case of beryllium sample irradiated at 380-420 °C, tritium release starts at a higher temperature ( Ts > Tann = 445 °C) and most of the tritium (˜99.8%) is released concurrently with helium which could be considered as evidence of co-existence of partial amounts of tritium and helium in common bubbles. Both the Be samples differ little in the upper temperatures of gas release: 745 and 775 °C for tritium; 1140 and 1160 °C for helium. Swelling of beryllium starts to play a key role in accelerating tritium release at Tann > 600 °C and in helium release - at Tann > 750 °C.

  11. DIANE, a simulation code for the interaction of neutrons with living tissues. Application to low doses of fast neutrons on human tumoral cells; DIANE, un code de simulation de l'interaction des neutrons avec la matiere vivante. Applications aux faibles doses de neutrons rapides sur des cellules tumorales humaines

    Energy Technology Data Exchange (ETDEWEB)

    Nenot, M.L

    2003-07-15

    Our work deals with the irradiation of cells and living tissues by 14 MeV neutrons at very low doses (a few 10{sup -2} Gy). Such experiments require an accurate knowledge of the values of neutron dose rates and fluences at the level of cell cultures. We have performed measurements of fluence rates through an activation method applied to gold and copper foils. The fluence rate is deduced from the gamma rays emitted by the irradiated foils. Neutron doses and dose rates have been measured through varied methods: PIN diodes, ionization tissue equivalent chambers, and Geiger-Mueller counters. We have designed the DIANE code to simulate the impact of energetic neutrons on cells. This code can be used with isolated cells or macroscopic tissues, it takes into account the roles of the ionisation electrons produced by recoil nuclei entering the cell. This point is all the more important since recent works have highlighted the impact of very low energy electrons on DNA. (A.C.)

  12. Evaluation of response function of moderating-type neutron detector and application to environmental neutron measurement

    International Nuclear Information System (INIS)

    Kosako, Toshiso; Nakamura, Takashi; Iwai, Satoshi; Katsuki, Shinji; Kamata, Masashi.

    1983-08-01

    The energy-dependent response function of a multi-cylinder moderating-type BF 3 counter, so-called Bonner counter, was calculated by the time-dependent multi-group Monte Carlo code, TMMCR. The calculated response function was evaluated experimentally for neutron energy below about 50 keV down to epithermal energy by the time-of-flight method combining with a large lead pile at the Nuclear Engineering Research Laboratory, University of Tokyo and also above 50 keV by using the monoenergetic neutron standard field a t the Electrotechnical Laboratory. The time delay in the polyethylene moderator of the Bonner counter due to multiple collisions with hydrogen was analyzed by the TMMCR code and used for the time-spectrum analysis of the time-of-flight measurement. The response function obtained by these two experiments showed good agreement with the calculated results. This Bonner counter having a response function evaluated from thermal to MeV energy range was used for spectrometry and dosimetry of environmental neutrons around some nuclear facilities. The neutron spectra and dose measured in the environment around a 252 Cf fission source, fast neutron source reactor and electron synchrotron were all in good agreement with the calculated results and the measured results with other neutron detectors. (author)

  13. Dose-response curve for blood exposed to gamma-neutron mixed field by conventional cytogenetic method

    International Nuclear Information System (INIS)

    Brandao, Jose Odinilson de C.; Souza, Priscilla L.G.; Santos, Joelan A.L.; Vilela, Eudice C.; Lima, Fabiana F.; Calixto, Merilane S.; Santos, Neide

    2009-01-01

    There is increasing concern about airline crew members (about one million worldwide) are exposed to measurable neutrons doses. Historically, cytogenetic biodosimetry assays have been based on quantifying asymmetrical chromosome alterations (dicentrics, centric rings and acentric fragments) in mytogen-stimulated T-lymphocytes in their first mitosis after radiation exposure. Increased levels of chromosome damage in peripheral blood lymphocytes are a sensitive indicator of radiation exposure and they are routinely exploited for assessing radiation absorbed dose after accidental or occupational exposure. Since radiological accidents are not common, not all nations feel that it is economically justified to maintain biodosimetry competence. However, dependable access to biological dosimetry capabilities is completely critical in event of an accident. In this paper the dose-response curve was measured for the induction of chromosomal alterations in peripheral blood lymphocytes after chronic exposure in vitro to neutron-gamma mixes field. Blood was obtained from one healthy donor and exposed to two neutron-gamma mixed field from sources 241 AmBe (20 Ci) at the Neutron Calibration Laboratory (NCL-CRCN/NE-PE-Brazil). The evaluated absorbed doses were 0.2 Gy; 1.0 Gy and 2.5 Gy. The dicentric chromosomes were observed at metaphase, following colcemid accumulation and 1000 well-spread metaphase figures were analyzed for the presence of dicentrics by two experienced scorers after painted by giemsa 5%. Our preliminary results showed a linear dependence between radiations absorbed dose and dicentric chromosomes frequencies. Dose-response curve described in this paper will contribute to the construction of calibration curve that will be used in our laboratory for biological dosimetry. (author)

