Electromagnetic Chirps from Neutron Star–Black Hole Mergers

NASA Astrophysics Data System (ADS)

Schnittman, Jeremy D.; Dal Canton, Tito; Camp, Jordan; Tsang, David; Kelly, Bernard J.

2018-02-01

We calculate the electromagnetic signal of a gamma-ray flare coming from the surface of a neutron star shortly before merger with a black hole companion. Using a new version of the Monte Carlo radiation transport code Pandurata that incorporates dynamic spacetimes, we integrate photon geodesics from the neutron star surface until they reach a distant observer or are captured by the black hole. The gamma-ray light curve is modulated by a number of relativistic effects, including Doppler beaming and gravitational lensing. Because the photons originate from the inspiraling neutron star, the light curve closely resembles the corresponding gravitational waveform: a chirp signal characterized by a steadily increasing frequency and amplitude. We propose to search for these electromagnetic chirps using matched filtering algorithms similar to those used in LIGO data analysis.

Electromagnetic Chirps from Neutron Star-Black Hole Mergers

NASA Technical Reports Server (NTRS)

Schnittman, Jeremy D.; Dal Canton, Tito; Camp, Jordan B.; Tsang, David; Kelly, Bernard J.

2018-01-01

We calculate the electromagnetic signal of a gamma-ray flare coming from the surface of a neutron star shortly before merger with a black hole companion. Using a new version of the Monte Carlo radiation transport code Pandurata that incorporates dynamic spacetimes, we integrate photon geodesics from the neutron star surface until they reach a distant observer or are captured by the black hole. The gamma-ray light curve is modulated by a number of relativistic effects, including Doppler beaming and gravitational lensing. Because the photons originate from the inspiraling neutron star, the light curve closely resembles the corresponding gravitational waveform: a chirp signal characterized by a steadily increasing frequency and amplitude. We propose to search for these electromagnetic chirps using matched filtering algorithms similar to those used in LIGO data analysis.

Reactor Dosimetry Applications Using RAPTOR-M3G:. a New Parallel 3-D Radiation Transport Code

NASA Astrophysics Data System (ADS)

Longoni, Gianluca; Anderson, Stanwood L.

2009-08-01

The numerical solution of the Linearized Boltzmann Equation (LBE) via the Discrete Ordinates method (SN) requires extensive computational resources for large 3-D neutron and gamma transport applications due to the concurrent discretization of the angular, spatial, and energy domains. This paper will discuss the development RAPTOR-M3G (RApid Parallel Transport Of Radiation - Multiple 3D Geometries), a new 3-D parallel radiation transport code, and its application to the calculation of ex-vessel neutron dosimetry responses in the cavity of a commercial 2-loop Pressurized Water Reactor (PWR). RAPTOR-M3G is based domain decomposition algorithms, where the spatial and angular domains are allocated and processed on multi-processor computer architectures. As compared to traditional single-processor applications, this approach reduces the computational load as well as the memory requirement per processor, yielding an efficient solution methodology for large 3-D problems. Measured neutron dosimetry responses in the reactor cavity air gap will be compared to the RAPTOR-M3G predictions. This paper is organized as follows: Section 1 discusses the RAPTOR-M3G methodology; Section 2 describes the 2-loop PWR model and the numerical results obtained. Section 3 addresses the parallel performance of the code, and Section 4 concludes this paper with final remarks and future work.

Monte Carlo closure for moment-based transport schemes in general relativistic radiation hydrodynamic simulations

NASA Astrophysics Data System (ADS)

Foucart, Francois

2018-04-01

General relativistic radiation hydrodynamic simulations are necessary to accurately model a number of astrophysical systems involving black holes and neutron stars. Photon transport plays a crucial role in radiatively dominated accretion discs, while neutrino transport is critical to core-collapse supernovae and to the modelling of electromagnetic transients and nucleosynthesis in neutron star mergers. However, evolving the full Boltzmann equations of radiative transport is extremely expensive. Here, we describe the implementation in the general relativistic SPEC code of a cheaper radiation hydrodynamic method that theoretically converges to a solution of Boltzmann's equation in the limit of infinite numerical resources. The algorithm is based on a grey two-moment scheme, in which we evolve the energy density and momentum density of the radiation. Two-moment schemes require a closure that fills in missing information about the energy spectrum and higher order moments of the radiation. Instead of the approximate analytical closure currently used in core-collapse and merger simulations, we complement the two-moment scheme with a low-accuracy Monte Carlo evolution. The Monte Carlo results can provide any or all of the missing information in the evolution of the moments, as desired by the user. As a first test of our methods, we study a set of idealized problems demonstrating that our algorithm performs significantly better than existing analytical closures. We also discuss the current limitations of our method, in particular open questions regarding the stability of the fully coupled scheme.

An abstract approach to evaporation models in rarefied gas dynamics

NASA Astrophysics Data System (ADS)

Greenberg, W.; van der Mee, C. V. M.

1984-03-01

Strong evaporation models involving 1D stationary problems with linear self-adjoint collision operators and solutions in abstract Hilbert spaces are investigated analytically. An efficient algorithm for locating the transition from existence to nonexistence of solutions is developed and applied to the 1D and 3D BGK model equations and the 3D BGK model in moment form, demonstrating the nonexistence of stationary evaporation states with supersonic drift velocities. Applications to similar models in electron and phonon transport, radiative transfer, and neutron transport are suggested.

Parareal in time 3D numerical solver for the LWR Benchmark neutron diffusion transient model

DOE Office of Scientific and Technical Information (OSTI.GOV)

Baudron, Anne-Marie, E-mail: anne-marie.baudron@cea.fr; CEA-DRN/DMT/SERMA, CEN-Saclay, 91191 Gif sur Yvette Cedex; Lautard, Jean-Jacques, E-mail: jean-jacques.lautard@cea.fr

2014-12-15

In this paper we present a time-parallel algorithm for the 3D neutrons calculation of a transient model in a nuclear reactor core. The neutrons calculation consists in numerically solving the time dependent diffusion approximation equation, which is a simplified transport equation. The numerical resolution is done with finite elements method based on a tetrahedral meshing of the computational domain, representing the reactor core, and time discretization is achieved using a θ-scheme. The transient model presents moving control rods during the time of the reaction. Therefore, cross-sections (piecewise constants) are taken into account by interpolations with respect to the velocity ofmore » the control rods. The parallelism across the time is achieved by an adequate use of the parareal in time algorithm to the handled problem. This parallel method is a predictor corrector scheme that iteratively combines the use of two kinds of numerical propagators, one coarse and one fine. Our method is made efficient by means of a coarse solver defined with large time step and fixed position control rods model, while the fine propagator is assumed to be a high order numerical approximation of the full model. The parallel implementation of our method provides a good scalability of the algorithm. Numerical results show the efficiency of the parareal method on large light water reactor transient model corresponding to the Langenbuch–Maurer–Werner benchmark.« less

An improved target velocity sampling algorithm for free gas elastic scattering

DOE Office of Scientific and Technical Information (OSTI.GOV)

Romano, Paul K.; Walsh, Jonathan A.

We present an improved algorithm for sampling the target velocity when simulating elastic scattering in a Monte Carlo neutron transport code that correctly accounts for the energy dependence of the scattering cross section. The algorithm samples the relative velocity directly, thereby avoiding a potentially inefficient rejection step based on the ratio of cross sections. Here, we have shown that this algorithm requires only one rejection step, whereas other methods of similar accuracy require two rejection steps. The method was verified against stochastic and deterministic reference results for upscattering percentages in 238U. Simulations of a light water reactor pin cell problemmore » demonstrate that using this algorithm results in a 3% or less penalty in performance when compared with an approximate method that is used in most production Monte Carlo codes« less

An improved target velocity sampling algorithm for free gas elastic scattering

DOE PAGES

Romano, Paul K.; Walsh, Jonathan A.

2018-02-03

We present an improved algorithm for sampling the target velocity when simulating elastic scattering in a Monte Carlo neutron transport code that correctly accounts for the energy dependence of the scattering cross section. The algorithm samples the relative velocity directly, thereby avoiding a potentially inefficient rejection step based on the ratio of cross sections. Here, we have shown that this algorithm requires only one rejection step, whereas other methods of similar accuracy require two rejection steps. The method was verified against stochastic and deterministic reference results for upscattering percentages in 238U. Simulations of a light water reactor pin cell problemmore » demonstrate that using this algorithm results in a 3% or less penalty in performance when compared with an approximate method that is used in most production Monte Carlo codes« less

A 3DHZETRN Code in a Spherical Uniform Sphere with Monte Carlo Verification

NASA Technical Reports Server (NTRS)

Wilson, John W.; Slaba, Tony C.; Badavi, Francis F.; Reddell, Brandon D.; Bahadori, Amir A.

2014-01-01

The computationally efficient HZETRN code has been used in recent trade studies for lunar and Martian exploration and is currently being used in the engineering development of the next generation of space vehicles, habitats, and extra vehicular activity equipment. A new version (3DHZETRN) capable of transporting High charge (Z) and Energy (HZE) and light ions (including neutrons) under space-like boundary conditions with enhanced neutron and light ion propagation is under development. In the present report, new algorithms for light ion and neutron propagation with well-defined convergence criteria in 3D objects is developed and tested against Monte Carlo simulations to verify the solution methodology. The code will be available through the software system, OLTARIS, for shield design and validation and provides a basis for personal computer software capable of space shield analysis and optimization.

Parallel and Portable Monte Carlo Particle Transport

NASA Astrophysics Data System (ADS)

Lee, S. R.; Cummings, J. C.; Nolen, S. D.; Keen, N. D.

1997-08-01

We have developed a multi-group, Monte Carlo neutron transport code in C++ using object-oriented methods and the Parallel Object-Oriented Methods and Applications (POOMA) class library. This transport code, called MC++, currently computes k and α eigenvalues of the neutron transport equation on a rectilinear computational mesh. It is portable to and runs in parallel on a wide variety of platforms, including MPPs, clustered SMPs, and individual workstations. It contains appropriate classes and abstractions for particle transport and, through the use of POOMA, for portable parallelism. Current capabilities are discussed, along with physics and performance results for several test problems on a variety of hardware, including all three Accelerated Strategic Computing Initiative (ASCI) platforms. Current parallel performance indicates the ability to compute α-eigenvalues in seconds or minutes rather than days or weeks. Current and future work on the implementation of a general transport physics framework (TPF) is also described. This TPF employs modern C++ programming techniques to provide simplified user interfaces, generic STL-style programming, and compile-time performance optimization. Physics capabilities of the TPF will be extended to include continuous energy treatments, implicit Monte Carlo algorithms, and a variety of convergence acceleration techniques such as importance combing.

The effects of nuclear data library processing on Geant4 and MCNP simulations of the thermal neutron scattering law

NASA Astrophysics Data System (ADS)

Hartling, K.; Ciungu, B.; Li, G.; Bentoumi, G.; Sur, B.

2018-05-01

Monte Carlo codes such as MCNP and Geant4 rely on a combination of physics models and evaluated nuclear data files (ENDF) to simulate the transport of neutrons through various materials and geometries. The grid representation used to represent the final-state scattering energies and angles associated with neutron scattering interactions can significantly affect the predictions of these codes. In particular, the default thermal scattering libraries used by MCNP6.1 and Geant4.10.3 do not accurately reproduce the ENDF/B-VII.1 model in simulations of the double-differential cross section for thermal neutrons interacting with hydrogen nuclei in a thin layer of water. However, agreement between model and simulation can be achieved within the statistical error by re-processing ENDF/B-VII.I thermal scattering libraries with the NJOY code. The structure of the thermal scattering libraries and sampling algorithms in MCNP and Geant4 are also reviewed.

Development of MCNPX-ESUT computer code for simulation of neutron/gamma pulse height distribution

NASA Astrophysics Data System (ADS)

Abolfazl Hosseini, Seyed; Vosoughi, Naser; Zangian, Mehdi

2015-05-01

In this paper, the development of the MCNPX-ESUT (MCNPX-Energy Engineering of Sharif University of Technology) computer code for simulation of neutron/gamma pulse height distribution is reported. Since liquid organic scintillators like NE-213 are well suited and routinely used for spectrometry in mixed neutron/gamma fields, this type of detectors is selected for simulation in the present study. The proposed algorithm for simulation includes four main steps. The first step is the modeling of the neutron/gamma particle transport and their interactions with the materials in the environment and detector volume. In the second step, the number of scintillation photons due to charged particles such as electrons, alphas, protons and carbon nuclei in the scintillator material is calculated. In the third step, the transport of scintillation photons in the scintillator and lightguide is simulated. Finally, the resolution corresponding to the experiment is considered in the last step of the simulation. Unlike the similar computer codes like SCINFUL, NRESP7 and PHRESP, the developed computer code is applicable to both neutron and gamma sources. Hence, the discrimination of neutron and gamma in the mixed fields may be performed using the MCNPX-ESUT computer code. The main feature of MCNPX-ESUT computer code is that the neutron/gamma pulse height simulation may be performed without needing any sort of post processing. In the present study, the pulse height distributions due to a monoenergetic neutron/gamma source in NE-213 detector using MCNPX-ESUT computer code is simulated. The simulated neutron pulse height distributions are validated through comparing with experimental data (Gohil et al. Nuclear Instruments and Methods in Physics Research Section A: Accelerators, Spectrometers, Detectors and Associated Equipment, 664 (2012) 304-309.) and the results obtained from similar computer codes like SCINFUL, NRESP7 and Geant4. The simulated gamma pulse height distribution for a 137Cs source is also compared with the experimental data.

High Performance Radiation Transport Simulations on TITAN

DOE Office of Scientific and Technical Information (OSTI.GOV)

Baker, Christopher G; Davidson, Gregory G; Evans, Thomas M

2012-01-01

In this paper we describe the Denovo code system. Denovo solves the six-dimensional, steady-state, linear Boltzmann transport equation, of central importance to nuclear technology applications such as reactor core analysis (neutronics), radiation shielding, nuclear forensics and radiation detection. The code features multiple spatial differencing schemes, state-of-the-art linear solvers, the Koch-Baker-Alcouffe (KBA) parallel-wavefront sweep algorithm for inverting the transport operator, a new multilevel energy decomposition method scaling to hundreds of thousands of processing cores, and a modern, novel code architecture that supports straightforward integration of new features. In this paper we discuss the performance of Denovo on the 10--20 petaflop ORNLmore » GPU-based system, Titan. We describe algorithms and techniques used to exploit the capabilities of Titan's heterogeneous compute node architecture and the challenges of obtaining good parallel performance for this sparse hyperbolic PDE solver containing inherently sequential computations. Numerical results demonstrating Denovo performance on early Titan hardware are presented.« less

3D Space Radiation Transport in a Shielded ICRU Tissue Sphere

NASA Technical Reports Server (NTRS)

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.

Evaluation of a new neutron energy spectrum unfolding code based on an Adaptive Neuro-Fuzzy Inference System (ANFIS).

PubMed

Hosseini, Seyed Abolfazl; Esmaili Paeen Afrakoti, Iman

2018-01-17

The purpose of the present study was to reconstruct the energy spectrum of a poly-energetic neutron source using an algorithm developed based on an Adaptive Neuro-Fuzzy Inference System (ANFIS). ANFIS is a kind of artificial neural network based on the Takagi-Sugeno fuzzy inference system. The ANFIS algorithm uses the advantages of both fuzzy inference systems and artificial neural networks to improve the effectiveness of algorithms in various applications such as modeling, control and classification. The neutron pulse height distributions used as input data in the training procedure for the ANFIS algorithm were obtained from the simulations performed by MCNPX-ESUT computational code (MCNPX-Energy engineering of Sharif University of Technology). Taking into account the normalization condition of each energy spectrum, 4300 neutron energy spectra were generated randomly. (The value in each bin was generated randomly, and finally a normalization of each generated energy spectrum was performed). The randomly generated neutron energy spectra were considered as output data of the developed ANFIS computational code in the training step. To calculate the neutron energy spectrum using conventional methods, an inverse problem with an approximately singular response matrix (with the determinant of the matrix close to zero) should be solved. The solution of the inverse problem using the conventional methods unfold neutron energy spectrum with low accuracy. Application of the iterative algorithms in the solution of such a problem, or utilizing the intelligent algorithms (in which there is no need to solve the problem), is usually preferred for unfolding of the energy spectrum. Therefore, the main reason for development of intelligent algorithms like ANFIS for unfolding of neutron energy spectra is to avoid solving the inverse problem. In the present study, the unfolded neutron energy spectra of 252Cf and 241Am-9Be neutron sources using the developed computational code were found to have excellent agreement with the reference data. Also, the unfolded energy spectra of the neutron sources as obtained using ANFIS were more accurate than the results reported from calculations performed using artificial neural networks in previously published papers. © The Author(s) 2018. Published by Oxford University Press on behalf of The Japan Radiation Research Society and Japanese Society for Radiation Oncology.

Rare Event Simulation in Radiation Transport

NASA Astrophysics Data System (ADS)

Kollman, Craig

This dissertation studies methods for estimating extremely small probabilities by Monte Carlo simulation. Problems in radiation transport typically involve estimating very rare events or the expected value of a random variable which is with overwhelming probability equal to zero. These problems often have high dimensional state spaces and irregular geometries so that analytic solutions are not possible. Monte Carlo simulation must be used to estimate the radiation dosage being transported to a particular location. If the area is well shielded the probability of any one particular particle getting through is very small. Because of the large number of particles involved, even a tiny fraction penetrating the shield may represent an unacceptable level of radiation. It therefore becomes critical to be able to accurately estimate this extremely small probability. Importance sampling is a well known technique for improving the efficiency of rare event calculations. Here, a new set of probabilities is used in the simulation runs. The results are multiplied by the likelihood ratio between the true and simulated probabilities so as to keep our estimator unbiased. The variance of the resulting estimator is very sensitive to which new set of transition probabilities are chosen. It is shown that a zero variance estimator does exist, but that its computation requires exact knowledge of the solution. A simple random walk with an associated killing model for the scatter of neutrons is introduced. Large deviation results for optimal importance sampling in random walks are extended to the case where killing is present. An adaptive "learning" algorithm for implementing importance sampling is given for more general Markov chain models of neutron scatter. For finite state spaces this algorithm is shown to give, with probability one, a sequence of estimates converging exponentially fast to the true solution. In the final chapter, an attempt to generalize this algorithm to a continuous state space is made. This involves partitioning the space into a finite number of cells. There is a tradeoff between additional computation per iteration and variance reduction per iteration that arises in determining the optimal grid size. All versions of this algorithm can be thought of as a compromise between deterministic and Monte Carlo methods, capturing advantages of both techniques.

A comparison of acceleration methods for solving the neutron transport k-eigenvalue problem

NASA Astrophysics Data System (ADS)

Willert, Jeffrey; Park, H.; Knoll, D. A.

2014-10-01

Over the past several years a number of papers have been written describing modern techniques for numerically computing the dominant eigenvalue of the neutron transport criticality problem. These methods fall into two distinct categories. The first category of methods rewrite the multi-group k-eigenvalue problem as a nonlinear system of equations and solve the resulting system using either a Jacobian-Free Newton-Krylov (JFNK) method or Nonlinear Krylov Acceleration (NKA), a variant of Anderson Acceleration. These methods are generally successful in significantly reducing the number of transport sweeps required to compute the dominant eigenvalue. The second category of methods utilize Moment-Based Acceleration (or High-Order/Low-Order (HOLO) Acceleration). These methods solve a sequence of modified diffusion eigenvalue problems whose solutions converge to the solution of the original transport eigenvalue problem. This second class of methods is, in our experience, always superior to the first, as most of the computational work is eliminated by the acceleration from the LO diffusion system. In this paper, we review each of these methods. Our computational results support our claim that the choice of which nonlinear solver to use, JFNK or NKA, should be secondary. The primary computational savings result from the implementation of a HOLO algorithm. We display computational results for a series of challenging multi-dimensional test problems.

Simulations of neutron transport at low energy: a comparison between GEANT and MCNP.

PubMed

Colonna, N; Altieri, S

2002-06-01

The use of the simulation tool GEANT for neutron transport at energies below 20 MeV is discussed, in particular with regard to shielding and dose calculations. The reliability of the GEANT/MICAP package for neutron transport in a wide energy range has been verified by comparing the results of simulations performed with this package in a wide energy range with the prediction of MCNP-4B, a code commonly used for neutron transport at low energy. A reasonable agreement between the results of the two codes is found for the neutron flux through a slab of material (iron and ordinary concrete), as well as for the dose released in soft tissue by neutrons. These results justify the use of the GEANT/MICAP code for neutron transport in a wide range of applications, including health physics problems.

Limits on the Efficiency of Event-Based Algorithms for Monte Carlo Neutron Transport

DOE Office of Scientific and Technical Information (OSTI.GOV)

Romano, Paul K.; Siegel, Andrew R.

The traditional form of parallelism in Monte Carlo particle transport simulations, wherein each individual particle history is considered a unit of work, does not lend itself well to data-level parallelism. Event-based algorithms, which were originally used for simulations on vector processors, may offer a path toward better utilizing data-level parallelism in modern computer architectures. In this study, a simple model is developed for estimating the efficiency of the event-based particle transport algorithm under two sets of assumptions. Data collected from simulations of four reactor problems using OpenMC was then used in conjunction with the models to calculate the speedup duemore » to vectorization as a function of the size of the particle bank and the vector width. When each event type is assumed to have constant execution time, the achievable speedup is directly related to the particle bank size. We observed that the bank size generally needs to be at least 20 times greater than vector size to achieve vector efficiency greater than 90%. Lastly, when the execution times for events are allowed to vary, the vector speedup is also limited by differences in execution time for events being carried out in a single event-iteration.« less

Limits on the Efficiency of Event-Based Algorithms for Monte Carlo Neutron Transport

DOE PAGES

Romano, Paul K.; Siegel, Andrew R.

2017-07-01

The traditional form of parallelism in Monte Carlo particle transport simulations, wherein each individual particle history is considered a unit of work, does not lend itself well to data-level parallelism. Event-based algorithms, which were originally used for simulations on vector processors, may offer a path toward better utilizing data-level parallelism in modern computer architectures. In this study, a simple model is developed for estimating the efficiency of the event-based particle transport algorithm under two sets of assumptions. Data collected from simulations of four reactor problems using OpenMC was then used in conjunction with the models to calculate the speedup duemore » to vectorization as a function of the size of the particle bank and the vector width. When each event type is assumed to have constant execution time, the achievable speedup is directly related to the particle bank size. We observed that the bank size generally needs to be at least 20 times greater than vector size to achieve vector efficiency greater than 90%. Lastly, when the execution times for events are allowed to vary, the vector speedup is also limited by differences in execution time for events being carried out in a single event-iteration.« less

A NUMERICAL ALGORITHM FOR MODELING MULTIGROUP NEUTRINO-RADIATION HYDRODYNAMICS IN TWO SPATIAL DIMENSIONS

DOE Office of Scientific and Technical Information (OSTI.GOV)

Swesty, F. Douglas; Myra, Eric S.

It is now generally agreed that multidimensional, multigroup, neutrino-radiation hydrodynamics (RHD) is an indispensable element of any realistic model of stellar-core collapse, core-collapse supernovae, and proto-neutron star instabilities. We have developed a new, two-dimensional, multigroup algorithm that can model neutrino-RHD flows in core-collapse supernovae. Our algorithm uses an approach similar to the ZEUS family of algorithms, originally developed by Stone and Norman. However, this completely new implementation extends that previous work in three significant ways: first, we incorporate multispecies, multigroup RHD in a flux-limited-diffusion approximation. Our approach is capable of modeling pair-coupled neutrino-RHD, and includes effects of Pauli blocking inmore » the collision integrals. Blocking gives rise to nonlinearities in the discretized radiation-transport equations, which we evolve implicitly in time. We employ parallelized Newton-Krylov methods to obtain a solution of these nonlinear, implicit equations. Our second major extension to the ZEUS algorithm is the inclusion of an electron conservation equation that describes the evolution of electron-number density in the hydrodynamic flow. This permits calculating deleptonization of a stellar core. Our third extension modifies the hydrodynamics algorithm to accommodate realistic, complex equations of state, including those having nonconvex behavior. In this paper, we present a description of our complete algorithm, giving sufficient details to allow others to implement, reproduce, and extend our work. Finite-differencing details are presented in appendices. We also discuss implementation of this algorithm on state-of-the-art, parallel-computing architectures. Finally, we present results of verification tests that demonstrate the numerical accuracy of this algorithm on diverse hydrodynamic, gravitational, radiation-transport, and RHD sample problems. We believe our methods to be of general use in a variety of model settings where radiation transport or RHD is important. Extension of this work to three spatial dimensions is straightforward.« less

Implementation and Initial Testing of Advanced Processing and Analysis Algorithms for Correlated Neutron Counting

DOE Office of Scientific and Technical Information (OSTI.GOV)

Santi, Peter Angelo; Cutler, Theresa Elizabeth; Favalli, Andrea

In order to improve the accuracy and capabilities of neutron multiplicity counting, additional quantifiable information is needed in order to address the assumptions that are present in the point model. Extracting and utilizing higher order moments (Quads and Pents) from the neutron pulse train represents the most direct way of extracting additional information from the measurement data to allow for an improved determination of the physical properties of the item of interest. The extraction of higher order moments from a neutron pulse train required the development of advanced dead time correction algorithms which could correct for dead time effects inmore » all of the measurement moments in a self-consistent manner. In addition, advanced analysis algorithms have been developed to address specific assumptions that are made within the current analysis model, namely that all neutrons are created at a single point within the item of interest, and that all neutrons that are produced within an item are created with the same energy distribution. This report will discuss the current status of implementation and initial testing of the advanced dead time correction and analysis algorithms that have been developed in an attempt to utilize higher order moments to improve the capabilities of correlated neutron measurement techniques.« less

Improved adaptive genetic algorithm with sparsity constraint applied to thermal neutron CT reconstruction of two-phase flow

NASA Astrophysics Data System (ADS)

Yan, Mingfei; Hu, Huasi; Otake, Yoshie; Taketani, Atsushi; Wakabayashi, Yasuo; Yanagimachi, Shinzo; Wang, Sheng; Pan, Ziheng; Hu, Guang

2018-05-01

Thermal neutron computer tomography (CT) is a useful tool for visualizing two-phase flow due to its high imaging contrast and strong penetrability of neutrons for tube walls constructed with metallic material. A novel approach for two-phase flow CT reconstruction based on an improved adaptive genetic algorithm with sparsity constraint (IAGA-SC) is proposed in this paper. In the algorithm, the neighborhood mutation operator is used to ensure the continuity of the reconstructed object. The adaptive crossover probability P c and mutation probability P m are improved to help the adaptive genetic algorithm (AGA) achieve the global optimum. The reconstructed results for projection data, obtained from Monte Carlo simulation, indicate that the comprehensive performance of the IAGA-SC algorithm exceeds the adaptive steepest descent-projection onto convex sets (ASD-POCS) algorithm in restoring typical and complex flow regimes. It especially shows great advantages in restoring the simply connected flow regimes and the shape of object. In addition, the CT experiment for two-phase flow phantoms was conducted on the accelerator-driven neutron source to verify the performance of the developed IAGA-SC algorithm.

Performance improvements of wavelength-shifting-fiber neutron detectors using high-resolution positioning algorithms

DOE PAGES

Wang, C. L.

2016-05-17

On the basis of FluoroBancroft linear-algebraic method [S.B. Andersson, Opt. Exp. 16, 18714 (2008)] three highly-resolved positioning methods were proposed for wavelength-shifting fiber (WLSF) neutron detectors. Using a Gaussian or exponential-decay light-response function (LRF), the non-linear relation of photon-number profiles vs. x-pixels was linearized and neutron positions were determined. The proposed algorithms give an average 0.03-0.08 pixel position error, much smaller than that (0.29 pixel) from a traditional maximum photon algorithm (MPA). The new algorithms result in better detector uniformity, less position misassignment (ghosting), better spatial resolution, and an equivalent or better instrument resolution in powder diffraction than the MPA.more » Moreover, these characters will facilitate broader applications of WLSF detectors at time-of-flight neutron powder diffraction beamlines, including single-crystal diffraction and texture analysis.« less

Population-based metaheuristic optimization in neutron optics and shielding design

NASA Astrophysics Data System (ADS)

DiJulio, D. D.; Björgvinsdóttir, H.; Zendler, C.; Bentley, P. M.

2016-11-01

Population-based metaheuristic algorithms are powerful tools in the design of neutron scattering instruments and the use of these types of algorithms for this purpose is becoming more and more commonplace. Today there exists a wide range of algorithms to choose from when designing an instrument and it is not always initially clear which may provide the best performance. Furthermore, due to the nature of these types of algorithms, the final solution found for a specific design scenario cannot always be guaranteed to be the global optimum. Therefore, to explore the potential benefits and differences between the varieties of these algorithms available, when applied to such design scenarios, we have carried out a detailed study of some commonly used algorithms. For this purpose, we have developed a new general optimization software package which combines a number of common metaheuristic algorithms within a single user interface and is designed specifically with neutronic calculations in mind. The algorithms included in the software are implementations of Particle-Swarm Optimization (PSO), Differential Evolution (DE), Artificial Bee Colony (ABC), and a Genetic Algorithm (GA). The software has been used to optimize the design of several problems in neutron optics and shielding, coupled with Monte-Carlo simulations, in order to evaluate the performance of the various algorithms. Generally, the performance of the algorithms depended on the specific scenarios, however it was found that DE provided the best average solutions in all scenarios investigated in this work.

Rare event simulation in radiation transport

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kollman, Craig

1993-10-01

This dissertation studies methods for estimating extremely small probabilities by Monte Carlo simulation. Problems in radiation transport typically involve estimating very rare events or the expected value of a random variable which is with overwhelming probability equal to zero. These problems often have high dimensional state spaces and irregular geometries so that analytic solutions are not possible. Monte Carlo simulation must be used to estimate the radiation dosage being transported to a particular location. If the area is well shielded the probability of any one particular particle getting through is very small. Because of the large number of particles involved,more » even a tiny fraction penetrating the shield may represent an unacceptable level of radiation. It therefore becomes critical to be able to accurately estimate this extremely small probability. Importance sampling is a well known technique for improving the efficiency of rare event calculations. Here, a new set of probabilities is used in the simulation runs. The results are multiple by the likelihood ratio between the true and simulated probabilities so as to keep the estimator unbiased. The variance of the resulting estimator is very sensitive to which new set of transition probabilities are chosen. It is shown that a zero variance estimator does exist, but that its computation requires exact knowledge of the solution. A simple random walk with an associated killing model for the scatter of neutrons is introduced. Large deviation results for optimal importance sampling in random walks are extended to the case where killing is present. An adaptive ``learning`` algorithm for implementing importance sampling is given for more general Markov chain models of neutron scatter. For finite state spaces this algorithm is shown to give with probability one, a sequence of estimates converging exponentially fast to the true solution.« less

Experimental validation of a coupled neutron-photon inverse radiation transport solver

NASA Astrophysics Data System (ADS)

Mattingly, John; Mitchell, Dean J.; Harding, Lee T.

2011-10-01

Sandia National Laboratories has developed an inverse radiation transport solver that applies nonlinear regression to coupled neutron-photon deterministic transport models. The inverse solver uses nonlinear regression to fit a radiation transport model to gamma spectrometry and neutron multiplicity counting measurements. The subject of this paper is the experimental validation of that solver. This paper describes a series of experiments conducted with a 4.5 kg sphere of α-phase, weapons-grade plutonium. The source was measured bare and reflected by high-density polyethylene (HDPE) spherical shells with total thicknesses between 1.27 and 15.24 cm. Neutron and photon emissions from the source were measured using three instruments: a gross neutron counter, a portable neutron multiplicity counter, and a high-resolution gamma spectrometer. These measurements were used as input to the inverse radiation transport solver to evaluate the solver's ability to correctly infer the configuration of the source from its measured radiation signatures.

Radiation shielding quality assurance

NASA Astrophysics Data System (ADS)

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.

Application of perturbation theory to lattice calculations based on method of cyclic characteristics

NASA Astrophysics Data System (ADS)

Assawaroongruengchot, Monchai

Perturbation theory is a technique used for the estimation of changes in performance functionals, such as linear reaction rate ratio and eigenvalue affected by small variations in reactor core compositions. Here the algorithm of perturbation theory is developed for the multigroup integral neutron transport problems in 2D fuel assemblies with isotropic scattering. The integral transport equation is used in the perturbative formulation because it represents the interconnecting neutronic systems of the lattice assemblies via the tracking lines. When the integral neutron transport equation is used in the formulation, one needs to solve the resulting integral transport equations for the flux importance and generalized flux importance functions. The relationship between the generalized flux importance and generalized source importance functions is defined in order to transform the generalized flux importance transport equations into the integro-differential equations for the generalized adjoints. Next we develop the adjoint and generalized adjoint transport solution algorithms based on the method of cyclic characteristics (MOCC) in DRAGON code. In the MOCC method, the adjoint characteristics equations associated with a cyclic tracking line are formulated in such a way that a closed form for the adjoint angular function can be obtained. The MOCC method then requires only one cycle of scanning over the cyclic tracking lines in each spatial iteration. We also show that the source importance function by CP method is mathematically equivalent to the adjoint function by MOCC method. In order to speed up the MOCC solution algorithm, a group-reduction and group-splitting techniques based on the structure of the adjoint scattering matrix are implemented. A combined forward flux/adjoint function iteration scheme, based on the group-splitting technique and the common use of a large number of variables storing tracking-line data and exponential values, is proposed to reduce the computing time when both direct and adjoint solutions are required. A problem that arises for the generalized adjoint problem is that the direct use of the negative external generalized adjoint sources in the adjoint solution algorithm results in negative generalized adjoint functions. A coupled flux biasing/decontamination scheme is applied to make the generalized adjoint functions positive using the adjoint functions in such a way that it can be used for the multigroup rebalance technique. Next we consider the application of the perturbation theory to the reactor problems. Since the coolant void reactivity (CVR) is a important factor in reactor safety analysis, we have decided to select this parameter for optimization studies. We consider the optimization and adjoint sensitivity techniques for the adjustments of CVR at beginning of burnup cycle (BOC) and k eff at end of burnup cycle (EOC) for a 2D Advanced CANDU Reactor (ACR) lattice. The sensitivity coefficients are evaluated using the perturbation theory based on the integral transport equations. Three sets of parameters for CVR-BOC and keff-EOC adjustments are studied: (1) Dysprosium density in the central pin with Uranium enrichment in the outer fuel rings, (2) Dysprosium density and Uranium enrichment both in the central pin, and (3) the same parameters as in the first case but the objective is to obtain a negative checkerboard CVR at beginning of cycle (CBCVR-BOC). To approximate the sensitivity coefficient at EOC, we perform constant-power burnup/depletion calculations for 600 full power days (FPD) using a slightly perturbed nuclear library and the unperturbed neutron fluxes to estimate the variation of nuclide densities at EOC. Sensitivity analyses of CVR and eigenvalue are included in the study. In addition the optimization and adjoint sensitivity techniques are applied to the CBCVR-BOC and keff-EOC adjustment of the ACR lattices with Gadolinium in the central pin. Finally we apply these techniques to the CVR-BOC, CVR-EOC and keff-EOC adjustment of a CANDU lattice of which the burnup period is extended from 300 to 450 FPDs. The cases with the central pin containing either Dysprosium or Gadolinium in the natural Uranium are considered in our study. (Abstract shortened by UMI.)

Benchmark solution for the Spencer-Lewis equation of electron transport theory

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ganapol, B.D.

As integrated circuits become smaller, the shielding of these sensitive components against penetrating electrons becomes extremely critical. Monte Carlo methods have traditionally been the method of choice in shielding evaluations primarily because they can incorporate a wide variety of relevant physical processes. Recently, however, as a result of a more accurate numerical representation of the highly forward peaked scattering process, S/sub n/ methods for one-dimensional problems have been shown to be at least as cost-effective in comparison with Monte Carlo methods. With the development of these deterministic methods for electron transport, a need has arisen to assess the accuracy ofmore » proposed numerical algorithms and to ensure their proper coding. It is the purpose of this presentation to develop a benchmark to the Spencer-Lewis equation describing the transport of energetic electrons in solids. The solution will take advantage of the correspondence between the Spencer-Lewis equation and the transport equation describing one-group time-dependent neutron transport.« less

Importance biasing scheme implemented in the PRIZMA code

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kandiev, I.Z.; Malyshkin, G.N.

1997-12-31

PRIZMA code is intended for Monte Carlo calculations of linear radiation transport problems. The code has wide capabilities to describe geometry, sources, material composition, and to obtain parameters specified by user. There is a capability to calculate path of particle cascade (including neutrons, photons, electrons, positrons and heavy charged particles) taking into account possible transmutations. Importance biasing scheme was implemented to solve the problems which require calculation of functionals related to small probabilities (for example, problems of protection against radiation, problems of detection, etc.). The scheme enables to adapt trajectory building algorithm to problem peculiarities.

Study on beam geometry and image reconstruction algorithm in fast neutron computerized tomography at NECTAR facility

NASA Astrophysics Data System (ADS)

Guo, J.; Bücherl, T.; Zou, Y.; Guo, Z.

2011-09-01

Investigations on the fast neutron beam geometry for the NECTAR facility are presented. The results of MCNP simulations and experimental measurements of the beam distributions at NECTAR are compared. Boltzmann functions are used to describe the beam profile in the detection plane assuming the area source to be set up of large number of single neutron point sources. An iterative algebraic reconstruction algorithm is developed, realized and verified by both simulated and measured projection data. The feasibility for improved reconstruction in fast neutron computerized tomography at the NECTAR facility is demonstrated.

Limits on the Efficiency of Event-Based Algorithms for Monte Carlo Neutron Transport

DOE Office of Scientific and Technical Information (OSTI.GOV)

Romano, Paul K.; Siegel, Andrew R.

