WorldWideScience

Sample records for carlo adjoint shielding

  1. Automated variance reduction of Monte Carlo shielding calculations using the discrete ordinates adjoint function

    Wagner, J.C.; Haghighat, A.

    1998-01-01

    Although the Monte Carlo method is considered to be the most accurate method available for solving radiation transport problems, its applicability is limited by its computational expense. Thus, biasing techniques, which require intuition, guesswork, and iterations involving manual adjustments, are employed to make reactor shielding calculations feasible. To overcome this difficulty, the authors have developed a method for using the S N adjoint function for automated variance reduction of Monte Carlo calculations through source biasing and consistent transport biasing with the weight window technique. They describe the implementation of this method into the standard production Monte Carlo code MCNP and its application to a realistic calculation, namely, the reactor cavity dosimetry calculation. The computational effectiveness of the method, as demonstrated through the increase in calculational efficiency, is demonstrated and quantified. Important issues associated with this method and its efficient use are addressed and analyzed. Additional benefits in terms of the reduction in time and effort required of the user are difficult to quantify but are possibly as important as the computational efficiency. In general, the automated variance reduction method presented is capable of increases in computational performance on the order of thousands, while at the same time significantly reducing the current requirements for user experience, time, and effort. Therefore, this method can substantially increase the applicability and reliability of Monte Carlo for large, real-world shielding applications

  2. Adjoint electron Monte Carlo calculations

    Jordan, T.M.

    1986-01-01

    Adjoint Monte Carlo is the most efficient method for accurate analysis of space systems exposed to natural and artificially enhanced electron environments. Recent adjoint calculations for isotropic electron environments include: comparative data for experimental measurements on electronics boxes; benchmark problem solutions for comparing total dose prediction methodologies; preliminary assessment of sectoring methods used during space system design; and total dose predictions on an electronics package. Adjoint Monte Carlo, forward Monte Carlo, and experiment are in excellent agreement for electron sources that simulate space environments. For electron space environments, adjoint Monte Carlo is clearly superior to forward Monte Carlo, requiring one to two orders of magnitude less computer time for relatively simple geometries. The solid-angle sectoring approximations used for routine design calculations can err by more than a factor of 2 on dose in simple shield geometries. For critical space systems exposed to severe electron environments, these potential sectoring errors demand the establishment of large design margins and/or verification of shield design by adjoint Monte Carlo/experiment

  3. A User's Manual for MASH V1.5 - A Monte Carlo Adjoint Shielding Code System

    C. O. Slater; J. M. Barnes; J. O. Johnson; J.D. Drischler

    1998-10-01

    The Monte Carlo ~djoint ~ielding Code System, MASH, calculates neutron and gamma- ray environments and radiation protection factors for armored military vehicles, structures, trenches, and other shielding configurations by coupling a forward discrete ordinates air- over-ground transport calculation with an adjoint Monte Carlo treatment of the shielding geometry. Efficiency and optimum use of computer time are emphasized. The code system includes the GRTUNCL and DORT codes for air-over-ground transport calculations, the MORSE code with the GIFT5 combinatorial geometry package for adjoint shielding calculations, and several peripheral codes that perform the required data preparations, transformations, and coupling functions. The current version, MASH v 1.5, is the successor to the original MASH v 1.0 code system initially developed at Oak Ridge National Laboratory (ORNL). The discrete ordinates calculation determines the fluence on a coupling surface surrounding the shielding geometry due to an external neutron/gamma-ray source. The Monte Carlo calculation determines the effectiveness of the fluence at that surface in causing a response in a detector within the shielding geometry, i.e., the "dose importance" of the coupling surface fluence. A coupling code folds the fluence together with the dose importance, giving the desired dose response. The coupling code can determine the dose response as a function of the shielding geometry orientation relative to the source, distance from the source, and energy response of the detector. This user's manual includes a short description of each code, the input required to execute the code along with some helpful input data notes, and a representative sample problem.

  4. Monte Carlo shielding analyses using an automated biasing procedure

    Tang, J.S.; Hoffman, T.J.

    1988-01-01

    A systematic and automated approach for biasing Monte Carlo shielding calculations is described. In particular, adjoint fluxes from a one-dimensional discrete ordinates calculation are used to generate biasing parameters for a Monte Carlo calculation. The entire procedure of adjoint calculation, biasing parameters generation, and Monte Carlo calculation has been automated. The automated biasing procedure has been applied to several realistic deep-penetration shipping cask problems. The results obtained for neutron and gamma-ray transport indicate that with the automated biasing procedure Monte Carlo shielding calculations of spent-fuel casks can be easily performed with minimum effort and that accurate results can be obtained at reasonable computing cost

  5. Adjoint acceleration of Monte Carlo simulations using TORT/MCNP coupling approach: A case study on the shielding improvement for the cyclotron room of the Buddhist Tzu Chi General Hospital

    Sheu, R. J.; Sheu, R. D.; Jiang, S. H.; Kao, C. H.

    2005-01-01

    Full-scale Monte Carlo simulations of the cyclotron room of the Buddhist Tzu Chi General Hospital were carried out to improve the original inadequate maze design. Variance reduction techniques are indispensable in this study to facilitate the simulations for testing a variety of configurations of shielding modification. The TORT/MCNP manual coupling approach based on the Consistent Adjoint Driven Importance Sampling (CADIS) methodology has been used throughout this study. The CADIS utilises the source and transport biasing in a consistent manner. With this method, the computational efficiency was increased significantly by more than two orders of magnitude and the statistical convergence was also improved compared to the unbiased Monte Carlo run. This paper describes the shielding problem encountered, the procedure for coupling the TORT and MCNP codes to accelerate the calculations and the calculation results for the original and improved shielding designs. In order to verify the calculation results and seek additional accelerations, sensitivity studies on the space-dependent and energy-dependent parameters were also conducted. (authors)

  6. New Monte Carlo approach to the adjoint Boltzmann equation

    De Matteis, A.; Simonini, R.

    1978-01-01

    A class of stochastic models for the Monte Carlo integration of the adjoint neutron transport equation is described. Some current general methods are brought within this class, thus preparing the ground for subsequent comparisons. Monte Carlo integration of the adjoint Boltzmann equation can be seen as a simulation of the transport of mathematical particles with reaction kernels not normalized to unity. This last feature is a source of difficulty: It can influence the variance of the result negatively and also often leads to preparation of special ''libraries'' consisting of tables of normalization factors as functions of energy, presently used by several methods. These are the two main points that are discussed and that are taken into account to devise a nonmultigroup method of solution for a certain class of problems. Reactions considered in detail are radiative capture, elastic scattering, discrete levels and continuum inelastic scattering, for which the need for tables has been almost completely eliminated. The basic policy pursued to avoid a source of statistical fluctuations is to try to make the statistical weight of the traveling particle dependent only on its starting and current energies, at least in simple cases. The effectiveness of the sampling schemes proposed is supported by numerical comparison with other more general adjoint Monte Carlo methods. Computation of neutron flux at a point by means of an adjoint formulation is the problem taken as a test for numerical experiments. Very good results have been obtained in the difficult case of resonant cross sections

  7. Practical adjoint Monte Carlo technique for fixed-source and eigenfunction neutron transport problems

    Hoogenboom, J.E.

    1981-01-01

    An adjoint Monte Carlo technique is described for the solution of neutron transport problems. The optimum biasing function for a zero-variance collision estimator is derived. The optimum treatment of an analog of a non-velocity thermal group has also been derived. The method is extended to multiplying systems, especially for eigenfunction problems to enable the estimate of averages over the unknown fundamental neutron flux distribution. A versatile computer code, FOCUS, has been written, based on the described theory. Numerical examples are given for a shielding problem and a critical assembly, illustrating the performance of the FOCUS code. 19 refs

  8. Lecture 1. Monte Carlo basics. Lecture 2. Adjoint Monte Carlo. Lecture 3. Coupled Forward-Adjoint calculations

    Hoogenboom, J.E. [Delft University of Technology, Interfaculty Reactor Institute, Delft (Netherlands)

    2000-07-01

    The Monte Carlo method is a statistical method to solve mathematical and physical problems using random numbers. The principle of the methods will be demonstrated for a simple mathematical problem and for neutron transport. Various types of estimators will be discussed, as well as generally applied variance reduction methods like splitting, Russian roulette and importance biasing. The theoretical formulation for solving eigenvalue problems for multiplying systems will be shown. Some reflections will be given about the applicability of the Monte Carlo method, its limitations and its future prospects for reactor physics calculations. Adjoint Monte Carlo is a Monte Carlo game to solve the adjoint neutron (or photon) transport equation. The adjoint transport equation can be interpreted in terms of simulating histories of artificial particles, which show properties of neutrons that move backwards in history. These particles will start their history at the detector from which the response must be estimated and give a contribution to the estimated quantity when they hit or pass through the neutron source. Application to multigroup transport formulation will be demonstrated Possible implementation for the continuous energy case will be outlined. The inherent advantages and disadvantages of the method will be discussed. The Midway Monte Carlo method will be presented for calculating a detector response due to a (neutron or photon) source. A derivation will be given of the basic formula for the Midway Monte Carlo method The black absorber technique, allowing for a cutoff of particle histories when reaching the midway surface in one of the calculations will be derived. An extension of the theory to coupled neutron-photon problems is given. The method will be demonstrated for an oil well logging problem, comprising a neutron source in a borehole and photon detectors to register the photons generated by inelastic neutron scattering. (author)

  9. Lecture 1. Monte Carlo basics. Lecture 2. Adjoint Monte Carlo. Lecture 3. Coupled Forward-Adjoint calculations

    Hoogenboom, J.E.

    2000-01-01

    The Monte Carlo method is a statistical method to solve mathematical and physical problems using random numbers. The principle of the methods will be demonstrated for a simple mathematical problem and for neutron transport. Various types of estimators will be discussed, as well as generally applied variance reduction methods like splitting, Russian roulette and importance biasing. The theoretical formulation for solving eigenvalue problems for multiplying systems will be shown. Some reflections will be given about the applicability of the Monte Carlo method, its limitations and its future prospects for reactor physics calculations. Adjoint Monte Carlo is a Monte Carlo game to solve the adjoint neutron (or photon) transport equation. The adjoint transport equation can be interpreted in terms of simulating histories of artificial particles, which show properties of neutrons that move backwards in history. These particles will start their history at the detector from which the response must be estimated and give a contribution to the estimated quantity when they hit or pass through the neutron source. Application to multigroup transport formulation will be demonstrated Possible implementation for the continuous energy case will be outlined. The inherent advantages and disadvantages of the method will be discussed. The Midway Monte Carlo method will be presented for calculating a detector response due to a (neutron or photon) source. A derivation will be given of the basic formula for the Midway Monte Carlo method The black absorber technique, allowing for a cutoff of particle histories when reaching the midway surface in one of the calculations will be derived. An extension of the theory to coupled neutron-photon problems is given. The method will be demonstrated for an oil well logging problem, comprising a neutron source in a borehole and photon detectors to register the photons generated by inelastic neutron scattering. (author)

  10. Continuous energy adjoint Monte Carlo for coupled neutron-photon transport

    Hoogenboom, J.E. [Delft Univ. of Technology (Netherlands). Interfaculty Reactor Inst.

    2001-07-01

    Although the theory for adjoint Monte Carlo calculations with continuous energy treatment for neutrons as well as for photons is known, coupled neutron-photon transport problems present fundamental difficulties because of the discrete energies of the photons produced by neutron reactions. This problem was solved by forcing the energy of the adjoint photon to the required discrete value by an adjoint Compton scattering reaction or an adjoint pair production reaction. A mathematical derivation shows the exact procedures to follow for the generation of an adjoint neutron and its statistical weight. A numerical example demonstrates that correct detector responses are obtained compared to a standard forward Monte Carlo calculation. (orig.)

  11. FOCUS: a non-multigroup adjoint Monte Carlo code with improved variance reduction

    Hoogenboom, J.E.

    1974-01-01

    A description is given of the selection mechanism in the adjoint Monte Carlo code FOCUS in which the energy is treated as a continuous variable. The method of Kalos who introduced the idea of adjoint cross sections is followed to derive a sampling scheme for the adjoint equation solved in FOCUS which is in most aspects analogous to the normal Monte Carlo game. The disadvantages of the use of these adjoint cross sections are removed to some extent by introduction of a new definition for the adjoint cross sections resulting in appreciable variance reduction. At the cost of introducing a weight factor slightly different from unity, the direction and energy are selected in a simple way without the need of two-dimensional probability tables. Finally the handling of geometry and cross section in FOCUS is briefly discussed. 6 references. (U.S.)

  12. Memory-efficient calculations of adjoint-weighted tallies by the Monte Carlo Wielandt method

    Choi, Sung Hoon; Shim, Hyung Jin

    2016-01-01

    Highlights: • The MC Wielandt method is applied to reduce memory for the adjoint estimation. • The adjoint-weighted kinetics parameters are estimated in the MC Wielandt calculations. • The MC S/U analyses are conducted in the MC Wielandt calculations. - Abstract: The current Monte Carlo (MC) adjoint-weighted tally techniques based on the iterated fission probability (IFP) concept require a memory amount which is proportional to the numbers of the adjoint-weighted tallies and histories per cycle to store history-wise tally estimates during the convergence of the adjoint flux. Especially the conventional MC adjoint-weighted perturbation (AWP) calculations for the nuclear data sensitivity and uncertainty (S/U) analysis suffer from the huge memory consumption to realize the IFP concept. In order to reduce the memory requirement drastically, we present a new adjoint estimation method in which the memory usage is irrelevant to the numbers of histories per cycle by applying the IFP concept for the MC Wielandt calculations. The new algorithms for the adjoint-weighted kinetics parameter estimation and the AWP calculations in the MC Wielandt method are implemented in a Seoul National University MC code, McCARD and its validity is demonstrated in critical facility problems. From the comparison of the nuclear data S/U analyses, it is demonstrated that the memory amounts to store the sensitivity estimates in the proposed method become negligibly small.

  13. Exploring the use of a deterministic adjoint flux calculation in criticality Monte Carlo simulations

    Jinaphanh, A.; Miss, J.; Richet, Y.; Martin, N.; Hebert, A.

    2011-01-01

    The paper presents a preliminary study on the use of a deterministic adjoint flux calculation to improve source convergence issues by reducing the number of iterations needed to reach the converged distribution in criticality Monte Carlo calculations. Slow source convergence in Monte Carlo eigenvalue calculations may lead to underestimate the effective multiplication factor or reaction rates. The convergence speed depends on the initial distribution and the dominance ratio. We propose using an adjoint flux estimation to modify the transition kernel according to the Importance Sampling technique. This adjoint flux is also used as the initial guess of the first generation distribution for the Monte Carlo simulation. Calculated Variance of a local estimator of current is being checked. (author)

  14. Estimation of Adjoint-Weighted Kinetics Parameters in Monte Carlo Wieland Calculations

    Choi, Sung Hoon; Shim, Hyung Jin

    2013-01-01

    The effective delayed neutron fraction, β eff , and the prompt neutron generation time, Λ, in the point kinetics equation are weighted by the adjoint flux to improve the accuracy of the reactivity estimate. Recently the Monte Carlo (MC) kinetics parameter estimation methods by using the self-consistent adjoint flux calculated in the MC forward simulations have been developed and successfully applied for the research reactor analyses. However these adjoint estimation methods based on the cycle-by-cycle genealogical table require a huge memory size to store the pedigree hierarchy. In this paper, we present a new adjoint estimation in which the pedigree of a single history is utilized by applying the MC Wielandt method. The effectiveness of the new method is demonstrated in the kinetics parameter estimations for infinite homogeneous two-group problems and the Godiva critical facility

  15. An Analysis on the Calculation Efficiency of the Responses Caused by the Biased Adjoint Fluxes in Hybrid Monte Carlo Simulation

    Khuat, Quang Huy; Kim, Song Hyun; Kim, Do Hyun; Shin, Chang Ho

    2015-01-01

    This technique is known as Consistent Adjoint Driven Importance Sampling (CADIS) method and it is implemented in SCALE code system. In the CADIS method, adjoint transport equation has to be solved to determine deterministic importance functions. Using the CADIS method, a problem was noted that the biased adjoint flux estimated by deterministic methods can affect the calculation efficiency and error. The biases of adjoint function are caused by the methodology, calculation strategy, tolerance of result calculated by the deterministic method and inaccurate multi-group cross section libraries. In this paper, a study to analyze the influence of the biased adjoint functions into Monte Carlo computational efficiency is pursued. In this study, a method to estimate the calculation efficiency was proposed for applying the biased adjoint fluxes in the CADIS approach. For a benchmark problem, the responses and FOMs using SCALE code system were evaluated as applying the adjoint fluxes. The results show that the biased adjoint fluxes significantly affects the calculation efficiencies

  16. Monte Carlo methods for shield design calculations

    Grimstone, M.J.

    1974-01-01

    A suite of Monte Carlo codes is being developed for use on a routine basis in commercial reactor shield design. The methods adopted for this purpose include the modular construction of codes, simplified geometries, automatic variance reduction techniques, continuous energy treatment of cross section data, and albedo methods for streaming. Descriptions are given of the implementation of these methods and of their use in practical calculations. 26 references. (U.S.)

  17. Estimation of ex-core detector responses by adjoint Monte Carlo

    Hoogenboom, J. E. [Delft Univ. of Technology, Mekelweg 15, 2629 JB Delft (Netherlands)

    2006-07-01

    Ex-core detector responses can be efficiently calculated by combining an adjoint Monte Carlo calculation with the converged source distribution of a forward Monte Carlo calculation. As the fission source distribution from a Monte Carlo calculation is given only as a collection of discrete space positions, the coupling requires a point flux estimator for each collision in the adjoint calculation. To avoid the infinite variance problems of the point flux estimator, a next-event finite-variance point flux estimator has been applied, witch is an energy dependent form for heterogeneous media of a finite-variance estimator known from the literature. To test the effects of this combined adjoint-forward calculation a simple geometry of a homogeneous core with a reflector was adopted with a small detector in the reflector. To demonstrate the potential of the method the continuous-energy adjoint Monte Carlo technique with anisotropic scattering was implemented with energy dependent absorption and fission cross sections and constant scattering cross section. A gain in efficiency over a completely forward calculation of the detector response was obtained, which is strongly dependent on the specific system and especially the size and position of the ex-core detector and the energy range considered. Further improvements are possible. The method works without problems for small detectors, even for a point detector and a small or even zero energy range. (authors)

  18. Methodology of Continuous-Energy Adjoint Monte Carlo for Neutron, Photon, and Coupled Neutron-Photon Transport

    Hoogenboom, J. Eduard

    2003-01-01

    Adjoint Monte Carlo may be a useful alternative to regular Monte Carlo calculations in cases where a small detector inhibits an efficient Monte Carlo calculation as only very few particle histories will cross the detector. However, in general purpose Monte Carlo codes, normally only the multigroup form of adjoint Monte Carlo is implemented. In this article the general methodology for continuous-energy adjoint Monte Carlo neutron transport is reviewed and extended for photon and coupled neutron-photon transport. In the latter cases the discrete photons generated by annihilation or by neutron capture or inelastic scattering prevent a direct application of the general methodology. Two successive reaction events must be combined in the selection process to accommodate the adjoint analog of a reaction resulting in a photon with a discrete energy. Numerical examples illustrate the application of the theory for some simplified problems

  19. Variational variance reduction for particle transport eigenvalue calculations using Monte Carlo adjoint simulation

    Densmore, Jeffery D.; Larsen, Edward W.

    2003-01-01

    The Variational Variance Reduction (VVR) method is an effective technique for increasing the efficiency of Monte Carlo simulations [Ann. Nucl. Energy 28 (2001) 457; Nucl. Sci. Eng., in press]. This method uses a variational functional, which employs first-order estimates of forward and adjoint fluxes, to yield a second-order estimate of a desired system characteristic - which, in this paper, is the criticality eigenvalue k. If Monte Carlo estimates of the forward and adjoint fluxes are used, each having global 'first-order' errors of O(1/√N), where N is the number of histories used in the Monte Carlo simulation, then the statistical error in the VVR estimation of k will in principle be O(1/N). In this paper, we develop this theoretical possibility and demonstrate with numerical examples that implementations of the VVR method for criticality problems can approximate O(1/N) convergence for significantly large values of N

  20. A study on the shielding element using Monte Carlo simulation

    Kim, Ki Jeong [Dept. of Radiology, Konkuk University Medical Center, Seoul (Korea, Republic of); Shim, Jae Goo [Dept. of Radiologic Technology, Daegu Health College, Daegu (Korea, Republic of)

    2017-06-15

    In this research, we simulated the elementary star shielding ability using Monte Carlo simulation to apply medical radiation shielding sheet which can replace existing lead. In the selection of elements, mainly elements and metal elements having a large atomic number, which are known to have high shielding performance, recently, various composite materials have improved shielding performance, so that weight reduction, processability, In consideration of activity etc., 21 elements were selected. The simulation tools were utilized Monte Carlo method. As a result of simulating the shielding performance by each element, it was estimated that the shielding ratio is the highest at 98.82% and 98.44% for tungsten and gold.

  1. The Adjoint Monte Carlo - a viable option for efficient radiotherapy treatment planning

    Goldstein, M [Israel Atomic Energy Commission, Beersheba (Israel). Nuclear Research Center-Negev

    1996-12-01

    In cancer therapy using collimated beams of photons, the radiation oncologist must determine a set of beams that delivers the required dose to each point in the tumor and minimizes the risk of damage to the healthy tissue and vital organs. Currently, the oncologist determines these beams iteratively, by using a sequence of dose calculations using approximate numerical methods. In this paper, a more accurate and potentially faster approach, based on the Adjoint Monte Carlo method, is presented (authors).

  2. FOCUS, Neutron Transport System for Complex Geometry Reactor Core and Shielding Problems by Monte-Carlo

    Hoogenboom, J.E.

    1980-01-01

    1 - Description of problem or function: FOCUS enables the calculation of any quantity related to neutron transport in reactor or shielding problems, but was especially designed to calculate differential quantities, such as point values at one or more of the space, energy, direction and time variables of quantities like neutron flux, detector response, reaction rate, etc. or averages of such quantities over a small volume of the phase space. Different types of problems can be treated: systems with a fixed neutron source which may be a mono-directional source located out- side the system, and Eigen function problems in which the neutron source distribution is given by the (unknown) fundamental mode Eigen function distribution. Using Monte Carlo methods complex 3- dimensional geometries and detailed cross section information can be treated. Cross section data are derived from ENDF/B, with anisotropic scattering and discrete or continuous inelastic scattering taken into account. Energy is treated as a continuous variable and time dependence may also be included. 2 - Method of solution: A transformed form of the adjoint Boltzmann equation in integral representation is solved for the space, energy, direction and time variables by Monte Carlo methods. Adjoint particles are defined with properties in some respects contrary to those of neutrons. Adjoint particle histories are constructed from which estimates are obtained of the desired quantity. Adjoint cross sections are defined with which the nuclide and reaction type are selected in a collision. The energy after a collision is selected from adjoint energy distributions calculated together with the adjoint cross sections in advance of the actual Monte Carlo calculation. For multiplying systems successive generations of adjoint particles are obtained which will die out for subcritical systems with a fixed neutron source and will be kept approximately stationary for Eigen function problems. Completely arbitrary problems can

  3. A midway forward-adjoint coupling method for neutron and photon Monte Carlo transport

    Serov, I.V.; John, T.M.; Hoogenboom, J.E.

    1999-01-01

    The midway Monte Carlo method for calculating detector responses combines a forward and an adjoint Monte Carlo calculation. In both calculations, particle scores are registered at a surface to be chosen by the user somewhere between the source and detector domains. The theory of the midway response determination is developed within the framework of transport theory for external sources and for criticality theory. The theory is also developed for photons, which are generated at inelastic scattering or capture of neutrons. In either the forward or the adjoint calculation a so-called black absorber technique can be applied; i.e., particles need not be followed after passing the midway surface. The midway Monte Carlo method is implemented in the general-purpose MCNP Monte Carlo code. The midway Monte Carlo method is demonstrated to be very efficient in problems with deep penetration, small source and detector domains, and complicated streaming paths. All the problems considered pose difficult variance reduction challenges. Calculations were performed using existing variance reduction methods of normal MCNP runs and using the midway method. The performed comparative analyses show that the midway method appears to be much more efficient than the standard techniques in an overwhelming majority of cases and can be recommended for use in many difficult variance reduction problems of neutral particle transport

  4. Performance of the improved version of Monte Carlo Code A3MCNP for cask shielding design

    Hasegawa, T.; Ueki, K.; Sato, O.; Sjoden, G.E.; Miyake, Y.; Ohmura, M.; Haghighat, A.

    2004-01-01

    A 3 MCNP (Automatic Adjoint Accelerated MCNP) is a revised version of the MCNP Monte Carlo code, that automatically prepares variance reduction parameters for the CADIS (Consistent Adjoint Driven Importance Sampling) methodology. Using a deterministic ''importance'' (or adjoint) function, CADIS performs source and transport biasing within the weight-window technique. The current version of A 3 MCNP uses the 3-D Sn transport TORT code to determine a 3-D importance function distribution. Based on simulation of several real-life problems, it is demonstrated that A3MCNP provides precise calculation results with a remarkably short computation time by using the proper and objective variance reduction parameters. However, since the first version of A 3 MCNP provided only a point source configuration option for large-scale shielding problems, such as spent-fuel transport casks, a large amount of memory may be necessary to store enough points to properly represent the source. Hence, we have developed an improved version of A 3 MCNP (referred to as A 3 MCNPV) which has a volumetric source configuration option. This paper describes the successful use of A 3 MCNPV for cask neutron and gamma-ray shielding problem

  5. Improvement of Monte Carlo code A3MCNP for large-scale shielding problems

    Miyake, Y.; Ohmura, M.; Hasegawa, T.; Ueki, K.; Sato, O.; Haghighat, A.; Sjoden, G.E.

    2004-01-01

    A 3 MCNP (Automatic Adjoint Accelerated MCNP) is a revised version of the MCNP Monte Carlo code, that automatically prepares variance reduction parameters for the CADIS (Consistent Adjoint Driven Importance Sampling) methodology. Using a deterministic 'importance' (or adjoint) function, CADIS performs source and transport biasing within the weight-window technique. The current version of A 3 MCNP uses the 3-D Sn transport TORT code to determine a 3-D importance function distribution. Based on simulation of several real-life problems, it is demonstrated that A 3 MCNP provides precise calculation results with a remarkably short computation time by using the proper and objective variance reduction parameters. However, since the first version of A 3 MCNP provided only a point source configuration option for large-scale shielding problems, such as spent-fuel transport casks, a large amount of memory may be necessary to store enough points to properly represent the source. Hence, we have developed an improved version of A 3 MCNP (referred to as A 3 MCNPV) which has a volumetric source configuration option. This paper describes the successful use of A 3 MCNPV for a concrete cask streaming problem and a PWR dosimetry problem. (author)

  6. Monte Carlo validation of self shielding and void effect calculations

    Tellier, H.; Coste, M.; Raepsaet, C.; Soldevila, M.; Van der Gucht, C.

    1995-01-01

    The self shielding validation and the void effect are studied with Monte Carlo method. The satisfactory comparison obtained between the APOLLO 2 results of the self shielding effect and the TRIPOLI and MCNP results allows us to be confident in the multigroup transport code. (K.A.)

  7. Problems in radiation shielding calculations with Monte Carlo methods

    Ueki, Kohtaro

    1985-01-01

    The Monte Carlo method is a very useful tool for solving a large class of radiation transport problem. In contrast with deterministic method, geometric complexity is a much less significant problem for Monte Carlo calculations. However, the accuracy of Monte Carlo calculations is of course, limited by statistical error of the quantities to be estimated. In this report, we point out some typical problems to solve a large shielding system including radiation streaming. The Monte Carlo coupling technique was developed to settle such a shielding problem accurately. However, the variance of the Monte Carlo results using the coupling technique of which detectors were located outside the radiation streaming, was still not enough. So as to bring on more accurate results for the detectors located outside the streaming and also for a multi-legged-duct streaming problem, a practicable way of ''Prism Scattering technique'' is proposed in the study. (author)

  8. Comparison of deterministic and Monte Carlo methods in shielding design.

    Oliveira, A D; Oliveira, C

    2005-01-01

    In shielding calculation, deterministic methods have some advantages and also some disadvantages relative to other kind of codes, such as Monte Carlo. The main advantage is the short computer time needed to find solutions while the disadvantages are related to the often-used build-up factor that is extrapolated from high to low energies or with unknown geometrical conditions, which can lead to significant errors in shielding results. The aim of this work is to investigate how good are some deterministic methods to calculating low-energy shielding, using attenuation coefficients and build-up factor corrections. Commercial software MicroShield 5.05 has been used as the deterministic code while MCNP has been used as the Monte Carlo code. Point and cylindrical sources with slab shield have been defined allowing comparison between the capability of both Monte Carlo and deterministic methods in a day-by-day shielding calculation using sensitivity analysis of significant parameters, such as energy and geometrical conditions.

  9. Comparison of deterministic and Monte Carlo methods in shielding design

    Oliveira, A. D.; Oliveira, C.

    2005-01-01

    In shielding calculation, deterministic methods have some advantages and also some disadvantages relative to other kind of codes, such as Monte Carlo. The main advantage is the short computer time needed to find solutions while the disadvantages are related to the often-used build-up factor that is extrapolated from high to low energies or with unknown geometrical conditions, which can lead to significant errors in shielding results. The aim of this work is to investigate how good are some deterministic methods to calculating low-energy shielding, using attenuation coefficients and build-up factor corrections. Commercial software MicroShield 5.05 has been used as the deterministic code while MCNP has been used as the Monte Carlo code. Point and cylindrical sources with slab shield have been defined allowing comparison between the capability of both Monte Carlo and deterministic methods in a day-by-day shielding calculation using sensitivity analysis of significant parameters, such as energy and geometrical conditions. (authors)

  10. Monte Carlo based radial shield design of typical PWR reactor

    Gul, Anas; Khan, Rustam; Qureshi, M. Ayub; Azeem, Muhammad Waqar; Raza, S.A. [Pakistan Institute of Engineering and Applied Sciences, Islamabad (Pakistan). Dept. of Nuclear Engineering; Stummer, Thomas [Technische Univ. Wien (Austria). Atominst.

    2016-11-15

    Neutron and gamma flux and dose equivalent rate distribution are analysed in radial and shields of a typical PWR type reactor based on the Monte Carlo radiation transport computer code MCNP5. The ENDF/B-VI continuous energy cross-section library has been employed for the criticality and shielding analysis. The computed results are in good agreement with the reference results (maximum difference is less than 56 %). It implies that MCNP5 a good tool for accurate prediction of neutron and gamma flux and dose rates in radial shield around the core of PWR type reactors.

  11. Monte Carlo applications to radiation shielding problems

    Subbaiah, K.V.

    2009-01-01

    Monte Carlo methods are a class of computational algorithms that rely on repeated random sampling of physical and mathematical systems to compute their results. However, basic concepts of MC are both simple and straightforward and can be learned by using a personal computer. Uses of Monte Carlo methods require large amounts of random numbers, and it was their use that spurred the development of pseudorandom number generators, which were far quicker to use than the tables of random numbers which had been previously used for statistical sampling. In Monte Carlo simulation of radiation transport, the history (track) of a particle is viewed as a random sequence of free flights that end with an interaction event where the particle changes its direction of movement, loses energy and, occasionally, produces secondary particles. The Monte Carlo simulation of a given experimental arrangement (e.g., an electron beam, coming from an accelerator and impinging on a water phantom) consists of the numerical generation of random histories. To simulate these histories we need an interaction model, i.e., a set of differential cross sections (DCS) for the relevant interaction mechanisms. The DCSs determine the probability distribution functions (pdf) of the random variables that characterize a track; 1) free path between successive interaction events, 2) type of interaction taking place and 3) energy loss and angular deflection in a particular event (and initial state of emitted secondary particles, if any). Once these pdfs are known, random histories can be generated by using appropriate sampling methods. If the number of generated histories is large enough, quantitative information on the transport process may be obtained by simply averaging over the simulated histories. The Monte Carlo method yields the same information as the solution of the Boltzmann transport equation, with the same interaction model, but is easier to implement. In particular, the simulation of radiation

  12. Performance of the improved version of Monte Carlo code A 3MCNP for large-scale shielding problems

    Omura, M.; Miyake, Y.; Hasegawa, T.; Ueki, K.; Sato, O.; Haghighat, A.; Sjoden, G. E.

    2005-01-01

    A 3MCNP (Automatic Adjoint Accelerated MCNP) is a revised version of the MCNP Monte Carlo code, which automatically prepares variance reduction parameters for the CADIS (Consistent Adjoint Driven Importance Sampling) methodology. Using a deterministic 'importance' (or adjoint) function, CADIS performs source and transport biasing within the weight-window technique. The current version of A 3MCNP uses the three-dimensional (3-D) Sn transport TORT code to determine a 3-D importance function distribution. Based on simulation of several real-life problems, it is demonstrated that A 3MCNP provides precise calculation results with a remarkably short computation time by using the proper and objective variance reduction parameters. However, since the first version of A 3MCNP provided only a point source configuration option for large-scale shielding problems, such as spent-fuel transport casks, a large amount of memory may be necessary to store enough points to properly represent the source. Hence, we have developed an improved version of A 3MCNP (referred to as A 3MCNPV) which has a volumetric source configuration option. This paper describes the successful use of A 3MCNPV for a concrete cask neutron and gamma-ray shielding problem, and a PWR dosimetry problem. (authors)

  13. Application of adjoint Monte Carlo to accelerate simulations of mono-directional beams in treatment planning for Boron Neutron Capture Therapy

    Nievaart, V. A.; Legrady, D.; Moss, R. L.; Kloosterman, J. L.; Hagen, T. H. J. J. van der; Dam, H. van

    2007-01-01

    This paper deals with the application of the adjoint transport theory in order to optimize Monte Carlo based radiotherapy treatment planning. The technique is applied to Boron Neutron Capture Therapy where most often mixed beams of neutrons and gammas are involved. In normal forward Monte Carlo simulations the particles start at a source and lose energy as they travel towards the region of interest, i.e., the designated point of detection. Conversely, with adjoint Monte Carlo simulations, the so-called adjoint particles start at the region of interest and gain energy as they travel towards the source where they are detected. In this respect, the particles travel backwards and the real source and real detector become the adjoint detector and adjoint source, respectively. At the adjoint detector, an adjoint function is obtained with which numerically the same result, e.g., dose or flux in the tumor, can be derived as with forward Monte Carlo. In many cases, the adjoint method is more efficient and by that is much quicker when, for example, the response in the tumor or organ at risk for many locations and orientations of the treatment beam around the patient is required. However, a problem occurs when the treatment beam is mono-directional as the probability of detecting adjoint Monte Carlo particles traversing the beam exit (detector plane in adjoint mode) in the negative direction of the incident beam is zero. This problem is addressed here and solved first with the use of next event estimators and second with the application of a Legendre expansion technique of the angular adjoint function. In the first approach, adjoint particles are tracked deterministically through a tube to a (adjoint) point detector far away from the geometric model. The adjoint particles will traverse the disk shaped entrance of this tube (the beam exit in the actual geometry) perpendicularly. This method is slow whenever many events are involved that are not contributing to the point

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

    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)

  15. An investigation of the adjoint method for external beam radiation therapy treatment planning using Monte Carlo transport

    Kowalok, M.; Mackie, T.R.

    2001-01-01

    A relatively new technique for achieving the right dose to the right tissue, is intensity modulated radiation therapy (IMRT). In this technique, a megavoltage x-ray beam is rotated around a patient, and the intensity and shape of the beam is modulated as a function of source position and patient anatomy. The relationship between beam-let intensity and patient dose can be expressed under a matrix form where the matrix D ij represents the dose delivered to voxel i by beam-let j per unit fluence. The D ij influence matrix is the key element that enables this approach. In this regard, sensitivity theory lends itself in a natural way to the process of computing beam weights for treatment planning. The solution of the adjoint form of the Boltzmann equation is an adjoint function that describes the importance of particles throughout the system in contributing to the detector response. In this case, adjoint methods can provide the sensitivity of the dose at a single point in the patient with respect to all points in the source field. The purpose of this study is to investigate the feasibility of using the adjoint method and Monte Carlo transport for radiation therapy treatment planning

  16. Monte Carlo based radial shield design of typical PWR reactor

    Gul, Anas; Khan, Rustam; Qureshi, M. Ayub; Azeem, Muhammad Waqar; Raza, S.A. [Pakistan Institute of Engineering and Applied Sciences, Islamabad (Pakistan). Dept. of Nuclear Engineering; Stummer, Thomas [Technische Univ. Wien (Austria). Atominst.

    2017-04-15

    This paper presents the radiation shielding model of a typical PWR (CNPP-II) at Chashma, Pakistan. The model was developed using Monte Carlo N Particle code [2], equipped with ENDF/B-VI continuous energy cross section libraries. This model was applied to calculate the neutron and gamma flux and dose rates in the radial direction at core mid plane. The simulated results were compared with the reference results of Shanghai Nuclear Engineering Research and Design Institute (SNERDI).

  17. Applications to shielding design and others of monte carlo method

    Ito, Daiichiro [Mitsui Engineering and Shipbuiding Co., Ltd., Tokyo (Japan)

    2001-01-01

    One-dimensional or two-dimensional Sn computer code (ANISN, DOT3.5, etc.) and a point attenuation kernel integral code (QAD, etc.) have been used widely for shielding design. Application examples of monte carlo method which could follow precisely the three-dimensional configuration of shielding structure are shown as follow: (1) CASTER cask has a complex structure which consists of a large number of fuel baskets (stainless steel), neutron moderators (polyethylene rods), the body (cast iron), and cooling fin. The R-{theta} model of Sn code DOT3.5 cannot follow closely the complex form of polyethylene rods and fuel baskets. A monte carlo code MORSE is used to ascertain the calculation results of DOT3.5. The discrepancy between the calculation results of DOT3.5 and MORSE was in 10% for dose rate at distance of 1 m from the cask surface. (2) The dose rates of an iron cell at 10 cm above the floor are calculated by the code QAD and the MORSE. The reflected components of gamma ray caused by the auxiliary floor shield (lead) are analyzed by the MORSE. (3) A monte carlo code MCNP4A is used for skyshine evaluation of spent fuel carrier ship 'ROKUEIMARU'. The direct and skyshine components of gamma ray and neutron flux are estimated at each center of engine room and wheel house. The skyshine dose rate of neutron flux is 5-15 times larger than the gamma ray. (M. Suetake)

  18. Application of adjoint Monte Carlo to accelerate simulations of mono-directional beams in treatment planning for Boron Neutron Capture Therapy

    Nievaart, V.A.; Legrady, D.; Moss, R.L.; Kloosterman, J.L.; Van der Hagen, T.H.; Van Dam, H.

    2007-01-01

    This paper deals with the application of the adjoint transport theory in order to optimize Monte Carlo based radiotherapy treatment planning. The technique is applied to Boron Neutron Capture Therapy where most often mixed beams of neutrons and gammas are involved. In normal forward Monte Carlo

  19. Shielding evaluation of neutron generator hall by Monte Carlo simulations

    Pujala, U.; Selvakumaran, T.S.; Baskaran, R.; Venkatraman, B. [Radiological Safety Division, Indira Gandhi Center for Atomic Research, Kalpakkam (India); Thilagam, L.; Mohapatra, D.K., E-mail: swathythila2@yahoo.com [Safety Research Institute, Atomic Energy Regulatory Board, Kalpakkam (India)

    2017-04-01

    A shielded hall was constructed for accommodating a D-D, D-T or D-Be based pulsed neutron generator (NG) with 4π yield of 10{sup 9} n/s. The neutron shield design of the facility was optimized using NCRP-51 methodology such that the total dose rates outside the hall areas are well below the regulatory limit for full occupancy criterion (1 μSv/h). However, the total dose rates at roof top, cooling room trench exit and labyrinth exit were found to be above this limit for the optimized design. Hence, additional neutron shielding arrangements were proposed for cooling room trench and labyrinth exits. The roof top was made inaccessible. The present study is an attempt to evaluate the neutron and associated capture gamma transport through the bulk shields for the complete geometry and materials of the NG-Hall using Monte Carlo (MC) codes MCNP and FLUKA. The neutron source terms of D-D, D-T and D-Be reactions are considered in the simulations. The effect of additional shielding proposed has been demonstrated through the simulations carried out with the consideration of the additional shielding for D-Be neutron source term. The results MC simulations using two different codes are found to be consistent with each other for neutron dose rate estimates. However, deviation up to 28% is noted between these two codes at few locations for capture gamma dose rate estimates. Overall, the dose rates estimated by MC simulations including additional shields shows that all the locations surrounding the hall satisfy the full occupancy criteria for all three types of sources. Additionally, the dose rates due to direct transmission of primary neutrons estimated by FLUKA are compared with the values calculated using the formula given in NCRP-51 which shows deviations up to 50% with each other. The details of MC simulations and NCRP-51 methodology for the estimation of primary neutron dose rate along with the results are presented in this paper. (author)

  20. Monte Carlo analysis of the effects of a blanket-shield penetration on the performance of a tokamak fusion reactor

    Santoro, R.T.; Tang, J.S.; Alsmiller, R.G. Jr.; Barnes, J.M.

    1977-05-01

    Adjoint Monte Carlo calculations have been carried out using the three-dimensional radiation transport code, MORSE, to estimate the nuclear heating and radiation damage in the toroidal field (TF) coils adjacent to a 28 x 68 cm 2 rectangular neutral beam injector duct that passes through the blanket and shield of a D-T burning Tokamak reactor. The plasma region, blanket, shield, and TF coils were represented in cylindrical geometry using the same dimensions and compositions as those of the Experimental Power Reactor. The radiation transport was accomplished using coupled 35-group neutron, 21-group gamma-ray cross sections obtained by collapsing the DLC-37 cross-section library. Nuclear heating and radiation damage rates were estimated using the latest available nuclear response functions. The presence of the neutral beam injector duct leads to increases in the nuclear heating rates in the TF coils ranging from a factor of 3 to a factor of 196 depending on the location. Increases in the radiation damage also result in the TF coils. The atomic displacement rates increase from factors of 2 to 138 and the hydrogen and helium gas production rates increase from factors of 11 to 7600 and from 15 to 9700, respectively

  1. Continuous-energy adjoint flux and perturbation calculation using the iterated fission probability method in Monte-Carlo code TRIPOLI-4 and underlying applications

    Truchet, G.; Leconte, P.; Peneliau, Y.; Santamarina, A.

    2013-01-01

    The first goal of this paper is to present an exact method able to precisely evaluate very small reactivity effects with a Monte Carlo code (<10 pcm). it has been decided to implement the exact perturbation theory in TRIPOLI-4 and, consequently, to calculate a continuous-energy adjoint flux. The Iterated Fission Probability (IFP) method was chosen because it has shown great results in some other Monte Carlo codes. The IFP method uses a forward calculation to compute the adjoint flux, and consequently, it does not rely on complex code modifications but on the physical definition of the adjoint flux as a phase-space neutron importance. In the first part of this paper, the IFP method implemented in TRIPOLI-4 is described. To illustrate the efficiency of the method, several adjoint fluxes are calculated and compared with their equivalent obtained by the deterministic code APOLLO-2. The new implementation can calculate angular adjoint flux. In the second part, a procedure to carry out an exact perturbation calculation is described. A single cell benchmark has been used to test the accuracy of the method, compared with the 'direct' estimation of the perturbation. Once again the method based on the IFP shows good agreement for a calculation time far more inferior to the 'direct' method. The main advantage of the method is that the relative accuracy of the reactivity variation does not depend on the magnitude of the variation itself, which allows us to calculate very small reactivity perturbations with high precision. It offers the possibility to split reactivity contributions on both isotopes and reactions. Other applications of this perturbation method are presented and tested like the calculation of exact kinetic parameters (βeff, Λeff) or sensitivity parameters

  2. SHIELD-HIT12A - a Monte Carlo particle transport program for ion therapy research

    Bassler, Niels; Hansen, David Christoffer; Lühr, Armin

    2014-01-01

    . We experienced that new users quickly learn to use SHIELD-HIT12A and setup new geometries. Contrary to previous versions of SHIELD-HIT, the 12A distribution comes along with easy-to-use example files and an English manual. A new implementation of Vavilov straggling resulted in a massive reduction......Abstract. Purpose: The Monte Carlo (MC) code SHIELD-HIT simulates the transport of ions through matter. Since SHIELD-HIT08 we added numerous features that improves speed, usability and underlying physics and thereby the user experience. The “-A” fork of SHIELD-HIT also aims to attach SHIELD....... It supports native formats compatible with the heavy ion treatment planning system TRiP. Stopping power files follow ICRU standard and are generated using the libdEdx library, which allows the user to choose from a multitude of stopping power tables. Results: SHIELD-HIT12A runs on Linux and Windows platforms...

  3. Development and application of the automated Monte Carlo biasing procedure in SAS4

    Tang, J.S.; Broadhead, B.L.

    1995-01-01

    An automated approach for biasing Monte Carlo shielding calculations is described. In particular, adjoint fluxes from a one-dimensional discrete-ordinates calculation are used to generate biasing parameters for a three-dimensional Monte Carlo calculation. The automated procedure consisting of cross-section processing, adjoint flux determination, biasing parameter generation, and the initiation of a MORSE-SGC/S Monte Carlo calculation has been implemented in the SAS4 module of the SCALE computer code system. (author)

  4. An Analysis on the Characteristic of Multi-response CADIS Method for the Monte Carlo Radiation Shielding Calculation

    Kim, Do Hyun; Shin, Chang Ho; Kim, Song Hyun

    2014-01-01

    It uses the deterministic method to calculate adjoint fluxes for the decision of the parameters used in the variance reductions. This is called as hybrid Monte Carlo method. The CADIS method, however, has a limitation to reduce the stochastic errors of all responses. The Forward Weighted CADIS (FW-CADIS) was introduced to solve this problem. To reduce the overall stochastic errors of the responses, the forward flux is used. In the previous study, the Multi-Response CADIS (MR-CAIDS) method was derived for minimizing sum of each squared relative error. In this study, the characteristic of the MR-CADIS method was evaluated and compared with the FW-CADIS method. In this study, how the CADIS, FW-CADIS, and MR-CADIS methods are applied to optimize and decide the parameters used in the variance reduction techniques was analyzed. The MR-CADIS Method uses a technique that the sum of squared relative error in each tally region was minimized to achieve uniform uncertainty. To compare the simulation efficiency of the methods, a simple shielding problem was evaluated. Using FW-CADIS method, it was evaluated that the average of the relative errors was minimized; however, MR-CADIS method gives a lowest variance of the relative errors. Analysis shows that, MR-CADIS method can efficiently and uniformly reduce the relative error of the plural response problem than FW-CADIS method

  5. Adjoint Monte-Carlo method with fictitious scattering in deep penetration and long-distance detector calculations

    Andreucci, N.

    1985-04-01

    Deep penetration transport problems in complex systems joint to heterogeneous source (Q) sampling give rise to some difficulties in evaluating leakage and fluxes on a detector point. To overcome these difficulties we have solved both the adjoint Boltzmann flux (phi*) equation and following scalar-dual equation: ∫Qphi* dP - ∫Q*phi dP = ∫phiphi* Ω . n dΣ dΩ dE dt + ∫ [phiphi*]sub(0)sup(T)/v dr dΩ dE D = (phase space). With a suitable choice for the domain D, for Q* and for the boundary conditions, an adjoint flux calculation allows us to obtain simultaneously the Q-source contribution and the detection (or leakage) spectrum. Compared to direct methods with importance sampling, the adjoint methods give very low-cost and faithful results

  6. DEMONR, Monte-Carlo Shielding Calculation for Neutron Flux and Neutron Spectra, Teaching Program

    Courtney, J. C.

    1987-01-01

    1 - Description of problem or function: DEMONR treats the behavior of neutrons in a slab shield. It is frequently used as a teaching tool. 2 - Method of solution: An unbiased Monte Carlo code calculates the number, energy, and direction of neutrons that penetrate or are reflected from a shield. 3 - Restrictions on the complexity of the problem: Only one shield may be used in each problem. The shield material may be a single element or a homogeneous mixture of elements with a single effective atomic weight. Only elastic scattering and neutron capture processes are allowed. The source is a point located on one face of the slab. It provides a cosine distribution of current. Monoenergetic or fission spectrum neutrons may be selected

  7. Radiation shielding design for DECY-13 cyclotron using Monte Carlo method

    Rasito T; Bunawas; Taufik; Sunardi; Hari Suryanto

    2016-01-01

    DECY-13 is a 13 MeV proton cyclotron with target H_2"1"8O. The bombarding of 13 MeV protons on target H_2"1"8O produce large amounts of neutrons and gamma radiation. It needs the efficient radiation shielding to reduce the level of neutrons and gamma rays to ensure safety for workers and public. Modeling and calculations have been carried out using Monte Carlo method with MCNPX code to optimize the thickness for the radiation shielding. The calculations were done for radiation shielding of rectangular space room type with the size of 5.5 m x 5 m x 3 m and thickness of 170 cm made from lightweight concrete types of portland. It was shown that with this shielding the dose rate outside the wall was reduced to 1 μSv/h. (author)

  8. The biological shield of a high-intensity spallation source: a monte Carlo design study

    Koprivnikar, I.; Schachinger, E.

    2004-01-01

    The design of high-intensity spallation sources requires the best possible estimates for the biological shield. The applicability of three-dimensional Monte Carlo simulation in the design of the biological shield of a spallation source will be discussed. In order to achieve reasonable computing times along with acceptable accuracy, biasing techniques are to be employed and it was the main purpose of this work to develop a strategy for an effective Monte Carlo simulation in shielding design. The most prominent MC computer codes, namely MCNPX and FLUKA99, have been applied to the same model spallation source (the European Spallation Source) and on the basis of the derived strategies, the design and characteristics of the target station shield are discussed. It is also the purpose of the paper to demonstrate the application of the developed strategy for the design of beam lines with their shielding using as an example the target-moderator-reflector complex of the ESS as the primary particle source. (author)

  9. The elimination of ray tracing in Monte Carlo shielding programs

    Bendall, D.E.

    1988-01-01

    The MONK6 code has clearly demonstrated the advantages of hole tracking, which was devised by Woodcock et at. for use in criticality codes from earlier work by Von Neumann. Hole tracking eliminates ray tracing by introducing, for all materials present in the problem, a pseudo scattering reaction that forward scatters without energy loss. The cross section for this reaction is chosen so that the total cross sections for all the materials are equal at a given energy. By this means, tracking takes place with a constant total cross section everywhere, so there is now no need to ray trace. The present work extends hole tracking to shielding codes, where it functions in tandem with Russian roulette and splitting. An algorithm has been evolved and its performance is compared with the ray-tracking code McBEND. A disadvantage with hole tracking occurs when there is a wide variation in total cross section for materials present. As the tracking uses the total cross section of the material that has the maximum cross section, there can be a large number of pseudo collisions in the materials with low total cross sections. In extreme cases, the advantages of hole tracking can be lost by the by the extra time taken in servicing these pseudo collisions; however, techniques for eliminating this problem are under consideration

  10. Investigation of Radiation Protection Methodologies for Radiation Therapy Shielding Using Monte Carlo Simulation and Measurement

    Tanny, Sean

    The advent of high-energy linear accelerators for dedicated medical use in the 1950's by Henry Kaplan and the Stanford University physics department began a revolution in radiation oncology. Today, linear accelerators are the standard of care for modern radiation therapy and can generate high-energy beams that can produce tens of Gy per minute at isocenter. This creates a need for a large amount of shielding material to properly protect members of the public and hospital staff. Standardized vault designs and guidance on shielding properties of various materials are provided by the National Council on Radiation Protection (NCRP) Report 151. However, physicists are seeking ways to minimize the footprint and volume of shielding material needed which leads to the use of non-standard vault configurations and less-studied materials, such as high-density concrete. The University of Toledo Dana Cancer Center has utilized both of these methods to minimize the cost and spatial footprint of the requisite radiation shielding. To ensure a safe work environment, computer simulations were performed to verify the attenuation properties and shielding workloads produced by a variety of situations where standard recommendations and guidance documents were insufficient. This project studies two areas of concern that are not addressed by NCRP 151, the radiation shielding workload for the vault door with a non-standard design, and the attenuation properties of high-density concrete for both photon and neutron radiation. Simulations have been performed using a Monte-Carlo code produced by the Los Alamos National Lab (LANL), Monte Carlo Neutrons, Photons 5 (MCNP5). Measurements have been performed using a shielding test port designed into the maze of the Varian Edge treatment vault.

  11. Monte Carlo simulation using MCNP4B for an optimal shielding design of a 252 Cf source

    Silva, Ademir X. da; Crispim, Verginia R.

    2001-01-01

    This study aim to investigate an optimum shielding design against neutrons and gamma-rays from a source of 252 Cf, using Monte Carlo simulation. The shielding materials studied were: borated polyethylene, borated-lead polyethylene and stainless steel. The Monte Carlo code MCNP, version 4B, was used to design shielding for 252 Cf based neutron irradiator systems. By normalizing the dose equivalent rate values presented to the neutron production rate of the source, the resulting calculations are independents of the intensity of actual 252 Cf source. The results shown what the total dose equivalent rates were reduced significantly by the shielding system optimization. (author)

  12. Using the Monte Carlo Coupling Technique to Evaluate the Shielding Ability of a Modular Shielding House to Accommodate Spent-Fuel Transportable Storage Casks

    Ueki, Kohtaro; Kawakami, Kazuo; Shimizu, Daisuke

    2003-01-01

    The Monte Carlo coupling technique with the coordinate transformation is used to evaluate the shielding ability of a modular shielding house that accommodates four spent-fuel transportable storage casks for two units. The effective dose rate distributions can be obtained as far as 300 m from the center of the shielding house. The coupling technique is created with the Surface Source Write (SSW) card and the Surface Source Read/Coordinate Transformation (SSR/CRT) card in the MCNP 4C continuous energy Monte Carlo code as the 'SSW-SSR/CRT calculation system'. In the present Monte Carlo coupling calculation, the total effective dose rates 100, 200, and 300 m from the center of the shielding house are estimated to be 1.69, 0.285, and 0.0826 (μSv/yr per four casks), respectively. Accordingly, if the distance between the center of the shielding house and the site boundary of the storage facility is kept at >300 m, approximately 2400 casks are able to be accommodated in the modular shielding houses, under the Japanese severe criterion of 50 μSv/yr at the site boundary. The shielding house alone satisfies not only the technical conditions but also the economic requirements.It became evident that secondary gamma rays account for >60% of the effective total dose rate at all the calculated points around the shielding house, most of which are produced from the water in the steel-water-steel shielding system of the shielding house. The remainder of the dose rate comes mostly from neutrons; the fission product and 60 Co activation gamma rays account for small percentages. Accordingly, reducing the secondary gamma rays is critical to improving not only the shielding ability but also the radiation safety of the shielding house

  13. Monte Carlo simulations of a D-T neutron generator shielding for landmine detection

    Reda, A.M.

    2011-01-01

    Shielding for a D-T sealed neutron generator has been designed using the MCNP5 Monte Carlo radiation transport code. The neutron generator will be used in field for the detection of explosives, landmines, drugs and other 'threat' materials. The optimization of the detection of buried objects was started by studying the signal-to-noise ratio for different geometric conditions. - Highlights: → A landmine detection system based on neutron fast/slow analysis has been designed. → Shielding for a D-T sealed neutron generator tube has been designed using Monte Carlo radiation transport code. → Detection of buried objects was started by studying the signal-to-noise ratio for different geometric conditions. → The signal-to-background ratio optimized at one position for all depths.

  14. A Monte Carlo study for the shielding of γ backgrounds induced by radionuclides for CDEX

    Li Lei; Tang Changjian; Yue Qian; Cheng Jianping; Kang Kejun; Li Jianmin; Li Jin; Li Yulan; Li Yuanjing; Ma Hao; Xue Tao; Zeng Zhi; Wong, H.T.

    2011-01-01

    The CDEX (China Dark matter EXperiment) Collaboration will carry out a direct search for WIMPs (Weakly Interacting Massive Particles) using an Ultra-Low Energy Threshold High Purity Germanium (ULE-HPGe) detector at the CJPL (China JinPing deep underground Laboratory). A complex shielding system was designed to reduce backgrounds and a detailed GEANT4 Monte Carlo simulation was performed to study the achievable reduction of γ rays induced by radionuclides and neutron backgrounds by D(γ,n)p reaction. Furthermore, the upper level of allowed radio purity of shielding materials was estimated under the constraint of the expected goal. Compared with the radio purity reported by other low-background rare-event experiments, it indicates that the shielding used in the CDEX can be made out of materials with obtainable radiopurity. (authors)

  15. Monteray Mark-I: Computer program (PC-version) for shielding calculation with Monte Carlo method

    Pudjijanto, M.S.; Akhmad, Y.R.

    1998-01-01

    A computer program for gamma ray shielding calculation using Monte Carlo method has been developed. The program is written in WATFOR77 language. The MONTERAY MARH-1 is originally developed by James Wood. The program was modified by the authors that the modified version is easily executed. Applying Monte Carlo method the program observe photon gamma transport in an infinity planar shielding with various thick. A photon gamma is observed till escape from the shielding or when its energy less than the cut off energy. Pair production process is treated as pure absorption process that annihilation photons generated in the process are neglected in the calculation. The out put data calculated by the program are total albedo, build-up factor, and photon spectra. The calculation result for build-up factor of a slab lead and water media with 6 MeV parallel beam gamma source shows that they are in agreement with published data. Hence the program is adequate as a shielding design tool for observing gamma radiation transport in various media

  16. Antiproton annihilation physics annihilation physics in the Monte Carlo particle transport code particle transport code SHIELD-HIT12A

    Taasti, Vicki Trier; Knudsen, Helge; Holzscheiter, Michael

    2015-01-01

    The Monte Carlo particle transport code SHIELD-HIT12A is designed to simulate therapeutic beams for cancer radiotherapy with fast ions. SHIELD-HIT12A allows creation of antiproton beam kernels for the treatment planning system TRiP98, but first it must be benchmarked against experimental data. An...

  17. Three-dimensional coupled Monte Carlo-discrete ordinates computational scheme for shielding calculations of large and complex nuclear facilities

    Chen, Y.; Fischer, U.

    2005-01-01

    Shielding calculations of advanced nuclear facilities such as accelerator based neutron sources or fusion devices of the tokamak type are complicated due to their complex geometries and their large dimensions, including bulk shields of several meters thickness. While the complexity of the geometry in the shielding calculation can be hardly handled by the discrete ordinates method, the deep penetration of radiation through bulk shields is a severe challenge for the Monte Carlo particle transport technique. This work proposes a dedicated computational scheme for coupled Monte Carlo-Discrete Ordinates transport calculations to handle this kind of shielding problems. The Monte Carlo technique is used to simulate the particle generation and transport in the target region with both complex geometry and reaction physics, and the discrete ordinates method is used to treat the deep penetration problem in the bulk shield. The coupling scheme has been implemented in a program system by loosely integrating the Monte Carlo transport code MCNP, the three-dimensional discrete ordinates code TORT and a newly developed coupling interface program for mapping process. Test calculations were performed with comparison to MCNP solutions. Satisfactory agreements were obtained between these two approaches. The program system has been chosen to treat the complicated shielding problem of the accelerator-based IFMIF neutron source. The successful application demonstrates that coupling scheme with the program system is a useful computational tool for the shielding analysis of complex and large nuclear facilities. (authors)

  18. A practical look at Monte Carlo variance reduction methods in radiation shielding

    Olsher, Richard H. [Los Alamos National Laboratory, Los Alamos (United States)

    2006-04-15

    With the advent of inexpensive computing power over the past two decades, applications of Monte Carlo radiation transport techniques have proliferated dramatically. At Los Alamos, the Monte Carlo codes MCNP5 and MCNPX are used routinely on personal computer platforms for radiation shielding analysis and dosimetry calculations. These codes feature a rich palette of Variance Reduction (VR) techniques. The motivation of VR is to exchange user efficiency for computational efficiency. It has been said that a few hours of user time often reduces computational time by several orders of magnitude. Unfortunately, user time can stretch into the many hours as most VR techniques require significant user experience and intervention for proper optimization. It is the purpose of this paper to outline VR strategies, tested in practice, optimized for several common radiation shielding tasks, with the hope of reducing user setup time for similar problems. A strategy is defined in this context to mean a collection of MCNP radiation transport physics options and VR techniques that work synergistically to optimize a particular shielding task. Examples are offered the areas of source definition, skyshine, streaming, and transmission.

  19. Monte-Carlo simulations of neutron shielding for the ATLAS forward region

    Stekl, I; Kovalenko, V E; Vorobel, V; Leroy, C; Piquemal, F; Eschbach, R; Marquet, C

    2000-01-01

    The effectiveness of different types of neutron shielding for the ATLAS forward region has been studied by means of Monte-Carlo simulations and compared with the results of an experiment performed at the CERN PS. The simulation code is based on GEANT, FLUKA, MICAP and GAMLIB. GAMLIB is a new library including processes with gamma-rays produced in (n, gamma), (n, n'gamma) neutron reactions and is interfaced to the MICAP code. The effectiveness of different types of shielding against neutrons and gamma-rays, composed from different types of material, such as pure polyethylene, borated polyethylene, lithium-filled polyethylene, lead and iron, were compared. The results from Monte-Carlo simulations were compared to the results obtained from the experiment. The simulation results reproduce the experimental data well. This agreement supports the correctness of the simulation code used to describe the generation, spreading and absorption of neutrons (up to thermal energies) and gamma-rays in the shielding materials....