  14. Dose-response curve for blood exposed to gamma-neutron mixed field by conventional cytogenetic method

    Energy Technology Data Exchange (ETDEWEB)

    Brandao, Jose Odinilson de C.; Souza, Priscilla L.G.; Santos, Joelan A.L.; Vilela, Eudice C.; Lima, Fabiana F., E-mail: jodinilson@cnen.gov.b, E-mail: fflima@cnen.gov.b, E-mail: jasantos@cnen.gov.b [Centro Regional de Ciencias Nucleares do Nordeste (CRCN-NE/CNEN-PE), Recife, PE (Brazil); Calixto, Merilane S.; Santos, Neide, E-mail: santos_neide@yahoo.com.b [Universidade Federal de Pernambuco (UFPE), Recife, PE (Brazil). Dept. de Genetica

    2009-07-01

    There is increasing concern about airline crew members (about one million worldwide) are exposed to measurable neutrons doses. Historically, cytogenetic biodosimetry assays have been based on quantifying asymmetrical chromosome alterations (dicentrics, centric rings and acentric fragments) in mytogen-stimulated T-lymphocytes in their first mitosis after radiation exposure. Increased levels of chromosome damage in peripheral blood lymphocytes are a sensitive indicator of radiation exposure and they are routinely exploited for assessing radiation absorbed dose after accidental or occupational exposure. Since radiological accidents are not common, not all nations feel that it is economically justified to maintain biodosimetry competence. However, dependable access to biological dosimetry capabilities is completely critical in event of an accident. In this paper the dose-response curve was measured for the induction of chromosomal alterations in peripheral blood lymphocytes after chronic exposure in vitro to neutron-gamma mixes field. Blood was obtained from one healthy donor and exposed to two neutron-gamma mixed field from sources {sup 241}AmBe (20 Ci) at the Neutron Calibration Laboratory (NCL-CRCN/NE-PE-Brazil). The evaluated absorbed doses were 0.2 Gy; 1.0 Gy and 2.5 Gy. The dicentric chromosomes were observed at metaphase, following colcemid accumulation and 1000 well-spread metaphase figures were analyzed for the presence of dicentrics by two experienced scorers after painted by giemsa 5%. Our preliminary results showed a linear dependence between radiations absorbed dose and dicentric chromosomes frequencies. Dose-response curve described in this paper will contribute to the construction of calibration curve that will be used in our laboratory for biological dosimetry. (author)

  15. Microstructure-tensile property correlation of 316SS in low dose neutron irradiations

    International Nuclear Information System (INIS)

    Yoshida, N.; Muroga, T.; Araki, K.; Heinisch, H.L.; Kiritani, M.

    1990-05-01

    The objective of this work is to determine the effects of the neutron spectrum on radiation-induced changes in mechanical properties for metals irradiated with fission and fusion neutrons. 10 refs., 6 figs

  16. Measuring element for the detection and determination of radiation doses of gamma radiation and neutrons

    International Nuclear Information System (INIS)

    Jahn, W.; Piesch, E.

    1975-01-01

    A measuring element detects and proves both gamma and neutron radiation. The element includes a photoluminescent material which stores gamma radiation and particles of arsenic and phosphorus embedded in the photoluminescent material for detecting neutron radiation. (U.S.)

  17. An Analytical Model of Leakage Neutron Equivalent Dose for Passively-Scattered Proton Radiotherapy and Validation with Measurements

    Energy Technology Data Exchange (ETDEWEB)

    Schneider, Christopher; Newhauser, Wayne, E-mail: newhauser@lsu.edu [Department of Physics and Astronomy, Louisiana State University and Agricultural and Mechanical College, 202 Nicholson Hall, Baton Rouge, LA 70803 (United States); Mary Bird Perkins Cancer Center, 4950 Essen Lane, Baton Rouge, LA 70809 (United States); Farah, Jad [Institut de Radioprotection et de Sûreté Nucléaire, Service de Dosimétrie Externe, BP-17, 92262 Fontenay-aux-Roses (France)

    2015-05-18

    Exposure to stray neutrons increases the risk of second cancer development after proton therapy. Previously reported analytical models of this exposure were difficult to configure and had not been investigated below 100 MeV proton energy. The purposes of this study were to test an analytical model of neutron equivalent dose per therapeutic absorbed dose (H/D) at 75 MeV and to improve the model by reducing the number of configuration parameters and making it continuous in proton energy from 100 to 250 MeV. To develop the analytical model, we used previously published H/D values in water from Monte Carlo simulations of a general-purpose beamline for proton energies from 100 to 250 MeV. We also configured and tested the model on in-air neutron equivalent doses measured for a 75 MeV ocular beamline. Predicted H/D values from the analytical model and Monte Carlo agreed well from 100 to 250 MeV (10% average difference). Predicted H/D values from the analytical model also agreed well with measurements at 75 MeV (15% average difference). The results indicate that analytical models can give fast, reliable calculations of neutron exposure after proton therapy. This ability is absent in treatment planning systems but vital to second cancer risk estimation.