The traditional form of parallelism in Monte Carlo particle transport simulations, wherein each individual particle history is considered a unit of work, does not lend itself well to data-level parallelism. Event-based algorithms, which were originally used for simulations on vector processors, may offer a path toward better utilizing data-level parallelism in modern computer architectures. In this study, a simple model is developed for estimating the efficiency of the event-based particle transport algorithm under two sets of assumptions. Data collected from simulations of four reactor problems using OpenMC was then used in conjunction with the models to calculate the speedup duemore » to vectorization as a function of two parameters: the size of the particle bank and the vector width. When each event type is assumed to have constant execution time, the achievable speedup is directly related to the particle bank size. We observed that the bank size generally needs to be at least 20 times greater than vector size in order to achieve vector efficiency greater than 90%. When the execution times for events are allowed to vary, however, the vector speedup is also limited by differences in execution time for events being carried out in a single event-iteration. For some problems, this implies that vector effciencies over 50% may not be attainable. While there are many factors impacting performance of an event-based algorithm that are not captured by our model, it nevertheless provides insights into factors that may be limiting in a real implementation.« less

Total variation-based neutron computed tomography

NASA Astrophysics Data System (ADS)

Barnard, Richard C.; Bilheux, Hassina; Toops, Todd; Nafziger, Eric; Finney, Charles; Splitter, Derek; Archibald, Rick

2018-05-01

We perform the neutron computed tomography reconstruction problem via an inverse problem formulation with a total variation penalty. In the case of highly under-resolved angular measurements, the total variation penalty suppresses high-frequency artifacts which appear in filtered back projections. In order to efficiently compute solutions for this problem, we implement a variation of the split Bregman algorithm; due to the error-forgetting nature of the algorithm, the computational cost of updating can be significantly reduced via very inexact approximate linear solvers. We present the effectiveness of the algorithm in the significantly low-angular sampling case using synthetic test problems as well as data obtained from a high flux neutron source. The algorithm removes artifacts and can even roughly capture small features when an extremely low number of angles are used.

Development of Neutron Energy Spectral Signatures for Passive Monitoring of Spent Nuclear Fuels in Dry Cask Storage

NASA Astrophysics Data System (ADS)

Harkness, Ira; Zhu, Ting; Liang, Yinong; Rauch, Eric; Enqvist, Andreas; Jordan, Kelly A.

2018-01-01

Demand for spent nuclear fuel dry casks as an interim storage solution has increased globally and the IAEA has expressed a need for robust safeguards and verification technologies for ensuring the continuity of knowledge and the integrity of radioactive materials inside spent fuel casks. Existing research has been focusing on "fingerprinting" casks based on count rate statistics to represent radiation emission signatures. The current research aims to expand to include neutron energy spectral information as part of the fuel characteristics. First, spent fuel composition data are taken from the Next Generation Safeguards Initiative Spent Fuel Libraries, representative for Westinghouse 17ˣ17 PWR assemblies. The ORIGEN-S code then calculates the spontaneous fission and (α,n) emissions for individual fuel rods, followed by detailed MCNP simulations of neutrons transported through the fuel assemblies. A comprehensive database of neutron energy spectral profiles is to be constructed, with different enrichment, burn-up, and cooling time conditions. The end goal is to utilize the computational spent fuel library, predictive algorithm, and a pressurized 4He scintillator to verify the spent fuel assemblies inside a cask. This work identifies neutron spectral signatures that correlate with the cooling time of spent fuel. Both the total and relative contributions from spontaneous fission and (α,n) change noticeably with respect to cooling time, due to the relatively short half-life (18 years) of the major neutron source 244Cm. Identification of this and other neutron spectral signatures allows the characterization of spent nuclear fuels in dry cask storage.

MODA: a new algorithm to compute optical depths in multidimensional hydrodynamic simulations

NASA Astrophysics Data System (ADS)

Perego, Albino; Gafton, Emanuel; Cabezón, Rubén; Rosswog, Stephan; Liebendörfer, Matthias

2014-08-01

Aims: We introduce the multidimensional optical depth algorithm (MODA) for the calculation of optical depths in approximate multidimensional radiative transport schemes, equally applicable to neutrinos and photons. Motivated by (but not limited to) neutrino transport in three-dimensional simulations of core-collapse supernovae and neutron star mergers, our method makes no assumptions about the geometry of the matter distribution, apart from expecting optically transparent boundaries. Methods: Based on local information about opacities, the algorithm figures out an escape route that tends to minimize the optical depth without assuming any predefined paths for radiation. Its adaptivity makes it suitable for a variety of astrophysical settings with complicated geometry (e.g., core-collapse supernovae, compact binary mergers, tidal disruptions, star formation, etc.). We implement the MODA algorithm into both a Eulerian hydrodynamics code with a fixed, uniform grid and into an SPH code where we use a tree structure that is otherwise used for searching neighbors and calculating gravity. Results: In a series of numerical experiments, we compare the MODA results with analytically known solutions. We also use snapshots from actual 3D simulations and compare the results of MODA with those obtained with other methods, such as the global and local ray-by-ray method. It turns out that MODA achieves excellent accuracy at a moderate computational cost. In appendix we also discuss implementation details and parallelization strategies.

Targeted Modification of Neutron Energy Spectra for National Security Applications

NASA Astrophysics Data System (ADS)

Bevins, James Edward

At its core, research represents an attempt to break from the "this is the way we have always done it" paradigm. This idea is evidenced from the start in this research effort by the problem formulation to develop a new way to generate synthetic debris that mimics the samples that would be collected for forensics purposes following a nuclear weapon attack on the U.S. or its allies. The philosophy is also demonstrated by the design methodology used to solve the synthetic debris problem, using methods not commonly applied to nuclear engineering problems. Through this research, the bounds of what is deemed possible in neutron spectral shaping are moved ever so slightly. A capability for the production of synthetic debris and fission products was developed for the National Ignition Facility (NIF). Synthetic debris has historically been made in a limited fashion using sample doping techniques since the cessation of nuclear weapons testing, but a more robust alternative approach using neutron spectral shaping was proposed and developed by the University of California-Berkeley and Lawrence Livermore National Laboratory (LLNL). Using NIF as a starting source spectrum, the energy tuning assembly (ETA) developed in this work can irradiate samples with a combined thermonuclear and prompt fission neutron spectrum (TN+PFNS). When used with fissile foils, this irradiation will produce a synthetic fission product distribution that is realistic across all mass chains. To design the ETA, traditional parametric point design approaches were discarded in favor of formal optimization techniques. Finding a lack of suitable algorithms in the literature, a metaheuristic-based optimization algorithm, Gnowee, was developed for rapid convergence to nearly globally optimum solutions for complex, constrained engineering problems with mixed-integer and combinatorial design vectors and high-cost, noisy, discontinuous, black box objective function evaluations. Comparisons between Gnowee and several well-established metaheuristic algorithms are made for a set of continuous, mixed-integer, and combinatorial benchmarks. These results demonstrated Gnoweee to have superior flexibility and convergence characteristics over a wide range of design spaces. The Gnowee algorithm was implemented in Coeus, a new piece of software, to perform optimization of design problems requiring radiation transport for the evaluation of their objective functions. Currently, Coeus solves ETA optimization problems using hybrid radiation transport (ADVANTG and MCNP) to assess design permutations developed by Gnowee. Future enhancements of Coeus will look to expand the geometries and objective functions considered to those beyond ETA design. Coeus was used to generate an ETA design for the TN+PFNS application on NIF. The design achieved a reasonable match with the objective TN+PFNS and associated fission product distributions within the size and weight constraints imposed by the NIF facility. The ETA design was built by American Elements, and initial validation tests were conducted at the Lawrence Berkeley National Laboratory's 88-Inch Cyclotron. These experiments used foil activation and pulse height spectroscopy to measure the ETA-modified spectrum. Additionally, pulse height spectroscopy measurements were taken as the ETA was built-up component-by-component to measure the impact of nuclear data on the ability to model the ETA performance. Some initial analysis of these results is included here. Finally, an integral validation experiment on NIF was proposed using the Coeus generated ETA design. A scoping study conducted by LLNL determined the proposed experiment and ETA design are within NIF facility limitations and current radio-chemistry capabilities. The study found that the proposed ETA experiment was "low risk," has "no show stoppers," and has a "reasonable cost." All that is needed is a sponsor to close the last funding gap and bring the experiment to fruition. This research broke with the current sample doping approach and applied neutron spectral shaping to design an ETA that can create realistic synthetic fission and activation products and improve technical nuclear forensics outcomes. However, the ETA presented in this research represents more than a stand alone point design with a limited scope and application. It is proof of a concept and the product of a unique capability that has a wide range of potential applications. This research demonstrates that the concept of neutron spectral shaping can be used to engineer complex neutron spectra within the confines of physics. There are many possible applications that could benefit from the ability to generate custom energy neutron spectra that fall outside of current sources and methods. The ETA is the product of a general-purpose optimization algorithm, Gnowee, and design framework, Coeus, which enables the use of Gnowee for complex nuclear design problems. Through Gnowee and Coeus, new ETA neutronics designs can be generated in days, not months or years, with a drastic reduction in the research effort required to do so. (Abstract shortened by ProQuest.).

Apparatus, Method and Program Storage Device for Determining High-Energy Neutron/Ion Transport to a Target of Interest

NASA Technical Reports Server (NTRS)

Wilson, John W. (Inventor); Tripathi, Ram K. (Inventor); Cucinotta, Francis A. (Inventor); Badavi, Francis F. (Inventor)

2012-01-01

An apparatus, method and program storage device for determining high-energy neutron/ion transport to a target of interest. Boundaries are defined for calculation of a high-energy neutron/ion transport to a target of interest; the high-energy neutron/ion transport to the target of interest is calculated using numerical procedures selected to reduce local truncation error by including higher order terms and to allow absolute control of propagated error by ensuring truncation error is third order in step size, and using scaling procedures for flux coupling terms modified to improve computed results by adding a scaling factor to terms describing production of j-particles from collisions of k-particles; and the calculated high-energy neutron/ion transport is provided to modeling modules to control an effective radiation dose at the target of interest.

ALGORITHM BASED ON ARTIFICIAL BEE COLONY FOR UNFOLDING OF NEUTRON SPECTRA OBTAINED WITH BONNER SPHERES.

PubMed

Silva, Everton R; Freitas, Bruno M; Santos, Denison S; Maurício, Cláudia L P

2018-04-13

Occupational neutron fields usually have energies from the thermal range to some MeV and the characterization of the spectra is essential for estimation of the radioprotection quantities. Thus, the spectrum must be unfolded based on a limited number of measurements. This study implemented an algorithm based on the bee colonies behavior, named Artificial Bee Colony (ABC), where the intelligent behavior of the bees in search of food is reproduced to perform the unfolding of neutron spectra. The experimental measurements used Bonner spheres and 6LiI (Eu) detector, with irradiations using a thermal neutron flux and three reference fields: 241Am-Be, 252Cf and 252Cf (D2O). The ABC obtained good estimation of the expected spectrum even without previous information and its results were closer to expected spectra than those obtained by the SPUNIT algorithm.

Fission Reaction Event Yield Algorithm

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hagmann, Christian; Verbeke, Jerome; Vogt, Ramona

FREYA (Fission Reaction Event Yield Algorithm) is a code that simulated the decay of a fissionable nucleus at specified excitation energy. In its present form, FREYA models spontaneous fission and neutron-induced fission up to 20 MeV. It includes the possibility of neutron emission from the nuclear prior to its fussion (nth chance fission).

In situ calibration of neutron activation system on the large helical device

NASA Astrophysics Data System (ADS)

Pu, N.; Nishitani, T.; Isobe, M.; Ogawa, K.; Kawase, H.; Tanaka, T.; Li, S. Y.; Yoshihashi, S.; Uritani, A.

2017-11-01

In situ calibration of the neutron activation system on the Large Helical Device (LHD) was performed by using an intense 252Cf neutron source. To simulate a ring-shaped neutron source, we installed a railway inside the LHD vacuum vessel and made a train loaded with the 252Cf source run along a typical magnetic axis position. Three activation capsules loaded with thirty pieces of indium foils stacked with total mass of approximately 18 g were prepared. Each capsule was irradiated over 15 h while the train was circulating. The activation response coefficient (9.4 ± 1.2) × 10-8 of 115In(n, n')115mIn reaction obtained from the experiment is in good agreement with results from three-dimensional neutron transport calculations using the Monte Carlo neutron transport simulation code 6. The activation response coefficients of 2.45 MeV birth neutron and secondary 14.1 MeV neutron from deuterium plasma were evaluated from the activation response coefficient obtained in this calibration experiment with results from three-dimensional neutron calculations using the Monte Carlo neutron transport simulation code 6.

Benchmark test of transport calculations of gold and nickel activation with implications for neutron kerma at Hiroshima.

PubMed

Hoshi, M; Hiraoka, M; Hayakawa, N; Sawada, S; Munaka, M; Kuramoto, A; Oka, T; Iwatani, K; Shizuma, K; Hasai, H

1992-11-01

A benchmark test of the Monte Carlo neutron and photon transport code system (MCNP) was performed using a 252Cf fission neutron source to validate the use of the code for the energy spectrum analyses of Hiroshima atomic bomb neutrons. Nuclear data libraries used in the Monte Carlo neutron and photon transport code calculation were ENDF/B-III, ENDF/B-IV, LASL-SUB, and ENDL-73. The neutron moderators used were granite (the main component of which is SiO2, with a small fraction of hydrogen), Newlight [polyethylene with 3.7% boron (natural)], ammonium chloride (NH4Cl), and water (H2O). Each moderator was 65 cm thick. The neutron detectors were gold and nickel foils, which were used to detect thermal and epithermal neutrons (4.9 eV) and fast neutrons (> 0.5 MeV), respectively. Measured activity data from neutron-irradiated gold and nickel foils in these moderators decreased to about 1/1,000th or 1/10,000th, which correspond to about 1,500 m ground distance from the hypocenter in Hiroshima. For both gold and nickel detectors, the measured activities and the calculated values agreed within 10%. The slopes of the depth-yield relations in each moderator, except granite, were similar for neutrons detected by the gold and nickel foils. From the results of these studies, the Monte Carlo neutron and photon transport code was verified to be accurate enough for use with the elements hydrogen, carbon, nitrogen, oxygen, silicon, chlorine, and cadmium, and for the incident 252Cf fission spectrum neutrons.

Verification of ARES transport code system with TAKEDA benchmarks

NASA Astrophysics Data System (ADS)

Zhang, Liang; Zhang, Bin; Zhang, Penghe; Chen, Mengteng; Zhao, Jingchang; Zhang, Shun; Chen, Yixue

2015-10-01

Neutron transport modeling and simulation are central to many areas of nuclear technology, including reactor core analysis, radiation shielding and radiation detection. In this paper the series of TAKEDA benchmarks are modeled to verify the critical calculation capability of ARES, a discrete ordinates neutral particle transport code system. SALOME platform is coupled with ARES to provide geometry modeling and mesh generation function. The Koch-Baker-Alcouffe parallel sweep algorithm is applied to accelerate the traditional transport calculation process. The results show that the eigenvalues calculated by ARES are in excellent agreement with the reference values presented in NEACRP-L-330, with a difference less than 30 pcm except for the first case of model 3. Additionally, ARES provides accurate fluxes distribution compared to reference values, with a deviation less than 2% for region-averaged fluxes in all cases. All of these confirms the feasibility of ARES-SALOME coupling and demonstrate that ARES has a good performance in critical calculation.

An algorithm for charge-integration, pulse-shape discrimination and estimation of neutron/photon misclassification in organic scintillators

NASA Astrophysics Data System (ADS)

Polack, J. K.; Flaska, M.; Enqvist, A.; Sosa, C. S.; Lawrence, C. C.; Pozzi, S. A.

2015-09-01

Organic scintillators are frequently used for measurements that require sensitivity to both photons and fast neutrons because of their pulse shape discrimination capabilities. In these measurement scenarios, particle identification is commonly handled using the charge-integration pulse shape discrimination method. This method works particularly well for high-energy depositions, but is prone to misclassification for relatively low-energy depositions. A novel algorithm has been developed for automatically performing charge-integration pulse shape discrimination in a consistent and repeatable manner. The algorithm is able to estimate the photon and neutron misclassification corresponding to the calculated discrimination parameters, and is capable of doing so using only the information measured by a single organic scintillator. This paper describes the algorithm and assesses its performance by comparing algorithm-estimated misclassification to values computed via a more traditional time-of-flight estimation. A single data set was processed using four different low-energy thresholds: 40, 60, 90, and 120 keVee. Overall, the results compared well between the two methods; in most cases, the algorithm-estimated values fell within the uncertainties of the TOF-estimated values.

Simultaneous parameter optimization of x-ray and neutron reflectivity data using genetic algorithms

DOE Office of Scientific and Technical Information (OSTI.GOV)

Singh, Surendra, E-mail: surendra@barc.gov.in; Basu, Saibal

2016-05-23

X-ray and neutron reflectivity are two non destructive techniques which provide a wealth of information on thickness, structure and interracial properties in nanometer length scale. Combination of X-ray and neutron reflectivity is well suited for obtaining physical parameters of nanostructured thin films and superlattices. Neutrons provide a different contrast between the elements than X-rays and are also sensitive to the magnetization depth profile in thin films and superlattices. The real space information is extracted by fitting a model for the structure of the thin film sample in reflectometry experiments. We have applied a Genetic Algorithms technique to extract depth dependentmore » structure and magnetic in thin film and multilayer systems by simultaneously fitting X-ray and neutron reflectivity data.« less

Chapman Enskog-maximum entropy method on time-dependent neutron transport equation

NASA Astrophysics Data System (ADS)

Abdou, M. A.

2006-09-01

The time-dependent neutron transport equation in semi and infinite medium with linear anisotropic and Rayleigh scattering is proposed. The problem is solved by means of the flux-limited, Chapman Enskog-maximum entropy for obtaining the solution of the time-dependent neutron transport. The solution gives the neutron distribution density function which is used to compute numerically the radiant energy density E(x,t), net flux F(x,t) and reflectivity Rf. The behaviour of the approximate flux-limited maximum entropy neutron density function are compared with those found by other theories. Numerical calculations for the radiant energy, net flux and reflectivity of the proposed medium are calculated at different time and space.

Feasibility of constraining the curvature parameter of the symmetry energy using elliptic flow data

NASA Astrophysics Data System (ADS)

Cozma, M. D.

2018-03-01

A QMD transport model that employs a modified momentum dependent interaction (MDI2) potential, supplemented by a phase-space coalescence model fitted to FOPI experimental multiplicities of free nucleons and light clusters is used to study the density dependence of the symmetry energy above the saturation point by a comparison with experimental elliptic flow ratios measured by the FOPI-LAND and ASYEOS Collaborations in 197Au + 197Au collisions at 400 MeV/nucleon impact energy. A previous calculation using the same model has proven that neutron-to-proton and neutron-to-charged-particles elliptic flow ratios probe on average different densities allowing in principle the extraction of both the slope L and curvature K_{sym} parameters of the symmetry energy. To make use of this result a Gogny interaction inspired potential is modified by the addition of a density dependent, momentum independent term, while enforcing a close description of the empirical nucleon optical potential, allowing independent modifications of L and Ksym. Comparing theoretical predictions with experimental data for neutron-to-proton and neutron-to-charged-particles elliptic flow ratios the following constraint is extracted: L = 85 ± 22(exp) ± 20(th) ± 12(sys) MeV and K_{sym} = 96 ± 315(exp) ± 170(th) ± 166(sys) MeV. Theoretical errors include effects due to uncertainties in the isoscalar part of the equation of state, value of the isovector neutron-proton effective mass splitting, in-medium effects on the elastic nucleon-nucleon cross-sections, Pauli blocking algorithm variants and scenario considered for the conservation of the total energy of the system. Systematical uncertainties are generated by the inability of the transport model to reproduce experimental light-cluster-to-proton multiplicity ratios. A value for L free of systematical theoretical uncertainties can be extracted from the neutron-to-proton elliptic flow ratio alone: L = 84 ± 30(exp) ± 19(th) MeV. It is demonstrated that elliptic flow ratios reach a maximum sensitivity on the K_{sym} parameter in heavy-ion collisions of about 250 MeV/nucleon impact energy, allowing a reduction of its experimental component of uncertainty to about 150 MeV.

Transmutation approximations for the application of hybrid Monte Carlo/deterministic neutron transport to shutdown dose rate analysis

DOE Office of Scientific and Technical Information (OSTI.GOV)

Biondo, Elliott D.; Wilson, Paul P. H.

In fusion energy systems (FES) neutrons born from burning plasma activate system components. The photon dose rate after shutdown from resulting radionuclides must be quantified. This shutdown dose rate (SDR) is calculated by coupling neutron transport, activation analysis, and photon transport. The size, complexity, and attenuating configuration of FES motivate the use of hybrid Monte Carlo (MC)/deterministic neutron transport. The Multi-Step Consistent Adjoint Driven Importance Sampling (MS-CADIS) method can be used to optimize MC neutron transport for coupled multiphysics problems, including SDR analysis, using deterministic estimates of adjoint flux distributions. When used for SDR analysis, MS-CADIS requires the formulation ofmore » an adjoint neutron source that approximates the transmutation process. In this work, transmutation approximations are used to derive a solution for this adjoint neutron source. It is shown that these approximations are reasonably met for typical FES neutron spectra and materials over a range of irradiation scenarios. When these approximations are met, the Groupwise Transmutation (GT)-CADIS method, proposed here, can be used effectively. GT-CADIS is an implementation of the MS-CADIS method for SDR analysis that uses a series of single-energy-group irradiations to calculate the adjoint neutron source. For a simple SDR problem, GT-CADIS provides speedups of 200 100 relative to global variance reduction with the Forward-Weighted (FW)-CADIS method and 9 ± 5 • 104 relative to analog. As a result, this work shows that GT-CADIS is broadly applicable to FES problems and will significantly reduce the computational resources necessary for SDR analysis.« less

Transmutation approximations for the application of hybrid Monte Carlo/deterministic neutron transport to shutdown dose rate analysis

DOE PAGES

Biondo, Elliott D.; Wilson, Paul P. H.

2017-05-08

In fusion energy systems (FES) neutrons born from burning plasma activate system components. The photon dose rate after shutdown from resulting radionuclides must be quantified. This shutdown dose rate (SDR) is calculated by coupling neutron transport, activation analysis, and photon transport. The size, complexity, and attenuating configuration of FES motivate the use of hybrid Monte Carlo (MC)/deterministic neutron transport. The Multi-Step Consistent Adjoint Driven Importance Sampling (MS-CADIS) method can be used to optimize MC neutron transport for coupled multiphysics problems, including SDR analysis, using deterministic estimates of adjoint flux distributions. When used for SDR analysis, MS-CADIS requires the formulation ofmore » an adjoint neutron source that approximates the transmutation process. In this work, transmutation approximations are used to derive a solution for this adjoint neutron source. It is shown that these approximations are reasonably met for typical FES neutron spectra and materials over a range of irradiation scenarios. When these approximations are met, the Groupwise Transmutation (GT)-CADIS method, proposed here, can be used effectively. GT-CADIS is an implementation of the MS-CADIS method for SDR analysis that uses a series of single-energy-group irradiations to calculate the adjoint neutron source. For a simple SDR problem, GT-CADIS provides speedups of 200 100 relative to global variance reduction with the Forward-Weighted (FW)-CADIS method and 9 ± 5 • 104 relative to analog. As a result, this work shows that GT-CADIS is broadly applicable to FES problems and will significantly reduce the computational resources necessary for SDR analysis.« less

Changes in prescribed doses for the Seattle neutron therapy system

NASA Astrophysics Data System (ADS)

Popescu, A.

2008-06-01

From the beginning of the neutron therapy program at the University of Washington Medical Center, the neutron dose distribution in tissue has been calculated using an in-house treatment planning system called PRISM. In order to increase the accuracy of the absorbed dose calculations, two main improvements were made to the PRISM treatment planning system: (a) the algorithm was changed by the addition of an analytical expression of the central axis wedge factor dependence with field size and depth developed at UWMC. Older versions of the treatment-planning algorithm used a constant central axis wedge factor; (b) a complete newly commissioned set of measured data was introduced in the latest version of PRISM. The new version of the PRISM algorithm allowed for the use of the wedge profiles measured at different depths instead of one wedge profile measured at one depth. The comparison of the absorbed dose calculations using the old and the improved algorithm showed discrepancies mainly due to the missing central axis wedge factor dependence with field size and depth and due to the absence of the wedge profiles at depths different from 10 cm. This study concludes that the previously reported prescribed doses for neutron therapy should be changed.

Energy spectra unfolding of fast neutron sources using the group method of data handling and decision tree algorithms

NASA Astrophysics Data System (ADS)

Hosseini, Seyed Abolfazl; Afrakoti, Iman Esmaili Paeen

2017-04-01

Accurate unfolding of the energy spectrum of a neutron source gives important information about unknown neutron sources. The obtained information is useful in many areas like nuclear safeguards, nuclear nonproliferation, and homeland security. In the present study, the energy spectrum of a poly-energetic fast neutron source is reconstructed using the developed computational codes based on the Group Method of Data Handling (GMDH) and Decision Tree (DT) algorithms. The neutron pulse height distribution (neutron response function) in the considered NE-213 liquid organic scintillator has been simulated using the developed MCNPX-ESUT computational code (MCNPX-Energy engineering of Sharif University of Technology). The developed computational codes based on the GMDH and DT algorithms use some data for training, testing and validation steps. In order to prepare the required data, 4000 randomly generated energy spectra distributed over 52 bins are used. The randomly generated energy spectra and the simulated neutron pulse height distributions by MCNPX-ESUT for each energy spectrum are used as the output and input data. Since there is no need to solve the inverse problem with an ill-conditioned response matrix, the unfolded energy spectrum has the highest accuracy. The 241Am-9Be and 252Cf neutron sources are used in the validation step of the calculation. The unfolded energy spectra for the used fast neutron sources have an excellent agreement with the reference ones. Also, the accuracy of the unfolded energy spectra obtained using the GMDH is slightly better than those obtained from the DT. The results obtained in the present study have good accuracy in comparison with the previously published paper based on the logsig and tansig transfer functions.

Results on the neutron energy distribution measurements at the RECH-1 Chilean nuclear reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Aguilera, P., E-mail: paguilera87@gmail.com; Romero-Barrientos, J.; Universidad de Chile, Dpto. de Física, Facultad de Ciencias, Las Palmeras 3425, Nuñoa, Santiago

2016-07-07

Neutron activations experiments has been perform at the RECH-1 Chilean Nuclear Reactor to measure its neutron flux energy distribution. Samples of pure elements was activated to obtain the saturation activities for each reaction. Using - ray spectroscopy we identify and measure the activity of the reaction product nuclei, obtaining the saturation activities of 20 reactions. GEANT4 and MCNP was used to compute the self shielding factor to correct the cross section for each element. With the Expectation-Maximization algorithm (EM) we were able to unfold the neutron flux energy distribution at dry tube position, near the RECH-1 core. In this work,more » we present the unfolding results using the EM algorithm.« less

A study on fast digital discrimination of neutron and gamma-ray for improvement neutron emission profile measurementa)

PubMed Central

Uchida, Y.; Takada, E.; Fujisaki, A.; Isobe, M.; Shinohara, K.; Tomita, H.; Kawarabayashi, J.; Iguchi, T.

2014-01-01

Neutron and γ-ray (n-γ) discrimination with a digital signal processing system has been used to measure the neutron emission profile in magnetic confinement fusion devices. However, a sampling rate must be set low to extend the measurement time because the memory storage is limited. Time jitter decreases a discrimination quality due to a low sampling rate. As described in this paper, a new charge comparison method was developed. Furthermore, automatic n-γ discrimination method was examined using a probabilistic approach. Analysis results were investigated using the figure of merit. Results show that the discrimination quality was improved. Automatic discrimination was applied using the EM algorithm and k-means algorithm. PMID:25430297

Comparison between calculation and measured data on secondary neutron energy spectra by heavy ion reactions from different thick targets.

PubMed

Iwase, H; Wiegel, B; Fehrenbacher, G; Schardt, D; Nakamura, T; Niita, K; Radon, T

2005-01-01

Measured neutron energy fluences from high-energy heavy ion reactions through targets several centimeters to several hundred centimeters thick were compared with calculations made using the recently developed general-purpose particle and heavy ion transport code system (PHITS). It was confirmed that the PHITS represented neutron production by heavy ion reactions and neutron transport in thick shielding with good overall accuracy.

Effects of Hot-Spot Geometry on Backscattering and Down-Scattering Neutron Spectra

NASA Astrophysics Data System (ADS)

Mohamed, Z. L.; Mannion, O. M.; Forrest, C. J.; Knauer, J. P.; Anderson, K. S.; Radha, P. B.

2017-10-01

The measured neutron spectrum produced by a fusion experiment plays a key role in inferring observable quantities. One important observable is the areal density of an implosion, which is inferred by measuring the scattering of neutrons. This project seeks to use particle-transport simulations to model the effects of hot-spot geometry on backscattering and down-scattering neutron spectra along different lines of sight. Implosions similar to those conducted at the Laboratory of Laser Energetics are modeled by neutron transport through a DT plasma and a DT ice shell using the particle transport codes MCNP and IRIS. Effects of hot-spot geometry are obtained by ``detecting'' scattered neutrons along different lines of sight. This process is repeated for various hot-spot geometries representing known shape distortions between the hot spot and the shell. This material is based upon work supported by the Department of Energy National Nuclear Security Administration under Award Number DE-NA0001944.

Further Improvement of the RITS Code for Pulsed Neutron Bragg-edge Transmission Imaging

NASA Astrophysics Data System (ADS)

Sato, H.; Watanabe, K.; Kiyokawa, K.; Kiyanagi, R.; Hara, K. Y.; Kamiyama, T.; Furusaka, M.; Shinohara, T.; Kiyanagi, Y.

The RITS code is a unique and powerful tool for a whole Bragg-edge transmission spectrum fitting analysis. However, it has had two major problems. Therefore, we have proposed methods to overcome these problems. The first issue is the difference in the crystallite size values between the diffraction and the Bragg-edge analyses. We found the reason was a different definition of the crystal structure factor. It affects the crystallite size because the crystallite size is deduced from the primary extinction effect which depends on the crystal structure factor. As a result of algorithm change, crystallite sizes obtained by RITS drastically approached to crystallite sizes obtained by Rietveld analyses of diffraction data; from 155% to 110%. The second issue is correction of the effect of background neutrons scattered from a specimen. Through neutron transport simulation studies, we found that the background components consist of forward Bragg scattering, double backward Bragg scattering, and thermal diffuse scattering. RITS with the background correction function which was developed through the simulation studies could well reconstruct various simulated and experimental transmission spectra, but refined crystalline microstructural parameters were often distorted. Finally, it was recommended to reduce the background by improving experimental conditions.

Estimation of neutron spectrum in the low-level gamma spectroscopy system using unfolding procedure

DOE Office of Scientific and Technical Information (OSTI.GOV)

Knežević, D., E-mail: david.knezevic@df.uns.ac.rs; Jovančević, N.; Krmar, M.

2016-03-25

The radiation resulting from neutron interactions with Ge nuclei in active volume of HPGe detectors is one of the main concerns in low-level gamma spectroscopy measurements [1,2]. It is usually not possible to measure directly spectrum of neutrons which strike detector. This paper explore the possibility of estimation of neutron spectrum using measured activities of certain Ge(n,γ) and Ge(n,n’) reactions (obtained from low-level gamma measurements), available ENDF cross section data and unfolding procedures. In this work HPGe detector with passive shield made from commercial low background lead was used for the measurement. The most important objective of this study wasmore » to reconstruct muon induced neutron spectrum created in the shield of the HPGe detector. MAXED [3] and GRAVEL [4] algorithms for neutron spectra unfolding were used. The results of those two algorithms were compared and we analyzed the sensitivity of the unfolding procedure to the various input parameters.« less

A Monte Carlo model system for core analysis and epithermal neutron beam design at the Washington State University Radiation Center

DOE Office of Scientific and Technical Information (OSTI.GOV)

Burns, T.D. Jr.

1996-05-01

The Monte Carlo Model System (MCMS) for the Washington State University (WSU) Radiation Center provides a means through which core criticality and power distributions can be calculated, as well as providing a method for neutron and photon transport necessary for BNCT epithermal neutron beam design. The computational code used in this Model System is MCNP4A. The geometric capability of this Monte Carlo code allows the WSU system to be modeled very accurately. A working knowledge of the MCNP4A neutron transport code increases the flexibility of the Model System and is recommended, however, the eigenvalue/power density problems can be run withmore » little direct knowledge of MCNP4A. Neutron and photon particle transport require more experience with the MCNP4A code. The Model System consists of two coupled subsystems; the Core Analysis and Source Plane Generator Model (CASP), and the BeamPort Shell Particle Transport Model (BSPT). The CASP Model incorporates the S({alpha}, {beta}) thermal treatment, and is run as a criticality problem yielding, the system eigenvalue (k{sub eff}), the core power distribution, and an implicit surface source for subsequent particle transport in the BSPT Model. The BSPT Model uses the source plane generated by a CASP run to transport particles through the thermal column beamport. The user can create filter arrangements in the beamport and then calculate characteristics necessary for assessing the BNCT potential of the given filter want. Examples of the characteristics to be calculated are: neutron fluxes, neutron currents, fast neutron KERMAs and gamma KERMAs. The MCMS is a useful tool for the WSU system. Those unfamiliar with the MCNP4A code can use the MCMS transparently for core analysis, while more experienced users will find the particle transport capabilities very powerful for BNCT filter design.« less

STABILITY OF THE NEUTRON DOSE DETERMINATION ALGORITHM FOR PERSONAL NEUTRON DOSEMETERS AT DIFFERENT RADON GAS EXPOSURES.

PubMed

Mayer, Sabine; Boschung, Markus; Butterweck, Gernot; Assenmacher, Frank; Hohmann, Eike

2016-09-01

Since 2008 the Paul Scherrer Institute (PSI) has been using a microscope-based automatic scanning system for assessing personal neutron doses with a dosemeter based on PADC. This scanning system, known as TASLImage, includes a comprehensive characterisation of tracks. The distributions of several specific track characteristics such as size, shape and optical density are compared with a reference set to discriminate tracks of alpha particles and non-track background. Due to the dosemeter design at PSI, it is anticipated that radon should not significantly contribute to the creation of additional tracks in the PADC detector. The present study tests the stability of the neutron dose determination algorithm of the personal neutron dosemeter system in operation at PSI at different radon gas exposures. © The Author 2015. Published by Oxford University Press. All rights reserved. For Permissions, please email: journals.permissions@oup.com.

Container Inspection Utilizing 14 MeV Neutrons

NASA Astrophysics Data System (ADS)

Valkovic, Vladivoj; Sudac, Davorin; Nad, Karlo; Obhodas, Jasmina

2016-06-01

A proposal for an autonomous and flexible ship container inspection system is presented. This could be accomplished by the incorporation of an inspection system on various container transportation devices (straddle carriers, yard gentry cranes, automated guided vehicles, trailers). The configuration is terminal specific and it should be defined by the container terminal operator. This enables that no part of the port operational area is used for inspection. The inspection scenario includes container transfer from ship to transportation device with the inspection unit mounted on it. The inspection is performed during actual container movement to the container location. A neutron generator without associated alpha particle detection is used. This allows the use of higher neutron intensities (5 × 109 - 1010 n/s in 4π). The inspected container is stationary in the “inspection position” on the transportation device while the “inspection unit” moves along its side. The following analytical methods will be used simultaneously: neutron radiography, X-ray radiography, neutron activation analysis, (n, γ) and (n,n'γ) reactions, neutron absorption. and scattering, X-ray backscattering. The neutron techniques will utilize “smart collimators” for neutrons and gamma rays, both emitted and detected. The inspected voxel is defined by the intersection of the neutron generator and the detectors solid angles. The container inspection protocol is based on identification of discrepancies between the cargo manifest, elemental “fingerprint” and radiography profiles. In addition, the information on container weight is obtained during the container transport and screening by measuring of density of material in the container.

Leveraging Anderson Acceleration for improved convergence of iterative solutions to transport systems

DOE Office of Scientific and Technical Information (OSTI.GOV)

Willert, Jeffrey; Taitano, William T.; Knoll, Dana

In this note we demonstrate that using Anderson Acceleration (AA) in place of a standard Picard iteration can not only increase the convergence rate but also make the iteration more robust for two transport applications. We also compare the convergence acceleration provided by AA to that provided by moment-based acceleration methods. Additionally, we demonstrate that those two acceleration methods can be used together in a nested fashion. We begin by describing the AA algorithm. At this point, we will describe two application problems, one from neutronics and one from plasma physics, on which we will apply AA. We provide computationalmore » results which highlight the benefits of using AA, namely that we can compute solutions using fewer function evaluations, larger time-steps, and achieve a more robust iteration.« less

QPO observations related to neutron star equations of state

NASA Astrophysics Data System (ADS)

Stuchlik, Zdenek; Urbanec, Martin; Török, Gabriel; Bakala, Pavel; Cermak, Petr

We apply a genetic algorithm method for selection of neutron star models relating them to the resonant models of the twin peak quasiperiodic oscillations observed in the X-ray neutron star binary systems. It was suggested that pairs of kilo-hertz peaks in the X-ray Fourier power density spectra of some neutron stars reflect a non-linear resonance between two modes of accretion disk oscillations. We investigate this concept for a specific neutron star source. Each neutron star model is characterized by the equation of state (EOS), rotation frequency Ω and central energy density ρc . These determine the spacetime structure governing geodesic motion and position dependent radial and vertical epicyclic oscillations related to the stable circular geodesics. Particular kinds of resonances (KR) between the oscillations with epicyclic frequencies, or the frequencies derived from them, can take place at special positions assigned ambiguously to the spacetime structure. The pairs of resonant eigenfrequencies relevant to those positions are therefore fully given by KR,ρc , Ω, EOS and can be compared to the observationally determined pairs of eigenfrequencies in order to eliminate the unsatisfactory sets (KR,ρc , Ω, EOS). For the elimination we use the advanced genetic algorithm. Genetic algorithm comes out from the method of natural selection when subjects with the best adaptation to assigned conditions have most chances to survive. The chosen genetic algorithm with sexual reproduction contains one chromosome with restricted lifetime, uniform crossing and genes of type 3/3/5. For encryption of physical description (KR,ρ, Ω, EOS) into chromosome we used Gray code. As a fitness function we use correspondence between the observed and calculated pairs of eigenfrequencies.