  20. A practical look at Monte Carlo variance reduction methods in radiation shielding

    Olsher, Richard H.

    2006-01-01

    With the advent of inexpensive computing power over the past two decades, applications of Monte Carlo radiation transport techniques have proliferated dramatically. At Los Alamos, the Monte Carlo codes MCNP5 and MCNPX are used routinely on personal computer platforms for radiation shielding analysis and dosimetry calculations. These codes feature a rich palette of Variance Reduction (VR) techniques. The motivation of VR is to exchange user efficiency for computational efficiency. It has been said that a few hours of user time often reduces computational time by several orders of magnitude. Unfortunately, user time can stretch into the many hours as most VR techniques require significant user experience and intervention for proper optimization. It is the purpose of this paper to outline VR strategies, tested in practice, optimized for several common radiation shielding tasks, with the hope of reducing user setup time for similar problems. A strategy is defined in this context to mean a collection of MCNP radiation transport physics options and VR techniques that work synergistically to optimize a particular shielding task. Examples are offered the areas of source definition, skyshine, streaming, and transmission

  1. Buildup factors for multilayer shieldings in deterministic methods and their comparison with Monte Carlo

    Listjak, M.; Slavik, O.; Kubovcova, D.; Vermeersch, F.

    2008-01-01

    In general there are two ways how to calculate effective doses. The first way is by use of deterministic methods like point kernel method which is implemented in Visiplan or Microshield. These kind of calculations are very fast, but they are not very convenient for a complex geometry with shielding composed of more then one material in meaning of result precision. In spite of this that programs are sufficient for ALARA optimisation calculations. On other side there are Monte Carlo methods which can be used for calculations. This way of calculation is quite precise in comparison with reality but calculation time is usually very large. Deterministic method like programs have one disadvantage -usually there is option to choose buildup factor (BUF) only for one material in multilayer stratified slabs shielding calculation problems even if shielding is composed from different materials. In literature there are proposed different formulas for multilayer BUF approximation. Aim of this paper was to examine these different formulas and their comparison with MCNP calculations. At first ware compared results of Visiplan and Microshield. Simple geometry was modelled - point source behind single and double slab shielding. For Build-up calculations was chosen Geometric Progression method (feature of the newest version of Visiplan) because there are lower deviations in comparison with Taylor fitting. (authors)

  2. Determination of shielding parameters for different types of concretes by Monte Carlo methods

    Aminian, A.; Nematollahi, M. R.

    2007-01-01

    The chose of a suitable concrete composition for a biological reactor shield remain as a research target up to now. In the present study the attempts has been made to estimate the influence of the concrete aggregates on the shielding parameters for three type of ordinary, serpentine and steel magnetite concrete by Monte Carlo N-Particle (MCNP ) transport code. MCNP calculations have been performed in order to obtain the leakage of neutrons, photons and electrons from dry shield. Also the mass attenuation coefficients and the liner attenuation coefficient are estimated for neutron and photon in those energies in range of actual energy which exist out of pressure vessel of power reactor in the cavity for the investigated concretes. The concrete densities ranged from 2.3 to 5.11 g/cm 3 . These calculations were done in the condition of a typical commercial Pressurized Water Reactor (PWR). The results show that Steel-magnetite concrete, with high density (5.11 g/cm 3 ) and constituents of relatively high atomic number, is an effective shield for both photons and neutrons

  3. Buildup factors for multilayer shieldings in deterministic methods and their comparison with Monte Carlo

    Listjak, M.; Slavik, O.; Kubovcova, D.; Vermeersch, F.

    2009-01-01

    In general there are two ways how to calculate effective doses. The first way is by use of deterministic methods like point kernel method which is implemented in Visiplan or Microshield. These kind of calculations are very fast, but they are not very convenient for a complex geometry with shielding composed of more then one material in meaning of result precision. In spite of this that programs are sufficient for ALARA optimisation calculations. On other side there are Monte Carlo methods which can be used for calculations. This way of calculation is quite precise in comparison with reality but calculation time is usually very large. Deterministic method like programs have one disadvantage -usually there is option to choose buildup factor (BUF) only for one material in multilayer stratified slabs shielding calculation problems even if shielding is composed from different materials. In literature there are proposed different formulas for multilayer BUF approximation. Aim of this paper was to examine these different formulas and their comparison with MCNP calculations. At first ware compared results of Visiplan and Microshield. Simple geometry was modelled - point source behind single and double slab shielding. For Build-up calculations was chosen Geometric Progression method (feature of the newest version of Visiplan) because there are lower deviations in comparison with Taylor fitting. (authors)

  4. Reliability of Monte Carlo simulations in modeling neutron yields from a shielded fission source

    McArthur, Matthew S., E-mail: matthew.s.mcarthur@gmail.com; Rees, Lawrence B., E-mail: Lawrence_Rees@byu.edu; Czirr, J. Bart, E-mail: czirr@juno.com

    2016-08-11

    Using the combination of a neutron-sensitive {sup 6}Li glass scintillator detector with a neutron-insensitive {sup 7}Li glass scintillator detector, we are able to make an accurate measurement of the capture rate of fission neutrons on {sup 6}Li. We used this detector with a {sup 252}Cf neutron source to measure the effects of both non-borated polyethylene and 5% borated polyethylene shielding on detection rates over a range of shielding thicknesses. Both of these measurements were compared with MCNP calculations to determine how well the calculations reproduced the measurements. When the source is highly shielded, the number of interactions experienced by each neutron prior to arriving at the detector is large, so it is important to compare Monte Carlo modeling with actual experimental measurements. MCNP reproduces the data fairly well, but it does generally underestimate detector efficiency both with and without polyethylene shielding. For non-borated polyethylene it underestimates the measured value by an average of 8%. This increases to an average of 11% for borated polyethylene.

  5. Measurements and Monte-Carlo simulations of the particle self-shielding effect of B4C grains in neutron shielding concrete

    DiJulio, D. D.; Cooper-Jensen, C. P.; Llamas-Jansa, I.; Kazi, S.; Bentley, P. M.

    2018-06-01

    A combined measurement and Monte-Carlo simulation study was carried out in order to characterize the particle self-shielding effect of B4C grains in neutron shielding concrete. Several batches of a specialized neutron shielding concrete, with varying B4C grain sizes, were exposed to a 2 Å neutron beam at the R2D2 test beamline at the Institute for Energy Technology located in Kjeller, Norway. The direct and scattered neutrons were detected with a neutron detector placed behind the concrete blocks and the results were compared to Geant4 simulations. The particle self-shielding effect was included in the Geant4 simulations by calculating effective neutron cross-sections during the Monte-Carlo simulation process. It is shown that this method well reproduces the measured results. Our results show that shielding calculations for low-energy neutrons using such materials would lead to an underestimate of the shielding required for a certain design scenario if the particle self-shielding effect is not included in the calculations.

  6. Development and application of the automated Monte Carlo biasing procedure in SAS4

    Tang, J.S.; Broadhead, B.L.

    1993-01-01

    An automated approach for biasing Monte Carlo shielding calculations is described. In particular, adjoint fluxes from a one-dimensional discrete-ordinates calculation are used to generate biasing parameters for a three-dimensional Monte Carlo calculation. The automated procedure consisting of cross-section processing, adjoint flux determination, biasing parameter generation, and the initiation of a MORSE-SGC/S Monte Carlo calculation has been implemented in the SAS4 module of the SCALE computer code system. The automated procedure has been used extensively in the investigation of both computational and experimental benchmarks for the NEACRP working group on shielding assessment of transportation packages. The results of these studies indicate that with the automated biasing procedure, Monte Carlo shielding calculations of spent fuel casks can be easily performed with minimum effort and that accurate results can be obtained at reasonable computing cost. The systematic biasing approach described in this paper can also be applied to other similar shielding problems

  7. Automated importance generation and biasing techniques for Monte Carlo shielding techniques by the TRIPOLI-3 code

    Both, J.P.; Nimal, J.C.; Vergnaud, T.

    1990-01-01

    We discuss an automated biasing procedure for generating the parameters necessary to achieve efficient Monte Carlo biasing shielding calculations. The biasing techniques considered here are exponential transform and collision biasing deriving from the concept of the biased game based on the importance function. We use a simple model of the importance function with exponential attenuation as the distance to the detector increases. This importance function is generated on a three-dimensional mesh including geometry and with graph theory algorithms. This scheme is currently being implemented in the third version of the neutron and gamma ray transport code TRIPOLI-3. (author)

  8. Monte Carlo simulations for the space radiation superconducting shield project (SR2S).

    Vuolo, M; Giraudo, M; Musenich, R; Calvelli, V; Ambroglini, F; Burger, W J; Battiston, R

    2016-02-01

    Astronauts on deep-space long-duration missions will be exposed for long time to galactic cosmic rays (GCR) and Solar Particle Events (SPE). The exposure to space radiation could lead to both acute and late effects in the crew members and well defined countermeasures do not exist nowadays. The simplest solution given by optimized passive shielding is not able to reduce the dose deposited by GCRs below the actual dose limits, therefore other solutions, such as active shielding employing superconducting magnetic fields, are under study. In the framework of the EU FP7 SR2S Project - Space Radiation Superconducting Shield--a toroidal magnetic system based on MgB2 superconductors has been analyzed through detailed Monte Carlo simulations using Geant4 interface GRAS. Spacecraft and magnets were modeled together with a simplified mechanical structure supporting the coils. Radiation transport through magnetic fields and materials was simulated for a deep-space mission scenario, considering for the first time the effect of secondary particles produced in the passage of space radiation through the active shielding and spacecraft structures. When modeling the structures supporting the active shielding systems and the habitat, the radiation protection efficiency of the magnetic field is severely decreasing compared to the one reported in previous studies, when only the magnetic field was modeled around the crew. This is due to the large production of secondary radiation taking place in the material surrounding the habitat. Copyright © 2016 The Committee on Space Research (COSPAR). Published by Elsevier Ltd. All rights reserved.

  9. The Adjoint Method for The Optimization of Brachytherapy and Radiotherapy Patient Treatment Planning Procedures Using Monte Carlo Calculations

    Henderson, D.L.; Yoo, S.; Kowalok, M.; Mackie, T.R.; Thomadsen, B.R.

    2001-01-01

    The goal of this project is to investigate the use of the adjoint method, commonly used in the reactor physics community, for the optimization of radiation therapy patient treatment plans. Two different types of radiation therapy are being examined, interstitial brachytherapy and radiotherapy. In brachytherapy radioactive sources are surgically implanted within the diseased organ such as the prostate to treat the cancerous tissue. With radiotherapy, the x-ray source is usually located at a distance of about 1-meter from the patient and focused on the treatment area. For brachytherapy the optimization phase of the treatment plan consists of determining the optimal placement of the radioactive sources, which delivers the prescribed dose to the disease tissue while simultaneously sparing (reducing) the dose to sensitive tissue and organs. For external beam radiation therapy the optimization phase of the treatment plan consists of determining the optimal direction and intensity of beam, which provides complete coverage of the tumor region with the prescribed dose while simultaneously avoiding sensitive tissue areas. For both therapy methods, the optimal treatment plan is one in which the diseased tissue has been treated with the prescribed dose and dose to the sensitive tissue and organs has been kept to a minimum

  10. Development of a shield based on Monte-Carlo studies for the COBRA experiment

    Heidrich, Nadine [Institut fuer Experimentalphysik, 22761 Hamburg (Germany); Collaboration: COBRA-Collaboration

    2013-07-01

    COBRA is a next-generation experiment searching for neutrinoless double beta decay using CdZnTe semiconductor detectors. The main focus is on {sup 116}Cd, with a decay energy of 2813.5 keV well above the highest naturally occurring gamma lines. The concept for a large scale set-up consists of an array of CdZnTe detectors with a total mass of 420 kg enriched in {sup 116}Cd up to 90 %. With a background rate in the order of 10{sup -3} counts/keV/kg/year, the experiment would be sensitive to a half-life larger than 10{sup 26} years, corresponding to a Majorana mass term m{sub ββ} smaller than 50 meV. To achieve the background level, an appropriate shield is necessary. The shield is developed based on Monte-Carlo simulations. For that, different materials and configurations are tested. In the talk the current status of the Monte-Carlo survey is presented and discussed.

  11. Monte Carlo simulation of photon buildup factors for shielding materials in diagnostic x-ray facilities

    Kharrati, Hedi; Agrebi, Amel; Karoui, Mohamed Karim

    2012-01-01

    Purpose: A simulation of buildup factors for ordinary concrete, steel, lead, plate glass, lead glass, and gypsum wallboard in broad beam geometry for photons energies from 10 keV to 150 keV at 5 keV intervals is presented. Methods: Monte Carlo N-particle radiation transport computer code has been used to determine the buildup factors for the studied shielding materials. Results: An example concretizing the use of the obtained buildup factors data in computing the broad beam transmission for tube potentials at 70, 100, 120, and 140 kVp is given. The half value layer, the tenth value layer, and the equilibrium tenth value layer are calculated from the broad beam transmission for these tube potentials. Conclusions: The obtained values compared with those calculated from the published data show the ability of these data to predict shielding transmission curves. Therefore, the buildup factors data can be combined with primary, scatter, and leakage x-ray spectra to provide a computationally based solution to broad beam transmission for barriers in shielding x-ray facilities.

  12. Efficient heterogeneous execution of Monte Carlo shielding calculations on a Beowulf cluster

    Dewar, D.; Hulse, P.; Cooper, A.; Smith, N.

    2005-01-01

    Recent work has been done in using a high-performance 'Beowulf' cluster computer system for the efficient distribution of Monte Carlo shielding calculations. This has enabled the rapid solution of complex shielding problems at low cost and with greater modularity and scalability than traditional platforms. The work has shown that a simple approach to distributing the workload is as efficient as using more traditional techniques such as PVM (Parallel Virtual Machine). In addition, when used in an operational setting this technique is fairer with the use of resources than traditional methods, in that it does not tie up a single computing resource but instead shares the capacity with other tasks. These developments in computing technology have enabled shielding problems to be solved that would have taken an unacceptably long time to run on traditional platforms. This paper discusses the BNFL Beowulf cluster and a number of tests that have recently been run to demonstrate the efficiency of the asynchronous technique in running the MCBEND program. The BNFL Beowulf currently consists of 84 standard PCs running RedHat Linux. Current performance of the machine has been estimated to be between 40 and 100 Gflop s -1 . When the whole system is employed on one problem up to four million particles can be tracked per second. There are plans to review its size in line with future business needs. (authors)

  13. Monte Carlo simulation of photon buildup factors for shielding materials in diagnostic x-ray facilities.

    Kharrati, Hedi; Agrebi, Amel; Karoui, Mohamed Karim

    2012-10-01

    A simulation of buildup factors for ordinary concrete, steel, lead, plate glass, lead glass, and gypsum wallboard in broad beam geometry for photons energies from 10 keV to 150 keV at 5 keV intervals is presented. Monte Carlo N-particle radiation transport computer code has been used to determine the buildup factors for the studied shielding materials. An example concretizing the use of the obtained buildup factors data in computing the broad beam transmission for tube potentials at 70, 100, 120, and 140 kVp is given. The half value layer, the tenth value layer, and the equilibrium tenth value layer are calculated from the broad beam transmission for these tube potentials. The obtained values compared with those calculated from the published data show the ability of these data to predict shielding transmission curves. Therefore, the buildup factors data can be combined with primary, scatter, and leakage x-ray spectra to provide a computationally based solution to broad beam transmission for barriers in shielding x-ray facilities.

  14. Toolkit for high performance Monte Carlo radiation transport and activation calculations for shielding applications in ITER

    Serikov, A.; Fischer, U.; Grosse, D.; Leichtle, D.; Majerle, M.

    2011-01-01

    The Monte Carlo (MC) method is the most suitable computational technique of radiation transport for shielding applications in fusion neutronics. This paper is intended for sharing the results of long term experience of the fusion neutronics group at Karlsruhe Institute of Technology (KIT) in radiation shielding calculations with the MCNP5 code for the ITER fusion reactor with emphasizing on the use of several ITER project-driven computer programs developed at KIT. Two of them, McCad and R2S, seem to be the most useful in radiation shielding analyses. The McCad computer graphical tool allows to perform automatic conversion of the MCNP models from the underlying CAD (CATIA) data files, while the R2S activation interface couples the MCNP radiation transport with the FISPACT activation allowing to estimate nuclear responses such as dose rate and nuclear heating after the ITER reactor shutdown. The cell-based R2S scheme was applied in shutdown photon dose analysis for the designing of the In-Vessel Viewing System (IVVS) and the Glow Discharge Cleaning (GDC) unit in ITER. Newly developed at KIT mesh-based R2S feature was successfully tested on the shutdown dose rate calculations for the upper port in the Neutral Beam (NB) cell of ITER. The merits of McCad graphical program were broadly acknowledged by the neutronic analysts and its continuous improvement at KIT has introduced its stable and more convenient run with its Graphical User Interface. Detailed 3D ITER neutronic modeling with the MCNP Monte Carlo method requires a lot of computation resources, inevitably leading to parallel calculations on clusters. Performance assessments of the MCNP5 parallel runs on the JUROPA/HPC-FF supercomputer cluster permitted to find the optimal number of processors for ITER-type runs. (author)

  15. Calculation of extended shields in the Monte Carlo method using importance function (BRAND and DD code systems)

    Androsenko, A.A.; Androsenko, P.A.; Kagalenko, I.Eh.; Mironovich, Yu.N.

    1992-01-01

    Consideration is given of a technique and algorithms of constructing neutron trajectories in the Monte-Carlo method taking into account the data on adjoint transport equation solution. When simulating the transport part of transfer kernel the use is made of piecewise-linear approximation of free path length density along the particle motion direction. The approach has been implemented in programs within the framework of the BRAND code system. The importance is calculated in the multigroup P 1 -approximation within the framework of the DD-30 code system. The efficiency of the developed computation technique is demonstrated by means of solution of two model problems. 4 refs.; 2 tabs

  16. Monte Carlo-based development of a shield and total background estimation for the COBRA experiment

    Heidrich, Nadine

    2014-11-01

    The COBRA experiment aims for the measurement of the neutrinoless double beta decay and thus for the determination the effective Majorana mass of the neutrino. To be competitive with other next-generation experiments the background rate has to be in the order of 10 -3 counts/kg/keV/yr, which is a challenging criterion. This thesis deals with the development of a shield design and the calculation of the expected total background rate for the large scale COBRA experiment containing 13824 6 cm 3 CdZnTe detectors. For the development of a shield single-layer and multi-layer shields were investigated and a shield design was optimized concerning high-energy muon-induced neutrons. As the best design the combination of 10 cm boron doped polyethylene as outermost layer, 20 cm lead and 10 cm copper as innermost layer were determined. It showed the best performance regarding neutron attenuation as well as (n, γ) self-shielding effects leading to a negligible background rate of less than 2.10 -6 counts/kg/keV/yr. Additionally. the shield with a thickness of 40 cm is compact and costeffective. In the next step the expected total background rate was computed taking into account individual setup parts and various background sources including natural and man-made radioactivity, cosmic ray-induced background and thermal neutrons. Furthermore, a comparison of measured data from the COBRA demonstrator setup with Monte Carlo data was used to calculate reliable contamination levels of the single setup parts. The calculation was performed conservatively to prevent an underestimation. In addition, the contribution to the total background rate regarding the individual detector parts and background sources was investigated. The main portion arise from the Delrin support structure, the Glyptal lacquer followed by the circuit board of the high voltage supply. Most background events originate from particles with a quantity of 99 % in total. Regarding surface events a contribution of 26

  17. Monte Carlo-based development of a shield and total background estimation for the COBRA experiment

    Heidrich, Nadine

    2014-11-15

    The COBRA experiment aims for the measurement of the neutrinoless double beta decay and thus for the determination the effective Majorana mass of the neutrino. To be competitive with other next-generation experiments the background rate has to be in the order of 10{sup -3} counts/kg/keV/yr, which is a challenging criterion. This thesis deals with the development of a shield design and the calculation of the expected total background rate for the large scale COBRA experiment containing 13824 6 cm{sup 3} CdZnTe detectors. For the development of a shield single-layer and multi-layer shields were investigated and a shield design was optimized concerning high-energy muon-induced neutrons. As the best design the combination of 10 cm boron doped polyethylene as outermost layer, 20 cm lead and 10 cm copper as innermost layer were determined. It showed the best performance regarding neutron attenuation as well as (n, γ) self-shielding effects leading to a negligible background rate of less than 2.10{sup -6} counts/kg/keV/yr. Additionally. the shield with a thickness of 40 cm is compact and costeffective. In the next step the expected total background rate was computed taking into account individual setup parts and various background sources including natural and man-made radioactivity, cosmic ray-induced background and thermal neutrons. Furthermore, a comparison of measured data from the COBRA demonstrator setup with Monte Carlo data was used to calculate reliable contamination levels of the single setup parts. The calculation was performed conservatively to prevent an underestimation. In addition, the contribution to the total background rate regarding the individual detector parts and background sources was investigated. The main portion arise from the Delrin support structure, the Glyptal lacquer followed by the circuit board of the high voltage supply. Most background events originate from particles with a quantity of 99 % in total. Regarding surface events a

  18. Evaluation of the shield calculation adequacy of radiotherapy rooms through Monte Carlo Method and experimental measures

    Meireles, Ramiro Conceicao

    2016-01-01

    The shielding calculation methodology for radiotherapy services adopted in Brazil and in several countries is that described in publication 151 of the National Council on Radiation Protection and Measurements (NCRP 151). This methodology however, markedly employs several approaches that can impact both in the construction cost and in the radiological safety of the facility. Although this methodology is currently well established by the high level of use, some parameters employed in the calculation methodology did not undergo to a detailed assessment to evaluate the impact of the various approaches considered. In this work the MCNP5 Monte Carlo code was used with the purpose of evaluating the above mentioned approaches. TVLs values were obtained for photons in conventional concrete (2.35g / cm 3 ), considering the energies of 6, 10 and 25 MeV, respectively, first considering an isotropic radiation source impinging perpendicular to the barriers, and subsequently a lead head shielding emitting a shaped beam, in the format of a pyramid trunk. Primary barriers safety margins, taking in account the head shielding emitting photon beam pyramid-shaped in the energies of 6, 10, 15 and 18 MeV were assessed. A study was conducted considering the attenuation provided by the patient's body in the energies of 6,10, 15 and 18 MeV, leading to new attenuation factors. Experimental measurements were performed in a real radiotherapy room, in order to map the leakage radiation emitted by the accelerator head shielding and the results obtained were employed in the Monte Carlo simulation, as well as to validate the entire study. The study results indicate that the TVLs values provided by (NCRP, 2005) show discrepancies in comparison with the values obtained by simulation and that there may be some barriers that are calculated with insufficient thickness. Furthermore, the simulation results show that the additional safety margins considered when calculating the width of the primary

  19. Application of variance reduction techniques of Monte-Carlo method to deep penetration shielding problems

    Rawat, K.K.; Subbaiah, K.V.

    1996-01-01

    General purpose Monte Carlo code MCNP is being widely employed for solving deep penetration problems by applying variance reduction techniques. These techniques depend on the nature and type of the problem being solved. Application of geometry splitting and implicit capture method are examined to study the deep penetration problems of neutron, gamma and coupled neutron-gamma in thick shielding materials. The typical problems chosen are: i) point isotropic monoenergetic gamma ray source of 1 MeV energy in nearly infinite water medium, ii) 252 Cf spontaneous source at the centre of 140 cm thick water and concrete and iii) 14 MeV fast neutrons incident on the axis of 100 cm thick concrete disk. (author). 7 refs., 5 figs

  20. Photon dose estimation from ultraintense laser–solid interactions and shielding calculation with Monte Carlo simulation

    Yang, Bo; Qiu, Rui; Li, JunLi; Lu, Wei; Wu, Zhen; Li, Chunyan

    2017-01-01

    When a strong laser beam irradiates a solid target, a hot plasma is produced and high-energy electrons are usually generated (the so-called “hot electrons”). These energetic electrons subsequently generate hard X-rays in the solid target through the Bremsstrahlung process. To date, only limited studies have been conducted on this laser-induced radiological protection issue. In this study, extensive literature reviews on the physics and properties of hot electrons have been conducted. On the basis of these information, the photon dose generated by the interaction between hot electrons and a solid target was simulated with the Monte Carlo code FLUKA. With some reasonable assumptions, the calculated dose can be regarded as the upper boundary of the experimental results over the laser intensity ranging from 10 19 to 10 21 W/cm 2 . Furthermore, an equation to estimate the photon dose generated from ultraintense laser–solid interactions based on the normalized laser intensity is derived. The shielding effects of common materials including concrete and lead were also studied for the laser-driven X-ray source. The dose transmission curves and tenth-value layers (TVLs) in concrete and lead were calculated through Monte Carlo simulations. These results could be used to perform a preliminary and fast radiation safety assessment for the X-rays generated from ultraintense laser–solid interactions. - Highlights: • The laser–driven X-ray ionizing radiation source was analyzed in this study. • An equation to estimate the photon dose based on the laser intensity is given. • The shielding effects of concrete and lead were studied for this new X-ray source. • The aim of this study is to analyze and mitigate the laser–driven X-ray hazard.

  1. Dose rate evaluation of body phantom behind ITER bio-shield wall using Monte Carlo method

    Beheshti, A.; Jabbari, I.; Karimian, A.; Abdi, M.

    2012-01-01

    One of the most critical risks to humans in reactors environment is radiation exposure. Around the tokamak hall personnel are exposed to a wide range of particles, including neutrons and photons. International Thermonuclear Experimental Reactor (ITER) is a nuclear fusion research and engineering project, which is the most advanced experimental tokamak nuclear fusion reactor. Dose rates assessment and photon radiation due to the neutron activation of the solid structures in ITER is important from the radiological point of view. Therefore, the dosimetry considered in this case is based on the Deuterium-Tritium (DT) plasma burning with neutrons production rate at 14.1 MeV. The aim of this study is assessment the amount of radiation behind bio-shield wall that a human received during normal operation of ITER by considering neutron activation and delay gammas. To achieve the aim, the ITER system and its components were simulated by Monte Carlo method. Also to increase the accuracy and precision of the absorbed dose assessment a body phantom were considered in the simulation. The results of this research showed that total dose rates level near the outside of bio-shield wall of the tokamak hall is less than ten percent of the annual occupational dose limits during normal operation of ITER and It is possible to learn how long human beings can remain in that environment before the body absorbs dangerous levels of radiation. (authors)

  2. CASIM, High Energy Cascades in Shields of Arbitrary Geometry Using Monte-Carlo Method

    Van Ginneken, A.

    1987-01-01

    1 - Description of problem or function: CASIM is a Monte Carlo program to study the average development of high energy cascades in large targets (shields) of arbitrary geometry and composition. The program is best suited for incident energies in the range 20-1000 GeV. 2 - Method of solution: The simulation makes extensive use of weighting techniques to avoid difficulties encountered in sampling complicated distributions and to allow the user to introduce bias in the sampling. The program uses the particle production model (in the form of a set of inclusive distributions) to compute (a) star densities (i.e. nuclear interaction densities) as a function of location and particle type throughout the target. From these star densities, estimates of a number of quantities of radiobiological interest can be obtained; (b) momentum spectra of particles interacting in the shield also as a function of location and type; (c) energy deposited by the cascade. This quantity is a useful measure of target heating embedded in the target (ionization calorimeter). 3 - Restrictions on the complexity of the problem: The program does not study transport of low momentum particles (less than or equal to 0.3 GeV/c)

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

    Ghassoun, Jillali; Jehoauni, Abdellatif

    2000-01-01

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

  4. Electron dose distributions caused by the contact-type metallic eye shield: Studies using Monte Carlo and pencil beam algorithms.

    Kang, Sei-Kwon; Yoon, Jai-Woong; Hwang, Taejin; Park, Soah; Cheong, Kwang-Ho; Han, Tae Jin; Kim, Haeyoung; Lee, Me-Yeon; Kim, Kyoung Ju; Bae, Hoonsik

    2015-01-01

    A metallic contact eye shield has sometimes been used for eyelid treatment, but dose distribution has never been reported for a patient case. This study aimed to show the shield-incorporated CT-based dose distribution using the Pinnacle system and Monte Carlo (MC) calculation for 3 patient cases. For the artifact-free CT scan, an acrylic shield machined as the same size as that of the tungsten shield was used. For the MC calculation, BEAMnrc and DOSXYZnrc were used for the 6-MeV electron beam of the Varian 21EX, in which information for the tungsten, stainless steel, and aluminum material for the eye shield was used. The same plan was generated on the Pinnacle system and both were compared. The use of the acrylic shield produced clear CT images, enabling delineation of the regions of interest, and yielded CT-based dose calculation for the metallic shield. Both the MC and the Pinnacle systems showed a similar dose distribution downstream of the eye shield, reflecting the blocking effect of the metallic eye shield. The major difference between the MC and the Pinnacle results was the target eyelid dose upstream of the shield such that the Pinnacle system underestimated the dose by 19 to 28% and 11 to 18% for the maximum and the mean doses, respectively. The pattern of dose difference between the MC and the Pinnacle systems was similar to that in the previous phantom study. In conclusion, the metallic eye shield was successfully incorporated into the CT-based planning, and the accurate dose calculation requires MC simulation. Copyright © 2015 American Association of Medical Dosimetrists. Published by Elsevier Inc. All rights reserved.

  5. Electron dose distributions caused by the contact-type metallic eye shield: Studies using Monte Carlo and pencil beam algorithms

    Kang, Sei-Kwon; Yoon, Jai-Woong; Hwang, Taejin; Park, Soah; Cheong, Kwang-Ho; Jin Han, Tae; Kim, Haeyoung; Lee, Me-Yeon; Ju Kim, Kyoung, E-mail: kjkim@hallym.or.kr; Bae, Hoonsik

    2015-10-01

    A metallic contact eye shield has sometimes been used for eyelid treatment, but dose distribution has never been reported for a patient case. This study aimed to show the shield-incorporated CT-based dose distribution using the Pinnacle system and Monte Carlo (MC) calculation for 3 patient cases. For the artifact-free CT scan, an acrylic shield machined as the same size as that of the tungsten shield was used. For the MC calculation, BEAMnrc and DOSXYZnrc were used for the 6-MeV electron beam of the Varian 21EX, in which information for the tungsten, stainless steel, and aluminum material for the eye shield was used. The same plan was generated on the Pinnacle system and both were compared. The use of the acrylic shield produced clear CT images, enabling delineation of the regions of interest, and yielded CT-based dose calculation for the metallic shield. Both the MC and the Pinnacle systems showed a similar dose distribution downstream of the eye shield, reflecting the blocking effect of the metallic eye shield. The major difference between the MC and the Pinnacle results was the target eyelid dose upstream of the shield such that the Pinnacle system underestimated the dose by 19 to 28% and 11 to 18% for the maximum and the mean doses, respectively. The pattern of dose difference between the MC and the Pinnacle systems was similar to that in the previous phantom study. In conclusion, the metallic eye shield was successfully incorporated into the CT-based planning, and the accurate dose calculation requires MC simulation.

  6. Monte Carlo analysis of helium production in the ITER shielding blanket module

    Sato, S.

    1999-01-01

    In order to examine the shielding performances of the inboard blanket module in the international thermonuclear experimental reactor (ITER), shielding calculations have been carried out using a three-dimensional Monte Carlo method. The impact of radiation streaming through the front access holes and gaps between adjacent blanket modules on the helium gas production in the branch pipe weld locations and back plate have been estimated. The three-dimensional model represents an 18 sector of the overall torus region and includes the vacuum vessel, inboard blanket and back plate, plasma region, and outboard reflecting medium. And it includes the 1 m high inboard mid-plane module and the 20 mm wide gaps between adjacent modules. From the calculated results for the reference design, it has been found that the helium production at the plug of the branch pipe is four to five times higher than the design goal of 1 appm for a neutron fluence of 0.9 MW a m -2 at the inboard mid-plane first wall. Also, it has been found that the helium production at the back plate behind the horizontal gap is about three times higher than the design goal. In the reference design, the stainless steel (SS):H 2 O composition in the blanket module is 80:20%. Shielding calculations also have been carried out for the SS:H 2 O composition of 70:30, 60:40, 50:50 and 40:60%. From the evaluated results for their design, it has been found that the dependence of helium production on the SS:H 2 170 mm will reduce helium production to satisfy the design goal and not have a significant impact on weight limitations imposed by remote maintenance handling limitations. Also based on the calculated results, about 200 mm thick shields such as a key structure in the vertical gap are suggested to be installed in the horizontal gap as well to reduce the helium production at the back plate and to satisfy the design goal. (orig.)

  7. The Monte Carlo method for shielding calculations analysis by MORSE code of a streaming case in the CAORSO BWR power reactor shielding (Italy)

    Zitouni, Y.

    1987-04-01

    In the field of shielding, the requirement of radiation transport calculations in severe conditions, characterized by irreducible three-dimensional geometries has increased the use of the Monte Carlo method. The latter has proved to be the only rigorous and appropriate calculational method in such conditions. However, further efforts at optimization are still necessary to render the technique practically efficient, despite recent improvements in the Monte Carlo codes, the progress made in the field of computers and the availability of accurate nuclear data. Moreover, the personal experience acquired in the field and the control of sophisticated calculation procedures are of the utmost importance. The aim of the work which has been carried out is the gathering of all the necessary elements and features that would lead to an efficient utilization of the Monte Carlo method used in connection with shielding problems. The study of the general aspects of the method and the exploitation techniques of the MORSE code, which has proved to be one of the most comprehensive of the Monte Carlo codes, lead to a successful analysis of an actual case. In fact, the severe conditions and difficulties met have been overcome using such a stochastic simulation code. Finally, a critical comparison between calculated and high-accuracy experimental results has allowed the final confirmation of the methodology used by us

  8. A Monte Carlo Method for the Analysis of Gamma Radiation Transport from Distributed Sources in Laminated Shields

    Leimdoerfer, M.

    1964-02-01

    A description is given of a method for calculating the penetration and energy deposition of gamma radiation, based on Monte Carlo techniques. The essential feature is the application of the exponential transformation to promote the transport of penetrating quanta and to balance the steep spatial variations of the source distributions which appear in secondary gamma emission problems. The estimated statistical errors in a number of sample problems, involving concrete shields with thicknesses up to 500 cm, are shown to be quite favorable, even at relatively short computing times. A practical reactor shielding problem is also shown and the predictions compared with measurements

  9. A Monte Carlo Method for the Analysis of Gamma Radiation Transport from Distributed Sources in Laminated Shields

    Leimdoerfer, M

    1964-02-15

    A description is given of a method for calculating the penetration and energy deposition of gamma radiation, based on Monte Carlo techniques. The essential feature is the application of the exponential transformation to promote the transport of penetrating quanta and to balance the steep spatial variations of the source distributions which appear in secondary gamma emission problems. The estimated statistical errors in a number of sample problems, involving concrete shields with thicknesses up to 500 cm, are shown to be quite favorable, even at relatively short computing times. A practical reactor shielding problem is also shown and the predictions compared with measurements.

  10. Modelling of neutron and photon transport in iron and concrete radiation shieldings by the Monte Carlo method - Version 2

    Žukauskaite, A; Plukiene, R; Plukis, A

    2007-01-01

    Particle accelerators and other high energy facilities produce penetrating ionizing radiation (neutrons and γ-rays) that must be shielded. The objective of this work was to model photon and neutron transport in various materials, usually used as shielding, such as concrete, iron or graphite. Monte Carlo method allows obtaining answers by simulating individual particles and recording some aspects of their average behavior. In this work several nuclear experiments were modeled: AVF 65 – γ-ray beams (1-10 MeV), HIMAC and ISIS-800 – high energy neutrons (20-800 MeV) transport in iron and concrete. The results were then compared with experimental data.

  11. JMCT Monte Carlo simulation analysis of full core PWR Pin-By-Pin and shielding

    Deng, L.; Li, G.; Zhang, B.; Shangguan, D.; Ma, Y.; Hu, Z.; Fu, Y.; Li, R.; Hu, X.; Cheng, T.; Shi, D.

    2015-01-01

    This paper describes the application of the JMCT Monte Carlo code to the simulation of Kord Smith Challenge H-M model, BEAVRS model and Chinese SG-III model. For H-M model, the 6.3624 millions tally regions and the 98.3 billion neutron histories do. The detailed pin flux and energy deposition densities obtain. 95% regions have less 1% standard deviation. For BEAVRS model, firstly, we performed the neutron transport calculation of 398 axial planes in the Hot Zero Power (HZP) status. Almost the same results with MC21 and OpenMC results are achieved. The detailed pin-power density distribution and standard deviation are shown. Then, we performed the calculation of ten depletion steps in 30 axial plane cases. The depletion regions exceed 1.5 million and 12,000 processors uses. Finally, the Chinese SG-III laser model is simulated. The neutron and photon flux distributions are given, respectively. The results show that the JMCT code well suits for extremely large reactor and shielding simulation. (author)

  12. Improved Monte Carlo modelling of multi-energy a-rays penetration through thick stratified shielding slabs

    Bakos, G.C.

    2001-01-01

    This paper deals with the application of Monte Carlo method for the calculation of dose build up factor of, mixed 1.37 and 2.75 MeV, a-rays penetration through stratified shielding slabs. Six double layer shielding slabs namely, 12 A l+Fe, 12 A l+Pb, 6 F e+Al, 6 F e+Pb, 4 P b+Al, 4 P b+Fe were examined. Furthermore, experimental and theoretical results are also presented. The experimental results were taken from the experimental facility installed at the Universities Research reactor Center (Risley, UK). Activated Na2SO3 solution provided a uniform Na-24 disc source of a-rays at both energies (1.37 and 2.75 MeV) with equal intensity. The theoretical results were calculated using the Bowman and Trubey formula. This formula takes into account an exponentially decaying function of the shield thickness (in mfp) to the end point of the multi-layer slab. The experimental and theoretical results were used to evaluate the simulation results produced from a Monte Carlo program (DUTMONCA code) which was developed in Democritus University of Thrace (Xanthi, Greece). The DUTMONCA code was written in Pascal language and run on an Intel PIII-800 microprocessor. The developed code (which is an improved version of an existing Monte Carlo program) has the ability to produce good results for thick shielding slabs overcoming the problems encountered in older version program. The simulation results are compared with experimental and theoretical results. Good agreement can be observed, even for thick layer shielding slabs, although there are some wayward experimental values which are due to sources of error associated with the experimental procedure

  13. Re-Shielding of Cobalt-60 Teletherapy Rooms for Tomotherapy and Conventional Linear Accelerators using Monte Carlo Simulations

    Çeçen, Yiğit; Yazgan, Çağrı

    2017-09-01

    Purpose. Nearly all Cobalt-60 teletherapy machines were removed around the world during the last two decades. The remaining ones are being used for experimental purposes. However, the rooms of these teletherapy machines are valuable because of lack of space in radiotherapy clinics. In order to place a new technology treatment machine in one of these rooms, one should re-shield the room since it was designed only for 1.25 MeV gamma beams on average. Mostly, the vendor of the new machine constructs the new shielding of the room using their experience. However, every radiotherapy room has different surrounding work areas and it would be wise to shield the room considering these special conditions. Also, the shield design goal of the clinic may be much lower than the International Atomic Energy Agency (IAEA) or the local association accepts. The study shows re-shielding of a Cobalt-60 room, specific to the clinic, using Monte Carlo simulations. Materials & Methods: First, a 6 MV Tomotherapy machine, then a 10 MV conventional linear accelerator (LINAC) was placed inside the Cobalt-60 teletherapy room. The photon flux outside the room was simulated using Monte Carlo N-Particle (MCNP6.1) code before and after re-shielding. For the Tomotherapy simulation, flux distributions around the machine were obtained from the vendor and implemented as the source of the model. The LINAC model was more generic with the 10 MeV electron source, the tungsten target, first and secondary collimators. The aim of the model was to obtain the maximum (40x40 cm2) open field at the isocenter. Two different simulations were carried out for gantry angles 90o and 270o. The LINAC was placed in the room such that the primary walls were A' (Gantry 270o) and C' (Gantry 90o) (figure 1). The second part of the study was to model the re-shielding of the room for Tomotherapy and for the conventional LINAC, separately. The aim was to investigate the recommended shielding by the vendors. Left side of the room

  14. Introduction to Adjoint Models

    Errico, Ronald M.

    2015-01-01

    In this lecture, some fundamentals of adjoint models will be described. This includes a basic derivation of tangent linear and corresponding adjoint models from a parent nonlinear model, the interpretation of adjoint-derived sensitivity fields, a description of methods of automatic differentiation, and the use of adjoint models to solve various optimization problems, including singular vectors. Concluding remarks will attempt to correct common misconceptions about adjoint models and their utilization.

  15. The adjoint string at finite temperature

    Damgaard, P.H.

    1986-10-01

    Expectations for the behavior of the adjoint string at finite temperature are presented. In the Migdal-Kadanoff approximation a real-space renormalization group study of the effective Polyakov like action predicts a deconfinement-like crossover for adjoint sources at a temperature slightly below the deconfinement temperature of fundamental sources. This prediction is compared with a Monte Carlo simulation of SU(2) lattice gauge theory on an 8 3 x2 lattice. (orig.)

  16. COG10, Multiparticle Monte Carlo Code System for Shielding and Criticality Use

    2007-01-01

    1 - Description of program or function: COG is a modern, full-featured Monte Carlo radiation transport code which provides accurate answers to complex shielding, criticality, and activation problems. COG was written to be state-of-the-art and free of physics approximations and compromises found in earlier codes. COG is fully 3-D, uses point-wise cross sections and exact angular scattering, and allows a full range of biasing options to speed up solutions for deep penetration problems. Additionally, a criticality option is available for computing Keff for assemblies of fissile materials. ENDL or ENDFB cross section libraries may be used. COG home page: http://www-phys.llnl.gov/N_Div/COG/. Cross section libraries are included in the package. COG can use either the LLNL ENDL-90 cross section set or the ENDFB/VI set. Analytic surfaces are used to describe geometric boundaries. Parts (volumes) are described by a method of Constructive Solid Geometry. Surface types include surfaces of up to fourth order, and pseudo-surfaces such as boxes, finite cylinders, and figures of revolution. Repeated assemblies need be defined only once. Parts are visualized in cross-section and perspective picture views. Source and random-walk biasing techniques may be selected to improve solution statistics. These include source angular biasing, importance weighting, particle splitting and Russian roulette, path-length stretching, point detectors, scattered direction biasing, and forced collisions. Criticality - For a fissioning system, COG will compute Keff by transporting batches of neutrons through the system. Activation - COG can compute gamma-ray doses due to neutron-activated materials, starting with just a neutron source. Coupled Problems - COG can solve coupled problems involving neutrons, photons, and electrons. 2 - Methods:COG uses Monte Carlo methods to solve the Boltzmann transport equation for particles traveling through arbitrary 3-dimensional geometries. Neutrons, photons

  17. Modeling of neutron and photon transport in iron and concrete radiation shields by using Monte Carlo method

    Žukauskaitėa, A; Plukienė, R; Ridikas, D

    2007-01-01

    Particle accelerators and other high energy facilities produce penetrating ionizing radiation (neutrons and γ-rays) that must be shielded. The objective of this work was to model photon and neutron transport in various materials, usually used as shielding, such as concrete, iron or graphite. Monte Carlo method allows obtaining answers by simulating individual particles and recording some aspects of their average behavior. In this work several nuclear experiments were modeled: AVF 65 (AVF cyclotron of Research Center of Nuclear Physics, Osaka University, Japan) – γ-ray beams (1-10 MeV), HIMAC (heavy-ion synchrotron of the National Institute of Radiological Sciences in Chiba, Japan) and ISIS-800 (ISIS intensive spallation neutron source facility of the Rutherford Appleton laboratory, UK) – high energy neutron (20-800 MeV) transport in iron and concrete. The calculation results were then compared with experimental data.compared with experimental data.

  18. Radiation shielding techniques and applications. 4. Two-Phase Monte Carlo Approach to Photon Streaming Through Three-Legged Penetrations

    White, Travis; Hack, Joe; Nathan, Steve; Barnett, Marvin

    2001-01-01

    solutions for scattering of neutrons through multi-legged penetrations are readily available in the literature; similar analytical solutions for photon scattering through penetrations, however, are not. Therefore, computer modeling must be relied upon to perform our analyses. The computer code typically used by Westinghouse SMS in the evaluation of photon transport through complex geometries is the MCNP Monte Carlo computer code. Yet, geometries of this nature can cause problems even with the Monte Carlo codes. Striking a balance between how the code handles bulk transport through the wall with transport through the penetration void, particularly with the use of typical variance reduction methods, is difficult when trying to ensure that all the important regions of the model are sampled appropriately. The problem was broken down into several roughly independent cases. First, scatter through the penetration was considered. Second, bulk transport through the hot leg of the duct and then through the remaining thickness of wall was calculated to determine the amount of supplemental shielding required in the wall. Similar analyses were performed for the middle and cold legs of the penetration. Finally, additional external shielding from radiation streaming through the duct was determined for cases where the minimum offset distance was not feasible. Each case was broken down further into two phases. In the first phase of each case, photons were transported from the source material to an area at the face of the wall, or the opening of the duct, where photon energy and angular distributions were tallied, representing the source incident on the wall or opening. Then, a simplified model for each case was developed and analyzed using the data from the first phase and the new source term. (authors)

  19. A study on optimization of photoneutron shielding in a medical accelerator room by using Monte Carlo simulation

    Kim, Yong Nam; Jeong, Kyoungkeun; Kim, Joo Young; Lee, Chang Geol; Seong, Jinsil; Choi, Sang Hyun; Kim, Chan Hyeong

    2008-01-01

    Medical linear accelerators operating above 10 MV require door shielding for neutrons in addition to photons. A criterion for choice of optimal configuration of lamination of BPE (Borated Polyethylene) and lead is not clear. Moreover, optimal configuration cannot be determined by the conventional method using an analytical formula and simple measurement. This study performs Monte Carlo simulation of radiation field in a commercial LINAC room with 15 MV X-ray sources. Considering two configuration of lamination such as 'lead-BPE' and 'lead-BPE-lead', dose equivalents are calculated by using the MCNPX code and comparative analyses are performed with each other. The obtained results show that there is no significant difference in neuron shielding between both configurations, whereas lead-BPE-lead is more effective for photon shielding. It is also noted that the absolute values of neutron doses are much greater than that of photon doses outside as well as inside the door, by three orders of magnitude. As a conclusion, the laminating of lead-BPE is suggested as the optimal configuration from the viewpoint of simplicity in fabrication and handling, even though it has no significant difference from lead-BPE-lead in terms of total dose equivalent. (author)

  20. Verification and Validation of Monte Carlo n-Particle Code 6 (MCNP6) with Neutron Protection Factor Measurements of an Iron Box

    2014-03-27

    Vehicle Code System (VCS), the Monte Carlo Adjoint SHielding (MASH), and the Monte Carlo n- Particle ( MCNP ) code. Of the three, the oldest and still most...widely utilized radiation transport code is MCNP . First created at Los Alamos National Laboratory (LANL) in 1957, the code simulated neutral...particle types, and previous versions of MCNP were repeatedly validated using both simple and complex 10 geometries [12, 13]. Much greater discussion and

  1. Monte Carlo model for a prototype CT-compatible, anatomically adaptive, shielded intracavitary brachytherapy applicator for the treatment of cervical cancer

    Price, Michael J.; Gifford, Kent A.; Horton, John L. Jr.; Eifel, Patricia J.; Gillin, Michael T.; Lawyer, Ann A.; Mourtada, Firas [Department of Radiation Physics, University of Texas M. D. Anderson Cancer Center, 1220 Holcombe Boulevard, Houston, Texas 77030 and Graduate School of Biomedical Sciences, University of Texas-Houston, 6767 Bertner Avenue, Houston, Texas 77030 (United States); Department of Radiation Physics, University of Texas M. D. Anderson Cancer Center, 1220 Holcombe Boulevard, Houston, Texas 77030 (United States); Division of Radiation Oncology, University of Texas M. D. Anderson Cancer Center, 1220 Holcombe Boulevard, Houston, Texas 77030 and Graduate School of Biomedical Sciences, University of Texas-Houston, 6767 Bertner Avenue, Houston, Texas 77030 (United States); Department of Radiation Physics, University of Texas M. D. Anderson Cancer Center, 1220 Holcombe Boulevard, Houston, Texas 77030 and Graduate School of Biomedical Sciences, University of Texas-Houston, 6767 Bertner Avenue, Houston, Texas 77030 (United States); Department of Radiation Physics, University of Texas M. D. Anderson Cancer Center, 1220 Holcombe Boulevard, Houston, Texas 77030 (United States); Department of Radiation Physics, University of Texas M. D. Anderson Cancer Center, 1220 Holcombe Boulevard, Houston, Texas 77030 and Graduate School of Biomedical Sciences, University of Texas-Houston, 6767 Bertner Avenue, Houston, Texas 77030 (United States)

    2009-09-15

    Purpose: Current, clinically applicable intracavitary brachytherapy applicators that utilize shielded ovoids contain a pair of tungsten-alloy shields which serve to reduce dose delivered to the rectum and bladder during source afterloading. After applicator insertion, these fixed shields are not necessarily positioned to provide optimal shielding of these critical structures due to variations in patient anatomies. The authors present a dosimetric evaluation of a novel prototype intracavitary brachytherapy ovoid [anatomically adaptive applicator (A{sup 3})], featuring a single shield whose position can be adjusted with two degrees of freedom: Rotation about and translation along the long axis of the ovoid. Methods: The dosimetry of the device for a HDR {sup 192}Ir was characterized using radiochromic film measurements for various shield orientations. A MCNPX Monte Carlo model was developed of the prototype ovoid and integrated with a previously validated model of a v2 mHDR {sup 192}Ir source (Nucletron Co.). The model was validated for three distinct shield orientations using film measurements. Results: For the most complex case, 91% of the absolute simulated and measured dose points agreed within 2% or 2 mm and 96% agreed within 10% or 2 mm. Conclusions: Validation of the Monte Carlo model facilitates future investigations into any dosimetric advantages the use of the A{sup 3} may have over the current state of art with respect to optimization and customization of dose delivery as a function of patient anatomical geometries.

  2. Monte Carlo model for a prototype CT-compatible, anatomically adaptive, shielded intracavitary brachytherapy applicator for the treatment of cervical cancer

    Price, Michael J.; Gifford, Kent A.; Horton, John L. Jr.; Eifel, Patricia J.; Gillin, Michael T.; Lawyer, Ann A.; Mourtada, Firas

    2009-01-01

    Purpose: Current, clinically applicable intracavitary brachytherapy applicators that utilize shielded ovoids contain a pair of tungsten-alloy shields which serve to reduce dose delivered to the rectum and bladder during source afterloading. After applicator insertion, these fixed shields are not necessarily positioned to provide optimal shielding of these critical structures due to variations in patient anatomies. The authors present a dosimetric evaluation of a novel prototype intracavitary brachytherapy ovoid [anatomically adaptive applicator (A 3 )], featuring a single shield whose position can be adjusted with two degrees of freedom: Rotation about and translation along the long axis of the ovoid. Methods: The dosimetry of the device for a HDR 192 Ir was characterized using radiochromic film measurements for various shield orientations. A MCNPX Monte Carlo model was developed of the prototype ovoid and integrated with a previously validated model of a v2 mHDR 192 Ir source (Nucletron Co.). The model was validated for three distinct shield orientations using film measurements. Results: For the most complex case, 91% of the absolute simulated and measured dose points agreed within 2% or 2 mm and 96% agreed within 10% or 2 mm. Conclusions: Validation of the Monte Carlo model facilitates future investigations into any dosimetric advantages the use of the A 3 may have over the current state of art with respect to optimization and customization of dose delivery as a function of patient anatomical geometries.

  3. Shielding requirements for constant-potential diagnostic x-ray beams determined by a Monte Carlo calculation

    Simpkin, D.J.

    1989-01-01

    A Monte Carlo calculation has been performed to determine the transmission of broad constant-potential x-ray beams through Pb, concrete, gypsum wallboard, steel and plate glass. The EGS4 code system was used with a simple broad-beam geometric model to generate exposure transmission curves for published 70, 100, 120 and 140-kVcp x-ray spectra. These curves are compared to measured three-phase generated x-ray transmission data in the literature and found to be reasonable. For calculation ease the data are fit to an equation previously shown to describe such curves quite well. These calculated transmission data are then used to create three-phase shielding tables for Pb and concrete, as well as other materials not available in Report No. 49 of the NCRP

  4. Shielding requirements for constant-potential diagnostic x-ray beams determined by a Monte Carlo calculation.

    Simpkin, D J

    1989-02-01

    A Monte Carlo calculation has been performed to determine the transmission of broad constant-potential x-ray beams through Pb, concrete, gypsum wallboard, steel and plate glass. The EGS4 code system was used with a simple broad-beam geometric model to generate exposure transmission curves for published 70, 100, 120 and 140-kVcp x-ray spectra. These curves are compared to measured three-phase generated x-ray transmission data in the literature and found to be reasonable. For calculation ease the data are fit to an equation previously shown to describe such curves quite well. These calculated transmission data are then used to create three-phase shielding tables for Pb and concrete, as well as other materials not available in Report No. 49 of the NCRP.

  5. SU-E-T-556: Monte Carlo Generated Dose Distributions for Orbital Irradiation Using a Single Anterior-Posterior Electron Beam and a Hanging Lens Shield

    Duwel, D; Lamba, M; Elson, H; Kumar, N

    2015-01-01

    Purpose: Various cancers of the eye are successfully treated with radiotherapy utilizing one anterior-posterior (A/P) beam that encompasses the entire content of the orbit. In such cases, a hanging lens shield can be used to spare dose to the radiosensitive lens of the eye to prevent cataracts. Methods: This research focused on Monte Carlo characterization of dose distributions resulting from a single A-P field to the orbit with a hanging shield in place. Monte Carlo codes were developed which calculated dose distributions for various electron radiation energies, hanging lens shield radii, shield heights above the eye, and beam spoiler configurations. Film dosimetry was used to benchmark the coding to ensure it was calculating relative dose accurately. Results: The Monte Carlo dose calculations indicated that lateral and depth dose profiles are insensitive to changes in shield height and electron beam energy. Dose deposition was sensitive to shield radius and beam spoiler composition and height above the eye. Conclusion: The use of a single A/P electron beam to treat cancers of the eye while maintaining adequate lens sparing is feasible. Shield radius should be customized to have the same radius as the patient’s lens. A beam spoiler should be used if it is desired to substantially dose the eye tissues lying posterior to the lens in the shadow of the lens shield. The compromise between lens sparing and dose to diseased tissues surrounding the lens can be modulated by varying the beam spoiler thickness, spoiler material composition, and spoiler height above the eye. The sparing ratio is a metric that can be used to evaluate the compromise between lens sparing and dose to surrounding tissues. The higher the ratio, the more dose received by the tissues immediately posterior to the lens relative to the dose received by the lens

  6. Accurate Monte Carlo modeling of cyclotrons for optimization of shielding and activation calculations in the biomedical field

    Infantino, Angelo; Marengo, Mario; Baschetti, Serafina; Cicoria, Gianfranco; Longo Vaschetto, Vittorio; Lucconi, Giulia; Massucci, Piera; Vichi, Sara; Zagni, Federico; Mostacci, Domiziano

    2015-11-01

    Biomedical cyclotrons for production of Positron Emission Tomography (PET) radionuclides and radiotherapy with hadrons or ions are widely diffused and established in hospitals as well as in industrial facilities and research sites. Guidelines for site planning and installation, as well as for radiation protection assessment, are given in a number of international documents; however, these well-established guides typically offer analytic methods of calculation of both shielding and materials activation, in approximate or idealized geometry set up. The availability of Monte Carlo codes with accurate and up-to-date libraries for transport and interactions of neutrons and charged particles at energies below 250 MeV, together with the continuously increasing power of nowadays computers, makes systematic use of simulations with realistic geometries possible, yielding equipment and site specific evaluation of the source terms, shielding requirements and all quantities relevant to radiation protection. In this work, the well-known Monte Carlo code FLUKA was used to simulate two representative models of cyclotron for PET radionuclides production, including their targetry; and one type of proton therapy cyclotron including the energy selection system. Simulations yield estimates of various quantities of radiological interest, including the effective dose distribution around the equipment, the effective number of neutron produced per incident proton and the activation of target materials, the structure of the cyclotron, the energy degrader, the vault walls and the soil. The model was validated against experimental measurements and comparison with well-established reference data. Neutron ambient dose equivalent H*(10) was measured around a GE PETtrace cyclotron: an average ratio between experimental measurement and simulations of 0.99±0.07 was found. Saturation yield of 18F, produced by the well-known 18O(p,n)18F reaction, was calculated and compared with the IAEA recommended

  7. Automatic mesh adaptivity for hybrid Monte Carlo/deterministic neutronics modeling of difficult shielding problems

    Ibrahim, Ahmad M.; Wilson, Paul P.H.; Sawan, Mohamed E.; Mosher, Scott W.; Peplow, Douglas E.; Wagner, John C.; Evans, Thomas M.; Grove, Robert E.

    2015-01-01

    The CADIS and FW-CADIS hybrid Monte Carlo/deterministic techniques dramatically increase the efficiency of neutronics modeling, but their use in the accurate design analysis of very large and geometrically complex nuclear systems has been limited by the large number of processors and memory requirements for their preliminary deterministic calculations and final Monte Carlo calculation. Three mesh adaptivity algorithms were developed to reduce the memory requirements of CADIS and FW-CADIS without sacrificing their efficiency improvement. First, a macromaterial approach enhances the fidelity of the deterministic models without changing the mesh. Second, a deterministic mesh refinement algorithm generates meshes that capture as much geometric detail as possible without exceeding a specified maximum number of mesh elements. Finally, a weight window coarsening algorithm decouples the weight window mesh and energy bins from the mesh and energy group structure of the deterministic calculations in order to remove the memory constraint of the weight window map from the deterministic mesh resolution. The three algorithms were used to enhance an FW-CADIS calculation of the prompt dose rate throughout the ITER experimental facility. Using these algorithms resulted in a 23.3% increase in the number of mesh tally elements in which the dose rates were calculated in a 10-day Monte Carlo calculation and, additionally, increased the efficiency of the Monte Carlo simulation by a factor of at least 3.4. The three algorithms enabled this difficult calculation to be accurately solved using an FW-CADIS simulation on a regular computer cluster, eliminating the need for a world-class super computer

  8. Monte Carlo simulations for the shielding of the future high-intensity accelerator facility fair at GSI

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

    2005-01-01

    The Gesellschaft fuer Schwerionenforschung (GSI) is planning a significant expansion of its accelerator facilities. Compared to the present GSI facility, a factor of 100 in primary beam intensities and up to a factor of 10,000 in secondary radioactive beam intensities are key technical goals of the proposal. The second branch of the so-called Facility for Antiproton and Ion Research (FAIR) is the production of antiprotons and their storage in rings and traps. The facility will provide beam energies a factor of ∼15 higher than presently available at the GSI for all ions, from protons to uranium. The shielding design of the synchrotron SIS 100/300 is shown exemplarily by using Monte Carlo calculations with the FLUKA code. The experimental area serving the investigation of compressed baryonic matter is analysed in the same way. In addition, a dose comparison is made for an experimental area operated with medium energy heavy-ion beams. Here, Monte Carlo calculations are performed by using either heavy-ion primary particles or proton beams with intensities scaled by the mass number of the corresponding heavy-ion beam. (authors)

  9. Monte Carlo simulations for the shielding of the future high-intensity accelerator facility FAIR at GSI.

    Radon, T; Gutermuth, F; Fehrenbacher, G

    2005-01-01

    The Gesellschaft für Schwerionenforschung (GSI) is planning a significant expansion of its accelerator facilities. Compared to the present GSI facility, a factor of 100 in primary beam intensities and up to a factor of 10,000 in secondary radioactive beam intensities are key technical goals of the proposal. The second branch of the so-called Facility for Antiproton and Ion Research (FAIR) is the production of antiprotons and their storage in rings and traps. The facility will provide beam energies a factor of approximately 15 higher than presently available at the GSI for all ions, from protons to uranium. The shielding design of the synchrotron SIS 100/300 is shown exemplarily by using Monte Carlo calculations with the FLUKA code. The experimental area serving the investigation of compressed baryonic matter is analysed in the same way. In addition, a dose comparison is made for an experimental area operated with medium energy heavy-ion beams. Here, Monte Carlo calculations are performed by using either heavy-ion primary particles or proton beams with intensities scaled by the mass number of the corresponding heavy-ion beam.

  10. Monte Carlo simulation of calibration of shadow shield scanning bed whole body monitor using different size BOMAB phantoms

    Bhati, S.; Patni, H.K.; Singh, I.S.; Garg, S.P.

    2005-01-01

    A shadow shield scanning bed whole body monitor incorporating a (102 mm dia x 76 mm thick) NaI(Tl) detector, is employed for assessment of high-energy photon emitters at BARC. The monitor is calibrated using a Reference BOMAB phantom representative of an average Indian radiation worker. However to account for the size variation in the physique of workers, it is required to calibrate the system with different size BOMAB phantoms which is both difficult and expensive. Therefore, a theoretical approach based on Monte Carlo techniques has been employed to calibrate the system with BOMAB phantoms of different sizes for several radionuclides of interest. A computer program developed for this purpose, simulates the scanning geometry of the whole body monitor and computes detection efficiencies for the BARC Reference phantom (63 kg/168 cm), ICRP Reference phantom (70 kg/170 cm) and several of its scaled versions covering a wide range of body builds. The detection efficiencies computed for different photon energies for BARC Reference phantom were found to be in very good agreement with experimental data, thus validating the Monte Carlo scheme used in the computer code. The results from this study could be used for assessment of internal contamination due to high-energy photon emitters for radiation workers of different physiques. (author)

  11. Monte Carlo based demonstration of sufficiently dimensioned shielding for a Co-60 testing facility

    Wind, Michael; Beck, Peter; Latocha, Marcin

    2015-01-01

    The electrical properties of electronic equipment can be changed in an ionized radiation field. The knowledge of these changes is necessary for applications in space, in air traffic and nuclear medicine. Experimental tests will be performed in Co-60 radiation fields in the irradiation facility (TEC facility) of the Seibersdorf Labor GmbH that is in construction. The contribution deals with a simulation that is aimed to calculate the local dose rate within and outside the building for demonstration of sufficient dimensioning of the shielding in compliance with the legal dose rate limits.