  18. An Analytical Model of Leakage Neutron Equivalent Dose for Passively-Scattered Proton Radiotherapy and Validation with Measurements

    International Nuclear Information System (INIS)

    Schneider, Christopher; Newhauser, Wayne; Farah, Jad

    2015-01-01

    Exposure to stray neutrons increases the risk of second cancer development after proton therapy. Previously reported analytical models of this exposure were difficult to configure and had not been investigated below 100 MeV proton energy. The purposes of this study were to test an analytical model of neutron equivalent dose per therapeutic absorbed dose (H/D) at 75 MeV and to improve the model by reducing the number of configuration parameters and making it continuous in proton energy from 100 to 250 MeV. To develop the analytical model, we used previously published H/D values in water from Monte Carlo simulations of a general-purpose beamline for proton energies from 100 to 250 MeV. We also configured and tested the model on in-air neutron equivalent doses measured for a 75 MeV ocular beamline. Predicted H/D values from the analytical model and Monte Carlo agreed well from 100 to 250 MeV (10% average difference). Predicted H/D values from the analytical model also agreed well with measurements at 75 MeV (15% average difference). The results indicate that analytical models can give fast, reliable calculations of neutron exposure after proton therapy. This ability is absent in treatment planning systems but vital to second cancer risk estimation

  19. An analytical model of leakage neutron equivalent dose for passively-scattered proton radiotherapy and validation with measurements.

    Science.gov (United States)

    Schneider, Christopher; Newhauser, Wayne; Farah, Jad

    2015-05-18

    Exposure to stray neutrons increases the risk of second cancer development after proton therapy. Previously reported analytical models of this exposure were difficult to configure and had not been investigated below 100 MeV proton energy. The purposes of this study were to test an analytical model of neutron equivalent dose per therapeutic absorbed dose  at 75 MeV and to improve the model by reducing the number of configuration parameters and making it continuous in proton energy from 100 to 250 MeV. To develop the analytical model, we used previously published H/D values in water from Monte Carlo simulations of a general-purpose beamline for proton energies from 100 to 250 MeV. We also configured and tested the model on in-air neutron equivalent doses measured for a 75 MeV ocular beamline. Predicted H/D values from the analytical model and Monte Carlo agreed well from 100 to 250 MeV (10% average difference). Predicted H/D values from the analytical model also agreed well with measurements at 75 MeV (15% average difference). The results indicate that analytical models can give fast, reliable calculations of neutron exposure after proton therapy. This ability is absent in treatment planning systems but vital to second cancer risk estimation.

  20. The new remcounter LB6411: Measurement of neutron ambient dose equivalent H*(10) according to ICRP60 with high sensitivity

    International Nuclear Information System (INIS)

    Klett, A.; Burgkhardt, B.

    1996-01-01

    Since the International Commission on Radiological Protection has issued in publication ICRP60 new recommendations on radiation protection quantities, in neutron monitoring there is now increasing Interest in commercially available instruments optimized and calibrated for the measurement of ambient dose equivalent H*(10). Therefore within a joint cooperation between the Research Center Karlsruhe and EG ampersand G Berthold the neutron-dose-rate meter LB6411 was newly developed. The detector system with integrated electronics has a 3 He proportional counter tube centered in a moderating sphere. The response between thermal energies and 20 MeV was optimized with the help of extensive MCNP Monte-Carlo calculations. The instrument has extremely high sensitivity of approximately 3 counts per nSv and can be used both as a portable or as a stationary neutron monitor. Fluence responses and angular dependencies had been measured in monoenergetic neutron beams provided by the Physikalisch-Technische Bundesanstalt (PTB) in Braunschweig, Germany. The ambient dose equivalent response of the LB6411 is reported over the whole energy range

  1. Application of the Monte Carlo method to estimate doses due to neutron activation of different materials in a nuclear reactor

    Science.gov (United States)

    Ródenas, José

    2017-11-01

    All materials exposed to some neutron flux can be activated independently of the kind of the neutron source. In this study, a nuclear reactor has been considered as neutron source. In particular, the activation of control rods in a BWR is studied to obtain the doses produced around the storage pool for irradiated fuel of the plant when control rods are withdrawn from the reactor and installed into this pool. It is very important to calculate these doses because they can affect to plant workers in the area. The MCNP code based on the Monte Carlo method has been applied to simulate activation reactions produced in the control rods inserted into the reactor. Obtained activities are introduced as input into another MC model to estimate doses produced by them. The comparison of simulation results with experimental measurements allows the validation of developed models. The developed MC models have been also applied to simulate the activation of other materials, such as components of a stainless steel sample introduced into a training reactors. These models, once validated, can be applied to other situations and materials where a neutron flux can be found, not only nuclear reactors. For instance, activation analysis with an Am-Be source, neutrography techniques in both medical applications and non-destructive analysis of materials, civil engineering applications using a Troxler, analysis of materials in decommissioning of nuclear power plants, etc.