Neutron star equation of state and QPO observations

NASA Astrophysics Data System (ADS)

Urbanec, Martin; Stuchlík, Zdeněk; Török, Gabriel; Bakala, Pavel; Čermák, Petr

2007-12-01

Assuming a resonant origin of the twin peak quasiperiodic oscillations observed in the X-ray neutron star binary systems, we apply a genetic algorithm method for selection of neutron star models. It was suggested that pairs of kilohertz peaks in the X-ray Fourier power density spectra of some neutron stars reflect a non-linear resonance between two modes of accretion disk oscillations. We investigate this concept for a specific neutron star source. Each neutron star model is characterized by the equation of state (EOS), rotation frequency Ω and central energy density rho_{c}. These determine the spacetime structure governing geodesic motion and position dependent radial and vertical epicyclic oscillations related to the stable circular geodesics. Particular kinds of resonances (KR) between the oscillations with epicyclic frequencies, or the frequencies derived from them, can take place at special positions assigned ambiguously to the spacetime structure. The pairs of resonant eigenfrequencies relevant to those positions are therefore fully given by KR, rho_{c}, Ω, EOS and can be compared to the observationally determined pairs of eigenfrequencies in order to eliminate the unsatisfactory sets (KR, rho_{c}, Ω, EOS). For the elimination we use the advanced genetic algorithm. Genetic algorithm comes out from the method of natural selection when subjects with the best adaptation to assigned conditions have most chances to survive. The chosen genetic algorithm with sexual reproduction contains one chromosome with restricted lifetime, uniform crossing and genes of type 3/3/5. For encryption of physical description (KR, rho_{c}, Ω, EOS) into the chromosome we use the Gray code. As a fitness function we use correspondence between the observed and calculated pairs of eigenfrequencies.

An assessment of coupling algorithms for nuclear reactor core physics simulations

DOE PAGES

Hamilton, Steven; Berrill, Mark; Clarno, Kevin; ...

2016-04-01

This paper evaluates the performance of multiphysics coupling algorithms applied to a light water nuclear reactor core simulation. The simulation couples the k-eigenvalue form of the neutron transport equation with heat conduction and subchannel flow equations. We compare Picard iteration (block Gauss–Seidel) to Anderson acceleration and multiple variants of preconditioned Jacobian-free Newton–Krylov (JFNK). The performance of the methods are evaluated over a range of energy group structures and core power levels. A novel physics-based approximation to a Jacobian-vector product has been developed to mitigate the impact of expensive on-line cross section processing steps. Furthermore, numerical simulations demonstrating the efficiency ofmore » JFNK and Anderson acceleration relative to standard Picard iteration are performed on a 3D model of a nuclear fuel assembly. Both criticality (k-eigenvalue) and critical boron search problems are considered.« less

An assessment of coupling algorithms for nuclear reactor core physics simulations

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hamilton, Steven; Berrill, Mark; Clarno, Kevin

This paper evaluates the performance of multiphysics coupling algorithms applied to a light water nuclear reactor core simulation. The simulation couples the k-eigenvalue form of the neutron transport equation with heat conduction and subchannel flow equations. We compare Picard iteration (block Gauss–Seidel) to Anderson acceleration and multiple variants of preconditioned Jacobian-free Newton–Krylov (JFNK). The performance of the methods are evaluated over a range of energy group structures and core power levels. A novel physics-based approximation to a Jacobian-vector product has been developed to mitigate the impact of expensive on-line cross section processing steps. Furthermore, numerical simulations demonstrating the efficiency ofmore » JFNK and Anderson acceleration relative to standard Picard iteration are performed on a 3D model of a nuclear fuel assembly. Both criticality (k-eigenvalue) and critical boron search problems are considered.« less

An assessment of coupling algorithms for nuclear reactor core physics simulations

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hamilton, Steven, E-mail: hamiltonsp@ornl.gov; Berrill, Mark, E-mail: berrillma@ornl.gov; Clarno, Kevin, E-mail: clarnokt@ornl.gov

This paper evaluates the performance of multiphysics coupling algorithms applied to a light water nuclear reactor core simulation. The simulation couples the k-eigenvalue form of the neutron transport equation with heat conduction and subchannel flow equations. We compare Picard iteration (block Gauss–Seidel) to Anderson acceleration and multiple variants of preconditioned Jacobian-free Newton–Krylov (JFNK). The performance of the methods are evaluated over a range of energy group structures and core power levels. A novel physics-based approximation to a Jacobian-vector product has been developed to mitigate the impact of expensive on-line cross section processing steps. Numerical simulations demonstrating the efficiency of JFNKmore » and Anderson acceleration relative to standard Picard iteration are performed on a 3D model of a nuclear fuel assembly. Both criticality (k-eigenvalue) and critical boron search problems are considered.« less

BOXER: Fine-flux Cross Section Condensation, 2D Few Group Diffusion and Transport Burnup Calculations

DOE Office of Scientific and Technical Information (OSTI.GOV)

2010-02-01

Neutron transport, calculation of multiplication factor and neutron fluxes in 2-D configurations: cell calculations, 2-D diffusion and transport, and burnup. Preparation of a cross section library for the code BOXER from a basic library in ENDF/B format (ETOBOX).

Probabilistic n/γ discrimination with robustness against outliers for use in neutron profile monitors

NASA Astrophysics Data System (ADS)

Uchida, Y.; Takada, E.; Fujisaki, A.; Kikuchi, T.; Ogawa, K.; Isobe, M.

2017-08-01

A method to stochastically discriminate neutron and γ-ray signals measured with a stilbene organic scintillator is proposed. Each pulse signal was stochastically categorized into two groups: neutron and γ-ray. In previous work, the Expectation Maximization (EM) algorithm was used with the assumption that the measured data followed a Gaussian mixture distribution. It was shown that probabilistic discrimination between these groups is possible. Moreover, by setting the initial parameters for the Gaussian mixture distribution with a k-means algorithm, the possibility of automatic discrimination was demonstrated. In this study, the Student's t-mixture distribution was used as a probabilistic distribution with the EM algorithm to improve the robustness against the effect of outliers caused by pileup of the signals. To validate the proposed method, the figures of merit (FOMs) were compared for the EM algorithm assuming a t-mixture distribution and a Gaussian mixture distribution. The t-mixture distribution resulted in an improvement of the FOMs compared with the Gaussian mixture distribution. The proposed data processing technique is a promising tool not only for neutron and γ-ray discrimination in fusion experiments but also in other fields, for example, homeland security, cancer therapy with high energy particles, nuclear reactor decommissioning, pattern recognition, and so on.

Validating the performance of correlated fission multiplicity implementation in radiation transport codes with subcritical neutron multiplication benchmark experiments

DOE PAGES

Arthur, Jennifer; Bahran, Rian; Hutchinson, Jesson; ...

2018-06-14

Historically, radiation transport codes have uncorrelated fission emissions. In reality, the particles emitted by both spontaneous and induced fissions are correlated in time, energy, angle, and multiplicity. This work validates the performance of various current Monte Carlo codes that take into account the underlying correlated physics of fission neutrons, specifically neutron multiplicity distributions. The performance of 4 Monte Carlo codes - MCNP®6.2, MCNP®6.2/FREYA, MCNP®6.2/CGMF, and PoliMi - was assessed using neutron multiplicity benchmark experiments. In addition, MCNP®6.2 simulations were run using JEFF-3.2 and JENDL-4.0, rather than ENDF/B-VII.1, data for 239Pu and 240Pu. The sensitive benchmark parameters that in this workmore » represent the performance of each correlated fission multiplicity Monte Carlo code include the singles rate, the doubles rate, leakage multiplication, and Feynman histograms. Although it is difficult to determine which radiation transport code shows the best overall performance in simulating subcritical neutron multiplication inference benchmark measurements, it is clear that correlations exist between the underlying nuclear data utilized by (or generated by) the various codes, and the correlated neutron observables of interest. This could prove useful in nuclear data validation and evaluation applications, in which a particular moment of the neutron multiplicity distribution is of more interest than the other moments. It is also quite clear that, because transport is handled by MCNP®6.2 in 3 of the 4 codes, with the 4th code (PoliMi) being based on an older version of MCNP®, the differences in correlated neutron observables of interest are most likely due to the treatment of fission event generation in each of the different codes, as opposed to the radiation transport.« less

Validating the performance of correlated fission multiplicity implementation in radiation transport codes with subcritical neutron multiplication benchmark experiments

DOE Office of Scientific and Technical Information (OSTI.GOV)

Arthur, Jennifer; Bahran, Rian; Hutchinson, Jesson

Historically, radiation transport codes have uncorrelated fission emissions. In reality, the particles emitted by both spontaneous and induced fissions are correlated in time, energy, angle, and multiplicity. This work validates the performance of various current Monte Carlo codes that take into account the underlying correlated physics of fission neutrons, specifically neutron multiplicity distributions. The performance of 4 Monte Carlo codes - MCNP®6.2, MCNP®6.2/FREYA, MCNP®6.2/CGMF, and PoliMi - was assessed using neutron multiplicity benchmark experiments. In addition, MCNP®6.2 simulations were run using JEFF-3.2 and JENDL-4.0, rather than ENDF/B-VII.1, data for 239Pu and 240Pu. The sensitive benchmark parameters that in this workmore » represent the performance of each correlated fission multiplicity Monte Carlo code include the singles rate, the doubles rate, leakage multiplication, and Feynman histograms. Although it is difficult to determine which radiation transport code shows the best overall performance in simulating subcritical neutron multiplication inference benchmark measurements, it is clear that correlations exist between the underlying nuclear data utilized by (or generated by) the various codes, and the correlated neutron observables of interest. This could prove useful in nuclear data validation and evaluation applications, in which a particular moment of the neutron multiplicity distribution is of more interest than the other moments. It is also quite clear that, because transport is handled by MCNP®6.2 in 3 of the 4 codes, with the 4th code (PoliMi) being based on an older version of MCNP®, the differences in correlated neutron observables of interest are most likely due to the treatment of fission event generation in each of the different codes, as opposed to the radiation transport.« less

Preliminary On-Orbit Neutron Dose Equivalent and Energy Spectrum Results from the ISS-RAD Fast Neutron Detector (FND)

NASA Technical Reports Server (NTRS)

Semones, Edward; Leitgab, Martin

2016-01-01

The ISS-RAD instrument was activated on ISS on February 1st, 2016. Integrated in ISS-RAD, the Fast Neutron Detector (FND) performs, for the first time on ISS, routine and precise direct neutron measurements between 0.5 and 8 MeV. Preliminary results for neutron dose equivalent and neutron flux energy distributions from online/on-board algorithms and offline ground analyses will be shown, along with comparisons to simulated data and previously measured neutron spectral data. On-orbit data quality and pre-launch analysis validation results will be discussed as well.

Container inspection in the port container terminal by using 14 MeV neutrons

DOE Office of Scientific and Technical Information (OSTI.GOV)

Valkovic, Vladivoj; Kvinticka 62, Zagreb; Sudac, Davorin

A proposal for an autonomous and flexible ship container inspection system is presented. This could be accomplished by the incorporation of inspection system on the container transportation devices (straddle carriers, yard gentry cranes automated guided vehicles, trailers). This configuration is terminal specific and it will be decided by container terminal operator. In such a way no part of port operational area will be used for inspection. The inspection scenario will include container transfer from ship to transportation device with inspection unit mounted on it, inspection during container movement to the container location. A neutron generator without associated alpha particle detectionmore » will be used. This will allow the use of higher neutron intensity (5x10{sup 9} - 10{sup 10} n/s in 4π). The inspected container will be stationary in the 'inspection position' on the transportation device while the 'inspection unit' will move along its side. Following analytical methods will be used simultaneously: neutron radiography, X-ray radiography, neutron activation analysis, (n,γ) and (n,n'γ) reactions, neutron absorption, and scattering, X-ray backscattering, Neutron techniques will take the advantage of using 'smart collimators' for neutrons and gammas, both emitted and detected. The inspected voxel will be defined by intersections/union of neutron generator and detectors solid angles. The container inspection protocol will be based on identification of discrepancies between its cargo manifest and its elemental 'fingerprint' and radiography profiles. In addition, the information on container weight will be obtained during the container transport and foreseen screening from the measurement of density of material in the container. (authors)« less

Stability and accuracy of 3D neutron transport simulations using the 2D/1D method in MPACT

DOE Office of Scientific and Technical Information (OSTI.GOV)

Collins, Benjamin, E-mail: collinsbs@ornl.gov; Stimpson, Shane, E-mail: stimpsonsg@ornl.gov; Kelley, Blake W., E-mail: kelleybl@umich.edu

2016-12-01

A consistent “2D/1D” neutron transport method is derived from the 3D Boltzmann transport equation, to calculate fuel-pin-resolved neutron fluxes for realistic full-core Pressurized Water Reactor (PWR) problems. The 2D/1D method employs the Method of Characteristics to discretize the radial variables and a lower order transport solution to discretize the axial variable. This paper describes the theory of the 2D/1D method and its implementation in the MPACT code, which has become the whole-core deterministic neutron transport solver for the Consortium for Advanced Simulations of Light Water Reactors (CASL) core simulator VERA-CS. Several applications have been performed on both leadership-class and industry-classmore » computing clusters. Results are presented for whole-core solutions of the Watts Bar Nuclear Power Station Unit 1 and compared to both continuous-energy Monte Carlo results and plant data.« less

Stability and accuracy of 3D neutron transport simulations using the 2D/1D method in MPACT

DOE PAGES

Collins, Benjamin; Stimpson, Shane; Kelley, Blake W.; ...

2016-08-25

We derived a consistent “2D/1D” neutron transport method from the 3D Boltzmann transport equation, to calculate fuel-pin-resolved neutron fluxes for realistic full-core Pressurized Water Reactor (PWR) problems. The 2D/1D method employs the Method of Characteristics to discretize the radial variables and a lower order transport solution to discretize the axial variable. Our paper describes the theory of the 2D/1D method and its implementation in the MPACT code, which has become the whole-core deterministic neutron transport solver for the Consortium for Advanced Simulations of Light Water Reactors (CASL) core simulator VERA-CS. We also performed several applications on both leadership-class and industry-classmore » computing clusters. Results are presented for whole-core solutions of the Watts Bar Nuclear Power Station Unit 1 and compared to both continuous-energy Monte Carlo results and plant data.« less

Coupled Neutron Transport for HZETRN

NASA Technical Reports Server (NTRS)

Slaba, Tony C.; Blattnig, Steve R.

2009-01-01

Exposure estimates inside space vehicles, surface habitats, and high altitude aircrafts exposed to space radiation are highly influenced by secondary neutron production. The deterministic transport code HZETRN has been identified as a reliable and efficient tool for such studies, but improvements to the underlying transport models and numerical methods are still necessary. In this paper, the forward-backward (FB) and directionally coupled forward-backward (DC) neutron transport models are derived, numerical methods for the FB model are reviewed, and a computationally efficient numerical solution is presented for the DC model. Both models are compared to the Monte Carlo codes HETC-HEDS, FLUKA, and MCNPX, and the DC model is shown to agree closely with the Monte Carlo results. Finally, it is found in the development of either model that the decoupling of low energy neutrons from the light particle transport procedure adversely affects low energy light ion fluence spectra and exposure quantities. A first order correction is presented to resolve the problem, and it is shown to be both accurate and efficient.

Neutron Transport Models and Methods for HZETRN and Coupling to Low Energy Light Ion Transport

NASA Technical Reports Server (NTRS)

Blattnig, S.R.; Slaba, T.C.; Heinbockel, J.H.

2008-01-01

Exposure estimates inside space vehicles, surface habitats, and high altitude aircraft exposed to space radiation are highly influenced by secondary neutron production. The deterministic transport code HZETRN has been identified as a reliable and efficient tool for such studies, but improvements to the underlying transport models and numerical methods are still necessary. In this paper, the forward-backward (FB) and directionally coupled forward-backward (DC) neutron transport models are derived, numerical methods for the FB model are reviewed, and a computationally efficient numerical solution is presented for the DC model. Both models are compared to the Monte Carlo codes HETCHEDS and FLUKA, and the DC model is shown to agree closely with the Monte Carlo results. Finally, it is found in the development of either model that the decoupling of low energy neutrons from the light ion (A<4) transport procedure adversely affects low energy light ion fluence spectra and exposure quantities. A first order correction is presented to resolve the problem, and it is shown to be both accurate and efficient.

Digital pile-up rejection for plutonium experiments with solution-grown stilbene

NASA Astrophysics Data System (ADS)

Bourne, M. M.; Clarke, S. D.; Paff, M.; DiFulvio, A.; Norsworthy, M.; Pozzi, S. A.

2017-01-01

A solution-grown stilbene detector was used in several experiments with plutonium samples including plutonium oxide, mixed oxide, and plutonium metal samples. Neutrons from different reactions and plutonium isotopes are accompanied by numerous gamma rays especially by the 59-keV gamma ray of 241Am. Identifying neutrons correctly is important for nuclear nonproliferation applications and makes neutron/gamma discrimination and pile-up rejection necessary. Each experimental dataset is presented with and without pile-up filtering using a previously developed algorithm. The experiments were simulated using MCNPX-PoliMi, a Monte Carlo code designed to accurately model scintillation detector response. Collision output from MCNPX-PoliMi was processed using the specialized MPPost post-processing code to convert neutron energy depositions event-by-event into light pulses. The model was compared to experimental data after pulse-shape discrimination identified waveforms as gamma ray or neutron interactions. We show that the use of the digital pile-up rejection algorithm allows for accurate neutron counting with stilbene to within 2% even when not using lead shielding.

The potential of detecting intermediate-scale biomass and canopy interception in a coniferous forest using cosmic-ray neutron intensity measurements and neutron transport modeling

NASA Astrophysics Data System (ADS)

Andreasen, M.; Looms, M. C.; Bogena, H. R.; Desilets, D.; Zreda, M. G.; Sonnenborg, T. O.; Jensen, K. H.

2014-12-01

The water stored in the various compartments of the terrestrial ecosystem (in snow, canopy interception, soil and litter) controls the exchange of the water and energy between the land surface and the atmosphere. Therefore, measurements of the water stored within these pools are critical for the prediction of e.g. evapotranspiration and groundwater recharge. The detection of cosmic-ray neutron intensity is a novel non-invasive method for the quantification of continuous intermediate-scale soil moisture. The footprint of the cosmic-ray neutron probe is a hemisphere of a few hectometers and subsurface depths of 10-70 cm depending on wetness. The cosmic-ray neutron method offers measurements at a scale between the point-scale measurements and large-scale satellite retrievals. The cosmic-ray neutron intensity is inversely correlated to the hydrogen stored within the footprint. Overall soil moisture represents the largest pool of hydrogen and changes in the soil moisture clearly affect the cosmic-ray neutron signal. However, the neutron intensity is also sensitive to variations of hydrogen in snow, canopy interception and biomass offering the potential to determine water content in such pools from the signal. In this study we tested the potential of determining canopy interception and biomass using cosmic-ray neutron intensity measurements within the framework of the Danish Hydrologic Observatory (HOBE) and the Terrestrial Environmental Observatories (TERENO). Continuous measurements at the ground and the canopy level, along with profile measurements were conducted at towers at forest field sites. Field experiments, including shielding the cosmic-ray neutron probes with cadmium foil (to remove lower-energy neutrons) and measuring reference intensity rates at complete water saturated conditions (on the sea close to the HOBE site), were further conducted to obtain an increased understanding of the physics controlling the cosmic-ray neutron transport and the equipment used. Additionally, neutron transport modeling, using the extended version of the Monte Carlo N-Particle Transport Code, was conducted. The responses of the reference condition, different amounts of biomass, soil moisture and canopy interception on the cosmic-ray neutron intensity were simulated and compared to the measurements.

Adaptive algorithms of position and energy reconstruction in Anger-camera type detectors: experimental data processing in ANTS

NASA Astrophysics Data System (ADS)

Morozov, A.; Defendi, I.; Engels, R.; Fraga, F. A. F.; Fraga, M. M. F. R.; Gongadze, A.; Guerard, B.; Jurkovic, M.; Kemmerling, G.; Manzin, G.; Margato, L. M. S.; Niko, H.; Pereira, L.; Petrillo, C.; Peyaud, A.; Piscitelli, F.; Raspino, D.; Rhodes, N. J.; Sacchetti, F.; Schooneveld, E. M.; Solovov, V.; Van Esch, P.; Zeitelhack, K.

2013-05-01

The software package ANTS (Anger-camera type Neutron detector: Toolkit for Simulations), developed for simulation of Anger-type gaseous detectors for thermal neutron imaging was extended to include a module for experimental data processing. Data recorded with a sensor array containing up to 100 photomultiplier tubes (PMT) or silicon photomultipliers (SiPM) in a custom configuration can be loaded and the positions and energies of the events can be reconstructed using the Center-of-Gravity, Maximum Likelihood or Least Squares algorithm. A particular strength of the new module is the ability to reconstruct the light response functions and relative gains of the photomultipliers from flood field illumination data using adaptive algorithms. The performance of the module is demonstrated with simulated data generated in ANTS and experimental data recorded with a 19 PMT neutron detector. The package executables are publicly available at http://coimbra.lip.pt/~andrei/

Measurement of high-energy neutron flux above ground utilizing a spallation based multiplicity technique

DOE PAGES

Roecker, Caleb; Bernstein, Adam; Marleau, Peter; ...

2016-11-14

Cosmogenic high-energy neutrons are a ubiquitous, difficult to shield, poorly measured background. Above ground the high-energy neutron energy-dependent flux has been measured, with significantly varying results. Below ground, high-energy neutron fluxes are largely unmeasured. Here we present a reconstruction algorithm to unfold the incident neutron energy-dependent flux measured using the Multiplicity and Recoil Spectrometer (MARS), simulated test cases to verify the algorithm, and provide a new measurement of the above ground high-energy neutron energy-dependent flux with a detailed systematic uncertainty analysis. Uncertainty estimates are provided based upon the measurement statistics, the incident angular distribution, the surrounding environment of the Montemore » Carlo model, and the MARS triggering efficiency. Quantified systematic uncertainty is dominated by the assumed incident neutron angular distribution and surrounding environment of the Monte Carlo model. The energy-dependent neutron flux between 90 MeV and 400 MeV is reported. Between 90 MeV and 250 MeV the MARS results are comparable to previous Bonner sphere measurements. Over the total energy regime measured, the MARS result are located within the span of previous measurements. Lastly, these results demonstrate the feasibility of future below ground measurements with MARS.« less

Measurement of high-energy neutron flux above ground utilizing a spallation based multiplicity technique

DOE Office of Scientific and Technical Information (OSTI.GOV)

Roecker, Caleb; Bernstein, Adam; Marleau, Peter

Cosmogenic high-energy neutrons are a ubiquitous, difficult to shield, poorly measured background. Above ground the high-energy neutron energy-dependent flux has been measured, with significantly varying results. Below ground, high-energy neutron fluxes are largely unmeasured. Here we present a reconstruction algorithm to unfold the incident neutron energy-dependent flux measured using the Multiplicity and Recoil Spectrometer (MARS), simulated test cases to verify the algorithm, and provide a new measurement of the above ground high-energy neutron energy-dependent flux with a detailed systematic uncertainty analysis. Uncertainty estimates are provided based upon the measurement statistics, the incident angular distribution, the surrounding environment of the Montemore » Carlo model, and the MARS triggering efficiency. Quantified systematic uncertainty is dominated by the assumed incident neutron angular distribution and surrounding environment of the Monte Carlo model. The energy-dependent neutron flux between 90 MeV and 400 MeV is reported. Between 90 MeV and 250 MeV the MARS results are comparable to previous Bonner sphere measurements. Over the total energy regime measured, the MARS result are located within the span of previous measurements. Lastly, these results demonstrate the feasibility of future below ground measurements with MARS.« less

Transportable, Low-Dose Active Fast-Neutron Imaging

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mihalczo, John T.; Wright, Michael C.; McConchie, Seth M.

2017-08-01

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

A comparison of the COG and MCNP codes in computational neutron capture therapy modeling, Part I: boron neutron capture therapy models.

PubMed

Culbertson, C N; Wangerin, K; Ghandourah, E; Jevremovic, T

2005-08-01

The goal of this study was to evaluate the COG Monte Carlo radiation transport code, developed and tested by Lawrence Livermore National Laboratory, for neutron capture therapy related modeling. A boron neutron capture therapy model was analyzed comparing COG calculational results to results from the widely used MCNP4B (Monte Carlo N-Particle) transport code. The approach for computing neutron fluence rate and each dose component relevant in boron neutron capture therapy is described, and calculated values are shown in detail. The differences between the COG and MCNP predictions are qualified and quantified. The differences are generally small and suggest that the COG code can be applied for BNCT research related problems.

Three-Dimensional General Relativistic Monte Carlo Neutrino Transport in Neutron Star Mergers

NASA Astrophysics Data System (ADS)

Richers, Sherwood; Radice, David

2018-06-01

How neutrinos interact with the debris ejected from merging neutron stars determines how much matter escapes, how hot the matter is, and the relative amounts of neutrons and protons. This makes understanding neutrino irradiation of ejected matter a necessary part of interpreting recent and future observations of so-called "kilonovae" to determine whether neutron star mergers can be the origin of heavy elements in the universe. I will discuss a new Monte Carlo method for simulating neutrino transport in these highly relativistic, multi-dimensional environments. I will use this tool to estimate how well approximate transport methods capture the neutrino irradiation and propose improvements to approximate methods that will aid in accurate modeling and interpretation of kilonovae.

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hep, J.; Konecna, A.; Krysl, V.

2011-07-01

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

PROTEUS-SN User Manual

DOE Office of Scientific and Technical Information (OSTI.GOV)

Shemon, Emily R.; Smith, Micheal A.; Lee, Changho

2016-02-16

PROTEUS-SN is a three-dimensional, highly scalable, high-fidelity neutron transport code developed at Argonne National Laboratory. The code is applicable to all spectrum reactor transport calculations, particularly those in which a high degree of fidelity is needed either to represent spatial detail or to resolve solution gradients. PROTEUS-SN solves the second order formulation of the transport equation using the continuous Galerkin finite element method in space, the discrete ordinates approximation in angle, and the multigroup approximation in energy. PROTEUS-SN’s parallel methodology permits the efficient decomposition of the problem by both space and angle, permitting large problems to run efficiently on hundredsmore » of thousands of cores. PROTEUS-SN can also be used in serial or on smaller compute clusters (10’s to 100’s of cores) for smaller homogenized problems, although it is generally more computationally expensive than traditional homogenized methodology codes. PROTEUS-SN has been used to model partially homogenized systems, where regions of interest are represented explicitly and other regions are homogenized to reduce the problem size and required computational resources. PROTEUS-SN solves forward and adjoint eigenvalue problems and permits both neutron upscattering and downscattering. An adiabatic kinetics option has recently been included for performing simple time-dependent calculations in addition to standard steady state calculations. PROTEUS-SN handles void and reflective boundary conditions. Multigroup cross sections can be generated externally using the MC2-3 fast reactor multigroup cross section generation code or internally using the cross section application programming interface (API) which can treat the subgroup or resonance table libraries. PROTEUS-SN is written in Fortran 90 and also includes C preprocessor definitions. The code links against the PETSc, METIS, HDF5, and MPICH libraries. It optionally links against the MOAB library and is a part of the SHARP multi-physics suite for coupled multi-physics analysis of nuclear reactors. This user manual describes how to set up a neutron transport simulation with the PROTEUS-SN code. A companion methodology manual describes the theory and algorithms within PROTEUS-SN.« less

Application of Neutron Tomography in Culture Heritage research.

PubMed

Mongy, T

2014-02-01

Neutron Tomography (NT) investigation of Culture Heritages (CH) is an efficient tool for understanding the culture of ancient civilizations. Neutron imaging (NI) is a-state-of-the-art non-destructive tool in the area of CH and plays an important role in the modern archeology. The NI technology can be widely utilized in the field of elemental analysis. At Egypt Second Research Reactor (ETRR-2), a collimated Neutron Radiography (NR) beam is employed for neutron imaging purposes. A digital CCD camera is utilized for recording the beam attenuation in the sample. This helps for the detection of hidden objects and characterization of material properties. Research activity can be extended to use computer software for quantitative neutron measurement. Development of image processing algorithms can be used to obtain high quality images. In this work, full description of ETRR-2 was introduced with up to date neutron imaging system as well. Tomographic investigation of a clay forged artifact represents CH object was studied by neutron imaging methods in order to obtain some hidden information and highlight some attractive quantitative measurements. Computer software was used for imaging processing and enhancement. Also the Astra Image 3.0 Pro software was employed for high precise measurements and imaging enhancement using advanced algorithms. This work increased the effective utilization of the ETRR-2 Neutron Radiography/Tomography (NR/T) technique in Culture Heritages activities. © 2013 Elsevier Ltd. All rights reserved.

Treating electron transport in MCNP{sup trademark}

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hughes, H.G.

1996-12-31

The transport of electrons and other charged particles is fundamentally different from that of neutrons and photons. A neutron, in aluminum slowing down from 0.5 MeV to 0.0625 MeV will have about 30 collisions; a photon will have fewer than ten. An electron with the same energy loss will undergo 10{sup 5} individual interactions. This great increase in computational complexity makes a single- collision Monte Carlo approach to electron transport unfeasible for many situations of practical interest. Considerable theoretical work has been done to develop a variety of analytic and semi-analytic multiple-scattering theories for the transport of charged particles. Themore » theories used in the algorithms in MCNP are the Goudsmit-Saunderson theory for angular deflections, the Landau an theory of energy-loss fluctuations, and the Blunck-Leisegang enhancements of the Landau theory. In order to follow an electron through a significant energy loss, it is necessary to break the electron`s path into many steps. These steps are chosen to be long enough to encompass many collisions (so that multiple-scattering theories are valid) but short enough that the mean energy loss in any one step is small (for the approximations in the multiple-scattering theories). The energy loss and angular deflection of the electron during each step can then be sampled from probability distributions based on the appropriate multiple- scattering theories. This subsumption of the effects of many individual collisions into single steps that are sampled probabilistically constitutes the ``condensed history`` Monte Carlo method. This method is exemplified in the ETRAN series of electron/photon transport codes. The ETRAN codes are also the basis for the Integrated TIGER Series, a system of general-purpose, application-oriented electron/photon transport codes. The electron physics in MCNP is similar to that of the Integrated TIGER Series.« less

Heat transport system

DOEpatents

Harkness, S.D.

A falling bed of ceramic particles receives neutron irradiation from a neutron-producing plasma and thereby transports energy as heat from the plasma to a heat exchange location where the ceramic particles are cooled by a gas flow. The cooled ceramic particles are elevated to a location from which they may again pass by gravity through the region where they are exposed to neutron radiation. Ceramic particles of alumina, magnesia, silica and combinations of these materials are contemplated as high-temperature materials that will accept energy from neutron irradiation. Separate containers of material incorporating lithium are exposed to the neutron flux for the breeding of tritium that may subsequently be used in neutron-producing reactions. The falling bed of ceramic particles includes velocity partitioning between compartments near to the neutron-producing plasma and compartments away from the plasma to moderate the maximum temperature in the bed.

Heat transport system

DOEpatents

Harkness, Samuel D.

1982-01-01

A falling bed of ceramic particles receives neutron irradiation from a neutron-producing plasma and thereby transports energy as heat from the plasma to a heat exchange location where the ceramic particles are cooled by a gas flow. The cooled ceramic particles are elevated to a location from which they may again pass by gravity through the region where they are exposed to neutron radiation. Ceramic particles of alumina, magnesia, silica and combinations of these materials are contemplated as high-temperature materials that will accept energy from neutron irradiation. Separate containers of material incorporating lithium are exposed to the neutron flux for the breeding of tritium that may subsequently be used in neutron-producing reactions. The falling bed of ceramic particles includes velocity partitioning between compartments near to the neutron-producing plasma and compartments away from the plasma to moderate the maximum temperature in the bed.

Radiation Protection Effectiveness of Polymeric Based Shielding Materials at Low Earth Orbit

NASA Technical Reports Server (NTRS)

Badavi, Francis F.; Stewart-Sloan, Charlotte R.; Wilson, John W.; Adams, Daniel O.

2008-01-01

Correlations of limited ionizing radiation measurements onboard the Space Transportation System (STS; shuttle) and the International Space Station (ISS) with numerical simulations of charged particle transport through spacecraft structure have indicated that usage of hydrogen rich polymeric materials improves the radiation shielding performance of space structures as compared to the traditionally used aluminum alloys. We discuss herein the radiation shielding correlations between measurements on board STS-81 (Atlantis, 1997) using four polyethylene (PE) spheres of varying radii, and STS-89 (Endeavour, 1998) using aluminum alloy spheres; with numerical simulations of charged particle transport using the Langley Research Center (LaRC)-developed High charge (Z) and Energy TRaNsport (HZETRN) algorithm. In the simulations, the Galactic Cosmic Ray (GCR) component of the ionizing radiation environment at Low Earth Orbit (LEO) covering ions in the 1< or equals Z< or equals 28 range is represented by O'Neill's (2004) model. To compute the transmission coefficient for GCR ions at LEO, O'Neill's model is coupled with the angular dependent LaRC cutoff model. The trapped protons/electrons component of LEO environment is represented by a LaRC-developed time dependent procedure which couples the AP8min/AP8max, Deep River Neutron Monitor (DRNM) and F10.7 solar radio frequency measurements. The albedo neutron environment resulting from interaction of GCR ions with upper atmosphere is modeled through extrapolation of the Atmospheric Ionizing Radiation (AIR) measurements. With the validity of numerical simulations through correlation with PE and aluminum spheres measurements established, we further present results from the expansion of the simulations through the selection of high hydrogen content commercially available polymeric constituents such as PE foam core and Spectra fiber(Registered TradeMark) composite face sheet to assess their radiation shield properties as compared to generic PE.

Normalization of a collimated 14.7 MeV neutron source in a neutron spectrometry system for benchmark experiments

NASA Astrophysics Data System (ADS)

Ofek, R.; Tsechanski, A.; Shani, G.

1988-05-01

In the present study a method used to normalize a collimated 14.7 MeV neutron beam is introduced. It combined a measurement of the fast neutron scalar flux passing through the collimator, using a copper foil activation, with a neutron transport calculation of the foil activation per unit source neutron, carried out by the discrete-ordinates transport code DOT 4.2. The geometry of the collimated neutron beam is composed of a D-T neutron source positioned 30 cm in front of a 6 cm diameter collimator, through a 120 cm thick paraffin wall. The neutron flux emitted from the D-T source was counted by an NE-213 scintillator, simultaneously with the irradiation of the copper foil. Thus, the determination of the normalization factor of the D-T source is used for an absolute flux calibration of the NE-213 scintillator. The major contributions to the uncertainty in the determination of the normalization factor, and their origins, are discussed.

The infinite medium Green's function for neutron transport in plane geometry 40 years later

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ganapol, B.D.

1993-01-01

In 1953, the first of what was supposed to be two volumes on neutron transport theory was published. The monograph, entitled [open quotes]Introduction to the Theory of Neutron Diffusion[close quotes] by Case et al., appeared as a Los Alamos National Laboratory report and was to be followed by a second volume, which never appeared as intended because of the death of Placzek. Instead, Case and Zweifel collaborated on the now classic work entitled Linear Transport Theory 2 in which the underlying mathematical theory of linear transport was presented. The initial monograph, however, represented the coming of age of neutron transportmore » theory, which had its roots in radiative transfer and kinetic theory. In addition, it provided the first benchmark results along with the mathematical development for several fundamental neutron transport problems. In particular, one-dimensional infinite medium Green's functions for the monoenergetic transport equation in plane and spherical geometries were considered complete with numerical results to be used as standards to guide code development for applications. Unfortunately, because of the limited computational resources of the day, some numerical results were incorrect. Also, only conventional mathematics and numerical methods were used because the transport theorists of the day were just becoming acquainted with more modern mathematical approaches. In this paper, Green's function solution is revisited in light of modern numerical benchmarking methods with an emphasis on evaluation rather than theoretical results. The primary motivation for considering the Green's function at this time is its emerging use in solving finite and heterogeneous media transport problems.« less

Recent on-beam tests of wide angle neutron polarization analysis with a 3He spin filter: Magic PASTIS on V20 at HZB

NASA Astrophysics Data System (ADS)

Babcock, E.; Salhi, Z.; Gainov, R.; Woracek, R.; Soltner, H.; Pistel, P.; Beule, F.; Bussmann, K.; Heynen, A.; Kämmerling, H.; Suxdorf, F.; Strobl, M.; Russina, M.; Voigt, J.; Ioffe, A.

2017-06-01

A complete XYZ polarization analysis solution is under development for the new thermal time of flight spectrometer TOPAS [1], to be operated in the coming east neutron guide hall at the MLZ. Polarization Analysis Studies on a Thermal Inelastic Spectrometer, commonly called PASTIS [2], is based on polarized 3He neutron spin filters and an XYZ field configuration for the sample environment and a polarization-preserving neutron guide field. The complete system was designed to provide adiabatic transport of the neutron polarization to the sample position while maintaining the homogeneity of the XYZ field. This system has now been tested on the polarized time-of-flight ESS test beam line V20 at HZB [3]. Down to the minimum wavelength of 1.6 Å on the instrument, the magnetic configuration worked ideally for neutron spin transport while giving full experimental freedom to change between the X, Y or Z field configuration. The 3He cell used was polarized at the 3He lab of the JCNS at the MLZ in Garching and transported to HZB in Berlin via car showing that such a transport is indeed feasible for such experiments. We present results of this test and the next steps forward.

Monte Carlo calculations of thermal neutron capture in gadolinium: a comparison of GEANT4 and MCNP with measurements.

PubMed

Enger, Shirin A; Munck af Rosenschöld, Per; Rezaei, Arash; Lundqvist, Hans

2006-02-01

GEANT4 is a Monte Carlo code originally implemented for high-energy physics applications and is well known for particle transport at high energies. The capacity of GEANT4 to simulate neutron transport in the thermal energy region is not equally well known. The aim of this article is to compare MCNP, a code commonly used in low energy neutron transport calculations and GEANT4 with experimental results and select the suitable code for gadolinium neutron capture applications. To account for the thermal neutron scattering from chemically bound atoms [S(alpha,beta)] in biological materials a comparison of thermal neutron fluence in tissue-like poly(methylmethacrylate) phantom is made with MCNP4B, GEANT4 6.0 patch1, and measurements from the neutron capture therapy (NCT) facility at the Studsvik, Sweden. The fluence measurements agreed with MCNP calculated results considering S(alpha,beta). The location of the thermal neutron peak calculated with MCNP without S(alpha,beta) and GEANT4 is shifted by about 0.5 cm towards a shallower depth and is 25%-30% lower in amplitude. Dose distribution from the gadolinium neutron capture reaction is then simulated by MCNP and compared with measured data. The simulations made by MCNP agree well with experimental results. As long as thermal neutron scattering from chemically bound atoms are not included in GEANT4 it is not suitable for NCT applications.