  12. Evaluation using Monte Carlo simulations, of the effect of a shielding, called external shielding, for fotoneutrons generated in linear accelerators, using the computational model of Varian accelerator 2300 C/D operating in eight rotation angles of the GA

    Silva, Hugo R.; Silva, Ademir X.; Rebello, Wilson F.; Silva, Maria G.

    2011-01-01

    This paper aims to present the results obtained by Monte Carlo simulation of the effect of shielding against neutrons, called External Shielding, to be placed on the heads of linear accelerators used in radiotherapy. For this, it was used the radiation transport code Monte Carlo N-Particle - MCNPX, in which were developed computational model of the head of the linear accelerator Varian 2300 C/D. The equipment was simulated within a bunker, operating at energies of 10, 15 and 18 MV, considering the rotation of the gantry at eight different angles ( 0 deg, 45 deg, 90 deg, 135 deg, 180 deg, 225 deg, 270 deg and 315 deg), in all cases, the equipment was modeled without and with the shielding positioned attached to the head of the accelerator on its bottom. In each of these settings, it was calculated the Ambient Dose Equivalent due to neutron H * (10)n on points situated in the region of the patient (region of interest for evaluation of undesirable neutron doses on the patient) and in the maze of radiotherapy room (region of interest for shielding the access door to the bunker). It was observed for all energies of equipment operation as well as for all angles of inclination of the gantry, a significant reduction in the values of H * (10) n when the equipment operated with the external shielding, both in the region of the patient as in the region of the maze. (author)

  13. Simulation of scanning geometry for Shadow Shield counter using Monte Carlo method

    Deepu, R.; Manohari, M.; Mathiyarasu, R.

    2018-01-01

    The whole body counting facility at RSD, IGCAR employs a Shadow Shield Counter (SSC) for the assessment of internal exposure of radiation workers from fission and activation products. The SSC system is routinely calibrated using an in-house built Masonite cut sheet phantom, loaded with standard point sources, in scanning mode. The system is capable of measuring gamma energy from 200 keV to 3 MeV. The standard sources available have energies in the range of 300 - 1500 keV. In order to extend the energy range, numerical simulation can be carried out. As, numerical methods cannot simulate scanning mode directly, a novel scheme is attempted to generate the efficiency value for scanning mode through the use of multiple static mode. The same method was also verified through measurement. The efficiency values of the simulation as well two measurements techniques were compared

  14. Monte Carlo simulations and measurements for efficiency determination of lead shielded plastic scintillator detectors

    Yasin, Zafar; Negoita, Florin; Tabbassum, Sana; Borcea, Ruxandra; Kisyov, Stanimir

    2017-12-01

    The plastic scintillators are used in different areas of science and technology. One of the use of these scintillator detectors is as beam loss monitors (BLM) for new generation of high intensity heavy ion in superconducting linear accelerators. Operated in pulse counting mode with rather high thresholds and shielded by few centimeters of lead in order to cope with radiofrequency noise and X-ray background emitted by accelerator cavities, they preserve high efficiency for high energy gamma ray and neutrons produced in the nuclear reactions of lost beam particles with accelerator components. Efficiency calculation and calibration of detectors is very important before their practical usage. In the present work, the efficiency of plastic scintillator detectors is simulated using FLUKA for different gamma and neutron sources like, 60Co, 137Cs and 238Pu-Be. The sources are placed at different positions around the detector. Calculated values are compared with the measured values and a reasonable agreement is observed.

  15. Monte-Carlo calculation of irradiation dose content beyond shielding of high-energy accelerators

    Mokhov, N.V.; Frolov, V.V.

    1975-01-01

    The MARS programme, designed for calculating the three-dimensional internuclear cascade in defence of the accelerators by the Monte Carlo method, is described. The methods used to reduce the dispersion and the system of semi-empirical formulas made it possible to exceed the parameters of the existing programmes. By means of a synthesis of the results, registered by MARS and HAMLET programmes, the dosage fields for homogeneous and heterogeneous defence were evaluated. The results of the calculated absorbed and equivalent dose behind the barrier, irradiated by a proton beam, having the energy of Esub(o)=1/1000 GeV are exposed. The dependence of the high- and low-energy neutron, proton, pion, kaon, muonium and γ-quantum dosage on the initial energy and thickness, on the material and the composition of the defence is investigated

  16. Evaluation of backscatter dose from internal lead shielding in clinical electron beams using EGSnrc Monte Carlo simulations.

    De Vries, Rowen J; Marsh, Steven

    2015-11-08

    Internal lead shielding is utilized during superficial electron beam treatments of the head and neck, such as lip carcinoma. Methods for predicting backscattered dose include the use of empirical equations or performing physical measurements. The accuracy of these empirical equations required verification for the local electron beams. In this study, a Monte Carlo model of a Siemens Artiste linac was developed for 6, 9, 12, and 15 MeV electron beams using the EGSnrc MC package. The model was verified against physical measurements to an accuracy of better than 2% and 2mm. Multiple MC simulations of lead interfaces at different depths, corresponding to mean electron energies in the range of 0.2-14 MeV at the interfaces, were performed to calculate electron backscatter values. The simulated electron backscatter was compared with current empirical equations to ascertain their accuracy. The major finding was that the current set of backscatter equations does not accurately predict electron backscatter, particularly in the lower energies region. A new equation was derived which enables estimation of electron backscatter factor at any depth upstream from the interface for the local treatment machines. The derived equation agreed to within 1.5% of the MC simulated electron backscatter at the lead interface and upstream positions. Verification of the equation was performed by comparing to measurements of the electron backscatter factor using Gafchromic EBT2 film. These results show a mean value of 0.997 ± 0.022 to 1σ of the predicted values of electron backscatter. The new empirical equation presented can accurately estimate electron backscatter factor from lead shielding in the range of 0.2 to 14 MeV for the local linacs.

  17. Neutron shielding calculations in a proton therapy facility based on Monte Carlo simulations and analytical models: Criterion for selecting the method of choice

    Titt, U.; Newhauser, W. D.

    2005-01-01

    Proton therapy facilities are shielded to limit the amount of secondary radiation to which patients, occupational workers and members of the general public are exposed. The most commonly applied shielding design methods for proton therapy facilities comprise semi-empirical and analytical methods to estimate the neutron dose equivalent. This study compares the results of these methods with a detailed simulation of a proton therapy facility by using the Monte Carlo technique. A comparison of neutron dose equivalent values predicted by the various methods reveals the superior accuracy of the Monte Carlo predictions in locations where the calculations converge. However, the reliability of the overall shielding design increases if simulation results, for which solutions have not converged, e.g. owing to too few particle histories, can be excluded, and deterministic models are being used at these locations. Criteria to accept or reject Monte Carlo calculations in such complex structures are not well understood. An optimum rejection criterion would allow all converging solutions of Monte Carlo simulation to be taken into account, and reject all solutions with uncertainties larger than the design safety margins. In this study, the optimum rejection criterion of 10% was found. The mean ratio was 26, 62% of all receptor locations showed a ratio between 0.9 and 10, and 92% were between 1 and 100. (authors)

  18. Shielding calculations for neutron calibration bunker using Monte Carlo code MCNP-4C

    Suman, H.; Kharita, M. H.; Yousef, S.

    2008-02-01

    In this work, the dose arising from an Am-Be source of 10 8 neutron/sec strength located inside the newly constructed neutron calibration bunker in the National Radiation Metrology Laboratories, was calculated using MCNP-4C code. It was found that the shielding of the neutron calibration bunker is sufficient. As the calculated dose is not expected to exceed in inhabited areas 0.183 μSv/hr, which is 10 times smaller than the regulatory dose constraints. Hence, it can be concluded that the calibration bunker can house - from the external exposure point of view - an Am-Be neutron source of 10 9 neutron/sec strength. It turned out that the neutron dose from the source is few times greater than the photon dose. The sky shine was found to contribute significantly to the total dose. This contribution was estimated to be 60% of the neutron dose and 10% of the photon dose. The systematic uncertainties due to various factors have been assessed and was found to be between 4 and 10% due to concrete density variations; 15% due to the dose estimation method; 4 -10% due to weather variations (temperature and moisture). The calculated dose was highly sensitive to the changes in source spectra. The uncertainty due to the use of two different neutron spectra is about 70%.(author)

  19. Iteration of adjoint equations

    Lewins, J.D.

    1994-01-01

    Adjoint functions are the basis of variational methods and now widely used for perturbation theory and its extension to higher order theory as used, for example, in modelling fuel burnup and optimization. In such models, the adjoint equation is to be solved in a critical system with an adjoint source distribution that is not zero but has special properties related to ratios of interest in critical systems. Consequently the methods of solving equations by iteration and accumulation are reviewed to show how conventional methods may be utilized in these circumstances with adequate accuracy. (author). 3 refs., 6 figs., 3 tabs

  20. Adjoint entropy vs topological entropy

    Giordano Bruno, Anna

    2012-01-01

    Recently the adjoint algebraic entropy of endomorphisms of abelian groups was introduced and studied. We generalize the notion of adjoint entropy to continuous endomorphisms of topological abelian groups. Indeed, the adjoint algebraic entropy is defined using the family of all finite-index subgroups, while we take only the subfamily of all open finite-index subgroups to define the topological adjoint entropy. This allows us to compare the (topological) adjoint entropy with the known topologic...

  1. Shielding calculations for industrial 5/7.5MeV electron accelerators using the MCNP Monte Carlo Code

    Peri, Eyal; Orion, Itzhak

    2017-09-01

    High energy X-rays from accelerators are used to irradiate food ingredients to prevent growth and development of unwanted biological organisms in food, and by that extend the shelf life of the products. The production of X-rays is done by accelerating 5 MeV electrons and bombarding them into a heavy target (high Z). Since 2004, the FDA has approved using 7.5 MeV energy, providing higher production rates with lower treatments costs. In this study we calculated all the essential data needed for a straightforward concrete shielding design of typical food accelerator rooms. The following evaluation is done using the MCNP Monte Carlo code system: (1) Angular dependence (0-180°) of photon dose rate for 5 MeV and 7.5 MeV electron beams bombarding iron, aluminum, gold, tantalum, and tungsten targets. (2) Angular dependence (0-180°) spectral distribution simulations of bremsstrahlung for gold, tantalum, and tungsten bombarded by 5 MeV and 7.5 MeV electron beams. (3) Concrete attenuation calculations in several photon emission angles for the 5 MeV and 7.5 MeV electron beams bombarding a tantalum target. Based on the simulation, we calculated the expected increase in dose rate for facilities intending to increase the energy from 5 MeV to 7.5 MeV, and the concrete width needed to be added in order to keep the existing dose rate unchanged.

  2. FIFRELIN - TRIPOLI-4® coupling for Monte Carlo simulations with a fission model. Application to shielding calculations

    Petit, Odile; Jouanne, Cédric; Litaize, Olivier; Serot, Olivier; Chebboubi, Abdelhazize; Pénéliau, Yannick

    2017-09-01

    TRIPOLI-4® Monte Carlo transport code and FIFRELIN fission model have been coupled by means of external files so that neutron transport can take into account fission distributions (multiplicities and spectra) that are not averaged, as is the case when using evaluated nuclear data libraries. Spectral effects on responses in shielding configurations with fission sampling are then expected. In the present paper, the principle of this coupling is detailed and a comparison between TRIPOLI-4® fission distributions at the emission of fission neutrons is presented when using JEFF-3.1.1 evaluated data or FIFRELIN data generated either through a n/g-uncoupled mode or through a n/g-coupled mode. Finally, an application to a modified version of the ASPIS benchmark is performed and the impact of using FIFRELIN data on neutron transport is analyzed. Differences noticed on average reaction rates on the surfaces closest to the fission source are mainly due to the average prompt fission spectrum. Moreover, when working with the same average spectrum, a complementary analysis based on non-average reaction rates still shows significant differences that point out the real impact of using a fission model in neutron transport simulations.

  3. Preliminary shielding analysis of VHTR reactors

    Flaspoehler, Timothy M.; Petrovic, Bojan

    2011-01-01

    Over the last 20 years a number of methods have been established for automated variance reduction in Monte Carlo shielding simulations. Hybrid methods rely on deterministic adjoint and/or forward calculations to generate these parameters. In the present study, we use the FWCADIS method implemented in MAVRIC sequence of the SCALE6 package to perform preliminary shielding analyses of a VHTR reactor. MAVRIC has been successfully used by a number of researchers for a range of shielding applications, including modeling of LWRs, spent fuel storage, radiation field throughout a nuclear power plant, study of irradiation facilities, and others. However, experience in using MAVRIC for shielding studies of VHTRs is more limited. Thus, the objective of this work is to contribute toward validating MAVRIC for such applications, and identify areas for potential improvement. A simplified model of a prismatic VHTR has been devised, based on general features of the 600 MWt reactor considered as one of the NGNP options. Fuel elements have been homogenized, and the core region is represented as an annulus. However, the overall mix of materials and the relatively large dimensions of the spatial domain challenging the shielding simulations have been preserved. Simulations are performed to evaluate fast neutron fluence, dpa, and other parameters of interest at relevant positions. The paper will investigate and discuss both the effectiveness of the automated variance reduction, as well as applicability of physics model from the standpoint of specific VHTR features. (author)

  4. Analysis of the ITER computational shielding benchmark with the Monte Carlo TRIPOLI-4{sup ®} neutron gamma coupled calculations

    Lee, Yi-Kang, E-mail: yi-kang.lee@cea.fr

    2016-11-01

    Highlights: • Verification and validation of TRIPOLI-4 radiation transport calculations for ITER shielding benchmark. • Evaluation of CEA-V5.1.1 and FENDL-3.0 nuclear data libraries on D–T fusion neutron continuous energy transport calculations. • Advances in nuclear analyses for nuclear heating and radiation damage in iron. • This work also demonstrates that the “safety factors” concept is necessary in the nuclear analyses of ITER. - Abstract: With the growing interest in using the continuous-energy TRIPOLI-4{sup ®} Monte Carlo radiation transport code for ITER applications, a key issue that arises is whether or not the released TRIPOLI-4 code and its associated nuclear data libraries are verified and validated for the D–T fusion neutronics calculations. Previous published benchmark results of TRIPOLI-4 code on the ITER related activities have concentrated on the first wall loading, the reactor dosimetry, the nuclear heating, and the tritium breeding ratio. To enhance the TRIPOLI-4 verification and validation on neutron-gamma coupled calculations for fusion device application, the computational ITER shielding benchmark of M. E. Sawan was performed in this work by using the 2013 released TRIPOLI-4.9S code and the associated CEA-V5.1.1 data library. First wall, blanket, vacuum vessel and toroidal field magnet of the inboard and outboard components were fully modelled in this 1-D toroidal cylindrical benchmark. The 14.1 MeV source neutrons were sampled from a uniform isotropic distribution in the plasma zone. Nuclear responses including neutron and gamma fluxes, nuclear heating, and material damage indicator were benchmarked against previous published results. The capabilities of the TRIPOLI-4 code on the evaluation of above physics parameters were presented. The nuclear data library from the new FENDL-3.0 evaluation was also benchmarked against the CEA-V5.1.1 results for the neutron transport calculations. The results show that both data libraries

  5. Analysis of the ITER computational shielding benchmark with the Monte Carlo TRIPOLI-4® neutron gamma coupled calculations

    Lee, Yi-Kang

    2016-01-01

    Highlights: • Verification and validation of TRIPOLI-4 radiation transport calculations for ITER shielding benchmark. • Evaluation of CEA-V5.1.1 and FENDL-3.0 nuclear data libraries on D–T fusion neutron continuous energy transport calculations. • Advances in nuclear analyses for nuclear heating and radiation damage in iron. • This work also demonstrates that the “safety factors” concept is necessary in the nuclear analyses of ITER. - Abstract: With the growing interest in using the continuous-energy TRIPOLI-4 ® Monte Carlo radiation transport code for ITER applications, a key issue that arises is whether or not the released TRIPOLI-4 code and its associated nuclear data libraries are verified and validated for the D–T fusion neutronics calculations. Previous published benchmark results of TRIPOLI-4 code on the ITER related activities have concentrated on the first wall loading, the reactor dosimetry, the nuclear heating, and the tritium breeding ratio. To enhance the TRIPOLI-4 verification and validation on neutron-gamma coupled calculations for fusion device application, the computational ITER shielding benchmark of M. E. Sawan was performed in this work by using the 2013 released TRIPOLI-4.9S code and the associated CEA-V5.1.1 data library. First wall, blanket, vacuum vessel and toroidal field magnet of the inboard and outboard components were fully modelled in this 1-D toroidal cylindrical benchmark. The 14.1 MeV source neutrons were sampled from a uniform isotropic distribution in the plasma zone. Nuclear responses including neutron and gamma fluxes, nuclear heating, and material damage indicator were benchmarked against previous published results. The capabilities of the TRIPOLI-4 code on the evaluation of above physics parameters were presented. The nuclear data library from the new FENDL-3.0 evaluation was also benchmarked against the CEA-V5.1.1 results for the neutron transport calculations. The results show that both data libraries can be

  6. A three-dimensional computed tomography-assisted Monte Carlo evaluation of ovoid shielding on the dose to the bladder and rectum in intracavitary radiotherapy for cervical cancer

    Gifford, Kent A.; Horton, John L.; Pelloski, Christopher E.; Jhingran, Anuja; Court, Laurence E.; Mourtada, Firas; Eifel, Patricia J.

    2005-01-01

    Purpose: To determine the effects of Fletcher Suit Delclos ovoid shielding on dose to the bladder and rectum during intracavitary radiotherapy for cervical cancer. Methods and Materials: The Monte Carlo method was used to calculate the dose in 12 patients receiving low-dose-rate intracavitary radiotherapy with both shielded and unshielded ovoids. Cumulative dose-difference surface histograms were computed for the bladder and rectum. Doses to the 2-cm 3 and 5-cm 3 volumes of highest dose were computed for the bladder and rectum with and without shielding. Results: Shielding affected dose to the 2-cm 3 and 5-cm 3 volumes of highest dose for the rectum (10.1% and 11.1% differences, respectively). Shielding did not have a major impact on the dose to the 2-cm 3 and 5-cm 3 volumes of highest dose for the bladder. The average dose reduction to 5% of the surface area of the bladder was 53 cGy. Reductions as large as 150 cGy were observed to 5% of the surface area of the bladder. The average dose reduction to 5% of the surface area of the rectum was 195 cGy. Reductions as large as 405 cGy were observed to 5% of the surface area of the rectum. Conclusions: Our data suggest that the ovoid shields can greatly reduce the radiation dose delivered to the rectum. We did not find the same degree of effect on the dose to the bladder. To calculate the dose accurately, however, the ovoid shields must be included in the dose model

  7. The adjoint variational nodal method

    Laurin-Kovitz, K.; Lewis, E.E.

    1993-01-01

    The widespread use of nodal methods for reactor core calculations in both diffusion and transport approximations has created a demand for the corresponding adjoint solutions as a prerequisite for performing perturbation calculations. With some computational methods, however, the solution of the adjoint problem presents a difficulty; the physical adjoint obtained by discretizing the adjoint equation is not the same as the mathematical adjoint obtained by taking the transpose of the coefficient matrix, which results from the discretization of the forward equation. This difficulty arises, in particular, when interface current nodal methods based on quasi-one-dimensional solution of the diffusion or transport equation are employed. The mathematical adjoint is needed to perform perturbation calculations. The utilization of existing nodal computational algorithms, however, requires the physical adjoint. As a result, similarity transforms or related techniques must be utilized to relate physical and mathematical adjoints. Thus far, such techniques have been developed only for diffusion theory

  8. Monte carlo calculation of the neutron effective dose rate at the outer surface of the biological shield of HTR-10 reactor

    Remetti, Romolo; Andreoli, Giulio; Keshishian, Silvina

    2012-01-01

    Highlights: ► We deal with HTR-10, that is a helium-cooled graphite-moderated pebble bed reactor. ► We carried out Monte Carlo simulation of the core by MCNP5. ► Extensive use of MCNP5 variance reduction methods has been done. ► We calculated the trend of neutron flux within the biological shield. ► We calculated neutron effective dose at the outer surface of biological shield. - Abstract: Research on experimental reactors, such as HTR-10, provide useful data about potentialities of very high temperature gas-cooled reactors (VHTR). The latter is today rated as one of the six nuclear reactor types involved in the Generation-IV International Forum (GIF) Initiative. In this study, the MCNP5 code has been employed to evaluate the neutron radiation trend vs. the biological shield's thickness and to calculate the neutron effective dose rate at the outer surface. The reactor's geometry has been completely modeled by means of lattices and universes provided by MCNP, even though some approximations were required. Monte Carlo calculations have been performed by means of a simple PC and, as a consequence, in order to obtain acceptable run times, it was made an extensive recourse to variance reduction methods.

  9. Self-shielding factors

    Kaul, D.C.

    1982-01-01

    Throughout the last two decades many efforts have been made to estimate the effect of body self-shielding on organ doses from externally incident neutrons and gamma rays. These began with the use of simple geometry phantoms and have culminated in the use of detailed anthropomorphic phantoms. In a recent effort, adjoint Monte Carlo analysis techniques have been used to determine dose and dose equivalent to the active marrow as a function of energy and angle of neutron fluence externally incident on an anthropomorphic phantom. When combined with fluences from actual nuclear devices, these dose-to-fluence factors result in marrow dose values that demonstrate great sensitivity to variations in device type, range, and body orientation. Under a state-of-the-art radiation transport analysis demonstration program for the Japanese cities, sponsored by the Defense Nuclear Agency at the request of the National Council on Radiation Protection and Measurements, the marrow dose study referred to above is being repeated to obtain spectral distributions within the marrow for externally incident neutrons and gamma rays of arbitrary energy and angle. This is intended to allow radiobiologists and epidemiologists to select and to modify numbers of merit for correlation with health effects and to permit a greater understanding of the relationship between human and laboratory subject dosimetry

  10. Study on shielding design method of radiation streaming in a tokamak-type DT fusion reactor based on Monte Carlo calculation

    Sato, Satoshi

    2003-09-01

    In tokamak-type DT nuclear fusion reactor, there are various type slits and ducts in the blanket and the vacuum vessel. The helium production in the rewelding location of the blanket and the vacuum vessel, the nuclear properties in the super-conductive TF coil, e.g. the nuclear heating rate in the coil winding pack, are enhanced by the radiation streaming through the slits and ducts, and they are critical concern in the shielding design. The decay gamma ray dose rate around the duct penetrating the blanket and the vacuum vessel is also enhanced by the radiation streaming through the duct, and they are also critical concern from the view point of the human access to the cryostat during maintenance. In order to evaluate these nuclear properties with good accuracy, three dimensional Monte Carlo calculation is required but requires long calculation time. Therefore, the development of the effective simple design evaluation method for radiation streaming is substantially important. This study aims to establish the systematic evaluation method for the nuclear properties of the blanket, the vacuum vessel and the Toroidal Field (TF) coil taking into account the radiation streaming through various types of slits and ducts, based on three dimensional Monte Carlo calculation using the MNCP code, and for the decay gamma ray dose rates penetrated around the ducts. The present thesis describes three topics in five chapters as follows; 1) In Chapter 2, the results calculated by the Monte Carlo code, MCNP, are compared with those by the Sn code, DOT3.5, for the radiation streaming in the tokamak-type nuclear fusion reactor, for validating the results of the Sn calculation. From this comparison, the uncertainties of the Sn calculation results coming from the ray-effect and the effect due to approximation of the geometry are investigated whether the two dimensional Sn calculation can be applied instead of the Monte Carlo calculation. Through the study, it can be concluded that the

  11. Monte Carlo computation of Bremsstrahlung intensity and energy spectrum from a 15 MV linear electron accelerator tungsten target to optimise LINAC head shielding

    Biju, K.; Sharma, Amiya; Yadav, R.K.; Kannan, R.; Bhatt, B.C.

    2003-01-01

    The knowledge of exact photon intensity and energy distributions from the target of an electron target is necessary while designing the shielding for the accelerator head from radiation safety point of view. The computations were carried out for the intensity and energy distribution of photon spectrum from a 0.4 cm thick tungsten target in different angular directions for 15 MeV electrons using a validated Monte Carlo code MCNP4A. Similar results were computed for 30 MeV electrons and found agreeing with the data available in literature. These graphs and the TVT values in lead help to suggest an optimum shielding thickness for 15 MV Linac head. (author)

  12. Shielding benchmark problems

    Tanaka, Shun-ichi; Sasamoto, Nobuo; Oka, Yoshiaki; Kawai, Masayoshi; Nakazawa, Masaharu.

    1978-09-01

    Shielding benchmark problems were prepared by the Working Group of Assessment of Shielding Experiments in the Research Comittee on Shielding Design of the Atomic Energy Society of Japan, and compiled by the Shielding Laboratory of Japan Atomic Energy Research Institute. Twenty-one kinds of shielding benchmark problems are presented for evaluating the calculational algorithm and the accuracy of computer codes based on the discrete ordinates method and the Monte Carlo method and for evaluating the nuclear data used in the codes. (author)

  13. MORSE/STORM: A generalized albedo option for Monte Carlo calculations

    Gomes, I.C.; Stevens, P.N.

    1991-09-01

    The advisability of using the albedo procedure for the Monte Carlo solution of deep penetration shielding problems that have ducts and other penetrations has been investigated. The use of albedo data can dramatically improve the computational efficiency of certain Monte Carlo calculations. However, the accuracy of these results may be unacceptable because of lost information during the albedo event and serious errors in the available differential albedo data. This study was done to evaluate and appropriately modify the MORSE/BREESE package, to develop new methods for generating the required albedo data, and to extend the adjoint capability to the albedo-modified calculations. Major modifications to MORSE/BREESE include an option to save for further use information that would be lost at the albedo event, an option to displace the point of emergence during an albedo event, and an option to use spatially dependent albedo data for both forward and adjoint calculations, which includes the point of emergence as a new random variable to be selected during an albedo event. The theoretical basis for using TORT-generated forward albedo information to produce adjuncton albedos was derived. The MORSE/STORM package was developed to perform both forward and adjoint modes of analysis using spatially dependent albedo data. Results obtained with MORSE/STORM for both forward and adjoint modes were compared with benchmark solutions. Excellent agreement and improved computational efficiency were achieved, demonstrating the full utilization of the albedo option in the MORSE code. 7 refs., 17 figs., 15 tabs

  14. Shielding performances analysis for the IFMIF test facility based on high-fidelity Monte Carlo neutronic calculations

    Kondo, Keitaro, E-mail: kondo.keitaro@jaea.go.jp; Arbeiter, Frederik; Fischer, Ulrich; Lu, Lei; Qiu, Yuefeng; Tian, Kuo

    2015-10-15

    Highlights: • A detailed geometry model with pipe penetrations and gaps was prepared for the IFMIF test cell. • The neutron streaming effect due to gaps and pipes with shielding plugs was investigated. • The present analysis revealed that the streaming effect can be mitigated by some counter measures. • Occupational workers can access to the room above the test cell during operation. - Abstract: The IFMIF Test Cell (TC) design was developed and optimized in the EVEDA phase, and finally the reference TC design was proposed. The present study is devoted to further investigations of open issues on the reference TC design. In order to examine the neutron streaming effect caused by pipe penetrations and gaps around removable shielding plugs, a new geometry model for neutronic analyses has been prepared directly from engineering CAD data by utilizing the McCad conversion software. All removable shielding plugs are separately described in the model and a detailed description of pipes was incorporated into the model. The calculation result suggests that the streaming effect is mitigated if the pipe penetration is designed appropriately, while the gaps around the shielding plugs above the TC have large impact on the radiation dose in the access cell. The concept of the reference TC design has been basically validated from the neutronics point of view, although the streaming effect should be compensated by the shielding capability of the test cell cover plate so that occupational workers can access to the access cell during operation.

  15. Shielding performances analysis for the IFMIF test facility based on high-fidelity Monte Carlo neutronic calculations

    Kondo, Keitaro; Arbeiter, Frederik; Fischer, Ulrich; Lu, Lei; Qiu, Yuefeng; Tian, Kuo

    2015-01-01

    Highlights: • A detailed geometry model with pipe penetrations and gaps was prepared for the IFMIF test cell. • The neutron streaming effect due to gaps and pipes with shielding plugs was investigated. • The present analysis revealed that the streaming effect can be mitigated by some counter measures. • Occupational workers can access to the room above the test cell during operation. - Abstract: The IFMIF Test Cell (TC) design was developed and optimized in the EVEDA phase, and finally the reference TC design was proposed. The present study is devoted to further investigations of open issues on the reference TC design. In order to examine the neutron streaming effect caused by pipe penetrations and gaps around removable shielding plugs, a new geometry model for neutronic analyses has been prepared directly from engineering CAD data by utilizing the McCad conversion software. All removable shielding plugs are separately described in the model and a detailed description of pipes was incorporated into the model. The calculation result suggests that the streaming effect is mitigated if the pipe penetration is designed appropriately, while the gaps around the shielding plugs above the TC have large impact on the radiation dose in the access cell. The concept of the reference TC design has been basically validated from the neutronics point of view, although the streaming effect should be compensated by the shielding capability of the test cell cover plate so that occupational workers can access to the access cell during operation.

  16. Implementation and display of Computer Aided Design (CAD) models in Monte Carlo radiation transport and shielding applications

    Burns, T.J.

    1994-01-01

    An Xwindow application capable of importing geometric information directly from two Computer Aided Design (CAD) based formats for use in radiation transport and shielding analyses is being developed at ORNL. The application permits the user to graphically view the geometric models imported from the two formats for verification and debugging. Previous models, specifically formatted for the radiation transport and shielding codes can also be imported. Required extensions to the existing combinatorial geometry analysis routines are discussed. Examples illustrating the various options and features which will be implemented in the application are presented. The use of the application as a visualization tool for the output of the radiation transport codes is also discussed

  17. Shielding Calculations for Industrial 5/7.5MeV Electron Accelerators Using the MCNP Monte Carlo Code

    Peri, E.; Orion, I.

    2014-01-01

    High energy X-rays from accelerators are used to irradiate food ingredients to prevent growth and development of unwanted biological organisms in food, in order to extend the shelf life of products. High energy photons can cause food activation due to (D 3 ,n) reactions. Until 2004, to eliminate the possibility of food activation, the electron energy was limited to 5 MeV X-rays for food irradiation. In 2004, the FDA approved the usage of up to 7.5 MeV, but only with tantalum and gold targets (1). Higher X-ray energy results an increased flux of X-rays in the forward direction, increased penetration, and higher photon dose rate due to better electron-to-photon conversion. These improvements could decrease the irradiation time and allow irradiation of larger packages, thereby providing higher production rates with lower treatment cost. Medical accelerators usually work with 6-18 MV electron energy with tungsten target to convert the electron beam to X-rays. In order to protect the patients, the accelerator head is protected with a heavy lead shielding; therefore, the bremsstrahlung is emitted only in the forward direction. There are many publications and standards that guide how to design optimal shielding for medical accelerator rooms. The shielding data for medical accelerators is not applicable for industrial accelerators, since the data is for different conversion targets, different X-Ray energies, and only for the forward direction. Collimators are not always in use in industrial accelerators, and therefore bremsstrahlung photons can be emitted in all directions. The bremsstrahlung spectrum and dose rate change as a function of the emission angle. The dose rate decreases from maximum in the forward direction (0°) to minimum at 180° by 1-2 orders of magnitude. In order to design and calculate optimal shielding for food accelerator rooms, there is a need to have the bremsstrahlung spectrum data, dose rates and concrete attenuation data in all emission directions

  18. Monte Carlo calculations of neutron and gamm-ray energy spectra for fusion-reactor shield design: comparison with experiment

    Santoro, R.T.; Barnes, J.M.

    1983-08-01

    Neutron and gamma-ray spectra resulting from the interactions of approx. 14-MeV neutrons in laminated slabs of stainless steel type-304 and borated polyethylene have been calculated using the Monte Carlo code MCNP. The calculated spectra are compared with measured data as a function of slab thickness and material composition and as a function of detector location behind the slabs. Comparisons of the differential energy spectra are made for neutrons with energies above 850 keV and for gamma rays with energies above 750 keV. The measured neutron spectra and those calculated using Monte Carlo methods agree within 5% to 50% depending on the slab thickness and composition and neutron energy. The agreement between the measured and calculated gamma-ray energy spectra is also within this range. The MCNP data are also in favorable agreement with attenuated data calculated previously by discrete ordinates transport methods and the Monte Carlo code SAM-CE

  19. Evaluation of the shield calculation adequacy of radiotherapy rooms through Monte Carlo Method and experimental measures; Avaliacao da adequacao do calculo de blindagens de salas de radioterapia atraves do metodo de Monte Carlos e medidas experimentais

    Meireles, Ramiro Conceicao

    2016-07-01

    The shielding calculation methodology for radiotherapy services adopted in Brazil and in several countries is that described in publication 151 of the National Council on Radiation Protection and Measurements (NCRP 151). This methodology however, markedly employs several approaches that can impact both in the construction cost and in the radiological safety of the facility. Although this methodology is currently well established by the high level of use, some parameters employed in the calculation methodology did not undergo to a detailed assessment to evaluate the impact of the various approaches considered. In this work the MCNP5 Monte Carlo code was used with the purpose of evaluating the above mentioned approaches. TVLs values were obtained for photons in conventional concrete (2.35g / cm{sup 3}), considering the energies of 6, 10 and 25 MeV, respectively, first considering an isotropic radiation source impinging perpendicular to the barriers, and subsequently a lead head shielding emitting a shaped beam, in the format of a pyramid trunk. Primary barriers safety margins, taking in account the head shielding emitting photon beam pyramid-shaped in the energies of 6, 10, 15 and 18 MeV were assessed. A study was conducted considering the attenuation provided by the patient's body in the energies of 6,10, 15 and 18 MeV, leading to new attenuation factors. Experimental measurements were performed in a real radiotherapy room, in order to map the leakage radiation emitted by the accelerator head shielding and the results obtained were employed in the Monte Carlo simulation, as well as to validate the entire study. The study results indicate that the TVLs values provided by (NCRP, 2005) show discrepancies in comparison with the values obtained by simulation and that there may be some barriers that are calculated with insufficient thickness. Furthermore, the simulation results show that the additional safety margins considered when calculating the width of the

  20. Development of a Matlab/Simulink tool to facilitate system analysis and simulation via the adjoint and covariance methods

    Bucco, D.; Weiss, M.

    2007-01-01

    The COVariance and ADjoint Analysis Tool (COVAD) is a specially designed software tool, written for the Matlab/Simulink environment, which allows the user the capability to carry out system analysis and simulation using the adjoint, covariance or Monte Carlo methods. This paper describes phase one

  1. Monte Carlo simulation of x-ray buildup factors of lead and its applications in shielding of diagnostic x-ray facilities

    Kharrati, Hedi; Agrebi, Amel; Karaoui, Mohamed-Karim

    2007-01-01

    X-ray buildup factors of lead in broad beam geometry for energies from 15 to 150 keV are determined using the general purpose Monte Carlo N-particle radiation transport computer code (MCNP4C). The obtained buildup factors data are fitted to a modified three parameter Archer et al. model for ease in calculating the broad beam transmission with computer at any tube potentials/filters combinations in diagnostic energies range. An example for their use to compute the broad beam transmission at 70, 100, 120, and 140 kVp is given. The calculated broad beam transmission is compared to data derived from literature, presenting good agreement. Therefore, the combination of the buildup factors data as determined and a mathematical model to generate x-ray spectra provide a computationally based solution to broad beam transmission for lead barriers in shielding x-ray facilities

  2. Neutron Skyshine in shielding projects of radiotherapy: comparison between theoretical approach and simulation by Monte Carlo method; 'Skyshine' de neutrons em projetos de blindagens de radioterapia: comparacao entre abordagem teorica e simulacao por metodo de Monte Carlo

    Falcao, R.C.; Facure, A. [Comissao Nacional de Energia Nuclear (CNEN), Rio de Janeiro, RJ (Brazil); Santini, E.S. [Comissao Nacional de Energia Nuclear (CNEN), Rio de Janeiro, RJ (Brazil); Centro Brasileiro de Pesquisas Fisicas (CBPF), Rio de Janeiro, RJ (Brazil); Silva, A.X. [Coordenacao dos Programas de Pos-Graduacao de Engenharia (PEN/COPPE/UFRJ), RJ (Brazil). Programa de Engenharia Nuclear

    2005-07-01

    In this work, the MCNP code is used to simulate the transport of neutrons in a room of radiotherapy, whose shieldings are designed according to the method of skyshine (scattering in the atmosphere). The simulations are compared with the results obtained from empirically established expressions, which are normally used for designing the ceilings of the rooms facilities, ensuring that dose rates (neutrons + photons) around them do not exceed the maximum limits allowed by the standards of the CNEN. Good agreement is observed between the doses calculated according to these expressions and those obtained through simulation by Monte Carlo in the case of rooms without ceiling, and an overestimate of the calculations by a factor 2 or 3 in relation to the simulations, in the case of rooms with ceiling.

  3. Solving the transport equation by Monte Carlo method and application for biological shield calculations; Resavanje transportne jednacine Monte Carlo metodom i njena primena u zadacima proracuna bioloskog stita

    Kocic, A [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1977-07-01

    General sampling Monte Carlo scheme for neutron transport equation has been described. Programme TRANSFER for neutron beam transmission analysis has been used to calculate the neutron leakage spectrum, detector efficiency and neutron angular distribution of the example problem (author) [Serbo-Croat] U radu se najpre razmatraju osnovni problemi resavanja transportne jednacine i nacin kako Monte Karlo metoda omogucuje da se prevazidju neki od njih: visedimenzionalnost zadatka, problem dubokog prodiranja i dovoljno fino tretiranje efikasnih preseka. Dalje, govori se o iskustvima sa primenom Monte Karlo metode u Laboratoriji za nuklearnu energetiku i tehnicku fiziku i o primeni ove metode na probleme zastite. Na kraju dati su i analizirani ilustrativni primeri proracuna transporta neutrona kroz ravan sloj zastitnog materijala koriscenjem Monte Karlo programa TRANSFER (author)

  4. Monte Carlo analysis of the effects of penetrations on the performance of a tokamak fusion reactor

    Santoro, R.T.; Tang, J.S.; Alsmiller, R.G. Jr.; Barnes, J.M.

    1977-01-01

    Adjoint Monte Carlo calculations have been carried out to estimate the nuclear heating and radiation damage in the toroidal field (TF) coils adjacent to a 28 x 68 cm 2 rectangular neutral beam injector duct that passes through the blanket and shield of a D-T burning Tokamak reactor. The plasma region, blanket, shield, and TF coils were represented in cylindrical geometry using the same dimensions and compositions as those of the Experimental Power Reactor. The radiation transport was accomplished using coupled 35-group neutron, 21-group gamma-ray cross sections and the nuclear heating and radiation damage were obtained using the latest available response functions. The presence of the neutral beam injector duct leads to increases in the nuclear heating rates in the TF coils ranging from a factor of 3 to a factor of 196 greater than in the fully shielded coils depending on the location. Substantial increases in the radation damage were also noted

  5. Numerical study of dense adjoint 2-color matter

    Hands, S.; Scorzato, L.; Oevers, M.

    2000-11-01

    We study the global symmetries of SU(2) gauge theory with N flavors of staggered fermions in the presence of a chemical potential. We motivate the special interest of the case N=1 (staggered) with fermions in the adjoint representation of the gauge group. We present results from numerical simulations with both hybrid Monte Carlo and the two-step multi-bosonic algorithm. (orig.)

  6. Shielding benchmark problems, (2)

    Tanaka, Shun-ichi; Sasamoto, Nobuo; Oka, Yoshiaki; Shin, Kazuo; Tada, Keiko.

    1980-02-01

    Shielding benchmark problems prepared by Working Group of Assessment of Shielding Experiments in the Research Committee on Shielding Design in the Atomic Energy Society of Japan were compiled by Shielding Laboratory in Japan Atomic Energy Research Institute. Fourteen shielding benchmark problems are presented newly in addition to twenty-one problems proposed already, for evaluating the calculational algorithm and accuracy of computer codes based on discrete ordinates method and Monte Carlo method and for evaluating the nuclear data used in codes. The present benchmark problems are principally for investigating the backscattering and the streaming of neutrons and gamma rays in two- and three-dimensional configurations. (author)

  7. Monte Carlo techniques for analyzing deep-penetration problems

    Cramer, S.N.; Gonnord, J.; Hendricks, J.S.

    1986-01-01

    Current methods and difficulties in Monte Carlo deep-penetration calculations are reviewed, including statistical uncertainty and recent adjoint optimization of splitting, Russian roulette, and exponential transformation biasing. Other aspects of the random walk and estimation processes are covered, including the relatively new DXANG angular biasing technique. Specific items summarized are albedo scattering, Monte Carlo coupling techniques with discrete ordinates and other methods, adjoint solutions, and multigroup Monte Carlo. The topic of code-generated biasing parameters is presented, including the creation of adjoint importance functions from forward calculations. Finally, current and future work in the area of computer learning and artificial intelligence is discussed in connection with Monte Carlo applications

  8. Estimation of the heat generation in vitrified waste product and shield thickness of the cask for the transportation of vitrified waste product using Monte Carlo technique

    Deepa, A.K.; Jakhete, A.P.; Mehta, D.; Kaushik, C.P.

    2011-01-01

    High Level Liquid waste (HLW) generated during reprocessing of spent fuel contains most of the radioactivity present in the spent fuel resulting in the need for isolation and surveillance for extended period of time. Major components in HLW are the corrosion products, fission products such as 137 Cs, 90 Sr, 106 Ru, 144 Ce, 125 Sb etc, actinides and various chemicals used during reprocessing of spent fuel. Fresh HLW having an activity concentration of around 100Ci/l is to be vitrified into borosilicate glass and packed in canisters which are placed in S.S overpacks for better confinement. These overpacks contain around 0.7 Million Curies of activity. Characterisation of activity in HLW and activity profile of radionuclides for various cooling periods sets the base for the study. For transporting the vitrified waste product (VWP), two most important parameters is the shield thickness of the transportation cask and the heat generation in the waste product. This paper describes the methodology used in the estimation of lead thickness for the transportation cask using the Monte Carlo Technique. Heat generation due to decay of fission products results in the increase in temperature of the vitrified waste product during interim storage and disposal. Glass being the material, not having very high thermal conductivity, temperature difference between the canister and surrounding bears significance in view of the possibility of temperature based devitrification of VWP. The heat generation in the canister and the overpack containing vitrified glass is also estimated using MCNP. (author)

  9. Development and validation of continuous energy adjoint-weighted calculations

    Truchet, Guillaume

    2015-01-01

    A key issue in nowadays Reactor Physics is to propagate input data uncertainties (e.g. nuclear data, manufacturing tolerances, etc.) to nuclear codes final results (e.g. k(eff), reaction rate, etc.). In order to propagate uncertainties, one typically assumes small variations around a reference and evaluates at first sensitivity profiles. Problem is that nuclear Monte Carlo codes are not - or were not until very recently - able to straightforwardly process such sensitivity profiles, even thought they are considered as reference codes. First goal of this PhD thesis is to implement a method to calculate k(eff)-sensitivity profiles to nuclear data or any perturbations in TRIPOLI-4, the CEA Monte Carlo neutrons transport code. To achieve such a goal, a method has first been developed to calculate the adjoint flux using the Iterated Fission Probability (IFP) principle that states that the adjoint flux at a given phase space point is proportional to the neutron importance in a just critical core after several power iterations. Thanks to our developments, it has been made possible, for the fist time, to calculate the continuous adjoint flux for an actual and complete reactor core configuration. From that new feature, we have elaborated a new method able to forwardly apply the exact perturbation theory in Monte Carlo codes. Exact perturbation theory does not rely on small variations which makes possible to calculate very complex experiments. Finally and after a deep analysis of the IFP method, this PhD thesis also reproduces and improves an already used method to calculate adjoint weighted kinetic parameters as well as reference migrations areas. (author) [fr

  10. MMW [multimegawatt] shielding design and analysis

    Olson, A.P.

    1988-01-01

    Reactor shielding for multimegawatt (MMW) space power must satisfy a mass constraint as well as performance specifications for neutron fluence and gamma dose. A minimum mass shield is helpful in attaining the launch mass goal for the entire vehicle, because the shield comprises about 1% to 2% of the total vehicle mass. In addition, the shield internal heating must produce tolerable temperatures. The analysis of shield performance for neutrons and gamma rays is emphasized. Topics addressed include cross section preparation for multigroup 2D S/sub n/-transport analyses, and the results of parametric design studies on shadow shield performance and mass versus key shield design variables such as cone angle, number, placement, and thickness of layers of tungsten, and shield top radius. Finally, adjoint methods are applied to the shield in order to spatially map its relative contribution to dose reduction, and to provide insight into further design optimization. 7 refs., 2 figs., 3 tabs

  11. Novel hybrid Monte Carlo/deterministic technique for shutdown dose rate analyses of fusion energy systems

    Ibrahim, Ahmad M.; Peplow, Douglas E.; Peterson, Joshua L.; Grove, Robert E.

    2014-01-01

    Highlights: •Develop the novel Multi-Step CADIS (MS-CADIS) hybrid Monte Carlo/deterministic method for multi-step shielding analyses. •Accurately calculate shutdown dose rates using full-scale Monte Carlo models of fusion energy systems. •Demonstrate the dramatic efficiency improvement of the MS-CADIS method for the rigorous two step calculations of the shutdown dose rate in fusion reactors. -- Abstract: The rigorous 2-step (R2S) computational system uses three-dimensional Monte Carlo transport simulations to calculate the shutdown dose rate (SDDR) in fusion reactors. Accurate full-scale R2S calculations are impractical in fusion reactors because they require calculating space- and energy-dependent neutron fluxes everywhere inside the reactor. The use of global Monte Carlo variance reduction techniques was suggested for accelerating the R2S neutron transport calculation. However, the prohibitive computational costs of these approaches, which increase with the problem size and amount of shielding materials, inhibit their ability to accurately predict the SDDR in fusion energy systems using full-scale modeling of an entire fusion plant. This paper describes a novel hybrid Monte Carlo/deterministic methodology that uses the Consistent Adjoint Driven Importance Sampling (CADIS) method but focuses on multi-step shielding calculations. The Multi-Step CADIS (MS-CADIS) methodology speeds up the R2S neutron Monte Carlo calculation using an importance function that represents the neutron importance to the final SDDR. Using a simplified example, preliminary results showed that the use of MS-CADIS enhanced the efficiency of the neutron Monte Carlo simulation of an SDDR calculation by a factor of 550 compared to standard global variance reduction techniques, and that the efficiency enhancement compared to analog Monte Carlo is higher than a factor of 10,000

  12. Comparison and physical interpretation of MCNP and TART neutron and γ Monte Carlo shielding calculations for a heavy-ion ICF system

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

    2004-01-01

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

  13. Radiation shielding calculation using MCNP

    Masukawa, Fumihiro

    2001-01-01

    To verify the Monte Carlo code MCNP4A as a tool to generate the reference data in the shielding designs and the safety evaluations, various shielding benchmark experiments were analyzed using this code. These experiments were categorized in three types of the shielding subjects; bulk shielding, streaming, and skyshine. For the variance reduction technique, which is indispensable to get meaningful results with the Monte Carlo shielding calculation, we mainly used the weight window, the energy dependent Russian roulette and spitting. As a whole, our analyses performed enough small statistical errors and showed good agreements with these experiments. (author)

  14. Radiation shielding lead shield

    Dei, Shoichi.

    1991-01-01

    The present invention concerns lead shields for radiation shielding. Shield boxes are disposed so as to surround a pipeline through which radioactive liquids, mists or like other objects are passed. Flanges are formed to each of the end edges of the shield boxes and the shield boxes are connected to each other by the flanges. Upon installation, empty shield boxes not charged with lead particles and iron plate shields are secured at first at the periphery of the pipeline. Then, lead particles are charged into the shield boxes. This attains a state as if lead plate corresponding to the depth of the box is disposed. Accordingly, operations for installation, dismantling and restoration can be conducted in an empty state with reduced weight to facilitate the operations. (I.S.)

  15. Approximation for the adjoint neutron spectrum

    Suster, Luis Carlos; Martinez, Aquilino Senra; Silva, Fernando Carvalho da

    2002-01-01

    The proposal of this work is the determination of an analytical approximation which is capable to reproduce the adjoint neutron flux for the energy range of the narrow resonances (NR). In a previous work we developed a method for the calculation of the adjoint spectrum which was calculated from the adjoint neutron balance equations, that were obtained by the collision probabilities method, this method involved a considerable quantity of numerical calculation. In the analytical method some approximations were done, like the multiplication of the escape probability in the fuel by the adjoint flux in the moderator, and after these approximations, taking into account the case of the narrow resonances, were substituted in the adjoint neutron balance equation for the fuel, resulting in an analytical approximation for the adjoint flux. The results obtained in this work were compared to the results generated with the reference method, which demonstrated a good and precise results for the adjoint neutron flux for the narrow resonances. (author)

  16. Weak self-adjoint differential equations

    Gandarias, M L

    2011-01-01

    The concepts of self-adjoint and quasi self-adjoint equations were introduced by Ibragimov (2006 J. Math. Anal. Appl. 318 742-57; 2007 Arch. ALGA 4 55-60). In Ibragimov (2007 J. Math. Anal. Appl. 333 311-28), a general theorem on conservation laws was proved. In this paper, we generalize the concept of self-adjoint and quasi self-adjoint equations by introducing the definition of weak self-adjoint equations. We find a class of weak self-adjoint quasi-linear parabolic equations. The property of a differential equation to be weak self-adjoint is important for constructing conservation laws associated with symmetries of the differential equation. (fast track communication)

  17. SCALE6.1 Hybrid Shielding Methodology For The Spent Fuel Dry Storage

    Matijevic, M.; Pevec, D.; Trontl, K.

    2015-01-01

    The SCALE6.1/MAVRIC hybrid deterministic-stochastic shielding methodology was used for dose rates calculation of the generic spent fuel dry storage installation. The neutron-gamma dose rates around the cask array were calculated over a large problem domain in order to determine the boundary of the controlled area. The FW-CADIS methodology, based on the deterministic forward and adjoint solution over the phase - space, was used for optimized, global Monte Carlo results over the mesh tally. The cask inventory was modeled as homogenized material corresponding to 20 fuel assemblies from a standard mid - sized PWR reactor. The global simulation model was an array of 32 casks in 2 rows with concrete foundations and external air, which makes a large spatial domain for shielding calculations. The dose rates around the casks were determined using FW-CADIS method with weighted adjoint source and mesh tally covering a portion of spatial domain of interest. The conservatively obtained dose rates give the upper boundary, since the activation reduction of sources was not taken into account when sequential filling of the dry storage will start. The effective area of the dry storage installation can be additionally reduced with lowering concrete foundation under the ground, embankment raising, and with extra concrete walls, that would additionally lower the dominant gamma dose rates. (author).

  18. Optimization of a neutron detector design using adjoint transport simulation

    Yi, C.; Manalo, K.; Huang, M.; Chin, M.; Edgar, C.; Applegate, S.; Sjoden, G.

    2012-01-01

    A synthetic aperture approach has been developed and investigated for Special Nuclear Materials (SNM) detection in vehicles passing a checkpoint at highway speeds. SNM is postulated to be stored in a moving vehicle and detector assemblies are placed on the road-side or in chambers embedded below the road surface. Neutron and gamma spectral awareness is important for the detector assembly design besides high efficiencies, so that different SNMs can be detected and identified with various possible shielding settings. The detector assembly design is composed of a CsI gamma-ray detector block and five neutron detector blocks, with peak efficiencies targeting different energy ranges determined by adjoint simulations. In this study, formulations are derived using adjoint transport simulations to estimate detector efficiencies. The formulations is applied to investigate several neutron detector designs for Block IV, which has its peak efficiency in the thermal range, and Block V, designed to maximize the total neutron counts over the entire energy spectrum. Other Blocks detect different neutron energies. All five neutron detector blocks and the gamma-ray block are assembled in both MCNP and deterministic simulation models, with detector responses calculated to validate the fully assembled design using a 30-group library. The simulation results show that the 30-group library, collapsed from an 80-group library using an adjoint-weighting approach with the YGROUP code, significantly reduced the computational cost while maintaining accuracy. (authors)

  19. Application of the adjoint function methodology for neutron fluence determination

    Haghighat, A.; Nanayakkara, B.; Livingston, J.; Mahgerefteh, M.; Luoma, J.

    1991-01-01

    In previous studies, the neutron fluence at a reactor pressure vessel has been estimated based on consolidation of transport theory calculations and experimental data obtained from in-vessel capsules and/or cavity dosimeters. Normally, a forward neutron transport calculation is performed for each fuel cycle and the neutron fluxes are integrated over the reactor operating time to estimate the neutron fluence. Such calculations are performed for a geometrical model which is composed of one-eighth (0 to 45 deg) of the reactor core and its surroundings; i.e., core barrel, thermal shield, downcomer, reactor vessel, cavity region, concrete wall, and instrumentation well. Because the model is large, transport theory calculations generally require a significant amount of computer memory and time; hence, more efficient methodologies such as the adjoint transport approach have been proposed. These studies, however, do not address the necessary sensitivity studies needed for adjoint function calculations. The adjoint methodology has been employed to estimate the activity of a cavity dosimeter and that of an in-vessel capsule. A sensitivity study has been performed on the mesh distribution used in and around the cavity dosimeter and the in-vessel capsule. Further, since a major portion of the detector response is due to the neutrons originated in the peripheral fuel assemblies, a study on the use of a smaller calculational model has been performed

  20. Wake Shield Target Protection

    Valmianski, Emanuil I.; Petzoldt, Ronald W.; Alexander, Neil B.

    2003-01-01

    The heat flux from both gas convection and chamber radiation on a direct drive target must be limited to avoid target damage from excessive D-T temperature increase. One of the possibilities of protecting the target is a wake shield flying in front of the target. A shield will also reduce drag force on the target, thereby facilitating target tracking and position prediction. A Direct Simulation Monte Carlo (DSMC) code was used to calculate convection heat loads as boundary conditions input into ANSYS thermal calculations. These were used for studying the quality of target protection depending on various shapes of shields, target-shield distance, and protective properties of the shield moving relative to the target. The results show that the shield can reduce the convective heat flux by a factor of 2 to 5 depending on pressure, temperature, and velocity. The protective effect of a shield moving relative to the target is greater than the protective properties of a fixed shield. However, the protective effect of a shield moving under the drag force is not sufficient for bringing the heat load on the target down to the necessary limit. Some other ways of diminishing heat flux using a protective shield are discussed

  1. SHREDI a removal diffusion shielding code for x-y and r-z geometries

    Daneri, A.; Toselli, G.

    1974-01-01

    The SHREDI, a removal diffusion neutron shielding code written in FORTRAN for IBM 370/165, is presented. The code computes neutron fluxes or adjoint fluxes and activations in bidimensional sections of the shield. It is also possible to consider shielding points with the same coordinate (y or z) (monodimensional problems)

  2. Validation of a new 39 neutron group self-shielded library based on the nucleonics analysis of the Lotus fusion-fission hybrid test facility performed with the Monte Carlo code

    Pelloni, S.; Cheng, E.T.

    1985-02-01

    The Swiss LOTUS fusion-fission hybrid test facility was used to investigate the influence of the self-shielding of resonance cross sections on the tritium breeding and on the thorium ratios. Nucleonic analyses were performed using the discrete-ordinates transport codes ANISN and ONEDANT, the surface-flux code SURCU, and the version 3 of the MCNP code for the Li 2 CO 3 and the Li 2 O blanket designs with lead, thorium and beryllium multipliers. Except for the MCNP calculation which bases on the ENDF/B-V files, all nuclear data are generated from the ENDF/B-IV basic library. For the deterministic methods three NJOY group libraries were considered. The first, a 39 neutron group self-shielded library, was generated at EIR. The second bases on the same group structure as the first does and consists of infinitely diluted cross sections. Finally the third library was processed at LANL and consists of coupled 30+12 neutron and gamma groups; these cross sections are not self-shielded. The Monte Carlo analysis bases on a continuous and on a discrete 262 group library from the ENDF/B-V evaluation. It is shown that the results agree well within 3% between the unshielded libraries and between the different transport codes and theories. The self-shielding of resonance cross sections results in a decrease of the thorium capture rate and in an increase of the tritium breeding of about 6%. The remaining computed ratios are not affected by the self-shielding of cross sections. (Auth.)

  3. Calculation of the BREN house shielding experiments

    Woolson, William A.; Gritzner, Michael L.

    1987-01-01

    The BREN house transmission experiments provide an excellent set of measurements to validate the calculational procedures that will be used to derive house shielding estimates for the revised dosimetry of the survivors of the Hiroshima and Nagasaki A-bombs. The BREN experiments were performed in realistic full scale models of Japanese residences. Although the radiation spectra and relative intensities of neutrons and gamma rays incident on the houses from the HPRR and the 60 Co source are not appropriate for direct application to the A-bomb survivors, they cover the full energy range of importance. The codes and calculations required to compare with BREN experiments are the same as those needed for the A-bomb dosimetry. They consist of a two-dimensional discrete-ordinates calculation of the free field coupled to an adjoint Monte Carlo calculation in detailed house geometry. The agreement obtained between calculations and the experiments is excellent for neutrons and 60 Co gamma rays. Every house transmission calculation spanning simple to complex configurations and detector locations for the 60 Co and HPRR was within an acceptable margin of error. The gamma-ray TF calculations for the reactor source did not agree well with the experiments. Analysis of this discrepancy, however, strongly indicates that the problem probably does not reside in the calculational procedure but in the measurements themselves. In conclusion, it is believed that the excellent agreement of our calculations with the BREN experiments validates the calculational procedure which is planed to be applied o estimating the house shielding for survivors of the Hiroshima and Nagasaki A-bombs. Certainly, the calculations for Hiroshima and Nagasaki will involve modifications to the code used for the computations reported here, but to the extent that these modifications involve increased calculational complexity to treat more realistic materials and configurations, the benchmark established by these

  4. Characterization by Monte Carlo of the dose after a glass shield lead for gamma ray; Caracterizacion por Monte Carlo de la dosis tras un blindaje de vidrio de plomo para rayos gamma

    Esteve Sanchez, S.; Gil Conde, M.; Contreras Gonzalez, J. L.; Rosado, J.; Pazyi, V.

    2013-07-01

    When a gamma-ray beam crosses the border between two media characterized by atomic number very different is they produce effects on the distribution of doses near the border difficult to predict with simple models. The case of rays gamma affecting a lead glass is particularly interesting for its application to shielding of common use. interested in studying the importance of the residual dose after the shield. (Author)

  5. The Laplace transformation of adjoint transport equations

    Hoogenboom, J.E.

    1985-01-01

    A clarification is given of the difference between the equation adjoint to the Laplace-transformed time-dependent transport equation and the Laplace-transformed time-dependent adjoint transport equation. Proper procedures are derived to obtain the Laplace transform of the instantaneous detector response. (author)

  6. The adjoint space in heat transport theory

    Dam, H. van; Hoogenboom, J.E.

    1980-01-01

    The mathematical concept of adjoint operators is applied to the heat transport equation and an adjoint equation is defined with a detector function as source term. The physical meaning of the solutions for the latter equation is outlined together with an application in the field of perturbation analysis. (author)

  7. Tax Shield, Insolvenz und Zinsschranke

    Arnold, Sven; Lahmann, Alexander; Schwetzler, Bernhard

    2010-01-01

    Dieser Beitrag analysiert den Wertbeitrag fremdfinanzierungsbedingter Steuervorteile (Tax Shield) unter realistischen Bedingungen (keine Negativsteuer; mögliche Insolvenz) für unterschiedliche Finanzierungspolitiken. Zusätzlich wird der Effekt der sogenannten Zinsschranke auf den Wert des Tax Shield ermittelt. Die Bewertung des Tax Shield mit und ohne Zinsschranke findet im einperiodigen Fall auf der Basis von Optionspreismodellen und im mehrperiodigen Fall auf der Basis von Monte Carlo Simul...

  8. Running coupling from gluon and ghost propagators in the Landau gauge: Yang-Mills theories with adjoint fermions

    Bergner, Georg; Piemonte, Stefano

    2018-04-01

    Non-Abelian gauge theories with fermions transforming in the adjoint representation of the gauge group (AdjQCD) are a fundamental ingredient of many models that describe the physics beyond the Standard Model. Two relevant examples are N =1 supersymmetric Yang-Mills (SYM) theory and minimal walking technicolor, which are gauge theories coupled to one adjoint Majorana and two adjoint Dirac fermions, respectively. While confinement is a property of N =1 SYM, minimal walking technicolor is expected to be infrared conformal. We study the propagators of ghost and gluon fields in the Landau gauge to compute the running coupling in the MiniMom scheme. We analyze several different ensembles of lattice Monte Carlo simulations for the SU(2) adjoint QCD with Nf=1 /2 ,1 ,3 /2 , and 2 Dirac fermions. We show how the running of the coupling changes as the number of interacting fermions is increased towards the conformal window.