  2. Dose and dose rate extrapolation factors for malignant and non-malignant health endpoints after exposure to gamma and neutron radiation

    Energy Technology Data Exchange (ETDEWEB)

    Tran, Van; Little, Mark P. [National Cancer Institute, Radiation Epidemiology Branch, Rockville, MD (United States)

    2017-11-15

    Murine experiments were conducted at the JANUS reactor in Argonne National Laboratory from 1970 to 1992 to study the effect of acute and protracted radiation dose from gamma rays and fission neutron whole body exposure. The present study reports the reanalysis of the JANUS data on 36,718 mice, of which 16,973 mice were irradiated with neutrons, 13,638 were irradiated with gamma rays, and 6107 were controls. Mice were mostly Mus musculus, but one experiment used Peromyscus leucopus. For both types of radiation exposure, a Cox proportional hazards model was used, using age as timescale, and stratifying on sex and experiment. The optimal model was one with linear and quadratic terms in cumulative lagged dose, with adjustments to both linear and quadratic dose terms for low-dose rate irradiation (<5 mGy/h) and with adjustments to the dose for age at exposure and sex. After gamma ray exposure there is significant non-linearity (generally with upward curvature) for all tumours, lymphoreticular, respiratory, connective tissue and gastrointestinal tumours, also for all non-tumour, other non-tumour, non-malignant pulmonary and non-malignant renal diseases (p < 0.001). Associated with this the low-dose extrapolation factor, measuring the overestimation in low-dose risk resulting from linear extrapolation is significantly elevated for lymphoreticular tumours 1.16 (95% CI 1.06, 1.31), elevated also for a number of non-malignant endpoints, specifically all non-tumour diseases, 1.63 (95% CI 1.43, 2.00), non-malignant pulmonary disease, 1.70 (95% CI 1.17, 2.76) and other non-tumour diseases, 1.47 (95% CI 1.29, 1.82). However, for a rather larger group of malignant endpoints the low-dose extrapolation factor is significantly less than 1 (implying downward curvature), with central estimates generally ranging from 0.2 to 0.8, in particular for tumours of the respiratory system, vasculature, ovary, kidney/urinary bladder and testis. For neutron exposure most endpoints, malignant and

  3. CTD: a computer program to solve the three dimensional multi-group diffusion equation in X, Y, Z, and triangular Z geometries

    Energy Technology Data Exchange (ETDEWEB)

    Fletcher, J K

    1973-05-01

    CTD is a computer program written in Fortran 4 to solve the multi-group diffusion theory equations in X, Y, Z and triangular Z geometries. A power print- out neutron balance and breeding gain are also produced. 4 references. (auth)

  4. CINESP - computational program of spatial kinetics for nuclear reactors in the one-two dimension multigroup diffusion theory

    International Nuclear Information System (INIS)

    Santos, R.S. dos

    1993-01-01

    This paper presents a computational program to solve numerically the reactor kinetics equations in the multigroup diffusion theory. One or two-dimensional problems in cylindrical or Cartesian geometries, with any number of energy and delayed-neutron precursors groups are dealt with. The main input and output of the program are briefly discussed. Various results demonstrate the accuracy and versatility of the program, when compared with other kinetics programs. (author)

  5. Distributions of neutron yields and doses around a water phantom bombarded with 290-MeV/nucleon and 430-MeV/nucleon carbon ions

    Energy Technology Data Exchange (ETDEWEB)

    Satoh, D., E-mail: satoh.daiki@jaea.go.jp [Japan Atomic Energy Agency, Tokai-mura, Naka-gun, Ibaraki 319-1195 (Japan); Kajimoto, T. [Hiroshima University, Kagamiyama, Higashi-Hiroshima-shi, Hiroshima 739-8527 (Japan); Shigyo, N.; Itashiki, Y.; Imabayashi, Y. [Kyushu University, Motooka, Nishi-ku, Fukuoka 819-0395 (Japan); Koba, Y.; Matsufuji, N. [National Institute of Radiological Sciences, Anagawa, Inage-ku, Chiba 263-8555 (Japan); Sanami, T. [High Energy Accelerator Research Organization, Oho-cho, Tsukuba-shi, Ibaraki 305-0801 (Japan); Nakao, N. [Shimizu Corporation, Etchujima, Koto-ku, Tokyo 135-8530 (Japan); Uozumi, Y. [Kyushu University, Motooka, Nishi-ku, Fukuoka 819-0395 (Japan)

    2016-11-15

    Double-differential neutron yields from a water phantom bombarded with 290-MeV/nucleon and 430-MeV/nucleon carbon ions were measured at emission angles of 15°, 30°, 45°, 60°, 75°, and 90°, and angular distributions of neutron yields and doses around the phantom were obtained. The experimental data were compared with results of the Monte-Carlo simulation code PHITS. The PHITS results showed good agreement with the measured data. On the basis of the PHITS simulation, we estimated the angular distributions of neutron yields and doses from 0° to 180° including thermal neutrons.