Asymptotic Analysis of Time-Dependent Neutron Transport Coupled with Isotopic Depletion and Radioactive Decay

DOE Office of Scientific and Technical Information (OSTI.GOV)

Brantley, P S

2006-09-27

We describe an asymptotic analysis of the coupled nonlinear system of equations describing time-dependent three-dimensional monoenergetic neutron transport and isotopic depletion and radioactive decay. The classic asymptotic diffusion scaling of Larsen and Keller [1], along with a consistent small scaling of the terms describing the radioactive decay of isotopes, is applied to this coupled nonlinear system of equations in a medium of specified initial isotopic composition. The analysis demonstrates that to leading order the neutron transport equation limits to the standard time-dependent neutron diffusion equation with macroscopic cross sections whose number densities are determined by the standard system of ordinarymore » differential equations, the so-called Bateman equations, describing the temporal evolution of the nuclide number densities.« less

The intensive DT neutron generator of TU Dresden

NASA Astrophysics Data System (ADS)

Klix, Axel; DÖring, Toralf; Domula, Alexander; Zuber, Kai

2018-01-01

TU Dresden operates an accelerator-based intensive DT neutron generator. Experimental activities comprise investigation into material activation and decay, neutron and photon transport in matter and R&D work on radiation detectors for harsh environments. The intense DT neutron generator is capable to produce a maximum of 1012 n/s. The neutron source is a solid-type water-cooled tritium target based on a titanium matrix on a copper carrier. The neutron yield at a typical deuteron beam current of 1 mA is of the order of 1011 n/s in 4Π. A pneumatic sample transport system is available for short-time irradiations and connected to wo high-purity germanium detector spectrometers for the measurement of induced activities. The overall design of the experimental hall with the neutron generator allows a flexible setup of experiments including the possibility of investigating larger structures and cooled samples or samples at high temperatures.

Analytical three-dimensional neutron transport benchmarks for verification of nuclear engineering codes. Final report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ganapol, B.D.; Kornreich, D.E.

Because of the requirement of accountability and quality control in the scientific world, a demand for high-quality analytical benchmark calculations has arisen in the neutron transport community. The intent of these benchmarks is to provide a numerical standard to which production neutron transport codes may be compared in order to verify proper operation. The overall investigation as modified in the second year renewal application includes the following three primary tasks. Task 1 on two dimensional neutron transport is divided into (a) single medium searchlight problem (SLP) and (b) two-adjacent half-space SLP. Task 2 on three-dimensional neutron transport covers (a) pointmore » source in arbitrary geometry, (b) single medium SLP, and (c) two-adjacent half-space SLP. Task 3 on code verification, includes deterministic and probabilistic codes. The primary aim of the proposed investigation was to provide a suite of comprehensive two- and three-dimensional analytical benchmarks for neutron transport theory applications. This objective has been achieved. The suite of benchmarks in infinite media and the three-dimensional SLP are a relatively comprehensive set of one-group benchmarks for isotropically scattering media. Because of time and resource limitations, the extensions of the benchmarks to include multi-group and anisotropic scattering are not included here. Presently, however, enormous advances in the solution for the planar Green`s function in an anisotropically scattering medium have been made and will eventually be implemented in the two- and three-dimensional solutions considered under this grant. Of particular note in this work are the numerical results for the three-dimensional SLP, which have never before been presented. The results presented were made possible only because of the tremendous advances in computing power that have occurred during the past decade.« less

Pulse shape discrimination of Cs2LiYCl6:Ce3+ detectors at high count rate based on triangular and trapezoidal filters

NASA Astrophysics Data System (ADS)

Wen, Xianfei; Enqvist, Andreas

2017-09-01

Cs2LiYCl6:Ce3+ (CLYC) detectors have demonstrated the capability to simultaneously detect γ-rays and thermal and fast neutrons with medium energy resolution, reasonable detection efficiency, and substantially high pulse shape discrimination performance. A disadvantage of CLYC detectors is the long scintillation decay times, which causes pulse pile-up at moderate input count rate. Pulse processing algorithms were developed based on triangular and trapezoidal filters to discriminate between neutrons and γ-rays at high count rate. The algorithms were first tested using low-rate data. They exhibit a pulse-shape discrimination performance comparable to that of the charge comparison method, at low rate. Then, they were evaluated at high count rate. Neutrons and γ-rays were adequately identified with high throughput at rates of up to 375 kcps. The algorithm developed using the triangular filter exhibits discrimination capability marginally higher than that of the trapezoidal filter based algorithm irrespective of low or high rate. The algorithms exhibit low computational complexity and are executable on an FPGA in real-time. They are also suitable for application to other radiation detectors whose pulses are piled-up at high rate owing to long scintillation decay times.

Simulated and measured neutron/gamma light output distribution for poly-energetic neutron/gamma sources

NASA Astrophysics Data System (ADS)

Hosseini, S. A.; Zangian, M.; Aghabozorgi, S.

2018-03-01

In the present paper, the light output distribution due to poly-energetic neutron/gamma (neutron or gamma) source was calculated using the developed MCNPX-ESUT-PE (MCNPX-Energy engineering of Sharif University of Technology-Poly Energetic version) computational code. The simulation of light output distribution includes the modeling of the particle transport, the calculation of scintillation photons induced by charged particles, simulation of the scintillation photon transport and considering the light resolution obtained from the experiment. The developed computational code is able to simulate the light output distribution due to any neutron/gamma source. In the experimental step of the present study, the neutron-gamma discrimination based on the light output distribution was performed using the zero crossing method. As a case study, 241Am-9Be source was considered and the simulated and measured neutron/gamma light output distributions were compared. There is an acceptable agreement between the discriminated neutron/gamma light output distributions obtained from the simulation and experiment.

Cyclotron-based neutron source for BNCT

NASA Astrophysics Data System (ADS)

Mitsumoto, T.; Yajima, S.; Tsutsui, H.; Ogasawara, T.; Fujita, K.; Tanaka, H.; Sakurai, Y.; Maruhashi, A.

2013-04-01

Kyoto University Research Reactor Institute (KURRI) and Sumitomo Heavy Industries, Ltd. (SHI) have developed a cyclotron-based neutron source for Boron Neutron Capture Therapy (BNCT). It was installed at KURRI in Osaka prefecture. The neutron source consists of a proton cyclotron named HM-30, a beam transport system and an irradiation & treatment system. In the cyclotron, H- ions are accelerated and extracted as 30 MeV proton beams of 1 mA. The proton beams is transported to the neutron production target made by a beryllium plate. Emitted neutrons are moderated by lead, iron, aluminum and calcium fluoride. The aperture diameter of neutron collimator is in the range from 100 mm to 250 mm. The peak neutron flux in the water phantom is 1.8×109 neutrons/cm2/sec at 20 mm from the surface at 1 mA proton beam. The neutron source have been stably operated for 3 years with 30 kW proton beam. Various pre-clinical tests including animal tests have been done by using the cyclotron-based neutron source with 10B-p-Borono-phenylalanine. Clinical trials of malignant brain tumors will be started in this year.

Neutronic analysis of the 1D and 1E banks reflux detection system

DOE Office of Scientific and Technical Information (OSTI.GOV)

Blanchard, A.

1999-12-21

Two H Canyon neutron monitoring systems for early detection of postulated abnormal reflux conditions in the Second Uranium Cycle 1E and 1D Mixer-Settle Banks have been designed and built. Monte Carlo neutron transport simulations using the general purpose, general geometry, n-particle MCNP code have been performed to model expected response of the monitoring systems to varying conditions.The confirmatory studies documented herein conclude that the 1E and 1D neutron monitoring systems are able to achieve adequate neutron count rates for various neutron source and detector configurations, thereby eliminating excessive integration count time. Neutron count rate sensitivity studies are also performed. Conversely,more » the transport studies concluded that the neutron count rates are statistically insensitive to nitric acid content in the aqueous region and to the transition region length. These studies conclude that the 1E and 1D neutron monitoring systems are able to predict the postulated reflux conditions for all examined perturbations in the neutron source and detector configurations. In the cases examined, the relative change in the neutron count rates due to postulated transitions from normal {sup 235}U concentration levels to reflux levels remain satisfactory detectable.« less

Application of the MCNPX-McStas interface for shielding calculations and guide design at ESS

NASA Astrophysics Data System (ADS)

Klinkby, E. B.; Knudsen, E. B.; Willendrup, P. K.; Lauritzen, B.; Nonbøl, E.; Bentley, P.; Filges, U.

2014-07-01

Recently, an interface between the Monte Carlo code MCNPX and the neutron ray-tracing code MCNPX was developed [1, 2]. Based on the expected neutronic performance and guide geometries relevant for the ESS, the combined MCNPX-McStas code is used to calculate dose rates along neutron beam guides. The generation and moderation of neutrons is simulated using a full scale MCNPX model of the ESS target monolith. Upon entering the neutron beam extraction region, the individual neutron states are handed to McStas via the MCNPX-McStas interface. McStas transports the neutrons through the beam guide, and by using newly developed event logging capability, the neutron state parameters corresponding to un-reflected neutrons are recorded at each scattering. This information is handed back to MCNPX where it serves as neutron source input for a second MCNPX simulation. This simulation enables calculation of dose rates in the vicinity of the guide. In addition the logging mechanism is employed to record the scatterings along the guides which is exploited to simulate the supermirror quality requirements (i.e. m-values) needed at different positions along the beam guide to transport neutrons in the same guide/source setup.

An Information-Theoretical Approach to Image Resolution Applied to Neutron Imaging Detectors Based Upon Individual Discriminator Signals

NASA Astrophysics Data System (ADS)

Clergeau, Jean-François; Ferraton, Matthieu; Guérard, Bruno; Khaplanov, Anton; Piscitelli, Francesco; Platz, Martin; Rigal, Jean-Marie; Van Esch, Patrick; Daullé, Thibault

2017-01-01

1D or 2D neutron position sensitive detectors with individual wire or strip readout using discriminators have the advantage of being able to treat several neutron impacts partially overlapping in time, hence reducing global dead time. A single neutron impact usually gives rise to several discriminator signals. In this paper, we introduce an information-theoretical definition of image resolution. Two point-like spots of neutron impacts with a given distance between them act as a source of information (each neutron hit belongs to one spot or the other), and the detector plus signal treatment is regarded as an imperfect communication channel that transmits this information. The maximal mutual information obtained from this channel as a function of the distance between the spots allows to define a calibration-independent measure of position resolution. We then apply this measure to quantify the power of position resolution of different algorithms treating these individual discriminator signals which can be implemented in firmware. The method is then applied to different detectors existing at the ILL. Center-of-gravity methods usually improve the position resolution over best-wire algorithms which are the standard way of treating these signals.

Innovative and Advanced Coupled Neutron Transport and Thermal Hydraulic Method (Tool) for the Design, Analysis and Optimization of VHTR/NGNP Prismatic Reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rahnema, Farzad; Garimeela, Srinivas; Ougouag, Abderrafi

2013-11-29

This project will develop a 3D, advanced coarse mesh transport method (COMET-Hex) for steady- state and transient analyses in advanced very high-temperature reactors (VHTRs). The project will lead to a coupled neutronics and thermal hydraulic (T/H) core simulation tool with fuel depletion capability. The computational tool will be developed in hexagonal geometry, based solely on transport theory without (spatial) homogenization in complicated 3D geometries. In addition to the hexagonal geometry extension, collaborators will concurrently develop three additional capabilities to increase the code’s versatility as an advanced and robust core simulator for VHTRs. First, the project team will develop and implementmore » a depletion method within the core simulator. Second, the team will develop an elementary (proof-of-concept) 1D time-dependent transport method for efficient transient analyses. The third capability will be a thermal hydraulic method coupled to the neutronics transport module for VHTRs. Current advancements in reactor core design are pushing VHTRs toward greater core and fuel heterogeneity to pursue higher burn-ups, efficiently transmute used fuel, maximize energy production, and improve plant economics and safety. As a result, an accurate and efficient neutron transport, with capabilities to treat heterogeneous burnable poison effects, is highly desirable for predicting VHTR neutronics performance. This research project’s primary objective is to advance the state of the art for reactor analysis.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Deyglun, C.; Simony, B.; Perot, B.

The quantification of radioactive material is essential in the fields of safeguards, criticality control of nuclear processes, dismantling of nuclear facilities and components, or radioactive waste characterization. The Nuclear Measurement Laboratory (LMN) of CEA is involved in the development of time-correlated neutron detection techniques using plastic scintillators. Usually, 3He proportional counters are used for passive neutron coincidence counting owing to their high thermal neutron capture efficiency and gamma insensitivity. However, the global {sup 3}He shortage in the past few years has made these detectors extremely expensive. In addition, contrary to {sup 3}He counters for which a few tens of microsecondsmore » are needed to thermalize fast neutrons, in view to maximize the {sup 3}He(n,p){sup 3}H capture cross section, plastic scintillators are based on elastic scattering and therefore the light signal is formed within a few nanoseconds, correlated pulses being detected within a few dozen- or hundred nanoseconds. This time span reflects fission particles time of flight, which allows reducing accordingly the duration of the coincidence gate and thus the rate of random coincidences, which may totally blind fission coincidences when using {sup 3}He counters in case of a high (α,n) reaction rate. However, plastic scintillators are very sensitive to gamma rays, requiring the use of a thick metallic shield to reduce the corresponding background. Cross talk between detectors is also a major issue, which consists on the detection of one particle by several detectors due to elastic or inelastic scattering, leading to true but undesired coincidences. Data analysis algorithms are tested to minimize cross-talk in simultaneously activated detectors. The distinction between useful fission coincidences and the correlated background due to cross-talk, (α,n) and induced (n,2n) or (n,n'γ) reactions, is achieved by measuring 3-fold coincidences. The performances of a passive neutron coincidence counting system for radioactive waste drums using plastic scintillators have been studied using the Monte Carlo radiation transport code MCNPX-PoliMi v2.0 coupled to data processing algorithms developed with ROOT data analysis software. In addition to the correlated background, accidental coincidences are taken into account in the simulation by randomly merging pulses from different calculations with fission and (α,n) sources. (authors)« less

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

NASA Astrophysics Data System (ADS)

Lee, Yi-Kang

2017-09-01

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

Determination of the Spectral Index in the Fission Spectrum Energy Regime

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lee, Amy Sarah

2016-05-16

Neutron reaction cross sections play a vital role in tracking the production and destruction of isotopes exposed to neutron fluence. They are central to the process of reconciling the initial and final atom inventories. Measurements of irradiated samples by radiochemical methods in tangent with an algorithm are used to evaluate the fluence a sample is exposed to over the course of the irradiation. This algorithm is the Isotope Production Code (IPC) created and used by the radiochemistry data assessment team at Los Alamos National Laboratory (LANL). An integral result is calculated by varying the total neutron fluence seen by amore » sample. A sample, irradiated in a critical assembly, will be exposed to a unique neutron flux defined by the neutron source and distance of the sample from the source. Neutron cross sections utilized are a function of the hardness of the neutron spectrum at the location of irradiation. A spectral index is used an indicator of the hardness of the neutron spectrum. Cross sections fit forms applied in IPC are collapsed from a LANL 30-group energy structure. Several decades of research and development have been performed to formalize the current IPC cross section library. Basis of the current fission spectrum neutron reaction cross section library is rooted in critical assembly experiments performed from the 1950’s through the early 1970’s at LANL. The focus of this report is development of the spectral index used an indicator of the hardness of the neutron spectrum in the fission spectrum energy regime.« less

Constraining neutron guide optimizations with phase-space considerations

NASA Astrophysics Data System (ADS)

Bertelsen, Mads; Lefmann, Kim

2016-09-01

We introduce a method named the Minimalist Principle that serves to reduce the parameter space for neutron guide optimization when the required beam divergence is limited. The reduced parameter space will restrict the optimization to guides with a minimal neutron intake that are still theoretically able to deliver the maximal possible performance. The geometrical constraints are derived using phase-space propagation from moderator to guide and from guide to sample, while assuming that the optimized guides will achieve perfect transport of the limited neutron intake. Guide systems optimized using these constraints are shown to provide performance close to guides optimized without any constraints, however the divergence received at the sample is limited to the desired interval, even when the neutron transport is not limited by the supermirrors used in the guide. As the constraints strongly limit the parameter space for the optimizer, two control parameters are introduced that can be used to adjust the selected subspace, effectively balancing between maximizing neutron transport and avoiding background from unnecessary neutrons. One parameter is needed to describe the expected focusing abilities of the guide to be optimized, going from perfectly focusing to no correlation between position and velocity. The second parameter controls neutron intake into the guide, so that one can select exactly how aggressively the background should be limited. We show examples of guides optimized using these constraints which demonstrates the higher signal to noise than conventional optimizations. Furthermore the parameter controlling neutron intake is explored which shows that the simulated optimal neutron intake is close to the analytically predicted, when assuming that the guide is dominated by multiple scattering events.

Development of a digital method for neutron/gamma-ray discrimination based on matched filtering

NASA Astrophysics Data System (ADS)

Korolczuk, S.; Linczuk, M.; Romaniuk, R.; Zychor, I.

2016-09-01

Neutron/gamma-ray discrimination is crucial for measurements with detectors sensitive to both neutron and gamma-ray radiation. Different techniques to discriminate between neutrons and gamma-rays based on pulse shape analysis are widely used in many applications, e.g., homeland security, radiation dosimetry, environmental monitoring, fusion experiments, nuclear spectroscopy. A common requirement is to improve a radiation detection level with a high detection reliability. Modern electronic components, such as high speed analog to digital converters and powerful programmable digital circuits for signal processing, allow us to develop a fully digital measurement system. With this solution it is possible to optimize digital signal processing algorithms without changing any electronic components in an acquisition signal path. We report on results obtained with a digital acquisition system DNG@NCBJ designed at the National Centre for Nuclear Research. A 2'' × 2'' EJ309 liquid scintillator was used to register mixed neutron and gamma-ray radiation from PuBe sources. A dedicated algorithm for pulse shape discrimination, based on real-time filtering, was developed and implemented in hardware.

Cyclotron-based neutron source for BNCT

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mitsumoto, T.; Yajima, S.; Tsutsui, H.

2013-04-19

Kyoto University Research Reactor Institute (KURRI) and Sumitomo Heavy Industries, Ltd. (SHI) have developed a cyclotron-based neutron source for Boron Neutron Capture Therapy (BNCT). It was installed at KURRI in Osaka prefecture. The neutron source consists of a proton cyclotron named HM-30, a beam transport system and an irradiation and treatment system. In the cyclotron, H- ions are accelerated and extracted as 30 MeV proton beams of 1 mA. The proton beams is transported to the neutron production target made by a beryllium plate. Emitted neutrons are moderated by lead, iron, aluminum and calcium fluoride. The aperture diameter of neutronmore » collimator is in the range from 100 mm to 250 mm. The peak neutron flux in the water phantom is 1.8 Multiplication-Sign 109 neutrons/cm{sup 2}/sec at 20 mm from the surface at 1 mA proton beam. The neutron source have been stably operated for 3 years with 30 kW proton beam. Various pre-clinical tests including animal tests have been done by using the cyclotron-based neutron source with {sup 10}B-p-Borono-phenylalanine. Clinical trials of malignant brain tumors will be started in this year.« less

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

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

1975-10-01

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

Calculation of Spectra of Neutrons and Charged Particles Produced in a Target of a Neutron Generator

NASA Astrophysics Data System (ADS)

Gaganov, V. V.

2017-12-01

An algorithm for calculating the spectra of neutrons and associated charged particles produced in the target of a neutron generator is detailed. The products of four nuclear reactions 3H( d, n)4He, 2H( d, n)3He, 2H( d, p)3H, and 3He( d, p)4He are analyzed. The results of calculations are presented in the form of neutron spectra for several emission angles and spectra of associated charged particles emitted at an angle of 180° for a deuteron initial energy of 0.13 MeV.

The improvement of the method of equivalent cross section in HTR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Guo, J.; Li, F.

The Method of Equivalence Cross-Sections (MECS) is a combined transport-diffusion method. By appropriately adjusting the diffusion coefficient of homogenized absorber region, the diffusion theory could yield satisfactory results for the full core model with strong neutron absorber material, for example the control rod in High temperature gas cooled reactor (HTR). Original implementation of MECS based on 1-D cell transport model has some limitation on accuracy and applicability, a new implementation of MECS based on 2-D transport model are proposed and tested in this paper. This improvement can extend the MECS to the calculation of twin small absorber ball system whichmore » have a non-circular boring in graphite reflector and different radial position. A least-square algorithm for the calculation of equivalent diffusion coefficient is adopted, and special treatment for diffusion coefficient for higher energy group is proposed in the case that absorber is absent. Numerical results to adopt MECS into control rod calculation in HTR are encouraging. However, there are some problems left. (authors)« less

Effect of Different Structural Materials on Neutronic Performance of a Hybrid Reactor

NASA Astrophysics Data System (ADS)

Übeyli, Mustafa; Tel, Eyyüp

2003-06-01

Selection of structural material for a fusion-fission (hybrid) reactor is very important by taking into account of neutronic performance of the blanket. Refractory metals and alloys have much higher operating temperatures and neutron wall load (NWL) capabilities than low activation materials (ferritic/martensitic steels, vanadium alloys and SiC/SiC composites) and austenitic stainless steels. In this study, effect of primary candidate refractory alloys, namely, W-5Re, T111, TZM and Nb-1Zr on neutronic performance of the hybrid reactor was investigated. Neutron transport calculations were conducted with the help of SCALE 4.3 System by solving the Boltzmann transport equation with code XSDRNPM. Among the investigated structural materials, tantalum had the worst performance due to the fact that it has higher neutron absorption cross section than others. And W-5Re and TZM having similar results showed the best performance.

The Multi-Step CADIS method for shutdown dose rate calculations and uncertainty propagation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ibrahim, Ahmad M.; Peplow, Douglas E.; Grove, Robert E.

2015-12-01

Shutdown dose rate (SDDR) analysis requires (a) a neutron transport calculation to estimate neutron flux fields, (b) an activation calculation to compute radionuclide inventories and associated photon sources, and (c) a photon transport calculation to estimate final SDDR. In some applications, accurate full-scale Monte Carlo (MC) SDDR simulations are needed for very large systems with massive amounts of shielding materials. However, these simulations are impractical because calculation of space- and energy-dependent neutron fluxes throughout the structural materials is needed to estimate distribution of radioisotopes causing the SDDR. Biasing the neutron MC calculation using an importance function is not simple becausemore » it is difficult to explicitly express the response function, which depends on subsequent computational steps. Furthermore, the typical SDDR calculations do not consider how uncertainties in MC neutron calculation impact SDDR uncertainty, even though MC neutron calculation uncertainties usually dominate SDDR uncertainty.« less

Neutron beam characterization measurements at the Manuel Lujan Jr. neutron scattering center

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mocko, Michal; Muhrer, Guenter; Daemen, Luke L

We have measured the neutron beam characteristics of neutron moderators at the Manuel Lujan Jr. Neutron Scattering Center at LANSCE. The absolute thermal neutron flux, energy spectra and time emission spectra were measured for the high resolution and high intensity decoupled water, partially coupled liquid hydrogen and partially coupled water moderators. The results of our experimental study will provide an insight into aging of different target-moderator-reflector-shield components as well as new experimental data for benchmarking of neutron transport codes.

Neutron Transport Simulations for NIST Neutron Lifetime Experiment

NASA Astrophysics Data System (ADS)

Li, Fangchen; BL2 Collaboration Collaboration

2016-09-01

Neutrons in stable nuclei can exist forever; a free neutron lasts for about 15 minutes on average before it beta decays to a proton, an electron, and an antineutrino. Precision measurements of the neutron lifetime test the validity of weak interaction theory and provide input into the theory of the evolution of light elements in the early universe. There are two predominant ways of measuring the neutron lifetime: the bottle method and the beam method. The bottle method measures decays of ultracold neutrons that are stored in a bottle. The beam method measures decay protons in a beam of cold neutrons of known flux. An improved beam experiment is being prepared at the National Institute of Science and Technology (Gaithersburg, MD) with the goal of reducing statistical and systematic uncertainties to the level of 1 s. The purpose of my studies was to develop computer simulations of neutron transport to determine the beam collimation and study the neutron distribution's effect on systematic effects for the experiment, such as the solid angle of the neutron flux monitor. The motivation for the experiment and the results of this work will be presented. This work was supported, in part, by a Grant to Gettysburg College from the Howard Hughes Medical Institute through the Precollege and Undergraduate Science Education Program.

Direct Discrete Method for Neutronic Calculations

DOE Office of Scientific and Technical Information (OSTI.GOV)

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

The objective of this paper is to introduce a new direct method for neutronic calculations. This method which is named Direct Discrete Method, is simpler than the neutron Transport equation and also more compatible with physical meaning of problems. This method is based on physic of problem and with meshing of the desired geometry, writing the balance equation for each mesh intervals and with notice to the conjunction between these mesh intervals, produce the final discrete equations series without production of neutron transport differential equation and mandatory passing from differential equation bridge. We have produced neutron discrete equations for amore » cylindrical shape with two boundary conditions in one group energy. The correction of the results from this method are tested with MCNP-4B code execution. (authors)« less

NEUTRON STAR MASS–RADIUS CONSTRAINTS USING EVOLUTIONARY OPTIMIZATION

DOE Office of Scientific and Technical Information (OSTI.GOV)

Stevens, A. L.; Morsink, S. M.; Fiege, J. D.

The equation of state of cold supra-nuclear-density matter, such as in neutron stars, is an open question in astrophysics. A promising method for constraining the neutron star equation of state is modeling pulse profiles of thermonuclear X-ray burst oscillations from hot spots on accreting neutron stars. The pulse profiles, constructed using spherical and oblate neutron star models, are comparable to what would be observed by a next-generation X-ray timing instrument like ASTROSAT , NICER , or a mission similar to LOFT . In this paper, we showcase the use of an evolutionary optimization algorithm to fit pulse profiles to determinemore » the best-fit masses and radii. By fitting synthetic data, we assess how well the optimization algorithm can recover the input parameters. Multiple Poisson realizations of the synthetic pulse profiles, constructed with 1.6 million counts and no background, were fitted with the Ferret algorithm to analyze both statistical and degeneracy-related uncertainty and to explore how the goodness of fit depends on the input parameters. For the regions of parameter space sampled by our tests, the best-determined parameter is the projected velocity of the spot along the observer’s line of sight, with an accuracy of ≤3% compared to the true value and with ≤5% statistical uncertainty. The next best determined are the mass and radius; for a neutron star with a spin frequency of 600 Hz, the best-fit mass and radius are accurate to ≤5%, with respective uncertainties of ≤7% and ≤10%. The accuracy and precision depend on the observer inclination and spot colatitude, with values of ∼1% achievable in mass and radius if both the inclination and colatitude are ≳60°.« less

FASTER 3: A generalized-geometry Monte Carlo computer program for the transport of neutrons and gamma rays. Volume 1: Summary report

NASA Technical Reports Server (NTRS)

Jordan, T. M.

1970-01-01

The theory used in FASTER-III, a Monte Carlo computer program for the transport of neutrons and gamma rays in complex geometries, is outlined. The program includes the treatment of geometric regions bounded by quadratic and quadric surfaces with multiple radiation sources which have specified space, angle, and energy dependence. The program calculates, using importance sampling, the resulting number and energy fluxes at specified point, surface, and volume detectors. It can also calculate minimum weight shield configuration meeting a specified dose rate constraint. Results are presented for sample problems involving primary neutron, and primary and secondary photon, transport in a spherical reactor shield configuration.

Correlated Production and Analog Transport of Fission Neutrons and Photons using Fission Models FREYA, FIFRELIN and the Monte Carlo Code TRIPOLI-4® .

NASA Astrophysics Data System (ADS)

Verbeke, Jérôme M.; Petit, Odile; Chebboubi, Abdelhazize; Litaize, Olivier

2018-01-01

Fission modeling in general-purpose Monte Carlo transport codes often relies on average nuclear data provided by international evaluation libraries. As such, only average fission multiplicities are available and correlations between fission neutrons and photons are missing. Whereas uncorrelated fission physics is usually sufficient for standard reactor core and radiation shielding calculations, correlated fission secondaries are required for specialized nuclear instrumentation and detector modeling. For coincidence counting detector optimization for instance, precise simulation of fission neutrons and photons that remain correlated in time from birth to detection is essential. New developments were recently integrated into the Monte Carlo transport code TRIPOLI-4 to model fission physics more precisely, the purpose being to access event-by-event fission events from two different fission models: FREYA and FIFRELIN. TRIPOLI-4 simulations can now be performed, either by connecting via an API to the LLNL fission library including FREYA, or by reading external fission event data files produced by FIFRELIN beforehand. These new capabilities enable us to easily compare results from Monte Carlo transport calculations using the two fission models in a nuclear instrumentation application. In the first part of this paper, broad underlying principles of the two fission models are recalled. We then present experimental measurements of neutron angular correlations for 252Cf(sf) and 240Pu(sf). The correlations were measured for several neutron kinetic energy thresholds. In the latter part of the paper, simulation results are compared to experimental data. Spontaneous fissions in 252Cf and 240Pu are modeled by FREYA or FIFRELIN. Emitted neutrons and photons are subsequently transported to an array of scintillators by TRIPOLI-4 in analog mode to preserve their correlations. Angular correlations between fission neutrons obtained independently from these TRIPOLI-4 simulations, using either FREYA or FIFRELIN, are compared to experimental results. For 240Pu(sf), the measured correlations were used to tune the model parameters.

Dynamics of fragment formation in neutron-rich matter

NASA Astrophysics Data System (ADS)

Alcain, P. N.; Dorso, C. O.

2018-01-01

Background: Neutron stars are astronomical systems with nucleons subjected to extreme conditions. Due to the longer range Coulomb repulsion between protons, the system has structural inhomogeneities. Several interactions tailored to reproduce nuclear matter plus a screened Coulomb term reproduce these inhomogeneities known as nuclear pasta. These structural inhomogeneities, located in the crusts of neutron stars, can also arise in expanding systems depending on the thermodynamic conditions (temperature, proton fraction, etc.) and the expansion velocity. Purpose: We aim to find the dynamics of the fragment formation for expanding systems simulated according to the little big bang model. This expansion resembles the evolution of merging neutron stars. Method: We study the dynamics of the nucleons with semiclassical molecular dynamics models. Starting with an equilibrium configuration, we expand the system homogeneously until we arrive at an asymptotic configuration (i.e., very low final densities). We study, with four different cluster recognition algorithms, the fragment distribution throughout this expansion and the dynamics of the cluster formation. Results: Studying the topology of the equilibrium states, before the expansion, we reproduced the known pasta phases plus a novel phase we called pregnocchi, consisting of proton aggregates embedded in a neutron sea. We have identified different fragmentation regimes, depending on the initial temperature and fragment velocity. In particular, for the already mentioned pregnocchi, a neutron cloud surrounds the clusters during the early stages of the expansion, resulting in systems that give rise to configurations compatible with the emergence of the r process. Conclusions: We showed that a proper identification of the cluster distribution is highly dependent on the cluster recognition algorithm chosen, and found that the early cluster recognition algorithm (ECRA) was the most stable one. This approach allowed us to identify the dynamics of the fragment formation. These calculations pave the way to a comparison between Earth experiments and neutron star studies.

Dibaryons in neutron stars

NASA Technical Reports Server (NTRS)

Olinto, Angela V.; Haensel, Pawel; Frieman, Joshua A.

1991-01-01

The effects are studied of H-dibaryons on the structure of neutron stars. It was found that H particles could be present in neutron stars for a wide range of dibaryon masses. The appearance of dibaryons softens the equations of state, lowers the maximum neutron star mass, and affects the transport properties of dense matter. The parameter space is constrained for dibaryons by requiring that a 1.44 solar mass neutron star be gravitationally stable.

Probing multiscale transport and inhomogeneity in a lithium-ion cells using in-situ neutron methods

DOE PAGES

Zhou, Hui; An, Ke; Allu, Srikanth; ...

2016-01-01

Here, we demonstrate for the first time the lithiation process in graphitic anodes using insitu neutron radiography in a pouch cell format. The neutron absorption contrast shows a direct correlation between degree of lithiation and the discharge voltage plateau. Furthermore, we provide a semi-quantitative comparison between the observed spatial variations of neutron attenuation line profile across the graphite electrode and the calculated lithium concentration profiles computed under similar electrochemical discharge conditions. In conjunction, in situ neutron diffraction of a similar pouch cell under identical test protocol was carried to obtain information about the local phase changes upon lithiation. Combined in-situmore » radiography and diffraction opens up a powerful nondestructive method to understand the multi-scale nature of lithium transport and degradation in practical lithium-ion cells.« less

Neutron Angular Scatter Effects in 3DHZETRN: Quasi-Elastic

NASA Technical Reports Server (NTRS)

Wilson, John W.; Werneth, Charles M.; Slaba, Tony C.; Badavi, Francis F.; Reddell, Brandon D.; Bahadori, Amir A.

2017-01-01

The current 3DHZETRN code has a detailed three dimensional (3D) treatment of neutron transport based on a forward/isotropic assumption and has been compared to Monte Carlo (MC) simulation codes in various geometries. In most cases, it has been found that 3DHZETRN agrees with the MC codes to the extent they agree with each other. However, a recent study of neutron leakage from finite geometries revealed that further improvements to the 3DHZETRN formalism are needed. In the present report, angular scattering corrections to the neutron fluence are provided in an attempt to improve fluence estimates from a uniform sphere. It is found that further developments in the nuclear production models are required to fully evaluate the impact of transport model updates. A model for the quasi-elastic neutron production spectra is therefore developed and implemented into 3DHZETRN.

A computationally simple model for determining the time dependent spectral neutron flux in a nuclear reactor core

NASA Astrophysics Data System (ADS)

Schneider, E. A.; Deinert, M. R.; Cady, K. B.

2006-10-01

The balance of isotopes in a nuclear reactor core is key to understanding the overall performance of a given fuel cycle. This balance is in turn most strongly affected by the time and energy-dependent neutron flux. While many large and involved computer packages exist for determining this spectrum, a simplified approach amenable to rapid computation is missing from the literature. We present such a model, which accepts as inputs the fuel element/moderator geometry and composition, reactor geometry, fuel residence time and target burnup and we compare it to OECD/NEA benchmarks for homogeneous MOX and UOX LWR cores. Collision probability approximations to the neutron transport equation are used to decouple the spatial and energy variables. The lethargy dependent neutron flux, governed by coupled integral equations for the fuel and moderator/coolant regions is treated by multigroup thermalization methods, and the transport of neutrons through space is modeled by fuel to moderator transport and escape probabilities. Reactivity control is achieved through use of a burnable poison or adjustable control medium. The model calculates the buildup of 24 actinides, as well as fission products, along with the lethargy dependent neutron flux and the results of several simulations are compared with benchmarked standards.

Semiconductor neutron detector

DOEpatents

Ianakiev, Kiril D [Los Alamos, NM; Littlewood, Peter B [Cambridge, GB; Blagoev, Krastan B [Arlington, VA; Swinhoe, Martyn T [Los Alamos, NM; Smith, James L [Los Alamos, NM; Sullivan, Clair J [Los Alamos, NM; Alexandrov, Boian S [Los Alamos, NM; Lashley, Jason Charles [Santa Fe, NM

2011-03-08

A neutron detector has a compound of lithium in a single crystal form as a neutron sensor element. The lithium compound, containing improved charge transport properties, is either lithium niobate or lithium tantalate. The sensor element is in direct contact with a monitor that detects an electric current. A signal proportional to the electric current is produced and is calibrated to indicate the neutrons sensed. The neutron detector is particularly useful for detecting neutrons in a radiation environment. Such radiation environment may, e.g. include gamma radiation and noise.

Hybrid reduced order modeling for assembly calculations

DOE PAGES

Bang, Youngsuk; Abdel-Khalik, Hany S.; Jessee, Matthew A.; ...

2015-08-14

While the accuracy of assembly calculations has greatly improved due to the increase in computer power enabling more refined description of the phase space and use of more sophisticated numerical algorithms, the computational cost continues to increase which limits the full utilization of their effectiveness for routine engineering analysis. Reduced order modeling is a mathematical vehicle that scales down the dimensionality of large-scale numerical problems to enable their repeated executions on small computing environment, often available to end users. This is done by capturing the most dominant underlying relationships between the model's inputs and outputs. Previous works demonstrated the usemore » of the reduced order modeling for a single physics code, such as a radiation transport calculation. This paper extends those works to coupled code systems as currently employed in assembly calculations. Finally, numerical tests are conducted using realistic SCALE assembly models with resonance self-shielding, neutron transport, and nuclides transmutation/depletion models representing the components of the coupled code system.« less

The Interplay between Proto--Neutron Star Convection and Neutrino Transport in Core-Collapse Supernovae

NASA Astrophysics Data System (ADS)

Mezzacappa, A.; Calder, A. C.; Bruenn, S. W.; Blondin, J. M.; Guidry, M. W.; Strayer, M. R.; Umar, A. S.

1998-01-01

We couple two-dimensional hydrodynamics to realistic one-dimensional multigroup flux-limited diffusion neutrino transport to investigate proto-neutron star convection in core-collapse supernovae, and more specifically, the interplay between its development and neutrino transport. Our initial conditions, time-dependent boundary conditions, and neutrino distributions for computing neutrino heating, cooling, and deleptonization rates are obtained from one-dimensional simulations that implement multigroup flux-limited diffusion and one-dimensional hydrodynamics. The development and evolution of proto-neutron star convection are investigated for both 15 and 25 M⊙ models, representative of the two classes of stars with compact and extended iron cores, respectively. For both models, in the absence of neutrino transport, the angle-averaged radial and angular convection velocities in the initial Ledoux unstable region below the shock after bounce achieve their peak values in ~20 ms, after which they decrease as the convection in this region dissipates. The dissipation occurs as the gradients are smoothed out by convection. This initial proto-neutron star convection episode seeds additional convectively unstable regions farther out beneath the shock. The additional proto-neutron star convection is driven by successive negative entropy gradients that develop as the shock, in propagating out after core bounce, is successively strengthened and weakened by the oscillating inner core. The convection beneath the shock distorts its sphericity, but on the average the shock radius is not boosted significantly relative to its radius in our corresponding one-dimensional models. In the presence of neutrino transport, proto-neutron star convection velocities are too small relative to bulk inflow velocities to result in any significant convective transport of entropy and leptons. This is evident in our two-dimensional entropy snapshots, which in this case appear spherically symmetric. The peak angle-averaged radial and angular convection velocities are orders of magnitude smaller than they are in the corresponding ``hydrodynamics-only'' models. A simple analytical model supports our numerical results, indicating that the inclusion of neutrino transport reduces the entropy-driven (lepton-driven) convection growth rates and asymptotic velocities by a factor ~3 (50) at the neutrinosphere and a factor ~250 (1000) at ρ = 1012 g cm-3, for both our 15 and 25 M⊙ models. Moreover, when transport is included, the initial postbounce entropy gradient is smoothed out by neutrino diffusion, whereas the initial lepton gradient is maintained by electron capture and neutrino escape near the neutrinosphere. Despite the maintenance of the lepton gradient, proto-neutron star convection does not develop over the 100 ms duration typical of all our simulations, except in the instance where ``low-test'' intial conditions are used, which are generated by core-collapse and bounce simulations that neglect neutrino-electron scattering and ion-ion screening corrections to neutrino-nucleus elastic scattering. Models favoring the development of proto-neutron star convection either by starting with more favorable, albeit artificial (low-test), initial conditions or by including transport corrections that were ignored in our ``fiducial'' models were considered. Our conclusions nonetheless remained the same. Evidence of proto-neutron star convection in our two-dimensional entropy snapshots was minimal, and, as in our fiducial models, the angle-averaged convective velocities when neutrino transport was included remained orders of magnitude smaller than their counterparts in the corresponding hydrodynamics-only models.