  9. Double-layer neutron shield design as neutron shielding application

    Sariyer, Demet; Küçer, Rahmi

    2018-02-01

    The shield design in particle accelerators and other high energy facilities are mainly connected to the high-energy neutrons. The deep penetration of neutrons through massive shield has become a very serious problem. For shielding to be efficient, most of these neutrons should be confined to the shielding volume. If the interior space will become limited, the sufficient thickness of multilayer shield must be used. Concrete and iron are widely used as a multilayer shield material. Two layers shield material was selected to guarantee radiation safety outside of the shield against neutrons generated in the interaction of the different proton energies. One of them was one meter of concrete, the other was iron-contained material (FeB, Fe2B and stainless-steel) to be determined shield thicknesses. FLUKA Monte Carlo code was used for shield design geometry and required neutron dose distributions. The resulting two layered shields are shown better performance than single used concrete, thus the shield design could leave more space in the interior shielded areas.

  10. Shielding Design and Radiation Shielding Evaluation for LSDS System Facility

    Kim, Younggook; Kim, Jeongdong; Lee, Yongdeok

    2015-01-01

    As the system characteristics, the target in the spectrometer emits approximately 1012 neutrons/s. To efficiently shield the neutron, the shielding door designs are proposed for the LSDS system through a comparison of the direct shield and maze designs. Hence, to guarantee the radiation safety for the facility, the door design is a compulsory course of the development of the LSDS system. To improve the shielding rates, 250x250 covering structure was added as a subsidiary around the spectrometer. In this study, the evaluations of the suggested shielding designs were conducted using MCNP code. The suggested door design and covering structures can shield the neutron efficiently, thus all evaluations of all conditions are satisfied within the public dose limits. From the Monte Carlo code simulation, Resin(Indoor type) and Tungsten(Outdoor type) were selected as the shielding door materials. From a comparative evaluation of the door thickness, In and Out door thickness was selected 50 cm

  11. Shielding practice

    Sauermann, P.F.

    1985-08-01

    The basis of shielding practice against external irradiation is shown in a simple way. For most sources of radiation (point sources) occurring in shielding practice, the basic data are given, mainly in the form of tables, which are required to solve the shielding problems. The application of these data is explained and discussed using practical examples. Thickness of shielding panes of glove boxes for α and β radiation; shielding of sealed γ-radiography sources; shielding of a Co-60 radiation source, and of the manipulator panels for hot cells; damping factors for γ radiation and neutrons; shielding of fast and thermal neutrons, and of bremsstrahlung (X-ray tubes, Kr-85 pressure gas cylinders, 42 MeV betatrons, 20 MeV linacs); two-fold shielding (lead glass windows for hot cells, 14 MeV neutron generators); shielding against scattered radiation. (orig./HP) [de

  12. Numerical study of dense adjoint matter in two color QCD

    Hands, S.; Morrison, S.; Scorzato, L.; Oevers, M.

    2000-06-01

    We identify the global symmetries of SU(2) lattice gauge theory with N flavors of staggered fermion in the presence of a quark chemical potential μ, for fermions in both fundamental and adjoint representations, and anticipate likely patterns of symmetry breaking at both low and high densities. Results from numerical simulations of the model with N=1 adjoint flavor on a 4 3 x 8 lattice are presented, using both hybrid Monte Carlo and two-step multi-boson algorithms. It is shown that the sign of the fermion determinant starts to fluctuate once the model enters a phase with non-zero baryon charge density. HMC simulations are not ergodic in this regime, but TSMB simulations retain ergodicity even in the dense phase, and in addition appear to show superior decorrelation. The HMC results for the equation of state and the pion mass show good quantitative agreement with the predictions of chiral perturbation theory, which should hold only for N≥2. The TSMB results incorporating the sign of the determinant support a delayed onset transition, consistent with the pattern of symmetry breaking expected for N=1. (orig.)

  13. Multigroup adjoint transport solution using the method of cyclic characteristics

    Assawaroongruengchot, M.; Marleau, G.

    2005-01-01

    The adjoint transport solution algorithm based on the method of cyclic characteristics (MOCC) is developed for the heterogeneous 2-dimensional geometries. The adjoint characteristics equation associated with a cyclic tracking line is formulated, then a closed form for adjoint angular flux can be determined. The acceleration techniques are implemented using the group-reduction and group-splitting techniques. To demonstrate the efficacy of the algorithm, the calculations are performed on the 17*17 PWR and Watanabe-Maynard benchmark problems. Comparisons of adjoint flux and k eff results obtained by MOCC and collision probability (CP) methods are performed. The mathematical relationship between pseudo-adjoint flux obtained by CP method and adjoint flux by MOCC method is presented. It appears that the pseudo-adjoint flux by CP method is equivalent to the adjoint flux by MOCC method and that the MOCC method requires lower computing time than the CP method for a single adjoint flux calculation

  14. Multigroup adjoint transport solution using the method of cyclic characteristics

    Assawaroongruengchot, M.; Marleau, G. [Ecole Polytechnique de Montreal, Institut de Genie Nucleaire, Montreal, Quebec (Canada)

    2005-07-01

    The adjoint transport solution algorithm based on the method of cyclic characteristics (MOCC) is developed for the heterogeneous 2-dimensional geometries. The adjoint characteristics equation associated with a cyclic tracking line is formulated, then a closed form for adjoint angular flux can be determined. The acceleration techniques are implemented using the group-reduction and group-splitting techniques. To demonstrate the efficacy of the algorithm, the calculations are performed on the 17*17 PWR and Watanabe-Maynard benchmark problems. Comparisons of adjoint flux and k{sub eff} results obtained by MOCC and collision probability (CP) methods are performed. The mathematical relationship between pseudo-adjoint flux obtained by CP method and adjoint flux by MOCC method is presented. It appears that the pseudo-adjoint flux by CP method is equivalent to the adjoint flux by MOCC method and that the MOCC method requires lower computing time than the CP method for a single adjoint flux calculation.

  15. Monte Carlo techniques for analyzing deep penetration problems

    Cramer, S.N.; Gonnord, J.; Hendricks, J.S.

    1985-01-01

    A review of current methods and difficulties in Monte Carlo deep-penetration calculations is presented. Statistical uncertainty is discussed, and recent adjoint optimization of splitting, Russian roulette, and exponential transformation biasing is reviewed. Other aspects of the random walk and estimation processes are covered, including the relatively new DXANG angular biasing technique. Specific items summarized are albedo scattering, Monte Carlo coupling techniques with discrete ordinates and other methods, adjoint solutions, and multi-group Monte Carlo. The topic of code-generated biasing parameters is presented, including the creation of adjoint importance functions from forward calculations. Finally, current and future work in the area of computer learning and artificial intelligence is discussed in connection with Monte Carlo applications

  16. Monte Carlo techniques for analyzing deep penetration problems

    Cramer, S.N.; Gonnord, J.; Hendricks, J.S.

    1985-01-01

    A review of current methods and difficulties in Monte Carlo deep-penetration calculations is presented. Statistical uncertainty is discussed, and recent adjoint optimization of splitting, Russian roulette, and exponential transformation biasing is reviewed. Other aspects of the random walk and estimation processes are covered, including the relatively new DXANG angular biasing technique. Specific items summarized are albedo scattering, Monte Carlo coupling techniques with discrete ordinates and other methods, adjoint solutions, and multi-group Monte Carlo. The topic of code-generated biasing parameters is presented, including the creation of adjoint importance functions from forward calculations. Finally, current and future work in the area of computer learning and artificial intelligence is discussed in connection with Monte Carlo applications. 29 refs

  17. Monte Carlo based demonstration of sufficiently dimensioned shielding for a Co-60 testing facility; Monte-Carlo-basierter Nachweis der ausreichend dimensionierten Abschirmung einer {sup 60}Co-Pruefanlage

    Wind, Michael; Beck, Peter; Latocha, Marcin [Seibersdorf Labor GmbH, Radiation Protection Dosimetry, Seibersdorf (Austria)

    2015-07-01

    The electrical properties of electronic equipment can be changed in an ionized radiation field. The knowledge of these changes is necessary for applications in space, in air traffic and nuclear medicine. Experimental tests will be performed in Co-60 radiation fields in the irradiation facility (TEC facility) of the Seibersdorf Labor GmbH that is in construction. The contribution deals with a simulation that is aimed to calculate the local dose rate within and outside the building for demonstration of sufficient dimensioning of the shielding in compliance with the legal dose rate limits.

  18. Electromagnetic shielding

    Tzeng, Wen-Shian V.

    1991-01-01

    Electromagnetic interference (EMI) shielding materials are well known in the art in forms such as gaskets, caulking compounds, adhesives, coatings and the like for a variety of EMI shielding purposes. In the past, where high shielding performance is necessary, EMI shielding has tended to use silver particles or silver coated copper particles dispersed in a resin binder. More recently, aluminum core silver coated particles have been used to reduce costs while maintaining good electrical and physical properties. (author). 8 figs

  19. Recent Improvements in the SHIELD-HIT Code

    Hansen, David Christoffer; Lühr, Armin Christian; Herrmann, Rochus

    2012-01-01

    Purpose: The SHIELD-HIT Monte Carlo particle transport code has previously been used to study a wide range of problems for heavy-ion treatment and has been benchmarked extensively against other Monte Carlo codes and experimental data. Here, an improved version of SHIELD-HIT is developed concentra......Purpose: The SHIELD-HIT Monte Carlo particle transport code has previously been used to study a wide range of problems for heavy-ion treatment and has been benchmarked extensively against other Monte Carlo codes and experimental data. Here, an improved version of SHIELD-HIT is developed...

  20. Shielding calculations using FLUKA

    Yamaguchi, Chiri; Tesch, K.; Dinter, H.

    1988-06-01

    The dose equivalent on the surface of concrete shielding has been calculated using the Monte Carlo code FLUKA86 for incident proton energies from 10 to 800 GeV. The results have been compared with some simple equations. The value of the angular dependent parameter in Moyer's equation has been calculated from the locations where the values of the maximum dose equivalent occur. (author)

  1. Adjoint spectrum calculation in fuel heterogeneous cells

    Suster, Luis Carlos

    1998-01-01

    In most codes for cells calculation, the multigroup cross sections are generated taking into consideration the conservation of the reaction rates in the forward spectrum. However, for certain uses of the perturbation theory it's necessary to use the average of the parameters for energy macrogroups over the forward and the adjoint spectra. In this thesis the adjoint spectrum was calculated from the adjoint neutron balance equations, that were obtained for a heterogeneous unit cell. The collision probabilities method was used to obtain these equations. In order optimize the computational run-time, the Gaussian quadrature method was used in the calculation of the neutron balance equations, forward and adjoint. This method of integration was also used for the Doppler broadening functions calculation, necessary for obtaining the energy dependent cross sections. In order to calculate the reaction rates and the average cross sections, using both the forward and the adjoint neutron spectra, the most important resonances of the U 238 were considered. The results obtained with the method show significant differences for the different cross sections weighting schemes. (author)

  2. Verification of a hybrid adjoint methodology in Titan for single photon emission computed tomography - 316

    Royston, K.; Haghighat, A.; Yi, C.

    2010-01-01

    The hybrid deterministic transport code TITAN is being applied to a Single Photon Emission Computed Tomography (SPECT) simulation of a myocardial perfusion study. The TITAN code's hybrid methodology allows the use of a discrete ordinates solver in the phantom region and a characteristics method solver in the collimator region. Currently we seek to validate the adjoint methodology in TITAN for this application using a SPECT model that has been created in the MCNP5 Monte Carlo code. The TITAN methodology was examined based on the response of a single voxel detector placed in front of the heart with and without collimation. For the case without collimation, the TITAN response for single voxel-sized detector had a -9.96% difference relative to the MCNP5 response. To simulate collimation, the adjoint source was specified in directions located within the collimator acceptance angle. For a single collimator hole with a diameter matching the voxel dimension, a difference of -0.22% was observed. Comparisons to groupings of smaller collimator holes of two different sizes resulted in relative differences of 0.60% and 0.12%. The number of adjoint source directions within an acceptance angle was increased and showed no significant change in accuracy. Our results indicate that the hybrid adjoint methodology of TITAN yields accurate solutions greater than a factor of two faster than MCNP5. (authors)

  3. Optimization of path length stretching in Monte Carlo calculations for non-leakage problems

    Hoogenboom, J.E. [Delft Univ. of Technology (Netherlands)

    2005-07-01

    Path length stretching (or exponential biasing) is a well known variance reduction technique in Monte Carlo calculations. It can especially be useful in shielding problems where particles have to penetrate a lot of material before being tallied. Several authors sought for optimization of the path length stretching parameter for detection of the leakage of neutrons from a slab. Here the adjoint function behaves as a single exponential function and can well be used to determine the stretching parameter. In this paper optimization is sought for a detector embedded in the system, which changes the adjoint function in the detector drastically. From literature it is known that the combination of path length stretching and angular biasing can result in appreciable variance reduction. However, angular biasing is not generally available in general purpose Monte Carlo codes and therefore we want to restrict ourselves to the application of pure path length stretching and finding optimum parameters for that. Nonetheless, the starting point for our research is the zero-variance scheme. In order to study the solution in detail the simplified monoenergetic two-direction model is adopted, which allows analytical solutions and can still be used in a Monte Carlo simulation. Knowing the zero-variance solution analytically, it is shown how optimum path length stretching parameters can be derived from it. It results in path length shrinking in the detector. Results for the variance in the detector response are shown in comparison with other patterns for the stretching parameter. The effect of anisotropic scattering on the path length stretching parameter is taken into account. (author)

  4. The Hausdorff measure of chaotic sets of adjoint shift maps

    Wang Huoyun [Department of Mathematics of Guangzhou University, Guangzhou 510006 (China)]. E-mail: wanghuoyun@sina.com; Song Wangan [Department of Computer, Huaibei Coal Industry Teacher College, Huaibei 235000 (China)

    2006-11-15

    In this paper, the size of chaotic sets of adjoint shift maps is estimated by Hausdorff measure. We prove that for any adjoint shift map there exists a finitely chaotic set with full Hausdorff measure.

  5. ADGEN: ADjoint GENerator for computer models

    Worley, B.A.; Pin, F.G.; Horwedel, J.E.; Oblow, E.M.

    1989-05-01

    This paper presents the development of a FORTRAN compiler and an associated supporting software library called ADGEN. ADGEN reads FORTRAN models as input and produces and enhanced version of the input model. The enhanced version reproduces the original model calculations but also has the capability to calculate derivatives of model results of interest with respect to any and all of the model data and input parameters. The method for calculating the derivatives and sensitivities is the adjoint method. Partial derivatives are calculated analytically using computer calculus and saved as elements of an adjoint matrix on direct assess storage. The total derivatives are calculated by solving an appropriate adjoint equation. ADGEN is applied to a major computer model of interest to the Low-Level Waste Community, the PRESTO-II model. PRESTO-II sample problem results reveal that ADGEN correctly calculates derivatives of response of interest with respect to 300 parameters. The execution time to create the adjoint matrix is a factor of 45 times the execution time of the reference sample problem. Once this matrix is determined, the derivatives with respect to 3000 parameters are calculated in a factor of 6.8 that of the reference model for each response of interest. For a single 3000 for determining these derivatives by parameter perturbations. The automation of the implementation of the adjoint technique for calculating derivatives and sensitivities eliminates the costly and manpower-intensive task of direct hand-implementation by reprogramming and thus makes the powerful adjoint technique more amenable for use in sensitivity analysis of existing models. 20 refs., 1 fig., 5 tabs.

  6. Global Seismic Imaging Based on Adjoint Tomography

    Bozdag, E.; Lefebvre, M.; Lei, W.; Peter, D. B.; Smith, J. A.; Zhu, H.; Komatitsch, D.; Tromp, J.

    2013-12-01

    Our aim is to perform adjoint tomography at the scale of globe to image the entire planet. We have started elastic inversions with a global data set of 253 CMT earthquakes with moment magnitudes in the range 5.8 ≤ Mw ≤ 7 and used GSN stations as well as some local networks such as USArray, European stations, etc. Using an iterative pre-conditioned conjugate gradient scheme, we initially set the aim to obtain a global crustal and mantle model with confined transverse isotropy in the upper mantle. Global adjoint tomography has so far remained a challenge mainly due to computational limitations. Recent improvements in our 3D solvers (e.g., a GPU version) and access to high-performance computational centers (e.g., ORNL's Cray XK7 "Titan" system) now enable us to perform iterations with higher-resolution (T > 9 s) and longer-duration (200 min) simulations to accommodate high-frequency body waves and major-arc surface waves, respectively, which help improve data coverage. The remaining challenge is the heavy I/O traffic caused by the numerous files generated during the forward/adjoint simulations and the pre- and post-processing stages of our workflow. We improve the global adjoint tomography workflow by adopting the ADIOS file format for our seismic data as well as models, kernels, etc., to improve efficiency on high-performance clusters. Our ultimate aim is to use data from all available networks and earthquakes within the magnitude range of our interest (5.5 ≤ Mw ≤ 7) which requires a solid framework to manage big data in our global adjoint tomography workflow. We discuss the current status and future of global adjoint tomography based on our initial results as well as practical issues such as handling big data in inversions and on high-performance computing systems.

  7. ADGEN: ADjoint GENerator for computer models

    Worley, B.A.; Pin, F.G.; Horwedel, J.E.; Oblow, E.M.

    1989-05-01

    This paper presents the development of a FORTRAN compiler and an associated supporting software library called ADGEN. ADGEN reads FORTRAN models as input and produces and enhanced version of the input model. The enhanced version reproduces the original model calculations but also has the capability to calculate derivatives of model results of interest with respect to any and all of the model data and input parameters. The method for calculating the derivatives and sensitivities is the adjoint method. Partial derivatives are calculated analytically using computer calculus and saved as elements of an adjoint matrix on direct assess storage. The total derivatives are calculated by solving an appropriate adjoint equation. ADGEN is applied to a major computer model of interest to the Low-Level Waste Community, the PRESTO-II model. PRESTO-II sample problem results reveal that ADGEN correctly calculates derivatives of response of interest with respect to 300 parameters. The execution time to create the adjoint matrix is a factor of 45 times the execution time of the reference sample problem. Once this matrix is determined, the derivatives with respect to 3000 parameters are calculated in a factor of 6.8 that of the reference model for each response of interest. For a single 3000 for determining these derivatives by parameter perturbations. The automation of the implementation of the adjoint technique for calculating derivatives and sensitivities eliminates the costly and manpower-intensive task of direct hand-implementation by reprogramming and thus makes the powerful adjoint technique more amenable for use in sensitivity analysis of existing models. 20 refs., 1 fig., 5 tabs

  8. System of adjoint P1 equations for neutron moderation

    Martinez, Aquilino Senra; Silva, Fernando Carvalho da; Cardoso, Carlos Eduardo Santos

    2000-01-01

    In some applications of perturbation theory, it is necessary know the adjoint neutron flux, which is obtained by the solution of adjoint neutron diffusion equation. However, the multigroup constants used for this are weighted in only the direct neutron flux, from the solution of direct P1 equations. In this work, this procedure is questioned and the adjoint P1 equations are derived by the neutron transport equation, the reversion operators rules and analogies between direct and adjoint parameters. (author)

  9. Adjoint-Based Uncertainty Quantification with MCNP

    Seifried, Jeffrey E. [Univ. of California, Berkeley, CA (United States)

    2011-09-01

    This work serves to quantify the instantaneous uncertainties in neutron transport simulations born from nuclear data and statistical counting uncertainties. Perturbation and adjoint theories are used to derive implicit sensitivity expressions. These expressions are transformed into forms that are convenient for construction with MCNP6, creating the ability to perform adjoint-based uncertainty quantification with MCNP6. These new tools are exercised on the depleted-uranium hybrid LIFE blanket, quantifying its sensitivities and uncertainties to important figures of merit. Overall, these uncertainty estimates are small (< 2%). Having quantified the sensitivities and uncertainties, physical understanding of the system is gained and some confidence in the simulation is acquired.

  10. Shielding plugs

    Makishima, Kenji.

    1986-01-01

    Purpose: In shielding plugs of an LMFBR type reactor, to restrain natural convection of heat in an annular space between a thermal shield layer and a shield shell, to prevent the lowering of heat-insulation performance, and to alleviate a thermal stress in a reactor container and the shield shell. Constitution: A ring-like leaf spring split in the direction of height is disposed in an annular space between a thermal shield layer and a shield shell. In consequence, the space is partitioned in the direction of height and, therefore, if axial temperature conditions and space width are the same and the space is low, the natural convection is hard to occur. Thus the rise of upper surface temperature of the shielding plugs can prevent the lowering of the heat insulation performance which will result in the increment of shielding plug cooling capacity, thereby improving reliability. In the meantime, since there is mounted an earthquake-resisting support, the thermal shield layer will move for a slight gap in case of an earthquake, being supported by the earthquake-resisting support, and the movement of the thermal shield layer is restricted, thereby maintaining integrity without increasing the stroke of the ring-like spring. (Kawakami, Y.)

  11. Monte Carlo transport model comparison with 1A GeV accelerated iron experiment: heavy-ion shielding evaluation of NASA space flight-crew foodstuff

    Stephens, D. L. Jr; Townsend, L. W.; Miller, J.; Zeitlin, C.; Heilbronn, L.

    2002-01-01

    Deep-space manned flight as a reality depends on a viable solution to the radiation problem. Both acute and chronic radiation health threats are known to exist, with solar particle events as an example of the former and galactic cosmic rays (GCR) of the latter. In this experiment Iron ions of 1A GeV are used to simulate GCR and to determine the secondary radiation field created as the GCR-like particles interact with a thick target. A NASA prepared food pantry locker was subjected to the iron beam and the secondary fluence recorded. A modified version of the Monte Carlo heavy ion transport code developed by Zeitlin at LBNL is compared with experimental fluence. The foodstuff is modeled as mixed nuts as defined by the 71st edition of the Chemical Rubber Company (CRC) Handbook of Physics and Chemistry. The results indicate a good agreement between the experimental data and the model. The agreement between model and experiment is determined using a linear fit to ordered pairs of data. The intercept is forced to zero. The slope fit is 0.825 and the R2 value is 0.429 over the resolved fluence region. The removal of an outlier, Z=14, gives values of 0.888 and 0.705 for slope and R2 respectively. c2002 COSPAR. Published by Elsevier Science Ltd. All rights reserved.

  12. Double-Difference Global Adjoint Tomography

    Orsvuran, R.; Bozdag, E.; Lei, W.; Tromp, J.

    2017-12-01

    The adjoint method allows us to incorporate full waveform simulations in inverse problems. Misfit functions play an important role in extracting the relevant information from seismic waveforms. In this study, our goal is to apply the Double-Difference (DD) methodology proposed by Yuan et al. (2016) to global adjoint tomography. Dense seismic networks, such as USArray, lead to higher-resolution seismic images underneath continents. However, the imbalanced distribution of stations and sources poses challenges in global ray coverage. We adapt double-difference multitaper measurements to global adjoint tomography. We normalize each DD measurement by its number of pairs, and if a measurement has no pair, as may frequently happen for data recorded at oceanic stations, classical multitaper measurements are used. As a result, the differential measurements and pair-wise weighting strategy help balance uneven global kernel coverage. Our initial experiments with minor- and major-arc surface waves show promising results, revealing more pronounced structure near dense networks while reducing the prominence of paths towards cluster of stations. We have started using this new measurement in global adjoint inversions, addressing azimuthal anisotropy in upper mantle. Meanwhile, we are working on combining the double-difference approach with instantaneous phase measurements to emphasize contributions of scattered waves in global inversions and extending it to body waves. We will present our results and discuss challenges and future directions in the context of global tomographic inversions.

  13. Dual of QCD with One Adjoint Fermion

    Mojaza, Matin; Nardecchia, Marco; Pica, Claudio

    2011-01-01

    We construct the magnetic dual of QCD with one adjoint Weyl fermion. The dual is a consistent solution of the 't Hooft anomaly matching conditions, allows for flavor decoupling and remarkably constitutes the first nonsupersymmetric dual valid for any number of colors. The dual allows to bound...

  14. Adjoint P1 equations solution for neutron slowing down

    Cardoso, Carlos Eduardo Santos; Martinez, Aquilino Senra; Silva, Fernando Carvalho da

    2002-01-01

    In some applications of perturbation theory, it is necessary know the adjoint neutron flux, which is obtained by the solution of adjoint neutron diffusion equation. However, the multigroup constants used for this are weighted in only the direct neutron flux, from the solution of direct P1 equations. In this work, the adjoint P1 equations are derived by the neutron transport equation, the reversion operators rules and analogies between direct and adjoint parameters. The direct and adjoint neutron fluxes resulting from the solution of P 1 equations were used to three different weighting processes, to obtain the macrogroup macroscopic cross sections. It was found out noticeable differences among them. (author)

  15. Shielding container

    Darling, K.A.M.

    1981-01-01

    A shielding container incorporates a dense shield, for example of depleted uranium, cast around a tubular member of curvilinear configuration for accommodating a radiation source capsule. A lining for the tubular member, in the form of a close-coiled flexible guide, provides easy replaceability to counter wear while the container is in service. Container life is extended, and maintenance costs are reduced. (author)

  16. Fully automatic adjoints: a robust and efficient mechanism for generating adjoint ocean models

    Ham, D. A.; Farrell, P. E.; Funke, S. W.; Rognes, M. E.

    2012-04-01

    The problem of generating and maintaining adjoint models is sufficiently difficult that typically only the most advanced and well-resourced community ocean models achieve it. There are two current technologies which each suffer from their own limitations. Algorithmic differentiation, also called automatic differentiation, is employed by models such as the MITGCM [2] and the Alfred Wegener Institute model FESOM [3]. This technique is very difficult to apply to existing code, and requires a major initial investment to prepare the code for automatic adjoint generation. AD tools may also have difficulty with code employing modern software constructs such as derived data types. An alternative is to formulate the adjoint differential equation and to discretise this separately. This approach, known as the continuous adjoint and employed in ROMS [4], has the disadvantage that two different model code bases must be maintained and manually kept synchronised as the model develops. The discretisation of the continuous adjoint is not automatically consistent with that of the forward model, producing an additional source of error. The alternative presented here is to formulate the flow model in the high level language UFL (Unified Form Language) and to automatically generate the model using the software of the FEniCS project. In this approach it is the high level code specification which is differentiated, a task very similar to the formulation of the continuous adjoint [5]. However since the forward and adjoint models are generated automatically, the difficulty of maintaining them vanishes and the software engineering process is therefore robust. The scheduling and execution of the adjoint model, including the application of an appropriate checkpointing strategy is managed by libadjoint [1]. In contrast to the conventional algorithmic differentiation description of a model as a series of primitive mathematical operations, libadjoint employs a new abstraction of the simulation

  17. A new approach for developing adjoint models

    Farrell, P. E.; Funke, S. W.

    2011-12-01

    Many data assimilation algorithms rely on the availability of gradients of misfit functionals, which can be efficiently computed with adjoint models. However, the development of an adjoint model for a complex geophysical code is generally very difficult. Algorithmic differentiation (AD, also called automatic differentiation) offers one strategy for simplifying this task: it takes the abstraction that a model is a sequence of primitive instructions, each of which may be differentiated in turn. While extremely successful, this low-level abstraction runs into time-consuming difficulties when applied to the whole codebase of a model, such as differentiating through linear solves, model I/O, calls to external libraries, language features that are unsupported by the AD tool, and the use of multiple programming languages. While these difficulties can be overcome, it requires a large amount of technical expertise and an intimate familiarity with both the AD tool and the model. An alternative to applying the AD tool to the whole codebase is to assemble the discrete adjoint equations and use these to compute the necessary gradients. With this approach, the AD tool must be applied to the nonlinear assembly operators, which are typically small, self-contained units of the codebase. The disadvantage of this approach is that the assembly of the discrete adjoint equations is still very difficult to perform correctly, especially for complex multiphysics models that perform temporal integration; as it stands, this approach is as difficult and time-consuming as applying AD to the whole model. In this work, we have developed a library which greatly simplifies and automates the alternate approach of assembling the discrete adjoint equations. We propose a complementary, higher-level abstraction to that of AD: that a model is a sequence of linear solves. The developer annotates model source code with library calls that build a 'tape' of the operators involved and their dependencies, and

  18. Radiation shielding analysis

    Moon, S.H.; Ha, C.W.; Kwon, S.K.; Lee, J.K.; Choi, H.S.

    1982-01-01

    The theoretical bases of radiation streaming analysis in power reactors, such as ducts or reactor cavity, have been investigated. Discrete ordinates-Monte Carlo or Monte Carlo-Monte Carlo coupling techniques are suggested for the streaming analysis of ducts or reactor cavity. Single albedo scattering approximation code (SINALB) has been developed for simple and quick estimation of gamma-ray ceiling scattering, where the ceiling is assumed to be semi-infinite medium. This code has been employed to calculate the gamma-ray ceiling scattering effects in the laboratory containing a Co-60 source. The SINALB is applicable to gamma-ray scattering, only where the ceiling is thicker than Σsup(-1) and the height is at least twice higher than the shield wall. This code can be used for the purpose of preliminary radiation shield design. The MORSE code has been improved to analyze the gamma-ray scattering problem with on approximation method in respect to the random walk and estimation processes. This improved MORSE code has been employed to the gamma-ray ceiling scattering problem. The results of the improved MORSE calculation are in good agreement with the SINALB and standard MORSE. (Author)

  19. Validation of a new midway forward-adjoint coupling option in MCNP

    Serov, I.V.; John, T.M.; Hoogenboom, J.E. [Technische Univ. Delft (Netherlands). Interfacultair Reactor Inst.

    1996-09-01

    The new midway Monte Carlo is based on the coupling of scores from a forward and an adjoint Monte Carlo calculation on a surface in between the source and the detector. The method is implemented in MCNP. The utilization of the method is fairly straight-forward and does not require any substantial expertise. The midway Monte Carlo method was tested against the gamma-ray skyshine MCNP benchmark problem. This problem involves deep penetration and streaming along complicated paths. The midway method supplied results, which agree with the results of the reference calculation within the limits of the estimated statistical uncertainties. The efficiency of the easy-to-implement midway calculation is higher than the efficiency of the reference calculation which is already optimized by use of an importance function. The midway method proves to be efficient in problems with complicated streaming paths towards small detectors. (author)

  20. Validation of a new midway forward-adjoint coupling option in MCNP

    Serov, I.V.; John, T.M.; Hoogenboom, J.E.

    1996-01-01

    The new midway Monte Carlo is based on the coupling of scores from a forward and an adjoint Monte Carlo calculation on a surface in between the source and the detector. The method is implemented in MCNP. The utilization of the method is fairly straight-forward and does not require any substantial expertise. The midway Monte Carlo method was tested against the gamma-ray skyshine MCNP benchmark problem. This problem involves deep penetration and streaming along complicated paths. The midway method supplied results, which agree with the results of the reference calculation within the limits of the estimated statistical uncertainties. The efficiency of the easy-to-implement midway calculation is higher than the efficiency of the reference calculation which is already optimized by use of an importance function. The midway method proves to be efficient in problems with complicated streaming paths towards small detectors. (author)

  1. Monte Carlo methods for the reliability analysis of Markov systems

    Buslik, A.J.

    1985-01-01

    This paper presents Monte Carlo methods for the reliability analysis of Markov systems. Markov models are useful in treating dependencies between components. The present paper shows how the adjoint Monte Carlo method for the continuous time Markov process can be derived from the method for the discrete-time Markov process by a limiting process. The straightforward extensions to the treatment of mean unavailability (over a time interval) are given. System unavailabilities can also be estimated; this is done by making the system failed states absorbing, and not permitting repair from them. A forward Monte Carlo method is presented in which the weighting functions are related to the adjoint function. In particular, if the exact adjoint function is known then weighting factors can be constructed such that the exact answer can be obtained with a single Monte Carlo trial. Of course, if the exact adjoint function is known, there is no need to perform the Monte Carlo calculation. However, the formulation is useful since it gives insight into choices of the weight factors which will reduce the variance of the estimator

  2. Adjoint string breaking in the pseudoparticle approach

    Szasz, Christian; Wagner, Marc

    2008-01-01

    We apply the pseudoparticle approach to SU(2) Yang-Mills theory and perform a detailed study of the potential between two static charges for various representations. Whereas for charges in the fundamental representation we find a linearly rising confining potential, we clearly observe string breaking, when considering charges in the adjoint representation. We also demonstrate Casimir scaling and compute gluelump masses for different spin and parity. Numerical results are in qualitative agreement with lattice results.

  3. Radiation shielding for neutron guides

    Ersez, T.; Braoudakis, G.; Osborn, J.C.

    2005-01-01

    Full text: Models of the neutron guide shielding for the out of bunker guides on the thermal and cold neutron beam lines of the OPAL Reactor (ANSTO) were constructed using the Monte Carlo code MCNP 4B. The neutrons that were not reflected inside the guides but were absorbed by the supermirror (SM) layers were noted to be a significant source of gammas. Gammas also arise from neutrons absorbed by the B, Si, Na and K contained in the glass. The proposed shielding design has produced compact shielding assemblies. These arrangements are consistent with safety requirements, floor load limits, and cost constraints. To verify the design a prototype was assembled consisting of 120mm thick Pb(96%)Sb(4%) walls resting on a concrete block. There was good agreement between experimental measurements and calculated dose rates for bulk shield regions. (authors)

  4. Radiation shielding for neutron guides

    Ersez, T.; Braoudakis, G.; Osborn, J.C.

    2006-01-01

    Models of the neutron guide shielding for the out of bunker guides on the thermal and cold neutron beam lines of the OPAL Reactor (ANSTO) were constructed using the Monte Carlo code MCNP 4B. The neutrons that were not reflected inside the guides but were absorbed by the supermirror (SM) layers were noted to be a significant source of gammas. Gammas also arise from neutrons absorbed by the B, Si, Na and K contained in the glass. The proposed shielding design has produced compact shielding assemblies. These arrangements are consistent with safety requirements, floor load limits, and cost constraints. To verify the design a prototype was assembled consisting of 120 mm thick Pb(96%)Sb(4%) walls resting on a concrete block. There was good agreement between experimental measurements and calculated dose rates for bulk shield regions

  5. Electromagnetic shield

    Miller, J.S.

    1987-01-01

    An electromagnetic shield is described comprising: closed, electrically-conductive rings, each having an open center; and binder means for arranging the rings in a predetermined, fixed relationship relative to each other, the so-arranged rings and binder means defining an outer surface; wherein electromagnetic energy received by the shield from a source adjacent its outer surface induces an electrical current to flow in a predetermined direction adjacent and parallel to the outer surface, through the rings; and wherein each ring is configured to cause source-induced alternating current flowing through the portion of the ring closest to the outer surface to electromagnetically induce an oppositely-directed current in the portion of the ring furthest from the surface, such oppositely-directed current bucking any source-induced current in the latter ring portion and thus reducing the magnitude of current flowing through it, whereby the electromagnetic shielding effected by the shield is enhanced

  6. Neutron shieldings

    Tarutani, Kohei

    1979-01-01

    Purpose: To decrease the stresses resulted by the core bendings to the base of an entrance nozzle. Constitution: Three types of round shielding rods of different diameter are arranged in a hexagonal tube. The hexagonal tube is provided with several spacer pads receiving the loads from the core constrain mechanism at its outer circumference, a handling head for a fuel exchanger at its top and an entrance nozzle for self-holding the neutron shieldings and flowing heat-removing coolants at its bottom. The diameters for R 1 , R 2 and R 3 for the round shielding rods are designed as: 0.1 R 1 2 1 and 0.2 R 1 2 1 . Since a plurality of shielding rods of small diameter are provided, soft structure are obtained and a plurality of coolant paths are formed. (Furukawa, Y.)

  7. Nuclear shields

    Linares, R.C.; Nienart, L.F.; Toelcke, G.A.

    1976-01-01

    A process is described for preparing melt-processable nuclear shielding compositions from chloro-fluoro substituted ethylene polymers, particularly PCTFE and E-CTFE, containing 1 to 75 percent by weight of a gadolinium compound. 13 claims, no drawings

  8. REACTOR SHIELD

    Wigner, E.P.; Ohlinger, L.E.; Young, G.J.; Weinberg, A.M.

    1959-02-17

    Radiation shield construction is described for a nuclear reactor. The shield is comprised of a plurality of steel plates arranged in parallel spaced relationship within a peripheral shell. Reactor coolant inlet tubes extend at right angles through the plates and baffles are arranged between the plates at right angles thereto and extend between the tubes to create a series of zigzag channels between the plates for the circulation of coolant fluid through the shield. The shield may be divided into two main sections; an inner section adjacent the reactor container and an outer section spaced therefrom. Coolant through the first section may be circulated at a faster rate than coolant circulated through the outer section since the area closest to the reactor container is at a higher temperature and is more radioactive. The two sections may have separate cooling systems to prevent the coolant in the outer section from mixing with the more contaminated coolant in the inner section.

  9. Eguchi-Kawai reduction with one flavor of adjoint Moebius fermion

    Cunningham, William; Giedt, Joel

    2013-01-01

    We study the single site lattice gauge theory of SU(N) coupled to one Dirac flavor of fermion in the adjoint representation. We utilize M\\"obius fermions for this study, and accelerate the calculation with graphics processing units (GPUs). Our Monte Carlo simulations indicate that for sufficiently large inverse 't Hooft coupling b = 1/g^2 N, and for N \\leq 10 the distribution of traced Polyakov loops has "fingers" that extend from the origin. However, in the massless case the distribution of ...

  10. SCALE6 Hybrid Deterministic-Stochastic Shielding Methodology for PWR Containment Calculations

    Matijevic, Mario; Pevec, Dubravko; Trontl, Kresimir

    2014-01-01

    The capabilities and limitations of SCALE6/MAVRIC hybrid deterministic-stochastic shielding methodology (CADIS and FW-CADIS) are demonstrated when applied to a realistic deep penetration Monte Carlo (MC) shielding problem of full-scale PWR containment model. The ultimate goal of such automatic variance reduction (VR) techniques is to achieve acceptable precision for the MC simulation in reasonable time by preparation of phase-space VR parameters via deterministic transport theory methods (discrete ordinates SN) by generating space-energy mesh-based adjoint function distribution. The hybrid methodology generates VR parameters that work in tandem (biased source distribution and importance map) in automated fashion which is paramount step for MC simulation of complex models with fairly uniform mesh tally uncertainties. The aim in this paper was determination of neutron-gamma dose rate distribution (radiation field) over large portions of PWR containment phase-space with uniform MC uncertainties. The sources of ionizing radiation included fission neutrons and gammas (reactor core) and gammas from activated two-loop coolant. Special attention was given to focused adjoint source definition which gave improved MC statistics in selected materials and/or regions of complex model. We investigated benefits and differences of FW-CADIS over CADIS and manual (i.e. analog) MC simulation of particle transport. Computer memory consumption by deterministic part of hybrid methodology represents main obstacle when using meshes with millions of cells together with high SN/PN parameters, so optimization of control and numerical parameters of deterministic module plays important role for computer memory management. We investigated the possibility of using deterministic module (memory intense) with broad group library v7 2 7n19g opposed to fine group library v7 2 00n47g used with MC module to fully take effect of low energy particle transport and secondary gamma emission. Compared with

  11. On the non-uniqueness of the nodal mathematical adjoint

    Müller, Erwin

    2014-01-01

    Highlights: • We evaluate three CMFD schemes for computing the nodal mathematical adjoint. • The nodal mathematical adjoint is not unique and can be non-positive (nonphysical). • Adjoint and forward eigenmodes are compatible if produced by the same CMFD method. • In nodal applications the excited eigenmodes are purely mathematical entities. - Abstract: Computation of the neutron adjoint flux within the framework of modern nodal diffusion methods is often facilitated by reducing the nodal equation system for the forward flux into a simpler coarse-mesh finite-difference form and then transposing the resultant matrix equations. The solution to the transposed problem is known as the nodal mathematical adjoint. Since the coarse-mesh finite-difference reduction of a given nodal formulation can be obtained in a number of ways, different nodal mathematical adjoint solutions can be computed. This non-uniqueness of the nodal mathematical adjoint challenges the credibility of the reduction strategy and demands a verdict as to its suitability in practical applications. This is the matter under consideration in this paper. A selected number of coarse-mesh finite-difference reduction schemes are described and compared. Numerical calculations are utilised to illustrate the differences in the adjoint solutions as well as to appraise the impact on such common applications as the computation of core point kinetics parameters. Recommendations are made for the proper application of the coarse-mesh finite-difference reduction approach to the nodal mathematical adjoint problem

  12. Shielding features of quarry stone

    Hernandez V, C.; Contreras S, H.; Hernandez A, L.; Baltazar R, A.; Escareno J, E.; Mares E, C. A.; Vega C, H. R.

    2010-10-01

    Quarry stone lineal attenuation coefficient for gamma-rays has been obtained. In Zacatecas, quarry stone is widely utilized as a decorative item in buildings, however its shielding features against gamma-rays unknown. The aim of this work is to determine the shielding properties of quarry stone against γ-rays using Monte Carlo calculations where a detailed model of a good geometry experimental setup was carried out. In the calculations 10 pieces 10 X 10 cm 2 of different thickness were utilized to evaluate the photons transmission as the quarry stone thickness is increased. It was noticed that transmitted photons decay away as the shield thickness is increased, these results were fitted to an exponential function were the linear attenuation coefficient was estimated. Also, using XCOM code the linear attenuation coefficient from several keV up to 100 MeV was estimated. From the comparison between Monte Carlo results and XCOM calculations a good agreement was found. For 0.662 MeV γ-rays the attenuation coefficient of quarry stone, whose density is 2.413 g-cm -3 , is 0.1798 cm -1 , this mean a X 1/2 = 3.9 cm, X 1/4 = 7.7 cm, X 1/10 = 12.8 cm, and X 1/100 = 25.6 cm. Having the information of quarry stone performance as shielding give the chance to use this material to shield X and γ-ray facilities. (Author)

  13. Neutron shielding for a 252 Cf source

    Vega C, H.R.; Manzanares A, E.; Hernandez D, V.M.; Eduardo Gallego, Alfredo Lorente

    2006-01-01

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

  14. Doppler Temperature Coefficient Calculations Using Adjoint-Weighted Tallies and Continuous Energy Cross Sections in MCNP6

    Gonzales, Matthew Alejandro

    The calculation of the thermal neutron Doppler temperature reactivity feedback co-efficient, a key parameter in the design and safe operation of advanced reactors, using first order perturbation theory in continuous energy Monte Carlo codes is challenging as the continuous energy adjoint flux is not readily available. Traditional approaches of obtaining the adjoint flux attempt to invert the random walk process as well as require data corresponding to all temperatures and their respective temperature derivatives within the system in order to accurately calculate the Doppler temperature feedback. A new method has been developed using adjoint-weighted tallies and On-The-Fly (OTF) generated continuous energy cross sections within the Monte Carlo N-Particle (MCNP6) transport code. The adjoint-weighted tallies are generated during the continuous energy k-eigenvalue Monte Carlo calculation. The weighting is based upon the iterated fission probability interpretation of the adjoint flux, which is the steady state population in a critical nuclear reactor caused by a neutron introduced at that point in phase space. The adjoint-weighted tallies are produced in a forward calculation and do not require an inversion of the random walk. The OTF cross section database uses a high order functional expansion between points on a user-defined energy-temperature mesh in which the coefficients with respect to a polynomial fitting in temperature are stored. The coefficients of the fits are generated before run- time and called upon during the simulation to produce cross sections at any given energy and temperature. The polynomial form of the OTF cross sections allows the possibility of obtaining temperature derivatives of the cross sections on-the-fly. The use of Monte Carlo sampling of adjoint-weighted tallies and the capability of computing derivatives of continuous energy cross sections with respect to temperature are used to calculate the Doppler temperature coefficient in a research

  15. Radiation shielding

    Aitken, D.

    1979-01-01

    Shields for equipment in which ionising radiation is associated with high electrical gradients, for example X-ray tubes and particle accelerators, incorporate a radiation-absorbing metal, as such or as a compound, and are electrically non-conducting and can be placed in the high electrical gradient region of the equipment. Substances disclosed include dispersions of lead, tungsten, uranium or oxides of these in acrylics polyesters, PVC, ABS, polyamides, PTFE, epoxy resins, glass or ceramics. The material used may constitute an evacuable enclosure of the equipment or may be an external shield thereof. (U.K.)

  16. Igo - A Monte Carlo Code For Radiotherapy Planning

    Goldstein, M.; Regev, D.

    1999-01-01

    The goal of radiation therapy is to deliver a lethal dose to the tumor, while minimizing the dose to normal tissues and vital organs. To carry out this task, it is critical to calculate correctly the 3-D dose delivered. Monte Carlo transport methods (especially the Adjoint Monte Carlo have the potential to provide more accurate predictions of the 3-D dose the currently used methods. IG0 is a Monte Carlo code derived from the general Monte Carlo Program - MCNP, tailored specifically for calculating the effects of radiation therapy. This paper describes the IG0 transport code, the PIG0 interface and some preliminary results

  17. Variational Variance Reduction for Monte Carlo Criticality Calculations

    Densmore, Jeffery D.; Larsen, Edward W.

    2001-01-01

    A new variational variance reduction (VVR) method for Monte Carlo criticality calculations was developed. This method employs (a) a variational functional that is more accurate than the standard direct functional, (b) a representation of the deterministically obtained adjoint flux that is especially accurate for optically thick problems with high scattering ratios, and (c) estimates of the forward flux obtained by Monte Carlo. The VVR method requires no nonanalog Monte Carlo biasing, but it may be used in conjunction with Monte Carlo biasing schemes. Some results are presented from a class of criticality calculations involving alternating arrays of fuel and moderator regions

  18. Shielding Benchmark Computational Analysis

    Hunter, H.T.; Slater, C.O.; Holland, L.B.; Tracz, G.; Marshall, W.J.; Parsons, J.L.

    2000-01-01

    Over the past several decades, nuclear science has relied on experimental research to verify and validate information about shielding nuclear radiation for a variety of applications. These benchmarks are compared with results from computer code models and are useful for the development of more accurate cross-section libraries, computer code development of radiation transport modeling, and building accurate tests for miniature shielding mockups of new nuclear facilities. When documenting measurements, one must describe many parts of the experimental results to allow a complete computational analysis. Both old and new benchmark experiments, by any definition, must provide a sound basis for modeling more complex geometries required for quality assurance and cost savings in nuclear project development. Benchmarks may involve one or many materials and thicknesses, types of sources, and measurement techniques. In this paper the benchmark experiments of varying complexity are chosen to study the transport properties of some popular materials and thicknesses. These were analyzed using three-dimensional (3-D) models and continuous energy libraries of MCNP4B2, a Monte Carlo code developed at Los Alamos National Laboratory, New Mexico. A shielding benchmark library provided the experimental data and allowed a wide range of choices for source, geometry, and measurement data. The experimental data had often been used in previous analyses by reputable groups such as the Cross Section Evaluation Working Group (CSEWG) and the Organization for Economic Cooperation and Development/Nuclear Energy Agency Nuclear Science Committee (OECD/NEANSC)

  19. Shielding design of ITER pressure suppression system

    Yamauchi, Michinori; Sato, Satoshi; Nishitani, Takeo; Kawasaki, Hiromitsu

    2006-01-01

    The duct shield from streaming D-T neutrons has been designed for the ITER pressure suppression system. Streaming calculations are performed with the DUCT-III code for the region from the inlet of the pressure relief line to the rupture disk. Next, the neutron permeation through the shield is studied by Monte Carlo calculations with the MCNP code. It is found that 0.15 m thick iron shield is enough to suppress the permeating component from the outside. In addition, it is suggested that the volume of the shield can be reduced by about 30% if the optimized iron shield structure having localized thickness across intense permeation paths is employed to shield the pressure suppression line. (T.I.)

  20. Elementary operators on self-adjoint operators

    Molnar, Lajos; Semrl, Peter

    2007-03-01

    Let H be a Hilbert space and let and be standard *-operator algebras on H. Denote by and the set of all self-adjoint operators in and , respectively. Assume that and are surjective maps such that M(AM*(B)A)=M(A)BM(A) and M*(BM(A)B)=M*(B)AM*(B) for every pair , . Then there exist an invertible bounded linear or conjugate-linear operator and a constant c[set membership, variant]{-1,1} such that M(A)=cTAT*, , and M*(B)=cT*BT, .

  1. Spent Fuel Pool Dose Rate Calculations Using Point Kernel and Hybrid Deterministic-Stochastic Shielding Methods

    Matijevic, M.; Grgic, D.; Jecmenica, R.

    2016-01-01

    This paper presents comparison of the Krsko Power Plant simplified Spent Fuel Pool (SFP) dose rates using different computational shielding methodologies. The analysis was performed to estimate limiting gamma dose rates on wall mounted level instrumentation in case of significant loss of cooling water. The SFP was represented with simple homogenized cylinders (point kernel and Monte Carlo (MC)) or cuboids (MC) using uranium, iron, water, and dry-air as bulk region materials. The pool is divided on the old and new section where the old one has three additional subsections representing fuel assemblies (FAs) with different burnup/cooling time (60 days, 1 year and 5 years). The new section represents the FAs with the cooling time of 10 years. The time dependent fuel assembly isotopic composition was calculated using ORIGEN2 code applied to the depletion of one of the fuel assemblies present in the pool (AC-29). The source used in Microshield calculation is based on imported isotopic activities. The time dependent photon spectra with total source intensity from Microshield multigroup point kernel calculations was then prepared for two hybrid deterministic-stochastic sequences. One is based on SCALE/MAVRIC (Monaco and Denovo) methodology and another uses Monte Carlo code MCNP6.1.1b and ADVANTG3.0.1. code. Even though this model is a fairly simple one, the layers of shielding materials are thick enough to pose a significant shielding problem for MC method without the use of effective variance reduction (VR) technique. For that purpose the ADVANTG code was used to generate VR parameters (SB cards in SDEF and WWINP file) for MCNP fixed-source calculation using continuous energy transport. ADVATNG employs a deterministic forward-adjoint transport solver Denovo which implements CADIS/FW-CADIS methodology. Denovo implements a structured, Cartesian-grid SN solver based on the Koch-Baker-Alcouffe parallel transport sweep algorithm across x-y domain blocks. This was first

  2. Self-adjointness of the Gaffney Laplacian on Vector Bundles

    Bandara, Lashi, E-mail: lashi.bandara@chalmers.se [Chalmers University of Technology and University of Gothenburg, Mathematical Sciences (Sweden); Milatovic, Ognjen, E-mail: omilatov@unf.edu [University of North Florida, Department of Mathematics and Statistics (United States)

    2015-12-15

    We study the Gaffney Laplacian on a vector bundle equipped with a compatible metric and connection over a Riemannian manifold that is possibly geodesically incomplete. Under the hypothesis that the Cauchy boundary is polar, we demonstrate the self-adjointness of this Laplacian. Furthermore, we show that negligible boundary is a necessary and sufficient condition for the self-adjointness of this operator.

  3. A reduced adjoint approach to variational data assimilation

    Altaf, Muhammad; El Gharamti, Mohamad; Heemink, Arnold W.; Hoteit, Ibrahim

    2013-01-01

    The adjoint method has been used very often for variational data assimilation. The computational cost to run the adjoint model often exceeds several original model runs and the method needs significant programming efforts to implement the adjoint model code. The work proposed here is variational data assimilation based on proper orthogonal decomposition (POD) which avoids the implementation of the adjoint of the tangent linear approximation of the original nonlinear model. An ensemble of the forward model simulations is used to determine the approximation of the covariance matrix and only the dominant eigenvectors of this matrix are used to define a model subspace. The adjoint of the tangent linear model is replaced by the reduced adjoint based on this reduced space. Thus the adjoint model is run in reduced space with negligible computational cost. Once the gradient is obtained in reduced space it is projected back in full space and the minimization process is carried in full space. In the paper the reduced adjoint approach to variational data assimilation is introduced. The characteristics and performance of the method are illustrated with a number of data assimilation experiments in a ground water subsurface contaminant model. © 2012 Elsevier B.V.

  4. A reduced adjoint approach to variational data assimilation

    Altaf, Muhammad

    2013-02-01

    The adjoint method has been used very often for variational data assimilation. The computational cost to run the adjoint model often exceeds several original model runs and the method needs significant programming efforts to implement the adjoint model code. The work proposed here is variational data assimilation based on proper orthogonal decomposition (POD) which avoids the implementation of the adjoint of the tangent linear approximation of the original nonlinear model. An ensemble of the forward model simulations is used to determine the approximation of the covariance matrix and only the dominant eigenvectors of this matrix are used to define a model subspace. The adjoint of the tangent linear model is replaced by the reduced adjoint based on this reduced space. Thus the adjoint model is run in reduced space with negligible computational cost. Once the gradient is obtained in reduced space it is projected back in full space and the minimization process is carried in full space. In the paper the reduced adjoint approach to variational data assimilation is introduced. The characteristics and performance of the method are illustrated with a number of data assimilation experiments in a ground water subsurface contaminant model. © 2012 Elsevier B.V.

  5. Self-adjointness of the Gaffney Laplacian on Vector Bundles

    Bandara, Lashi; Milatovic, Ognjen

    2015-01-01

    We study the Gaffney Laplacian on a vector bundle equipped with a compatible metric and connection over a Riemannian manifold that is possibly geodesically incomplete. Under the hypothesis that the Cauchy boundary is polar, we demonstrate the self-adjointness of this Laplacian. Furthermore, we show that negligible boundary is a necessary and sufficient condition for the self-adjointness of this operator

  6. The dynamic adjoint as a Green’s function

    Pázsit, I.; Dykin, V.

    2015-01-01

    Highlight: • The relationship between the direct Green’s function and the dynamic adjoint function is discussed in two-group theory. • It is shown that the elements of the direct Greens’ function matrix are identical to those of the transpose of the adjoint Green’s function matrix, with an interchange of arguments. • It is also remarked how the dynamic adjoint function of van Dam can be given in terms of the direct Green’s function matrix. - Abstract: The concept of the dynamic adjoint was introduced by Hugo van Dam for calculating the in-core neutron noise in boiling water reactors in the mid-70’s. This successful approach found numerous applications for calculating the neutron noise in both PWRs and BWRs since then. Although the advantages and disadvantages of using the direct (forward) or the adjoint (backward) approach for the calculation of the neutron noise were analysed in a number of publications, the direct relationship between the forward Green’s function and the dynamic adjoint has not been discussed. On the other hand, in particle transport theory the relationship between the direct and adjoint Green’s function has been discussed in detail, in which Mike Williams has had many seminal contributions. In this note we analyse the relationship between the direct Green’s function and the dynamic adjoint in the spirit of Mike’s work in neutron transport and radiation damage theory. The paper is closed with some personal remarks and reminiscences.

  7. Simultaneous Optimization of Tallies in Difficult Shielding Problems

    Peplow, Douglas E.; Evans, Thomas M.; Wagner, John C.

    2008-01-01

    Monte Carlo is quite useful for calculating specific quantities in complex transport problems. Many variance reduction strategies have been developed that accelerate Monte Carlo calculations for specific tallies. However, when trying to calculate multiple tallies or a mesh tally, users have had to accept different levels of relative uncertainty among the tallies or run separate calculations optimized for each individual tally. To address this limitation, an extension of the CADIS (Consistent Adjoint Driven Importance Sampling) method, which is used for difficult source/detector problems, has been developed to optimize several tallies or the cells of a mesh tally simultaneously. The basis for this method is the development of an importance function that represents the importance of particles to the objective of uniform Monte Carlo particle density in the desired tally regions. This method utilizes the results of a forward discrete ordinates solution, which may be based on a quick, coarse-mesh calculation, to develop a forward-weighted source for the adjoint calculation. The importance map and the biased source computed from the adjoint flux are then used in the forward Monte Carlo calculation to obtain approximately uniform relative uncertainties for the desired tallies. This extension is called forward-weighted CADIS, or FW-CADIS

  8. Gonadal shield.

    Purdy, J A; Stiteler, R D; Glasgow, G P; Mill, W B

    1975-10-01

    A secondary gonadal shield for use in the pelvic irradiation of males was designed and built using material and apparatus available with the Cerrobend blocking system. The gonadal dose was reduced to approximately 1.5 to 2.5% of the given dose.

  9. Simulation of transport equations with Monte Carlo

    Matthes, W.

    1975-09-01

    The main purpose of the report is to explain the relation between the transport equation and the Monte Carlo game used for its solution. The introduction of artificial particles carrying a weight provides one with high flexibility in constructing many different games for the solution of the same equation. This flexibility opens a way to construct a Monte Carlo game for the solution of the adjoint transport equation. Emphasis is laid mostly on giving a clear understanding of what to do and not on the details of how to do a specific game

  10. Sensitivity analysis of predictive models with an automated adjoint generator

    Pin, F.G.; Oblow, E.M.

    1987-01-01

    The adjoint method is a well established sensitivity analysis methodology that is particularly efficient in large-scale modeling problems. The coefficients of sensitivity of a given response with respect to every parameter involved in the modeling code can be calculated from the solution of a single adjoint run of the code. Sensitivity coefficients provide a quantitative measure of the importance of the model data in calculating the final results. The major drawback of the adjoint method is the requirement for calculations of very large numbers of partial derivatives to set up the adjoint equations of the model. ADGEN is a software system that has been designed to eliminate this drawback and automatically implement the adjoint formulation in computer codes. The ADGEN system will be described and its use for improving performance assessments and predictive simulations will be discussed. 8 refs., 1 fig

  11. Development of neutron shielding concrete containing iron content materials

    Sariyer, Demet; Küçer, Rahmi

    2018-02-01

    Concrete is one of the most important construction materials which widely used as a neutron shielding. Neutron shield is obtained of interaction with matter depends on neutron energy and the density of the shielding material. Shielding properties of concrete could be improved by changing its composition and density. High density materials such as iron or high atomic number elements are added to concrete to increase the radiation resistance property. In this study, shielding properties of concrete were investigated by adding iron, FeB, Fe2B, stainless - steel at different ratios into concrete. Neutron dose distributions and shield design was obtained by using FLUKA Monte Carlo code. The determined shield thicknesses vary depending on the densities of the mixture formed by the additional material and ratio. It is seen that a combination of iron rich materials is enhanced the neutron shielding of capabilities of concrete. Also, the thicknesses of shield are reduced.

  12. Radiation shielding of the main injector

    Bhat, C.M.; Martin, P.S.

    1995-05-01

    The radiation shielding in the Fermilab Main Injector (FMI) complex has been carried out by adopting a number of prescribed stringent guidelines established by a previous safety analysis. Determination of the required amount of radiation shielding at various locations of the FMI has been done using Monte Carlo computations. A three dimensional ray tracing code as well as a code based upon empirical observations have been employed in certain cases

  13. Radiation shielding

    Yue, D.D.

    1979-01-01

    Details are given of a cylindrical electric penetration assembly for carrying instrumentation leads, used in monitoring the performance of a nuclear reactor, through the containment wall of the reactor. Effective yet economical shielding protection against both fast neutron and high-energy gamma radiation is provided. Adequate spacing within the assembly allows excessive heat to be efficiently dissipated and means of monitoring all potential radiation and gas leakage paths are provided. (UK)

  14. Shielded container

    Fries, B.A.

    1978-01-01

    A shielded container for transportation of radioactive materials is disclosed in which leakage from the container is minimized due to constructional features including, inter alia, forming the container of a series of telescoping members having sliding fits between adjacent side walls and having at least two of the members including machine sealed lids and at least two of the elements including hand-tightenable caps

  15. Automated Monte Carlo biasing for photon-generated electrons near surfaces.

    Franke, Brian Claude; Crawford, Martin James; Kensek, Ronald Patrick

    2009-09-01

    This report describes efforts to automate the biasing of coupled electron-photon Monte Carlo particle transport calculations. The approach was based on weight-windows biasing. Weight-window settings were determined using adjoint-flux Monte Carlo calculations. A variety of algorithms were investigated for adaptivity of the Monte Carlo tallies. Tree data structures were used to investigate spatial partitioning. Functional-expansion tallies were used to investigate higher-order spatial representations.

  16. GPU-accelerated adjoint algorithmic differentiation

    Gremse, Felix; Höfter, Andreas; Razik, Lukas; Kiessling, Fabian; Naumann, Uwe

    2016-03-01

    Many scientific problems such as classifier training or medical image reconstruction can be expressed as minimization of differentiable real-valued cost functions and solved with iterative gradient-based methods. Adjoint algorithmic differentiation (AAD) enables automated computation of gradients of such cost functions implemented as computer programs. To backpropagate adjoint derivatives, excessive memory is potentially required to store the intermediate partial derivatives on a dedicated data structure, referred to as the ;tape;. Parallelization is difficult because threads need to synchronize their accesses during taping and backpropagation. This situation is aggravated for many-core architectures, such as Graphics Processing Units (GPUs), because of the large number of light-weight threads and the limited memory size in general as well as per thread. We show how these limitations can be mediated if the cost function is expressed using GPU-accelerated vector and matrix operations which are recognized as intrinsic functions by our AAD software. We compare this approach with naive and vectorized implementations for CPUs. We use four increasingly complex cost functions to evaluate the performance with respect to memory consumption and gradient computation times. Using vectorization, CPU and GPU memory consumption could be substantially reduced compared to the naive reference implementation, in some cases even by an order of complexity. The vectorization allowed usage of optimized parallel libraries during forward and reverse passes which resulted in high speedups for the vectorized CPU version compared to the naive reference implementation. The GPU version achieved an additional speedup of 7.5 ± 4.4, showing that the processing power of GPUs can be utilized for AAD using this concept. Furthermore, we show how this software can be systematically extended for more complex problems such as nonlinear absorption reconstruction for fluorescence-mediated tomography.