  6. Using the probability method for multigroup calculations of reactor cells in a thermal energy range

    International Nuclear Information System (INIS)

    Rubin, I.E.; Pustoshilova, V.S.

    1984-01-01

    The possibility of using the transmission probability method with performance inerpolation for determining spatial-energy neutron flux distribution in cells of thermal heterogeneous reactors is considered. The results of multigroup calculations of several uranium-water plane and cylindrical cells with different fuel enrichment in a thermal energy range are given. A high accuracy of results is obtained with low computer time consumption. The use of the transmission probability method is particularly reasonable in algorithms of the programmes compiled computer with significant reserve of internal memory

  7. Mining the multigroup-discrete ordinates algorithm for high quality solutions

    International Nuclear Information System (INIS)

    Ganapol, B.D.; Kornreich, D.E.

    2005-01-01

    A novel approach to the numerical solution of the neutron transport equation via the discrete ordinates (SN) method is presented. The new technique is referred to as 'mining' low order (SN) numerical solutions to obtain high order accuracy. The new numerical method, called the Multigroup Converged SN (MGCSN) algorithm, is a combination of several sequence accelerators: Romberg and Wynn-epsilon. The extreme accuracy obtained by the method is demonstrated through self consistency and comparison to the independent semi-analytical benchmark BLUE. (authors)

  8. Specifications for a two-dimensional multi-group scattering code: ALCI

    International Nuclear Information System (INIS)

    Bayard, J.P.; Guillou, A.; Lago, B.; Bureau du Colombier, M.J.; Guillou, G.; Vasseur, Ch.

    1965-02-01

    This report describes the specifications of the ALCI programme. This programme resolves the system of difference equations similar to the homogeneous problem of multigroup neutron scattering, with two dimensions in space, in the three geometries XY, RZ, RΘ. It is possible with this method to calculate geometric and composition criticalities and also to calculate the accessory problem on demand. The maximum number of points dealt with is 6000. The maximum permissible number of groups is 12. The internal iterations are treated by the method of alternating directions. The external iterations are accelerated using the extrapolation method due to Tchebychev. (authors) [fr

  9. Development of a polynomial nodal model to the multigroup transport equation in one dimension

    International Nuclear Information System (INIS)

    Feiz, M.

    1986-01-01

    A polynomial nodal model that uses Legendre polynomial expansions was developed for the multigroup transport equation in one dimension. The development depends upon the least-squares minimization of the residuals using the approximate functions over the node. Analytical expressions were developed for the polynomial coefficients. The odd moments of the angular neutron flux over the half ranges were used at the internal interfaces, and the Marshak boundary condition was used at the external boundaries. Sample problems with fine-mesh finite-difference solutions of the diffusion and transport equations were used for comparison with the model

  10. The simulated measurements of area and personal neutron-gamma dose equivalent in the building of HWRR

    International Nuclear Information System (INIS)

    Chen Changmao; Wen Youqin; Su Jingling; Liu Shuying; Liu Nairong

    1988-01-01

    The measuring methods and results for area and personal n-γ dose equivalent in the building of HWRR of Institute of Atomic Energy were reported. The reactor operated 4440 hours during 1985, the average themal power was 11 MW. The average area n-γ dose equivalents of the basement, experimental hall, corridors and laboratories in the building were 12.2, 11.6, 0.45 and 0.23 cSv/a, respectively. The fraction of the neutron dose equivalent in any working area was less than 21%. The average personal n-γ dose equivalent to radiation workers in the building was about 0.49 cSv/a, the γ dose equivalent was a major component. The measuring methods were compared

  11. Dose Determination using alanine detectors in a Mixed Neutron and Gamma Field for Boron Neutron Capture Therapy of Liver Malignancies

    DEFF Research Database (Denmark)

    Schmitz, T.; Blaickner, M.; Ziegner, M.

    2011-01-01

    Introduction Boron Neutron Capture Therapy for liver malignancies is being investigated at the University of Mainz. One important aim is the set-up of a reliable dosimetry system. Alanine dosimeters have previously been applied for dosimetry of mixed radiation fields in antiproton therapy, and ma...