Fast neutron flux analyzer with real-time digital pulse shape discrimination

NASA Astrophysics Data System (ADS)

Ivanova, A. A.; Zubarev, P. V.; Ivanenko, S. V.; Khilchenko, A. D.; Kotelnikov, A. I.; Polosatkin, S. V.; Puryga, E. A.; Shvyrev, V. G.; Sulyaev, Yu. S.

2016-08-01

Investigation of subthermonuclear plasma confinement and heating in magnetic fusion devices such as GOL-3 and GDT at the Budker Institute (Novosibirsk, Russia) requires sophisticated equipment for neutron-, gamma- diagnostics and upgrading data acquisition systems with online data processing. Measurement of fast neutron flux with stilbene scintillation detectors raised the problem of discrimination of the neutrons (n) from background cosmic particles (muons) and neutron-induced gamma rays (γ). This paper describes a fast neutron flux analyzer with real-time digital pulse-shape discrimination (DPSD) algorithm FPGA-implemented for the GOL-3 and GDT devices. This analyzer was tested and calibrated with the help of 137Cs and 252Cf radiation sources. The Figures of Merit (FOM) calculated for different energy cuts are presented.

Benchmarking shielding simulations for an accelerator-driven spallation neutron source

DOE PAGES

Cherkashyna, Nataliia; Di Julio, Douglas D.; Panzner, Tobias; ...

2015-08-09

The shielding at an accelerator-driven spallation neutron facility plays a critical role in the performance of the neutron scattering instruments, the overall safety, and the total cost of the facility. Accurate simulation of shielding components is thus key for the design of upcoming facilities, such as the European Spallation Source (ESS), currently in construction in Lund, Sweden. In this paper, we present a comparative study between the measured and the simulated neutron background at the Swiss Spallation Neutron Source (SINQ), at the Paul Scherrer Institute (PSI), Villigen, Switzerland. The measurements were carried out at several positions along the SINQ monolithmore » wall with the neutron dosimeter WENDI-2, which has a well-characterized response up to 5 GeV. The simulations were performed using the Monte-Carlo radiation transport code Geant4, and include a complete transport from the proton beam to the measurement locations in a single calculation. An agreement between measurements and simulations is about a factor of 2 for the points where the measured radiation dose is above the background level, which is a satisfactory result for such simulations spanning many energy regimes, different physics processes and transport through several meters of shielding materials. The neutrons contributing to the radiation field emanating from the monolith were confirmed to originate from neutrons with energies above 1 MeV in the target region. The current work validates Geant4 as being well suited for deep-shielding calculations at accelerator-based spallation sources. We also extrapolate what the simulated flux levels might imply for short (several tens of meters) instruments at ESS.« less

Recent Research applications at the Athens Neutron Monitor Station

NASA Astrophysics Data System (ADS)

Mavromichalaki, H.; Gerontidou, M.; Paschalis, P.; Papaioannou, A.; Paouris, E.; Papailiou, M.; Souvatzoglou, G.

2015-08-01

The ground based neutron monitor measurements play a key role in the field of space physics, solar-terrestrial relations, and space weather applications. The Athens cosmic ray group has developed several research applications such as an optimized automated Ground Level Enhancement Alert (GLE Alert Plus) and a web interface, providing data from multiple Neutron Monitor stations (Multi-Station tool). These services are actually available via the Space Weather Portal operated by the European Space Agency (http://swe.ssa.esa.int). In addition, two simulation tools, based on Geant4, have also been implemented. The first one is for the simulation of the cosmic ray showers in the atmosphere (DYASTIMA) and the second one is for the simulation of the 6NM-64 neutron monitor. The contribution of the simulation tools to the calculations of the radiation dose received by air crews and passengers within the Earth's atmosphere and to the neutron monitor study is presented as well. Furthermore, the accurate calculation of the barometric coefficient and the primary data processing by filtering algorithms, such as the well known Median Editor and the developed by the Athens group ANN Algorithm and Edge Editor which contribute to the provision of high quality neutron monitor data are also discussed. Finally, a Space Weather Forecasting Center which provides a three day geomagnetic activity report on a daily basis has been set up and has been operating for the last two years at the Athens Neutron Monitor Station.

Development of a reliable estimation procedure of radioactivity inventory in a BWR plant due to neutron irradiation for decommissioning

NASA Astrophysics Data System (ADS)

Tanaka, Ken-ichi; Ueno, Jun

2017-09-01

Reliable information of radioactivity inventory resulted from the radiological characterization is important in order to plan decommissioning planning and is also crucial in order to promote decommissioning in effectiveness and in safe. The information is referred to by planning of decommissioning strategy and by an application to regulator. Reliable information of radioactivity inventory can be used to optimize the decommissioning processes. In order to perform the radiological characterization reliably, we improved a procedure of an evaluation of neutron-activated materials for a Boiling Water Reactor (BWR). Neutron-activated materials are calculated with calculation codes and their validity should be verified with measurements. The evaluation of neutron-activated materials can be divided into two processes. One is a distribution calculation of neutron-flux. Another is an activation calculation of materials. The distribution calculation of neutron-flux is performed with neutron transport calculation codes with appropriate cross section library to simulate neutron transport phenomena well. Using the distribution of neutron-flux, we perform distribution calculations of radioactivity concentration. We also estimate a time dependent distribution of radioactivity classification and a radioactive-waste classification. The information obtained from the evaluation is utilized by other tasks in the preparatory tasks to make the decommissioning plan and the activity safe and rational.

S4 solution of the transport equation for eigenvalues using Legendre polynomials

NASA Astrophysics Data System (ADS)

Öztürk, Hakan; Bülbül, Ahmet

2017-09-01

Numerical solution of the transport equation for monoenergetic neutrons scattered isotropically through the medium of a finite homogeneous slab is studied for the determination of the eigenvalues. After obtaining the discrete ordinates form of the transport equation, separated homogeneous and particular solutions are formed and then the eigenvalues are calculated using the Gauss-Legendre quadrature set. Then, the calculated eigenvalues for various values of the c0, the mean number of secondary neutrons per collision, are given in the tables.

New Frontier in Probing Fluid Transport in Low-Permeability Geomedia: Applications of Elastic and Inelastic Neutron Scattering

NASA Astrophysics Data System (ADS)

Ding, M.; Hjelm, R.; Sussman, A. J.

2016-12-01

Low-permeability geomedia are prevalent in subsurface environments. They have become increasingly important in a wide range of applications such as CO2-sequestration, hydrocarbon recovery, enhanced geothermal systems, legacy waste stewardship, high-level radioactive waste disposal, and global security. The flow and transport characteristics of low-permeability geomedia are dictated by their exceedingly low permeability values ranging from 10-6 to 10-12 darcy with porosities dominated by nanoscale pores. Developing new characterization methods and robust computational models that allow estimation of transport properties of low-permeability geomedia has been identified as a critical basic research and technology development need for controlling subsurface and fluids flow. Due to its sensibility to hydrogen and flexible sample environment, neutron based elastic and inelastic scattering can, through various techniques, interrogate all the nanoscale pores in the sample whether they are fluid accessible or not, and readily characterize interfacial waters. In this presentation, we will present two studies revealing the effects of nanoscale pore confinement on fluid dynamics in geomedia. In one study, we use combined (ultra-small)/small-angle elastic neutron scatterings to probe nanoporous features responses in geological materials to transport processes. In the other study, incoherent inelastic neutron scattering was used to distingwish between intergranular pore water and fluid inclusion moisture in bedded rock salt, and to explore their thermal stablibility. Our work demonstrates that neutron based elastic and inelastic scatterings are techniques of choice for in situ probing hydrocarbon and water behavior in nanoporous materials, providing new insights into water-rock interaction and fluids transport in low-permeability geomaterials.

A real-time neutron-gamma discriminator based on the support vector machine method for the time-of-flight neutron spectrometer

NASA Astrophysics Data System (ADS)

Wei, ZHANG; Tongyu, WU; Bowen, ZHENG; Shiping, LI; Yipo, ZHANG; Zejie, YIN

2018-04-01

A new neutron-gamma discriminator based on the support vector machine (SVM) method is proposed to improve the performance of the time-of-flight neutron spectrometer. The neutron detector is an EJ-299-33 plastic scintillator with pulse-shape discrimination (PSD) property. The SVM algorithm is implemented in field programmable gate array (FPGA) to carry out the real-time sifting of neutrons in neutron-gamma mixed radiation fields. This study compares the ability of the pulse gradient analysis method and the SVM method. The results show that this SVM discriminator can provide a better discrimination accuracy of 99.1%. The accuracy and performance of the SVM discriminator based on FPGA have been evaluated in the experiments. It can get a figure of merit of 1.30.

The "neutron channel design"—A method for gaining the desired neutrons

NASA Astrophysics Data System (ADS)

Hu, G.; Hu, H. S.; Wang, S.; Pan, Z. H.; Jia, Q. G.; Yan, M. F.

2016-12-01

The neutrons with desired parameters can be obtained after initial neutrons penetrating various structure and component of the material. A novel method, the "neutron channel design", is proposed in this investigation for gaining the desired neutrons. It is established by employing genetic algorithm (GA) combining with Monte Carlo software. This method is verified by obtaining 0.01eV to 1.0eV neutrons from the Compact Accelerator-driven Neutron Source (CANS). One layer polyethylene (PE) moderator was designed and installed behind the beryllium target in CANS. The simulations and the experiment for detection the neutrons were carried out. The neutron spectrum at 500cm from the PE moderator was simulated by MCNP and PHITS software. The counts of 0.01eV to 1.0eV neutrons were simulated by MCNP and detected by the thermal neutron detector in the experiment. These data were compared and analyzed. Then this method is researched on designing the complex structure of PE and the composite material consisting of PE, lead and zirconium dioxide.

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

NASA Astrophysics Data System (ADS)

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.

Introducing single-crystal scattering and optical potentials into MCNPX: Predicting neutron emission from a convoluted moderator

DOE PAGES

Gallmeier, F. X.; Iverson, E. B.; Lu, W.; ...

2016-01-08

Neutron transport simulation codes are an indispensable tool used for the design and construction of modern neutron scattering facilities and instrumentation. It has become increasingly clear that some neutron instrumentation has started to exploit physics that is not well-modelled by the existing codes. Particularly, the transport of neutrons through single crystals and across interfaces in MCNP(X), Geant4 and other codes ignores scattering from oriented crystals and refractive effects, and yet these are essential ingredients for the performance of monochromators and ultra-cold neutron transport respectively (to mention but two examples). In light of these developments, we have extended the MCNPX codemore » to include a single-crystal neutron scattering model and neutron reflection/refraction physics. Furthermore, we have also generated silicon scattering kernels for single crystals of definable orientation with respect to an incoming neutron beam. As a first test of these new tools, we have chosen to model the recently developed convoluted moderator concept, in which a moderating material is interleaved with layers of perfect crystals to provide an exit path for neutrons moderated to energies below the crystal s Bragg cut off at locations deep within the moderator. Studies of simple cylindrical convoluted moderator systems of 100 mm diameter and composed of polyethylene and single crystal silicon were performed with the upgraded MCNPX code and reproduced the magnitude of effects seen in experiments compared to homogeneous moderator systems. Applying different material properties for refraction and reflection, and by replacing the silicon in the models with voids, we show that the emission enhancements seen in recent experiments are primarily caused by the transparency of the silicon/void layers. Finally the convoluted moderator experiments described by Iverson et al. were simulated and we find satisfactory agreement between the measurement and the results of simulations performed using the tools we have developed.« less

Ageing of a neutron shielding used in transport/storage casks

DOE Office of Scientific and Technical Information (OSTI.GOV)

Nizeyiman, Fidele; Alami, Aatif; Issard, Herve

2012-07-11

In radioactive materials transport/storage casks, a mineral-filled vinylester composite is used for neutron shielding which relies on its hydrogen and boron atoms content. During cask service life, this composite is mainly subjected to three types of ageing: hydrothermal ageing, thermal oxidation and neutron irradiation. The aim of this study is to investigate the effect of hydrothermal ageing on the properties and chemical composition of this polymer composite. At high temperature (120 Degree-Sign C and 140 Degree-Sign C), the main consequence is the strong decrease of mechanical properties induced by the filler/matrix debonding.

VERA 3.6 Release Notes

DOE Office of Scientific and Technical Information (OSTI.GOV)

Williamson, Richard L.; Kochunas, Brendan; Adams, Brian M.

The Virtual Environment for Reactor Applications components included in this distribution include selected computational tools and supporting infrastructure that solve neutronics, thermal-hydraulics, fuel performance, and coupled neutronics-thermal hydraulics problems. The infrastructure components provide a simplified common user input capability and provide for the physics integration with data transfer and coupled-physics iterative solution algorithms.

Synthetic neutron camera and spectrometer in JET based on AFSI-ASCOT simulations

NASA Astrophysics Data System (ADS)

Sirén, P.; Varje, J.; Weisen, H.; Koskela, T.; contributors, JET

2017-09-01

The ASCOT Fusion Source Integrator (AFSI) has been used to calculate neutron production rates and spectra corresponding to the JET 19-channel neutron camera (KN3) and the time-of-flight spectrometer (TOFOR) as ideal diagnostics, without detector-related effects. AFSI calculates fusion product distributions in 4D, based on Monte Carlo integration from arbitrary reactant distribution functions. The distribution functions were calculated by the ASCOT Monte Carlo particle orbit following code for thermal, NBI and ICRH particle reactions. Fusion cross-sections were defined based on the Bosch-Hale model and both DD and DT reactions have been included. Neutrons generated by AFSI-ASCOT simulations have already been applied as a neutron source of the Serpent neutron transport code in ITER studies. Additionally, AFSI has been selected to be a main tool as the fusion product generator in the complete analysis calculation chain: ASCOT - AFSI - SERPENT (neutron and gamma transport Monte Carlo code) - APROS (system and power plant modelling code), which encompasses the plasma as an energy source, heat deposition in plant structures as well as cooling and balance-of-plant in DEMO applications and other reactor relevant analyses. This conference paper presents the first results and validation of the AFSI DD fusion model for different auxiliary heating scenarios (NBI, ICRH) with very different fast particle distribution functions. Both calculated quantities (production rates and spectra) have been compared with experimental data from KN3 and synthetic spectrometer data from ControlRoom code. No unexplained differences have been observed. In future work, AFSI will be extended for synthetic gamma diagnostics and additionally, AFSI will be used as part of the neutron transport calculation chain to model real diagnostics instead of ideal synthetic diagnostics for quantitative benchmarking.

Classifying threats with a 14-MeV neutron interrogation system.

PubMed

Strellis, Dan; Gozani, Tsahi

2005-01-01

SeaPODDS (Sea Portable Drug Detection System) is a non-intrusive tool for detecting concealed threats in hidden compartments of maritime vessels. This system consists of an electronic neutron generator, a gamma-ray detector, a data acquisition computer, and a laptop computer user-interface. Although initially developed to detect narcotics, recent algorithm developments have shown that the system is capable of correctly classifying a threat into one of four distinct categories: narcotic, explosive, chemical weapon, or radiological dispersion device (RDD). Detection of narcotics, explosives, and chemical weapons is based on gamma-ray signatures unique to the chemical elements. Elements are identified by their characteristic prompt gamma-rays induced by fast and thermal neutrons. Detection of RDD is accomplished by detecting gamma-rays emitted by common radioisotopes and nuclear reactor fission products. The algorithm phenomenology for classifying threats into the proper categories is presented here.

Deuterated scintillators and their application to neutron spectroscopy

NASA Astrophysics Data System (ADS)

Febbraro, M.; Lawrence, C. C.; Zhu, H.; Pierson, B.; Torres-Isea, R. O.; Becchetti, F. D.; Kolata, J. J.; Riggins, J.

2015-06-01

Deuterated scintillators have been used as a tool for neutron spectroscopy without Neutron Time-of-Flight (n-ToF) for more than 30 years. This article will provide a brief historical overview of the technique and current uses of deuterated scintillators in the UM-DSA and DESCANT arrays. Pulse-shape discrimination and spectrum unfolding with the maximum-likelihood expectation maximization algorithm will be discussed. Experimental unfolding and cross section results from measurements of (d,n), (3He,n) and (α,n) reactions are shown.

Characteristics comparison between a cyclotron-based neutron source and KUR-HWNIF for boron neutron capture therapy

NASA Astrophysics Data System (ADS)

Tanaka, H.; Sakurai, Y.; Suzuki, M.; Masunaga, S.; Kinashi, Y.; Kashino, G.; Liu, Y.; Mitsumoto, T.; Yajima, S.; Tsutsui, H.; Maruhashi, A.; Ono, K.

2009-06-01

At Kyoto University Research Reactor Institute (KURRI), 275 clinical trials of boron neutron capture therapy (BNCT) have been performed as of March 2006, and the effectiveness of BNCT has been revealed. In order to further develop BNCT, it is desirable to supply accelerator-based epithermal-neutron sources that can be installed near the hospital. We proposed the method of filtering and moderating fast neutrons, which are emitted from the reaction between a beryllium target and 30-MeV protons accelerated by a cyclotron accelerator, using an optimum moderator system composed of iron, lead, aluminum and calcium fluoride. At present, an epithermal-neutron source is under construction from June 2008. This system consists of a cyclotron accelerator, beam transport system, neutron-yielding target, filter, moderator and irradiation bed. In this article, an overview of this system and the properties of the treatment neutron beam optimized by the MCNPX Monte Carlo neutron transport code are presented. The distribution of biological effect weighted dose in a head phantom compared with that of Kyoto University Research Reactor (KUR) is shown. It is confirmed that for the accelerator, the biological effect weighted dose for a deeply situated tumor in the phantom is 18% larger than that for KUR, when the limit dose of the normal brain is 10 Gy-eq. The therapeutic time of the cyclotron-based neutron sources are nearly one-quarter of that of KUR. The cyclotron-based epithermal-neutron source is a promising alternative to reactor-based neutron sources for treatments by BNCT.

Radiation transport codes for potential applications related to radiobiology and radiotherapy using protons, neutrons, and negatively charged pions

NASA Technical Reports Server (NTRS)

Armstrong, T. W.

1972-01-01

Several Monte Carlo radiation transport computer codes are used to predict quantities of interest in the fields of radiotherapy and radiobiology. The calculational methods are described and comparisions of calculated and experimental results are presented for dose distributions produced by protons, neutrons, and negatively charged pions. Comparisons of calculated and experimental cell survival probabilities are also presented.

Space Radiation Transport Code Development: 3DHZETRN

NASA Technical Reports Server (NTRS)

Wilson, John W.; Slaba, Tony C.; Badavi, Francis F.; Reddell, Brandon D.; Bahadori, Amir A.

2015-01-01

The space radiation transport code, HZETRN, has been used extensively for research, vehicle design optimization, risk analysis, and related applications. One of the simplifying features of the HZETRN transport formalism is the straight-ahead approximation, wherein all particles are assumed to travel along a common axis. This reduces the governing equation to one spatial dimension allowing enormous simplification and highly efficient computational procedures to be implemented. Despite the physical simplifications, the HZETRN code is widely used for space applications and has been found to agree well with fully 3D Monte Carlo simulations in many circumstances. Recent work has focused on the development of 3D transport corrections for neutrons and light ions (Z < 2) for which the straight-ahead approximation is known to be less accurate. Within the development of 3D corrections, well-defined convergence criteria have been considered, allowing approximation errors at each stage in model development to be quantified. The present level of development assumes the neutron cross sections have an isotropic component treated within N explicit angular directions and a forward component represented by the straight-ahead approximation. The N = 1 solution refers to the straight-ahead treatment, while N = 2 represents the bi-directional model in current use for engineering design. The figure below shows neutrons, protons, and alphas for various values of N at locations in an aluminum sphere exposed to a solar particle event (SPE) spectrum. The neutron fluence converges quickly in simple geometry with N > 14 directions. The improved code, 3DHZETRN, transports neutrons, light ions, and heavy ions under space-like boundary conditions through general geometry while maintaining a high degree of computational efficiency. A brief overview of the 3D transport formalism for neutrons and light ions is given, and extensive benchmarking results with the Monte Carlo codes Geant4, FLUKA, and PHITS are provided for a variety of boundary conditions and geometries. Improvements provided by the 3D corrections are made clear in the comparisons. Developments needed to connect 3DHZETRN to vehicle design and optimization studies will be discussed. Future theoretical development will relax the forward plus isotropic interaction assumption to more general angular dependence.

Simulation des fuites neutroniques a l'aide d'un modele B1 heterogene pour des reacteurs a neutrons rapides et a eau legere

NASA Astrophysics Data System (ADS)

Faure, Bastien

The neutronic calculation of a reactor's core is usually done in two steps. After solving the neutron transport equation over an elementary domain of the core, a set of parameters, namely macroscopic cross sections and potentially diffusion coefficients, are defined in order to perform a full core calculation. In the first step, the cell or assembly is calculated using the "fundamental mode theory", the pattern being inserted in an infinite lattice of periodic structures. This simple representation allows a precise modeling for the geometry and the energy variable and can be treated within transport theory with minimalist approximations. However, it supposes that the reactor's core can be treated as a periodic lattice of elementary domains, which is already a big hypothesis, and cannot, at first sight, take into account neutron leakage between two different zones and out of the core. The leakage models propose to correct the transport equation with an additional leakage term in order to represent this phenomenon. For historical reasons, numerical methods for solving the transport equation being limited by computer's features (processor speeds and memory sizes), the leakage term is, in most cases, modeled by a homogeneous and isotropic probability within a "homogeneous leakage model". Driven by technological innovation in the computer science field, "heterogeneous leakage models" have been developed and implemented in several neutron transport calculation codes. This work focuses on a study of some of those models, including the TIBERE model from the DRAGON-3 code developed at Ecole Polytechnique de Montreal, as well as the heterogeneous model from the APOLLO-3 code developed at Commissariat a l'Energie Atomique et aux energies alternatives. The research based on sodium cooled fast reactors and light water reactors has allowed us to demonstrate the interest of those models compared to a homogeneous leakage model. In particular, it has been shown that a heterogeneous model has a significant impact on the calculation of the out of core leakage rate that permits a better estimation of the transport equation eigenvalue Keff . The neutron streaming between two zones of different compositions was also proven to be better calculated.

Localization and spectral isolation of special nuclear material using stochastic image reconstruction

NASA Astrophysics Data System (ADS)

Hamel, M. C.; Polack, J. K.; Poitrasson-Rivière, A.; Clarke, S. D.; Pozzi, S. A.

2017-01-01

In this work we present a technique for isolating the gamma-ray and neutron energy spectra from multiple radioactive sources localized in an image. Image reconstruction algorithms for radiation scatter cameras typically focus on improving image quality. However, with scatter cameras being developed for non-proliferation applications, there is a need for not only source localization but also source identification. This work outlines a modified stochastic origin ensembles algorithm that provides localized spectra for all pixels in the image. We demonstrated the technique by performing three experiments with a dual-particle imager that measured various gamma-ray and neutron sources simultaneously. We showed that we could isolate the peaks from 22Na and 137Cs and that the energy resolution is maintained in the isolated spectra. To evaluate the spectral isolation of neutrons, a 252Cf source and a PuBe source were measured simultaneously and the reconstruction showed that the isolated PuBe spectrum had a higher average energy and a greater fraction of neutrons at higher energies than the 252Cf. Finally, spectrum isolation was used for an experiment with weapons grade plutonium, 252Cf, and AmBe. The resulting neutron and gamma-ray spectra showed the expected characteristics that could then be used to identify the sources.

Conceptual design of a hybrid neutron-gamma detector for study of β-delayed neutrons at the RIB facility of RIKEN

NASA Astrophysics Data System (ADS)

Tarifeño-Saldivia, A.; Tain, J. L.; Domingo-Pardo, C.; Calviño, F.; Cortés, G.; Phong, V. H.; Riego, A.; Agramunt, J.; Algora, A.; Brewer, N.; Caballero-Folch, R.; Coleman-Smith, P. J.; Davinson, T.; Dillmann, I.; Estradé, A.; Griffin, C. J.; Grzywacz, R.; Harkness-Brennan, L. J.; Kiss, G. G.; Kogimtzis, M.; Labiche, M.; Lazarus, I. H.; Lorusso, G.; Matsui, K.; Miernik, K.; Montes, F.; Morales, A. I.; Nishimura, S.; Page, R. D.; Podolyák, Z. S.; Pucknell, V. F. E.; Rasco, B. C.; Regan, P.; Rubio, B.; Rykaczewski, K. P.; Saito, Y.; Sakurai, H.; Simpson, J.; Sokol, E.; Surman, R.; Svirkhin, A.; Thomas, S. L.; Tolosa, A.; Woods, P.

2017-04-01

The conceptual design of the BRIKEN neutron detector at the radioactive ion beam factory (RIBF) of the RIKEN Nishina Center is reported. The BRIKEN setup is a complex system aimed at detecting heavy-ion implants, β particles, γ rays and β-delayed neutrons. The whole setup includes the Advanced Implantation Detection Array (AIDA), two HPGe Clover detectors and up to 166 3He-filled counters embedded in a high-density polyethylene moderator. The design is quite complex due to the large number and different types of 3He-tubes involved and the additional constraints introduced by the ancillary detectors for charged particles and γ rays. This article reports on a novel methodology developed for the conceptual design and optimisation of the 3He-counter array, aiming for the best possible performance in terms of neutron detection. The algorithm is based on a geometric representation of two selected detector parameters of merit, namely, the average neutron detection efficiency and the efficiency flatness as a function of a reduced number of geometric variables. The response of the neutron detector is obtained from a systematic Monte Carlo simulation implemented in GEANT4. The robustness of the algorithm allowed us to design a versatile detection system, which operated in hybrid mode includes the full neutron counter and two clover detectors for high-precision gamma spectroscopy. In addition, the system can be reconfigured into a compact mode by removing the clover detectors and re-arranging the 3He tubes in order to maximize the neutron detection performance. Both operation modes shows a rather flat and high average efficiency. In summary, we have designed a system which shows an average efficiency for hybrid mode (3He tubes + clovers) of 68.6% and 64% for neutron energies up to 1 and 5 MeV, respectively. For compact mode (only 3He tubes), the average efficiency is 75.7% and 71% for neutron energies up to 1 and 5 MeV, respectively. The performance of the BRIKEN detection system has been also quantified by means of Monte Carlo simulations with different neutron energy distributions.

Neutron Capture Energies for Flux Normalization and Approximate Model for Gamma-Smeared Power

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kim, Kang Seog; Clarno, Kevin T.; Liu, Yuxuan

The Consortium for Advanced Simulation of Light Water Reactors (CASL) Virtual Environment for Reactor Applications (VERA) neutronics simulator MPACT has used a single recoverable fission energy for each fissionable nuclide assuming that all recoverable energies come only from fission reaction, for which capture energy is merged with fission energy. This approach includes approximations and requires improvement by separating capture energy from the merged effective recoverable energy. This report documents the procedure to generate recoverable neutron capture energies and the development of a program called CapKappa to generate capture energies. Recoverable neutron capture energies have been generated by using CapKappa withmore » the evaluated nuclear data file (ENDF)/B-7.0 and 7.1 cross section and decay libraries. The new capture kappas were compared to the current SCALE-6.2 and the CASMO-5 capture kappas. These new capture kappas have been incorporated into the Simplified AMPX 51- and 252-group libraries, and they can be used for the AMPX multigroup (MG) libraries and the SCALE code package. The CASL VERA neutronics simulator MPACT does not include a gamma transport capability, which limits it to explicitly estimating local energy deposition from fission, neutron, and gamma slowing down and capture. Since the mean free path of gamma rays is typically much longer than that for the neutron, and the total gamma energy is about 10% to the total energy, the gamma-smeared power distribution is different from the fission power distribution. Explicit local energy deposition through neutron and gamma transport calculation is significantly important in multi-physics whole core simulation with thermal-hydraulic feedback. Therefore, the gamma transport capability should be incorporated into the CASL neutronics simulator MPACT. However, this task will be timeconsuming in developing the neutron induced gamma production and gamma cross section libraries. This study is to investigate an approximate model to estimate gammasmeared power distribution without performing any gamma transport calculation. A simple approximate gamma smearing model has been investigated based on the facts that pinwise gamma energy depositions are almost flat over a fuel assembly, and assembly-wise gamma energy deposition is proportional to kappa-fission energy deposition. The approximate gamma smearing model works well for single assembly cases, and can partly improve the gamma smeared power distribution for the whole core model. Although the power distributions can be improved by the approximate gamma smearing model, still there is an issue to explicitly obtain local energy deposition. A new simple approach or gamma transport/diffusion capability may need to be incorporated into MPACT to estimate local energy deposition for more robust multi-physics simulation.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Uchida, Y., E-mail: h1312101@mailg.nc-toyama.ac.jp; Takada, E.; Fujisaki, A.

Neutron and γ-ray (n-γ) discrimination with a digital signal processing system has been used to measure the neutron emission profile in magnetic confinement fusion devices. However, a sampling rate must be set low to extend the measurement time because the memory storage is limited. Time jitter decreases a discrimination quality due to a low sampling rate. As described in this paper, a new charge comparison method was developed. Furthermore, automatic n-γ discrimination method was examined using a probabilistic approach. Analysis results were investigated using the figure of merit. Results show that the discrimination quality was improved. Automatic discrimination was appliedmore » using the EM algorithm and k-means algorithm.« less

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Burns, Kimberly A.

2009-08-01

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

The performance of the upgraded Los Alamos Neutron Source

NASA Astrophysics Data System (ADS)

Ito, Takeyasu; LANL UCN Source Collaboration

2017-09-01

Los Alamos National Laboratory has been operating an ultracold (UCN) source based on a solid deuterium (SD2) UCN converter driven by spallation neutrons for over 10 years. It has recently been successfully upgraded, by replacing the cryostat that contains the cold neutron moderator, SD2 volume, and vertical UCN guide. The horizontal UCN guide that transports UCN out of the radiation shield was also replaced. The new design reflects lessons learned from the 10+ year long operation of the previous version of the UCN source and is optimized to maximize the cold neutron flux at the SD2 volume, featuring a close coupled cold neutron moderator, and maximize the transport of the UCN to experiments. During the commissioning of the upgraded UCN source, data were collected to measure its performance, including cold neutron spectra as a function of the cold moderator temperature, and the UCN density in a vessel outside the source. In this talk, after a brief overview of the design of the upgraded source, the results of the performance tests and comparison to prediction will be presented. This work was funded by LANL LDRD.

Interfacing MCNPX and McStas for simulation of neutron transport

NASA Astrophysics Data System (ADS)

Klinkby, Esben; Lauritzen, Bent; Nonbøl, Erik; Kjær Willendrup, Peter; Filges, Uwe; Wohlmuther, Michael; Gallmeier, Franz X.

2013-02-01

Simulations of target-moderator-reflector system at spallation sources are conventionally carried out using Monte Carlo codes such as MCNPX (Waters et al., 2007 [1]) or FLUKA (Battistoni et al., 2007; Ferrari et al., 2005 [2,3]) whereas simulations of neutron transport from the moderator and the instrument response are performed by neutron ray tracing codes such as McStas (Lefmann and Nielsen, 1999; Willendrup et al., 2004, 2011a,b [4-7]). The coupling between the two simulation suites typically consists of providing analytical fits of MCNPX neutron spectra to McStas. This method is generally successful but has limitations, as it e.g. does not allow for re-entry of neutrons into the MCNPX regime. Previous work to resolve such shortcomings includes the introduction of McStas inspired supermirrors in MCNPX. In the present paper different approaches to interface MCNPX and McStas are presented and applied to a simple test case. The direct coupling between MCNPX and McStas allows for more accurate simulations of e.g. complex moderator geometries, backgrounds, interference between beam-lines as well as shielding requirements along the neutron guides.

Double differential neutron spectra generated by the interaction of a 12 MeV/nucleon 36S beam on a thick natCu target

NASA Astrophysics Data System (ADS)

Trinh, N. D.; Fadil, M.; Lewitowicz, M.; Ledoux, X.; Laurent, B.; Thomas, J.-C.; Clerc, T.; Desmezières, V.; Dupuis, M.; Madeline, A.; Dessay, E.; Grinyer, G. F.; Grinyer, J.; Menard, N.; Porée, F.; Achouri, L.; Delaunay, F.; Parlog, M.

2018-07-01

Double differential neutron spectra (energy, angle) originating from a thick natCu target bombarded by a 12 MeV/nucleon 36S16+ beam were measured by the activation method and the Time-of-flight technique at the Grand Accélérateur National d'Ions Lourds (GANIL). A neutron spectrum unfolding algorithm combining the SAND-II iterative method and Monte-Carlo techniques was developed for the analysis of the activation results that cover a wide range of neutron energies. It was implemented into a graphical user interface program, called GanUnfold. The experimental neutron spectra are compared to Monte-Carlo simulations performed using the PHITS and FLUKA codes.

3D-radiative transfer in terrestrial atmosphere: An efficient parallel numerical procedure

NASA Astrophysics Data System (ADS)

Bass, L. P.; Germogenova, T. A.; Nikolaeva, O. V.; Kokhanovsky, A. A.; Kuznetsov, V. S.

2003-04-01

Light propagation and scattering in terrestrial atmosphere is usually studied in the framework of the 1D radiative transfer theory [1]. However, in reality particles (e.g., ice crystals, solid and liquid aerosols, cloud droplets) are randomly distributed in 3D space. In particular, their concentrations vary both in vertical and horizontal directions. Therefore, 3D effects influence modern cloud and aerosol retrieval procedures, which are currently based on the 1D radiative transfer theory. It should be pointed out that the standard radiative transfer equation allows to study these more complex situations as well [2]. In recent year the parallel version of the 2D and 3D RADUGA code has been developed. This version is successfully used in gammas and neutrons transport problems [3]. Applications of this code to radiative transfer in atmosphere problems are contained in [4]. Possibilities of code RADUGA are presented in [5]. The RADUGA code system is an universal solver of radiative transfer problems for complicated models, including 2D and 3D aerosol and cloud fields with arbitrary scattering anisotropy, light absorption, inhomogeneous underlying surface and topography. Both delta type and distributed light sources can be accounted for in the framework of the algorithm developed. The accurate numerical procedure is based on the new discrete ordinate SWDD scheme [6]. The algorithm is specifically designed for parallel supercomputers. The version RADUGA 5.1(P) can run on MBC1000M [7] (768 processors with 10 Gb of hard disc memory for each processor). The peak productivity is equal 1 Tfl. Corresponding scalar version RADUGA 5.1 is working on PC. As a first example of application of the algorithm developed, we have studied the shadowing effects of clouds on neighboring cloudless atmosphere, depending on the cloud optical thickness, surface albedo, and illumination conditions. This is of importance for modern satellite aerosol retrieval algorithms development. [1] Sobolev, V. V., 1972: Light scattering in planetary atmosphere, M.:Nauka. [2] Evans, K. F., 1998: The spherical harmonic discrete ordinate method for three dimensional atmospheric radiative transfer, J. Atmos. Sci., 55, 429 446. [3] L.P. Bass, T.A. Germogenova, V.S. Kuznetsov, O.V. Nikolaeva. RADUGA 5.1 and RADUGA 5.1(P) codes for stationary transport equation solution in 2D and 3D geometries on one and multiprocessors computers. Report on seminar “Algorithms and Codes for neutron physical of nuclear reactor calculations” (Neutronica 2001), Obninsk, Russia, 30 October 2 November 2001. [4] T.A. Germogenova, L.P. Bass, V.S. Kuznetsov, O.V. Nikolaeva. Mathematical modeling on parallel computers solar and laser radiation transport in 3D atmosphere. Report on International Symposium CIS countries “Atmosphere radiation”, 18 21 June 2002, St. Peterburg, Russia, p. 15 16. [5] L.P. Bass, T.A. Germogenova, O.V. Nikolaeva, V.S. Kuznetsov. Radiative Transfer Universal 2D 3D Code RADUGA 5.1(P) for Multiprocessor Computer. Abstract. Poster report on this Meeting. [6] L.P. Bass, O.V. Nikolaeva. Correct calculation of Angular Flux Distribution in Strongly Heterogeneous Media and Voids. Proc. of Joint International Conference on Mathematical Methods and Supercomputing for Nuclear Applications, Saratoga Springs, New York, October 5 9, 1997, p. 995 1004. [7] http://www/jscc.ru

New estimation method of neutron skyshine for a high-energy particle accelerator

NASA Astrophysics Data System (ADS)

Oh, Joo-Hee; Jung, Nam-Suk; Lee, Hee-Seock; Ko, Seung-Kook

2016-09-01

A skyshine is the dominant component of the prompt radiation at off-site. Several experimental studies have been done to estimate the neutron skyshine at a few accelerator facilities. In this work, the neutron transports from a source place to off-site location were simulated using the Monte Carlo codes, FLUKA and PHITS. The transport paths were classified as skyshine, direct (transport), groundshine and multiple-shine to understand the contribution of each path and to develop a general evaluation method. The effect of each path was estimated in the view of the dose at far locations. The neutron dose was calculated using the neutron energy spectra obtained from each detector placed up to a maximum of 1 km from the accelerator. The highest altitude of the sky region in this simulation was set as 2 km from the floor of the accelerator facility. The initial model of this study was the 10 GeV electron accelerator, PAL-XFEL. Different compositions and densities of air, soil and ordinary concrete were applied in this calculation, and their dependences were reviewed. The estimation method used in this study was compared with the well-known methods suggested by Rindi, Stevenson and Stepleton, and also with the simple code, SHINE3. The results obtained using this method agreed well with those using Rindi's formula.