  17. Phenomenology of spinless adjoints in two universal extra dimensions

    Ghosh, Kirtiman; Datta, Anindya

    2008-01-01

    We discuss the phenomenology of (1,1)-mode adjoint scalars in the framework of two Universal Extra Dimensions. The Kaluza-Klein (KK) towers of these adjoint scalars arise in the 4-dimensional effective theory from the 6th component of the gauge fields after compactification. Adjoint scalars can have KK-number conserving as well as KK-number violating interactions. We calculate the KK-number violating operators involving these scalars and two Standard Model fields. Decay widths of these scalars into different channels have been estimated. We have also briefly discussed pair-production and single production of such scalars at the Large Hadron Collider

  18. Adjoint sensitivity analysis of plasmonic structures using the FDTD method.

    Zhang, Yu; Ahmed, Osman S; Bakr, Mohamed H

    2014-05-15

    We present an adjoint variable method for estimating the sensitivities of arbitrary responses with respect to the parameters of dispersive discontinuities in nanoplasmonic devices. Our theory is formulated in terms of the electric field components at the vicinity of perturbed discontinuities. The adjoint sensitivities are computed using at most one extra finite-difference time-domain (FDTD) simulation regardless of the number of parameters. Our approach is illustrated through the sensitivity analysis of an add-drop coupler consisting of a square ring resonator between two parallel waveguides. The computed adjoint sensitivities of the scattering parameters are compared with those obtained using the accurate but computationally expensive central finite difference approach.

  19. Shielding analyses: the rabbit vs the turtle?

    Broadhead, B.L.

    1996-01-01

    This paper compares solutions using Monte Carlo and discrete- ordinates methods applied to two actual shielding situations in order to make some general observations concerning the efficiency and advantages/disadvantages of the two approaches. The discrete- ordinates solutions are performed using two-dimensional geometries, while the Monte Carlo approaches utilize three-dimensional geometries with both multigroup and point cross-section data

  20. A code for leakage neutron spectra through thick shields

    Nagarajan, P.S.; Sethulakshmi, P.; Raghavendran, C.P.

    1975-01-01

    An exponential transform Monte Carlo code has been developed for deep penetration of neutrons and the results of leakage neutron spectra of this code have been compared with those of a basic Monte Carlo code for small thickness. The development of the code and optimisation of certain transform parameters are discussed and results are presented for a few thick shields of concrete and water in the context of neutron monitoring in the environs of accelerator and reactor shields. (author)

  1. Radiation shielding quality assurance

    Um, Dallsun

    For the radiation shielding quality assurance, the validity and reliability of the neutron transport code MCNP, which is now one of the most widely used radiation shielding analysis codes, were checked with lot of benchmark experiments. And also as a practical example, follows were performed in this thesis. One integral neutron transport experiment to measure the effect of neutron streaming in iron and void was performed with Dog-Legged Void Assembly in Knolls Atomic Power Laboratory in 1991. Neutron flux was measured six different places with the methane detectors and a BF-3 detector. The main purpose of the measurements was to provide benchmark against which various neutron transport calculation tools could be compared. Those data were used in verification of Monte Carlo Neutron & Photon Transport Code, MCNP, with the modeling for that. Experimental results and calculation results were compared in both ways, as the total integrated value of neutron fluxes along neutron energy range from 10 KeV to 2 MeV and as the neutron spectrum along with neutron energy range. Both results are well matched with the statistical error +/-20%. MCNP results were also compared with those of TORT, a three dimensional discrete ordinates code which was developed by Oak Ridge National Laboratory. MCNP results are superior to the TORT results at all detector places except one. This means that MCNP is proved as a very powerful tool for the analysis of neutron transport through iron & air and further it could be used as a powerful tool for the radiation shielding analysis. For one application of the analysis of variance (ANOVA) to neutron and gamma transport problems, uncertainties for the calculated values of critical K were evaluated as in the ANOVA on statistical data.

  2. Neutron shielding performance of water-extended polyester

    Vega Carrillo, H.R.; Manzanares-Acuna, E.; Hernandez-Davila, V.M.; Vega Carrillo, H.R.; Hernandez-Davila, V.M.; Gallego, E.; Lorente, A.

    2006-01-01

    A Monte Carlo study to determine the shielding features to neutrons of water-extended polyester (WEP) was carried out. Materials with low atomic number are predominantly used for neutron shielding because these materials effectively attenuate neutrons, mainly through elastic and inelastic collisions. In addition to neutron attenuation properties, other desirable properties for neutron shielding materials include mechanical strength, stability, low cost, and ease of handling. During the selection of materials to design a neutron shield, prompt gamma production as well as radionuclide induced by neutron activation must be considered. In this investigation the Monte Carlo method (MCNP code) was used to evaluate the performance of a water-extended polyester shield designed for the transportation, storage, and use of a 252 Cf isotopic neutron source, for comparison the calculations were extended to water shielding, the bare source in vacuum and in air. (authors)

  3. Neutron shielding performance of water-extended polyester

    Vega Carrillo, H.R.; Manzanares-Acuna, E.; Hernandez-Davila, V.M.; Vega Carrillo, H.R.; Gallegoc, E.; Lorentec, A.; Hernandez-Davila, V.M.

    2006-01-01

    A Monte Carlo study to determine the shielding features to neutrons of water-extended polyester was carried out. Materials with low atomic number are predominantly used for neutron shielding because these materials effectively attenuate neutrons, mainly through elastic and inelastic collisions. In addition to neutron attenuation properties, other desirable properties for neutron shielding materials include mechanical strength, stability, low cost, and ease of handling. During the selection of materials to design a neutron shield, prompt gamma production as well as radionuclide induced by neutron activation must be considered. In this investigation the Monte Carlo method (M.C.N.P. code) was used to evaluate the performance of a water-extended polyester shield designed for the transportation, storage, and use of a 252 Cf isotopic neutron source, for comparison the calculations were extended to water shielding, the bare source in vacuum and in air. (authors)

  4. Neutron shielding performance of water-extended polyester

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

    2006-07-01

    A Monte Carlo study to determine the shielding features to neutrons of water-extended polyester (WEP) was carried out. Materials with low atomic number are predominantly used for neutron shielding because these materials effectively attenuate neutrons, mainly through elastic and inelastic collisions. In addition to neutron attenuation properties, other desirable properties for neutron shielding materials include mechanical strength, stability, low cost, and ease of handling. During the selection of materials to design a neutron shield, prompt gamma production as well as radionuclide induced by neutron activation must be considered. In this investigation the Monte Carlo method (MCNP code) was used to evaluate the performance of a water-extended polyester shield designed for the transportation, storage, and use of a {sup 252}Cf isotopic neutron source, for comparison the calculations were extended to water shielding, the bare source in vacuum and in air. (authors)

  5. Response decomposition with Monte Carlo correlated coupling

    Ueki, T.; Hoogenboom, J.E.; Kloosterman, J.L.

    2001-01-01

    Particle histories that contribute to a detector response are categorized according to whether they are fully confined inside a source-detector enclosure or cross and recross the same enclosure. The contribution from the confined histories is expressed using a forward problem with the external boundary condition on the source-detector enclosure. The contribution from the crossing and recrossing histories is expressed as the surface integral at the same enclosure of the product of the directional cosine and the fluxes in the foregoing forward problem and the adjoint problem for the whole spatial domain. The former contribution can be calculated by a standard forward Monte Carlo. The latter contribution can be calculated by correlated coupling of forward and adjoint histories independently of the former contribution. We briefly describe the computational method and discuss its application to perturbation analysis for localized material changes. (orig.)

  6. Response decomposition with Monte Carlo correlated coupling

    Ueki, T.; Hoogenboom, J.E.; Kloosterman, J.L. [Delft Univ. of Technology (Netherlands). Interfaculty Reactor Inst.

    2001-07-01

    Particle histories that contribute to a detector response are categorized according to whether they are fully confined inside a source-detector enclosure or cross and recross the same enclosure. The contribution from the confined histories is expressed using a forward problem with the external boundary condition on the source-detector enclosure. The contribution from the crossing and recrossing histories is expressed as the surface integral at the same enclosure of the product of the directional cosine and the fluxes in the foregoing forward problem and the adjoint problem for the whole spatial domain. The former contribution can be calculated by a standard forward Monte Carlo. The latter contribution can be calculated by correlated coupling of forward and adjoint histories independently of the former contribution. We briefly describe the computational method and discuss its application to perturbation analysis for localized material changes. (orig.)

  7. Adjoint-Based Aerodynamic Design of Complex Aerospace Configurations

    Nielsen, Eric J.

    2016-01-01

    An overview of twenty years of adjoint-based aerodynamic design research at NASA Langley Research Center is presented. Adjoint-based algorithms provide a powerful tool for efficient sensitivity analysis of complex large-scale computational fluid dynamics (CFD) simulations. Unlike alternative approaches for which computational expense generally scales with the number of design parameters, adjoint techniques yield sensitivity derivatives of a simulation output with respect to all input parameters at the cost of a single additional simulation. With modern large-scale CFD applications often requiring millions of compute hours for a single analysis, the efficiency afforded by adjoint methods is critical in realizing a computationally tractable design optimization capability for such applications.

  8. Applications guide to the MORSE Monte Carlo code

    Cramer, S.N.

    1985-08-01

    A practical guide for the implementation of the MORESE-CG Monte Carlo radiation transport computer code system is presented. The various versions of the MORSE code are compared and contrasted, and the many references dealing explicitly with the MORSE-CG code are reviewed. The treatment of angular scattering is discussed, and procedures for obtaining increased differentiality of results in terms of reaction types and nuclides from a multigroup Monte Carlo code are explained in terms of cross-section and geometry data manipulation. Examples of standard cross-section data input and output are shown. Many other features of the code system are also reviewed, including (1) the concept of primary and secondary particles, (2) fission neutron generation, (3) albedo data capability, (4) DOMINO coupling, (5) history file use for post-processing of results, (6) adjoint mode operation, (7) variance reduction, and (8) input/output. In addition, examples of the combinatorial geometry are given, and the new array of arrays geometry feature (MARS) and its three-dimensional plotting code (JUNEBUG) are presented. Realistic examples of user routines for source, estimation, path-length stretching, and cross-section data manipulation are given. A deatiled explanation of the coupling between the random walk and estimation procedure is given in terms of both code parameters and physical analogies. The operation of the code in the adjoint mode is covered extensively. The basic concepts of adjoint theory and dimensionality are discussed and examples of adjoint source and estimator user routines are given for all common situations. Adjoint source normalization is explained, a few sample problems are given, and the concept of obtaining forward differential results from adjoint calculations is covered. Finally, the documentation of the standard MORSE-CG sample problem package is reviewed and on-going and future work is discussed

  9. Self-Adjointness Criterion for Operators in Fock Spaces

    Falconi, Marco

    2015-01-01

    In this paper we provide a criterion of essential self-adjointness for operators in the tensor product of a separable Hilbert space and a Fock space. The class of operators we consider may contain a self-adjoint part, a part that preserves the number of Fock space particles and a non-diagonal part that is at most quadratic with respect to the creation and annihilation operators. The hypotheses of the criterion are satisfied in several interesting applications

  10. Shielding design method for LMFBR validation on the Phenix factor

    Cabrillat, J.C.; Crouzet, J.; Misrakis, J.; Salvatores, M.; Rado, V.; Palmiotti, G.

    1983-05-01

    Shielding design methods, developed at CEA for shielding calculations find a global validation by the means of Phenix power reactor (250 MWe) measurements. Particularly, the secondary sodium activation of pool type LMFBR such as Super Phenix (1200 MWe) which is subject to strict safety limitation is well calculated by the adapted scheme, i.e. a two dimension transport calculation of shielding coupled to a Monte-Carlo calculation of secondary sodium activation

  11. Four-Dimensional Data Assimilation Using the Adjoint Method

    Bao, Jian-Wen

    The calculus of variations is used to confirm that variational four-dimensional data assimilation (FDDA) using the adjoint method can be implemented when the numerical model equations have a finite number of first-order discontinuous points. These points represent the on/off switches associated with physical processes, for which the Jacobian matrix of the model equation does not exist. Numerical evidence suggests that, in some situations when the adjoint method is used for FDDA, the temperature field retrieved using horizontal wind data is numerically not unique. A physical interpretation of this type of non-uniqueness of the retrieval is proposed in terms of energetics. The adjoint equations of a numerical model can also be used for model-parameter estimation. A general computational procedure is developed to determine the size and distribution of any internal model parameter. The procedure is then applied to a one-dimensional shallow -fluid model in the context of analysis-nudging FDDA: the weighting coefficients used by the Newtonian nudging technique are determined. The sensitivity of these nudging coefficients to the optimal objectives and constraints is investigated. Experiments of FDDA using the adjoint method are conducted using the dry version of the hydrostatic Penn State/NCAR mesoscale model (MM4) and its adjoint. The minimization procedure converges and the initialization experiment is successful. Temperature-retrieval experiments involving an assimilation of the horizontal wind are also carried out using the adjoint of MM4.

  12. Research on Monte Carlo improved quasi-static method for reactor space-time dynamics

    Xu Qi; Wang Kan; Li Shirui; Yu Ganglin

    2013-01-01

    With large time steps, improved quasi-static (IQS) method can improve the calculation speed for reactor dynamic simulations. The Monte Carlo IQS method was proposed in this paper, combining the advantages of both the IQS method and MC method. Thus, the Monte Carlo IQS method is beneficial for solving space-time dynamics problems of new concept reactors. Based on the theory of IQS, Monte Carlo algorithms for calculating adjoint neutron flux, reactor kinetic parameters and shape function were designed and realized. A simple Monte Carlo IQS code and a corresponding diffusion IQS code were developed, which were used for verification of the Monte Carlo IQS method. (authors)

  13. Investigating the minimum achievable variance in a Monte Carlo criticality calculation

    Christoforou, Stavros; Eduard Hoogenboom, J. [Delft University of Technology, Mekelweg 15, 2629 JB Delft (Netherlands)

    2008-07-01

    The sources of variance in a Monte Carlo criticality calculation are identified and their contributions analyzed. A zero-variance configuration is initially simulated using analytically calculated adjoint functions for biasing. From there, the various sources are analyzed. It is shown that the minimum threshold comes from the fact that the fission source is approximated. In addition, the merits of a simple variance reduction method, such as implicit capture, are shown when compared to an analog simulation. Finally, it is shown that when non-exact adjoint functions are used for biasing, the variance reduction is rather insensitive to the quality of the adjoints, suggesting that the generation of the adjoints should have as low CPU cost as possible, in order to o et the CPU cost in the implementation of the biasing of a simulation. (authors)

  14. Global adjoint tomography: first-generation model

    Bozdağ, Ebru

    2016-09-23

    We present the first-generation global tomographic model constructed based on adjoint tomography, an iterative full-waveform inversion technique. Synthetic seismograms were calculated using GPU-accelerated spectral-element simulations of global seismic wave propagation, accommodating effects due to 3-D anelastic crust & mantle structure, topography & bathymetry, the ocean load, ellipticity, rotation, and self-gravitation. Fréchet derivatives were calculated in 3-D anelastic models based on an adjoint-state method. The simulations were performed on the Cray XK7 named \\'Titan\\', a computer with 18 688 GPU accelerators housed at Oak Ridge National Laboratory. The transversely isotropic global model is the result of 15 tomographic iterations, which systematically reduced differences between observed and simulated three-component seismograms. Our starting model combined 3-D mantle model S362ANI with 3-D crustal model Crust2.0. We simultaneously inverted for structure in the crust and mantle, thereby eliminating the need for widely used \\'crustal corrections\\'. We used data from 253 earthquakes in the magnitude range 5.8 ≤ M ≤ 7.0. We started inversions by combining ~30 s body-wave data with ~60 s surface-wave data. The shortest period of the surface waves was gradually decreased, and in the last three iterations we combined ~17 s body waves with ~45 s surface waves. We started using 180 min long seismograms after the 12th iteration and assimilated minor- and major-arc body and surface waves. The 15th iteration model features enhancements of well-known slabs, an enhanced image of the Samoa/Tahiti plume, as well as various other plumes and hotspots, such as Caroline, Galapagos, Yellowstone and Erebus. Furthermore, we see clear improvements in slab resolution along the Hellenic and Japan Arcs, as well as subduction along the East of Scotia Plate, which does not exist in the starting model. Point-spread function tests demonstrate that we are approaching the

  15. Infrared conformality and bulk critical points: SU(2) with heavy adjoint quarks

    Lucini, Biagio; Rago, Antonio; Rinaldi, Enrico

    2013-01-01

    The lattice phase structure of a gauge theory can be a serious obstruction to Monte Carlo studies of its continuum behaviour. This issue is particularly delicate when numerical studies are performed to determine whether a theory is in a (near-)conformal phase. In this work we investigate the heavy mass limit of the SU(2) gauge theory with Nf=2 adjoint fermions and its lattice phase diagram, showing the presence of a critical point ending a line of first order bulk phase transition. The relevant gauge observables and the low-lying spectrum are monitored in the vicinity of the critical point with very good control over different systematic effects. The scaling properties of masses and susceptibilities open the possibility that the effective theory at criticality is a scalar theory in the universality class of the four-dimensional Gaussian model. This behaviour is clearly different from what is observed for SU(2) gauge theory with two dynamical adjoint fermions, whose (near-)conformal numerical signature is henc...

  16. Manipulating Rayleigh-Taylor Growth Using Adjoints

    Kord, Ali; Capecelatro, Jesse

    2017-11-01

    It has been observed that initial interfacial perturbations affect the growth of Rayleigh-Taylor (RT) instabilities. However, it remains to be seen to what extent the perturbations alter the RT growth rate. Direct numerical simulations (DNS) provide a powerful means for studying the effects of initial conditions (IC) on the growth rate. However, a brute-force approach for identifying optimal initial perturbations is not practical via DNS. In addition, identifying sensitivity of the RT growth to the large number of parameters used in defining the IC is computationally expensive. A discrete adjoint is formulated to measure sensitivities of multi-mode RT growth to ICs in a high-order finite difference framework. The sensitivity is used as a search direction for adjusting the initial perturbations to both maximize and suppress the RT growth rate during its non-linear regime. The modes that contribute the greatest sensitivity are identified, and optimized perturbation energy spectrum are reported. PhD Student, Department of Mechanical Engineering, University of Michigan, Ann Arbor, MI.

  17. Shielding calculational system for plutonium

    Zimmerman, M.G.; Thomsen, D.H.

    1975-08-01

    A computer calculational system has been developed and assembled specifically for calculating dose rates in AEC plutonium fabrication facilities. The system consists of two computer codes and all nuclear data necessary for calculation of neutron and gamma dose rates from plutonium. The codes include the multigroup version of the Battelle Monte Carlo code for solution of general neutron and gamma shielding problems and the PUSHLD code for solution of shielding problems where low energy gamma and x-rays are important. The nuclear data consists of built in neutron and gamma yields and spectra for various plutonium compounds, an automatic calculation of age effects and all cross-sections commonly used. Experimental correlations have been performed to verify portions of the calculational system. (23 tables, 7 figs, 16 refs) (U.S.)

  18. An eddy-permitting, dynamically consistent adjoint-based assimilation system for the tropical Pacific: Hindcast experiments in 2000

    Hoteit, Ibrahim; Cornuelle, B.; Heimbach, P.

    2010-01-01

    An eddy-permitting adjoint-based assimilation system has been implemented to estimate the state of the tropical Pacific Ocean. The system uses the Massachusetts Institute of Technology's general circulation model and its adjoint. The adjoint method

  19. About the Scythian Shields

    About the Scythian Shields

    2017-10-01

    Full Text Available Shields played major role in the armament system of the Scythians. Made from organic materials, they are poorly traced on the materials of archaeological excavations. Besides, scaly surface of shields was often perceived in practice as the remnants of the scaly armor. E. V. Chernenko was able to discern the difference between shields’ scaly plates and armor scales. The top edge of the scales was bent inwards, and shield plates had a wire fixation. These observations let significantly increase the number of shields, found in the burial complexes of the Scythians. The comparison of archaeological materials and the images of Scythian warriors allow distinguishing the main forms of Scythian shields. All shields are divided into fencing shields and cover shields. The fencing shields include round wooden shields, reinforced with bronze sheet, and round moon-shaped shields with a notch at the top, with a metal scaly surface. They came to the Scythians under the Greek influence and are known in the monuments of the 4th century BC. Oval shields with scaly surface (back cover shields were used by the Scythian cavalry. They protected the rider in case of frontal attack, and moved back in case of maneuver or closein fighting. Scythian battle tactics were based on rapid approaching the enemy and throwing spears and further rapid withdrawal. Spears stuck in the shields of enemies, forcing them to drop the shields, uncover, and in this stage of the battle the archers attacked the disorganized ranks of the enemy. That was followed by the stage of close fight. Oval form of a wooden shield with leather covering was used by the Scythian infantry and spearmen. Rectangular shields, including wooden shields and the shields pleached from rods, represented a special category. The top of such shield was made of wood, and a pleached pad on leather basis was attached to it. This shield could be a reliable protection from arrows, but it could not protect against javelins

  20. Adjoint current-based approaches to prostate brachytherapy optimization

    Roberts, J. A.; Henderson, D. L.

    2009-01-01

    This paper builds on previous work done at the Univ. of Wisconsin - Madison to employ the adjoint concept of nuclear reactor physics in the so-called greedy heuristic of brachytherapy optimization. Whereas that previous work focused on the adjoint flux, i.e. the importance, this work has included use of the adjoint current to increase the amount of information available in optimizing. Two current-based approaches were developed for 2-D problems, and each was compared to the most recent form of the flux-based methodology. The first method aimed to take a treatment plan from the flux-based greedy heuristic and adjust via application of the current-displacement, or a vector displacement based on a combination of tissue (adjoint) and seed (forward) currents acting as forces on a seed. This method showed promise in improving key urethral and rectal dosimetric quantities. The second method uses the normed current-displacement as the greedy criterion such that seeds are placed in regions of least force. This method, coupled with the dose-update scheme, generated treatment plans with better target irradiation and sparing of the urethra and normal tissues than the flux-based approach. Tables of these parameters are given for both approaches. In summary, these preliminary results indicate adjoint current methods are useful in optimization and further work in 3-D should be performed. (authors)

  1. Implementation of Generalized Adjoint Equation Solver for DeCART

    Han, Tae Young; Cho, Jin Young; Lee, Hyun Chul; Noh, Jae Man

    2013-01-01

    In this paper, the generalized adjoint solver based on the generalized perturbation theory is implemented on DeCART and the verification calculations were carried out. As the results, the adjoint flux for the general response coincides with the reference solution and it is expected that the solver could produce the parameters for the sensitivity and uncertainty analysis. Recently, MUSAD (Modules of Uncertainty and Sensitivity Analysis for DeCART) was developed for the uncertainty analysis of PMR200 core and the fundamental adjoint solver was implemented into DeCART. However, the application of the code was limited to the uncertainty to the multiplication factor, k eff , because it was based on the classical perturbation theory. For the uncertainty analysis to the general response as like the power density, it is necessary to develop the analysis module based on the generalized perturbation theory and it needs the generalized adjoint solutions from DeCART. In this paper, the generalized adjoint solver is implemented on DeCART and the calculation results are compared with the results by TSUNAMI of SCALE 6.1

  2. Sensitivity Analysis for Steady State Groundwater Flow Using Adjoint Operators

    Sykes, J. F.; Wilson, J. L.; Andrews, R. W.

    1985-03-01

    Adjoint sensitivity theory is currently being considered as a potential method for calculating the sensitivity of nuclear waste repository performance measures to the parameters of the system. For groundwater flow systems, performance measures of interest include piezometric heads in the vicinity of a waste site, velocities or travel time in aquifers, and mass discharge to biosphere points. The parameters include recharge-discharge rates, prescribed boundary heads or fluxes, formation thicknesses, and hydraulic conductivities. The derivative of a performance measure with respect to the system parameters is usually taken as a measure of sensitivity. To calculate sensitivities, adjoint sensitivity equations are formulated from the equations describing the primary problem. The solution of the primary problem and the adjoint sensitivity problem enables the determination of all of the required derivatives and hence related sensitivity coefficients. In this study, adjoint sensitivity theory is developed for equations of two-dimensional steady state flow in a confined aquifer. Both the primary flow equation and the adjoint sensitivity equation are solved using the Galerkin finite element method. The developed computer code is used to investigate the regional flow parameters of the Leadville Formation of the Paradox Basin in Utah. The results illustrate the sensitivity of calculated local heads to the boundary conditions. Alternatively, local velocity related performance measures are more sensitive to hydraulic conductivities.

  3. Adjoint-consistent formulations of slip models for coupled electroosmotic flow systems

    Garg, Vikram V; Prudhomme, Serge; van der Zee, Kris G; Carey, Graham F

    2014-01-01

    Models based on the Helmholtz `slip' approximation are often used for the simulation of electroosmotic flows. The objectives of this paper are to construct adjoint-consistent formulations of such models, and to develop adjoint

  4. Adjoint-consistent formulations of slip models for coupled electroosmotic flow systems

    Garg, Vikram V

    2014-09-27

    Background Models based on the Helmholtz `slip\\' approximation are often used for the simulation of electroosmotic flows. The objectives of this paper are to construct adjoint-consistent formulations of such models, and to develop adjoint-based numerical tools for adaptive mesh refinement and parameter sensitivity analysis. Methods We show that the direct formulation of the `slip\\' model is adjoint inconsistent, and leads to an ill-posed adjoint problem. We propose a modified formulation of the coupled `slip\\' model, which is shown to be well-posed, and therefore automatically adjoint-consistent. Results Numerical examples are presented to illustrate the computation and use of the adjoint solution in two-dimensional microfluidics problems. Conclusions An adjoint-consistent formulation for Helmholtz `slip\\' models of electroosmotic flows has been proposed. This formulation provides adjoint solutions that can be reliably used for mesh refinement and sensitivity analysis.

  5. Does one see gluon condensation after subtraction of mean field perturbation theory from Monte Carlo data

    Schlichting, H.

    1985-01-01

    We do a linearised mean field calculation in axial gauge for the four dimensional mixed fundamental adjoint SU(2) lattice gauge theory and extract the gluon condensate parameter from the expectation values of the plaquette and the action by subtracting mean field perturbation theory from Monte Carlo data. (orig.)

  6. Radiation shielding device

    Nakagawa, Takahiro; Yamagami, Makoto.

    1996-01-01

    A fixed shielding member made of a radiation shielding material is constituted in perpendicular to an opening formed on radiation shielding walls. The fixed shielding member has one side opened and has other side, the upper portion and the lower portion disposed in close contact with the radiation shielding walls. Movable shielding members made of a radiation shielding material are each disposed openably on both side of the fixed shielding member. The movable shielding member has a shaft as a fulcrum on one side thereof for connecting it to the radiation shielding walls. The other side has a handle attached for opening/closing the movable shielding member. Upon access of an operator, when each one of the movable shielding members is opened/closed on every time, leakage of linear or scattered radiation can be prevented. Even when both of the movable shielding members are opened simultaneously, the fixed shielding member and the movable shielding members form labyrinth to prevent leakage of linear radioactivity. (I.N.)

  7. Normal and adjoint integral and integrodifferential neutron transport equations. Pt. 2

    Velarde, G.

    1976-01-01

    Using the simplifying hypotheses of the integrodifferential Boltzmann equations of neutron transport, given in JEN 334 report, several integral equations, and theirs adjoint ones, are obtained. Relations between the different normal and adjoint eigenfunctions are established and, in particular, proceeding from the integrodifferential Boltzmann equation it's found out the relation between the solutions of the adjoint equation of its integral one, and the solutions of the integral equation of its adjoint one (author)

  8. Cosmic Ray Interactions in Shielding Materials

    Aguayo Navarrete, Estanislao; Kouzes, Richard T.; Ankney, Austin S.; Orrell, John L.; Berguson, Timothy J.; Troy, Meredith D.

    2011-01-01

    This document provides a detailed study of materials used to shield against the hadronic particles from cosmic ray showers at Earth's surface. This work was motivated by the need for a shield that minimizes activation of the enriched germanium during transport for the MAJORANA collaboration. The materials suitable for cosmic-ray shield design are materials such as lead and iron that will stop the primary protons, and materials like polyethylene, borated polyethylene, concrete and water that will stop the induced neutrons. The interaction of the different cosmic-ray components at ground level (protons, neutrons, muons) with their wide energy range (from kilo-electron volts to giga-electron volts) is a complex calculation. Monte Carlo calculations have proven to be a suitable tool for the simulation of nucleon transport, including hadron interactions and radioactive isotope production. The industry standard Monte Carlo simulation tool, Geant4, was used for this study. The result of this study is the assertion that activation at Earth's surface is a result of the neutronic and protonic components of the cosmic-ray shower. The best material to shield against these cosmic-ray components is iron, which has the best combination of primary shielding and minimal secondary neutron production.

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

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

    2006-07-01

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

  10. Methods for calculating radiation attenuation in shields

    Butler, J; Bueneman, D; Etemad, A; Lafore, P; Moncassoli, A M; Penkuhn, H; Shindo, M; Stoces, B

    1964-10-01

    In recent years the development of high-speed digital computers of large capacity has revolutionized the field of reactor shield design. For compact special-purpose reactor shields, Monte-Carlo codes in two- and three dimensional geometries are now available for the proper treatment of both the neutron and gamma- ray problems. Furthermore, techniques are being developed for the theoretical optimization of minimum-weight shield configurations for this type of reactor system. In the design of land-based power reactors, on the other hand, there is a strong incentive to reduce the capital cost of the plant, and economic considerations are also relevant to reactors designed for merchant ship propulsion. In this context simple methods are needed which are economic in their data input and computing time requirements and which, at the same time, are sufficiently accurate for design work. In general the computing time required for Monte-Carlo calculations in complex geometry is excessive for routine design calculations and the capacity of the present codes is inadequate for the proper treatment of large reactor shield systems in three dimensions. In these circumstances a wide range of simpler techniques are currently being employed for design calculations. The methods of calculation for neutrons in reactor shields fall naturally into four categories: Multigroup diffusion theory; Multigroup diffusion with removal sources; Transport codes; and Monte Carlo methods. In spite of the numerous Monte- Carlo techniques which are available for penetration and back scattering, serious problems are still encountered in practice with the scattering of gamma rays from walls of buildings which contain critical facilities and also concrete-lined discharge shafts containing irradiated fuel elements. The considerable volume of data in the unclassified literature on the solution of problems of this type in civil defence work appears not to have been evaluated for reactor shield design. In

  11. FLUKA shielding calculations for the FAIR project

    Fehrenbacher, Georg; Kozlova, Ekaterina; Radon, Torsten; Sokolov, Alexey

    2015-01-01

    FAIR is an international accelerator project being in construction at GSI Helmholtz center for heavy ion research in Darmstadt. The Monte Carlo program FLUKA is used to study radiation protection problems. The contribution deals with general application possibilities of FLUKA and for FAIR with respect the radiation protection planning. The necessity to simulate the radiation transport through shielding of several meters thickness and to determine the equivalent doses outside the shielding with sufficient accuracy is demonstrated using two examples under consideration of the variance reduction. Results of simulation calculations for activation estimation in accelerator facilities are presented.

  12. Implementation of an approximate zero-variance scheme in the TRIPOLI Monte Carlo code

    Christoforou, S.; Hoogenboom, J. E. [Delft Univ. of Technology, Mekelweg 15, 2629 JB Delft (Netherlands); Dumonteil, E.; Petit, O.; Diop, C. [Commissariat a l' Energie Atomique CEA, Gif-sur-Yvette (France)

    2006-07-01

    In an accompanying paper it is shown that theoretically a zero-variance Monte Carlo scheme can be devised for criticality calculations if the space, energy and direction dependent adjoint function is exactly known. This requires biasing of the transition and collision kernels with the appropriate adjoint function. In this paper it is discussed how an existing general purpose Monte Carlo code like TRIPOLI can be modified to approach the zero-variance scheme. This requires modifications for reading in the adjoint function obtained from a separate deterministic calculation for a number of space intervals, energy groups and discrete directions. Furthermore, a function has to be added to supply the direction dependent and the averaged adjoint function at a specific position in the system by interpolation. The initial particle weights of a certain batch must be set inversely proportional to the averaged adjoint function and proper normalization of the initial weights must be secured. The sampling of the biased transition kernel requires cumulative integrals of the biased kernel along the flight path until a certain value, depending on a selected random number is reached to determine a new collision site. The weight of the particle must be adapted accordingly. The sampling of the biased collision kernel (in a multigroup treatment) is much more like the normal sampling procedure. A numerical example is given for a 3-group calculation with a simplified transport model (two-direction model), demonstrating that the zero-variance scheme can be approximated quite well for this simplified case. (authors)

  13. The Spallation Neutron Source (SNS) conceptual design shielding analysis

    Johnson, J.O.; Odano, N.; Lillie, R.A.

    1998-03-01

    The shielding design is important for the construction of an intense high-energy accelerator facility like the proposed Spallation Neutron Source (SNS) due to its impact on conventional facility design, maintenance operations, and since the cost for the radiation shielding shares a considerable part of the total facility costs. A calculational strategy utilizing coupled high energy Monte Carlo calculations and multi-dimensional discrete ordinates calculations, along with semi-empirical calculations, was implemented to perform the conceptual design shielding assessment of the proposed SNS. Biological shields have been designed and assessed for the proton beam transport system and associated beam dumps, the target station, and the target service cell and general remote maintenance cell. Shielding requirements have been assessed with respect to weight, space, and dose-rate constraints for operating, shutdown, and accident conditions. A discussion of the proposed facility design, conceptual design shielding requirements calculational strategy, source terms, preliminary results and conclusions, and recommendations for additional analyses are presented

  14. Non-self-adjoint hamiltonians defined by Riesz bases

    Bagarello, F., E-mail: fabio.bagarello@unipa.it [Dipartimento di Energia, Ingegneria dell' Informazione e Modelli Matematici, Facoltà di Ingegneria, Università di Palermo, I-90128 Palermo, Italy and INFN, Università di Torino, Torino (Italy); Inoue, A., E-mail: a-inoue@fukuoka-u.ac.jp [Department of Applied Mathematics, Fukuoka University, Fukuoka 814-0180 (Japan); Trapani, C., E-mail: camillo.trapani@unipa.it [Dipartimento di Matematica e Informatica, Università di Palermo, I-90123 Palermo (Italy)

    2014-03-15

    We discuss some features of non-self-adjoint Hamiltonians with real discrete simple spectrum under the assumption that the eigenvectors form a Riesz basis of Hilbert space. Among other things, we give conditions under which these Hamiltonians can be factorized in terms of generalized lowering and raising operators.

  15. Nefness of adjoint bundles for ample vector bundles

    Hidetoshi Maeda

    1995-11-01

    Full Text Available Let E be an ample vector bundle of rank >1 on a smooth complex projective variety X of dimension n. This paper gives a classification of pairs (X,E whose adjoint bundles K_X+det E are not nef in the case when  r=n-2.

  16. Adjoint sensitivity analysis of high frequency structures with Matlab

    Bakr, Mohamed; Demir, Veysel

    2017-01-01

    This book covers the theory of adjoint sensitivity analysis and uses the popular FDTD (finite-difference time-domain) method to show how wideband sensitivities can be efficiently estimated for different types of materials and structures. It includes a variety of MATLAB® examples to help readers absorb the content more easily.

  17. Assimilating Remote Ammonia Observations with a Refined Aerosol Thermodynamics Adjoint"

    Ammonia emissions parameters in North America can be refined in order to improve the evaluation of modeled concentrations against observations. Here, we seek to do so by developing and applying the GEOS-Chem adjoint nested over North America to conductassimilation of observations...

  18. Toward regional-scale adjoint tomography in the deep earth

    Masson, Y.; Romanowicz, B. A.

    2013-12-01

    Thanks to the development of efficient numerical computation methods, such as the Spectral Element Method (SEM) and to the increasing power of computer clusters, it is now possible to obtain regional-scale images of the Earth's interior using adjoint-tomography (e.g. Tape, C., et al., 2009). As for now, these tomographic models are limited to the upper layers of the earth, i.e., they provide us with high-resolution images of the crust and the upper part of the mantle. Given the gigantic amount of calculation it represents, obtaing similar models at the global scale (i.e. images of the entire Earth) seems out of reach at the moment. Furthermore, it's likely that the first generation of such global adjoint tomographic models will have a resolution significantly smaller than the current regional models. In order to image regions of interests in the deep Earth, such as plumes, slabs or large low shear velocity provinces (LLSVPs), while keeping the computation tractable, we are developing new tools that will allow us to perform regional-scale adjoint-tomography at arbitrary depths. In a recent study (Masson et al., 2013), we showed that a numerical equivalent of the time reversal mirrors used in experimental acoustics permits to confine the wave propagation computations (i.e. using SEM simulations) inside the region to be imaged. With this ability to limit wave propagation modeling inside a region of interest, obtaining the adjoint sensitivity kernels needed for tomographic imaging is only two steps further. First, the local wavefield modeling needs to be coupled with field extrapolation techniques in order to obtain synthetic seismograms at the surface of the earth. These seismograms will account for the 3D structure inside the region of interest in a quasi-exact manner. We will present preliminary results where the field-extrapolation is performed using Green's function computed in a 1D Earth model thanks to the Direct Solution Method (DSM). Once synthetic seismograms

  19. Handout on shielding calculation

    Heilbron Filho, P.F.L.

    1991-01-01

    In order to avoid the difficulties of the radioprotection supervisors in the tasks related to shielding calculations, is presented in this paper the basic concepts of shielding theory. It also includes exercises and examples. (author)

  20. Approximation for the adjoint neutron spectrum; Aproximacao para o espectro adjunto de neutrons

    Suster, Luis Carlos; Martinez, Aquilino Senra; Silva, Fernando Carvalho da [Universidade Federal, Rio de Janeiro, RJ (Brazil). Coordenacao dos Programas de Pos-graduacao de Engenharia. Programa de Engenharia Nuclear

    2002-07-01

    The proposal of this work is the determination of an analytical approximation which is capable to reproduce the adjoint neutron flux for the energy range of the narrow resonances (NR). In a previous work we developed a method for the calculation of the adjoint spectrum which was calculated from the adjoint neutron balance equations, that were obtained by the collision probabilities method, this method involved a considerable quantity of numerical calculation. In the analytical method some approximations were done, like the multiplication of the escape probability in the fuel by the adjoint flux in the moderator, and after these approximations, taking into account the case of the narrow resonances, were substituted in the adjoint neutron balance equation for the fuel, resulting in an analytical approximation for the adjoint flux. The results obtained in this work were compared to the results generated with the reference method, which demonstrated a good and precise results for the adjoint neutron flux for the narrow resonances. (author)

  1. Discrete adjoint of fractional step Navier-Stokes solver in generalized coordinates

    Wang, Mengze; Mons, Vincent; Zaki, Tamer

    2017-11-01

    Optimization and control in transitional and turbulent flows require evaluation of gradients of the flow state with respect to the problem parameters. Using adjoint approaches, these high-dimensional gradients can be evaluated with a similar computational cost as the forward Navier-Stokes simulations. The adjoint algorithm can be obtained by discretizing the continuous adjoint Navier-Stokes equations or by deriving the adjoint to the discretized Navier-Stokes equations directly. The latter algorithm is necessary when the forward-adjoint relations must be satisfied to machine precision. In this work, our forward model is the fractional step solution to the Navier-Stokes equations in generalized coordinates, proposed by Rosenfeld, Kwak & Vinokur. We derive the corresponding discrete adjoint equations. We also demonstrate the accuracy of the combined forward-adjoint model, and its application to unsteady wall-bounded flows. This work has been partially funded by the Office of Naval Research (Grant N00014-16-1-2542).

  2. Design of emergency shield

    Soliman, S.E.

    1993-01-01

    Manufacturing of an emergency movable shield in the hot laboratories center is urgently needed for the safety of personnel in case of accidents or spilling of radioactive materials. In this report, a full design for an emergency shield is presented and the corresponding dose rates behind the shield for different activities (from 1 mCi to 5 Ci) was calculated by using micro shield computer code. 4 figs., 1 tab

  3. A study on the apron shielding ratio according to electromagnetic radiation energy

    Jang, Dong Gun; Lee, Sang Ho; Choi, Hyung Seok; Son, Joo Chul; Yoon, Chang Yong; Ji, Yung Sik; Cho, Yong In; Lee, Hong Je; Yang, Seoung Oh

    2014-01-01

    The medical institution has been used electromagnetic radiation of various energy. But researchers are divided on whether using apron for radiation shielding will be effective or not. The purpose of present study was to analyze electromagnetic radiation shielding effect of apron by using Monte Carlo simulation. 1 MBq electromagnetic radiation was emitted from 10-500 keV at 10 keV increments in Monte Carlo simulation. Then shielded radiation dose difference was confirmed, when 0.25 mmPb shield use for shielding. As a results, shielding ratio was markedly decreased in high energy electromagnetic radiation. The radiation dose was inversely increased with 0.25 mmPb shielding

  4. A study on the apron shielding ratio according to electromagnetic radiation energy

    Jang, Dong Gun; Lee, Sang Ho; Choi, Hyung Seok; Son, Joo Chul; Yoon, Chang Yong; Ji, Yung Sik; Cho, Yong In; Lee, Hong Je; Yang, Seoung Oh [Dept. of Nuclear Medicine, Dongnam Institute of Radiological and Medical Sciences Cancer Center, Busan (Korea, Republic of)

    2014-12-15

    The medical institution has been used electromagnetic radiation of various energy. But researchers are divided on whether using apron for radiation shielding will be effective or not. The purpose of present study was to analyze electromagnetic radiation shielding effect of apron by using Monte Carlo simulation. 1 MBq electromagnetic radiation was emitted from 10-500 keV at 10 keV increments in Monte Carlo simulation. Then shielded radiation dose difference was confirmed, when 0.25 mmPb shield use for shielding. As a results, shielding ratio was markedly decreased in high energy electromagnetic radiation. The radiation dose was inversely increased with 0.25 mmPb shielding.

  5. Electromagnetically shielded building

    Takahashi, T.; Nakamura, M.; Yabana, Y.; Ishikawa, T.; Nagata, K.

    1992-01-01

    This invention relates to a building having an electromagnetic shield structure well-suited for application to an information network system utilizing electromagnetic waves, and more particularly to an electromagnetically shielded building for enhancing the electromagnetic shielding performance of an external wall. 6 figs

  6. Electromagnetically shielded building

    Takahashi, T; Nakamura, M; Yabana, Y; Ishikawa, T; Nagata, K

    1992-04-21

    This invention relates to a building having an electromagnetic shield structure well-suited for application to an information network system utilizing electromagnetic waves, and more particularly to an electromagnetically shielded building for enhancing the electromagnetic shielding performance of an external wall. 6 figs.

  7. Electromagnetic shielding formulae

    Dahlberg, E.

    1979-02-01

    This addendum to an earlier collection of electromagnetic shielding formulae (TRITA-EPP-75-27) contains simple transfer matrices suitable for calculating the quasistatic shielding efficiency for multiple transverse-field and axial-field cylindrical and spherical shields, as well as for estimating leakage fields from long coaxial cables and the normal-incidence transmission of a plane wave through a multiple plane shield. The differences and similarities between these cases are illustrated by means of equivalent circuits and transmission line analogies. The addendum also includes a discussion of a possible heuristic improvement of some shielding formulae. (author)

  8. The Active Muon Shield

    Bezshyiko, Iaroslava

    2016-01-01

    In the SHiP beam-dump of the order of 1011 muons will be produced per second. An active muon-shield is used to magnetically deflect these muons out of the acceptance of the spectrom- eter. This note describes how this shield is modelled and optimized. The SHiP spectrometer is being re-optimized using a conical decay-vessel, and utilizing the possibility to magnetize part of the beam-dump shielding iron. A shield adapted to these new conditions is presented which is significantly shorter and lighter than the shield used in the Technical Proposal (TP), while showing a similar performance.

  9. Computational methods for high-energy source shielding

    Armstrong, T.W.; Cloth, P.; Filges, D.

    1983-01-01

    The computational methods for high-energy radiation transport related to shielding of the SNQ-spallation source are outlined. The basic approach is to couple radiation-transport computer codes which use Monte Carlo methods and discrete ordinates methods. A code system is suggested that incorporates state-of-the-art radiation-transport techniques. The stepwise verification of that system is briefly summarized. The complexity of the resulting code system suggests a more straightforward code specially tailored for thick shield calculations. A short guide line to future development of such a Monte Carlo code is given

  10. Convergence acceleration in the Monte-Carlo particle transport code TRIPOLI-4 in criticality

    Dehaye, Benjamin

    2014-01-01

    Fields such as criticality studies need to compute some values of interest in neutron physics. Two kind of codes may be used: deterministic ones and stochastic ones. The stochastic codes do not require approximation and are thus more exact. However, they may require a lot of time to converge with a sufficient precision.The work carried out during this thesis aims to build an efficient acceleration strategy in the TRIPOLI-4. We wish to implement the zero variance game. To do so, the method requires to compute the adjoint flux. The originality of this work is to directly compute the adjoint flux directly from a Monte-Carlo simulation without using external codes thanks to the fission matrix method. This adjoint flux is then used as an importance map to bias the simulation. (author) [fr

  11. Perturbation of self-adjoint operators by Dirac distributions

    Zorbas, J.

    1980-01-01

    The existence of a family of self-adjoint Hamiltonians H/sub theta/, theta element of [0, 2π), corresponding to the formal expression H 0 +νdelta (x) is shown for a general class of self-adjoint operators H 0 . Expressions for the Green's function and wavefunction corresponding to H/sub theta/ are obtained in terms of the Green's function and wavefunction corresponding to H 0 . Similar results are shown for the perturbation of H 0 by a finite sum of Dirac distributions. A prescription is given for obtaining H/sub theta/ as the strong resolvent limit of a family of momentum cutoff Hamiltonians H/sup N/. The relationship between the scattering theories corresponding to H/sup N/ and H/sub theta/ is examined

  12. Forward and adjoint sensitivity computation of chaotic dynamical systems

    Wang, Qiqi, E-mail: qiqi@mit.edu [Department of Aeronautics and Astronautics, MIT, 77 Mass Ave., Cambridge, MA 02139 (United States)

    2013-02-15

    This paper describes a forward algorithm and an adjoint algorithm for computing sensitivity derivatives in chaotic dynamical systems, such as the Lorenz attractor. The algorithms compute the derivative of long time averaged “statistical” quantities to infinitesimal perturbations of the system parameters. The algorithms are demonstrated on the Lorenz attractor. We show that sensitivity derivatives of statistical quantities can be accurately estimated using a single, short trajectory (over a time interval of 20) on the Lorenz attractor.

  13. Global Linear Representations of Nonlinear Systems and the Adjoint Map

    Banks, S.P.

    1988-01-01

    In this paper we shall study the global linearization of nonlinear systems on a manifold by two methods. The first consists of an expansion of the vector field in the space of square integrable vector fields. In the second method we use the adjoint representation of the Lie algebra vector fields to obtain an infinite-dimensional matrix representation of the system. A connection between the two approaches will be developed.

  14. Four-fermi anomalous dimension with adjoint fermions

    Del Debbio, Luigi; Ruano, Carlos Pena

    2014-01-01

    The four-fermi interaction can play an important role in models of strong dynamical EW sym- metry breaking if the anomalous dimensions of the four-fermi operators become large in the IR. We discuss a number of issues that are relevant for the nonperturbative computation of the four- fermi anomalous dimensions for the SU(2) gauge theory with two flavors of Dirac fermions in the adjoint representation, using a Schrödinger functional formalism.

  15. Shielding for neutrons produced by medical linear accelerators

    Rebello, Wilson F.; Silva, Ademir X.

    2007-01-01

    The shielding system called Multileaf Shielding (MLS) was designed in Brazil to be used for protection patients, who undergo radiotherapy treatment, against undesired neutrons produced in the medical linear accelerator heads. During the conceiving of the MLS it was necessary to evaluate its efficiency. For that purpose, several simulations using the Monte Carlo N-particle radiation transport code, MCNP5, were made, in order to evaluate the response of the new shielding system. The results showed a significant neutron dose reduction after the inclusion of the MLS. This work aims to presenting these simulation results. (author)

  16. Shielding calculations for the Intense Neutron Source Facility. Final report

    Battat, M.E.; Henninger, R.J.; Macdonald, J.L.; Dudziak, D.J.

    1978-06-01

    Results of shielding calculations for the Intnse Neutron Source (INS) facility are presented. The INS facility is designed to house two sources, each of which will produce D--T neutrons with intensities in the range from 1 to 3 x 10 15 n/s on a continuous basis. Topics covered include the design of the biological shield, use of two-dimensional discrete-ordinates results to specify the source terms for a Monte Carlo skyshine calculation, air activation, and dose rates in the source cell (after shutdown) due to activation of the biological shield

  17. Automatic generation of biasing parameters for MORSE shielding problems

    Hoogenboom, J.E.

    1995-01-01

    It would be favourable if the biasing functions could be obtained from the Monte Carlo calculation itself. This is discussed in this paper as well as the way to obtain biasing parameters from it for splitting, Russian roulette and path length stretching. The method is demonstrated for a shielding problem solved with the MORSE-SGC/S Monte Carlo code of the SCALE-system. (K.A.)

  18. Determination of the potential for fundamental- and adjoint-representation sources in SU(2) in three dimensions

    Mawhinney, R.D.

    1990-01-01

    Pure SU(2) lattice gauge theory in three dimensions is studied by Monte Carlo simulation with a determination of the potential between fundamental- and adjoint-representation sources as a major goal. A 32 3 lattice is used and Wilson loops up to 16 by 16 are measured using a modification to the standard multihit variance reduction which improves the statistics by at least a factor of 3 at β=6.0. The integrated autocorrelation times measured for the loops show a peak for loops of size β by β. The fundamental- and adjoint-representation potentials are seen to have the same functional form to very high accuracy and their numerical values are in the ratio of their Casimir operators. The string tension is extracted and scaling is seen to within a few percent over a range of couplings which correspond to a factor of 2 change in the glueball mass. Correlated errors have been taken into account in the extraction of the potentials from the Wilson-loop values

  19. Detector placement optimization for cargo containers using deterministic adjoint transport examination for SNM detection

    McLaughlin, Trevor D.; Sjoden, Glenn E.; Manalo, Kevin L.

    2011-01-01

    With growing concerns over port security and the potential for illicit trafficking of SNM through portable cargo shipping containers, efforts are ongoing to reduce the threat via container monitoring. This paper focuses on answering an important question of how many detectors are necessary for adequate coverage of a cargo container considering the detection of neutrons and gamma rays. Deterministic adjoint transport calculations are performed with compressed helium- 3 polyethylene moderated neutron detectors and sodium activated cesium-iodide gamma-ray scintillation detectors on partial and full container models. Results indicate that the detector capability is dependent on source strength and potential shielding. Using a surrogate weapons grade plutonium leakage source, it was determined that for a 20 foot ISO container, five neutron detectors and three gamma detectors are necessary for adequate coverage. While a large CsI(Na) gamma detector has the potential to monitor the entire height of the container for SNM, the He-3 neutron detector is limited to roughly 1.25 m in depth. Detector blind spots are unavoidable inside the container volume unless additional measures are taken for adequate coverage. (author)

  20. Spectral monodromy of non-self-adjoint operators

    Phan, Quang Sang

    2014-01-01

    In the present paper, we build a combinatorial invariant, called the “spectral monodromy” from the spectrum of a single (non-self-adjoint) h-pseudodifferential operator with two degrees of freedom in the semi-classical limit. Our inspiration comes from the quantum monodromy defined for the joint spectrum of an integrable system of n commuting self-adjoint h-pseudodifferential operators, given by S. Vu Ngoc [“Quantum monodromy in integrable systems,” Commun. Math. Phys. 203(2), 465–479 (1999)]. The first simple case that we treat in this work is a normal operator. In this case, the discrete spectrum can be identified with the joint spectrum of an integrable quantum system. The second more complex case we propose is a small perturbation of a self-adjoint operator with a classical integrability property. We show that the discrete spectrum (in a small band around the real axis) also has a combinatorial monodromy. The main difficulty in this case is that we do not know the description of the spectrum everywhere, but only in a Cantor type set. In addition, we also show that the corresponding monodromy can be identified with the classical monodromy, defined by J. Duistermaat [“On global action-angle coordinates,” Commun. Pure Appl. Math. 33(6), 687–706 (1980)

  1. Spectral monodromy of non-self-adjoint operators

    Phan, Quang Sang, E-mail: quang.phan@uj.edu.pl [Université de Rennes 1, Institut de Recherche Mathématique de Rennes (UMR 6625), Campus de Beaulieu, 35042 Rennes (France)

    2014-01-15

    In the present paper, we build a combinatorial invariant, called the “spectral monodromy” from the spectrum of a single (non-self-adjoint) h-pseudodifferential operator with two degrees of freedom in the semi-classical limit. Our inspiration comes from the quantum monodromy defined for the joint spectrum of an integrable system of n commuting self-adjoint h-pseudodifferential operators, given by S. Vu Ngoc [“Quantum monodromy in integrable systems,” Commun. Math. Phys. 203(2), 465–479 (1999)]. The first simple case that we treat in this work is a normal operator. In this case, the discrete spectrum can be identified with the joint spectrum of an integrable quantum system. The second more complex case we propose is a small perturbation of a self-adjoint operator with a classical integrability property. We show that the discrete spectrum (in a small band around the real axis) also has a combinatorial monodromy. The main difficulty in this case is that we do not know the description of the spectrum everywhere, but only in a Cantor type set. In addition, we also show that the corresponding monodromy can be identified with the classical monodromy, defined by J. Duistermaat [“On global action-angle coordinates,” Commun. Pure Appl. Math. 33(6), 687–706 (1980)].

  2. Spectral monodromy of non-self-adjoint operators

    Phan, Quang Sang

    2014-01-01

    In the present paper, we build a combinatorial invariant, called the "spectral monodromy" from the spectrum of a single (non-self-adjoint) h-pseudodifferential operator with two degrees of freedom in the semi-classical limit. Our inspiration comes from the quantum monodromy defined for the joint spectrum of an integrable system of n commuting self-adjoint h-pseudodifferential operators, given by S. Vu Ngoc ["Quantum monodromy in integrable systems," Commun. Math. Phys. 203(2), 465-479 (1999)]. The first simple case that we treat in this work is a normal operator. In this case, the discrete spectrum can be identified with the joint spectrum of an integrable quantum system. The second more complex case we propose is a small perturbation of a self-adjoint operator with a classical integrability property. We show that the discrete spectrum (in a small band around the real axis) also has a combinatorial monodromy. The main difficulty in this case is that we do not know the description of the spectrum everywhere, but only in a Cantor type set. In addition, we also show that the corresponding monodromy can be identified with the classical monodromy, defined by J. Duistermaat ["On global action-angle coordinates," Commun. Pure Appl. Math. 33(6), 687-706 (1980)].

  3. Onboard radiation shielding estimates for interplanetary manned missions

    Totemeier, A.; Jevremovic, T.; Hounshel, D.

    2004-01-01

    The main focus of space related shielding design is to protect operating systems, personnel and key structural components from outer space and onboard radiation. This paper summarizes the feasibility of a lightweight neutron radiation shield design for a nuclear powered, manned space vehicle. The Monte Carlo code MCNP5 is used to determine radiation transport characteristics of the different materials and find the optimized shield configuration. A phantom torso encased in air is used to determine a dose rate for a crew member on the ship. Calculation results indicate that onboard shield against neutron radiation coming from nuclear engine can be achieved with very little addition of weight to the space vehicle. The selection of materials and neutron transport analysis as presented in this paper are useful starting data to design shield against neutrons generated when high-energy particles from outer space interact with matter on the space vehicle. (authors)

  4. Solving the multigroup adjoint transport equations using the method of cyclic characteristics

    Assawaroongruengchot, M.; Marleau, G. [Ecole Polytechnique de Montreal, Inst. de genie nucleaire, Montreal, Quebec (Canada)]. E-mail: monchai.assawar@polymtl.ca

    2005-07-01

    The adjoint transport solution algorithm based on the method of cyclic characteristics (MOCC) is developed for the heterogeneous 2D geometries. The adjoint characteristics equation associated with a cyclic tracking line is formulated, then a closed form for adjoint angular flux can be determined. The acceleration techniques are implemented using the group-reduction and group-splitting techniques. To demonstrate the efficacy of the algorithm, the calculations are performed on the 37 pin CANDU cell and on the Watanabe-Maynard benchmark problem. Comparisons of adjoint flux and k{sub eff} results obtained by MOCC and collision probability (CP) methods are performed. The mathematical relationship between pseudo-adjoint flux obtained by CP method and adjoint flux by MOCC method is presented. (author)

  5. Solving the multigroup adjoint transport equations using the method of cyclic characteristics

    Assawaroongruengchot, M.; Marleau, G.

    2005-01-01

    The adjoint transport solution algorithm based on the method of cyclic characteristics (MOCC) is developed for the heterogeneous 2D geometries. The adjoint characteristics equation associated with a cyclic tracking line is formulated, then a closed form for adjoint angular flux can be determined. The acceleration techniques are implemented using the group-reduction and group-splitting techniques. To demonstrate the efficacy of the algorithm, the calculations are performed on the 37 pin CANDU cell and on the Watanabe-Maynard benchmark problem. Comparisons of adjoint flux and k eff results obtained by MOCC and collision probability (CP) methods are performed. The mathematical relationship between pseudo-adjoint flux obtained by CP method and adjoint flux by MOCC method is presented. (author)

  6. The discrete adjoint method for parameter identification in multibody system dynamics.

    Lauß, Thomas; Oberpeilsteiner, Stefan; Steiner, Wolfgang; Nachbagauer, Karin

    2018-01-01

    The adjoint method is an elegant approach for the computation of the gradient of a cost function to identify a set of parameters. An additional set of differential equations has to be solved to compute the adjoint variables, which are further used for the gradient computation. However, the accuracy of the numerical solution of the adjoint differential equation has a great impact on the gradient. Hence, an alternative approach is the discrete adjoint method , where the adjoint differential equations are replaced by algebraic equations. Therefore, a finite difference scheme is constructed for the adjoint system directly from the numerical time integration method. The method provides the exact gradient of the discretized cost function subjected to the discretized equations of motion.

  7. Radiation shielding concrete

    Kunishima, Shigeru.

    1990-01-01

    The radiation shielding concretes comprise water, cement, fine aggregates consisting of serpentines and blown mist slags, coarse aggregates consisting of serpentines and kneading materials. Since serpentines containing a relatively great amount of water of crystallization in rocks as coarse aggregates and fine aggregates, the hydrogen content in the radiation shielding concretes is increased and the neutron shielding effect is improved. In addition, since serpentines are added as the fine aggregates and blown mists slags of a great specific gravity are used, the specific gravity of the shielding concretes is increased to improve the γ-ray shielding effect. Further, by the use of the kneading material having a water reducing effect and fluidizing effect, and by the bearing effect of the spherical blown mist slags used as the fine aggregates, concrete fluidity can be increased. Accordingly, workability of the radiation shielding concretes can be improved. (T.M.)

  8. Critical and shielding parametric studies with the Monte Carlo code TRIPOLI to identify the key points to take into account during the transportation of blanket assemblies with high ratio of americium

    Gosmain, Cecile-Aline

    2011-01-01

    In the framework of French research program on Generation IV sodium cooled fast reactor, one possible option consists in burning minor actinides in this kind of Advanced Sodium Technological Reactor. Two types of transmutation mode are studied in the world : the homogeneous mode of transmutation where actinides are scattered with very low enrichment ratio in fissile assemblies and the heterogeneous mode where fissile core is surrounded by blanket assemblies filled with minor actinides with ratio of incorporated actinides up to 20%. Depending on which element is considered to be burnt and on its content, these minor actinides contents imply constraints on assemblies' transportation between Nuclear Power Plants and fuel cycle facilities. In this study, we present some academic studies in order to identify some key constraints linked to the residual power and neutron/gamma load of such kind of blanket assemblies. To simplify the approach, we considered a modeling of a 'model cask' dedicated to the transportation of a unique irradiated blanket assembly loaded with 20% of Americium and basically inspired from an existent cask designed initially for the damaged fissile Superphenix assembly transport. Thermal calculations performed with EDF-SYRTHES code have shown that due to thermal limitations on cladding temperature, the decay time to be considered before transportation is 20 years. This study is based on explicit 3D representations of the cask and the contained blanket assembly with the Monte Carlo code TRIPOLI/JEFF3.1.1 library and concludes that after such a decay time, the transportation of a unique Americium radial blanket is feasible only if the design of our model cask is modified in order to comply with the dose limitation criterion. (author)

  9. Shielding design for positron emission tomography facility

    Abdallah, I.I.