  12. Macroscopic multigroup constants for accelerator driven system core calculation

    International Nuclear Information System (INIS)

    Heimlich, Adino; Santos, Rubens Souza dos

    2011-01-01

    The high-level wastes stored in facilities above ground or shallow repositories, in close connection with its nuclear power plant, can take almost 106 years before the radiotoxicity became of the order of the background. While the disposal issue is not urgent from a technical viewpoint, it is recognized that extended storage in the facilities is not acceptable since these ones cannot provide sufficient isolation in the long term and neither is it ethical to leave the waste problem to future generations. A technique to diminish this time is to transmute these long-lived elements into short-lived elements. The approach is to use an Accelerator Driven System (ADS), a sub-critical arrangement which uses a Spallation Neutron Source (SNS), after separation the minor actinides and the long-lived fission products (LLFP), to convert them to short-lived isotopes. As an advanced reactor fuel, still today, there is a few data around these type of core systems. In this paper we generate macroscopic multigroup constants for use in calculations of a typical ADS fuel, take into consideration, the ENDF/BVI data file. Four energy groups are chosen to collapse the data from ENDF/B-VI data file by PREPRO code. A typical MOX fuel cell is used to validate the methodology. The results are used to calculate one typical subcritical ADS core. (author)

  13. The equidosemeter ED-02 as a device for dose equivalent measurements in mixed neutron and photon radiation fields

    International Nuclear Information System (INIS)

    Abrosimov, A.I.; Alekseev, A.G.; Antipov, V.A.; Golovachik, V.T.

    1985-01-01

    The equidosemeter ED-02 is to be used for simultaneous measurements of the dose equivalent, absorbed dose, and mean quality factor of mixed radiations. The detector is a tissue equivalent spherical low-pressure proportional counter tube the signal of which is simultaneously recorded in two channels - a current channel and a pulse one. The current channel is linear and its response proportional to the absorbed dose. The pulse channel includes a nonlinear pulse amplitude converter the characteristic of which, taking into account the required dependence of the mean quality factor on linear energy transfer, has been chosen in such a way that in final counting the pulse channel response is proportional to the difference between dose equivalent and absorbed dose. On the basis of calculations of event spectra in the sensitive volume of the detector, the energy dependence of the dosemeter sensitivity is analysed for neutron energies up to 20 MeV. The characteristic of the nonlinear converter has been calculated on the basis of the construction parameters of the detector and optimized with respect to a representative sample of neutron spectra beyond the shields of nuclear plants. The heterogeneity of the detector, i.e. the difference between the atomic composition of wall and filling and the composition of soft biological tissue as well as the effect of the conducting coating of the case cathode, has been taken into consideration. Moreover, the test results of the device in mixed neutron-photon fields of 60 Co, 239 Pu-α-Be and 252 Cf radioisotope sources are presented. The main measuring error of dose characteristics is shown to be less than 20% in the dose range 1 x 10 -3 to 4 x 10 -3 Sv/h. (author)

  14. Introduction of corrections taking into account interdependence of multigroup constants to the results of multigroup perturbation theory calculations

    International Nuclear Information System (INIS)

    Raskach, K. F.

    2012-01-01

    In multigroup calculations of reactivity and sensitivity coefficients, methodical errors can appear if the interdependence of multigroup constants is not taken into account. For this effect to be taken into account, so-called implicit components of the aforementioned values are introduced. A simple technique for computing these values is proposed. It is based on the use of subgroup parameters.

  15. An investigation of methods for neutron dose measurement in high temperature irradiation fields

    Energy Technology Data Exchange (ETDEWEB)

    Kosako, Toshisou; Sugiura, Nobuyuki [Tokyo Univ. (Japan); Kudo, Kazuhiko [Kyushu Univ., Fukuoka (Japan)] [and others

    2000-10-01

    The Japan Atomic Energy Research Institute (JAERI) has been conducting the innovative basic research on high temperature since 1994, which is a series of high temperature irradiation studies using the High Temperature Engineering Test Reactor (HTTR). 'The Task Group for Evaluation of Irradiation Dose under High Temperature Radiation' was founded in the HTTR Utilization Research Committee, which is the promoting body of the innovative basic research. The present report is a summary of investigation which has been made by the Task Group on the present status and subjects of research and development of neutron detectors in high temperature irradiation fields, in view of contributing to high temperature irradiation research using the HTTR. Detectors investigated here in the domestic survey are the following five kinds of in-core detectors: 1) small fission counter, 2) small fission chamber, 3) self-powered detector, 4) activation detector, and 5) optical fiber. In addition, the research and development status in Russia has been investigated. The present report will also be useful as nuclear instrumentation of high temperature gas-cooled reactors. (author)

  16. Survival of tumor bearing mice by sequencing of low dose rate (LDR) neutron and photon radiation

    International Nuclear Information System (INIS)

    Onomura, C.I.; Feola, J.M.; Maruyama, Y.

    1984-01-01

    Cf-252 neutron radiation (NT) has been shown to be effective therapy for bulky, hypoxic human tumor and to produce consistent rapid clearance and 5 year cures. NT has been found to be more or less effective depending upon the schedule in which it is used and upon mixing with photon radiation. In an effort to study this scheduling and photon effect, LSA tumor was irradiated in vivo in a hypoxic, advanced state, in different schedules in combination of NT with Co-60 photons. The LSA lymphoma of C57BL/ym mice represents an accurate system to assess dose-response of tumor cells in vivo. Mean survival time was used as endpoint. A high RBE for LDR Cf-252 NT was observed with a RBE(n) of -- 5.0. The effect was not greatly sensitive to sequence in which photons were used. Comparison studies were also tested relative to LDR Cs-137 photon radiation. The results support the high efficacy of LDR NT for destruction of hypoxic tumor in vivo

  17. Neutron equivalent dose rates at the surroundings of the electron linear accelerator operated by the university of Sao Paulo - Physics institute

    International Nuclear Information System (INIS)

    Yanagihara, L.S.