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

NASA Astrophysics Data System (ADS)

Burns, Kimberly Ann

The accurate and efficient simulation of coupled neutron-photon problems is necessary for several important radiation detection applications. Examples include the detection of nuclear threats concealed in cargo containers and prompt gamma neutron activation analysis for nondestructive determination of elemental composition of unknown samples. In these applications, high-resolution gamma-ray spectrometers are used to preserve as much information as possible about the emitted photon flux, which consists of both continuum and characteristic gamma rays with discrete energies. Monte Carlo transport is the most commonly used modeling tool for this type of problem, but computational times for many problems can be prohibitive. This work explores the use of coupled Monte Carlo-deterministic methods for the simulation of neutron-induced photons for high-resolution gamma-ray spectroscopy applications. RAdiation Detection Scenario Analysis Toolbox (RADSAT), a code which couples deterministic and Monte Carlo transport to perform radiation detection scenario analysis in three dimensions [1], was used as the building block for the methods derived in this work. RADSAT was capable of performing coupled deterministic-Monte Carlo simulations for gamma-only and neutron-only problems. The purpose of this work was to develop the methodology necessary to perform coupled neutron-photon calculations and add this capability to RADSAT. Performing coupled neutron-photon calculations requires four main steps: the deterministic neutron transport calculation, the neutron-induced photon spectrum calculation, the deterministic photon transport calculation, and the Monte Carlo detector response calculation. The necessary requirements for each of these steps were determined. A major challenge in utilizing multigroup deterministic transport methods for neutron-photon problems was maintaining the discrete neutron-induced photon signatures throughout the simulation. Existing coupled neutron-photon cross-section libraries and the methods used to produce neutron-induced photons were unsuitable for high-resolution gamma-ray spectroscopy applications. Central to this work was the development of a method for generating multigroup neutron-photon cross-sections in a way that separates the discrete and continuum photon emissions so the neutron-induced photon signatures were preserved. The RADSAT-NG cross-section library was developed as a specialized multigroup neutron-photon cross-section set for the simulation of high-resolution gamma-ray spectroscopy applications. The methodology and cross sections were tested using code-to-code comparison with MCNP5 [2] and NJOY [3]. A simple benchmark geometry was used for all cases compared with MCNP. The geometry consists of a cubical sample with a 252Cf neutron source on one side and a HPGe gamma-ray spectrometer on the opposing side. Different materials were examined in the cubical sample: polyethylene (C2H4), P, N, O, and Fe. The cross sections for each of the materials were compared to cross sections collapsed using NJOY. Comparisons of the volume-averaged neutron flux within the sample, volume-averaged photon flux within the detector, and high-purity gamma-ray spectrometer response (only for polyethylene) were completed using RADSAT and MCNP. The code-to-code comparisons show promising results for the coupled Monte Carlo-deterministic method. The RADSAT-NG cross-section production method showed good agreement with NJOY for all materials considered although some additional work is needed in the resonance region and in the first and last energy bin. Some cross section discrepancies existed in the lowest and highest energy bin, but the overall shape and magnitude of the two methods agreed. For the volume-averaged photon flux within the detector, typically the five most intense lines agree to within approximately 5% of the MCNP calculated flux for all of materials considered. The agreement in the code-to-code comparisons cases demonstrates a proof-of-concept of the method for use in RADSAT for coupled neutron-photon problems in high-resolution gamma-ray spectroscopy applications. One of the primary motivators for using the coupled method over pure Monte Carlo method is the potential for significantly lower computational times. For the code-to-code comparison cases, the run times for RADSAT were approximately 25--500 times shorter than for MCNP, as shown in Table 1. This was assuming a 40 mCi 252Cf neutron source and 600 seconds of "real-world" measurement time. The only variance reduction technique implemented in the MCNP calculation was forward biasing of the source toward the sample target. Improved MCNP runtimes could be achieved with the addition of more advanced variance reduction techniques.

Application of a Java-based, univel geometry, neutral particle Monte Carlo code to the searchlight problem

DOE Office of Scientific and Technical Information (OSTI.GOV)

Charles A. Wemple; Joshua J. Cogliati

2005-04-01

A univel geometry, neutral particle Monte Carlo transport code, written entirely in the Java programming language, is under development for medical radiotherapy applications. The code uses ENDF-VI based continuous energy cross section data in a flexible XML format. Full neutron-photon coupling, including detailed photon production and photonuclear reactions, is included. Charged particle equilibrium is assumed within the patient model so that detailed transport of electrons produced by photon interactions may be neglected. External beam and internal distributed source descriptions for mixed neutron-photon sources are allowed. Flux and dose tallies are performed on a univel basis. A four-tap, shift-register-sequence random numbermore » generator is used. Initial verification and validation testing of the basic neutron transport routines is underway. The searchlight problem was chosen as a suitable first application because of the simplicity of the physical model. Results show excellent agreement with analytic solutions. Computation times for similar numbers of histories are comparable to other neutron MC codes written in C and FORTRAN.« less

Reanalysis of tritium production in a sphere of /sup 6/LiD irradiated by 14-MeV neutrons

DOE Office of Scientific and Technical Information (OSTI.GOV)

Fawcett, L.R. Jr.

1985-08-01

Tritium production and activation of radiochemical detector foils in a sphere of /sup 6/LiD irradiated by a central source of 14-MeV neutrons has been reanalyzed. The /sup 6/LiD sphere consisted of 10 solid hemispherical nested shells with ampules of /sup 6/LiH, /sup 7/LiH, and activation foils located 2.2, 5, 7.7, 12.6, 20, and 30 cm from the center. The Los Alamos Monte Carlo Neutron Photon Transport Code (MCNP) was used to calculate neutron transport through the /sup 6/LiD, tritium production in the ampules, and foil activation. The MCNP input model was three-dimensional and employed ENDF/B-V cross sections for transport, tritiummore » production, and (where available) foil activation. The reanalyzed experimentally observed-to-calculated values of tritium production were 1.053 +- 2.1% in /sup 6/LiH and 0.999 +- 2.1% in /sup 7/LiH. The recalculated foil activation observed-to-calculated ratios were not generally improved over those reported in the original analysis.« less

Nuclear radiation analysis

NASA Technical Reports Server (NTRS)

Knies, R. J.; Byrn, N. R.; Smith, H. T.

1972-01-01

A study program of radiation shielding against the deleterious effects of nuclear radiation on man and equipment is reported. The methods used to analyze the radiation environment from bremsstrahlung photons are discussed along with the methods employed by transport code users. The theory and numerical methods used to solve transport of neutrons and gammas are described, and the neutron and cosmic fluxes that would be present on the gamma-ray telescope were analyzed.

Development of a point-kinetic verification scheme for nuclear reactor applications

DOE Office of Scientific and Technical Information (OSTI.GOV)

Demazière, C., E-mail: demaz@chalmers.se; Dykin, V.; Jareteg, K.

In this paper, a new method that can be used for checking the proper implementation of time- or frequency-dependent neutron transport models and for verifying their ability to recover some basic reactor physics properties is proposed. This method makes use of the application of a stationary perturbation to the system at a given frequency and extraction of the point-kinetic component of the system response. Even for strongly heterogeneous systems for which an analytical solution does not exist, the point-kinetic component follows, as a function of frequency, a simple analytical form. The comparison between the extracted point-kinetic component and its expectedmore » analytical form provides an opportunity to verify and validate neutron transport solvers. The proposed method is tested on two diffusion-based codes, one working in the time domain and the other working in the frequency domain. As long as the applied perturbation has a non-zero reactivity effect, it is demonstrated that the method can be successfully applied to verify and validate time- or frequency-dependent neutron transport solvers. Although the method is demonstrated in the present paper in a diffusion theory framework, higher order neutron transport methods could be verified based on the same principles.« less

Calculating the Responses of Self-Powered Radiation Detectors.

NASA Astrophysics Data System (ADS)

Thornton, D. A.

Available from UMI in association with The British Library. The aim of this research is to review and develop the theoretical understanding of the responses of Self -Powered Radiation Detectors (SPDs) in Pressurized Water Reactors (PWRs). Two very different models are considered. A simple analytic model of the responses of SPDs to neutrons and gamma radiation is presented. It is a development of the work of several previous authors and has been incorporated into a computer program (called GENSPD), the predictions of which have been compared with experimental and theoretical results reported in the literature. Generally, the comparisons show reasonable consistency; where there is poor agreement explanations have been sought and presented. Two major limitations of analytic models have been identified; neglect of current generation in insulators and over-simplified electron transport treatments. Both of these are developed in the current work. A second model based on the Explicit Representation of Radiation Sources and Transport (ERRST) is presented and evaluated for several SPDs in a PWR at beginning of life. The model incorporates simulation of the production and subsequent transport of neutrons, gamma rays and electrons, both internal and external to the detector. Neutron fluxes and fuel power ratings have been evaluated with core physics calculations. Neutron interaction rates in assembly and detector materials have been evaluated in lattice calculations employing deterministic transport and diffusion methods. The transport of the reactor gamma radiation has been calculated with Monte Carlo, adjusted diffusion and point-kernel methods. The electron flux associated with the reactor gamma field as well as the internal charge deposition effects of the transport of photons and electrons have been calculated with coupled Monte Carlo calculations of photon and electron transport. The predicted response of a SPD is evaluated as the sum of contributions from individual response mechanisms.

Study of the response of a lithium yttrium borate scintillator based neutron rem counter by Monte Carlo radiation transport simulations

NASA Astrophysics Data System (ADS)

Sunil, C.; Tyagi, Mohit; Biju, K.; Shanbhag, A. A.; Bandyopadhyay, T.

2015-12-01

The scarcity and the high cost of 3He has spurred the use of various detectors for neutron monitoring. A new lithium yttrium borate scintillator developed in BARC has been studied for its use in a neutron rem counter. The scintillator is made of natural lithium and boron, and the yield of reaction products that will generate a signal in a real time detector has been studied by FLUKA Monte Carlo radiation transport code. A 2 cm lead introduced to enhance the gamma rejection shows no appreciable change in the shape of the fluence response or in the yield of reaction products. The fluence response when normalized at the average energy of an Am-Be neutron source shows promise of being used as rem counter.

Measurement of the Neutron Beta Decay Lifetime using Magnetically Trapped Ultracold Neutrons

NASA Astrophysics Data System (ADS)

Adamek, Evan Robert

The neutron lifetime is an important parameter in the Standard Model of particle physics, with influences on the electroweak interaction and on Big Bang nucleosynthesis. Measurements of this quantity in cold beam experiments and in experiments using ultracold neutrons (UCN) disagree; this discrepancy may indicate that these measurements possess unaccounted-for systematic errors. The UCNtau experiment at Los Alamos Neutron Science Center (LANSCe) utilizes an asymmetrical magneto-gravitational storage volume with an in-situ vanadium detector. This setup is designed to either avoid or control many of the weaknesses that reduce systematic precision in other UCN lifetime experiments. Controlling for the many measurable errors requires detailed calculation and simulation, aided, for example, by the Geant4 Monte Carlo particle transport toolkit, which has been used to create a high fidelity model of the UCNtau experiment for modeling UCN transport, storage, and detection. Through the course of running the experiment, improvements in knowledge of particle measurement have led to improvements to the transport and to the detectors used in various parts of the experiment. With the experimental setup optimized to account for the subtleties of the measurement, the 2014-2015 beam period at LANSCe generated 85 measurement runs from which we could calculate the storage lifetime. Careful analysis of the effects of background on the vanadium detector assembly allowed for elimination of undesired signal and allowed for the extraction of a preliminary value for the neutron lifetime and the determination of areas to improve for the following run cycle.

Neutrons in proton pencil beam scanning: parameterization of energy, quality factors and RBE

NASA Astrophysics Data System (ADS)

Schneider, Uwe; Hälg, Roger A.; Baiocco, Giorgio; Lomax, Tony

2016-08-01

The biological effectiveness of neutrons produced during proton therapy in inducing cancer is unknown, but potentially large. In particular, since neutron biological effectiveness is energy dependent, it is necessary to estimate, besides the dose, also the energy spectra, in order to obtain quantities which could be a measure of the biological effectiveness and test current models and new approaches against epidemiological studies on cancer induction after proton therapy. For patients treated with proton pencil beam scanning, this work aims to predict the spatially localized neutron energies, the effective quality factor, the weighting factor according to ICRP, and two RBE values, the first obtained from the saturation corrected dose mean lineal energy and the second from DSB cluster induction. A proton pencil beam was Monte Carlo simulated using GEANT. Based on the simulated neutron spectra for three different proton beam energies a parameterization of energy, quality factors and RBE was calculated. The pencil beam algorithm used for treatment planning at PSI has been extended using the developed parameterizations in order to calculate the spatially localized neutron energy, quality factors and RBE for each treated patient. The parameterization represents the simple quantification of neutron energy in two energy bins and the quality factors and RBE with a satisfying precision up to 85 cm away from the proton pencil beam when compared to the results based on 3D Monte Carlo simulations. The root mean square error of the energy estimate between Monte Carlo simulation based results and the parameterization is 3.9%. For the quality factors and RBE estimates it is smaller than 0.9%. The model was successfully integrated into the PSI treatment planning system. It was found that the parameterizations for neutron energy, quality factors and RBE were independent of proton energy in the investigated energy range of interest for proton therapy. The pencil beam algorithm has been extended using the developed parameterizations in order to calculate the neutron energy, quality factor and RBE.

Neutrons in proton pencil beam scanning: parameterization of energy, quality factors and RBE.

PubMed

Schneider, Uwe; Hälg, Roger A; Baiocco, Giorgio; Lomax, Tony

2016-08-21

The biological effectiveness of neutrons produced during proton therapy in inducing cancer is unknown, but potentially large. In particular, since neutron biological effectiveness is energy dependent, it is necessary to estimate, besides the dose, also the energy spectra, in order to obtain quantities which could be a measure of the biological effectiveness and test current models and new approaches against epidemiological studies on cancer induction after proton therapy. For patients treated with proton pencil beam scanning, this work aims to predict the spatially localized neutron energies, the effective quality factor, the weighting factor according to ICRP, and two RBE values, the first obtained from the saturation corrected dose mean lineal energy and the second from DSB cluster induction. A proton pencil beam was Monte Carlo simulated using GEANT. Based on the simulated neutron spectra for three different proton beam energies a parameterization of energy, quality factors and RBE was calculated. The pencil beam algorithm used for treatment planning at PSI has been extended using the developed parameterizations in order to calculate the spatially localized neutron energy, quality factors and RBE for each treated patient. The parameterization represents the simple quantification of neutron energy in two energy bins and the quality factors and RBE with a satisfying precision up to 85 cm away from the proton pencil beam when compared to the results based on 3D Monte Carlo simulations. The root mean square error of the energy estimate between Monte Carlo simulation based results and the parameterization is 3.9%. For the quality factors and RBE estimates it is smaller than 0.9%. The model was successfully integrated into the PSI treatment planning system. It was found that the parameterizations for neutron energy, quality factors and RBE were independent of proton energy in the investigated energy range of interest for proton therapy. The pencil beam algorithm has been extended using the developed parameterizations in order to calculate the neutron energy, quality factor and RBE.

High-energy neutron depth-dose distribution experiment.

PubMed

Ferenci, M S; Hertel, N E

2003-01-01

A unique set of high-energy neutron depth-dose benchmark experiments were performed at the Los Alamos Neutron Science Center/Weapons Neutron Research (LANSCE/WNR) complex. The experiments consisted of filtered neutron beams with energies up to 800 MeV impinging on a 30 x 30 x 30 cm3 liquid, tissue-equivalent phantom. The absorbed dose was measured in the phantom at various depths with tissue-equivalent ion chambers. This experiment is intended to serve as a benchmark experiment for the testing of high-energy radiation transport codes for the international radiation protection community.

Strategy for the absolute neutron emission measurement on ITER.

PubMed

Sasao, M; Bertalot, L; Ishikawa, M; Popovichev, S

2010-10-01

Accuracy of 10% is demanded to the absolute fusion measurement on ITER. To achieve this accuracy, a functional combination of several types of neutron measurement subsystem, cross calibration among them, and in situ calibration are needed. Neutron transport calculation shows the suitable calibration source is a DT/DD neutron generator of source strength higher than 10(10) n/s (neutron/second) for DT and 10(8) n/s for DD. It will take eight weeks at the minimum with this source to calibrate flux monitors, profile monitors, and the activation system.

Study of neutron generation in the compact tokamak TUMAN-3M in support of a tokamak-based fusion neutron source

NASA Astrophysics Data System (ADS)

Kornev, V. A.; Askinazi, L. G.; Belokurov, A. A.; Chernyshev, F. V.; Lebedev, S. V.; Melnik, A. D.; Shabelsky, A. A.; Tukachinsky, A. S.; Zhubr, N. A.

2017-12-01

The paper presents DD neutron flux measurements in neutron beam injection (NBI) experiments aimed at the optimization of target plasma and heating beam parameters to achieve maximum neutron flux in the TUMAN-3M compact tokamak. Two ion sources of different design were used, which allowed the separation of the beam’s energy and power influence on the neutron rate. Using the database of experiments performed with the two ion sources, an empirical scaling was derived describing the neutron rate dependence on the target plasma and heating beam parameters. Numerical modeling of the neutron rate in the NBI experiments performed using the ASTRA transport code showed good agreement with the scaling.

WINDOWS: a program for the analysis of spectral data foil activation measurements

DOE Office of Scientific and Technical Information (OSTI.GOV)

Stallmann, F.W.; Eastham, J.F.; Kam, F.B.K.

The computer program WINDOWS together with its subroutines is described for the analysis of neutron spectral data foil activation measurements. In particular, the unfolding of the neutron differential spectrum, estimated windows and detector contributions, upper and lower bounds for an integral response, and group fluxes obtained from neutron transport calculations. 116 references. (JFP)

Time dependent worldwide distribution of atmospheric neutrons and of their products. I, II, III.

NASA Technical Reports Server (NTRS)

Merker, M.; Light, E. S.; Verschell, H. J.; Mendell, R. B.; Korff, S. A.

1973-01-01

Review of the experimental results obtained in a series of measurements of the fast neutron cosmic ray spectrum by means of high-altitude balloons and aircraft. These results serve as a basis for checking a Monte Carlo calculation of the entire neutron distribution and its products. A calculation of neutron production and transport in the earth's atmosphere is then discussed for the purpose of providing a detailed description of the morphology of secondary neutron components. Finally, an analysis of neutron observations during solar particle events is presented. The Monte Carlo output is used to estimate the contribution of flare particles to fluctuations in the steady state neutron distributions.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Marleau, Peter; Monterial, Mateusz; Clarke, Shaun

A Bayesian approach is proposed for pulse shape discrimination of photons and neutrons in liquid organic scinitillators. Instead of drawing a decision boundary, each pulse is assigned a photon or neutron confidence probability. In addition, this allows for photon and neutron classification on an event-by-event basis. The sum of those confidence probabilities is used to estimate the number of photon and neutron instances in the data. An iterative scheme, similar to an expectation-maximization algorithm for Gaussian mixtures, is used to infer the ratio of photons-to-neutrons in each measurement. Therefore, the probability space adapts to data with varying photon-to-neutron ratios. Amore » time-correlated measurement of Am–Be and separate measurements of 137Cs, 60Co and 232Th photon sources were used to construct libraries of neutrons and photons. These libraries were then used to produce synthetic data sets with varying ratios of photons-to-neutrons. Probability weighted method that we implemented was found to maintain neutron acceptance rate of up to 90% up to photon-to-neutron ratio of 2000, and performed 9% better than the decision boundary approach. Furthermore, the iterative approach appropriately changed the probability space with an increasing number of photons which kept the neutron population estimate from unrealistically increasing.« less

Development of a transportable neutron activation analysis system to quantify manganese in bone in vivo: feasibility and methodology.

PubMed

Liu, Yingzi; Koltick, David; Byrne, Patrick; Wang, Haoyu; Zheng, Wei; Nie, Linda H

2013-12-01

This study was conducted to investigate the methodology and feasibility of developing a transportable neutron activation analysis (NAA) system to quantify manganese (Mn) in bone using a portable deuterium-deuterium (DD) neutron generator as the neutron source. Since a DD neutron generator was not available in our laboratory, a deuterium-tritium (DT) neutron generator was used to obtain experimental data and validate the results from Monte Carlo (MC) simulations. After validation, MC simulations using a DD generator as the neutron source were then conducted. Different types of moderators and reflectors were simulated, and the optimal thicknesses for the moderator and reflector were determined. To estimate the detection limit (DL) of the system, and to observe the interference of the magnesium (Mg) γ line at 844 keV to the Mn γ line at 847 keV, three hand phantoms with Mn concentrations of 30 parts per million (ppm), 150 ppm, and 500 ppm were made and irradiated by the DT generator system. The Mn signals in these phantoms were then measured using a 50% high-efficiency high-purity germanium (HPGe) detector. The DL was calculated to be about 4.4 ppm for the chosen irradiation, decay, and measurement time. This was calculated to be equivalent to a DL of about 3.3 ppm for the DD generator system. To achieve this DL with one 50% high-efficiency HPGe detector, the dose to the hand was simulated to be about 37 mSv, with the total body equivalent dose being about 23µSv. In conclusion, it is feasible to develop a transportable NAA system to quantify Mn in bone in vivo with an acceptable radiation exposure to the subject.

Development of a transportable neutron activation analysis system to quantify manganese in bone in vivo: feasibility and methodology

PubMed Central

Liu, Yingzi; Koltick, David; Byrne, Patrick; Wang, Haoyu; Zheng, Wei; Nie, Linda H

2014-01-01

This study was conducted to investigate the methodology and feasibility of developing a transportable neutron activation analysis (NAA) system to quantify manganese (Mn) in bone using a portable deuterium–deuterium (DD) neutron generator as the neutron source. Since a DD neutron generator was not available in our laboratory, a deuterium–tritium (DT) neutron generator was used to obtain experimental data and validate the results from Monte Carlo (MC) simulations. After validation, MC simulations using a DD generator as the neutron source were then conducted. Different types of moderators and reflectors were simulated, and the optimal thicknesses for the moderator and reflector were determined. To estimate the detection limit (DL) of the system, and to observe the interference of the magnesium (Mg) γ line at 844 keV to the Mn γ line at 847 keV, three hand phantoms with Mn concentrations of 30 parts per million (ppm), 150 ppm, and 500 ppm were made and irradiated by the DT generator system. The Mn signals in these phantoms were then measured using a 50% high-efficiency high-purity germanium (HPGe) detector. The DL was calculated to be about 4.4 ppm for the chosen irradiation, decay, and measurement time. This was calculated to be equivalent to a DL of about 3.3 ppm for the DD generator system. To achieve this DL with one 50% high-efficiency HPGe detector, the dose to the hand was simulated to be about 37 mSv, with the total body equivalent dose being about 23μSv. In conclusion, it is feasible to develop a transportable NAA system to quantify Mn in bone in vivo with an acceptable radiation exposure to the subject. PMID:24165395

Benchmark test of neutron transport calculations: indium, nickel, gold, europium, and cobalt activation with and without energy moderated fission neutrons by iron simulating the Hiroshima atomic bomb casing.

PubMed

Iwatani, K; Hoshi, M; Shizuma, K; Hiraoka, M; Hayakawa, N; Oka, T; Hasai, H

1994-10-01

A benchmark test of the Monte Carlo neutron and photon transport code system (MCNP) was performed using a bare- and energy-moderated 252Cf fission neutron source which was obtained by transmission through 10-cm-thick iron. An iron plate was used to simulate the effect of the Hiroshima atomic bomb casing. This test includes the activation of indium and nickel for fast neutrons and gold, europium, and cobalt for thermal and epithermal neutrons, which were inserted in the moderators. The latter two activations are also to validate 152Eu and 60Co activity data obtained from the atomic bomb-exposed specimens collected at Hiroshima and Nagasaki, Japan. The neutron moderators used were Lucite and Nylon 6 and the total thickness of each moderator was 60 cm or 65 cm. Measured activity data (reaction yield) of the neutron-irradiated detectors in these moderators decreased to about 1/1,000th or 1/10,000th, which corresponds to about 1,500 m ground distance from the hypocenter in Hiroshima. For all of the indium, nickel, and gold activity data, the measured and calculated values agreed within 25%, and the corresponding values for europium and cobalt were within 40%. From this study, the MCNP code was found to be accurate enough for the bare- and energy-moderated 252Cf neutron activation calculations of these elements using moderators containing hydrogen, carbon, nitrogen, and oxygen.

Spallation neutron production and the current intra-nuclear cascade and transport codes

NASA Astrophysics Data System (ADS)

Filges, D.; Goldenbaum, F.; Enke, M.; Galin, J.; Herbach, C.-M.; Hilscher, D.; Jahnke, U.; Letourneau, A.; Lott, B.; Neef, R.-D.; Nünighoff, K.; Paul, N.; Péghaire, A.; Pienkowski, L.; Schaal, H.; Schröder, U.; Sterzenbach, G.; Tietze, A.; Tishchenko, V.; Toke, J.; Wohlmuther, M.

A recent renascent interest in energetic proton-induced production of neutrons originates largely from the inception of projects for target stations of intense spallation neutron sources, like the planned European Spallation Source (ESS), accelerator-driven nuclear reactors, nuclear waste transmutation, and also from the application for radioactive beams. In the framework of such a neutron production, of major importance is the search for ways for the most efficient conversion of the primary beam energy into neutron production. Although the issue has been quite successfully addressed experimentally by varying the incident proton energy for various target materials and by covering a huge collection of different target geometries --providing an exhaustive matrix of benchmark data-- the ultimate challenge is to increase the predictive power of transport codes currently on the market. To scrutinize these codes, calculations of reaction cross-sections, hadronic interaction lengths, average neutron multiplicities, neutron multiplicity and energy distributions, and the development of hadronic showers are confronted with recent experimental data of the NESSI collaboration. Program packages like HERMES, LCS or MCNPX master the prevision of reaction cross-sections, hadronic interaction lengths, averaged neutron multiplicities and neutron multiplicity distributions in thick and thin targets for a wide spectrum of incident proton energies, geometrical shapes and materials of the target generally within less than 10% deviation, while production cross-section measurements for light charged particles on thin targets point out that appreciable distinctions exist within these models.

Neutron response function characterization of 4He scintillation detectors

DOE PAGES

Kelley, Ryan P.; Rolison, Lucas M.; Lewis, Jason M.; ...

2015-04-15

Time-of-flight measurements were conducted to characterize the neutron energy response of pressurized 4He fast neutron scintillation detectors for the first time, using the Van de Graaff generator at Ohio University. The time-of-flight spectra and pulse height distributions were measured. This data was used to determine the light output response function, which was found to be linear at energies below 3.5 MeV. The intrinsic efficiency of the detector as a function of incident energy was also calculated: the average efficiency up to 10 MeV was 3.1%, with a maximum efficiency of 6.6% at 1.05 MeV. Furthermore, these results will enable developmentmore » of neutron spectrum unfolding algorithms for neutron spectroscopy applications with these detectors.« less

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

PubMed

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

2016-12-01

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

Uncertainty in the delayed neutron fraction in fuel assembly depletion calculations

NASA Astrophysics Data System (ADS)

Aures, Alexander; Bostelmann, Friederike; Kodeli, Ivan A.; Velkov, Kiril; Zwermann, Winfried

2017-09-01

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

Sensitive and transportable gadolinium-core plastic scintillator sphere for neutron detection and counting

NASA Astrophysics Data System (ADS)

Dumazert, Jonathan; Coulon, Romain; Carrel, Frédérick; Corre, Gwenolé; Normand, Stéphane; Méchin, Laurence; Hamel, Matthieu

2016-08-01

Neutron detection forms a critical branch of nuclear-related issues, currently driven by the search for competitive alternative technologies to neutron counters based on the helium-3 isotope. The deployment of plastic scintillators shows a high potential for efficient detectors, safer and more reliable than liquids, more easily scalable and cost-effective than inorganic. In the meantime, natural gadolinium, through its 155 and mostly 157 isotopes, presents an exceptionally high interaction probability with thermal neutrons. This paper introduces a dual system including a metal gadolinium core inserted at the center of a high-scale plastic scintillator sphere. Incident fast neutrons are thermalized by the scintillator shell and then may be captured with a significant probability by gadolinium 155 and 157 nuclei in the core. The deposition of a sufficient fraction of the capture high-energy prompt gamma signature inside the scintillator shell will then allow discrimination from background radiations by energy threshold, and therefore neutron detection. The scaling of the system with the Monte Carlo MCNPX2.7 code was carried out according to a tradeoff between the moderation of incident fast neutrons and the probability of slow neutron capture by a moderate-cost metal gadolinium core. Based on the parameters extracted from simulation, a first laboratory prototype for the assessment of the detection method principle has been synthetized. The robustness and sensitivity of the neutron detection principle are then assessed by counting measurement experiments. Experimental results confirm the potential for a stable, highly sensitive, transportable and cost-efficient neutron detector and orientate future investigation toward promising axes.

Measurement and Interpretation of DT Neutron Emission from Tftr.

NASA Astrophysics Data System (ADS)

McCauley, John Scott, Jr.

A fast-ion diffusion coefficient of 0.1 +/- 0.1 m^2s ^{-1} has been deduced from the triton burnup neutron emission profile measured by a collimated array of helium-4 spectrometers. The experiment was performed with high-power deuterium discharges produced by Princeton University's Tokamak Fusion Test Reactor (TFTR). The fast ions monitored were the 1.0 MeV tritons produced from the d(d,t)p triton burnup reaction. These tritons "burn up" with deuterons and emit a 14 MeV neutron by the d(t, alpha)n reaction. The measured radial profiles of DT emission were compared with the predictions of a computer transport code. The ratio of the measured-to -calculated DT yield is typically 70%. The measured DT profile width is typically 5 cm larger than predicted by the transport code. The radial 14 MeV neutron profile was measured by a radial array of helium-4 recoil neutron spectrometers installed in the TFTR Multichannel Neutron Collimator (MCNC). The spectrometers are capable of measuring the primary and secondary neutron fluxes from deuterium discharges. The response to 14 MeV neutrons of the array has been measured by cross calibrating with the MCNC ZnS detector array when the emission from TFTR is predominantly DT neutrons. The response was also checked by comparing a model of the recoil spectrum based on nuclear physics data to the observed spectrum from ^{252 }Cf, ^{238}Pu -Be, and DT neutron sources. Extensions of this diagnostic to deuterium-tritium plasma and the implications for fusion research are discussed.

Parallel simulations of Grover's algorithm for closest match search in neutron monitor data

NASA Astrophysics Data System (ADS)

Kussainov, Arman; White, Yelena

We are studying the parallel implementations of Grover's closest match search algorithm for neutron monitor data analysis. This includes data formatting, and matching quantum parameters to a conventional structure of a chosen programming language and selected experimental data type. We have employed several workload distribution models based on acquired data and search parameters. As a result of these simulations, we have an understanding of potential problems that may arise during configuration of real quantum computational devices and the way they could run tasks in parallel. The work was supported by the Science Committee of the Ministry of Science and Education of the Republic of Kazakhstan Grant #2532/GF3.

On three-dimensional reconstruction of a neutron/x-ray source from very few two-dimensional projections

DOE PAGES

Volegov, P. L.; Danly, C. R.; Merrill, F. E.; ...

2015-11-24

The neutron imaging system at the National Ignition Facility is an important diagnostic tool for measuring the two-dimensional size and shape of the source of neutrons produced in the burning deuterium-tritium plasma during the stagnation phase of inertial confinement fusion implosions. Few two-dimensional projections of neutronimages are available to reconstruct the three-dimensionalneutron source. In our paper, we present a technique that has been developed for the 3Dreconstruction of neutron and x-raysources from a minimal number of 2D projections. Here, we present the detailed algorithms used for this characterization and the results of reconstructedsources from experimental data collected at Omega.

Experimental characterization of the Advanced Liquid Hydrogen Cold Neutron Source spectrum of the NBSR reactor at the NIST Center for Neutron Research

NASA Astrophysics Data System (ADS)

Cook, J. C.; Barker, J. G.; Rowe, J. M.; Williams, R. E.; Gagnon, C.; Lindstrom, R. M.; Ibberson, R. M.; Neumann, D. A.

2015-08-01

The recent expansion of the National Institute of Standards and Technology (NIST) Center for Neutron Research facility has offered a rare opportunity to perform an accurate measurement of the cold neutron spectrum at the exit of a newly-installed neutron guide. Using a combination of a neutron time-of-flight measurement, a gold foil activation measurement, and Monte Carlo simulation of the neutron guide transmission, we obtain the most reliable experimental characterization of the Advanced Liquid Hydrogen Cold Neutron Source brightness to date. Time-of-flight measurements were performed at three distinct fuel burnup intervals, including one immediately following reactor startup. Prior to the latter measurement, the hydrogen was maintained in a liquefied state for an extended period in an attempt to observe an initial radiation-induced increase of the ortho (o)-hydrogen fraction. Since para (p)-hydrogen has a small scattering cross-section for neutron energies below 15 meV (neutron wavelengths greater than about 2.3 Å), changes in the o- p hydrogen ratio and in the void distribution in the boiling hydrogen influence the spectral distribution. The nature of such changes is simulated with a continuous-energy, Monte Carlo radiation-transport code using 20 K o and p hydrogen scattering kernels and an estimated hydrogen density distribution derived from an analysis of localized heat loads. A comparison of the transport calculations with the mean brightness function resulting from the three measurements suggests an overall o- p ratio of about 17.5(±1) % o- 82.5% p for neutron energies<15 meV, a significantly lower ortho concentration than previously assumed.

Secondary neutron-production cross sections from heavy-ioninteractions in composite targets.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Heilbronn, L.; Iwata, Y.; Iwase,H.

Secondary neutron-production cross-sections have been measured from interactions of 290 MeV/nucleon C and 600 MeV/nucleon Ne in a target composed of simulated Martian regolith and polyethylene, and from 400 MeV/nucleon Ne interactions in wall material from the International Space Station. The data were measured between 5 and 80 deg in the laboratory. We report the double-differential cross sections, angular distributions, and total neutron-production cross sections from all three systems. The spectra from all three systems exhibit behavior previously reported in other heavy-ion, neutron production experiments; namely, a peak at forward angles near the energy corresponding to the beam velocity, withmore » the remaining spectra generated by pre-equilibrium and equilibrium processes. The double differential cross sections are fitted with a moving-source parameterization. Also reported are the data without corrections for neutron flux attenuation in the target and other intervening materials, and for neutron production in non-target materials near the target position. These uncorrected spectra are compared with SHIELD-HIT and PHITS transport model calculations. The transport model calculations reproduce the spectral shapes well, but, on average, underestimate the magnitudes of the cross sections.« less

Secondary neutron-production cross sections from heavy-ion interactions in composite targets

DOE Office of Scientific and Technical Information (OSTI.GOV)

Heilbronn, L.; Iwata, Y.; Murakami, T.

Secondary neutron-production cross sections have been measured from interactions of 290 MeV/nucleon C and 600 MeV/nucleon Ne in a target composed of simulated Martian regolith and polyethylene, and from 400 MeV/nucleon Ne interactions in wall material from the International Space Station. The data were measured between 5 deg. and 80 deg. in the laboratory. We report the double-differential cross sections, angular distributions, and total neutron-production cross sections from all three systems. The spectra from all three systems exhibit behavior previously reported in other heavy-ion neutron-production experiments, namely, a peak at forward angles near the energy corresponding to the beam velocity,more » with the remaining spectra generated by pre-equilibrium and equilibrium processes. The double-differential cross sections are fitted with a moving-source parametrization. Also reported are the data without corrections for neutron flux attenuation in the target and other intervening materials and for neutron production in nontarget materials near the target position. These uncorrected spectra are compared with SHIELD-HIT and PHITS transport model calculations. The transport model calculations reproduce the spectral shapes well but, on average, underestimate the magnitudes of the cross sections.« less

Neutron Imaging Reveals Internal Plant Hydraulic Dynamics

DOE Office of Scientific and Technical Information (OSTI.GOV)

Warren, Jeffrey; Bilheux, Hassina Z; Kang, Misun

2013-01-01

Many terrestrial ecosystem processes are constrained by water availability and transport within the soil. Knowledge of plant water fluxes is thus critical for assessing mechanistic processes linked to biogeochemical cycles, yet resolution of root structure and xylem water transport dynamics has been a particularly daunting task for the ecologist. Through neutron imaging, we demonstrate the ability to non-invasively monitor individual root functionality and water fluxes within Zea mays L. (maize) and Panicum virgatum L. (switchgrass) seedlings growing in a sandy medium. Root structure and growth were readily imaged by neutron radiography and neutron computed tomography. Seedlings were irrigated with watermore » or deuterium oxide and imaged through time as a growth lamp was cycled on to alter leaf demand for water. Sub-millimeter scale resolution reveals timing and magnitudes of root water uptake, redistribution within the roots, and root-shoot hydraulic linkages, relationships not well characterized by other techniques.« less

Thermal neutron filter design for the neutron radiography facility at the LVR-15 reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Soltes, Jaroslav; Faculty of Nuclear Sciences and Physical Engineering, CTU in Prague,; Viererbl, Ladislav

2015-07-01

In 2011 a decision was made to build a neutron radiography facility at one of the unused horizontal channels of the LVR-15 research reactor in Rez, Czech Republic. One of the key conditions for operating an effective radiography facility is the delivery of a high intensity, homogeneous and collimated thermal neutron beam at the sample location. Additionally the intensity of fast neutrons has to be kept as low as possible as the fast neutrons may damage the detectors used for neutron imaging. As the spectrum in the empty horizontal channel roughly copies the spectrum in the reactor core, which hasmore » a high ratio of fast neutrons, neutron filter components have to be installed inside the channel in order to achieve desired beam parameters. As the channel design does not allow the instalment of complex filters and collimators, an optimal solution represent neutron filters made of large single-crystal ingots of proper material composition. Single-crystal silicon was chosen as a favorable filter material for its wide availability in sufficient dimensions. Besides its ability to reasonably lower the ratio of fast neutrons while still keeping high intensities of thermal neutrons, due to its large dimensions, it suits as a shielding against gamma radiation from the reactor core. For designing the necessary filter dimensions the Monte-Carlo MCNP transport code was used. As the code does not provide neutron cross-section libraries for thermal neutron transport through single-crystalline silicon, these had to be created by approximating the theory of thermal neutron scattering and modifying the original cross-section data which are provided with the code. Carrying out a series of calculations the filter thickness of 1 m proved good for gaining a beam with desired parameters and a low gamma background. After mounting the filter inside the channel several measurements of the neutron field were realized at the beam exit. The results have justified the expected calculated values. After the successful filter installing and a series of measurements, first test neutron radiography attempts with test samples could been carried out. (authors)« less

Applicability of a Bonner Shere technique for pulsed neutron in 120 GeV proton facility

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sanami, T.; Hagiwara, M.; Iwase, H.