    2007-01-01

    With the recent advent of readily available tracer isotopes, there has been marked increase in the number of hospital-based and free-standing positron emission tomography (PET) clinics. PET facilities employ relatively large activities of high-energy photon emitting isotopes, which can be dangerous to the health of humans and animals. This coupled with the current dose limits for radiation worker and members of the public can result in shielding requirements. This research contributes to the calculation of the appropriate shielding to keep the level of radiation within an acceptable recommended limit. Two different methods were used including measurements made at selected points of an operating PET facility and computer simulations by using Monte Carlo Transport Code. The measurements mainly concerned the radiation exposure at different points around facility using the survey meter detectors and Thermoluminescent Dosimeters (TLD). Then the set of manual calculation procedures were used to estimate the shielding requirements for a newly built PEF facility. The results from the measurement and the computer simulation were compared to the results obtained from the set manual calculation procedure. In general, the estimated weekly dose at the points of interest is lower than the regulatory limits for the little company of Mary Hospital. Furthermore, the density and the HVL for normal strength concrete and clay bricks are almost similar. In conclusion, PET facilities present somewhat different design requirements and are more likely to require additional radiation shielding. Therefore, existing shields at the little Company of Mary Hospital are in general found to be adequate and satisfactory and additional shielding was found necessary at the new PET facility in the department of Nuclear Medicine of the Dr. George Mukhari Hospital. By use of appropriate design, by implying specific shielding requirements and by maintaining good operating practices, radiation doses to

  10. Accelerator shielding benchmark problems

    Hirayama, H.; Ban, S.; Nakamura, T.

    1993-01-01

    Accelerator shielding benchmark problems prepared by Working Group of Accelerator Shielding in the Research Committee on Radiation Behavior in the Atomic Energy Society of Japan were compiled by Radiation Safety Control Center of National Laboratory for High Energy Physics. Twenty-five accelerator shielding benchmark problems are presented for evaluating the calculational algorithm, the accuracy of computer codes and the nuclear data used in codes. (author)

  11. I2S-LWR Activation Analysis of Heat Exchangers Using Hybrid Shielding Methodology with SCALE6.1

    Matijevic, M.; Pevec, D.; Jecmenica, R.

    2016-01-01

    The Integral Inherently Safe Light Water Reactor (I2S-LWR) concept developed by Georgia Tech is a novel PWR reactor delivering electric power of 1000 MWe while implementing inherent safety features typical for Generation III+ small modular reactors. The main safety feature is based on integral primary circuit configuration, bringing together compact design of the reactor core with 121 fuel assembly (FA), control rod drive mechanism (CRDM), 8 primary heat exchangers (PHE), 4 passive decay heat removal systems (DHRS), 8 pumps, and other integral components. A high power density core based on silicide fuel is selected to achieve a high thermal power which is extracted with PHEs placed in the annual region between the barrel and the vessel. The complex and integrated design of I2S-LWR leads to activation of integral components, mainly made from stainless steel, so accurate and precise Monte Carlo (MC) simulations are needed to quantify potential dose rates to personnel during routine maintenance operation. This shielding problem is therefore very challenging one, posing a non-trivial neutron flux solution in a phase space. This paper presents the performance of the hybrid shielding methodologies CADIS/FW-CADIS implemented in the MAVRIC sequence of the SCALE6.1 code package. The main objective was to develop a detailed MC shielding model of the I2S-LWR reactor along with effective variance reduction (VR) parameters and to calculate neutron fluence rates inside PHEs. Such results are then utilized to find neutron activation rate distribution via 60Co generation inside of a stack of microchannel heat exchangers (MCHX), which will be periodically withdrawn for the maintenance. 59Co impurities are the main cause of (n,gamma) radiative gamma dose to personnel via neutron activation since 60Co has half-life of 5.27 years and is emitting high energy gamma rays (1.17 MeV and 1.33 MeV). The developed MC model was successfully used to find converged fluxes inside all 8 stacks of

  12. Self-adjointness of the fast flux in a pressurized water reactor

    Mosteller, R.D.

    1985-01-01

    Most computer codes for the analysis of systems transients rely on a simplified representation of the active core, typically employing either a one-dimensional or a point kinetics model. The collapsing of neutronics data from multidimensional steady-state calculations normally employs flux/flux-adjoint weighting. The multidimensional calculations, however, usually are performed only for the forward problem, not the adjoint. The collapsing methodologies employed in generating the neutronics input for transient codes typically construct adjoint fluxes from the assumption that the fast flux is self-adjoint. Until now, no further verification of this assumption has been undertaken for thermal reactors. As part of the verification effort for EPRI's reactor analysis support package, the validity of this assumption now has been investigated for a modern pressurized water reactor (PWR). The PDQ-7 code was employed to perform two-group fine-mesh forward and adjoint calculations for a two-dimensional representation of Zion Unit 2 at beginning of life, based on the standard PWR ARMP model. It has been verified that the fast flux is very nearly self-adjoint in a PWR. However, a significant error can arise during the subsequent construction of the thermal adjoint flux unless allowance is made for the difference between the forward and adjoint thermal buckling terms. When such a difference is included, the thermal adjoint flux can be estimated very accurately

  13. Review and comparison of effective delayed neutron fraction calculation methods with Monte Carlo codes

    Bécares, V.; Pérez-Martín, S.; Vázquez-Antolín, M.; Villamarín, D.; Martín-Fuertes, F.; González-Romero, E.M.; Merino, I.

    2014-01-01

    Highlights: • Review of several Monte Carlo effective delayed neutron fraction calculation methods. • These methods have been implemented with the Monte Carlo code MCNPX. • They have been benchmarked against against some critical and subcritical systems. • Several nuclear data libraries have been used. - Abstract: The calculation of the effective delayed neutron fraction, β eff , with Monte Carlo codes is a complex task due to the requirement of properly considering the adjoint weighting of delayed neutrons. Nevertheless, several techniques have been proposed to circumvent this difficulty and obtain accurate Monte Carlo results for β eff without the need of explicitly determining the adjoint flux. In this paper, we make a review of some of these techniques; namely we have analyzed two variants of what we call the k-eigenvalue technique and other techniques based on different interpretations of the physical meaning of the adjoint weighting. To test the validity of all these techniques we have implemented them with the MCNPX code and we have benchmarked them against a range of critical and subcritical systems for which either experimental or deterministic values of β eff are available. Furthermore, several nuclear data libraries have been used in order to assess the impact of the uncertainty in nuclear data in the calculated value of β eff

  14. INTOR radiation shielding for personnel access

    Gohar, Y.; Abdou, M.

    1981-01-01

    The INTOR reactor shield system consists of the blanket, bulk shield, penetration shield, component shield, and biological shield. The bulk shield consists of two parts: (a) the inboard shield; and (b) the outboard shield. The distinction between the different components of the shield system is essential to satisfy the different design constraints and achieve various objectives

  15. A contribution Monte Carlo method

    Aboughantous, C.H.

    1994-01-01

    A Contribution Monte Carlo method is developed and successfully applied to a sample deep-penetration shielding problem. The random walk is simulated in most of its parts as in conventional Monte Carlo methods. The probability density functions (pdf's) are expressed in terms of spherical harmonics and are continuous functions in direction cosine and azimuthal angle variables as well as in position coordinates; the energy is discretized in the multigroup approximation. The transport pdf is an unusual exponential kernel strongly dependent on the incident and emergent directions and energies and on the position of the collision site. The method produces the same results obtained with the deterministic method with a very small standard deviation, with as little as 1,000 Contribution particles in both analog and nonabsorption biasing modes and with only a few minutes CPU time

  16. Adjoint sensitivity analysis applied on a model of irradiation assisted degradation of metals in aqueous systems

    Simonson, S.A.; Ballinger, R.G.; Christensen, R.A.

    1990-01-01

    Irradiation of an aqueous environment results, in general, in a steady state concentration of oxidizing chemical species in solution. Although the effect may be beneficial to the metal in contact with the solution in some cases, say by producing a more protective film, it is generally believed to be detrimental. The ability to predict the concentrations of the oxidizing species and from this begin to analyze the detrimental behavior on the metals requires computer codes that model the chemical reactions, production rates, and diffusion characteristics of the species being produced by irradiation. The large number of parameters and the complexity of the interactions involved in the predictions of irradiation effects on metals degradation requires a more sophisticated approach to determining the sensitivities of the final results. Monte Carlo techniques are too computationally intensive for practical use in determining sensitivities. The paper presents an approach, adjoint sensitivity analysis, that is more practical, i.e., three computer runs versus thousands, and also a more accurate measure of the sensitivities of the model

  17. Deterministic calculation of the effective delayed neutron fraction without using the adjoint neutron flux - 299

    Talamo, A.; Gohar, Y.; Aliberti, G.; Zhong, Z.; Bournos, V.; Fokov, Y.; Kiyavitskaya, H.; Routkovskaya, C.; Serafimovich, I.

    2010-01-01

    In 1997, Bretscher calculated the effective delayed neutron fraction by the k-ratio method. The Bretscher's approach is based on calculating the multiplication factor of a nuclear reactor core with and without the contribution of delayed neutrons. The multiplication factor set by the delayed neutrons (the delayed multiplication factor) is obtained as the difference between the total and the prompt multiplication factors. Bretscher evaluated the effective delayed neutron fraction as the ratio between the delayed and total multiplication factors (therefore the method is often referred to as k-ratio method). In the present work, the k-ratio method is applied by deterministic nuclear codes. The ENDF/B nuclear data library of the fuel isotopes ( 238 U and 238 U) have been processed by the NJOY code with and without the delayed neutron data to prepare multigroup WIMSD nuclear data libraries for the DRAGON code. The DRAGON code has been used for preparing the PARTISN macroscopic cross sections. This calculation methodology has been applied to the YALINA-Thermal assembly of Belarus. The assembly has been modeled and analyzed using PARTISN code with 69 energy groups and 60 different material zones. The deterministic and Monte Carlo results for the effective delayed neutron fraction obtained by the k-ratio method agree very well. The results also agree with the values obtained by using the adjoint flux. (authors)

  18. Adjoint-based global variance reduction approach for reactor analysis problems

    Zhang, Qiong; Abdel-Khalik, Hany S.

    2011-01-01

    A new variant of a hybrid Monte Carlo-Deterministic approach for simulating particle transport problems is presented and compared to the SCALE FW-CADIS approach. The new approach, denoted by the Subspace approach, optimizes the selection of the weight windows for reactor analysis problems where detailed properties of all fuel assemblies are required everywhere in the reactor core. Like the FW-CADIS approach, the Subspace approach utilizes importance maps obtained from deterministic adjoint models to derive automatic weight-window biasing. In contrast to FW-CADIS, the Subspace approach identifies the correlations between weight window maps to minimize the computational time required for global variance reduction, i.e., when the solution is required everywhere in the phase space. The correlations are employed to reduce the number of maps required to achieve the same level of variance reduction that would be obtained with single-response maps. Numerical experiments, serving as proof of principle, are presented to compare the Subspace and FW-CADIS approaches in terms of the global reduction in standard deviation. (author)

  19. Shielded regenerative neutron detector

    Terhune, J.H.; Neissel, J.P.

    1978-01-01

    An ion chamber type neutron detector is disclosed which has a greatly extended lifespan. The detector includes a fission chamber containing a mixture of active and breeding material and a neutron shielding material. The breeding and shielding materials are selected to have similar or substantially matching neutron capture cross-sections so that their individual effects on increased detector life are mutually enhanced

  20. Neutron streaming analysis for shield design of FMIT Facility

    Carter, L.L.

    1980-12-01

    Applications of the Monte Carlo method have been summarized relevant to neutron streaming problems of interest in the shield design for the FMIT Facility. An improved angular biasing method has been implemented to further optimize the calculation of streaming and this method has been applied to calculate streaming within a double bend pipe

  1. Carlos Romero

    2008-05-01

    Full Text Available Entrevista (en español Presentación Carlos Romero, politólogo, es profesor-investigador en el Instituto de Estudios Políticos de la Facultad de Ciencias Jurídicas y Políticas de la Universidad Central de Venezuela, en donde se ha desempeñado como coordinador del Doctorado, subdirector y director del Centro de Estudios de Postgrado. Cuenta con ocho libros publicados sobre temas de análisis político y relaciones internacionales, siendo uno de los últimos Jugando con el globo. La política exter...

  2. Radiation shielding plate

    Kobayashi, Torakichi; Sugawara, Takeo.

    1983-01-01

    Purpose: To reduce the weight and stabilize the configuration of a radiation shielding plate which is used in close contact with an object to be irradiated with radiation rays. Constitution: The radiation shielding plate comprises a substrate made of lead glass and a metallic lead coating on the surface of the substrate by means of plating, vapor deposition or the like. Apertures for permeating radiation rays are formed to the radiation shielding plate. Since the shielding plate is based on a lead glass plate, a sufficient mechanical strength can be obtained with a thinner structure as compared with the conventional plate made of metallic lead. Accordingly, if the shielding plate is disposed on a soft object to be irradiated with radiation rays, the object and the plate itself less deform to obtain a radiation irradiation pattern with distinct edges. (Moriyama, K.)

  3. Adjoint optimization of natural convection problems: differentially heated cavity

    Saglietti, Clio; Schlatter, Philipp; Monokrousos, Antonios; Henningson, Dan S.

    2017-12-01

    Optimization of natural convection-driven flows may provide significant improvements to the performance of cooling devices, but a theoretical investigation of such flows has been rarely done. The present paper illustrates an efficient gradient-based optimization method for analyzing such systems. We consider numerically the natural convection-driven flow in a differentially heated cavity with three Prandtl numbers (Pr=0.15{-}7) at super-critical conditions. All results and implementations were done with the spectral element code Nek5000. The flow is analyzed using linear direct and adjoint computations about a nonlinear base flow, extracting in particular optimal initial conditions using power iteration and the solution of the full adjoint direct eigenproblem. The cost function for both temperature and velocity is based on the kinetic energy and the concept of entransy, which yields a quadratic functional. Results are presented as a function of Prandtl number, time horizons and weights between kinetic energy and entransy. In particular, it is shown that the maximum transient growth is achieved at time horizons on the order of 5 time units for all cases, whereas for larger time horizons the adjoint mode is recovered as optimal initial condition. For smaller time horizons, the influence of the weights leads either to a concentric temperature distribution or to an initial condition pattern that opposes the mean shear and grows according to the Orr mechanism. For specific cases, it could also been shown that the computation of optimal initial conditions leads to a degenerate problem, with a potential loss of symmetry. In these situations, it turns out that any initial condition lying in a specific span of the eigenfunctions will yield exactly the same transient amplification. As a consequence, the power iteration converges very slowly and fails to extract all possible optimal initial conditions. According to the authors' knowledge, this behavior is illustrated here for

  4. Advances in Global Adjoint Tomography -- Massive Data Assimilation

    Ruan, Y.; Lei, W.; Bozdag, E.; Lefebvre, M. P.; Smith, J. A.; Krischer, L.; Tromp, J.

    2015-12-01

    Azimuthal anisotropy and anelasticity are key to understanding a myriad of processes in Earth's interior. Resolving these properties requires accurate simulations of seismic wave propagation in complex 3-D Earth models and an iterative inversion strategy. In the wake of successes in regional studies(e.g., Chen et al., 2007; Tape et al., 2009, 2010; Fichtner et al., 2009, 2010; Chen et al.,2010; Zhu et al., 2012, 2013; Chen et al., 2015), we are employing adjoint tomography based on a spectral-element method (Komatitsch & Tromp 1999, 2002) on a global scale using the supercomputer ''Titan'' at Oak Ridge National Laboratory. After 15 iterations, we have obtained a high-resolution transversely isotropic Earth model (M15) using traveltime data from 253 earthquakes. To obtain higher resolution images of the emerging new features and to prepare the inversion for azimuthal anisotropy and anelasticity, we expanded the original dataset with approximately 4,220 additional global earthquakes (Mw5.5-7.0) --occurring between 1995 and 2014-- and downloaded 300-minute-long time series for all available data archived at the IRIS Data Management Center, ORFEUS, and F-net. Ocean Bottom Seismograph data from the last decade are also included to maximize data coverage. In order to handle the huge dataset and solve the I/O bottleneck in global adjoint tomography, we implemented a python-based parallel data processing workflow based on the newly developed Adaptable Seismic Data Format (ASDF). With the help of the data selection tool MUSTANG developed by IRIS, we cleaned our dataset and assembled event-based ASDF files for parallel processing. We have started Centroid Moment Tensors (CMT) inversions for all 4,220 earthquakes with the latest model M15, and selected high-quality data for measurement. We will statistically investigate each channel using synthetic seismograms calculated in M15 for updated CMTs and identify problematic channels. In addition to data screening, we also modified

  5. The Adjoint Method for the Inverse Problem of Option Pricing

    Shou-Lei Wang

    2014-01-01

    Full Text Available The estimation of implied volatility is a typical PDE inverse problem. In this paper, we propose the TV-L1 model for identifying the implied volatility. The optimal volatility function is found by minimizing the cost functional measuring the discrepancy. The gradient is computed via the adjoint method which provides us with an exact value of the gradient needed for the minimization procedure. We use the limited memory quasi-Newton algorithm (L-BFGS to find the optimal and numerical examples shows the effectiveness of the presented method.

  6. Application of a dummy eye shield for electron treatment planning

    Kang, Sei-Kwon; Park, Soah; Hwang, Taejin; Cheong, Kwang-Ho; Han, Taejin; Kim, Haeyoung; Lee, Me-Yeon; Kim, Kyoung Ju; Oh, Do Hoon; Bae, Hoonsik

    2013-01-01

    Metallic eye shields have been widely used for near-eye treatments to protect critical regions, but have never been incorporated into treatment plans because of the unwanted appearance of the metal artifacts on CT images. The purpose of this work was to test the use of an acrylic dummy eye shield as a substitute for a metallic eye shield during CT scans. An acrylic dummy shield of the same size as the tungsten eye shield was machined and CT scanned. The BEAMnrc and the DOSXYZnrc were used for the Monte Carlo (MC) simulation, with the appropriate material information and density for the aluminum cover, steel knob and tungsten body of the eye shield. The Pinnacle adopting the Hogstrom electron pencil-beam algorithm was used for the one-port 6-MeV beam plan after delineation and density override of the metallic parts. The results were confirmed with the metal oxide semiconductor field effect transistor (MOSFET) detectors and the Gafchromic EBT2 film measurements. For both the maximum eyelid dose over the shield and the maximum dose under the shield, the MC results agreed with the EBT2 measurements within 1.7%. For the Pinnacle plan, the maximum dose under the shield agreed with the MC within 0.3%; however, the eyelid dose differed by -19.3%. The adoption of the acrylic dummy eye shield was successful for the treatment plan. However, the Pinnacle pencil-beam algorithm was not sufficient to predict the eyelid dose on the tungsten shield, and more accurate algorithms like MC should be considered for a treatment plan. (author)

  7. Cross-section uncertainty study of the NET shielding blanket

    Jaeger, J.F.

    1990-11-01

    The Next European Torus (NET) is foreseen as the next step in the European development towards the controlled use of thermonuclear fusion. Detail design of the shielding blanket protecting the peripherals, more especially the super-conducting coils, is well advanced. A cross-section uncertainty, i.e. a study of the expected inaccuracy due to the nuclear cross-section data, has been done for the neutron-gamma reactions in the insulation of the coils for such a design. As an extension of previous work on the NET shielding blanket (e.g. MCNP calculations), it was deemed necessary to estimate the accuracy attainable with transport codes in view of the uncertainties in microscopic cross-sections. The code used, SENSIBL, is based on perturbation theory and uses covariance files, COVFILS-2, for the cross-section data. This necessitates forward and adjoint flux calculations with a transport code (e.g. ONEDANT, TRISM) and folding the information contained in these coupled fluxes with the accuracy estimates of the evaluators of the ENDF/B-V files. Transport, P 5 S 12 , calculations were done with the ONEDANT code, for a shielding blanket design with 714 MW plasma fusion power. Several runs were done to obtain well converged forward and adjoint fluxes (ca. 1%). The forward and adjoint integral responses agree to 2%, which is consistent with the above accuracy. The n-γ response was chosen as it is typical of the general accuracy and is available for all materials considered. The present version of SENSIBL allows direct use of the geometric files of ONEDANT (or TRISM) which simplifies the input. Covariance data is not available at present in COVFILS-2 for all of the materials considered. Only H, C, N, O, Al, Si, Fe, Ni, and Pb could be considered, the big absentee being copper. The resulting uncertainty for the neutron-gamma reactions in the insulation of the coil was found to be 17%. Simulating copper by aluminium produces a negligible increase in the uncertainty, mainly

  8. HPGe detector shielding adjustment

    Trnkova, L.; Rulik, P.

    2008-01-01

    Low-level background shielding of HPGe detectors is used mainly for environmental samples with very low content of radionuclides. National Radiation Protection Institute (SURO) in Prague is equipped with 14 HPGe detectors with relative efficiency up to 150%. The detectors are placed in a room built from materials with low content of natural radionuclides and equipped with a double isolation of the floor against radon. Detectors themselves are placed in lead or steel shielding. Steel shielding with one of these detectors with relative efficiency of 100% was chosen to be rebuilt to achieve lower minimum detectable activity (MDA). Additional lead and copper shielding was built up inside the original steel shielding to reduce the volume of the inner space and filled with nitrogen by means of evaporating liquid nitrogen. The additional lead and copper shielding, consequent reduction of the inner volume and supply of evaporated nitrogen, caused a decrease of the background count and accordingly MDA values as well. The effect of nitrogen evaporation on the net areas of peaks belonging to radon daughters is significant. The enhanced shielding adjustment has the biggest influence in low energy range, what can be seen in collected data. MDA values in energy range from 30 keV to 400 keV decreased to 0.65-0.85 of original value, in energy range from 400 keV to 2 MeV they fell to 0.70-0.97 of original value. (authors)

  9. Shielding in experimental areas

    Stevens, A.; Tarnopolsky, G.; Thorndike, A.; White, S.

    1979-01-01

    The amount of shielding necessary to protect experimental detectors from various sources of background radiation is discussed. As illustrated an experiment has line of sight to sources extending approx. 90 m upstream from the intersection point. Packing a significant fraction of this space with shielding blocks would in general be unacceptable because primary access to the ring tunnel is from the experimental halls. (1) From basic machine design considerations and the inherent necessity to protect superconducting magnets it is expected that experimental areas in general will be cleaner than at any existing accelerator. (2) Even so, it will likely be necessary to have some shielding blocks available to protect experimental apparatus, and it may well be necessary to have a large amount of shielding available in the WAH. (3) Scraping will likely have some influence on all halls, and retractable apparatus may sometimes be necessary. (4) If access to any tunnel is needed to replace a magnet, one has 96 h (4 days) available to move shielding away to permit access without additional downtime. This (the amount of shielding one can shuffle about in 96 h) is a reasonable upper limit to shielding necessary in a hall

  10. Construction of adjoint operators for coupled equations depending on different variables

    Hoogenboom, J.E.

    1986-01-01

    A procedure is described for the construction of the adjoint operator matrix in case of coupled equations defining quantities that depend on different sets of variables. This case is not properly treated in the literature. From this procedure a simple rule can be deduced for the construction of such adjoint operator matrices

  11. Convergence problems associated with the iteration of adjoint equations in nuclear reactor theory

    Ngcobo, E.

    2003-01-01

    Convergence problems associated with the iteration of adjoint equations based on two-group neutron diffusion theory approximations in slab geometry are considered. For this purpose first-order variational techniques are adopted to minimise numerical errors involved. The importance of deriving the adjoint source from a breeding ratio is illustrated. The results obtained are consistent with the expected improvement in accuracy

  12. Solution of the mathematical adjoint equations for an interface current nodal formulation

    Yang, W.S.; Taiwo, T.A.; Khalil, H.

    1994-01-01

    Two techniques for solving the mathematical adjoint equations of an interface current nodal method are described. These techniques are the ''similarity transformation'' procedure and a direct solution scheme. A theoretical basis is provided for the similarity transformation procedure originally proposed by Lawrence. It is shown that the matrices associated with the mathematical and physical adjoint equations are similar to each other for the flat transverse leakage approximation but not for the quadratic leakage approximation. It is also shown that a good approximate solution of the mathematical adjoint for the quadratic transverse leakage approximation is obtained by applying the similarity transformation for the flat transverse leakage approximation to the physical adjoint solution. The direct solution scheme, which was developed as an alternative to the similarity transformation procedure, yields the correct mathematical adjoint solution for both flat and quadratic transverse leakage approximations. In this scheme, adjoint nodal equations are cast in a form very similar to that of the forward equations by employing a linear transformation of the adjoint partial currents. This enables the use of the forward solution algorithm with only minor modifications for solving the mathematical adjoint equations. By using the direct solution scheme as a reference method, it is shown that while the results computed with the similarity transformation procedure are approximate, they are sufficiently accurate for calculations of global and local reactivity changes resulting from coolant voiding in a liquid-metal reactor

  13. Forward-weighted CADIS method for variance reduction of Monte Carlo calculations of distributions and multiple localized quantities

    Wagner, J. C.; Blakeman, E. D.; Peplow, D. E.

    2009-01-01

    This paper presents a new hybrid (Monte Carlo/deterministic) method for increasing the efficiency of Monte Carlo calculations of distributions, such as flux or dose rate distributions (e.g., mesh tallies), as well as responses at multiple localized detectors and spectra. This method, referred to as Forward-Weighted CADIS (FW-CADIS), is a variation on the Consistent Adjoint Driven Importance Sampling (CADIS) method, which has been used for some time to very effectively improve the efficiency of Monte Carlo calculations of localized quantities, e.g., flux, dose, or reaction rate at a specific location. The basis of this method is the development of an importance function that represents the importance of particles to the objective of uniform Monte Carlo particle density in the desired tally regions. Implementation of this method utilizes the results from a forward deterministic calculation to develop a forward-weighted source for a deterministic adjoint calculation. The resulting adjoint function is then used to generate consistent space- and energy-dependent source biasing parameters and weight windows that are used in a forward Monte Carlo calculation to obtain approximately uniform statistical uncertainties in the desired tally regions. The FW-CADIS method has been implemented in the ADVANTG/MCNP framework and has been fully automated within the MAVRIC sequence of SCALE 6. Results of the application of the method to enabling the calculation of dose rates throughout an entire full-scale pressurized-water reactor facility are presented and discussed. (authors)

  14. Ford motor company NDE facility shielding design

    Metzger, R. L.; Van Riper, K. A.; Jones, M. H.

    2005-01-01

    Ford Motor Company proposed the construction of a large non-destructive evaluation laboratory for radiography of automotive power train components. The authors were commissioned to design the shielding and to survey the completed facility for compliance with radiation doses for occupationally and non-occupationally exposed personnel. The two X-ray sources are Varian Linatron 3000 accelerators operating at 9-11 MV. One performs computed tomography of automotive transmissions, while the other does real-time radiography of operating engines and transmissions. The shield thickness for the primary barrier and all secondary barriers were determined by point-kernel techniques. Point-kernel techniques did not work well for skyshine calculations and locations where multiple sources (e.g. tube head leakage and various scatter fields) impacted doses. Shielding for these areas was determined using transport calculations. A number of MCNP [Briesmeister, J. F. MCNPCA general Monte Carlo N-particle transport code version 4B. Los Alamos National Laboratory Manual (1997)] calculations focused on skyshine estimates and the office areas. Measurements on the operational facility confirmed the shielding calculations. (authors)

  15. Ford Motor Company NDE facility shielding design.

    Metzger, Robert L; Van Riper, Kenneth A; Jones, Martin H

    2005-01-01

    Ford Motor Company proposed the construction of a large non-destructive evaluation laboratory for radiography of automotive power train components. The authors were commissioned to design the shielding and to survey the completed facility for compliance with radiation doses for occupationally and non-occupationally exposed personnel. The two X-ray sources are Varian Linatron 3000 accelerators operating at 9-11 MV. One performs computed tomography of automotive transmissions, while the other does real-time radiography of operating engines and transmissions. The shield thickness for the primary barrier and all secondary barriers were determined by point-kernel techniques. Point-kernel techniques did not work well for skyshine calculations and locations where multiple sources (e.g. tube head leakage and various scatter fields) impacted doses. Shielding for these areas was determined using transport calculations. A number of MCNP [Briesmeister, J. F. MCNPCA general Monte Carlo N-particle transport code version 4B. Los Alamos National Laboratory Manual (1997)] calculations focused on skyshine estimates and the office areas. Measurements on the operational facility confirmed the shielding calculations.

  16. Track 4: basic nuclear science variance reduction for Monte Carlo criticality simulations. 6. Variational Variance Reduction for Monte Carlo Criticality Calculations

    Densmore, Jeffery D.; Larsen, Edward W.

    2001-01-01

    Recently, it has been shown that the figure of merit (FOM) of Monte Carlo source-detector problems can be enhanced by using a variational rather than a direct functional to estimate the detector response. The direct functional, which is traditionally employed in Monte Carlo simulations, requires an estimate of the solution of the forward problem within the detector region. The variational functional is theoretically more accurate than the direct functional, but it requires estimates of the solutions of the forward and adjoint source-detector problems over the entire phase-space of the problem. In recent work, we have performed Monte Carlo simulations using the variational functional by (a) approximating the adjoint solution deterministically and representing this solution as a function in phase-space and (b) estimating the forward solution using Monte Carlo. We have called this general procedure variational variance reduction (VVR). The VVR method is more computationally expensive per history than traditional Monte Carlo because extra information must be tallied and processed. However, the variational functional yields a more accurate estimate of the detector response. Our simulations have shown that the VVR reduction in variance usually outweighs the increase in cost, resulting in an increased FOM. In recent work on source-detector problems, we have calculated the adjoint solution deterministically and represented this solution as a linear-in-angle, histogram-in-space function. This procedure has several advantages over previous implementations: (a) it requires much less adjoint information to be stored and (b) it is highly efficient for diffusive problems, due to the accurate linear-in-angle representation of the adjoint solution. (Traditional variance-reduction methods perform poorly for diffusive problems.) Here, we extend this VVR method to Monte Carlo criticality calculations, which are often diffusive and difficult for traditional variance-reduction methods

  17. System of adjoint P1 equations for neutron moderation; Sistema de equacoes P1 adjuntas para a moderacao de neutrons

    Martinez, Aquilino Senra; Silva, Fernando Carvalho da; Cardoso, Carlos Eduardo Santos [Universidade Federal, Rio de Janeiro, RJ (Brazil). Coordenacao dos Programas de Pos-graduacao de Engenharia. Programa de Engenharia Nuclear

    2000-07-01

    In some applications of perturbation theory, it is necessary know the adjoint neutron flux, which is obtained by the solution of adjoint neutron diffusion equation. However, the multigroup constants used for this are weighted in only the direct neutron flux, from the solution of direct P1 equations. In this work, this procedure is questioned and the adjoint P1 equations are derived by the neutron transport equation, the reversion operators rules and analogies between direct and adjoint parameters. (author)

  18. Future directions in shielding methods and analysis

    Goldstein, H.

    1987-01-01

    Over the nearly half century history of shielding against reactor radiation, there has been a see-saw battle between theory and measurement. During that period the capability and accuracy of calculational methods have been enormously improved. The microscopic cross sections needed as input to the theoretical computations are now also known to adequate accuracy (with certain exceptions). Nonetheless, there remain substantial classes of shielding problems not yet accessible to satisfactory computational methods, particularly where three-dimensional geometries are involved. This paper discusses promising avenues to approach such problems, especially in the light of recent and expected advances in supercomputers. In particular, it seems that Monte Carlo methods should be much more advantageous in the new computer environment than they have been in the past

  19. Monte Carlo technique for local perturbations in multiplying systems

    Bernnat, W.

    1974-01-01

    The use of the Monte Carlo method for the calculation of reactivity perturbations in multiplying systems due to changes in geometry or composition requires a correlated sampling technique to make such calculations economical or in the case of very small perturbations even feasible. The technique discussed here is suitable for local perturbations. Very small perturbation regions will be treated by an adjoint mode. The perturbation of the source distribution due to the changed system and its reaction on the reactivity worth or other values of interest is taken into account by a fission matrix method. The formulation of the method and its application are discussed. 10 references. (U.S.)

  20. Adaptively Learning an Importance Function Using Transport Constrained Monte Carlo

    Booth, T.E.

    1998-01-01

    It is well known that a Monte Carlo estimate can be obtained with zero-variance if an exact importance function for the estimate is known. There are many ways that one might iteratively seek to obtain an ever more exact importance function. This paper describes a method that has obtained ever more exact importance functions that empirically produce an error that is dropping exponentially with computer time. The method described herein constrains the importance function to satisfy the (adjoint) Boltzmann transport equation. This constraint is provided by using the known form of the solution, usually referred to as the Case eigenfunction solution

  1. Self-adjoint extensions and spectral analysis in the Calogero problem

    Gitman, D M; Tyutin, I V; Voronov, B L

    2010-01-01

    In this paper, we present a mathematically rigorous quantum-mechanical treatment of a one-dimensional motion of a particle in the Calogero potential αx -2 . Although the problem is quite old and well studied, we believe that our consideration based on a uniform approach to constructing a correct quantum-mechanical description for systems with singular potentials and/or boundaries, proposed in our previous works, adds some new points to its solution. To demonstrate that a consideration of the Calogero problem requires mathematical accuracy, we discuss some 'paradoxes' inherent in the 'naive' quantum-mechanical treatment. Using a self-adjoint extension method, we construct and study all possible self-adjoint operators (self-adjoint Hamiltonians) associated with a formal differential expression for the Calogero Hamiltonian. In particular, we discuss a spontaneous scale-symmetry breaking associated with self-adjoint extensions. A complete spectral analysis of all self-adjoint Hamiltonians is presented.

  2. Probability density adjoint for sensitivity analysis of the Mean of Chaos

    Blonigan, Patrick J., E-mail: blonigan@mit.edu; Wang, Qiqi, E-mail: qiqi@mit.edu

    2014-08-01

    Sensitivity analysis, especially adjoint based sensitivity analysis, is a powerful tool for engineering design which allows for the efficient computation of sensitivities with respect to many parameters. However, these methods break down when used to compute sensitivities of long-time averaged quantities in chaotic dynamical systems. This paper presents a new method for sensitivity analysis of ergodic chaotic dynamical systems, the density adjoint method. The method involves solving the governing equations for the system's invariant measure and its adjoint on the system's attractor manifold rather than in phase-space. This new approach is derived for and demonstrated on one-dimensional chaotic maps and the three-dimensional Lorenz system. It is found that the density adjoint computes very finely detailed adjoint distributions and accurate sensitivities, but suffers from large computational costs.

  3. Variation estimation of the averaged cross sections in the direct and adjoint fluxes

    Cardoso, Carlos Eduardo Santos; Martinez, Aquilino Senra; Silva, Fernando Carvalho da

    1995-01-01

    There are several applications of the perturbation theory to specifics problems of reactor physics, such as nonuniform fuel burnup, nonuniform poison accumulation and evaluations of Doppler effects on reactivity. The neutron fluxes obtained from the solutions of direct and adjoint diffusion equations, are used in these applications. In the adjoint diffusion equation has been used the group constants averaged in the energy-dependent direct neutron flux, that it is not theoretically consistent. In this paper it is presented a method to calculate the energy-dependent adjoint neutron flux, to obtain the average group-constant that will be used in the adjoint diffusion equation. The method is based on the solution of the adjoint neutron balance equations, that were derived for a two regions cell. (author). 5 refs, 2 figs, 1 tab

  4. Numerical solution of multi group-Two dimensional- Adjoint equation with finite element method

    Poursalehi, N.; Khalafi, H.; Shahriari, M.; Minoochehr

    2008-01-01

    Adjoint equation is used for perturbation theory in nuclear reactor design. For numerical solution of adjoint equation, usually two methods are applied. These are Finite Element and Finite Difference procedures. Usually Finite Element Procedure is chosen for solving of adjoint equation, because it is more use able in variety of geometries. In this article, Galerkin Finite Element method is discussed. This method is applied for numerical solving multi group, multi region and two dimensional (X, Y) adjoint equation. Typical reactor geometry is partitioned with triangular meshes and boundary condition for adjoint flux is considered zero. Finally, for a case of defined parameters, Finite Element Code was applied and results were compared with Citation Code

  5. Radiological Shielding Design for the Neutron High-Resolution Backscattering Spectrometer EMU at the OPAL Reactor

    Ersez Tunay

    2017-01-01

    Full Text Available The shielding for the neutron high-resolution backscattering spectrometer (EMU located at the OPAL reactor (ANSTO was designed using the Monte Carlo code MCNP 5-1.60. The proposed shielding design has produced compact shielding assemblies, such as the neutron pre-monochromator bunker with sliding cylindrical block shields to accommodate a range of neutron take-off angles, and in the experimental area - shielding of neutron focusing guides, choppers, flight tube, backscattering monochromator, and additional shielding elements inside the Scattering Tank. These shielding assemblies meet safety and engineering requirements and cost constraints. The neutron dose rates around the EMU instrument were reduced to < 0.5 µSv/h and the gamma dose rates to a safe working level of ≤ 3 µSv/h.

  6. Radiological Shielding Design for the Neutron High-Resolution Backscattering Spectrometer EMU at the OPAL Reactor

    Ersez, Tunay; Esposto, Fernando; Souza, Nicolas R. de

    2017-09-01

    The shielding for the neutron high-resolution backscattering spectrometer (EMU) located at the OPAL reactor (ANSTO) was designed using the Monte Carlo code MCNP 5-1.60. The proposed shielding design has produced compact shielding assemblies, such as the neutron pre-monochromator bunker with sliding cylindrical block shields to accommodate a range of neutron take-off angles, and in the experimental area - shielding of neutron focusing guides, choppers, flight tube, backscattering monochromator, and additional shielding elements inside the Scattering Tank. These shielding assemblies meet safety and engineering requirements and cost constraints. The neutron dose rates around the EMU instrument were reduced to < 0.5 µSv/h and the gamma dose rates to a safe working level of ≤ 3 µSv/h.

  7. Shielding member for thermonuclear device

    Onozuka, Masanori

    1997-06-30

    In a thermonuclear device for shielding fast neutrons by shielding members disposed in a shielding vessel (vacuum vessel and structures such as a blanket disposed in the vacuum vessel), the shielding member comprises a large number of shielding wires formed fine and short so as to have elasticity. The shielding wires are sealed in a shielding vessel together with water, and when the width of the shielding vessel is changed, the shielding wires follow after the change of the width while elastically deforming in the shielding vessel, so that great stress and deformation are not formed thereby enabling to improve reliability. In addition, the length, the diameter and the shape of each of the shielding wires can be selected in accordance with the shielding space of the shielding vessel. Even if the shape of the shielding vessel is complicated, the shielding wires can be inserted easily. Accordingly, the filling rate of the shielding members can be changed easily. It can be produced more easily compared with a conventional spherical pebbles. It can be produced more easily than existent spherical shielding pebbles thereby enabling to reduce the production cost. (N.H.)

  8. Monte Carlo Calculation of Sensitivities to Secondary Angular Distributions. Theory and Validation

    Perell, R. L.

    2002-01-01

    The basic methods for solution of the transport equation that are in practical use today are the discrete ordinates (SN) method, and the Monte Carlo (Monte Carlo) method. While the SN method is typically less computation time consuming, the Monte Carlo method is often preferred for detailed and general description of three-dimensional geometries, and for calculations using cross sections that are point-wise energy dependent. For analysis of experimental and calculated results, sensitivities are needed. Sensitivities to material parameters in general, and to the angular distribution of the secondary (scattered) neutrons in particular, can be calculated by well known SN methods, using the fluxes obtained from solution of the direct and the adjoint transport equations. Algorithms to calculate sensitivities to cross-sections with Monte Carlo methods have been known for quite a time. However, only just recently we have developed a general Monte Carlo algorithm for the calculation of sensitivities to the angular distribution of the secondary neutrons

  9. Oscillator representations for self-adjoint Calogero Hamiltonians

    Gitman, D M [Institute of Physics, University of Sao Paulo (Brazil); Tyutin, I V; Voronov, B L, E-mail: gitman@dfn.if.usp.br, E-mail: tyutin@lpi.ru, E-mail: voronov@lpi.ru [Lebedev Physical Institute, Moscow (Russian Federation)

    2011-10-21

    In Gitman et al (2010 J. Phys. A: Math. Theor. 43 145205), we presented a mathematically rigorous quantum-mechanical treatment of a one-dimensional motion of a particle in the Calogero potential V(x) = {alpha}x{sup -2}. We described all possible self-adjoint (s.a.) operators (s.a. Hamiltonians) associated with the differential operation H=-d{sub x}{sup 2}+{alpha}x{sup -2} for the Calogero Hamiltonian. Here, we discuss a new aspect of the problem, the so-called oscillator representations for the Calogero Hamiltonians. As is known, operators of the form N-hat = a-hat{sup +} a-hat and A-hat = a-hat a-hat{sup +} are called operators of oscillator type. Oscillator-type operators possess a number of useful properties in the case when the elementary operators a-hat are closed. It turns out that some s.a. Calogero Hamiltonians allow oscillator-type representations. We describe such Hamiltonians and find the corresponding mutually adjoint elementary operators a-hat and a-hat{sup +}. An oscillator-type representation for a given Hamiltonian is generally not unique. (paper)

  10. Oscillator representations for self-adjoint Calogero Hamiltonians

    Gitman, D M; Tyutin, I V; Voronov, B L

    2011-01-01

    In Gitman et al (2010 J. Phys. A: Math. Theor. 43 145205), we presented a mathematically rigorous quantum-mechanical treatment of a one-dimensional motion of a particle in the Calogero potential V(x) = αx -2 . We described all possible self-adjoint (s.a.) operators (s.a. Hamiltonians) associated with the differential operation H=-d x 2 +αx -2 for the Calogero Hamiltonian. Here, we discuss a new aspect of the problem, the so-called oscillator representations for the Calogero Hamiltonians. As is known, operators of the form N-hat = a-hat + a-hat and A-hat = a-hat a-hat + are called operators of oscillator type. Oscillator-type operators possess a number of useful properties in the case when the elementary operators a-hat are closed. It turns out that some s.a. Calogero Hamiltonians allow oscillator-type representations. We describe such Hamiltonians and find the corresponding mutually adjoint elementary operators a-hat and a-hat + . An oscillator-type representation for a given Hamiltonian is generally not unique. (paper)

  11. Passive control of thermoacoustic oscillations with adjoint methods

    Aguilar, Jose; Juniper, Matthew

    2017-11-01

    Strict pollutant regulations are driving gas turbine manufacturers to develop devices that operate under lean premixed conditions, which produce less NOx but encourage thermoacoustic oscillations. These are a form of unstable combustion that arise due to the coupling between the acoustic field and the fluctuating heat release in a combustion chamber. In such devices, in which safety is paramount, thermoacoustic oscillations must be eliminated passively, rather than through feedback control. The ideal way to eliminate thermoacoustic oscillations is by subtly changing the shape of the device. To achieve this, one must calculate the sensitivity of each unstable thermoacoustic mode to every geometric parameter. This is prohibitively expensive with standard methods, but is relatively cheap with adjoint methods. In this study we first present low-order network models as a tool to model and study the thermoacoustic behaviour of combustion chambers. Then we compute the continuous adjoint equations and the sensitivities to relevant parameters. With this, we run an optimization routine that modifies the parameters in order to stabilize all the resonant modes of a laboratory combustor rig.

  12. Solar wind reconstruction from magnetosheath data using an adjoint approach

    Nabert, C.; Othmer, C.

    2015-01-01

    We present a new method to reconstruct solar wind conditions from spacecraft data taken during magnetosheath passages, which can be used to support, e.g., magnetospheric models. The unknown parameters of the solar wind are used as boundary conditions of an MHD (magnetohydrodynamics) magnetosheath model. The boundary conditions are varied until the spacecraft data matches the model predictions. The matching process is performed using a gradient-based minimization of the misfit between data and model. To achieve this time-consuming procedure, we introduce the adjoint of the magnetosheath model, which allows efficient calculation of the gradients. An automatic differentiation tool is used to generate the adjoint source code of the model. The reconstruction method is applied to THEMIS (Time History of Events and Macroscale Interactions during Substorms) data to calculate the solar wind conditions during spacecraft magnetosheath transitions. The results are compared to actual solar wind data. This allows validation of our reconstruction method and indicates the limitations of the MHD magnetosheath model used.

  13. Solar wind reconstruction from magnetosheath data using an adjoint approach

    Nabert, C.; Othmer, C. [Technische Univ. Braunschweig (Germany). Inst. fuer Geophysik und extraterrestrische Physik; Glassmeier, K.H. [Technische Univ. Braunschweig (Germany). Inst. fuer Geophysik und extraterrestrische Physik; Max Planck Institute for Solar System Research, Goettingen (Germany)

    2015-07-01

    We present a new method to reconstruct solar wind conditions from spacecraft data taken during magnetosheath passages, which can be used to support, e.g., magnetospheric models. The unknown parameters of the solar wind are used as boundary conditions of an MHD (magnetohydrodynamics) magnetosheath model. The boundary conditions are varied until the spacecraft data matches the model predictions. The matching process is performed using a gradient-based minimization of the misfit between data and model. To achieve this time-consuming procedure, we introduce the adjoint of the magnetosheath model, which allows efficient calculation of the gradients. An automatic differentiation tool is used to generate the adjoint source code of the model. The reconstruction method is applied to THEMIS (Time History of Events and Macroscale Interactions during Substorms) data to calculate the solar wind conditions during spacecraft magnetosheath transitions. The results are compared to actual solar wind data. This allows validation of our reconstruction method and indicates the limitations of the MHD magnetosheath model used.

  14. Self-adjoint oscillator operator from a modified factorization

    Reyes, Marco A. [Departamento de Fisica, DCI Campus Leon, Universidad de Guanajuato, Apdo. Postal E143, 37150 Leon, Gto. (Mexico); Rosu, H.C., E-mail: hcr@ipicyt.edu.mx [IPICyT, Instituto Potosino de Investigacion Cientifica y Tecnologica, Apdo. Postal 3-74 Tangamanga, 78231 San Luis Potosi, S.L.P. (Mexico); Gutierrez, M. Ranferi [Departamento de Fisica, DCI Campus Leon, Universidad de Guanajuato, Apdo. Postal E143, 37150 Leon, Gto. (Mexico)

    2011-05-30

    By using an alternative factorization, we obtain a self-adjoint oscillator operator of the form L{sub δ}=d/(dx) (p{sub δ}(x)d/(dx) )-((x{sup 2})/(p{sub δ}(x)) +p{sub δ}(x)-1), where p{sub δ}(x)=1+δe{sup -x{sup 2}}, with δ element of (-1,∞) an arbitrary real factorization parameter. At positive values of δ, this operator interpolates between the quantum harmonic oscillator Hamiltonian for δ=0 and a scaled Hermite operator at high values of δ. For the negative values of δ, the eigenfunctions look like deformed quantum mechanical Hermite functions. Possible applications are mentioned. -- Highlights: → We present a generalization of the Mielnik factorization. → We study the case of linear relationship between the factorization coefficients. → We introduce a new one-parameter self-adjoint oscillator operator. → We show its properties depending on the values of the parameter.

  15. Nucleonic analysis of a preliminary design for the ETF neutral-beam-injector duct shielding

    Urban, W.T.; Seed, T.J.; Dudziak, D.J.

    1980-01-01

    A nucleonic analysis of the Engineering Test Facility Neutral-Beam-Injector duct shielding has been made using a hybrid Monte Carlo/discrete-ordinates method. This method used Monte Carlo to determine internal and external boundary surface sources for a subsequent discrete-ordinates calculation of the neutron and gamma-ray transport through the shield. The analysis also included determination of the energy and angular distribution of neutrons and gamma rays entering the duct from the torus plasma chamber. Confidence in the hybrid method and the results obtained were provided through a comparison with three-dimensional Monte Carlo results

  16. Neutron shielding verification measurements and simulations for a 235-MeV proton therapy center

    Newhauser, W.D.; Titt, U.; Dexheimer, D.; Yan, X.; Nill, S.

    2002-01-01

    The neutron shielding at the Massachusetts General Hospital's 235-MeV proton therapy facility was investigated with measurements, analytical calculations, and realistic three-dimensional Monte Carlo simulations. In 37 of 40 cases studied, the analytical calculations predicted higher neutron dose equivalent rates outside the shielding than the measured, typically by more than a factor of 10, and in some cases more than 100. Monte Carlo predictions of dose equivalent at three locations are, on average, 1.1 times the measured values. Except at one location, all of the analytical model predictions and Monte Carlo simulations overestimate neutron dose equivalent

  17. Importance estimation in Monte Carlo modelling of neutron and photon transport

    Mickael, M.W.

    1992-01-01

    The estimation of neutron and photon importance in a three-dimensional geometry is achieved using a coupled Monte Carlo and diffusion theory calculation. The parameters required for the solution of the multigroup adjoint diffusion equation are estimated from an analog Monte Carlo simulation of the system under investigation. The solution of the adjoint diffusion equation is then used as an estimate of the particle importance in the actual simulation. This approach provides an automated and efficient variance reduction method for Monte Carlo simulations. The technique has been successfully applied to Monte Carlo simulation of neutron and coupled neutron-photon transport in the nuclear well-logging field. The results show that the importance maps obtained in a few minutes of computer time using this technique are in good agreement with Monte Carlo generated importance maps that require prohibitive computing times. The application of this method to Monte Carlo modelling of the response of neutron porosity and pulsed neutron instruments has resulted in major reductions in computation time. (Author)

  18. Estimating ISABELLE shielding requirements

    Stevens, A.J.; Thorndike, A.M.

    1976-01-01

    Estimates were made of the shielding thicknesses required at various points around the ISABELLE ring. Both hadron and muon requirements are considered. Radiation levels at the outside of the shield and at the BNL site boundary are kept at or below 1000 mrem per year and 5 mrem/year respectively. Muon requirements are based on the Wang formula for pion spectra, and the hadron requirements on the hadron cascade program CYLKAZ of Ranft. A muon shield thickness of 77 meters of sand is indicated outside the ring in one area, and hadron shields equivalent to from 2.7 to 5.6 meters in thickness of sand above the ring. The suggested safety allowance would increase these values to 86 meters and 4.0 to 7.2 meters respectively. There are many uncertainties in such estimates, but these last figures are considered to be rather conservative

  19. Scintillation counter, segmented shield

    Olson, R.E.; Thumim, A.D.

    1975-01-01

    A scintillation counter, particularly for counting gamma ray photons, includes a massive lead radiation shield surrounding a sample-receiving zone. The shield is disassembleable into a plurality of segments to allow facile installation and removal of a photomultiplier tube assembly, the segments being so constructed as to prevent straight-line access of external radiation through the shield into radiation-responsive areas. Provisions are made for accurately aligning the photomultiplier tube with respect to one or more sample-transmitting bores extending through the shield to the sample receiving zone. A sample elevator, used in transporting samples into the zone, is designed to provide a maximum gamma-receiving aspect to maximize the gamma detecting efficiency. (U.S.)

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

    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

  1. Shields for nuclear reactors

    Aspden, G.J.

    1984-01-01

    The patent concerns shields for nuclear reactors. The roof shield comprises a normally fixed radial outer portion, a radial inner portion rotatable about a vertical axis, and a connection between the inner and outer portions. In the event of hypothecal core disruption conditions, a cantilever system on the inner wall allows the upward movement of the inner wall, in order to prevent loss of containment. (UK)

  2. Radiation shielding curtain

    Winkler, N.T.

    1976-01-01

    A radiation shield is described in the form of a stranded curtain made up of bead-chains whose material and geometry are selected to produce a cross-sectional density that is the equivalent of 0.25 mm or more of lead and which curtain may be mounted on various radiological devices to shield against scattered radiation while offering a minimum of obstruction to the radiologist

  3. Shielded cells transfer automation

    Fisher, J.J.

    1984-01-01

    Nuclear waste from shielded cells is removed, packaged, and transferred manually in many nuclear facilities. Radiation exposure is absorbed by operators during these operations and limited only through procedural controls. Technological advances in automation using robotics have allowed a production waste removal operation to be automated to reduce radiation exposure. The robotic system bags waste containers out of glove box and transfers them to a shielded container. Operators control the system outside the system work area via television cameras. 9 figures

  4. Effect of lattice-level adjoint-weighting on the kinetics parameters of CANDU reactors

    Nichita, Eleodor

    2009-01-01

    Space-time kinetics calculations for CANDU reactors are routinely performed using the Improved Quasistatic (IQS) method. The IQS method calculates kinetics parameters such as the effective delayed-neutron fraction and generation time using adjoint weighting. In the current implementation of IQS, the direct flux, as well as the adjoint, is calculated using a two-group cell-homogenized reactor model which is inadequate for capturing the effect of the softer energy spectrum of the delayed neutrons. Additionally, there may also be fine spatial effects that are lost because the intra-cell adjoint shape is ignored. The purpose of this work is to compare the kinetics parameters calculated using the two-group cell-homogenized model with those calculated using lattice-level fine-group heterogeneous adjoint weighting and to assess whether the differences are large enough to justify further work on incorporating lattice-level adjoint weighting into the IQS method. A second goal is to evaluate whether the use of a fine-group cell-homogenized lattice-level adjoint, such as is the current practice for Light Water Reactors (LWRs), is sufficient to capture the lattice effects in question. It is found that, for CANDU lattices, the generation time is almost unaffected by the type of adjoint used to calculate it, but that the effective delayed-neutron fraction is affected by the type of adjoint used. The effective delayed-neutron fraction calculated using the two-group cell-homogenized adjoint is 5.2% higher than the 'best' effective delayed-neutron fraction value obtained using the detailed lattice-level fine-group heterogeneous adjoint. The effective delayed-neutron fraction calculated using the fine-group cell-homogenized adjoint is only 1.7% higher than the 'best' effective delayed-neutron fraction value but is still not equal to it. This situation is different from that encountered in LWRs where weighting by a fine-group cell-homogenized adjoint is sufficient to calculate the

  5. The influence of Shelter's FCM on the shield efficiency at there of containing

    Gorbachev, B.I.

    2000-01-01

    The reasonable detailed quantitative estimations of the influence of the γ-radiation capture and scattering processes in the Shelter's FCM material on the shield precautions efficiency at there of containing for the further shelf purpose. The Monte-Carlo calculations was carry out by the software Micro Shield 4.00 serial 4.00-00283, which make it possible correctly to account for the radiation shielding geometry, the radiation sources geometry, the radiation sources spectrums and the processes of the gamma-rays multi scattering in 'thick' shielding. Results presented in the tables, which is convenient to use. 3 refs., 18 tab

  6. Adjoint P1 equations solution for neutron slowing down; Solucao das equacoes P1 adjuntas para moderacao de neutrons

    Cardoso, Carlos Eduardo Santos; Martinez, Aquilino Senra; Silva, Fernando Carvalho da [Universidade Federal, Rio de Janeiro, RJ (Brazil). Coordenacao dos Programas de Pos-graduacao de Engenharia. Programa de Engenharia Nuclear

    2002-07-01

    In some applications of perturbation theory, it is necessary know the adjoint neutron flux, which is obtained by the solution of adjoint neutron diffusion equation. However, the multigroup constants used for this are weighted in only the direct neutron flux, from the solution of direct P1 equations. In this work, the adjoint P1 equations are derived by the neutron transport equation, the reversion operators rules and analogies between direct and adjoint parameters. The direct and adjoint neutron fluxes resulting from the solution of P{sub 1} equations were used to three different weighting processes, to obtain the macrogroup macroscopic cross sections. It was found out noticeable differences among them. (author)

  7. Deterministic sensitivity analysis of two-phase flow systems: forward and adjoint methods. Final report

    Cacuci, D.G.

    1984-07-01

    This report presents a self-contained mathematical formalism for deterministic sensitivity analysis of two-phase flow systems, a detailed application to sensitivity analysis of the homogeneous equilibrium model of two-phase flow, and a representative application to sensitivity analysis of a model (simulating pump-trip-type accidents in BWRs) where a transition between single phase and two phase occurs. The rigor and generality of this sensitivity analysis formalism stem from the use of Gateaux (G-) differentials. This report highlights the major aspects of deterministic (forward and adjoint) sensitivity analysis, including derivation of the forward sensitivity equations, derivation of sensitivity expressions in terms of adjoint functions, explicit construction of the adjoint system satisfied by these adjoint functions, determination of the characteristics of this adjoint system, and demonstration that these characteristics are the same as those of the original quasilinear two-phase flow equations. This proves that whenever the original two-phase flow problem is solvable, the adjoint system is also solvable and, in principle, the same numerical methods can be used to solve both the original and adjoint equations

  8. Unsteady adjoint for large eddy simulation of a coupled turbine stator-rotor system

    Talnikar, Chaitanya; Wang, Qiqi; Laskowski, Gregory

    2016-11-01

    Unsteady fluid flow simulations like large eddy simulation are crucial in capturing key physics in turbomachinery applications like separation and wake formation in flow over a turbine vane with a downstream blade. To determine how sensitive the design objectives of the coupled system are to control parameters, an unsteady adjoint is needed. It enables the computation of the gradient of an objective with respect to a large number of inputs in a computationally efficient manner. In this paper we present unsteady adjoint solutions for a coupled turbine stator-rotor system. As the transonic fluid flows over the stator vane, the boundary layer transitions to turbulence. The turbulent wake then impinges on the rotor blades, causing early separation. This coupled system exhibits chaotic dynamics which causes conventional adjoint solutions to diverge exponentially, resulting in the corruption of the sensitivities obtained from the adjoint solutions for long-time simulations. In this presentation, adjoint solutions for aerothermal objectives are obtained through a localized adjoint viscosity injection method which aims to stabilize the adjoint solution and maintain accurate sensitivities. Preliminary results obtained from the supercomputer Mira will be shown in the presentation.

  9. Neutron shielding properties of a new high-density concrete

    Lorente, A.; Gallego, E.; Vega Carrillo, H.R.; Mendez, R.

    2008-01-01

    The neutron shielding properties of a new high-density concrete (commercially available under the name Hormirad TM , developed in Spain by the company CT-RAD) have been characterized both experimentally and by Monte Carlo calculations. The shielding properties of this concrete against photons were previously studied and the material is being used to build bunkers, mazes and doors in medical accelerator facilities with good overall results. In this work, the objective was to characterize the material behaviour against neutrons, as well as to test alternative mixings including boron compounds in an effort to improve neutron shielding efficiency. With that purpose, Hormirad TM slabs of different thicknesses were exposed to an 241 Am-Be neutron source under controlled conditions in the neutron measurements laboratory of the Nuclear Engineering Department at UPM. The original mix, which includes a high fraction of magnetite, was then modified by adding different proportions of anhydrous borax (Na 2 B 4 O 7 ). In order to have a reference against common concrete used to shield medical accelerator facilities, the same experiment was repeated with ordinary (HA-25) concrete slabs. In parallel to the experiments, Monte Carlo calculations of the experiments were performed with MCNP5. The experimental results agree reasonably well with the Monte Carlo calculations. Therefore, the first and equilibrium tenth-value layers have been determined for the different types of concrete tested. The results show an advantageous behaviour of the Hormirad TM concrete, in terms of neutron attenuation against real thickness of the shielding. Borated concretes seem less practical since they did not show better neutron attenuation with respect to real thickness and their structural properties are worse. The neutron attenuation properties of Hormirad TM for typical neutron spectra in clinical LINAC accelerators rooms have been also characterized by Monte Carlo calculation. (author)

  10. Shielding measurements for a 230 MeV proton beam

    Siebers, J.V.

    1990-01-01

    Energetic secondary neutrons produced as protons interact with accelerator components and patients dominate the radiation shielding environment for proton radiotherapy facilities. Due to the scarcity of data describing neutron production, attenuation, absorbed dose, and dose equivalent values, these parameters were measured for 230 MeV proton bombardment of stopping length Al, Fe, and Pb targets at emission angles of 0 degree, 22 degree, 45 degree, and 90 degree in a thick concrete shield. Low pressure tissue-equivalent proportional counters with volumes ranging from 1 cm 3 to 1000 cm 3 were used to obtain microdosimetric spectra from which absorbed dose and radiation quality are deduced. Does equivalent values and attenuation lengths determined at depth in the shield were found to vary sharply with angle, but were found to be independent of target material. Neutron dose and radiation length values are compared with Monte Carlo neutron transport calculations performed using the Los Alamos High Energy Transport Code (LAHET). Calculations used 230 MeV protons incident upon an Fe target in a shielding geometry similar to that used in the experiment. LAHET calculations overestimated measured attenuation values at 0 degree, 22 degree, and 45 degree, yet correctly predicted the attenuation length at 90 degree. Comparison of the mean radiation quality estimated with the Monte Carlo calculations with measurements suggest that neutron quality factors should be increased by a factor of 1.4. These results are useful for the shielding design of new facilities as well as for testing neutron production and transport calculations

  11. Sensitivity kernels for viscoelastic loading based on adjoint methods

    Al-Attar, David; Tromp, Jeroen

    2014-01-01

    Observations of glacial isostatic adjustment (GIA) allow for inferences to be made about mantle viscosity, ice sheet history and other related parameters. Typically, this inverse problem can be formulated as minimizing the misfit between the given observations and a corresponding set of synthetic data. When the number of parameters is large, solution of such optimization problems can be computationally challenging. A practical, albeit non-ideal, solution is to use gradient-based optimization. Although the gradient of the misfit required in such methods could be calculated approximately using finite differences, the necessary computation time grows linearly with the number of model parameters, and so this is often infeasible. A far better approach is to apply the `adjoint method', which allows the exact gradient to be calculated from a single solution of the forward problem, along with one solution of the associated adjoint problem. As a first step towards applying the adjoint method to the GIA inverse problem, we consider its application to a simpler viscoelastic loading problem in which gravitationally self-consistent ocean loading is neglected. The earth model considered is non-rotating, self-gravitating, compressible, hydrostatically pre-stressed, laterally heterogeneous and possesses a Maxwell solid rheology. We determine adjoint equations and Fréchet kernels for this problem based on a Lagrange multiplier method. Given an objective functional J defined in terms of the surface deformation fields, we show that its first-order perturbation can be written δ J = int _{MS}K_{η }δ ln η dV +int _{t0}^{t1}int _{partial M}K_{dot{σ }} δ dot{σ } dS dt, where δ ln η = δη/η denotes relative viscosity variations in solid regions MS, dV is the volume element, δ dot{σ } is the perturbation to the time derivative of the surface load which is defined on the earth model's surface ∂M and for times [t0, t1] and dS is the surface element on ∂M. The `viscosity

  12. Adjoint sensitivity analysis of the thermomechanical behavior of repositories

    Wilson, J.L.; Thompson, B.M.