    1984-01-01

    For the determination of the neutron dose rates at the surroundings of an electron linear accelerators it is necessary the knowledge of the neutron spectrum or its mean energy, because the conversion factor of the flux in equivalent dose rates, is strongly dependent on the neutron energy. Taking this fact into consideration, equivalent dose rates were determined in the three representative sites of the IF/USP Linear Electron Accelerator. Also, due to the radiation field be pulsed, a theoretical and experimental study has been realized to evaluate the effect produced by the variation of the field on the detector. (author)

  18. Dose-effect relationships for fife shortening, tumorigenesis, and systemic injuries in mice irradiated with fission neutron or 60Co gamma radiation

    International Nuclear Information System (INIS)

    Ainsworth, E.J.; Fry, R.J.M.; Williamson, F.S.; Brennan, P.C.; Stearner, S.P.; Yang, V.V.; Crouse, D.A.; Rust, J.H.; Borak, T.B.

    1977-01-01

    The objective of this research is to provide additional data on life shortening, neoplastic and non-neoplastic diseases, and other systematic injuries necessary for the determination of dose-response relationships. The data are used to test existing predictive models and formulate new models which may assist with radiation risk assessment. Late somatic effects of fission neutrons from the JANUS reactor or from cobalt-60 gamma radiation are evaluated in young adult B6CF 1 mice that receive either a range of single doses or protracted doses at low dose rates; the protracted irradiation is administered over a 6-month period. After single doses of gamma radiation the relationship between radiation dose and percent life shortening appears linear whereas after single doses of fission spectrum neutrons a non-linear dose response is observed. These results suggest that estimates of radiation risk for fission spectrum neutrons should take into account the following: the curvilinearity of the neutron dose-response curve for life shortening, and the increased life shortening produced by neutron dose fractionation

  19. SU-E-T-611: Photon and Neutron Peripheral Dose Ratio for Low (6 MV) and High (15 MV) Energy for Treatment Selection

    Energy Technology Data Exchange (ETDEWEB)

    Irazola, L; Sanchez-Doblado, F [Departamento de Fisiologia Medica y Biofisica, Universidad de Seville (Spain); Servicio de Radiofisica, Hospital Universitario Virgen Macarena, Seville (Spain); Terron, J; Ortiz-Seidel, M [Servicio de Radiofisica, Hospital Universitario Virgen Macarena, Seville (Spain); Departamento de Fisiologia Medica y Biofisica, Universidad de Seville (Spain); Sanchez-Nieto, B [Instituto de Fisica, Pontificia Universidad Catolica de Chile, Santiago (Chile)

    2015-06-15

    Purpose: Differences between radiotherapy techniques and energies, can offer improvements in tumor coverage and organs at risk preservation. However, a more complete decision should include peripheral doses delivered to the patient. The purpose of this work is the balance of photon and neutron peripheral doses for a prostate case solved with 6 different treatment modalities. Methods: Inverse and Forward IMRT and 3D-CRT in 6 and 15 MV for a Siemens Primus linac, using the same CT data set and contours. The methodology described in [1], was used with the TNRD thermal neutron detector [2] for neutron peripheral dose estimation at 7 relevant organs (colon, esophagus, stomach, liver, lung, thyroid and skin). Photon doses were estimated for these organs by terms of the algorithm proposed in [3]. Plans were optimized with the same restrictions and limited to 30 segments in the Inverse case. Results: A similar photon peripheral dose was found comparing 6 and 15 MV cases with slightly higher values of (1.9 ± 1.6) % in mean, for the 6 MV cases. Neutron presence when using 15 MV, represents an increase in peripheral dose of (18 ± 17) % in average. Due to the higher number of MU used in Inverse IMRT, an increasing of (22 ± 3) % in neutron dose is found related to Forward and 3D-CRT plans. This corresponds to photon doses within 44 and 255 mSv along the organs, for a dose prescription of 68 Gy at the isocenter. Conclusion: Neutron and photon peripheral doses for a prostate treatment planified in 6 different techniques have been analyzed. 6 MV plans are slightly more demanding in terms of photon peripheral doses. Inverse technique in 15 MV has Result to be the most demanding one in terms of total peripheral doses, including neutrons and photons.