2008-02-01

The data on neutron spectra and intensity behind shielding are important for radiation safety design of high-energy accelerators since neutrons are capable of penetrating thick shielding and activating materials. Corresponding particle transport codes--that involve physics models of neutron and other particle production, transportation, and interaction--have been developed and used world-wide [1-8]. The results of these codes have been ensured through plenty of comparisons with experimental results taken in simple geometries. For neutron generation and transport, several related experiments have been performed to measure neutron spectra, attenuation length and reaction rates behind shielding walls of various thicknesses and materials in energymore » range up to several hundred of MeV [9-11]. The data have been used to benchmark--and modify if needed--the simulation modes and parameters in the codes, as well as the reference data for radiation safety design. To obtain such kind of data above several hundred of MeV, Japan-Fermi National Accelerator Laboratory (FNAL) collaboration for shielding experiments has been started in 2007, based on suggestion from the specialist meeting of shielding, Shielding Aspects of Target, Irradiation Facilities (SATIF), because of very limited data available in high-energy region (see, for example, [12]). As a part of this shielding experiment, a set of Bonner sphere (BS) was tested at the antiproton production target facility (pbar target station) at FNAL to obtain neutron spectra induced by a 120-GeV proton beam in concrete and iron shielding. Generally, utilization of an active detector around high-energy accelerators requires an improvement on its readout to overcome burst of secondary radiation since the accelerator delivers an intense beam to a target in a short period after relatively long acceleration period. In this paper, we employ BS for a spectrum measurement of neutrons that penetrate the shielding wall of the pbar target station in FNAL.« less

The ENEA criticality accident dosimetry system: a contribution to the 2002 international intercomparison at the SILENE reactor.

PubMed

Gualdrini, G; Bedogni, R; Fantuzzi, E; Mariotti, F

2004-01-01

The present paper summarises the activity carried out at the ENEA Radiation Protection Institute for updating the methodologies employed for the evaluation of the neutron and photon dose to the exposed workers in case of a criticality accident, in the framework of the 'International Intercomparison of Criticality Accident Dosimetry Systems' (Silène reactor, IRSN-CEA-Valduc June 2002). The evaluation of the neutron spectra and the neutron dosimetric quantities relies on activation detectors and on unfolding algorithms. Thermoluminescent detectors are employed for the gamma dose measurement. The work is aimed at accurately characterising the measurement system and, at the same time, testing the algorithms. Useful spectral information were included, based on Monte Carlo simulations, to take into account the potential accident scenarios of practical interest. All along this exercise intercomparison a particular attention was devoted to the 'traceability' of all the experimental and computational parameters and therefore, aimed at an easy treatment by the user.

A comparative study of history-based versus vectorized Monte Carlo methods in the GPU/CUDA environment for a simple neutron eigenvalue problem

NASA Astrophysics Data System (ADS)

Liu, Tianyu; Du, Xining; Ji, Wei; Xu, X. George; Brown, Forrest B.

2014-06-01

For nuclear reactor analysis such as the neutron eigenvalue calculations, the time consuming Monte Carlo (MC) simulations can be accelerated by using graphics processing units (GPUs). However, traditional MC methods are often history-based, and their performance on GPUs is affected significantly by the thread divergence problem. In this paper we describe the development of a newly designed event-based vectorized MC algorithm for solving the neutron eigenvalue problem. The code was implemented using NVIDIA's Compute Unified Device Architecture (CUDA), and tested on a NVIDIA Tesla M2090 GPU card. We found that although the vectorized MC algorithm greatly reduces the occurrence of thread divergence thus enhancing the warp execution efficiency, the overall simulation speed is roughly ten times slower than the history-based MC code on GPUs. Profiling results suggest that the slow speed is probably due to the memory access latency caused by the large amount of global memory transactions. Possible solutions to improve the code efficiency are discussed.

LASER APPLICATIONS AND OTHER TOPICS IN QUANTUM ELECTRONICS On control of kinematic parameters of ultracold neutrons in waveguides

NASA Astrophysics Data System (ADS)

Rivlin, Lev A.

2010-10-01

The possibility of controlling the kinematic parameters of ultracold neutrons (UCNs) is analysed by the example of a waveguide transfer and transformation of 2D images in ultracold neutrons and by the example of an increase in the concentration and deceleration/acceleration of ultracold neutrons during their transport in the waveguide with a variable cross section. The critical parameters of the problem are estimated, which indicates both consistency of the proposed approach and the emerging experimental limitations.

Bayesian approach to the analysis of neutron Brillouin scattering data on liquid metals

NASA Astrophysics Data System (ADS)

De Francesco, A.; Guarini, E.; Bafile, U.; Formisano, F.; Scaccia, L.

2016-08-01

When the dynamics of liquids and disordered systems at mesoscopic level is investigated by means of inelastic scattering (e.g., neutron or x ray), spectra are often characterized by a poor definition of the excitation lines and spectroscopic features in general and one important issue is to establish how many of these lines need to be included in the modeling function and to estimate their parameters. Furthermore, when strongly damped excitations are present, commonly used and widespread fitting algorithms are particularly affected by the choice of initial values of the parameters. An inadequate choice may lead to an inefficient exploration of the parameter space, resulting in the algorithm getting stuck in a local minimum. In this paper, we present a Bayesian approach to the analysis of neutron Brillouin scattering data in which the number of excitation lines is treated as unknown and estimated along with the other model parameters. We propose a joint estimation procedure based on a reversible-jump Markov chain Monte Carlo algorithm, which efficiently explores the parameter space, producing a probabilistic measure to quantify the uncertainty on the number of excitation lines as well as reliable parameter estimates. The method proposed could turn out of great importance in extracting physical information from experimental data, especially when the detection of spectral features is complicated not only because of the properties of the sample, but also because of the limited instrumental resolution and count statistics. The approach is tested on generated data set and then applied to real experimental spectra of neutron Brillouin scattering from a liquid metal, previously analyzed in a more traditional way.

BINP accelerator based epithermal neutron source.

PubMed

Aleynik, V; Burdakov, A; Davydenko, V; Ivanov, A; Kanygin, V; Kuznetsov, A; Makarov, A; Sorokin, I; Taskaev, S

2011-12-01

Innovative facility for neutron capture therapy has been built at BINP. This facility is based on compact vacuum insulation tandem accelerator designed to produce proton current up to 10 mA. Epithermal neutrons are proposed to be generated by 1.915-2.5 MeV protons bombarding a lithium target using (7)Li(p,n)(7)Be threshold reaction. In the article, diagnostic techniques for proton beam and neutrons developed are described, results of experiments on proton beam transport and neutron generation are shown, discussed, and plans are presented. Copyright © 2011 Elsevier Ltd. All rights reserved.

EXCALIBUR-at-CALIBAN: a neutron transmission experiment for {sup 238}U(n,n'{sub continuum}γ) nuclear data validation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bernard, David; Leconte, Pierre; Destouches, Christophe

2015-07-01

Two recent papers justified a new experimental program to give a new basis for the validation of {sup 238}U nuclear data, namely neutron induced inelastic scattering and transport codes at neutron fission energies. The general idea is to perform a neutron transmission experiment through natural uranium material. As shown by Hans Bethe, neutron transmissions measured by dosimetric responses are linked to inelastic cross sections. This paper describes the principle and the results of such an experience called EXCALIBUR performed recently (January and October 2014) at the CALIBAN reactor facility. (authors)

Electrostatic design and beam transport for a folded tandem electrostatic quadrupole accelerator facility for accelerator-based boron neutron capture therapy.

PubMed

Vento, V Thatar; Bergueiro, J; Cartelli, D; Valda, A A; Kreiner, A J

2011-12-01

Within the frame of an ongoing project to develop a folded Tandem-Electrostatic-Quadrupole (TESQ) accelerator facility for Accelerator-Based Boron Neutron Capture Therapy (AB-BNCT), we discuss here the electrostatic design of the machine, including the accelerator tubes with electrostatic quadrupoles and the simulations for the transport and acceleration of a high intensity beam. Copyright © 2011 Elsevier Ltd. All rights reserved.

Synergism of the method of characteristics and CAD technology for neutron transport calculation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Chen, Z.; Wang, D.; He, T.

2013-07-01

The method of characteristics (MOC) is a very popular methodology in neutron transport calculation and numerical simulation in recent decades for its unique advantages. One of the key problems determining whether the MOC can be applied in complicated and highly heterogeneous geometry is how to combine an effective geometry processing method with MOC. Most of the existing MOC codes describe the geometry by lines and arcs with extensive input data, such as circles, ellipses, regular polygons and combination of them. Thus they have difficulty in geometry modeling, background meshing and ray tracing for complicated geometry domains. In this study, amore » new idea making use of a CAD solid modeler MCAM which is a CAD/Image-based Automatic Modeling Program for Neutronics and Radiation Transport developed by FDS Team in China was introduced for geometry modeling and ray tracing of particle transport to remove these geometrical limitations mentioned above. The diamond-difference scheme was applied to MOC to reduce the spatial discretization error of the flat flux approximation in theory. Based on MCAM and MOC, a new MOC code was developed and integrated into SuperMC system, which is a Super Multi-function Computational system for neutronics and radiation simulation. The numerical testing results demonstrated the feasibility and effectiveness of the new idea for geometry treatment in SuperMC. (authors)« less

A Hybrid Method for Calculating TiO2 Concentrations Using Clementine UVVIS Data, and Verified with Lunar Prospector Neutron Spectrometer Data

NASA Technical Reports Server (NTRS)

Gillis, J. J.; Jolliff, B. L.; Elphic, R. C.; Maurice, S.; Feldman, W. C.; Lawrence, D. J.

2001-01-01

We present a new algorithm for extracting TiO2 concentrations from Clementine UVVIS data, which accounts for soil darkness and UV/VIS ratio. The accuracy of these TiO2 estimates are examined with Lunar Prospector thermal/epithermal neutron flux data. Additional information is contained in the original extended abstract.

Estimation of neutron energy distributions from prompt gamma emissions

NASA Astrophysics Data System (ADS)

Panikkath, Priyada; Udupi, Ashwini; Sarkar, P. K.

2017-11-01

A technique of estimating the incident neutron energy distribution from emitted prompt gamma intensities from a system exposed to neutrons is presented. The emitted prompt gamma intensities or the measured photo peaks in a gamma detector are related to the incident neutron energy distribution through a convolution of the response of the system generating the prompt gammas to mono-energetic neutrons. Presently, the system studied is a cylinder of high density polyethylene (HDPE) placed inside another cylinder of borated HDPE (BHDPE) having an outer Pb-cover and exposed to neutrons. The emitted five prompt gamma peaks from hydrogen, boron, carbon and lead can be utilized to unfold the incident neutron energy distribution as an under-determined deconvolution problem. Such an under-determined set of equations are solved using the genetic algorithm based Monte Carlo de-convolution code GAMCD. Feasibility of the proposed technique is demonstrated theoretically using the Monte Carlo calculated response matrix and intensities of emitted prompt gammas from the Pb-covered BHDPE-HDPE system in the case of several incident neutron spectra spanning different energy ranges.

High reliability - low noise radionuclide signature identification algorithms for border security applications

NASA Astrophysics Data System (ADS)

Lee, Sangkyu

Illicit trafficking and smuggling of radioactive materials and special nuclear materials (SNM) are considered as one of the most important recent global nuclear threats. Monitoring the transport and safety of radioisotopes and SNM are challenging due to their weak signals and easy shielding. Great efforts worldwide are focused at developing and improving the detection technologies and algorithms, for accurate and reliable detection of radioisotopes of interest in thus better securing the borders against nuclear threats. In general, radiation portal monitors enable detection of gamma and neutron emitting radioisotopes. Passive or active interrogation techniques, present and/or under the development, are all aimed at increasing accuracy, reliability, and in shortening the time of interrogation as well as the cost of the equipment. Equally important efforts are aimed at advancing algorithms to process the imaging data in an efficient manner providing reliable "readings" of the interiors of the examined volumes of various sizes, ranging from cargos to suitcases. The main objective of this thesis is to develop two synergistic algorithms with the goal to provide highly reliable - low noise identification of radioisotope signatures. These algorithms combine analysis of passive radioactive detection technique with active interrogation imaging techniques such as gamma radiography or muon tomography. One algorithm consists of gamma spectroscopy and cosmic muon tomography, and the other algorithm is based on gamma spectroscopy and gamma radiography. The purpose of fusing two detection methodologies per algorithm is to find both heavy-Z radioisotopes and shielding materials, since radionuclides can be identified with gamma spectroscopy, and shielding materials can be detected using muon tomography or gamma radiography. These combined algorithms are created and analyzed based on numerically generated images of various cargo sizes and materials. In summary, the three detection methodologies are fused into two algorithms with mathematical functions providing: reliable identification of radioisotopes in gamma spectroscopy; noise reduction and precision enhancement in muon tomography; and the atomic number and density estimation in gamma radiography. It is expected that these new algorithms maybe implemented at portal scanning systems with the goal to enhance the accuracy and reliability in detecting nuclear materials inside the cargo containers.

Investigation of Secondary Neutron Production in Large Space Vehicles for Deep Space

NASA Technical Reports Server (NTRS)

Rojdev, Kristina; Koontz, Steve; Reddell, Brandon; Atwell, William; Boeder, Paul

2016-01-01

Future NASA missions will focus on deep space and Mars surface operations with large structures necessary for transportation of crew and cargo. In addition to the challenges of manufacturing these large structures, there are added challenges from the space radiation environment and its impacts on the crew, electronics, and vehicle materials. Primary radiation from the sun (solar particle events) and from outside the solar system (galactic cosmic rays) interact with materials of the vehicle and the elements inside the vehicle. These interactions lead to the primary radiation being absorbed or producing secondary radiation (primarily neutrons). With all vehicles, the high-energy primary radiation is of most concern. However, with larger vehicles, there is more opportunity for secondary radiation production, which can be significant enough to cause concern. In a previous paper, we embarked upon our first steps toward studying neutron production from large vehicles by validating our radiation transport codes for neutron environments against flight data. The following paper will extend the previous work to focus on the deep space environment and the resulting neutron flux from large vehicles in this deep space environment.

A versatile UHV transport and measurement chamber for neutron reflectometry under UHV conditions

NASA Astrophysics Data System (ADS)

Syed Mohd, A.; Pütter, S.; Mattauch, S.; Koutsioubas, A.; Schneider, H.; Weber, A.; Brückel, T.

2016-12-01

We report on a versatile mini ultra-high vacuum (UHV) chamber which is designed to be used on the MAgnetic Reflectometer with high Incident Angle of the Jülich Centre for Neutron Science at Heinz Maier-Leibnitz Zentrum in Garching, Germany. Samples are prepared in the adjacent thin film laboratory by molecular beam epitaxy and moved into the compact chamber for transfer without exposure to ambient air. The chamber is based on DN 40 CF flanges and equipped with sapphire view ports, a small getter pump, and a wobble stick, which serves also as sample holder. Here, we present polarized neutron reflectivity measurements which have been performed on Co thin films at room temperature in UHV and in ambient air in a magnetic field of 200 mT and in the Q-range of 0.18 Å-1. The results confirm that the Co film is not contaminated during the polarized neutron reflectivity measurement. Herewith it is demonstrated that the mini UHV transport chamber also works as a measurement chamber which opens new possibilities for polarized neutron measurements under UHV conditions.

A versatile UHV transport and measurement chamber for neutron reflectometry under UHV conditions.

PubMed

Syed Mohd, A; Pütter, S; Mattauch, S; Koutsioubas, A; Schneider, H; Weber, A; Brückel, T

2016-12-01

We report on a versatile mini ultra-high vacuum (UHV) chamber which is designed to be used on the MAgnetic Reflectometer with high Incident Angle of the Jülich Centre for Neutron Science at Heinz Maier-Leibnitz Zentrum in Garching, Germany. Samples are prepared in the adjacent thin film laboratory by molecular beam epitaxy and moved into the compact chamber for transfer without exposure to ambient air. The chamber is based on DN 40 CF flanges and equipped with sapphire view ports, a small getter pump, and a wobble stick, which serves also as sample holder. Here, we present polarized neutron reflectivity measurements which have been performed on Co thin films at room temperature in UHV and in ambient air in a magnetic field of 200 mT and in the Q-range of 0.18 Å -1 . The results confirm that the Co film is not contaminated during the polarized neutron reflectivity measurement. Herewith it is demonstrated that the mini UHV transport chamber also works as a measurement chamber which opens new possibilities for polarized neutron measurements under UHV conditions.

Application of Bayes' theorem for pulse shape discrimination

NASA Astrophysics Data System (ADS)

Monterial, Mateusz; Marleau, Peter; Clarke, Shaun; Pozzi, Sara

2015-09-01

A Bayesian approach is proposed for pulse shape discrimination of photons and neutrons in liquid organic scinitillators. Instead of drawing a decision boundary, each pulse is assigned a photon or neutron confidence probability. This allows for photon and neutron classification on an event-by-event basis. The sum of those confidence probabilities is used to estimate the number of photon and neutron instances in the data. An iterative scheme, similar to an expectation-maximization algorithm for Gaussian mixtures, is used to infer the ratio of photons-to-neutrons in each measurement. Therefore, the probability space adapts to data with varying photon-to-neutron ratios. A time-correlated measurement of Am-Be and separate measurements of 137Cs, 60Co and 232Th photon sources were used to construct libraries of neutrons and photons. These libraries were then used to produce synthetic data sets with varying ratios of photons-to-neutrons. Probability weighted method that we implemented was found to maintain neutron acceptance rate of up to 90% up to photon-to-neutron ratio of 2000, and performed 9% better than the decision boundary approach. Furthermore, the iterative approach appropriately changed the probability space with an increasing number of photons which kept the neutron population estimate from unrealistically increasing.

Application of Bayes' theorem for pulse shape discrimination

DOE PAGES

Marleau, Peter; Monterial, Mateusz; Clarke, Shaun; ...

2015-06-14

A Bayesian approach is proposed for pulse shape discrimination of photons and neutrons in liquid organic scinitillators. Instead of drawing a decision boundary, each pulse is assigned a photon or neutron confidence probability. In addition, this allows for photon and neutron classification on an event-by-event basis. The sum of those confidence probabilities is used to estimate the number of photon and neutron instances in the data. An iterative scheme, similar to an expectation-maximization algorithm for Gaussian mixtures, is used to infer the ratio of photons-to-neutrons in each measurement. Therefore, the probability space adapts to data with varying photon-to-neutron ratios. Amore » time-correlated measurement of Am–Be and separate measurements of 137Cs, 60Co and 232Th photon sources were used to construct libraries of neutrons and photons. These libraries were then used to produce synthetic data sets with varying ratios of photons-to-neutrons. Probability weighted method that we implemented was found to maintain neutron acceptance rate of up to 90% up to photon-to-neutron ratio of 2000, and performed 9% better than the decision boundary approach. Furthermore, the iterative approach appropriately changed the probability space with an increasing number of photons which kept the neutron population estimate from unrealistically increasing.« less

Secondary Neutron Doses to Pediatric Patients During Intracranial Proton Therapy: Monte Carlo Simulation of the Neutron Energy Spectrum and its Organ Doses.

PubMed

Matsumoto, Shinnosuke; Koba, Yusuke; Kohno, Ryosuke; Lee, Choonsik; Bolch, Wesley E; Kai, Michiaki

2016-04-01

Proton therapy has the physical advantage of a Bragg peak that can provide a better dose distribution than conventional x-ray therapy. However, radiation exposure of normal tissues cannot be ignored because it is likely to increase the risk of secondary cancer. Evaluating secondary neutrons generated by the interaction of the proton beam with the treatment beam-line structure is necessary; thus, performing the optimization of radiation protection in proton therapy is required. In this research, the organ dose and energy spectrum were calculated from secondary neutrons using Monte Carlo simulations. The Monte Carlo code known as the Particle and Heavy Ion Transport code System (PHITS) was used to simulate the transport proton and its interaction with the treatment beam-line structure that modeled the double scattering body of the treatment nozzle at the National Cancer Center Hospital East. The doses of the organs in a hybrid computational phantom simulating a 5-y-old boy were calculated. In general, secondary neutron doses were found to decrease with increasing distance to the treatment field. Secondary neutron energy spectra were characterized by incident neutrons with three energy peaks: 1×10, 1, and 100 MeV. A block collimator and a patient collimator contributed significantly to organ doses. In particular, the secondary neutrons from the patient collimator were 30 times higher than those from the first scatter. These results suggested that proactive protection will be required in the design of the treatment beam-line structures and that organ doses from secondary neutrons may be able to be reduced.

Sensitivity Analysis of Cf-252 (sf) Neutron and Gamma Observables in CGMF

DOE Office of Scientific and Technical Information (OSTI.GOV)

Carter, Austin Lewis; Talou, Patrick; Stetcu, Ionel

CGMF is a Monte Carlo code that simulates the decay of primary fission fragments by emission of neutrons and gamma rays, according to the Hauser-Feshbach equations. As the CGMF code was recently integrated into the MCNP6.2 transport code, great emphasis has been placed on providing optimal parameters to CGMF such that many different observables are accurately represented. Of these observables, the prompt neutron spectrum, prompt neutron multiplicity, prompt gamma spectrum, and prompt gamma multiplicity are crucial for accurate transport simulations of criticality and nonproliferation applications. This contribution to the ongoing efforts to improve CGMF presents a study of the sensitivitymore » of various neutron and gamma observables to several input parameters for Californium-252 spontaneous fission. Among the most influential parameters are those that affect the input yield distributions in fragment mass and total kinetic energy (TKE). A new scheme for representing Y(A,TKE) was implemented in CGMF using three fission modes, S1, S2 and SL. The sensitivity profiles were calculated for 17 total parameters, which show that the neutron multiplicity distribution is strongly affected by the TKE distribution of the fragments. The total excitation energy (TXE) of the fragments is shared according to a parameter RT, which is defined as the ratio of the light to heavy initial temperatures. The sensitivity profile of the neutron multiplicity shows a second order effect of RT on the mean neutron multiplicity. A final sensitivity profile was produced for the parameter alpha, which affects the spin of the fragments. Higher values of alpha lead to higher fragment spins, which inhibit the emission of neutrons. Understanding the sensitivity of the prompt neutron and gamma observables to the many CGMF input parameters provides a platform for the optimization of these parameters.« less

Neutron Capture Gamma-Ray Libraries for Nuclear Applications

NASA Astrophysics Data System (ADS)

Sleaford, B. W.; Firestone, R. B.; Summers, N.; Escher, J.; Hurst, A.; Krticka, M.; Basunia, S.; Molnar, G.; Belgya, T.; Revay, Z.; Choi, H. D.

2011-06-01

The neutron capture reaction is useful in identifying and analyzing the gamma-ray spectrum from an unknown assembly as it gives unambiguous information on its composition. This can be done passively or actively where an external neutron source is used to probe an unknown assembly. There are known capture gamma-ray data gaps in the ENDF libraries used by transport codes for various nuclear applications. The Evaluated Gamma-ray Activation file (EGAF) is a new thermal neutron capture database of discrete line spectra and cross sections for over 260 isotopes that was developed as part of an IAEA Coordinated Research Project. EGAF is being used to improve the capture gamma production in ENDF libraries. For medium to heavy nuclei the quasi continuum contribution to the gamma cascades is not experimentally resolved. The continuum contains up to 90% of all the decay energy and is modeled here with the statistical nuclear structure code DICEBOX. This code also provides a consistency check of the level scheme nuclear structure evaluation. The calculated continuum is of sufficient accuracy to include in the ENDF libraries. This analysis also determines new total thermal capture cross sections and provides an improved RIPL database. For higher energy neutron capture there is less experimental data available making benchmarking of the modeling codes more difficult. We are investigating the capture spectra from higher energy neutrons experimentally using surrogate reactions and modeling this with Hauser-Feshbach codes. This can then be used to benchmark CASINO, a version of DICEBOX modified for neutron capture at higher energy. This can be used to simulate spectra from neutron capture at incident neutron energies up to 20 MeV to improve the gamma-ray spectrum in neutron data libraries used for transport modeling of unknown assemblies.

Neutron counter based on beryllium activation

NASA Astrophysics Data System (ADS)

Bienkowska, B.; Prokopowicz, R.; Scholz, M.; Kaczmarczyk, J.; Igielski, A.; Karpinski, L.; Paducha, M.; Pytel, K.

2014-08-01

The fusion reaction occurring in DD plasma is followed by emission of 2.45 MeV neutrons, which carry out information about fusion reaction rate and plasma parameters and properties as well. Neutron activation of beryllium has been chosen for detection of DD fusion neutrons. The cross-section for reaction 9Be(n, α)6He has a useful threshold near 1 MeV, which means that undesirable multiple-scattered neutrons do not undergo that reaction and therefore are not recorded. The product of the reaction, 6He, decays with half-life T1/2 = 0.807 s emitting β- particles which are easy to detect. Large area gas sealed proportional detector has been chosen as a counter of β-particles leaving activated beryllium plate. The plate with optimized dimensions adjoins the proportional counter entrance window. Such set-up is also equipped with appropriate electronic components and forms beryllium neutron activation counter. The neutron flux density on beryllium plate can be determined from the number of counts. The proper calibration procedure needs to be performed, therefore, to establish such relation. The measurements with the use of known β-source have been done. In order to determine the detector response function such experiment have been modeled by means of MCNP5-the Monte Carlo transport code. It allowed proper application of the results of transport calculations of β- particles emitted from radioactive 6He and reaching proportional detector active volume. In order to test the counter system and measuring procedure a number of experiments have been performed on PF devices. The experimental conditions have been simulated by means of MCNP5. The correctness of simulation outcome have been proved by measurements with known radioactive neutron source. The results of the DD fusion neutron measurements have been compared with other neutron diagnostics.

Transport Properties in Nuclear Pasta

NASA Astrophysics Data System (ADS)

Caplan, Matthew; Horowitz, Charles; Berry, Donald; da Silva Schneider, Andre

2016-09-01

At the base of the inner crust of neutron stars, where matter is near the nuclear saturation density, nuclear matter arranges itself into exotic shapes such as cylinders and slabs, called `nuclear pasta.' Lepton scattering from these structures may govern the transport properties of the inner crust; electron scattering from protons in the pasta determines the thermal and electrical conductivity, as well as the shear viscosity of the inner crust. These properties may vary in pasta structures which form at various densities, temperatures, and proton fractions. In this talk, we report on our calculations of lepton transport in nuclear pasta and the implication for neutron star observables.

Developments in capture-γ libraries for nonproliferation applications

NASA Astrophysics Data System (ADS)

Hurst, A. M.; Firestone, R. B.; Sleaford, B. W.; Bleuel, D. L.; Basunia, M. S.; Bečvář, F.; Belgya, T.; Bernstein, L. A.; Carroll, J. J.; Detwiler, B.; Escher, J. E.; Genreith, C.; Goldblum, B. L.; Krtička, M.; Lerch, A. G.; Matters, D. A.; McClory, J. W.; McHale, S. R.; Révay, Zs.; Szentmiklosi, L.; Turkoglu, D.; Ureche, A.; Vujic, J.

2017-09-01

The neutron-capture reaction is fundamental for identifying and analyzing the γ-ray spectrum from an unknown assembly because it provides unambiguous information on the neutron-absorbing isotopes. Nondestructive-assay applications may exploit this phenomenon passively, for example, in the presence of spontaneous-fission neutrons, or actively where an external neutron source is used as a probe. There are known gaps in the Evaluated Nuclear Data File libraries corresponding to neutron-capture γ-ray data that otherwise limit transport-modeling applications. In this work, we describe how new thermal neutron-capture data are being used to improve information in the neutron-data libraries for isotopes relevant to nonproliferation applications. We address this problem by providing new experimentally-deduced partial and total neutron-capture reaction cross sections and then evaluate these data by comparison with statistical-model calculations.

Implementation of a tree algorithm in MCNP code for nuclear well logging applications.

PubMed

Li, Fusheng; Han, Xiaogang

2012-07-01

The goal of this paper is to develop some modeling capabilities that are missing in the current MCNP code. Those missing capabilities can greatly help for some certain nuclear tools designs, such as a nuclear lithology/mineralogy spectroscopy tool. The new capabilities to be developed in this paper include the following: zone tally, neutron interaction tally, gamma rays index tally and enhanced pulse-height tally. The patched MCNP code also can be used to compute neutron slowing-down length and thermal neutron diffusion length. Copyright © 2011 Elsevier Ltd. All rights reserved.

Effect of random inhomogeneities in the spatial distribution of radiation-induced defect clusters on carrier transport through the thin base of a heterojunction bipolar transistor upon neutron irradiation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Puzanov, A. S.; Obolenskiy, S. V., E-mail: obolensk@rf.unn.ru; Kozlov, V. A.

We analyze the electron transport through the thin base of a GaAs heterojunction bipolar transistor with regard to fluctuations in the spatial distribution of defect clusters induced by irradiation with a fissionspectrum fast neutron flux. We theoretically demonstrate that the homogeneous filling of the working region with radiation-induced defect clusters causes minimum degradation of the dc gain of the heterojunction bipolar transistor.

Estimation of neutron dose equivalent at the mezzanine of the Advanced Light Source and the laboratory boundary using the ORNL program MORSE.

PubMed

Sun, R K

1990-12-01

To investigate the radiation effect of neutrons near the Advanced Light Source (ALS) at Lawrence Berkeley Laboratory (LBL) with respect to the neutron dose equivalents in nearby occupied areas and at the site boundary, the neutron transport code MORSE, from Oak Ridge National Laboratory (ORNL), was used. These dose equivalents result from both skyshine neutrons transported by air scattering and direct neutrons penetrating the shielding. The ALS neutron sources are a 50-MeV linear accelerator and its transfer line, a 1.5-GeV booster, a beam extraction line, and a 1.9-GeV storage ring. The most conservative total occupational-dose-equivalent rate in the center of the ALS mezzanine, 39 m from the ALS center, was found to be 1.14 X 10(-3) Sv y-1 per 2000-h "occupational" year, and the total environmental-dose-equivalent rate at the ALS boundary, 125 m from the ALS center, was found to be 3.02 X 10(-4) Sv y-1 per 8760-h calendar year. More realistic dose-equivalent rates, using the nominal (expected) storage-ring current, were calculated to be 1.0 X 10(-4) Sv y-1 and 2.65 X 10(-5) Sv y-1 occupational year and calendar year, respectively, which are much lower than the DOE reporting levels.

A setup for active neutron analysis of the fissile material content in fuel assemblies of nuclear reactors

NASA Astrophysics Data System (ADS)

Bushuev, A. V.; Kozhin, A. F.; Aleeva, T. B.; Zubarev, V. N.; Petrova, E. V.; Smirnov, V. E.

2016-12-01

An active neutron method for measuring the residual mass of 235U in spent fuel assemblies (FAs) of the IRT MEPhI research reactor is presented. The special measuring stand design and uniform irradiation of the fuel with neutrons along the entire length of the active part of the FA provide high accuracy of determination of the residual 235U content. AmLi neutron sources yield a higher effect/background ratio than other types of sources and do not induce the fission of 238U. The proposed method of transfer of the isotope source in accordance with a given algorithm may be used in experiments where the studied object needs to be irradiated with a uniform fluence.

Computational Transport Modeling of High-Energy Neutrons Found in the Space Environment

NASA Technical Reports Server (NTRS)

Cox, Brad; Theriot, Corey A.; Rohde, Larry H.; Wu, Honglu

2012-01-01

The high charge and high energy (HZE) particle radiation environment in space interacts with spacecraft materials and the human body to create a population of neutrons encompassing a broad kinetic energy spectrum. As an HZE ion penetrates matter, there is an increasing chance of fragmentation as penetration depth increases. When an ion fragments, secondary neutrons are released with velocities up to that of the primary ion, giving some neutrons very long penetration ranges. These secondary neutrons have a high relative biological effectiveness, are difficult to effectively shield, and can cause more biological damage than the primary ions in some scenarios. Ground-based irradiation experiments that simulate the space radiation environment must account for this spectrum of neutrons. Using the Particle and Heavy Ion Transport Code System (PHITS), it is possible to simulate a neutron environment that is characteristic of that found in spaceflight. Considering neutron dosimetry, the focus lies on the broad spectrum of recoil protons that are produced in biological targets. In a biological target, dose at a certain penetration depth is primarily dependent upon recoil proton tracks. The PHITS code can be used to simulate a broad-energy neutron spectrum traversing biological targets, and it account for the recoil particle population. This project focuses on modeling a neutron beamline irradiation scenario for determining dose at increasing depth in water targets. Energy-deposition events and particle fluence can be simulated by establishing cross-sectional scoring routines at different depths in a target. This type of model is useful for correlating theoretical data with actual beamline radiobiology experiments. Other work exposed human fibroblast cells to a high-energy neutron source to study micronuclei induction in cells at increasing depth behind water shielding. Those findings provide supporting data describing dose vs. depth across a water-equivalent medium. This poster presents PHITS data suggesting an increase in dose, up to roughly 10 cm depth, followed by a continual decrease as neutrons come to a stop in the target.

Gravitational effects on planetary neutron flux spectra

NASA Astrophysics Data System (ADS)

Feldman, W. C.; Drake, D. M.; O'dell, R. D.; Brinkley, F. W.; Anderson, R. C.

1989-01-01

The effects of gravity on the planetary neutron flux spectra for planet Mars, and the lifetime of the neutron, were investigated using a modified one-dimensional diffusion accelerated neutral-particle transport code, coupled with a multigroup cross-section library tailored specifically for Mars. The results showed the presence of a qualitatively new feature in planetary neutron leakage spectra in the form of a component of returning neutrons with kinetic energies less than the gravitational binding energy (0.132 eV for Mars). The net effect is an enhancement in flux at the lowest energies that is largest at and above the outermost layer of planetary matter.

Reevaluation of secondary neutron spectra from thick targets upon heavy-ion bombardment

NASA Astrophysics Data System (ADS)

Satoh, D.; Kurosawa, T.; Sato, T.; Endo, A.; Takada, M.; Iwase, H.; Nakamura, T.; Niita, K.

2007-12-01

Previously published data of secondary neutron spectra from thick targets of C, Al, Cu and Pb bombarded with heavy ions from He to Xe are revised by using a new set of neutron-detection efficiency values for a liquid organic scintillator calculated with SCINFUL-QMD. Additional data have been measured for bombardment of C target by 400-MeV/nucleon C ions and 800-MeV/nucleon Si ions. The set of spectra are compared with the calculation results using a Monte-Carlo heavy-ion transport code, PHITS. It was found that PHITS is able to reproduce the secondary neutron spectra in a wide neutron-energy regime.

Determination of the fast-neutron-induced fission cross-section of 242Pu at nELBE

NASA Astrophysics Data System (ADS)

Kögler, Toni; Beyer, Roland; Junghans, Arnd R.; Schwengner, Ronald; Wagner, Andreas

2018-03-01

The fast-neutron-induced fission cross section of 242Pu was determined in the energy range of 0.5 MeV to 10MeV at the neutron time-of-flight facility nELBE. Using a parallel-plate fission ionization chamber this quantity was measured relative to 235U(n,f). The number of target nuclei was thereby calculated by means of measuring the spontaneous fission rate of 242Pu. An MCNP 6 neutron transport simulation was used to correct the relative cross section for neutron scattering. The determined results are in good agreement with current experimental and evaluated data sets.

Neutron-based nonintrusive inspection techniques

NASA Astrophysics Data System (ADS)

Gozani, Tsahi

1997-02-01

Non-intrusive inspection of large objects such as trucks, sea-going shipping containers, air cargo containers and pallets is gaining attention as a vital tool in combating terrorism, drug smuggling and other violation of international and national transportation and Customs laws. Neutrons are the preferred probing radiation when material specificity is required, which is most often the case. Great strides have been made in neutron based inspection techniques. Fast and thermal neutrons, whether in steady state or in microsecond, or even nanosecond pulses are being employed to interrogate, at high speeds, for explosives, drugs, chemical agents, and nuclear and many other smuggled materials. Existing neutron techniques will be compared and their current status reported.

Intensity modulated neutron radiotherapy optimization by photon proxy

DOE Office of Scientific and Technical Information (OSTI.GOV)

Snyder, Michael; Hammoud, Ahmad; Bossenberger, Todd

2012-08-15

Purpose: Introducing intensity modulation into neutron radiotherapy (IMNRT) planning has the potential to mitigate some normal tissue complications seen in past neutron trials. While the hardware to deliver IMNRT plans has been in use for several years, until recently the IMNRT planning process has been cumbersome and of lower fidelity than conventional photon plans. Our in-house planning system used to calculate neutron therapy plans allows beam weight optimization of forward planned segments, but does not provide inverse optimization capabilities. Commercial treatment planning systems provide inverse optimization capabilities, but currently cannot model our neutron beam. Methods: We have developed a methodologymore » and software suite to make use of the robust optimization in our commercial planning system while still using our in-house planning system to calculate final neutron dose distributions. Optimized multileaf collimator (MLC) leaf positions for segments designed in the commercial system using a 4 MV photon proxy beam are translated into static neutron ports that can be represented within our in-house treatment planning system. The true neutron dose distribution is calculated in the in-house system and then exported back through the MATLAB software into the commercial treatment planning system for evaluation. Results: The planning process produces optimized IMNRT plans that reduce dose to normal tissue structures as compared to 3D conformal plans using static MLC apertures. The process involves standard planning techniques using a commercially available treatment planning system, and is not significantly more complex than conventional IMRT planning. Using a photon proxy in a commercial optimization algorithm produces IMNRT plans that are more conformal than those previously designed at our center and take much less time to create. Conclusions: The planning process presented here allows for the optimization of IMNRT plans by a commercial treatment planning optimization algorithm, potentially allowing IMNRT to achieve similar conformality in treatment as photon IMRT. The only remaining requirements for the delivery of very highly modulated neutron treatments are incremental improvements upon already implemented hardware systems that should be readily achievable.« less

A Pulsed Sphere Tutorial

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cullen, Dermott E.

2017-01-30

Here I attempt to explain what physically happens when we pulse an object with neutrons, specifically what we expect the time dependent behavior of the neutron population to look like. Emphasis is on the time dependent emission of both prompt and delayed neutrons. I also describe how the TART Monte Carlo transport code models this situation; see the appendix for a complete description of the model used by TART. I will also show that, as we expect, MCNP and MERCURY, produce similar results using the same delayed neutron model (again, see the appendix).

Benchmarking PARTISN with Analog Monte Carlo: Moments of the Neutron Number and the Cumulative Fission Number Probability Distributions

DOE Office of Scientific and Technical Information (OSTI.GOV)

O'Rourke, Patrick Francis

The purpose of this report is to provide the reader with an understanding of how a Monte Carlo neutron transport code was written, developed, and evolved to calculate the probability distribution functions (PDFs) and their moments for the neutron number at a final time as well as the cumulative fission number, along with introducing several basic Monte Carlo concepts.