    1984-01-01

    The adjoint sensitivity method is applied to thermomechanical models for the first time. The method provides an efficient and inexpensive answer to the question: how sensitive are thermomechanical predictions to assumed parameters. The answer is exact, in the sense that it yields exact derivatives of response measures to parameters, and approximate, in the sense that projections of the response fo other parameter assumptions are only first order correct. The method is applied to linear finite element models of thermomechanical behavior. Extensions to more complicated models are straight-forward but often laborious. An illustration of the method with a two-dimensional repository corridor model reveals that the chosen stress response measure was most sensitive to Poisson's ratio for the rock matrix

  13. Neutron shielding material

    Nodaka, M.; Iida, T.; Taniuchi, H.; Yosimura, K.; Nagahama, H.

    1993-01-01

    From among the neutron shielding materials of the 'kobesh' series developed by Kobe Steel, Ltd. for transport and storage packagings, silicon rubber base type material has been tested for several items with a view to practical application and official authorization, and in order to determine its adaptability to actual vessels. Silicon rubber base type 'kobesh SR-T01' is a material in which, from among the silicone rubber based neutron shielding materials, the hydrogen content is highest and the boron content is most optimized. Its neutron shielding capability has been already described in the previous report (Taniuchi, 1986). The following tests were carried out to determine suitability for practical application; 1) Long-term thermal stability test 2) Pouring test on an actual-scale model 3) Fire test The experimental results showed that the silicone rubber based neutron shielding material has good neutron shielding capability and high long-term fire resistance, and that it can be applied to the advanced transport packaging. (author)

  14. Concrete radiation shielding

    Kaplan, M.F.

    1989-01-01

    The increased use of nuclear energy has given rise to a growth in the amount of artificially produced radiation and radioactive materials. The design and construction of shielding to protect people, equipment and structures from the effects of radiation has never been more important. Experience has shown that concrete is an effective, versatile and economical material for the construction of radiation shielding. This book provides information on the principles governing the interaction of radiation with matter and on relevant nuclear physics to give the engineer an understanding of the design and construction of concrete shielding. It covers the physical, mechanical and nuclear properties of concrete; the effects of elevated temperatures and possible damage to concrete due to radiation; basic procedures for the design of concrete radiation shields and finally the special problems associated with their construction and cost. Although written primarily for engineers concerned with the design and construction of concrete shielding, the book also reviews the widely scattered data and information available on this subject and should therefore be of interest to students and those wishing to research further in this field. (author)

  15. Method for dismantling shields

    Fukuzawa, Rokuro; Kondo, Nobuhiro; Kamiyama, Yoshinori; Kawasato, Ken; Hiraga, Tomoaki.

    1990-01-01

    The object of the present invention is to enable operators to dismantle shieldings contaminated by radioactivity easily and in a short period of time without danger of radiation exposure. A plurality of introduction pipes are embedded previously to the shielding walls of shielding members which contain a reactor core in a state where both ends of the introduction pipes are in communication with the outside. A wire saw is inserted into the introduction pipes to cut the shieldings upon dismantling. Then, shieldings can be dismantled easily in a short period of time with no radiation exposure to operator's. Further, according to the present invention, since the wire saw can be set easily and a large area can be cut at once, operation efficiency is improved. Further, since remote control is possible, cutting can be conducted in water and complicated places of the reactor. Biting upon starting the wire saw in the introduction pipe is reduced to facilitate startup for the rotation. (I.S.)

  16. Mechanical shielded hot cell

    Higgy, H.R.; Abdel-Rassoul, A.A.

    1983-01-01

    A plan to erect a mechanical shielded hot cell in the process hall of the Radiochemical Laboratory at Inchas is described. The hot cell is designed for safe handling of spent fuel bundles, from the Inchas reactor, and for dismantling and cutting the fuel rods in preparation for subsequent treatment. The biological shielding allows for the safe handling of a total radioactivity level up to 10,000 MeV-Ci. The hot cell consists of an α-tight stainless-steel box, connected to a γ-shielded SAS, through an air-lock containing a movable carriage. The α-box is tightly connected with six dry-storage cavities for adequate storage of the spent fuel bundles. Both the α-box, with the dry-storage cavities, and the SAS are surrounded by 200-mm thick biological lead shielding. The α-box is equipped with two master-slave manipulators, a lead-glass window, a monorail crane and Padirac and Minirag systems. The SAS is equipped with a lead-glass window, tong manipulator, a shielded pit and a mechanism for the entry of the spent fuel bundle. The hot cell is served by adequate ventilation and monitoring systems. (author)

  17. Visualising Earth's Mantle based on Global Adjoint Tomography

    Bozdag, E.; Pugmire, D.; Lefebvre, M. P.; Hill, J.; Komatitsch, D.; Peter, D. B.; Podhorszki, N.; Tromp, J.

    2017-12-01

    Recent advances in 3D wave propagation solvers and high-performance computing have enabled regional and global full-waveform inversions. Interpretation of tomographic models is often done on visually. Robust and efficient visualization tools are necessary to thoroughly investigate large model files, particularly at the global scale. In collaboration with Oak Ridge National Laboratory (ORNL), we have developed effective visualization tools and used for visualization of our first-generation global model, GLAD-M15 (Bozdag et al. 2016). VisIt (https://wci.llnl.gov/simulation/computer-codes/visit/) is used for initial exploration of the models and for extraction of seismological features. The broad capability of VisIt, and its demonstrated scalability proved valuable for experimenting with different visualization techniques, and in the creation of timely results. Utilizing VisIt's plugin-architecture, a data reader plugin was developed, which reads the ADIOS (https://www.olcf.ornl.gov/center-projects/adios/) format of our model files. Blender (https://www.blender.org) is used for the setup of lighting, materials, camera paths and rendering of geometry. Python scripting was used to control the orchestration of different geometries, as well as camera animation for 3D movies. While we continue producing 3D contour plots and movies for various seismic parameters to better visualize plume- and slab-like features as well as anisotropy throughout the mantle, our aim is to make visualization an integral part of our global adjoint tomography workflow to routinely produce various 2D cross-sections to facilitate examination of our models after each iteration. This will ultimately form the basis for use of pattern recognition techniques in our investigations. Simulations for global adjoint tomography are performed on ORNL's Titan system and visualization is done in parallel on ORNL's post-processing cluster Rhea.

  18. Time reversal imaging, Inverse problems and Adjoint Tomography}

    Montagner, J.; Larmat, C. S.; Capdeville, Y.; Kawakatsu, H.; Fink, M.

    2010-12-01

    With the increasing power of computers and numerical techniques (such as spectral element methods), it is possible to address a new class of seismological problems. The propagation of seismic waves in heterogeneous media is simulated more and more accurately and new applications developed, in particular time reversal methods and adjoint tomography in the three-dimensional Earth. Since the pioneering work of J. Claerbout, theorized by A. Tarantola, many similarities were found between time-reversal methods, cross-correlations techniques, inverse problems and adjoint tomography. By using normal mode theory, we generalize the scalar approach of Draeger and Fink (1999) and Lobkis and Weaver (2001) to the 3D- elastic Earth, for theoretically understanding time-reversal method on global scale. It is shown how to relate time-reversal methods on one hand, with auto-correlations of seismograms for source imaging and on the other hand, with cross-correlations between receivers for structural imaging and retrieving Green function. Time-reversal methods were successfully applied in the past to acoustic waves in many fields such as medical imaging, underwater acoustics, non destructive testing and to seismic waves in seismology for earthquake imaging. In the case of source imaging, time reversal techniques make it possible an automatic location in time and space as well as the retrieval of focal mechanism of earthquakes or unknown environmental sources . We present here some applications at the global scale of these techniques on synthetic tests and on real data, such as Sumatra-Andaman (Dec. 2004), Haiti (Jan. 2010), as well as glacial earthquakes and seismic hum.

  19. An Adjoint-Based Adaptive Ensemble Kalman Filter

    Song, Hajoon

    2013-10-01

    A new hybrid ensemble Kalman filter/four-dimensional variational data assimilation (EnKF/4D-VAR) approach is introduced to mitigate background covariance limitations in the EnKF. The work is based on the adaptive EnKF (AEnKF) method, which bears a strong resemblance to the hybrid EnKF/three-dimensional variational data assimilation (3D-VAR) method. In the AEnKF, the representativeness of the EnKF ensemble is regularly enhanced with new members generated after back projection of the EnKF analysis residuals to state space using a 3D-VAR [or optimal interpolation (OI)] scheme with a preselected background covariance matrix. The idea here is to reformulate the transformation of the residuals as a 4D-VAR problem, constraining the new member with model dynamics and the previous observations. This should provide more information for the estimation of the new member and reduce dependence of the AEnKF on the assumed stationary background covariance matrix. This is done by integrating the analysis residuals backward in time with the adjoint model. Numerical experiments are performed with the Lorenz-96 model under different scenarios to test the new approach and to evaluate its performance with respect to the EnKF and the hybrid EnKF/3D-VAR. The new method leads to the least root-mean-square estimation errors as long as the linear assumption guaranteeing the stability of the adjoint model holds. It is also found to be less sensitive to choices of the assimilation system inputs and parameters.

  20. An Adjoint-Based Adaptive Ensemble Kalman Filter

    Song, Hajoon; Hoteit, Ibrahim; Cornuelle, Bruce D.; Luo, Xiaodong; Subramanian, Aneesh C.

    2013-01-01

    A new hybrid ensemble Kalman filter/four-dimensional variational data assimilation (EnKF/4D-VAR) approach is introduced to mitigate background covariance limitations in the EnKF. The work is based on the adaptive EnKF (AEnKF) method, which bears a strong resemblance to the hybrid EnKF/three-dimensional variational data assimilation (3D-VAR) method. In the AEnKF, the representativeness of the EnKF ensemble is regularly enhanced with new members generated after back projection of the EnKF analysis residuals to state space using a 3D-VAR [or optimal interpolation (OI)] scheme with a preselected background covariance matrix. The idea here is to reformulate the transformation of the residuals as a 4D-VAR problem, constraining the new member with model dynamics and the previous observations. This should provide more information for the estimation of the new member and reduce dependence of the AEnKF on the assumed stationary background covariance matrix. This is done by integrating the analysis residuals backward in time with the adjoint model. Numerical experiments are performed with the Lorenz-96 model under different scenarios to test the new approach and to evaluate its performance with respect to the EnKF and the hybrid EnKF/3D-VAR. The new method leads to the least root-mean-square estimation errors as long as the linear assumption guaranteeing the stability of the adjoint model holds. It is also found to be less sensitive to choices of the assimilation system inputs and parameters.

  1. Objective-function Hybridization in Adjoint Seismic Tomography

    Yuan, Y. O.; Bozdag, E.; Simons, F.; Gao, F.

    2016-12-01

    In the realm of seismic tomography, we are at the threshold of a new era of huge seismic datasets. However, how to assimilate as much information as possible from every seismogram is still a challenge. Cross-correlation measurements are generally tailored to some window selection algorithms, such as FLEXWIN (Maggie et al. 2008), to balance amplitude differences between seismic phases. However, these measurements naturally favor maximum picks in selected windows. It is also difficult to select all usable portions of seismograms in an optimum way that lots of information is generally lost, particularly the scattered waves. Instantaneous phase type of misfits extract information from every wiggle without cutting seismograms into small pieces, however, dealing with cycle skips at short periods can be challenging. For this purpose, we introduce a flexible hybrid approach for adjoint seismic tomography, to combine various objective functions. We initially focus on phase measurements and propose using instantaneous phase to take into account relatively small-magnitude scattered waves at long periods while using cross-correlation measurements on FLEXWIN windows to select distinct body-wave arrivals without complicating measurements due to non-linearities at short periods. To better deal with cycle skips and reliably measure instantaneous phases we design a new misfit function that incorporates instantaneous phase information implicitly instead of measuring it explicitly, through using normalized analytic signals. We present in our synthetic experiments how instantaneous phase, cross-correlation and their hybridization affect tomographic results. The combination of two different phase measurements in a hybrid approach constitutes progress towards using "anything and everything" in a data set, addressing data quality and measurement challenges simultaneously. We further extend hybridisation of misfit functions for amplitude measurements such as cross-correlation amplitude

  2. Use of adjoint methods in the probabilistic finite element approach to fracture mechanics

    Liu, Wing Kam; Besterfield, Glen; Lawrence, Mark; Belytschko, Ted

    1988-01-01

    The adjoint method approach to probabilistic finite element methods (PFEM) is presented. When the number of objective functions is small compared to the number of random variables, the adjoint method is far superior to the direct method in evaluating the objective function derivatives with respect to the random variables. The PFEM is extended to probabilistic fracture mechanics (PFM) using an element which has the near crack-tip singular strain field embedded. Since only two objective functions (i.e., mode I and II stress intensity factors) are needed for PFM, the adjoint method is well suited.

  3. Status of multigroup cross-section data for shielding applications

    Roussin, R.W.; Maskewitz, B.F.; Trubey, D.K.

    1983-01-01

    Multigroup cross-section libraries for shielding applications in formats for direct use in discrete ordinates or Monte Carlo codes have long been a part of the Data Library Collection (DLC) of the Radiation Shielding Information Center (RSIC). In recent years libraries in more flexible and comprehensive formats, which allow the user to derive his own problem-dependent sets, have been added to the collection. The current status of both types is described, as well as projections for adding data libraries based on ENDF/B-V

  4. Shield support frame. Schildausbaugestell

    Plaga, K.

    1981-09-17

    A powered shield support frame for coal sheds is described comprising of two bottom sliding shoes, a large area gob shield and a larg area roof assembly, all joined movable together. The sliding shoes and the gob shield are joined by a lemniscate guide. Two hydraulic props are arranged at the face-side at one third of the length of the sliding shoes and at the goaf-side at one third of the length of the roof assembly. A nearly horizontal lying pushing prop unit joins the bottom wall sliding shoes to the goaf-side lemniscate guide. This assembly can be applied to seams with a thickness down to 45 cm. (OGR).

  5. Radiation shielding material

    Matsumoto, Akio; Isobe, Eiji.

    1976-01-01

    Purpose: To increase the shielding capacity of the radiation shielding material having an abundant flexibility. Constitution: A mat consisting of a lead or lead alloy fibrous material is covered with a cloth, and the two are made integral by sewing in a kilted fashion by using a yarn. Thereafter, the system is covered with a gas-tight film or sheet. The shielding material obtained in this way has, in addition to the above merits, advantages in that (1) it is free from restoration due to elasticity so that it can readily seal contaminants, (2) it can be used in a state consisting of a number of overlapped layers, (3) it fits the shoulder well and is readily portable and (4) it permits attachment of fasteners or the like. (Ikeda, J.)

  6. Hybrid Magnetic Shielding

    Royal, Kevin; Crawford, Christopher; Mullins, Andrew; Porter, Greg; Blanton, Hunter; Johnstone, Connor; Kistler, Ben; Olivera, Daniela

    2017-09-01

    The search for the electric dipole moment of the neutron requires the ambient magnetic field to be on the pT scale which is accomplished with large magnetic shielding rooms. These rooms are fitted with large mu-metal sheets to allow for passive cancellation of background magnetic fields. Active shielding technology cannot uniformly cancel background magnetic fields. These issues can be remedied by combining the methods into a hybrid system. The design used is composed of panels that have an active layer of cancellation between two sheets of mu-metal. The panels form a cube and draw in magnetic fields perpendicular to the surface which can then be reduced using active shielding. This work is supported by the Department of Energy under Contract DE-SC0008107.

  7. Reactor head shielding apparatus

    Schukei, G.E.; Roebelen, G.J.

    1992-01-01

    This patent describes a nuclear reactor head shielding apparatus for mounting on spaced reactor head lifting members radially inwardly of the head bolts. It comprises a frame of sections for mounting on the lifting members and extending around the top central area of the head, mounting means for so mounting the frame sections, including downwardly projecting members on the frame sections and complementary upwardly open recessed members for fastening to the lifting members for receiving the downwardly projecting members when the frame sections are lowered thereto with lead shielding supported thereby on means for hanging lead shielding on the frame to minimize radiation exposure or personnel working with the head bolts or in the vicinity thereof

  8. Benchmarking study and its application for shielding analysis of large accelerator facilities

    Lee, Hee-Seock; Kim, Dong-hyun; Oranj, Leila Mokhtari; Oh, Joo-Hee; Lee, Arim; Jung, Nam-Suk [POSTECH, Pohang (Korea, Republic of)

    2015-10-15

    Shielding Analysis is one of subjects which are indispensable to construct large accelerator facility. Several methods, such as the Monte Carlo, discrete ordinate, and simplified calculation, have been used for this purpose. The calculation precision is overcome by increasing the trial (history) numbers. However its accuracy is still a big issue in the shielding analysis. To secure the accuracy in the Monte Carlo calculation, the benchmarking study using experimental data and the code comparison are adopted fundamentally. In this paper, the benchmarking result for electrons, protons, and heavy ions are presented as well as the proper application of the results is discussed. The benchmarking calculations, which are indispensable in the shielding analysis were performed for different particles: proton, heavy ion and electron. Four different multi-particle Monte Carlo codes, MCNPX, FLUKA, PHITS, and MARS, were examined for higher energy range equivalent to large accelerator facility. The degree of agreement between the experimental data including the SINBAD database and the calculated results were estimated in the terms of secondary neutron production and attenuation through the concrete and iron shields. The degree of discrepancy and the features of Monte Carlo codes were investigated and the application way of the benchmarking results are discussed in the view of safety margin and selecting the code for the shielding analysis. In most cases, the tested Monte Carlo codes give proper credible results except of a few limitation of each codes.

  9. Benchmarking study and its application for shielding analysis of large accelerator facilities

    Lee, Hee-Seock; Kim, Dong-hyun; Oranj, Leila Mokhtari; Oh, Joo-Hee; Lee, Arim; Jung, Nam-Suk

    2015-01-01

    Shielding Analysis is one of subjects which are indispensable to construct large accelerator facility. Several methods, such as the Monte Carlo, discrete ordinate, and simplified calculation, have been used for this purpose. The calculation precision is overcome by increasing the trial (history) numbers. However its accuracy is still a big issue in the shielding analysis. To secure the accuracy in the Monte Carlo calculation, the benchmarking study using experimental data and the code comparison are adopted fundamentally. In this paper, the benchmarking result for electrons, protons, and heavy ions are presented as well as the proper application of the results is discussed. The benchmarking calculations, which are indispensable in the shielding analysis were performed for different particles: proton, heavy ion and electron. Four different multi-particle Monte Carlo codes, MCNPX, FLUKA, PHITS, and MARS, were examined for higher energy range equivalent to large accelerator facility. The degree of agreement between the experimental data including the SINBAD database and the calculated results were estimated in the terms of secondary neutron production and attenuation through the concrete and iron shields. The degree of discrepancy and the features of Monte Carlo codes were investigated and the application way of the benchmarking results are discussed in the view of safety margin and selecting the code for the shielding analysis. In most cases, the tested Monte Carlo codes give proper credible results except of a few limitation of each codes

  10. Shielding research in France

    Lafore, P

    1964-10-01

    Shielding research as an independent subject in France dates from 1956. The importance of these studies has been reflected in the contribution which they have made to power reactor design and in the resultant savings in expenditure for civil engineering and machinery for the removal of mobile shields. The Reactor Shielding Research Division numbers approximately 60 persons and uses several experimental facilities. These include: NAIADE I, installed near the ZOE reactor and operating with a natural uranium slab 2 cm thick (an effective diameter of 60 cm is the one most commonly used); the TRITON pool-type reactor, mainly used in shielding studies, includes an active-water loop, by means of which the secondary shields required for light-water reactors can be studied; core, NEREIDE, which is situated near a 2 m x 2 m aluminium window enables a large neutron source to be placed in a compartment without water in which large-scale mock-ups can be mounted for the study, in particular, of neutron diffusion in large cavities, and of reactor shielding of greater thickness than that in NAIADE I; SAMES 600 keV accelerator is used for monoenergetic neutron studies. Instrumentation studies are an important part of the work, mainly in the measurement of fast neutrons and their spectra by activation detectors. Of late, attention has been directed towards the use of (n, n') (rhodium) reactions and of heavy detectors for low-flux measurements. The simultaneous use of a large number of detectors poses automation problems. With our installation we can count 16 detectors simultaneously. Neutron spectrum studies are conducted with nuclear emulsions and a lithium-6 semiconductor spectrometer. As to the materials used, the research carried out in France involves chiefly graphite, iron and concrete at various temperatures up to 800 deg C. Different compounds, borated and non-borated and of densities up to between 1 and 9 are under consideration. Problems connected with applications are

  11. Radiation shielding bricks

    Crowe, G.J.W.

    1983-01-01

    A radiation shielding brick for use in building dry walls to form radiation proof enclosures and other structures is described. It is square in shape and comprises a sandwich of an inner layer of lead or similar shielding material between outer layers of plastics material, for structural stability. The ability to mechanically interlock adjacent bricks is provided by shaping the edges as cooperating external and internal V-sections. Relatively leak-free joints are ensured by enlarging the width of the inner layer in the edge region. (author)

  12. Shielding performance of the NET vacuum vessel

    Arkuszewski, J.J.; Jaeger, J.F.

    1988-01-01

    To corroborate 1-D deterministic shielding calculations on the Next European Torus (NET) vacuum vessel/shield and shielding blanket, 3-D Monte Carlo calculations have been done with the MCNP code. This should provide information on the poloidal and the toroidal variations. Plasma source simulation and the geometrical model are described, as are other assumptions. The calculations are based on the extended plasma power of 714 MW. The results reported here are the heat deposition in various parts of the device, on the one hand, and the neutron and photon currents at the outer boundary of the vacuum vessel, on the other hand. The latter are needed for the detailed design of the super-conducting magnetic coils. A reasonable statistics has been obtained on the outboard side of the torus, though this cannot be said for the inboard side. The inboard is, however, much more toroidally symmetric than the outboard, so that other methods could be applied such as 2-D deterministic calculations, for instance. (author) 4 refs., 44 figs., 42 tabs

  13. Hybrid Active-Passive Radiation Shielding System

    National Aeronautics and Space Administration — A radiation shielding system is proposed that integrates active magnetic fields with passive shielding materials. The objective is to increase the shielding...

  14. SP-100 GES/NAT radiation shielding systems design and development testing

    Disney, R.K.; Kulikowski, H.D.; McGinnis, C.A.; Reese, J.C.; Thomas, K.; Wiltshire, F.

    1991-01-01

    Advanced Energy Systems (AES) of Westinghouse Electric Corporation is under subcontract to the General Electric Company to supply nuclear radiation shielding components for the SP-100 Ground Engineering System (GES) Nuclear Assembly Test to be conducted at Westinghouse Hanford Company at Richland, Washington. The radiation shielding components are integral to the Nuclear Assembly Test (NAT) assembly and include prototypic and non-prototypic radiation shielding components which provide prototypic test conditions for the SP-100 reactor subsystem and reactor control subsystem components during the GES/NAT operations. W-AES is designing three radiation shield components for the NAT assembly; a prototypic Generic Flight System (GFS) shield, the Lower Internal Facility Shield (LIFS), and the Upper Internal Facility Shield (UIFS). This paper describes the design approach and development testing to support the design, fabrication, and assembly of these three shield components for use within the vacuum vessel of the GES/NAT. The GES/NAT shields must be designed to operate in a high vacuum which simulates space operations. The GFS shield and LIFS must provide prototypic radiation/thermal environments and mechanical interfaces for reactor system components. The NAT shields, in combination with the test facility shielding, must provide adequate radiation attenuation for overall test operations. Special design considerations account for the ground test facility effects on the prototypic GFS shield. Validation of the GFS shield design and performance will be based on detailed Monte Carlo analyses and developmental testing of design features. Full scale prototype testing of the shield subsystems is not planned

  15. Selection of important Monte Carlo histories

    Egbert, Stephen D.

    1987-01-01

    The 1986 Dosimetry System (DS86) for Japanese A-bomb survivors uses information describing the behavior of individual radiation particles, simulated by Monte Carlo methods, to calculate the transmission of radiation into structures and, thence, into humans. However, there are practical constraints on the number of such particle 'histories' that may be used. First, the number must be sufficiently high to provide adequate statistical precision fir any calculated quantity of interest. For integral quantities, such as dose or kerma, statistical precision of approximately 5% (standard deviation) is required to ensure that statistical uncertainties are not a major contributor to the overall uncertainty of the transmitted value. For differential quantities, such as scalar fluence spectra, 10 to 15% standard deviation on individual energy groups is adequate. Second, the number of histories cannot be so large as to require an unacceptably large amount of computer time to process the entire survivor data base. Given that there are approx. 30,000 survivors, each having 13 or 14 organs of interest, the number of histories per organ must be constrained to less than several ten's of thousands at the very most. Selection and use of the most important Monte Carlo leakage histories from among all those calculated allows the creation of an efficient house and organ radiation transmission system for use at RERF. While attempts have been made during the adjoint Monte Carlo calculation to bias the histories toward an efficient dose estimate, this effort has been far from satisfactory. Many of the adjoint histories on a typical leakage tape are either starting in an energy group in which there is very little kerma or dose or leaking into an energy group with very little free-field couple with. By knowing the typical free-field fluence and the fluence-to-dose factors with which the leaking histories will be used, one can select histories rom a leakage tape that will contribute to dose

  16. Discrete Adjoint-Based Design Optimization of Unsteady Turbulent Flows on Dynamic Unstructured Grids

    Nielsen, Eric J.; Diskin, Boris; Yamaleev, Nail K.

    2009-01-01

    An adjoint-based methodology for design optimization of unsteady turbulent flows on dynamic unstructured grids is described. The implementation relies on an existing unsteady three-dimensional unstructured grid solver capable of dynamic mesh simulations and discrete adjoint capabilities previously developed for steady flows. The discrete equations for the primal and adjoint systems are presented for the backward-difference family of time-integration schemes on both static and dynamic grids. The consistency of sensitivity derivatives is established via comparisons with complex-variable computations. The current work is believed to be the first verified implementation of an adjoint-based optimization methodology for the true time-dependent formulation of the Navier-Stokes equations in a practical computational code. Large-scale shape optimizations are demonstrated for turbulent flows over a tiltrotor geometry and a simulated aeroelastic motion of a fighter jet.

  17. Almost commuting self-adjoint matrices: The real and self-dual cases

    Loring, Terry A.; Sørensen, Adam P. W.

    2016-08-01

    We show that a pair of almost commuting self-adjoint, symmetric matrices is close to a pair of commuting self-adjoint, symmetric matrices (in a uniform way). Moreover, we prove that the same holds with self-dual in place of symmetric and also for paths of self-adjoint matrices. Since a symmetric, self-adjoint matrix is real, we get a real version of Huaxin Lin’s famous theorem on almost commuting matrices. Similarly, the self-dual case gives a version for matrices over the quaternions. To prove these results, we develop a theory of semiprojectivity for real C*-algebras and also examine various definitions of low-rank for real C*-algebras.

  18. Non-analogue Monte Carlo method, application to neutron simulation; Methode de Monte Carlo non analogue, application a la simulation des neutrons

    Morillon, B.

    1996-12-31

    With most of the traditional and contemporary techniques, it is still impossible to solve the transport equation if one takes into account a fully detailed geometry and if one studies precisely the interactions between particles and matters. Only the Monte Carlo method offers such a possibility. However with significant attenuation, the natural simulation remains inefficient: it becomes necessary to use biasing techniques where the solution of the adjoint transport equation is essential. The Monte Carlo code Tripoli has been using such techniques successfully for a long time with different approximate adjoint solutions: these methods require from the user to find out some parameters. If this parameters are not optimal or nearly optimal, the biases simulations may bring about small figures of merit. This paper presents a description of the most important biasing techniques of the Monte Carlo code Tripoli ; then we show how to calculate the importance function for general geometry with multigroup cases. We present a completely automatic biasing technique where the parameters of the biased simulation are deduced from the solution of the adjoint transport equation calculated by collision probabilities. In this study we shall estimate the importance function through collision probabilities method and we shall evaluate its possibilities thanks to a Monte Carlo calculation. We compare different biased simulations with the importance function calculated by collision probabilities for one-group and multigroup problems. We have run simulations with new biasing method for one-group transport problems with isotropic shocks and for multigroup problems with anisotropic shocks. The results show that for the one-group and homogeneous geometry transport problems the method is quite optimal without splitting and russian roulette technique but for the multigroup and heterogeneous X-Y geometry ones the figures of merit are higher if we add splitting and russian roulette technique.

  19. Radiation shielding cloth

    Ijiri, Yasuo; Fujinuma, Tadashi; Tamura, Shoji.

    1989-01-01

    Radiation shielding cloth having radiation shielding layers comprising a composition of inorganic powder of high specific gravity and rubber are excellentin flexibility and comfortable to put on. However, since they are heavy in the weight, operators are tired upon putting them for a long time. In view of the above, the radiation ray shielding layers are prepared by calendering sheets obtained by preliminary molding of the composition to set the variation of the thickness within a range of +15% to -0% of prescribed thickness. Since the composition of inorganic powder at high specific gravity and rubber used for radiation ray shielding comprises a great amount of inorganic powder at high specific gravity blended therein, it is generally poor in fabricability. Therefor, it is difficult to attain fine control for the sheet thickness by merely molding a composition block at once. Then, the composition is at first preliminarily molded into a sheet-like shape which is somewhat thickener than the final thickness and then finished by calendering, by which the thickness can be reduced in average as compared with conventional products while keeping the prescribed thickness and reducing the weight reduce by so much. (N.H.)

  20. Electrostatic shielding of transformers

    De Leon, Francisco

    2017-11-28

    Toroidal transformers are currently used only in low-voltage applications. There is no published experience for toroidal transformer design at distribution-level voltages. Toroidal transformers are provided with electrostatic shielding to make possible high voltage applications and withstand the impulse test.

  1. Penetration portion shielding structure

    Hayashi, Katsumi; Narita, Hitoshi; Handa, Hiroyuki; Takeuchi, Jun; Tozuka, Fumio.

    1994-01-01

    Openings of a plurality of shieldings for penetration members are aligned to each other, and penetration members are inserted from the openings. Then, the openings of the plurality of shielding members are slightly displaced with each other to make the penetration portions into a helical configuration, so that leakage of radiation is reduced. Upon removal of the members, reverse operation is conducted. When a flowable shielding material is used, the penetration portions are constituted with two plates having previously formed openings and pipes for connecting the openings with each other and a vessel covering the entire of them. After passing the penetration members such as a cable, the relative position of the two plates is changed by twisting, to form a helical configuration which reduces radiation leakage. Since they are bent into the helical configuration, shielding performance is extremely improved compared with a case that radiation leakage is caused from an opening of a straight pipe. In addition, since they can be returned to straight pipes, attachment, detachment and maintenance can be conducted easily. (N.H.)

  2. Dosimetry and shielding

    Farinelli, U.

    1977-01-01

    Today, reactor dosimetry and shielding have wide areas of overlap as concerns both problems and methods. Increased interchange of results and know-how would benefit both. The areas of common interest include calculational methods, sensitivity studies, theoretical and experimental benchmarks, cross sections and other nuclear data, multigroup libraries and procedures for their adjustment, experimental techniques and damage functions. This paper reviews the state-of-the-art and the latest development in each of these areas as far as shielding is concerned, and suggests a number of interactions that could be profitable for reactor dosimetry. Among them, re-evaluation of the potentialities of calculational methods (in view of the recent developments) in predicting radiation environments of interest; the application of sensitivity analysis to dosimetry problems; a common effort in the field of theoretical benchmarks; the use of the shielding one-material propagation experiments as reference spectra for detector cross sections; common standardization of the detector nuclear data used in both fields; the setting up of a common (or compatible) multigroup structure and library applicable to shielding, dosimetry and core physics; the exchange of information and experience in the fields of cross section errors, correlations and adjustment; and the intercomparison of experimental techniques

  3. Radiation shielding glass

    Kido, Kazuhiro; Ueda, Hajime.

    1997-01-01

    It was found that a glass composition comprising, as essential ingredients, SiO 2 , PbO, Gd 2 O 3 and alkali metal oxides can provide a shielding performance against electromagnetic waves, charged particles and neutrons. The present invention provides radiation shielding glass containing at least from 16 to 46wt% of SiO 2 , from 47 to 75wt% of PbO, from 1 to 10wt% of Gd 2 O 3 , from 0 to 3wt% of Li 2 O, from 0 to 7wt% of Na 2 O, from 0 to 7wt% of K 2 O provided that Li 2 O + Na 2 O + K 2 O is from 1 to 10wt%, B 2 O 3 is from 0 to 10wt%, CeO 2 is from 0 to 3wt%, As 2 O 3 is from 0 to 1wt% and Sb 2 O 3 is from 0 to 1wt%. Since the glass can shield electromagnetic waves, charged particles and neutrons simultaneously, radiation shielding windows can be designed and manufactured at a reduced thickness and by less constitutional numbers in a circumstance where they are present altogether. (T.M.)

  4. Characterization and uniqueness of distinguished self-adjoint extensions of dirac operators

    Klaus, M.; Wuest, R.; Princeton Univ., NJ

    1979-01-01

    Distinguished self-adjoint extensions of Dirac operators are characterized by Nenciu and constructed by means of cut-off potentials by Wuest. In this paper it is shown that the existence and a more explicit characterization of Nenciu's self-adjoint extensions can be obtained as a consequence from results of the cut-off method, that these extensions are the same as the extensions constructed with cut-off potentials and that they are unique in some sense. (orig.) [de

  5. Contaminant deposition building shielding factors for US residential structures

    Dickson, E D; Hamby, D M; Eckerman, K F

    2015-01-01

    This paper presents validated building shielding factors designed for contemporary US housing-stock under an idealized, yet realistic, exposure scenario from contaminant deposition on the roof and surrounding surfaces. The building shielding factors are intended for use in emergency planning and level three probabilistic risk assessments for a variety of postulated radiological events in which a realistic assessment is necessary to better understand the potential risks for accident mitigation and emergency response planning. Factors are calculated from detailed computational housing-units models using the general-purpose Monte Carlo N-Particle computational code, MCNP5, and are benchmarked from a series of narrow- and broad-beam measurements analyzing the shielding effectiveness of ten common general-purpose construction materials and ten shielding models representing the primary weather barriers (walls and roofs) of likely US housing-stock. Each model was designed to scale based on common residential construction practices and include, to the extent practical, all structurally significant components important for shielding against ionizing radiation. Calculations were performed for floor-specific locations from contaminant deposition on the roof and surrounding ground as well as for computing a weighted-average representative building shielding factor for single- and multi-story detached homes, both with and without basement as well for single-wide manufactured housing-unit. (paper)

  6. Cloud immersion building shielding factors for US residential structures

    Dickson, E D; Hamby, D M

    2014-01-01

    This paper presents validated building shielding factors designed for contemporary US housing-stock under an idealized, yet realistic, exposure scenario within a semi-infinite cloud of radioactive material. The building shielding factors are intended for use in emergency planning and level three probabilistic risk assessments for a variety of postulated radiological events in which a realistic assessment is necessary to better understand the potential risks for accident mitigation and emergency response planning. Factors are calculated from detailed computational housing-units models using the general-purpose Monte Carlo N-Particle computational code, MCNP5, and are benchmarked from a series of narrow- and broad-beam measurements analyzing the shielding effectiveness of ten common general-purpose construction materials and ten shielding models representing the primary weather barriers (walls and roofs) of likely US housing-stock. Each model was designed to scale based on common residential construction practices and include, to the extent practical, all structurally significant components important for shielding against ionizing radiation. Calculations were performed for floor-specific locations as well as for computing a weighted-average representative building shielding factor for single- and multi-story detached homes, both with and without basement, as well for single-wide manufactured housing-units. (paper)

  7. Gamma dose from activation of internal shields in IRIS reactor.

    Agosteo, Stefano; Cammi, Antonio; Garlati, Luisella; Lombardi, Carlo; Padovani, Enrico

    2005-01-01

    The International Reactor Innovative and Secure is a modular pressurised water reactor with an integral design. This means that all the primary system components, such as the steam generators, pumps, pressuriser and control rod drive mechanisms, are located inside the reactor vessel, which requires a large diameter. For the sake of better reliability and safety, it is desirable to achieve the reduction of vessel embrittlement as well as the lowering of the dose beyond the vessel. The former can be easily accomplished by the presence of a wide downcomer, filled with water, which surrounds the core region, while the latter needs the presence of additional internal shields. An optimal shielding configuration is under investigation, for reducing the ex-vessel dose due to activated internals and for limiting the amount of the biological shielding. MCNP 4C calculations were performed to evaluate the neutron and the gamma dose during operation and the 60Co activation of various shields configurations. The gamma dose beyond the vessel from activation of its structural components was estimated in a shutdown condition, with the Monte Carlo code FLUKA 2002 and the MicroShield software. The results of the two codes are in agreement and show that the dose is sufficiently low, even without an additional shield.

  8. Gamma dose from activation of internal shields in IRIS reactor

    Agosteo, S.; Cammi, A.; Garlati, L.; Lombardi, C.; Padovani, E.

    2005-01-01

    The International Reactor Innovative and Secure is a modular pressurised water reactor with an integral design. This means that all the primary system components, such as the steam generators, pumps, pressurizer and control rod drive mechanisms, are located inside the reactor vessel, which requires a large diameter. For the sake of better reliability and safety, it is desirable to achieve the reduction of vessel embrittlement as well as the lowering of the dose beyond the vessel. The former can be easily accomplished by the presence of a wide downcomer, filled with water, which surrounds the core region, while the latter needs the presence of additional internal shields. An optimal shielding configuration is under investigation, for reducing the ex-vessel dose due to activated internals and for limiting the amount of the biological shielding. MCNP 4C calculations were performed to evaluate the neutron and the gamma dose during operation and the 60 Co activation of various shields configurations. The gamma dose beyond the vessel from activation of its structural components was estimated in a shutdown condition, with the Monte Carlo code FLUKA 2002 and the MicroShield software. The results of the two codes are in agreement and show that the dose is sufficiently low, even without an additional shield. (authors)

  9. Contaminant deposition building shielding factors for US residential structures.

    Dickson, Elijah; Hamby, David; Eckerman, Keith

    2017-10-10

    This paper presents validated building shielding factors designed for contemporary US housing-stock under an idealized, yet realistic, exposure scenario from contaminant deposition on the roof and surrounding surfaces. The building shielding factors are intended for use in emergency planning and level three probabilistic risk assessments for a variety of postulated radiological events in which a realistic assessment is necessary to better understand the potential risks for accident mitigation and emergency response planning. Factors are calculated from detailed computational housing-units models using the general-purpose Monte Carlo N-Particle computational code, MCNP5, and are benchmarked from a series of narrow- and broad-beam measurements analyzing the shielding effectiveness of ten common general-purpose construction materials and ten shielding models representing the primary weather barriers (walls and roofs) of likely US housing-stock. Each model was designed to scale based on common residential construction practices and include, to the extent practical, all structurally significant components important for shielding against ionizing radiation. Calculations were performed for floor-specific locations from contaminant deposition on the roof and surrounding ground as well as for computing a weighted-average representative building shielding factor for single- and multi-story detached homes, both with and without basement as well for single-wide manufactured housing-unit. © 2017 IOP Publishing Ltd.

  10. Evaluation of Shielding Performance for Newly Developed Composite Materials

    Evans, Beren Richard

    This work details an investigation into the contributing factors behind the success of newly developed composite neutron shield materials. Monte Carlo simulation methods were utilized to assess the neutron shielding capabilities and secondary radiation production characteristics of aluminum boron carbide, tungsten boron carbide, bismuth borosilicate glass, and Metathene within various neutron energy spectra. Shielding performance and secondary radiation data suggested that tungsten boron carbide was the most effective composite material. An analysis of the macroscopic cross-section contributions from constituent materials and interaction mechanisms was then performed in an attempt to determine the reasons for tungsten boron carbide's success over the other investigated materials. This analysis determined that there was a positive correlation between a non-elastic interaction contribution towards a material's total cross-section and shielding performance within the thermal and epi-thermal energy regimes. This finding was assumed to be a result of the boron-10 absorption reaction. The analysis also determined that within the faster energy regions, materials featuring higher non-elastic interaction contributions were comparable to those exhibiting primarily elastic scattering via low Z elements. This allowed for the conclusion that composite shield success within higher energy neutron spectra does not necessitate the use elastic scattering via low Z elements. These findings suggest that the inclusion of materials featuring high thermal absorption properties is more critical to composite neutron shield performance than the presence of constituent materials more inclined to maximize elastic scattering energy loss.

  11. Spatial discretizations for self-adjoint forms of the radiative transfer equations

    Morel, Jim E.; Adams, B. Todd; Noh, Taewan; McGhee, John M.; Evans, Thomas M.; Urbatsch, Todd J.

    2006-01-01

    There are three commonly recognized second-order self-adjoint forms of the neutron transport equation: the even-parity equations, the odd-parity equations, and the self-adjoint angular flux equations. Because all of these equations contain second-order spatial derivatives and are self-adjoint for the mono-energetic case, standard continuous finite-element discretization techniques have proved quite effective when applied to the spatial variables. We first derive analogs of these equations for the case of time-dependent radiative transfer. The primary unknowns for these equations are functions of the angular intensity rather than the angular flux, hence the analog of the self-adjoint angular flux equation is referred to as the self-adjoint angular intensity equation. Then we describe a general, arbitrary-order, continuous spatial finite-element approach that is applied to each of the three equations in conjunction with backward-Euler differencing in time. We refer to it as the 'standard' technique. We also introduce an alternative spatial discretization scheme for the self-adjoint angular intensity equation that requires far fewer unknowns than the standard method, but appears to give comparable accuracy. Computational results are given that demonstrate the validity of both of these discretization schemes

  12. Gonad shielding in diagnostic radiology

    Anon.

    1975-01-01

    The use of gonad shielding is an important radiation protection technique, intended to reduce unnecessary x-ray exposure of the gonads of patients from diagnostic x-ray procedures. The types of gonad shields in use are discussed as are the types of diagnostic examinations that should include gonad shielding. It was found that when properly used, most shields provided substantial gonad dose reductions

  13. Comparison of calculational methods for liquid metal reactor shields

    Carter, L.L.; Moore, F.S.; Morford, R.J.; Mann, F.M.

    1985-09-01

    A one-dimensional comparison is made between Monte Carlo (MCNP), discrete ordinances (ANISN), and diffusion theory (MlDX) calculations of neutron flux and radiation damage from the core of the Fast Flux Test Facility (FFTF) out to the reactor vessel. Diffusion theory was found to be reasonably accurate for the calculation of both total flux and radiation damage. However, for large distances from the core, the calculated flux at very high energies is low by an order of magnitude or more when the diffusion theory is used. Particular emphasis was placed in this study on the generation of multitable cross sections for use in discrete ordinates codes that are self-shielded, consistent with the self-shielding employed in the generation of cross sections for use with diffusion theory. The Monte Carlo calculation, with a pointwise representation of the cross sections, was used as the benchmark for determining the limitations of the other two calculational methods. 12 refs., 33 figs

  14. Shielding experiments for accelerator facilities

    Nakashima, Hiroshi; Tanaka, Susumu; Sakamoto, Yukio [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment] [and others

    2000-06-01

    A series of shielding experiments was carried out by using AVF cyclotron accelerator of TIARA at JAERI in order to validate shielding design methods for accelerator facilities in intermediate energy region. In this paper neutron transmission experiment through thick shields and radiation streaming experiment through a labyrinth are reported. (author)

  15. Shielding experiments for accelerator facilities

    Nakashima, Hiroshi; Tanaka, Susumu; Sakamoto, Yukio

    2000-01-01

    A series of shielding experiments was carried out by using AVF cyclotron accelerator of TIARA at JAERI in order to validate shielding design methods for accelerator facilities in intermediate energy region. In this paper neutron transmission experiment through thick shields and radiation streaming experiment through a labyrinth are reported. (author)

  16. Data and Workflow Management Challenges in Global Adjoint Tomography

    Lei, W.; Ruan, Y.; Smith, J. A.; Modrak, R. T.; Orsvuran, R.; Krischer, L.; Chen, Y.; Balasubramanian, V.; Hill, J.; Turilli, M.; Bozdag, E.; Lefebvre, M. P.; Jha, S.; Tromp, J.

    2017-12-01

    It is crucial to take the complete physics of wave propagation into account in seismic tomography to further improve the resolution of tomographic images. The adjoint method is an efficient way of incorporating 3D wave simulations in seismic tomography. However, global adjoint tomography is computationally expensive, requiring thousands of wavefield simulations and massive data processing. Through our collaboration with the Oak Ridge National Laboratory (ORNL) computing group and an allocation on Titan, ORNL's GPU-accelerated supercomputer, we are now performing our global inversions by assimilating waveform data from over 1,000 earthquakes. The first challenge we encountered is dealing with the sheer amount of seismic data. Data processing based on conventional data formats and processing tools (such as SAC), which are not designed for parallel systems, becomes our major bottleneck. To facilitate the data processing procedures, we designed the Adaptive Seismic Data Format (ASDF) and developed a set of Python-based processing tools to replace legacy FORTRAN-based software. These tools greatly enhance reproducibility and accountability while taking full advantage of highly parallel system and showing superior scaling on modern computational platforms. The second challenge is that the data processing workflow contains more than 10 sub-procedures, making it delicate to handle and prone to human mistakes. To reduce human intervention as much as possible, we are developing a framework specifically designed for seismic inversion based on the state-of-the art workflow management research, specifically the Ensemble Toolkit (EnTK), in collaboration with the RADICAL team from Rutgers University. Using the initial developments of the EnTK, we are able to utilize the full computing power of the data processing cluster RHEA at ORNL while keeping human interaction to a minimum and greatly reducing the data processing time. Thanks to all the improvements, we are now able to

  17. Full Waveform Adjoint Seismic Tomography of the Antarctic Plate

    Lloyd, A. J.; Wiens, D.; Zhu, H.; Tromp, J.; Nyblade, A.; Anandakrishnan, S.; Aster, R. C.; Huerta, A. D.; Winberry, J. P.; Wilson, T. J.; Dalziel, I. W. D.; Hansen, S. E.; Shore, P.

    2017-12-01

    Recent studies investigating the response and influence of the solid Earth on the evolution of the cryosphere demonstrate the need to account for 3D rheological structure to better predict ice sheet dynamics, stability, and future sea level impact, as well as to improve glacial isostatic adjustment models and more accurately measure ice mass loss. Critical rheological properties like mantle viscosity and lithospheric thickness may be estimated from shear wave velocity models that, for Antarctica, would ideally possess regional-scale resolution extending down to at least the base of the transition zone (i.e. 670 km depth). However, current global- and continental-scale seismic velocity models are unable to obtain both the resolution and spatial coverage necessary, do not take advantage of the full set of available Antarctic data, and, in most instance, employ traditional seismic imaging techniques that utilize limited seismogram information. We utilize 3-component earthquake waveforms from almost 300 Antarctic broadband seismic stations and 26 southern mid-latitude stations from 270 earthquakes (5.5 ≤ Mw ≤ 7.0) between 2001-2003 and 2007-2016 to conduct a full-waveform adjoint inversion for Antarctica and surrounding regions of the Antarctic plate. Necessary forward and adjoint wavefield simulations are performed utilizing SPECFEM3D_GLOBE with the aid of the Texas Advanced Computing Center. We utilize phase observations from seismogram segments containing P, S, Rayleigh, and Love waves, including reflections and overtones, which are autonomously identified using FLEXWIN. The FLEXWIN analysis is carried out over a short (15-50 s) and long (initially 50-150 s) period band that target body waves, or body and surface waves, respectively. As our model is iteratively refined, the short-period corner of the long period band is gradually reduced to 25 s as the model converges over 20 linearized inversion iterations. We will briefly present this new high

  18. Big Data Challenges in Global Seismic 'Adjoint Tomography' (Invited)

    Tromp, J.; Bozdag, E.; Krischer, L.; Lefebvre, M.; Lei, W.; Smith, J.

    2013-12-01

    The challenge of imaging Earth's interior on a global scale is closely linked to the challenge of handling large data sets. The related iterative workflow involves five distinct phases, namely, 1) data gathering and culling, 2) synthetic seismogram calculations, 3) pre-processing (time-series analysis and time-window selection), 4) data assimilation and adjoint calculations, 5) post-processing (pre-conditioning, regularization, model update). In order to implement this workflow on modern high-performance computing systems, a new seismic data format is being developed. The Adaptable Seismic Data Format (ASDF) is designed to replace currently used data formats with a more flexible format that allows for fast parallel I/O. The metadata is divided into abstract categories, such as "source" and "receiver", along with provenance information for complete reproducibility. The structure of ASDF is designed keeping in mind three distinct applications: earthquake seismology, seismic interferometry, and exploration seismology. Existing time-series analysis tool kits, such as SAC and ObsPy, can be easily interfaced with ASDF so that seismologists can use robust, previously developed software packages. ASDF accommodates an automated, efficient workflow for global adjoint tomography. Manually managing the large number of simulations associated with the workflow can rapidly become a burden, especially with increasing numbers of earthquakes and stations. Therefore, it is of importance to investigate the possibility of automating the entire workflow. Scientific Workflow Management Software (SWfMS) allows users to execute workflows almost routinely. SWfMS provides additional advantages. In particular, it is possible to group independent simulations in a single job to fit the available computational resources. They also give a basic level of fault resilience as the workflow can be resumed at the correct state preceding a failure. Some of the best candidates for our particular workflow

  19. Improvement of top shield analysis technology for CANDU 6 reactor

    Kim, Kyo Yoon; Jin, Young Kwon; Lee, Sung Hee; Moon, Bok Ja; Kim, Yong Il

    1996-07-01

    As for Wolsung NPP unit 1, radiation shielding analysis was performed by using neutron diffusion codes, one-dimensional discrete ordinates code ANISN, and analytical methods. But for Wolsung NPP unit 2, 3, and 4, two-dimensional discrete ordinates code DOT substituted for neutron diffusion codes. In other words, the method of analysis and computer codes used for radiation shielding of CANDU 6 type reactor have been improved. Recently Monte Carlo MCNP code has been widely utilized in the field of radiation physics and other radiation related areas because it can describe an object sophisticately by use of three-dimensional modelling and can adopt continuous energy cross-section library. Nowadays Monte Carlo method has been reported to be competitive to discrete ordinate method in the field of radiation shielding and the former has been known to be superior to the latter for complex geometry problem. However, Monte Carlo method had not been used for radiation streaming calculation in the shielding design of CANDU type reactor. Neutron and gamma radiations are expected to be streamed from calandria through the penetrations to reactivity mechanism deck (R/M deck) because many reactivity control units which are established on R/M deck extend from R/M deck to calandria within penetrations, which are provided by guide tube extensions. More precise estimation of radiation streaming is required because R/M deck is classified as an accessible area where atomic worker can access when necessary. Therefore neutron and gamma dose rates were estimated using MCNP code on the R/M deck in the top shield system of CANDU 6 reactor. 9 tabs., 17 figs., 21 refs. (Author)

  20. Radiation shielding techniques and applications. 3. Analysis of Photon Streaming Through and Around Shield Doors

    Barnett, Marvin; Hack, Joe; Nathan, Steve; White, Travis

    2001-01-01

    Westinghouse Safety Management Solutions (Westinghouse SMS) has been tasked with providing radiological engineering design support for the new Commercial Light Water Reactor Tritium Extraction Facility (CLWR-TEF) being constructed at the Savannah River Site (SRS). The Remote Handling Building (RHB) of the CLWR-TEF will act as the receiving facility for irradiated targets used in the production of tritium for the U.S. Department of Energy (DOE). Because of the high dose rates, approaching 50 000 rads/h (500 Gy/h) from the irradiated target bundles, significant attention has been made to shielding structures within the facility. One aspect of the design that has undergone intense review is the shield doors. The RHB has six shield doors that needed to be studied with respect to photon streaming. Several aspects had to be examined to ensure that the design meets the radiation dose levels. Both the thickness and streaming issues around the door edges were designed and examined. Photon streaming through and around a shield door is a complicated problem, creating a reliance on computer modeling to perform the analyses. The computer code typically used by the Westinghouse SMS in the evaluation of photon transport through complex geometries is the MCNP Monte Carlo computer code. The complexity of the geometry within the problem can cause problems even with the Monte Carlo codes. Striking a balance between how the code handles transport through the shield door with transport through the streaming paths, particularly with the use of typical variance reduction methods, is difficult when trying to ensure that all important regions of the model are sampled appropriately. The thickness determination used a simple variance reduction technique. In construction, the shield door will not be flush against the wall, so a solid rectangular slab leaves streaming paths around the edges. Administrative controls could be used to control dose to workers; however, 10 CFR 835.1001 states

  1. 3D Space Radiation Transport in a Shielded ICRU Tissue Sphere

    Wilson, John W.; Slaba, Tony C.; Badavi, Francis F.; Reddell, Brandon D.; Bahadori, Amir A.

    2014-01-01

    A computationally efficient 3DHZETRN code capable of simulating High Charge (Z) and Energy (HZE) and light ions (including neutrons) under space-like boundary conditions with enhanced neutron and light ion propagation was recently developed for a simple homogeneous shield object. Monte Carlo benchmarks were used to verify the methodology in slab and spherical geometry, and the 3D corrections were shown to provide significant improvement over the straight-ahead approximation in some cases. In the present report, the new algorithms with well-defined convergence criteria are extended to inhomogeneous media within a shielded tissue slab and a shielded tissue sphere and tested against Monte Carlo simulation to verify the solution methods. The 3D corrections are again found to more accurately describe the neutron and light ion fluence spectra as compared to the straight-ahead approximation. These computationally efficient methods provide a basis for software capable of space shield analysis and optimization.

  2. Laboratory tests on neutron shields for gamma-ray detectors in space

    Hong, J; Hailey, C J

    2000-01-01

    Shields capable of suppressing neutron-induced background in new classes of gamma-ray detectors such as CdZnTe are becoming important for a variety of reasons. These include a high cross section for neutron interactions in new classes of detector materials as well as the inefficient vetoing of neutron-induced background in conventional active shields. We have previously demonstrated through Monte-Carlo simulations how our new approach, supershields, is superior to the monolithic, bi-atomic neutron shields which have been developed in the past. We report here on the first prototype models for supershields based on boron and hydrogen. We verify the performance of these supershields through laboratory experiments. These experimental results, as well as measurements of conventional monolithic neutron shields, are shown to be consistent with Monte-Carlo simulations. We discuss the implications of this experiment for designs of supershields in general and their application to future hard X-ray/gamma-ray experiments...

  3. Shielding calculations in support of the Spallation Neutron Source (SNS) proton beam transport system

    Johnson, Jeffrey O.; Gallmeier, Franz X.; Popova, Irina

    2002-01-01

    Determining the bulk shielding requirements for accelerator environments is generally an easy task compared to analyzing the radiation transport through the complex shield configurations and penetrations typically associated with the detailed Title II design efforts of a facility. Shielding calculations for penetrations in the SNS accelerator environment are presented based on hybrid Monte Carlo and discrete ordinates particle transport methods. This methodology relies on coupling tools that map boundary surface leakage information from the Monte Carlo calculations to boundary sources for one-, two-, and three-dimensional discrete ordinates calculations. The paper will briefly introduce the coupling tools for coupling MCNPX to the one-, two-, and three-dimensional discrete ordinates codes in the DOORS code suite. The paper will briefly present typical applications of these tools in the design of complex shield configurations and penetrations in the SNS proton beam transport system

  4. Shielding considerations for neutral-beam injection systems

    de Seynes, X.

    1983-03-01

    Results of a study on the geometry of an FED-A Neutral Beam Injector beamline duct shield are presented. Also included is a calculation of dose rates, as a function of time, from an activated NBI. The shielding investigations consisted of varying the parameters of the geometry and transporting particles through it using the MCNP Monte-Carlo code. The dose rates were calculated by the ACDOS3 code using realistic MCNP results. A final-to-incident flux ratio of 6.5 x 10 -7 can be achieved through the use of a 65.5 cm reentry duct. This is for a realistic source and pure water shielding material. The activated NBI produced a dose rate of 15.9 mrem/hr two and a half days after shutdown of the reactor

  5. The evaluation of the radiation shielding ability of lead glass

    Tsuda, Keisuke; Fukushi, Masahiro; Myojoyama, Atsushi; Kitamura, Hideaki; Nakaya, Giichiro; Hassan, Nabil; Inoue, Kazumasa; Kimura, Junichi; Sawaguchi, Masato; Kinase, Sakae; Saito, Kimiaki

    2008-01-01

    Positron emission tomography (PET) scanning with the tracer 2-[F-18] Fluoro-2deoxy-D-glucose (FDG) is widely used in the clinical PET. However, the photon energy used in the PET scans is considerably higher than that of the X-rays traditionally used in the diagnoses. The radiation protection in the PET institution, therefore, is the remaining problem. Meanwhile, lead glass has attracted considerable attention as a radiation-shielding material for the PET institution. The aim of the present study was to evaluate the radiation-shielding ability of the lead glass against the positron emitters. The shielding ability evaluations were done both in the actual experiments and in the Monte Carlo simulation. The lead glass, the object of evaluation in this study, proved to have sufficient protective effect. The development and the spread of a thinner and lighter lead glass with the same effective dose transmission factor should be expected in the near future. (author)

  6. Successful vectorization - reactor physics Monte Carlo code

    Martin, W.R.

    1989-01-01

    Most particle transport Monte Carlo codes in use today are based on the ''history-based'' algorithm, wherein one particle history at a time is simulated. Unfortunately, the ''history-based'' approach (present in all Monte Carlo codes until recent years) is inherently scalar and cannot be vectorized. In particular, the history-based algorithm cannot take advantage of vector architectures, which characterize the largest and fastest computers at the current time, vector supercomputers such as the Cray X/MP or IBM 3090/600. However, substantial progress has been made in recent years in developing and implementing a vectorized Monte Carlo algorithm. This algorithm follows portions of many particle histories at the same time and forms the basis for all successful vectorized Monte Carlo codes that are in use today. This paper describes the basic vectorized algorithm along with descriptions of several variations that have been developed by different researchers for specific applications. These applications have been mainly in the areas of neutron transport in nuclear reactor and shielding analysis and photon transport in fusion plasmas. The relative merits of the various approach schemes will be discussed and the present status of known vectorization efforts will be summarized along with available timing results, including results from the successful vectorization of 3-D general geometry, continuous energy Monte Carlo. (orig.)

  7. Neutron shielding and activation of the MASTU device and surrounds

    Taylor, David, E-mail: david.taylor@ccfe.ac.uk [EURATOM/CCFE Fusion Association, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Lilley, Steven; Turner, Andrew [EURATOM/CCFE Fusion Association, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Davis, Andrew [Now at College of Engineering, University of Wisconsin, Madison, WI 53706 (United States)

    2014-10-15

    Highlights: We model neutron shielding for the planned MASTU device; nadequacies in the existing shielding design are remedied; Levels of public exposure are considered; We model activated gamma emission for the device under a worst case scenario. Abstract: A significant functional upgrade is planned for the Mega Ampere Spherical Tokamak (MAST) device, located at Culham in the UK, including the implementation of a notably greater neutral beam injection power. This upgrade will cause the emission of a substantially increased intensity of neutron radiation for a substantially increased amount of time upon operation of the device. Existing shielding and activation precautions are shown to prove insufficient in some regards, and recommendations for improvements are made, including the following areas: shielding doors to MAST shielded facility enclosure (known as “the blockhouse”); north access tunnel; blockhouse roof; west cabling duct. In addition, some specific neutronic dose rate questions are addressed and answered; those discussed here relate to shielding penetrations and dose rate reflected from the air above the device (“skyshine”). It is shown that the alterations to shielding and area access reduce the dose rate in unrestricted areas from greater than 100 μSv/h to less than 2 μSv/h averaged over the working day. The tools used for this analysis are the MCNP (Monte Carlo N-particle) code, used to calculate the three-dimensional spatial distribution of neutron and photon dose rates in and around the device and its shields, and the nuclear inventory code FISPACT, run under the umbrella code MCR2S, used to calculate the time-dependent shutdown dose rate in the region of the device at several decay times.

  8. A demonstration of adjoint methods for multi-dimensional remote sensing of the atmosphere and surface

    Martin, William G.K.; Hasekamp, Otto P.

    2018-01-01

    Highlights: • We demonstrate adjoint methods for atmospheric remote sensing in a two-dimensional setting. • Searchlight functions are used to handle the singularity of measurement response functions. • Adjoint methods require two radiative transfer calculations to evaluate the measurement misfit function and its derivatives with respect to all unknown parameters. • Synthetic retrieval studies show the scalability of adjoint methods to problems with thousands of measurements and unknown parameters. • Adjoint methods and the searchlight function technique are generalizable to 3D remote sensing. - Abstract: In previous work, we derived the adjoint method as a computationally efficient path to three-dimensional (3D) retrievals of clouds and aerosols. In this paper we will demonstrate the use of adjoint methods for retrieving two-dimensional (2D) fields of cloud extinction. The demonstration uses a new 2D radiative transfer solver (FSDOM). This radiation code was augmented with adjoint methods to allow efficient derivative calculations needed to retrieve cloud and surface properties from multi-angle reflectance measurements. The code was then used in three synthetic retrieval studies. Our retrieval algorithm adjusts the cloud extinction field and surface albedo to minimize the measurement misfit function with a gradient-based, quasi-Newton approach. At each step we compute the value of the misfit function and its gradient with two calls to the solver FSDOM. First we solve the forward radiative transfer equation to compute the residual misfit with measurements, and second we solve the adjoint radiative transfer equation to compute the gradient of the misfit function with respect to all unknowns. The synthetic retrieval studies verify that adjoint methods are scalable to retrieval problems with many measurements and unknowns. We can retrieve the vertically-integrated optical depth of moderately thick clouds as a function of the horizontal coordinate. It is also

  9. A practical discrete-adjoint method for high-fidelity compressible turbulence simulations

    Vishnampet, Ramanathan; Bodony, Daniel J.; Freund, Jonathan B.