  20. Risk of Developing Second Cancer From Neutron Dose in Proton Therapy as Function of Field Characteristics, Organ, and Patient Age

    International Nuclear Information System (INIS)

    Zacharatou Jarlskog, Christina; Paganetti, Harald

    2008-01-01

    Purpose: To estimate the risk of a second malignancy after treatment of a primary brain cancer using passive scattered proton beam therapy. The focus was on the cancer risk caused by neutrons outside the treatment volume and the dependency on the patient's age. Methods and Materials: Organ-specific neutron-equivalent doses previously calculated for eight different proton therapy brain fields were considered. Organ-specific models were applied to assess the risk of developing solid cancers and leukemia. Results: The main contributors (>80%) to the neutron-induced risk are neutrons generated in the treatment head. Treatment volume can influence the risk by up to a factor of ∼2. Young patients are subject to significantly greater risks than are adult patients because of the geometric differences and age dependency of the risk models. Breast cancer should be the main concern for females. For males, the risks of lung cancer, leukemia, and thyroid cancer were significant for pediatric patients. In contrast, leukemia was the leading risk for an adult. Most lifetime risks were <1% (70-Gy treatment). The only exceptions were breast, thyroid, and lung cancer for females. For female thyroid cancer, the treatment risk can exceed the baseline risk. Conclusion: The risk of developing a second malignancy from neutrons from proton beam therapy of a brain lesion is small (i.e., presumably outweighed by the therapeutic benefit) but not negligible (i.e., potentially greater than the baseline risk). The patient's age at treatment plays a major role

  1. Subgroup approximation in Monte Carlo neutron transport calculation in resonance region; Primena podgrupe aproksimacije u Monte Carlo proracunima transporta neutrona u rezonantnoj oblasti energije

    Energy Technology Data Exchange (ETDEWEB)

    Belicev, P [Vojnotehnicki Inst., Belgrade (Yugoslavia)

    1988-07-01

    An outline of the problems encountered in the multigroup calculations of the neutron transport in the resonance region is given. The difference between subgroup and multigroup approximation is described briefly. The features of the Monte Carlo code SUBGR are presented. The results of the calculations of the neutron transmission and albedo for infinite iron slabs are given. (author)

  2. Development of a phoswich detector for neutron dose rate measurements in the Earth's atmosphere

    Energy Technology Data Exchange (ETDEWEB)

    Doensdorf, Esther Miriam

    2014-04-30

    The Earth is constantly exposed to a stream of energetic particles from outer space. Through the interaction of this radiation with the Earth's magnetosphere and atmosphere a complex radiation field is formed which varies with the location inside the Earth's atmosphere. This radiation field consists of charged and uncharged particles leading to the constant exposure of human beings to radiation. As this ionizing radiation can be harmful for humans, it is necessary to perform dose rate measurements in different altitudes in the Earth's atmosphere. Due to their higher biological effectiveness the exposure to neutrons is more harmful than the exposure to γ-rays and charged particles, which is why the determination of neutron dose rates is the focus of this work. In this work the prototype of a Phoswich detector called PING (Phoswich Instrument for Neutrons and Gammas) is developed to determine dose rates caused by neutrons in the Earth's atmosphere and to distinguish these from γ-rays. The instrument is composed of two different scintillators optically coupled to each other and read out by one common photomultiplier tube. The scintillator package consists of an inner plastic scintillator made of the material BC-412 and a surrounding anti-coincidence made of sodium doped caesium iodide (CsI(Na)). In this work the instrument is calibrated, tested and flown and a procedure for a pulse shape analysis for this instrument is developed. With this analysis it is possible to distinguish pulses from the plastic scintillator and pulses from the CsI(Na). The pulses from the plastic scintillator are mainly due to the interaction of neutrons but there is an energy-dependent contribution of γ-rays to these events. Measurements performed on board an airplane show that the dose rates measured with the developed detector are in the same order of magnitude as results of other instruments. During measurements on board stratospheric balloons the altitude dependence

  3. Calibration of a TLD system to estimate personal doses in fields of gamma-neutrons radiation

    International Nuclear Information System (INIS)

    Villegas, E.N.; Somarriba, S.I.

    2016-01-01

    Currently Nicaragua has no personal neutron dosimetry system. The calibration of a batch of albedo neutron dosimeters consisting of two pairs of "6LiF and "7LiF (Mg, Ti) detectors was done. The dosimeter and reader sensitivities were obtained using a "1"3"7Cs source, and a neutron calibration factor was found with a "2"4"1AmBe source. Reproducibility and homogeneity tests were performed, and the detection limit of the system was determined. This calibration will allow the beginning of neutron personal monitoring in the country. (author)

  4. A multigroup treatment of radiation transport

    International Nuclear Information System (INIS)

    Tahir, N.A.; Laing, E.W.; Nicholas, D.J.

    1980-12-01

    A multi-group radiation package is outlined which will accurately handle radiation transfer problems in laser-produced plasmas. Bremsstrahlung, recombination and line radiation are included as well as fast electron Bremsstrahlung radiation. The entire radiation field is divided into a large number of groups (typically 20), which diffuse radiation energy in real space as well as in energy space, the latter occurring via electron-radiation i