    2015-01-01

    Methods and computing hardware advances have enabled accurate predictions of complex compressible turbulence phenomena, such as the generation of jet noise that motivates the present effort. However, limited understanding of underlying physical mechanisms restricts the utility of such predictions since they do not, by themselves, indicate a route to design improvements. Gradient-based optimization using adjoints can circumvent the flow complexity to guide designs, though this is predicated on the availability of a sufficiently accurate solution of the forward and adjoint systems. These are challenging to obtain, since both the chaotic character of the turbulence and the typical use of discretizations near their resolution limits in order to efficiently represent its smaller scales will amplify any approximation errors made in the adjoint formulation. Formulating a practical exact adjoint that avoids such errors is especially challenging if it is to be compatible with state-of-the-art simulation methods used for the turbulent flow itself. Automatic differentiation (AD) can provide code to calculate a nominally exact adjoint, but existing general-purpose AD codes are inefficient to the point of being prohibitive for large-scale turbulence simulations. Here, we analyze the compressible flow equations as discretized using the same high-order workhorse methods used for many high-fidelity compressible turbulence simulations, and formulate a practical space–time discrete-adjoint method without changing the basic discretization. A key step is the definition of a particular discrete analog of the continuous norm that defines our cost functional; our selection leads directly to an efficient Runge–Kutta-like scheme, though it would be just first-order accurate if used outside the adjoint formulation for time integration, with finite-difference spatial operators for the adjoint system. Its computational cost only modestly exceeds that of the flow equations. We confirm that

  10. External dosimetry sources and shielding

    Calisto, Washington

    1994-01-01

    A definition of external dosimetry r external sources dosimetry,physical and mathematical treatment of the interaction of gamma radiation with a minimal area in that direction. Concept of attenuation coefficient, cumulated effect by polyenergetic sources, exposition rate, units, cumulated dose,shielding, foton shielding, depth calculation, materials used for shielding.Beta shielding, consideration of range and maximum β energy , low stopping radiation by use of low Z shielding. Tables for β energy of β emitters, I (tau) factor, energy-range curves for β emitters in aqueous media, gamma attenuation factors for U, W and Pb. Y factor for bone tissue,muscle and air, build-up factors

  11. An adjoint-based scheme for eigenvalue error improvement

    Merton, S.R.; Smedley-Stevenson, R.P.; Pain, C.C.; El-Sheikh, A.H.; Buchan, A.G.

    2011-01-01

    A scheme for improving the accuracy and reducing the error in eigenvalue calculations is presented. Using a rst order Taylor series expansion of both the eigenvalue solution and the residual of the governing equation, an approximation to the error in the eigenvalue is derived. This is done using a convolution of the equation residual and adjoint solution, which is calculated in-line with the primal solution. A defect correction on the solution is then performed in which the approximation to the error is used to apply a correction to the eigenvalue. The method is shown to dramatically improve convergence of the eigenvalue. The equation for the eigenvalue is shown to simplify when certain normalizations are applied to the eigenvector. Two such normalizations are considered; the rst of these is a fission-source type of normalisation and the second is an eigenvector normalisation. Results are demonstrated on a number of demanding elliptic problems using continuous Galerkin weighted nite elements. Moreover, the correction scheme may also be applied to hyperbolic problems and arbitrary discretization. This is not limited to spatial corrections and may be used throughout the phase space of the discrete equation. The applied correction not only improves fidelity of the calculation, it allows assessment of the reliability of numerical schemes to be made and could be used to guide mesh adaption algorithms or to automate mesh generation schemes. (author)

  12. Surface spectra of Weyl semimetals through self-adjoint extensions

    Seradjeh, Babak; Vennettilli, Michael

    2018-02-01

    We apply the method of self-adjoint extensions of Hermitian operators to the low-energy, continuum Hamiltonians of Weyl semimetals in bounded geometries and derive the spectrum of the surface states on the boundary. This allows for the full characterization of boundary conditions and the surface spectra on surfaces both normal to the Weyl node separation as well as parallel to it. We show that the boundary conditions for quadratic bulk dispersions are, in general, specified by a U (2 ) matrix relating the wave function and its derivatives normal to the surface. We give a general procedure to obtain the surface spectra from these boundary conditions and derive them in specific cases of bulk dispersion. We consider the role of global symmetries in the boundary conditions and their effect on the surface spectrum. We point out several interesting features of the surface spectra for different choices of boundary conditions, such as a Mexican-hat shaped dispersion on the surface normal to Weyl node separation. We find that the existence of bound states, Fermi arcs, and the shape of their dispersion, depend on the choice of boundary conditions. This illustrates the importance of the physics at and near the boundaries in the general statement of bulk-boundary correspondence.

  13. Shielding benchmark test

    Kawai, Masayoshi

    1984-01-01

    Iron data in JENDL-2 have been tested by analyzing shielding benchmark experiments for neutron transmission through iron block performed at KFK using CF-252 neutron source and at ORNL using collimated neutron beam from reactor. The analyses are made by a shielding analysis code system RADHEAT-V4 developed at JAERI. The calculated results are compared with the measured data. As for the KFK experiments, the C/E values are about 1.1. For the ORNL experiments, the calculated values agree with the measured data within an accuracy of 33% for the off-center geometry. The d-t neutron transmission measurements through carbon sphere made at LLNL are also analyzed preliminarily by using the revised JENDL data for fusion neutronics calculation. (author)

  14. Neutron shielding material

    Suzuki, Shigenori; Iimori, Hiroshi; Kobori, Junzo.

    1980-01-01

    Purpose: To provide a neutron shielding material which incorporates preferable shielding capacity, heat resistance, fire resistance and workability by employing a mixture of thermosetting resin, polyethylene and aluminium hydroxide in special range ratio and curing it. Constitution: A mixture containing 20 to 60% by weight of thermosetting resin having preferable heat resistance, 10 to 40% by weight of polyethylene powder having high hydrogen atom density and 1000 to 60000 of molecular weight, and 15 to 55% by weight of Al(OH) 3 for imparting fire resistance and self-fire extinguishing property thereto is cured. At this time approx. 0.5 to 5% of curing catalyst of the thermosetting resin is contained in 100 parts by weight of the mixture. (Sekiya, K.)

  15. Radiation shielding wall structure

    Nishimura, Yoshitaka; Oka, Shinji; Kan, Toshihiko; Misato, Takeshi.

    1990-01-01

    A space between a pair of vertical steel plates laterally disposed in parallel at an optional distance has a structure of a plurality of vertically extending tranks partitioned laterally by vertically placed steel plates. Then, cements are grouted to the tranks. Strip-like steel plates each having a thickness greater than the gap between the each of the vertically placed steel plates and the cement are bonded each at the surface for each of the vertically placed steel plates opposing to the cements. A protrusion of a strip width having radiation shielding performance substantially identical with that by the thickness of the cement is disposed in the strip-like steel plates. With such a constitution, a safety radiation shielding wall structure with no worry of radiation intrusion to gaps, if formed, between the steel plates and the grouted cements due to shrinkage of the cements. (I.N.)

  16. Radiation shielding material

    Kawakubo, Takamasa; Yamada, Fumiyuki; Nakazato, Kenjiro.

    1976-01-01

    Purpose: To provide a material, which is used for printing a samples name and date on an X-ray photographic film at the same time an X-ray radiography. Constitution: A radiation shielding material of a large mass absorption coefficient such as lead oxide, barium oxide, barium sulfate, etc. is added to a solution of a radiation permeable substance capable of imparting cold plastic fluidity (such as microcrystalline wax, paraffin, low molecular polyethylene, polyvinyl chloride, etc.). The resultant system is agitated and then cooled, and thereafter it is press fitted to or bonded to a base in the form of a film of a predetermined thickness. This radiation shielding layer is scraped off by using a writing tool to enter information to be printed in a photographic film, and then it is laid over the film and exposed to X-radiation to thereby print the information on the film. (Seki, T.)

  17. Multilayer radiation shield

    Urbahn, John Arthur; Laskaris, Evangelos Trifon

    2009-06-16

    A power generation system including: a generator including a rotor including a superconductive rotor coil coupled to a rotatable shaft; a first prime mover drivingly coupled to the rotatable shaft; and a thermal radiation shield, partially surrounding the rotor coil, including at least a first sheet and a second sheet spaced apart from the first sheet by centripetal force produced by the rotatable shaft. A thermal radiation shield for a generator including a rotor including a super-conductive rotor coil including: a first sheet having at least one surface formed from a low emissivity material; and at least one additional sheet having at least one surface formed from a low emissivity material spaced apart from the first sheet by centripetal force produced by the rotatable shaft, wherein each successive sheet is an incrementally greater circumferential arc length and wherein the centripetal force shapes the sheets into a substantially catenary shape.

  18. Light shielding apparatus

    Miller, Richard Dean; Thom, Robert Anthony

    2017-10-10

    A light shielding apparatus for blocking light from reaching an electronic device, the light shielding apparatus including left and right support assemblies, a cross member, and an opaque shroud. The support assemblies each include primary support structure, a mounting element for removably connecting the apparatus to the electronic device, and a support member depending from the primary support structure for retaining the apparatus in an upright orientation. The cross member couples the left and right support assemblies together and spaces them apart according to the size and shape of the electronic device. The shroud may be removably and adjustably connectable to the left and right support assemblies and configured to take a cylindrical dome shape so as to form a central space covered from above. The opaque shroud prevents light from entering the central space and contacting sensitive elements of the electronic device.

  19. Shielding container for radioactive isotopes

    Sumi, Tetsuo; Tosa, Masayoshi; Hatogai, Tatsuaki.

    1975-01-01

    Object: To effect opening and closing bidirectional radiation used particularly for a gamma densimeter or the like by one operation. Structure: This device comprises a rotatable shielding body for receiving radioactive isotope in the central portion thereof and having at least two radiation openings through which radiation is taken out of the isotope, and a shielding container having openings corresponding to the first mentioned radiation openings, respectively. The radioactive isotope is secured to a rotational shaft of the shielding body, and the shielding body is rotated to register the openings of the shielding container with the openings of the shielding body or to shield the openings, thereby effecting radiation and cut off of gamma ray in the bidirection by one operation. (Kamimura, M.)

  20. Primary shield displacement and bowing

    Scott, K.V.

    1978-01-01

    The reactor primary shield is constructed of high density concrete and surrounds the reactor core. The inlet, outlet and side primary shields were constructed in-place using 2.54 cm (1 in) thick steel plates as the forms. The plates remained as an integral part of the shields. The elongation of the pressure tubes due to thermal expansion and pressurization is not moving through the inlet nozzle hardware as designed but is accommodated by outward displacement and bowing of the inlet and outlet shields. Excessive distortion of the shields may result in gas seal failures, intolerable helium gas leaks, increased argon-41 emissions, and shield cooling tube failures. The shield surveillance and testing results are presented

  1. A spectral analysis of the domain decomposed Monte Carlo method for linear systems

    Slattery, S. R.; Wilson, P. P. H. [Engineering Physics Department, University of Wisconsin - Madison, 1500 Engineering Dr., Madison, WI 53706 (United States); Evans, T. M. [Oak Ridge National Laboratory, 1 Bethel Valley Road, Oak Ridge, TN 37830 (United States)

    2013-07-01

    The domain decomposed behavior of the adjoint Neumann-Ulam Monte Carlo method for solving linear systems is analyzed using the spectral properties of the linear operator. Relationships for the average length of the adjoint random walks, a measure of convergence speed and serial performance, are made with respect to the eigenvalues of the linear operator. In addition, relationships for the effective optical thickness of a domain in the decomposition are presented based on the spectral analysis and diffusion theory. Using the effective optical thickness, the Wigner rational approximation and the mean chord approximation are applied to estimate the leakage fraction of stochastic histories from a domain in the decomposition as a measure of parallel performance and potential communication costs. The one-speed, two-dimensional neutron diffusion equation is used as a model problem to test the models for symmetric operators. In general, the derived approximations show good agreement with measured computational results. (authors)

  2. A spectral analysis of the domain decomposed Monte Carlo method for linear systems

    Slattery, S. R.; Wilson, P. P. H.; Evans, T. M.

    2013-01-01

    The domain decomposed behavior of the adjoint Neumann-Ulam Monte Carlo method for solving linear systems is analyzed using the spectral properties of the linear operator. Relationships for the average length of the adjoint random walks, a measure of convergence speed and serial performance, are made with respect to the eigenvalues of the linear operator. In addition, relationships for the effective optical thickness of a domain in the decomposition are presented based on the spectral analysis and diffusion theory. Using the effective optical thickness, the Wigner rational approximation and the mean chord approximation are applied to estimate the leakage fraction of stochastic histories from a domain in the decomposition as a measure of parallel performance and potential communication costs. The one-speed, two-dimensional neutron diffusion equation is used as a model problem to test the models for symmetric operators. In general, the derived approximations show good agreement with measured computational results. (authors)

  3. Comparison of the Adjoint and Adjoint-Free 4dVar Assimilation of the Hydrographic and Velocity Observations in the Adriatic Sea

    2015-11-10

    approach is the ab- sence of the necessity to develop and maintain tangent linear and adjoint codes and its flexibility in adaptation to various...uadratic term in the right hand side of (B.10) is negligible. In the re- orted experiments we kept it in place since the value of ε was close o 0.01 and

  4. Muon shielding for PEP

    Jenkins, T.M.; Thomas, R.H.

    1974-01-01

    The first stage of construction of PEP will consist of electron and positron storage rings. At a later date a 200 GeV proton storage ring may be added. It is judicious therefore, to ensure that the first and second phases of construction are compatible with each other. One of several factors determining the elevation at which the storage rings will be constructed is the necessity to provide adequate radiation shielding. The overhead shielding of PEP is determined by the reproduction of neutrons in the hadron cascade generated by primary protons lost from the storage ring. The minimum overburden planned for PEP is 5.5 meters of earth (1100 gm cm/sup /minus/2/). To obtain a rough estimate of the magnitude of the muon radiation problem this note presents some preliminary calculations. Their purpose is intended merely to show that the presently proposed design for PEP will present no major shielding problems should the protons storage ring be installed. More detailed calculations will be made using muon yield computer codes developed at CERN and NAL and muon transport codes developed at SLAC, when details of the proton storage ring become settled. 9 refs., 4 figs

  5. Shielding calculations for NET

    Verschuur, K.A.; Hogenbirk, A.

    1991-05-01

    In the European Fusion Technology Programme there is only a small activity on research and development for fusion neutronics. Never-the-less, looking further than blanket design now, as ECN is getting involved in design of radiation shields for the coils and biological shields, it becomes apparent that fusion neutronics as a whole still needs substantial development. Existing exact codes for calculation of complex geometries like MCNP and DORT/TORT are put over the limits of their numerical capabilities, whilst approximate codes for complex geometries like FURNACE and MERCURE4 are put over the limits of their modelling capabilities. The main objective of this study is just to find out how far we can get with existing codes in obtaining reliable values for the radiation levels inside and outside the cryostat/shield during operation and after shut-down. Starting with a 1D torus model for preliminary parametric studies, more dimensional approximation of the torus or parts of it including the main heterogeneities should follow. Regular contacts with the NET-Team are kept, to be aware of main changes in NET design that might affect our calculation models. Work on the contract started 1 July 1990. The technical description of the contract is given. (author). 14 refs.; 4 figs.; 1 tab

  6. The computer code system for reactor radiation shielding in design of nuclear power plant

    Li Chunhuai; Fu Shouxin; Liu Guilian

    1995-01-01

    The computer code system used in reactor radiation shielding design of nuclear power plant includes the source term codes, discrete ordinate transport codes, Monte Carlo and Albedo Monte Carlo codes, kernel integration codes, optimization code, temperature field code, skyshine code, coupling calculation codes and some processing codes for data libraries. This computer code system has more satisfactory variety of codes and complete sets of data library. It is widely used in reactor radiation shielding design and safety analysis of nuclear power plant and other nuclear facilities

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

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

    2016-07-07

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

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

    2002-01-01

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

  9. A response matrix method for slab-geometry discrete ordinates adjoint calculations in energy-dependent source-detector problems

    Mansur, Ralph S.; Moura, Carlos A., E-mail: ralph@ime.uerj.br, E-mail: demoura@ime.uerj.br [Universidade do Estado do Rio de Janeiro (UERJ), RJ (Brazil). Departamento de Engenharia Mecanica; Barros, Ricardo C., E-mail: rcbarros@pq.cnpq.br [Universidade do Estado do Rio de Janeiro (UERJ), Nova Friburgo, RJ (Brazil). Departamento de Modelagem Computacional

    2017-07-01

    Presented here is an application of the Response Matrix (RM) method for adjoint discrete ordinates (S{sub N}) problems in slab geometry applied to energy-dependent source-detector problems. The adjoint RM method is free from spatial truncation errors, as it generates numerical results for the adjoint angular fluxes in multilayer slabs that agree with the numerical values obtained from the analytical solution of the energy multigroup adjoint SN equations. Numerical results are given for two typical source-detector problems to illustrate the accuracy and the efficiency of the offered RM computer code. (author)

  10. Self-consistent adjoint analysis for topology optimization of electromagnetic waves

    Deng, Yongbo; Korvink, Jan G.

    2018-05-01

    In topology optimization of electromagnetic waves, the Gâteaux differentiability of the conjugate operator to the complex field variable results in the complexity of the adjoint sensitivity, which evolves the original real-valued design variable to be complex during the iterative solution procedure. Therefore, the self-inconsistency of the adjoint sensitivity is presented. To enforce the self-consistency, the real part operator has been used to extract the real part of the sensitivity to keep the real-value property of the design variable. However, this enforced self-consistency can cause the problem that the derived structural topology has unreasonable dependence on the phase of the incident wave. To solve this problem, this article focuses on the self-consistent adjoint analysis of the topology optimization problems for electromagnetic waves. This self-consistent adjoint analysis is implemented by splitting the complex variables of the wave equations into the corresponding real parts and imaginary parts, sequentially substituting the split complex variables into the wave equations with deriving the coupled equations equivalent to the original wave equations, where the infinite free space is truncated by the perfectly matched layers. Then, the topology optimization problems of electromagnetic waves are transformed into the forms defined on real functional spaces instead of complex functional spaces; the adjoint analysis of the topology optimization problems is implemented on real functional spaces with removing the variational of the conjugate operator; the self-consistent adjoint sensitivity is derived, and the phase-dependence problem is avoided for the derived structural topology. Several numerical examples are implemented to demonstrate the robustness of the derived self-consistent adjoint analysis.

  11. Comparison of MCNP4C and experimental results on neutron and gamma ray shielding effects for materials

    Cha, Kyoon Ho; Lee, Eun Ki [KEPRI, Taejon (Korea, Republic of)

    2004-07-01

    MCNP code is a general-purpose Monte Carlo radiation transport code that can numerically simulate neutron, photon, and electron transport. Increasing the speed of computing machine is making numerical transport simulation more attractive and has led to the widespread use of such code. This code can be used for general radiation shielding and criticality accident alarm system related dose calculations, so that the version 4C2 of this code was used to evaluate the shielding effect against neutron and gamma ray experiments. The Ueki experiments were used for neutron shielding effects for materials, and the Kansas State University (KSU) photon skyshine experiments of 1977 were tested for gamma ray shielding effects.

  12. Design and Shielding of Radiotherapy Treatment Facilities; IPEM Report 75, 2nd Edition

    Horton, Patrick; Eaton, David

    2017-07-01

    Design and Shielding of Radiotherapy Treatment Facilities provides readers with a single point of reference for protection advice to the construction and modification of radiotherapy facilities. The book assembles a faculty of national and international experts on all modalities including megavoltage and kilovoltage photons, brachytherapy and high-energy particles, and on conventional and Monte Carlo shielding calculations. This book is a comprehensive reference for qualified experts and radiation-shielding designers in radiation physics and also useful to anyone involved in the design of radiotherapy facilities.

  13. A contribution to shielding effectiveness analysis of shielded tents

    Vranić Zoran M.

    2004-01-01

    Full Text Available An analysis of shielding effectiveness (SE of the shielded tents made of the metallised fabrics is given. First, two electromagnetic characteristic fundamental for coupling through electrically thin shield, the skin depth break frequency and the surface resistance or transfer impedance, is defined and analyzed. Then, the transfer function and the SE are analyzed regarding to the frequency range of interest to the Electromagnetic Compatibility (EMC Community.

  14. Self-adjoint extensions and spectral analysis in the Calogero problem

    Gitman, D M [Institute of Physics, University of Sao Paulo (Brazil); Tyutin, I V; Voronov, B L [Lebedev Physical Institute, Moscow (Russian Federation)], E-mail: gitman@dfn.if.usp.br, E-mail: tyutin@lpi.ru, E-mail: voronov@lpi.ru

    2010-04-09

    In this paper, we present a mathematically rigorous quantum-mechanical treatment of a one-dimensional motion of a particle in the Calogero potential {alpha}x{sup -2}. Although the problem is quite old and well studied, we believe that our consideration based on a uniform approach to constructing a correct quantum-mechanical description for systems with singular potentials and/or boundaries, proposed in our previous works, adds some new points to its solution. To demonstrate that a consideration of the Calogero problem requires mathematical accuracy, we discuss some 'paradoxes' inherent in the 'naive' quantum-mechanical treatment. Using a self-adjoint extension method, we construct and study all possible self-adjoint operators (self-adjoint Hamiltonians) associated with a formal differential expression for the Calogero Hamiltonian. In particular, we discuss a spontaneous scale-symmetry breaking associated with self-adjoint extensions. A complete spectral analysis of all self-adjoint Hamiltonians is presented.

  15. Evaluation of the performance of peridotite aggregates for radiation shielding concrete

    Wang, Jinjun; Li, Guofeng; Meng, Dechuan

    2014-01-01

    Highlights: • Using peridotite rich in crystal water as aggregates of radiation-shielding concrete. • Performance of peridotite concrete is simulated and compared with ordinary concrete. • Performance of concrete samples is tested. • Neutron shielding performance can be significantly enhanced by peridotite aggregates. - Abstract: Peridotite is a kind of material that is rich in crystal water. In this paper, peridotite is used as fine and coarse aggregates for radiation shielding concrete. The transmission data of different concrete thickness and different energy neutron are calculated using Monte-Carlo method. The neutron shielding performance of the peridotite concrete samples are tested using 241 Am-Be neutron source. The results show that the peridotite is an excellent neutron shielding material

  16. Deep-penetration calculations in concrete and iron for shielding of proton therapy accelerators

    Sheu, Rong-Jiun; Chen, Yen-Fu; Lin, Uei-Tyng; Jiang, Shiang-Huei

    2012-01-01

    Proton accelerators in the energy range of approximately 200 MeV have become increasingly popular for cancer treatment in recent years. These proton therapy facilities usually involve bulky concrete or iron in their shielding design or accelerator structure. Simple shielding data, such as source terms or attenuation lengths for various proton energies and materials are useful in designing accelerator shielding. Understanding the appropriateness or uncertainties associated with these data, which are largely generated from Monte Carlo simulations, is critical to the quality of a shielding design. This study demonstrated and investigated the problems of deep-penetration calculations on the estimation of shielding parameters through an extensive comparison between the FLUKA and MCNPX calculations for shielding against a 200-MeV proton beam hitting an iron target. Simulations of double-differential neutron production from proton bombardment were validated by comparison with experimental data. For the concrete shielding, the FLUKA calculated depth–dose distributions were consistent with the MCNPX results, except for some discrepancies in backward directions. However, for the iron shielding, if FLUKA is used inappropriately then overestimation of neutron attenuation can be expected as shown by this work because of the multigroup treatment for low-energy neutrons in FLUKA. Two neutron energy group structures, three degrees of self-shielding correction, and two iron compositions were considered in this study. Significant variation of the resulting attenuation lengths indicated the importance of problem-dependent multigroup cross sections and proper modeling of iron composition in deep-penetration calculations.

  17. Status of Monte Carlo at Los Alamos

    Thompson, W.L.; Cashwell, E.D.; Godfrey, T.N.K.; Schrandt, R.G.; Deutsch, O.L.; Booth, T.E.

    1980-05-01

    Four papers were presented by Group X-6 on April 22, 1980, at the Oak Ridge Radiation Shielding Information Center (RSIC) Seminar-Workshop on Theory and Applications of Monte Carlo Methods. These papers are combined into one report for convenience and because they are related to each other. The first paper (by Thompson and Cashwell) is a general survey about X-6 and MCNP and is an introduction to the other three papers. It can also serve as a resume of X-6. The second paper (by Godfrey) explains some of the details of geometry specification in MCNP. The third paper (by Cashwell and Schrandt) illustrates calculating flux at a point with MCNP; in particular, the once-more-collided flux estimator is demonstrated. Finally, the fourth paper (by Thompson, Deutsch, and Booth) is a tutorial on some variance-reduction techniques. It should be required for a fledging Monte Carlo practitioner

  18. Neutronic reactor thermal shield

    Lowe, P.E.

    1976-01-01

    A shield for a nuclear reactor includes at least two layers of alternating wide and narrow rectangular blocks so arranged that the spaces between blocks in adjacent layers are out of registry, each block having an opening therein equally spaced from the sides of the blocks and nearer the top of the block than the bottom, the distance from the top of the block to the opening in one layer being different from this distance in adjacent layers, openings in blocks in adjacent layers being in registry. 1 claim, 7 drawing figures

  19. A shield against distraction

    Halin, N.; Marsh, J.E.; Hellman, A.; Hellstrom, I.; Sörqvist, Patrik

    2014-01-01

    In this paper, we apply the basic idea of a trade-off between the level of concentration and distractibility to test whether a manipulation of task difficulty can shield against distraction. Participants read, either in quiet or with a speech noise background, texts that were displayed either in an easy-to-read or a hard-to-read font. Background speech impaired prose recall, but only when the text was displayed in the easy-to-read font. Most importantly, recall was better in the background sp...

  20. Measuring space radiation shielding effectiveness

    Bahadori Amir; Semones Edward; Ewert Michael; Broyan James; Walker Steven

    2017-01-01

    Passive radiation shielding is one strategy to mitigate the problem of space radiation exposure. While space vehicles are constructed largely of aluminum, polyethylene has been demonstrated to have superior shielding characteristics for both galactic cosmic rays and solar particle events due to the high hydrogen content. A method to calculate the shielding effectiveness of a material relative to reference material from Bragg peak measurements performed using energetic heavy charged particles ...

  1. Selective shielding device for scintiphotography

    Harper, J.W.; Kay, T.D.

    1976-01-01

    A selective shielding device to be used in combination with a scintillation camera is described. The shielding device is a substantially oval-shaped configuration removably secured to the scintillation camera. As a result of this combination scanning of preselected areas of a patient can be rapidly and accurately performed without the requirement of mounting any type of shielding paraphernalia on the patient. 1 claim, 2 drawing figures

  2. Hybrid SN/Monte Carlo research and results

    Baker, R.S.

    1993-01-01

    The neutral particle transport equation is solved by a hybrid method that iteratively couples regions where deterministic (S N ) and stochastic (Monte Carlo) methods are applied. The Monte Carlo and S N regions are fully coupled in the sense that no assumption is made about geometrical separation or decoupling. The hybrid Monte Carlo/S N method provides a new means of solving problems involving both optically thick and optically thin regions that neither Monte Carlo nor S N is well suited for by themselves. The hybrid method has been successfully applied to realistic shielding problems. The vectorized Monte Carlo algorithm in the hybrid method has been ported to the massively parallel architecture of the Connection Machine. Comparisons of performance on a vector machine (Cray Y-MP) and the Connection Machine (CM-2) show that significant speedups are obtainable for vectorized Monte Carlo algorithms on massively parallel machines, even when realistic problems requiring variance reduction are considered. However, the architecture of the Connection Machine does place some limitations on the regime in which the Monte Carlo algorithm may be expected to perform well

  3. SHIELD verification and validation report

    Boman, C.

    1992-02-01

    This document outlines the verification and validation effort for the SHIELD, SHLDED, GEDIT, GENPRT, FIPROD, FPCALC, and PROCES modules of the SHIELD system code. Along with its predecessors, SHIELD has been in use at the Savannah River Site (SRS) for more than ten years. During this time the code has been extensively tested and a variety of validation documents have been issued. The primary function of this report is to specify the features and capabilities for which SHIELD is to be considered validated, and to reference the documents that establish the validation

  4. Multifunctional Hot Structure Heat Shield

    National Aeronautics and Space Administration — This project is performing preliminary development of a Multifunctional Hot Structure (HOST) heat shield for planetary entry. Results of this development will...

  5. Radiation shield for nuclear reactors

    Weissenfluh, J.A.

    1978-01-01

    A shield for use with nuclear reactor systems to attenuate radiation resulting from reactor operation is described. The shield comprises a container preferably of a thin, flexible or elastic material, which may be in the form of a bag, a mattress, a toroidal segment or toroid or the like filled with radiation attenuating liuid. Means are provided in the container for filling and draining the container in place. Due to its flexibility, the shield readily conforms to irregularities in surfaces with which it may be in contact in a shielding position

  6. Shielding benchmark tests of JENDL-3

    Kawai, Masayoshi; Hasegawa, Akira; Ueki, Kohtaro; Yamano, Naoki; Sasaki, Kenji; Matsumoto, Yoshihiro; Takemura, Morio; Ohtani, Nobuo; Sakurai, Kiyoshi.

    1994-03-01

    The integral test of neutron cross sections for major shielding materials in JENDL-3 has been performed by analyzing various shielding benchmark experiments. For the fission-like neutron source problem, the following experiments are analyzed: (1) ORNL Broomstick experiments for oxygen, iron and sodium, (2) ASPIS deep penetration experiments for iron, (3) ORNL neutron transmission experiments for iron, stainless steel, sodium and graphite, (4) KfK leakage spectrum measurements from iron spheres, (5) RPI angular neutron spectrum measurements in a graphite block. For D-T neutron source problem, the following two experiments are analyzed: (6) LLNL leakage spectrum measurements from spheres of iron and graphite, and (7) JAERI-FNS angular neutron spectrum measurements on beryllium and graphite slabs. Analyses have been performed using the radiation transport codes: ANISN(1D Sn), DIAC(1D Sn), DOT3.5(2D Sn) and MCNP(3D point Monte Carlo). The group cross sections for Sn transport calculations are generated with the code systems PROF-GROUCH-G/B and RADHEAT-V4. The point-wise cross sections for MCNP are produced with NJOY. For comparison, the analyses with JENDL-2 and ENDF/B-IV have been also carried out. The calculations using JENDL-3 show overall agreement with the experimental data as well as those with ENDF/B-IV. Particularly, JENDL-3 gives better results than JENDL-2 and ENDF/B-IV for sodium. It has been concluded that JENDL-3 is very applicable for fission and fusion reactor shielding analyses. (author)

  7. Self-adjoint extensions and spectral analysis in the generalized Kratzer problem

    Baldiotti, M C; Gitman, D M; Tyutin, I V; Voronov, B L

    2011-01-01

    We present a mathematically rigorous quantum-mechanical treatment of a one-dimensional non-relativistic motion of a particle in the potential field V(x)=g 1 x -1 +g 2 x -2 , x is an element of R + = [0, ∞). For g 2 >0 and g 1 K (x) and is usually used to describe molecular energy and structure, interactions between different molecules and interactions between non-bonded atoms. We construct all self-adjoint Schroedinger operators with the potential V(x) and represent rigorous solutions of the corresponding spectral problems. Solving the first part of the problem, we use a method of specifying self-adjoint extensions by (asymptotic) self-adjoint boundary conditions. Solving spectral problems, we follow Krein's method of guiding functionals. This work is a continuation of our previous works devoted to the Coulomb, Calogero and Aharonov-Bohm potentials.

  8. Formulation of coarse mesh finite difference to calculate mathematical adjoint flux

    Pereira, Valmir; Martinez, Aquilino Senra; Silva, Fernando Carvalho da

    2002-01-01

    The objective of this work is the obtention of the mathematical adjoint flux, having as its support the nodal expansion method (NEM) for coarse mesh problems. Since there are difficulties to evaluate this flux by using NEM. directly, a coarse mesh finite difference program was developed to obtain this adjoint flux. The coarse mesh finite difference formulation (DFMG) adopted uses results of the direct calculation (node average flux and node face averaged currents) obtained by NEM. These quantities (flux and currents) are used to obtain the correction factors which modify the classical finite differences formulation . Since the DFMG formulation is also capable of calculating the direct flux it was also tested to obtain this flux and it was verified that it was able to reproduce with good accuracy both the flux and the currents obtained via NEM. In this way, only matrix transposition is needed to calculate the mathematical adjoint flux. (author)

  9. An Adjoint-Based Approach to Study a Flexible Flapping Wing in Pitching-Rolling Motion

    Jia, Kun; Wei, Mingjun; Xu, Min; Li, Chengyu; Dong, Haibo

    2017-11-01

    Flapping-wing aerodynamics, with advantages in agility, efficiency, and hovering capability, has been the choice of many flyers in nature. However, the study of bio-inspired flapping-wing propulsion is often hindered by the problem's large control space with different wing kinematics and deformation. The adjoint-based approach reduces largely the computational cost to a feasible level by solving an inverse problem. Facing the complication from moving boundaries, non-cylindrical calculus provides an easy extension of traditional adjoint-based approach to handle the optimization involving moving boundaries. The improved adjoint method with non-cylindrical calculus for boundary treatment is first applied on a rigid pitching-rolling plate, then extended to a flexible one with active deformation to further increase its propulsion efficiency. The comparison of flow dynamics with the initial and optimal kinematics and deformation provides a unique opportunity to understand the flapping-wing mechanism. Supported by AFOSR and ARL.

  10. An exact and consistent adjoint method for high-fidelity discretization of the compressible flow equations

    Subramanian, Ramanathan Vishnampet Ganapathi

    Methods and computing hardware advances have enabled accurate predictions of complex compressible turbulence phenomena, such as the generation of jet noise that motivates the present effort. However, limited understanding of underlying physical mechanisms restricts the utility of such predictions since they do not, by themselves, indicate a route to design improvement. Gradient-based optimization using adjoints can circumvent the flow complexity to guide designs. Such methods have enabled sensitivity analysis and active control of turbulence at engineering flow conditions by providing gradient information at computational cost comparable to that of simulating the flow. They accelerate convergence of numerical design optimization algorithms, though this is predicated on the availability of an accurate gradient of the discretized flow equations. This is challenging to obtain, since both the chaotic character of the turbulence and the typical use of discretizations near their resolution limits in order to efficiently represent its smaller scales will amplify any approximation errors made in the adjoint formulation. Formulating a practical exact adjoint that avoids such errors is especially challenging if it is to be compatible with state-of-the-art simulation methods used for the turbulent flow itself. Automatic differentiation (AD) can provide code to calculate a nominally exact adjoint, but existing general-purpose AD codes are inefficient to the point of being prohibitive for large-scale turbulence simulations. We analyze the compressible flow equations as discretized using the same high-order workhorse methods used for many high-fidelity compressible turbulence simulations, and formulate a practical space--time discrete-adjoint method without changing the basic discretization. A key step is the definition of a particular discrete analog of the continuous norm that defines our cost functional; our selection leads directly to an efficient Runge--Kutta-like scheme

  11. Adjoint Based A Posteriori Analysis of Multiscale Mortar Discretizations with Multinumerics

    Tavener, Simon

    2013-01-01

    In this paper we derive a posteriori error estimates for linear functionals of the solution to an elliptic problem discretized using a multiscale nonoverlapping domain decomposition method. The error estimates are based on the solution of an appropriately defined adjoint problem. We present a general framework that allows us to consider both primal and mixed formulations of the forward and adjoint problems within each subdomain. The primal subdomains are discretized using either an interior penalty discontinuous Galerkin method or a continuous Galerkin method with weakly imposed Dirichlet conditions. The mixed subdomains are discretized using Raviart- Thomas mixed finite elements. The a posteriori error estimate also accounts for the errors due to adjoint-inconsistent subdomain discretizations. The coupling between the subdomain discretizations is achieved via a mortar space. We show that the numerical discretization error can be broken down into subdomain and mortar components which may be used to drive adaptive refinement.Copyright © by SIAM.

  12. Intergenerational Correlation in Monte Carlo k-Eigenvalue Calculation

    Ueki, Taro

    2002-01-01

    This paper investigates intergenerational correlation in the Monte Carlo k-eigenvalue calculation of a neutron effective multiplicative factor. To this end, the exponential transform for path stretching has been applied to large fissionable media with localized highly multiplying regions because in such media an exponentially decaying shape is a rough representation of the importance of source particles. The numerical results show that the difference between real and apparent variances virtually vanishes for an appropriate value of the exponential transform parameter. This indicates that the intergenerational correlation of k-eigenvalue samples could be eliminated by the adjoint biasing of particle transport. The relation between the biasing of particle transport and the intergenerational correlation is therefore investigated in the framework of collision estimators, and the following conclusion has been obtained: Within the leading order approximation with respect to the number of histories per generation, the intergenerational correlation vanishes when immediate importance is constant, and the immediate importance under simulation can be made constant by the biasing of particle transport with a function adjoint to the source neutron's distribution, i.e., the importance over all future generations

  13. New Toroid shielding design

    Hedberg V

    On the 15th of June 2001 the EB approved a new conceptual design for the toroid shield. In the old design, shown in the left part of the figure above, the moderator part of the shielding (JTV) was situated both in the warm and cold areas of the forward toroid. It consisted both of rings of polyethylene and hundreds of blocks of polyethylene (or an epoxy resin) inside the toroid vacuum vessel. In the new design, shown to the right in the figure above, only the rings remain inside the toroid. To compensate for the loss of moderator in the toroid, the copper plug (JTT) has been reduced in radius so that a layer of borated polyethylene can be placed around it (see figure below). The new design gives significant cost-savings and is easier to produce in the tight time schedule of the forward toroid. Since the amount of copper is reduced the weight that has to be carried by the toroid is also reduced. Outgassing into the toroid vacuum was a potential problem in the old design and this is now avoided. The main ...

  14. SHIELDS Final Technical Report

    Jordanova, Vania Koleva [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-10-03

    Predicting variations in the near-Earth space environment that can lead to spacecraft damage and failure, i.e. “space weather”, remains a big space physics challenge. A new capability was developed at Los Alamos National Laboratory (LANL) to understand, model, and predict Space Hazards Induced near Earth by Large Dynamic Storms, the SHIELDS framework. This framework simulates the dynamics of the Surface Charging Environment (SCE), the hot (keV) electrons representing the source and seed populations for the radiation belts, on both macro- and micro-scale. In addition to using physics-based models (like RAM-SCB, BATS-R-US, and iPIC3D), new data assimilation techniques employing data from LANL instruments on the Van Allen Probes and geosynchronous satellites were developed. An order of magnitude improvement in the accuracy in the simulation of the spacecraft surface charging environment was thus obtained. SHIELDS also includes a post-processing tool designed to calculate the surface charging for specific spacecraft geometry using the Curvilinear Particle-In-Cell (CPIC) code and to evaluate anomalies' relation to SCE dynamics. Such diagnostics is critically important when performing forensic analyses of space-system failures.

  15. A demonstration of adjoint methods for multi-dimensional remote sensing of the atmosphere and surface

    Martin, William G. K.; Hasekamp, Otto P.

    2018-01-01

    In previous work, we derived the adjoint method as a computationally efficient path to three-dimensional (3D) retrievals of clouds and aerosols. In this paper we will demonstrate the use of adjoint methods for retrieving two-dimensional (2D) fields of cloud extinction. The demonstration uses a new 2D radiative transfer solver (FSDOM). This radiation code was augmented with adjoint methods to allow efficient derivative calculations needed to retrieve cloud and surface properties from multi-angle reflectance measurements. The code was then used in three synthetic retrieval studies. Our retrieval algorithm adjusts the cloud extinction field and surface albedo to minimize the measurement misfit function with a gradient-based, quasi-Newton approach. At each step we compute the value of the misfit function and its gradient with two calls to the solver FSDOM. First we solve the forward radiative transfer equation to compute the residual misfit with measurements, and second we solve the adjoint radiative transfer equation to compute the gradient of the misfit function with respect to all unknowns. The synthetic retrieval studies verify that adjoint methods are scalable to retrieval problems with many measurements and unknowns. We can retrieve the vertically-integrated optical depth of moderately thick clouds as a function of the horizontal coordinate. It is also possible to retrieve the vertical profile of clouds that are separated by clear regions. The vertical profile retrievals improve for smaller cloud fractions. This leads to the conclusion that cloud edges actually increase the amount of information that is available for retrieving the vertical profile of clouds. However, to exploit this information one must retrieve the horizontally heterogeneous cloud properties with a 2D (or 3D) model. This prototype shows that adjoint methods can efficiently compute the gradient of the misfit function. This work paves the way for the application of similar methods to 3D remote

  16. Usefulness of the Monte Carlo method in reliability calculations

    Lanore, J.M.; Kalli, H.

    1977-01-01

    Three examples of reliability Monte Carlo programs developed in the LEP (Laboratory for Radiation Shielding Studies in the Nuclear Research Center at Saclay) are presented. First, an uncertainty analysis is given for a simplified spray system; a Monte Carlo program PATREC-MC has been written to solve the problem with the system components given in the fault tree representation. The second program MONARC 2 has been written to solve the problem of complex systems reliability by the Monte Carlo simulation, here again the system (a residual heat removal system) is in the fault tree representation. Third, the Monte Carlo program MONARC was used instead of the Markov diagram to solve the simulation problem of an electric power supply including two nets and two stand-by diesels

  17. BPS Center Vortices in Nonrelativistic SU(N) Gauge Models with Adjoint Higgs Fields

    Oxman, L. E.

    2015-01-01

    We propose a class of SU(N) Yang-Mills models, with adjoint Higgs fields, that accept BPS center vortex equations. The lack of a local magnetic flux that could serve as an energy bound is circumvented by including a new term in the energy functional. This term tends to align, in the Lie algebra, the magnetic field and one of the adjoint Higgs fields. Finally, a reduced set of equations for the center vortex profile functions is obtained (for N=2,3). In particular, Z(3) BPS vortices come in three colours and three anticolours, obtained from an ansatz based on the defining representation and its conjugate.

  18. Adjoint shape optimization for fluid-structure interaction of ducted flows

    Heners, J. P.; Radtke, L.; Hinze, M.; Düster, A.

    2018-03-01

    Based on the coupled problem of time-dependent fluid-structure interaction, equations for an appropriate adjoint problem are derived by the consequent use of the formal Lagrange calculus. Solutions of both primal and adjoint equations are computed in a partitioned fashion and enable the formulation of a surface sensitivity. This sensitivity is used in the context of a steepest descent algorithm for the computation of the required gradient of an appropriate cost functional. The efficiency of the developed optimization approach is demonstrated by minimization of the pressure drop in a simple two-dimensional channel flow and in a three-dimensional ducted flow surrounded by a thin-walled structure.

  19. Adjoint sensitivity studies of loop current and eddy shedding in the Gulf of Mexico

    Gopalakrishnan, Ganesh; Cornuelle, Bruce D.; Hoteit, Ibrahim

    2013-01-01

    Adjoint model sensitivity analyses were applied for the loop current (LC) and its eddy shedding in the Gulf of Mexico (GoM) using the MIT general circulation model (MITgcm). The circulation in the GoM is mainly driven by the energetic LC and subsequent LC eddy separation. In order to understand which ocean regions and features control the evolution of the LC, including anticyclonic warm-core eddy shedding in the GoM, forward and adjoint sensitivities with respect to previous model state and atmospheric forcing were computed using the MITgcm and its adjoint. Since the validity of the adjoint model sensitivities depends on the capability of the forward model to simulate the real LC system and the eddy shedding processes, a 5 year (2004–2008) forward model simulation was performed for the GoM using realistic atmospheric forcing, initial, and boundary conditions. This forward model simulation was compared to satellite measurements of sea-surface height (SSH) and sea-surface temperature (SST), and observed transport variability. Despite realistic mean state, standard deviations, and LC eddy shedding period, the simulated LC extension shows less variability and more regularity than the observations. However, the model is suitable for studying the LC system and can be utilized for examining the ocean influences leading to a simple, and hopefully generic LC eddy separation in the GoM. The adjoint sensitivities of the LC show influences from the Yucatan Channel (YC) flow and Loop Current Frontal Eddy (LCFE) on both LC extension and eddy separation, as suggested by earlier work. Some of the processes that control LC extension after eddy separation differ from those controlling eddy shedding, but include YC through-flow. The sensitivity remains stable for more than 30 days and moves generally upstream, entering the Caribbean Sea. The sensitivities of the LC for SST generally remain closer to the surface and move at speeds consistent with advection by the high-speed core of

  20. On the use of flux-adjoint condensed nuclear data for 1-group AGR kinetics

    Hutt, P.K.

    1979-03-01

    Following previous work on the differences between one and two neutron group AGR kinetics the possible advantages of flux-adjoint condensed lattice data over the simple flux condensation procedure are investigated. Analytic arguments are given for expecting flux-adjoint condensation to give a better representation of rod worth slopes and flux shape changes associated with partially rodded cores. These areas have previously been found to yield most of the one to two neutron group differences. The validity of these arguments is demonstrated comparing various calculations. (U.K.)

  1. Adjoint sensitivity studies of loop current and eddy shedding in the Gulf of Mexico

    Gopalakrishnan, Ganesh

    2013-07-01

    Adjoint model sensitivity analyses were applied for the loop current (LC) and its eddy shedding in the Gulf of Mexico (GoM) using the MIT general circulation model (MITgcm). The circulation in the GoM is mainly driven by the energetic LC and subsequent LC eddy separation. In order to understand which ocean regions and features control the evolution of the LC, including anticyclonic warm-core eddy shedding in the GoM, forward and adjoint sensitivities with respect to previous model state and atmospheric forcing were computed using the MITgcm and its adjoint. Since the validity of the adjoint model sensitivities depends on the capability of the forward model to simulate the real LC system and the eddy shedding processes, a 5 year (2004–2008) forward model simulation was performed for the GoM using realistic atmospheric forcing, initial, and boundary conditions. This forward model simulation was compared to satellite measurements of sea-surface height (SSH) and sea-surface temperature (SST), and observed transport variability. Despite realistic mean state, standard deviations, and LC eddy shedding period, the simulated LC extension shows less variability and more regularity than the observations. However, the model is suitable for studying the LC system and can be utilized for examining the ocean influences leading to a simple, and hopefully generic LC eddy separation in the GoM. The adjoint sensitivities of the LC show influences from the Yucatan Channel (YC) flow and Loop Current Frontal Eddy (LCFE) on both LC extension and eddy separation, as suggested by earlier work. Some of the processes that control LC extension after eddy separation differ from those controlling eddy shedding, but include YC through-flow. The sensitivity remains stable for more than 30 days and moves generally upstream, entering the Caribbean Sea. The sensitivities of the LC for SST generally remain closer to the surface and move at speeds consistent with advection by the high-speed core of

  2. Ambient noise adjoint tomography for a linear array in North China

    Zhang, C.; Yao, H.; Liu, Q.; Yuan, Y. O.; Zhang, P.; Feng, J.; Fang, L.

    2017-12-01

    Ambient noise tomography based on dispersion data and ray theory has been widely utilized for imaging crustal structures. In order to improve the inversion accuracy, ambient noise tomography based on the 3D adjoint approach or full waveform inversion has been developed recently, however, the computational cost is tremendous. In this study we present 2D ambient noise adjoint tomography for a linear array in north China with significant computational efficiency compared to 3D ambient noise adjoint tomography. During the preprocessing, we first convert the observed data in 3D media, i.e., surface-wave empirical Green's functions (EGFs) from ambient noise cross-correlation, to the reconstructed EGFs in 2D media using a 3D/2D transformation scheme. Different from the conventional steps of measuring phase dispersion, the 2D adjoint tomography refines 2D shear wave speeds along the profile directly from the reconstructed Rayleigh wave EGFs in the period band 6-35s. With the 2D initial model extracted from the 3D model from traditional ambient noise tomography, adjoint tomography updates the model by minimizing the frequency-dependent Rayleigh wave traveltime misfits between the reconstructed EGFs and synthetic Green function (SGFs) in 2D media generated by the spectral-element method (SEM), with a preconditioned conjugate gradient method. The multitaper traveltime difference measurement is applied in four period bands during the inversion: 20-35s, 15-30s, 10-20s and 6-15s. The recovered model shows more detailed crustal structures with pronounced low velocity anomaly in the mid-lower crust beneath the junction of Taihang Mountains and Yin-Yan Mountains compared with the initial model. This low velocity structure may imply the possible intense crust-mantle interactions, probably associated with the magmatic underplating during the Mesozoic to Cenozoic evolution of the region. To our knowledge, it's first time that ambient noise adjoint tomography is implemented in 2D media

  3. Mass anomalous dimension of Adjoint QCD at large N from twisted volume reduction

    Pérez, Margarita García; Keegan, Liam; Okawa, Masanori

    2015-01-01

    In this work we consider the $SU(N)$ gauge theory with two Dirac fermions in the adjoint representation, in the limit of large $N$. In this limit the infinite-volume physics of this model can be studied by means of the corresponding twisted reduced model defined on a single site lattice. Making use of this strategy we study the reduced model for various values of $N$ up to 289. By analyzing the eigenvalue distribution of the adjoint Dirac operator we test the conformality of the theory and extract the corresponding mass anomalous dimension.

  4. Mass anomalous dimension of adjoint QCD at large N from twisted volume reduction

    Pérez, Margarita García [Instituto de Física Teórica UAM-CSIC, Nicolás Cabrera 13-15, Universidad Autónoma de Madrid,E-28049-Madrid (Spain); González-Arroyo, Antonio [Instituto de Física Teórica UAM-CSIC, Nicolás Cabrera 13-15, Universidad Autónoma de Madrid,E-28049-Madrid (Spain); Departamento de Física Teórica, C-XI, Universidad Autónoma de Madrid,E-28049-Madrid (Spain); Keegan, Liam [PH-TH, CERN,CH-1211 Geneva 23 (Switzerland); Okawa, Masanori [Graduate School of Science, Hiroshima University,Higashi-Hiroshima, Hiroshima 739-8526 (Japan); Core of Research for the Energetic Universe, Hiroshima University,Higashi-Hiroshima, Hiroshima 739-8526 (Japan)

    2015-08-07

    In this work we consider the SU(N) gauge theory with two Dirac fermions in the adjoint representation, in the limit of large N. In this limit the infinite-volume physics of this model can be studied by means of the corresponding twisted reduced model defined on a single site lattice. Making use of this strategy we study the reduced model for various values of N up to 289. By analyzing the eigenvalue distribution of the adjoint Dirac operator we test the conformality of the theory and extract the corresponding mass anomalous dimension.

  5. Application of adjoint sensitivity theory to performance assessment of hydrogeologic concerns

    Metcalfe, D.E.; Harper, W.V.

    1986-01-01

    Sensitivity and uncertainty analyses are important components of performance assessment activities for potential high-level radioactive waste repositories. The application of the adjoint sensitivity technique is demonstrated for the Leadville Limestone in the Paradox Basin, Utah. The adjoint technique is used sequentially to first assist in the calibration of the regional conceptual ground-water flow model to measured potentiometric data. Second, it is used to evaluate the sensitivities of the calculated pressures used to define local scale boundary conditions to regional parameters and boundary conditions

  6. Adaptive mesh refinement and adjoint methods in geophysics simulations

    Burstedde, Carsten

    2013-04-01

    required by human intervention and analysis. Specifying an objective functional that quantifies the misfit between the simulation outcome and known constraints and then minimizing it through numerical optimization can serve as an automated technique for parameter identification. As suggested by the similarity in formulation, the numerical algorithm is closely related to the one used for goal-oriented error estimation. One common point is that the so-called adjoint equation needs to be solved numerically. We will outline the derivation and implementation of these methods and discuss some of their pros and cons, supported by numerical results.

  7. Exploring Monte Carlo methods

    Dunn, William L

    2012-01-01

    Exploring Monte Carlo Methods is a basic text that describes the numerical methods that have come to be known as "Monte Carlo." The book treats the subject generically through the first eight chapters and, thus, should be of use to anyone who wants to learn to use Monte Carlo. The next two chapters focus on applications in nuclear engineering, which are illustrative of uses in other fields. Five appendices are included, which provide useful information on probability distributions, general-purpose Monte Carlo codes for radiation transport, and other matters. The famous "Buffon's needle proble

  8. Monte Carlo methods

    Bardenet Rémi

    2013-07-01

    Full Text Available Bayesian inference often requires integrating some function with respect to a posterior distribution. Monte Carlo methods are sampling algorithms that allow to compute these integrals numerically when they are not analytically tractable. We review here the basic principles and the most common Monte Carlo algorithms, among which rejection sampling, importance sampling and Monte Carlo Markov chain (MCMC methods. We give intuition on the theoretical justification of the algorithms as well as practical advice, trying to relate both. We discuss the application of Monte Carlo in experimental physics, and point to landmarks in the literature for the curious reader.

  9. Basic design of shield blocks for a spallation neutron source under the high-intensity proton accelerator project

    Yoshida, Katsuhiko; Maekawa, Fujio; Takada, Hiroshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2003-03-01

    Under the JAERI-KEK High-Intensity Proton Accelerator Project (J-PARC), a spallation neutron source driven by a 3 GeV-1 MW proton beam is planed to be constructed as a main part of the Materials and Life Science Facility. Overall dimensions of a biological shield of the neutron source had been determined by evaluation of shielding performance by Monte Carlo calculations. This report describes results of design studies on an optimum dividing scheme in terms of cost and treatment and mechanical strength of shield blocks for the biological shield. As for mechanical strength, it was studied whether the shield blocks would be stable, fall down or move to a horizontal direction in case of an earthquake of seismic intensity of 5.5 (250 Gal) as an abnormal load. For ceiling shielding blocks being supported by both ends of the long blocks, maximum bending moment and an amount of maximum deflection of their center were evaluated. (author)

  10. Basic design of shield blocks for a spallation neutron source under the high-intensity proton accelerator project

    Yoshida, K; Takada, H

    2003-01-01

    Under the JAERI-KEK High-Intensity Proton Accelerator Project (J-PARC), a spallation neutron source driven by a 3 GeV-1 MW proton beam is planed to be constructed as a main part of the Materials and Life Science Facility. Overall dimensions of a biological shield of the neutron source had been determined by evaluation of shielding performance by Monte Carlo calculations. This report describes results of design studies on an optimum dividing scheme in terms of cost and treatment and mechanical strength of shield blocks for the biological shield. As for mechanical strength, it was studied whether the shield blocks would be stable, fall down or move to a horizontal direction in case of an earthquake of seismic intensity of 5.5 (250 Gal) as an abnormal load. For ceiling shielding blocks being supported by both ends of the long blocks, maximum bending moment and an amount of maximum deflection of their center were evaluated.

  11. Radiation shield for PWR reactors

    Esenov, Amra; Pustovgar, Andrey

    2013-01-01

    One of the chief structures of a reactor pit is a 'dry' shield. Setting up a 'dry' shield includes the technologically complex process of thermal processing of serpentinite concrete. Modern advances in the area of materials technology permit avoiding this complex and demanding procedure, and this significantly decreases the duration, labor intensity, and cost of setting it up. (orig.)

  12. Nuclear data for radiation shielding

    Miyasaka, Shunichi; Takahashi, Hiroshi.

    1976-01-01

    The third shielding expert conference was convened in Paris in Oct. 1975 for exchanging informations about the sensitivity evaluation of nuclear data in shielding calculation and integral bench mark experiment. The requirements about nuclear data presented at present from the field of nuclear design do not reflect sufficiently the requirements of shielding design, therefore it was the object to gather the requirements about nuclear data from the field of shielding. The nuclides used for shielding are numerous, and the nuclear data on these isotopes are required. Some of them cannot be ignored as the source of secondary γ-ray or in view of the radioactivation of materials. The requirements for the nuclear data of neutrons in the field of shielding are those concerning the reaction cross sections producing secondary γ-ray, the reaction cross sections including the production of secondary neutrons, elastic scattering cross sections, and total cross sections. The topics in the Paris conference about neutron shielding data are described, such as the methodology of sensitivity evaluation, the standardization of group constant libraries, the bench mark experiment on iron and sodium, and the cross section of γ-ray production. In the shielding of nuclear fission reactors, the γ-ray production owing to nuclear fission reaction is also important. In (d, t) fusion reactors, high energy neutrons are generated, and high energy γ-ray is emitted through giant E1 resonance. (Kako, I.)

  13. Concrete shielding exterior to iron

    Yurista, P.; Cossairt, D.

    1983-08-01

    A rule of thumb at Fermilab has been to use 3 feet of concrete exterior to iron shielding. A recent design of a shield with a severe dimensional constraint has prompted a re-evaluation of this rule of thumb and has led to the following calculations of the concrete thickness required to nullify this problem. 4 references, 4 figures

  14. Gonad shielding in diagnostic radiology

    1975-06-01

    The use of gonad shielding is an important radiation protection technique, intended to reduce unnecessary x-ray exposure of the gonads of patients from diagnostic x-ray procedures. This pamphlet will provide physicians and radiologic technologists with information which will aid their appropriate use of gonad shielding

  15. Proposal of a Technical Design of the ATLAS Forward Region Shielding

    Leroy, C; Palla, J; Pospísil, S; Puchmajer, P; Sodomka, J; Stekl, I

    2002-01-01

    The aim of this note is to present a technical design of the ATLAS forward region. The concept is based on segmented shielding supported by the results of an experiment performed at CERN - PS and Monte Carlo simulations extending these results to ATLAS situation. This concept is translated into a practical engineering design.

  16. A detailed investigation of interactions within the shielding to HPGe detector response using MCNP code

    Thanh, Tran Thien; Tao, Chau Van; Loan, Truong Thi Hong; Nhon, Mai Van; Chuong, Huynh Dinh; Au, Bui Hai [Vietnam National Univ., Ho Chi Minh City (Viet Nam). Dept. of Nuclear Physics

    2012-12-15

    The accuracy of the coincidence-summing corrections in gamma spectrometry depends on the total efficiency calibration that is hardly obtained over the whole energy as the required experimental conditions are not easily attained. Monte Carlo simulations using MCNP5 code was performed in order to estimate the affect of the shielding to total efficiency. The effect of HPGe response are also shown. (orig.)

  17. Neutron and gamma ray transport calculations in shielding system

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

    1998-03-01

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

  18. The Imperial Shield

    Mortensen, Simon Valentin

    2006-01-01

      The title of this Ph.d. dissertation is "The Imperial Shield: Imperial Overstretch, Assured Destruction, and the ban on nationwide ABM-defense with particular emphasis on the Johnson and the Nixon Administration". The dissertation set out to explain the origins of the ABM Treaty's central meaning....... Domestic spending continued to increase by more in real terms than the GDP, and the Democratically controlled Congress also made some very expensive modifications in Nixon tax bill in the fall of 1969, once again plunging the budget into the red.The economic crisis was partly caused by, and partly...... the Administration debated the deployment of new ABM-sites in early 1970, Kissinger could not prevail against these forces, but had to settle for a compromise, which he regarded as less than a definite commitment to nationwide ABM-defense.The political developments were of even greater importance. A strong link has...

  19. Shielded Canister Transporter

    Eidem, G.G. Jr.; Fages, R.

    1993-01-01

    The Hanford Waste Vitrification Plant (HWVP) will produce canisters filled with high-level radioactive waste immobilized in borosilicate glass. This report discusses a Shielded Canister Transporter (SCT) which will provide the means for safe transportation and handling of the canisters from the Vitrification Building to the Canister Storage Building (CSB). The stainless steel canisters are 0.61 meters in diameter, 3.0 meters tall, and weigh approximately 2,135 kilograms, with a maximum exterior surface dose rate of 90,000 R/hr. The canisters are placed into storage tubes to a maximum of three tall (two for overpack canisters) with an impact limiter placed at the tube bottom and between each canister. A floor plug seals the top of the storage tube at the operating floor level of the CSB

  20. ITER shielding blanket

    Strebkov, Yu [ENTEK, Moscow (Russian Federation); Avsjannikov, A [ENTEK, Moscow (Russian Federation); Baryshev, M [NIAT, Moscow (Russian Federation); Blinov, Yu [ENTEK, Moscow (Russian Federation); Shatalov, G [KIAE, Moscow (Russian Federation); Vasiliev, N [KIAE, Moscow (Russian Federation); Vinnikov, A [ENTEK, Moscow (Russian Federation); Chernjagin, A [DYNAMICA, Moscow (Russian Federation)

    1995-03-01

    A reference non-breeding blanket is under development now for the ITER Basic Performance Phase for the purpose of high reliability during the first stage of ITER operation. More severe operation modes are expected in this stage with first wall (FW) local heat loads up to 100-300Wcm{sup -2}. Integration of a blanket design with protective and start limiters requires new solutions to achieve high reliability, and possible use of beryllium as a protective material leads to technologies. The rigid shielding blanket concept was developed in Russia to satisfy the above-mentioned requirements. The concept is based on a copper alloy FW, austenitic stainless steel blanket structure, water cooling. Beryllium protection is integrated in the FW design. Fabrication technology and assembly procedure are described in parallel with the equipment used. (orig.).