Preliminary estimates of nucleon fluxes in a water target exposed to solar-flare protons: BRYNTRN versus Monte Carlo code

NASA Technical Reports Server (NTRS)

Shinn, Judy L.; Wilson, John W.; Lone, M. A.; Wong, P. Y.; Costen, Robert C.

1994-01-01

A baryon transport code (BRYNTRN) has previously been verified using available Monte Carlo results for a solar-flare spectrum as the reference. Excellent results were obtained, but the comparisons were limited to the available data on dose and dose equivalent for moderate penetration studies that involve minor contributions from secondary neutrons. To further verify the code, the secondary energy spectra of protons and neutrons are calculated using BRYNTRN and LAHET (Los Alamos High-Energy Transport code, which is a Monte Carlo code). These calculations are compared for three locations within a water slab exposed to the February 1956 solar-proton spectrum. Reasonable agreement was obtained when various considerations related to the calculational techniques and their limitations were taken into account. Although the Monte Carlo results are preliminary, it appears that the neutron albedo, which is not currently treated in BRYNTRN, might be a cause for the large discrepancy seen at small penetration depths. It also appears that the nonelastic neutron production cross sections in BRYNTRN may underestimate the number of neutrons produced in proton collisions with energies below 200 MeV. The notion that the poor energy resolution in BRYNTRN may cause a large truncation error in neutron elastic scattering requires further study.

MCNP capabilities for nuclear well logging calculations

DOE Office of Scientific and Technical Information (OSTI.GOV)

Forster, R.A.; Little, R.C.; Briesmeister, J.F.

The Los Alamos Radiation Transport Code System (LARTCS) consists of state-of-the-art Monte Carlo and discrete ordinates transport codes and data libraries. This paper discusses how the general-purpose continuous-energy Monte Carlo code MCNP ({und M}onte {und C}arlo {und n}eutron {und p}hoton), part of the LARTCS, provides a computational predictive capability for many applications of interest to the nuclear well logging community. The generalized three-dimensional geometry of MCNP is well suited for borehole-tool models. SABRINA, another component of the LARTCS, is a graphics code that can be used to interactively create a complex MCNP geometry. Users can define many source and tallymore » characteristics with standard MCNP features. The time-dependent capability of the code is essential when modeling pulsed sources. Problems with neutrons, photons, and electrons as either single particle or coupled particles can be calculated with MCNP. The physics of neutron and photon transport and interactions is modeled in detail using the latest available cross-section data.« less

How to polarise all neutrons in one beam: a high performance polariser and neutron transport system

NASA Astrophysics Data System (ADS)

Rodriguez, D. Martin; Bentley, P. M.; Pappas, C.

2016-09-01

Polarised neutron beams are used in disciplines as diverse as magnetism,soft matter or biology. However, most of these applications often suffer from low flux also because the existing neutron polarising methods imply the filtering of one of the spin states, with a transmission of 50% at maximum. With the purpose of using all neutrons that are usually discarded, we propose a system that splits them according to their polarisation, flips them to match the spin direction, and then focuses them at the sample. Monte Carlo (MC) simulations show that this is achievable over a wide wavelength range and with an outstanding performance at the price of a more divergent neutron beam at the sample position.

Quasiparticle Interactions in Neutron Matter for Applications in Neutron Stars

NASA Technical Reports Server (NTRS)

Wambach, J.; Anisworth, T. L.; Pines, D.

1993-01-01

A microscopic model for the quaisiparticle interaction in neutron matter is presented. Both particle-particle (pp) and particle-hole (ph) correlation are are included. The pp correlations are treated in semi-empirical way, while ph correlations are incorporated by solving coupled two-body equations for the particle hole interaction and the scattering amplitude on the Fermi sphere. The resulting integral equations self-consistently sum the ph reducible diagrams. Antisymmetry is kept at all stages and hence the forward-scattering sum rules are obeyed. Results for Landau parameters and transport coefficients in a density regime representing the crust of a neutron star are presented. We also estimate the S-1 gap parameter for neutron superfluidity and comment briefly on neutron-star implications.

Quasiparticle Interactions in Neutron Matter for Applications in Neutron Stars

NASA Technical Reports Server (NTRS)

Wambach, J; Ainsworth, T. L.; Pines, D.

1993-01-01

A microscopic model for the quasiparticle interaction in neutron matter is presented. Both-particle (pp) and particle-hole (ph) correlations are included. The pp correlations are treated in semi-empirical way, while ph correlations are incorporated by solving coupled two-body equations for particle-hole interaction and the scattering amplitude of the Fermi sphere. The resulting integral equations self-consistently sum the ph reducible diagrams. Antisymmetry is kept at all stages and hence the forward-scattering sum rules for the scattering amplitude are obeyed. Results for Landau parameters and transport coefficients in a density regime representing the crust of a neutron star are presented. We also estimate the (1)S(sub 0) gap parameter for neutron superfluidity and comment briefly on neutron-star implications.

Feasibility studies on explosive detection and homeland security applications using a neutron and x-ray combined computed tomography system

NASA Astrophysics Data System (ADS)

Sinha, V.; Srivastava, A.; Lee, H. K.; Liu, X.

2013-05-01

The successful creation and operation of a neutron and X-ray combined computed tomography (NXCT) system has been demonstrated by researchers at the Missouri University of Science and Technology. The NXCT system has numerous applications in the field of material characterization and object identification in materials with a mixture of atomic numbers represented. Presently, the feasibility studies have been performed for explosive detection and homeland security applications, particularly in concealed material detection and determination of the light atomic number materials. These materials cannot be detected using traditional X-ray imaging. The new system has the capability to provide complete structural and compositional information due to the complementary nature of X-ray and neutron interactions with materials. The design of the NXCT system facilitates simultaneous and instantaneous imaging operation, promising enhanced detection capabilities of explosive materials, low atomic number materials and illicit materials for homeland security applications. In addition, a sample positioning system allowing the user to remotely and automatically manipulate the sample makes the system viable for commercial applications. Several explosives and weapon simulants have been imaged and the results are provided. The fusion algorithms which combine the data from the neutron and X-ray imaging produce superior images. This paper is a compete overview of the NXCT system for feasibility studies of explosive detection and homeland security applications. The design of the system, operation, algorithm development, and detection schemes are provided. This is the first combined neutron and X-ray computed tomography system in operation. Furthermore, the method of fusing neutron and X-ray images together is a new approach which provides high contrast images of the desired object. The system could serve as a standardized tool in nondestructive testing of many applications, especially in explosives detection and homeland security research.

Calculation of radiation therapy dose using all particle Monte Carlo transport

DOEpatents

Chandler, William P.; Hartmann-Siantar, Christine L.; Rathkopf, James A.

1999-01-01

The actual radiation dose absorbed in the body is calculated using three-dimensional Monte Carlo transport. Neutrons, protons, deuterons, tritons, helium-3, alpha particles, photons, electrons, and positrons are transported in a completely coupled manner, using this Monte Carlo All-Particle Method (MCAPM). The major elements of the invention include: computer hardware, user description of the patient, description of the radiation source, physical databases, Monte Carlo transport, and output of dose distributions. This facilitated the estimation of dose distributions on a Cartesian grid for neutrons, photons, electrons, positrons, and heavy charged-particles incident on any biological target, with resolutions ranging from microns to centimeters. Calculations can be extended to estimate dose distributions on general-geometry (non-Cartesian) grids for biological and/or non-biological media.

Calculation of radiation therapy dose using all particle Monte Carlo transport

DOEpatents

Chandler, W.P.; Hartmann-Siantar, C.L.; Rathkopf, J.A.

1999-02-09

The actual radiation dose absorbed in the body is calculated using three-dimensional Monte Carlo transport. Neutrons, protons, deuterons, tritons, helium-3, alpha particles, photons, electrons, and positrons are transported in a completely coupled manner, using this Monte Carlo All-Particle Method (MCAPM). The major elements of the invention include: computer hardware, user description of the patient, description of the radiation source, physical databases, Monte Carlo transport, and output of dose distributions. This facilitated the estimation of dose distributions on a Cartesian grid for neutrons, photons, electrons, positrons, and heavy charged-particles incident on any biological target, with resolutions ranging from microns to centimeters. Calculations can be extended to estimate dose distributions on general-geometry (non-Cartesian) grids for biological and/or non-biological media. 57 figs.

Simultaneous usage of pinhole and penumbral apertures for imaging small scale neutron sources from inertial confinement fusion experiments.

PubMed

Guler, N; Volegov, P; Danly, C R; Grim, G P; Merrill, F E; Wilde, C H

2012-10-01

Inertial confinement fusion experiments at the National Ignition Facility are designed to understand the basic principles of creating self-sustaining fusion reactions by laser driven compression of deuterium-tritium (DT) filled cryogenic plastic capsules. The neutron imaging diagnostic provides information on the distribution of the central fusion reaction region and the surrounding DT fuel by observing neutron images in two different energy bands for primary (13-17 MeV) and down-scattered (6-12 MeV) neutrons. From this, the final shape and size of the compressed capsule can be estimated and the symmetry of the compression can be inferred. These experiments provide small sources with high yield neutron flux. An aperture design that includes an array of pinholes and penumbral apertures has provided the opportunity to image the same source with two different techniques. This allows for an evaluation of these different aperture designs and reconstruction algorithms.

Three-dimensional reconstruction of neutron, gamma-ray, and x-ray sources using spherical harmonic decomposition

NASA Astrophysics Data System (ADS)

Volegov, P. L.; Danly, C. R.; Fittinghoff, D.; Geppert-Kleinrath, V.; Grim, G.; Merrill, F. E.; Wilde, C. H.

2017-11-01

Neutron, gamma-ray, and x-ray imaging are important diagnostic tools at the National Ignition Facility (NIF) for measuring the two-dimensional (2D) size and shape of the neutron producing region, for probing the remaining ablator and measuring the extent of the DT plasmas during the stagnation phase of Inertial Confinement Fusion implosions. Due to the difficulty and expense of building these imagers, at most only a few two-dimensional projections images will be available to reconstruct the three-dimensional (3D) sources. In this paper, we present a technique that has been developed for the 3D reconstruction of neutron, gamma-ray, and x-ray sources from a minimal number of 2D projections using spherical harmonics decomposition. We present the detailed algorithms used for this characterization and the results of reconstructed sources from experimental neutron and x-ray data collected at OMEGA and NIF.

A Monte Carlo simulation study for the gamma-ray/neutron dual-particle imager using rotational modulation collimator (RMC).

PubMed

Kim, Hyun Suk; Choi, Hong Yeop; Lee, Gyemin; Ye, Sung-Joon; Smith, Martin B; Kim, Geehyun

2018-03-01

The aim of this work is to develop a gamma-ray/neutron dual-particle imager, based on rotational modulation collimators (RMCs) and pulse shape discrimination (PSD)-capable scintillators, for possible applications for radioactivity monitoring as well as nuclear security and safeguards. A Monte Carlo simulation study was performed to design an RMC system for the dual-particle imaging, and modulation patterns were obtained for gamma-ray and neutron sources in various configurations. We applied an image reconstruction algorithm utilizing the maximum-likelihood expectation-maximization method based on the analytical modeling of source-detector configurations, to the Monte Carlo simulation results. Both gamma-ray and neutron source distributions were reconstructed and evaluated in terms of signal-to-noise ratio, showing the viability of developing an RMC-based gamma-ray/neutron dual-particle imager using PSD-capable scintillators.

Extremum seeking-based optimization of high voltage converter modulator rise-time

DOE Office of Scientific and Technical Information (OSTI.GOV)

Scheinker, Alexander; Bland, Michael; Krstic, Miroslav

2013-02-01

We digitally implement an extremum seeking (ES) algorithm, which optimizes the rise time of the output voltage of a high voltage converter modulator (HVCM) at the Los Alamos Neutron Science Center (LANSCE) HVCM test stand by iteratively, simultaneously tuning the first 8 switching edges of each of the three phase drive waveforms (24 variables total). We achieve a 50 μs rise time, which is reduction in half compared to the 100 μs achieved at the Spallation Neutron Source (SNS) at Oak Ridge National Laboratory. Considering that HVCMs typically operate with an output voltage of 100 kV, with a 60Hz repetitionmore » rate, the 50 μs rise time reduction will result in very significant energy savings. The ES algorithm will prove successful, despite the noisy measurements and cost calculations, confirming the theoretical results that the algorithm is not affected by noise whose frequency components are independent of the perturbing frequencies.« less

Transport of Neutrons and Photons Through Iron and Water Layers

NASA Astrophysics Data System (ADS)

Košťál, Michal; Cvachovec, František; Ošmera, Bohumil; Hansen, Wolfgang; Noack, Klaus

2009-08-01

The neutron and photon spectra were measured after iron and water plates placed at the horizontal channel of the Dresden University reactor AK-2. The measurements have been performed with the multiparameter spectrometer [1] with a stilbene cylindrical crystal, 10 × 10 mm or 45 × 45 mm; the neutron and photon spectra have been measured simultaneously. The calculations were performed with the MCNP code and nuclear data libraries ENDF/B VI.2, ENDF/BVII.0, JENDL 3.3 and JEFF 3.1. The measured channel leakage spectrum was used as the input spectrum for the transport calculation. Photons, the primary photons from the reactor - as well as the ones induced by neutron interaction - were calculated. The comparison of the measurements and calculations through 10 cm of iron and 20 cm thickness of water are presented. Besides that, the attenuation of the radiation mixed field by iron layers from 5 to 30 cm is presented; the measured and calculated data are compared.

Analysis of tritium production in concentric spheres of oralloy and /sup 6/LiD irradiated by 14-MeV neutrons

DOE Office of Scientific and Technical Information (OSTI.GOV)

Fawcett, L.R. Jr.; Roberts, R.R. II; Hunter, R.E.

1988-03-01

Tritium production and activation of radiochemical detector foils in a sphere of /sup 6/LiD with an oralloy core irradiated by a central source of 14-MeV neutrons have been calculated and compared with experimental measurements. The experimental assembly consisted of an oralloy sphere surrounded by three solid /sup 6/LiD concentric shells with ampules of /sup 6/LiH and /sup 7/LiH and activation foils located in several positions throughout the assembly. The Los Alamos Monte Carlo Neutron Photon Transport Code (MCNP) was used to calculate neutron transport throughout the system, tritium production in the ampules, and foil activation. The overall experimentally observed-to-calculated ratiosmore » of tritium production were 0.996 +- 2.5% in /sup 6/Li ampules and 0.903 +- 5.2% in /sup 7/Li ampules. Observed-to-calculated ratios for foil activation are also presented. 11 refs., 4 figs., 7 tabs.« less

MCNP (Monte Carlo Neutron Photon) capabilities for nuclear well logging calculations

DOE Office of Scientific and Technical Information (OSTI.GOV)

Forster, R.A.; Little, R.C.; Briesmeister, J.F.

The Los Alamos Radiation Transport Code System (LARTCS) consists of state-of-the-art Monte Carlo and discrete ordinates transport codes and data libraries. The general-purpose continuous-energy Monte Carlo code MCNP (Monte Carlo Neutron Photon), part of the LARTCS, provides a computational predictive capability for many applications of interest to the nuclear well logging community. The generalized three-dimensional geometry of MCNP is well suited for borehole-tool models. SABRINA, another component of the LARTCS, is a graphics code that can be used to interactively create a complex MCNP geometry. Users can define many source and tally characteristics with standard MCNP features. The time-dependent capabilitymore » of the code is essential when modeling pulsed sources. Problems with neutrons, photons, and electrons as either single particle or coupled particles can be calculated with MCNP. The physics of neutron and photon transport and interactions is modeled in detail using the latest available cross-section data. A rich collections of variance reduction features can greatly increase the efficiency of a calculation. MCNP is written in FORTRAN 77 and has been run on variety of computer systems from scientific workstations to supercomputers. The next production version of MCNP will include features such as continuous-energy electron transport and a multitasking option. Areas of ongoing research of interest to the well logging community include angle biasing, adaptive Monte Carlo, improved discrete ordinates capabilities, and discrete ordinates/Monte Carlo hybrid development. Los Alamos has requested approval by the Department of Energy to create a Radiation Transport Computational Facility under their User Facility Program to increase external interactions with industry, universities, and other government organizations. 21 refs.« less

In vivo prompt gamma neutron activation analysis for the screening of boron-10 distribution in a rabbit knee: a simulation study

NASA Astrophysics Data System (ADS)

Zhu, X.; Clackdoyle, R.; Shortkroff, S.; Yanch, J.

2008-05-01

Boron neutron capture synovectomy (BNCS) is under development as a potential treatment modality for rheumatoid arthritis (RA). RA is characterized by the inflammation of the synovium (the membrane lining articular joints), which leads to pain and a restricted range of motion. BNCS is a two-part procedure involving the injection of a boronated compound directly into the diseased joint followed by irradiation with a low-energy neutron beam. The neutron capture reactions taking place in the synovium deliver a local, high-linear energy transfer (LET) dose aimed at destroying the inflamed synovial membrane. For successful treatment via BNCS, a boron-labeled compound exhibiting both high synovial uptake and long retention time is necessary. Currently, the in vivo uptake behavior of potentially useful boronated compounds is evaluated in the knee joints of rabbits in which arthritis has been induced. This strategy involves the sacrifice and dissection of a large number of animals. An in vivo 10B screening approach is therefore under investigation with the goal of significantly reducing the number of animals needed for compound evaluation via dissection studies. The 'in vivo prompt gamma neutron activation analysis' (IVPGNAA) approach uses a narrow neutron beam to irradiate the knee from several angular positions following the intra-articular injection of a boronated compound whose uptake characteristics are unknown. A high-purity germanium detector collects the 478 keV gamma photons produced by the 10B capture reactions. The 10B distribution in the knee is then reconstructed by solving a system of simultaneous equations using a weighted least squares algorithm. To study the practical feasibility of IVPGNAA, simulation data were generated with the Monte Carlo N-particle transport code. The boron-containing region of a rabbit knee was partitioned into 8 compartments, and the 10B prompt gamma signals were tallied from 16 angular positions. Results demonstrate that for this level of spatial resolution, an estimate of 10B distribution inside the joint can be obtained to within 10% uncertainty, under ideal conditions. Variations of the anatomic dimensions among individual rabbit knees and potential knee positioning errors will result in an uncertainty of over 20%. IVPGNAA thus provides sufficient resolution and quantification regarding the in vivo uptake characteristics of boronated pharmaceuticals to serve as a useful means of screening new compounds of potential use in BNCS.

Low-energy beam transport studies supporting the spallation neutron source 1-MW beam operation.

PubMed

Han, B X; Kalvas, T; Tarvainen, O; Welton, R F; Murray, S N; Pennisi, T R; Santana, M; Stockli, M P

2012-02-01

The H(-) injector consisting of a cesium enhanced RF-driven ion source and a 2-lens electrostatic low-energy beam transport (LEBT) system supports the spallation neutron source 1 MW beam operation with ∼38 mA beam current in the linac at 60 Hz with a pulse length of up to ∼1.0 ms. In this work, two important issues associated with the low-energy beam transport are discussed: (1) inconsistent dependence of the post-radio frequency quadrupole accelerator beam current on the ion source tilt angle and (2) high power beam losses on the LEBT electrodes under some off-nominal conditions compromising their reliability.

Benchmarking of Neutron Production of Heavy-Ion Transport Codes

DOE Office of Scientific and Technical Information (OSTI.GOV)

Remec, Igor; Ronningen, Reginald M.; Heilbronn, Lawrence

Accurate prediction of radiation fields generated by heavy ion interactions is important in medical applications, space missions, and in design and operation of rare isotope research facilities. In recent years, several well-established computer codes in widespread use for particle and radiation transport calculations have been equipped with the capability to simulate heavy ion transport and interactions. To assess and validate these capabilities, we performed simulations of a series of benchmark-quality heavy ion experiments with the computer codes FLUKA, MARS15, MCNPX, and PHITS. We focus on the comparisons of secondary neutron production. Results are encouraging; however, further improvements in models andmore » codes and additional benchmarking are required.« less

Nonnegative Tensor Factorization Approach Applied to Fission Chamber’s Output Signals Blind Source Separation

NASA Astrophysics Data System (ADS)

Laassiri, M.; Hamzaoui, E.-M.; Cherkaoui El Moursli, R.

2018-02-01

Inside nuclear reactors, gamma-rays emitted from nuclei together with the neutrons introduce unwanted backgrounds in neutron spectra. For this reason, powerful extraction methods are needed to extract useful neutron signal from recorded mixture and thus to obtain clearer neutron flux spectrum. Actually, several techniques have been developed to discriminate between neutrons and gamma-rays in a mixed radiation field. Most of these techniques, tackle using analogue discrimination methods. Others propose to use some organic scintillators to achieve the discrimination task. Recently, systems based on digital signal processors are commercially available to replace the analog systems. As alternative to these systems, we aim in this work to verify the feasibility of using a Nonnegative Tensor Factorization (NTF) to blind extract neutron component from mixture signals recorded at the output of fission chamber (WL-7657). This last have been simulated through the Geant4 linked to Garfield++ using a 252Cf neutron source. To achieve our objective of obtaining the best possible neutron-gamma discrimination, we have applied the two different NTF algorithms, which have been found to be the best methods that allow us to analyse this kind of nuclear data.

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

PubMed Central

Eley, John; Newhauser, Wayne; Homann, Kenneth; Howell, Rebecca; Schneider, Christopher; Durante, Marco; Bert, Christoph

2015-01-01

Equivalent dose from neutrons produced during proton radiotherapy increases the predicted risk of radiogenic late effects. However, out-of-field neutron dose is not taken into account by commercial proton radiotherapy treatment planning systems. The purpose of this study was to demonstrate the feasibility of implementing an analytical model to calculate leakage neutron equivalent dose in a treatment planning system. Passive scattering proton treatment plans were created for a water phantom and for a patient. For both the phantom and patient, the neutron equivalent doses were small but non-negligible and extended far beyond the therapeutic field. The time required for neutron equivalent dose calculation was 1.6 times longer than that required for proton dose calculation, with a total calculation time of less than 1 h on one processor for both treatment plans. Our results demonstrate that it is feasible to predict neutron equivalent dose distributions using an analytical dose algorithm for individual patients with irregular surfaces and internal tissue heterogeneities. Eventually, personalized estimates of neutron equivalent dose to organs far from the treatment field may guide clinicians to create treatment plans that reduce the risk of late effects. PMID:25768061

Implementation of an analytical model for leakage neutron equivalent dose in a proton radiotherapy planning system.

PubMed

Eley, John; Newhauser, Wayne; Homann, Kenneth; Howell, Rebecca; Schneider, Christopher; Durante, Marco; Bert, Christoph

2015-03-11

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

Assessing and minimizing contamination in time of flight based validation data

NASA Astrophysics Data System (ADS)

Lennox, Kristin P.; Rosenfield, Paul; Blair, Brenton; Kaplan, Alan; Ruz, Jaime; Glenn, Andrew; Wurtz, Ronald

2017-10-01

Time of flight experiments are the gold standard method for generating labeled training and testing data for the neutron/gamma pulse shape discrimination problem. As the popularity of supervised classification methods increases in this field, there will also be increasing reliance on time of flight data for algorithm development and evaluation. However, time of flight experiments are subject to various sources of contamination that lead to neutron and gamma pulses being mislabeled. Such labeling errors have a detrimental effect on classification algorithm training and testing, and should therefore be minimized. This paper presents a method for identifying minimally contaminated data sets from time of flight experiments and estimating the residual contamination rate. This method leverages statistical models describing neutron and gamma travel time distributions and is easily implemented using existing statistical software. The method produces a set of optimal intervals that balance the trade-off between interval size and nuisance particle contamination, and its use is demonstrated on a time of flight data set for Cf-252. The particular properties of the optimal intervals for the demonstration data are explored in detail.

Parallel deterministic neutronics with AMR in 3D

DOE Office of Scientific and Technical Information (OSTI.GOV)

Clouse, C.; Ferguson, J.; Hendrickson, C.

1997-12-31

AMTRAN, a three dimensional Sn neutronics code with adaptive mesh refinement (AMR) has been parallelized over spatial domains and energy groups and runs on the Meiko CS-2 with MPI message passing. Block refined AMR is used with linear finite element representations for the fluxes, which allows for a straight forward interpretation of fluxes at block interfaces with zoning differences. The load balancing algorithm assumes 8 spatial domains, which minimizes idle time among processors.

A hybrid multigroup neutron-pattern model

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pogosbekyan, L.R.; Lysov, D.A.

In this paper, we use the general approach to construct a multigroup hybrid model for the neutron pattern. The equations are given together with a reasonably economic and simple iterative method of solving them. The algorithm can be used to calculate the pattern and the functionals as well as to correct the constants from the experimental data and to adapt the support over the constants to the engineering programs by reference to precision ones.

Post-merger evolution of a neutron star-black hole binary with neutrino transport

NASA Astrophysics Data System (ADS)

Foucart, Francois; O'Connor, Evan; Roberts, Luke; Duez, Matthew D.; Haas, Roland; Kidder, Lawrence E.; Ott, Christian D.; Pfeiffer, Harald P.; Scheel, Mark A.; Szilagyi, Bela

2015-06-01

We present a first simulation of the post-merger evolution of a black hole-neutron star binary in full general relativity using an energy-integrated general-relativistic truncated moment formalism for neutrino transport. We describe our implementation of the moment formalism and important tests of our code, before studying the formation phase of an accretion disk after a black hole-neutron star merger. We use as initial data an existing general-relativistic simulation of the merger of a neutron star of mass 1.4 M⊙ with a black hole of mass 7 M⊙ and dimensionless spin χBH=0.8 . Comparing with a simpler leakage scheme for the treatment of the neutrinos, we find noticeable differences in the neutron-to-proton ratio in and around the disk, and in the neutrino luminosity. We find that the electron neutrino luminosity is much lower in the transport simulations, and that both the disk and the disk outflows are less neutron rich. The spatial distribution of the neutrinos is significantly affected by relativistic effects, due to large velocities and curvature in the regions of strongest emission. Over the short time scale evolved, we do not observe purely neutrino-driven outflows. However, a small amount of material (3 ×10-4M⊙ ) is ejected in the polar region during the circularization of the disk. Most of that material is ejected early in the formation of the disk, and is fairly neutron rich (electron fraction Ye˜0.15 - 0.25 ). Through r-process nucleosynthesis, that material should produce high-opacity lanthanides in the polar region, and could thus affect the light curve of radioactively powered electromagnetic transients. We also show that by the end of the simulation, while the bulk of the disk remains neutron rich (Ye˜0.15 - 0.2 and decreasing), its outer layers have a higher electron fraction: 10% of the remaining mass has Ye>0.3 . As that material would be the first to be unbound by disk outflows on longer time scales, and as composition evolution is slower at later times, the changes in Ye experienced during the formation phase of the disk could have an impact on nucleosynthesis outputs from neutrino-driven and viscously driven outflows. Finally, we find that the effective viscosity due to momentum transport by neutrinos is unlikely to have a strong effect on the growth of the magnetorotational instability in the post-merger accretion disk.

NEUTRON PHYSICS DIVISION ANNUAL PROGRESS REPORT. Period Ending September 1, 1962

DOE Office of Scientific and Technical Information (OSTI.GOV)

None

1963-01-11

A total of 74 subsections are included in the report. The information in 4 subsections was previously abstracted in NSA. Separate abstracts were prepared for 38 of the subsections. Those sections for which no abstracts were prepared contain information on prompt neutron lifetime, Rover critical experiments, Pu/sup 239/ fission, neutron decay, the O5R code, alpha scattering, 8 and P wavelengths, proton scattering, deuteron scattering, local optical potentials, N. S. Savamah radiation leakage, reactor shielding, cross section data analysis, gamma transport, gamma energy deposition, gaussian integration, data interpolation, neutron scattering, neutron energy deposition, space vehicles, computer analyses, shielding, positron sources, andmore » secondary particles. (J.R.D.)« less

The effect of thermal neutron field slagging caused by cylindrical BF3 counters in diffusion media

NASA Technical Reports Server (NTRS)

Gorshkov, G. V.; Tsvetkov, O. S.; Yakovlev, R. M.

1975-01-01

Computations are carried out in transport approximation (first collision method) for the attenuation of the field of thermal neutrons formed in counters of the CHM-8 and CHMO-5 type. The deflection of the thermal neutron field is also obtained near the counters and in the air (shade effect) and in various decelerating media (water, paraffin, plexiglas) for which the calculations are carried out on the basis of diffusion theory. To verify the calculations, the distribution of the density of the thermal neutrons at various distances from the counter in the water is measured.

Improved Fission Neutron Data Base for Active Interrogation of Actinides

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pozzi, Sara; Czirr, J. Bart; Haight, Robert

2013-11-06

This project will develop an innovative neutron detection system for active interrogation measurements. Many active interrogation methods to detect fissionable material are based on the detection of neutrons from fission induced by fast neutrons or high-energy gamma rays. The energy spectrum of the fission neutrons provides data to identify the fissionable isotopes and materials such as shielding between the fissionable material and the detector. The proposed path for the project is as follows. First, the team will develop new neutron detection systems and algorithms by Monte Carlo simulations and bench-top experiments. Next, They will characterize and calibrate detection systems bothmore » with monoenergetic and white neutron sources. Finally, high-fidelity measurements of neutron emission from fissions induced by fast neutrons will be performed. Several existing fission chambers containing U-235, Pu-239, U-238, or Th-232 will be used to measure the neutron-induced fission neutron emission spectra. The challenge for making confident measurements is the detection of neutrons in the energy ranges of 0.01 – 1 MeV and above 8 MeV, regions where the basic data on the neutron energy spectrum emitted from fission is least well known. In addition, improvements in the specificity of neutron detectors are required throughout the complete energy range: they must be able to clearly distinguish neutrons from other radiations, in particular gamma rays and cosmic rays. The team believes that all of these challenges can be addressed successfully with emerging technologies under development by this collaboration. In particular, the collaboration will address the area of fission neutron emission spectra for isotopes of interest in the advanced fuel cycle initiative (AFCI).« less

Mobile Accelerator Neutron Radiography System

DTIC Science & Technology

1984-10-01

panel which is a lamination of ceramic tiles and metal skins, adhesively bonded. The x-ray images the tiles in high contrast, while the neutron...personnel and transported by flatbed truck to its site of operation inside Building 704, a large maintenance hangar located west of the main n:rth-south

Calculation and benchmarking of an azimuthal pressure vessel neutron fluence distribution using the BOXER code and scraping experiments

DOE Office of Scientific and Technical Information (OSTI.GOV)

Holzgrewe, F.; Hegedues, F.; Paratte, J.M.

1995-03-01

The light water reactor BOXER code was used to determine the fast azimuthal neutron fluence distribution at the inner surface of the reactor pressure vessel after the tenth cycle of a pressurized water reactor (PWR). Using a cross-section library in 45 groups, fixed-source calculations in transport theory and x-y geometry were carried out to determine the fast azimuthal neutron flux distribution at the inner surface of the pressure vessel for four different cycles. From these results, the fast azimuthal neutron fluence after the tenth cycle was estimated and compared with the results obtained from scraping test experiments. In these experiments,more » small samples of material were taken from the inner surface of the pressure vessel. The fast neutron fluence was then determined form the measured activity of the samples. Comparing the BOXER and scraping test results have maximal differences of 15%, which is very good, considering the factor of 10{sup 3} neutron attenuation between the reactor core and the pressure vessel. To compare the BOXER results with an independent code, the 21st cycle of the PWR was also calculated with the TWODANT two-dimensional transport code, using the same group structure and cross-section library. Deviations in the fast azimuthal flux distribution were found to be <3%, which verifies the accuracy of the BOXER results.« less

Energy deposition and neutron flux study in a gravity-driven dense granular target (DGT) with GEANT4 toolkit

NASA Astrophysics Data System (ADS)

Zhao, Qiang; Cui, Wenjuan; He, Zhiyong; Zhang, Xueying; Ma, Wenjing

2018-07-01

China initiative Accelerator Driven System (CiADS) has been approved as a strategic plan to build an ADS demonstration facility in the next few years. It proposed a new concept for a high-power spallation target: the gravity-driven dense granular target (DGT). As the same with a monolithic target (MT), both solid and liquid target, energy deposition and neutron flux are two critical issues. In this paper, we focus on these two issues and long for some valuable results for the project. Unlike a solid target, the internal geometry structure of a DGT is very complicated. To be as much as closer with the reality, we designed an algorithm and firstly packed the grains randomly in a cylindrical container in GEANT4 software. The packing result was in great agreement with the experimentally measured results. It shows that the algorithm is practicable. In the next step, all the simulations about energy deposition and neutron flux of a DGT were performed with the GEANT4 codes, and the results were compared with the data of a MT. Compared to a MT, a DGT has inarguable advantages in both terms of energy deposition and neutron flux. In addition, the simulations with different radius of grains were also performed. Finally, we found that both the energy deposition and neutron flux are nearly irrelevant to the radius of the grains in the range of 0.5 mm-5 mm when the packing density is same by analyzing the results meticulously.

Characteristic evaluation of a Lithium-6 loaded neutron coincidence spectrometer.

PubMed

Hayashi, M; Kaku, D; Watanabe, Y; Sagara, K

2007-01-01

Characteristics of a (6)Li-loaded neutron coincidence spectrometer were investigated from both measurements and Monte Carlo simulations. The spectrometer consists of three (6)Li-glass scintillators embedded in a liquid organic scintillator BC-501A, which can detect selectively neutrons that deposit the total energy in the BC-501A using a coincidence signal generated from the capture event of thermalised neutrons in the (6)Li-glass scintillators. The relative efficiency and the energy response were measured using 4.7, 7.2 and 9.0 MeV monoenergetic neutrons. The measured ones were compared with the Monte Carlo calculations performed by combining the neutron transport code PHITS and the scintillator response calculation code SCINFUL. The experimental light output spectra were in good agreement with the calculated ones in shape. The energy dependence of the detection efficiency was reproduced by the calculation. The response matrices for 1-10 MeV neutrons were finally obtained.

Neutron flux measurements on a mock-up of a storage cask for high-level nuclear waste using 2.5 MeV neutrons.

PubMed

Suárez, H Saurí; Becker, F; Klix, A; Pang, B; Döring, T

2018-06-07

To store and dispose spent nuclear fuel, shielding casks are employed to reduce the emitted radiation. To evaluate the exposure of employees handling such casks, Monte Carlo radiation transport codes can be employed. Nevertheless, to assess the reliability of these codes and nuclear data, experimental checks are required. In this study, a neutron generator (NG) producing neutrons of 2.5 MeV was employed to simulate neutrons produced in spent nuclear fuel. Different configurations of shielding layers of steel and polyethylene were positioned between the target of the NG and a NE-213 detector. The results of the measurements of neutron and γ radiation and the corresponding simulations with the code MCNP6 are presented. Details of the experimental set-up as well as neutron and photon flux spectra are provided as reference points for such NG investigations with shielding structures.

Measured and Calculated Neutron Spectra and Dose Equivalent Rates at High Altitudes; Relevance to SST Operations and Space Research

NASA Technical Reports Server (NTRS)

Foelsche, T.; Mendell, R. B.; Wilson, J. W.; Adams, R. R.

1974-01-01

Results of the NASA Langley-New York University high-altitude radiation study are presented. Measurements of the absorbed dose rate and of secondary fast neutrons (1 to 10 MeV energy) during the years 1965 to 1971 are used to determine the maximum radiation exposure from galactic and solar cosmic rays of supersonic transport (SST) and subsonic jet occupants. The maximum dose equivalent rates that the SST crews might receive turn out to be 13 to 20 percent of the maximum permissible dose rate (MPD) for radiation workers (5 rem/yr). The exposure of passengers encountering an intense giant-energy solar particle event could exceed the MPD for the general population (0.5 rem/yr), but would be within these permissible limits if in such rare cases the transport descends to subsonic altitude; it is in general less than 12 percent of the MPD. By Monte Carlo calculations of the transport and buildup of nucleons in air for incident proton energies E of 0.02 to 10 GeV, the measured neutron spectra were extrapolated to lower and higher energies and for galactic cosmic rays were found to continue with a relatively high intensity to energies greater than 400 MeV, in a wide altitude range. This condition, together with the measured intensity profiles of fast neutrons, revealed that the biologically important fast and energetic neutrons penetrate deep into the atmosphere and contribute approximately 50 percent of the dose equivalant rates at SST and present subsonic jet altitudes.

Neutron counter based on beryllium activation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bienkowska, B.; Prokopowicz, R.; Kaczmarczyk, J.

2014-08-21

The fusion reaction occurring in DD plasma is followed by emission of 2.45 MeV neutrons, which carry out information about fusion reaction rate and plasma parameters and properties as well. Neutron activation of beryllium has been chosen for detection of DD fusion neutrons. The cross-section for reaction {sup 9}Be(n, α){sup 6}He has a useful threshold near 1 MeV, which means that undesirable multiple-scattered neutrons do not undergo that reaction and therefore are not recorded. The product of the reaction, {sup 6}He, decays with half-life T{sub 1/2} = 0.807 s emitting β{sup −} particles which are easy to detect. Large areamore » gas sealed proportional detector has been chosen as a counter of β–particles leaving activated beryllium plate. The plate with optimized dimensions adjoins the proportional counter entrance window. Such set-up is also equipped with appropriate electronic components and forms beryllium neutron activation counter. The neutron flux density on beryllium plate can be determined from the number of counts. The proper calibration procedure needs to be performed, therefore, to establish such relation. The measurements with the use of known β–source have been done. In order to determine the detector response function such experiment have been modeled by means of MCNP5–the Monte Carlo transport code. It allowed proper application of the results of transport calculations of β{sup −} particles emitted from radioactive {sup 6}He and reaching proportional detector active volume. In order to test the counter system and measuring procedure a number of experiments have been performed on PF devices. The experimental conditions have been simulated by means of MCNP5. The correctness of simulation outcome have been proved by measurements with known radioactive neutron source. The results of the DD fusion neutron measurements have been compared with other neutron diagnostics.« less

Characterization of the neutron irradiation system for use in the Low-Dose-Rate Irradiation Facility at Sandia National Laboratories.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Franco, Manuel

The objective of this work was to characterize the neutron irradiation system consisting of americium-241 beryllium (241AmBe) neutron sources placed in a polyethylene shielding for use at Sandia National Laboratories (SNL) Low Dose Rate Irradiation Facility (LDRIF). With a total activity of 0.3 TBq (9 Ci), the source consisted of three recycled 241AmBe sources of different activities that had been combined into a single source. The source in its polyethylene shielding will be used in neutron irradiation testing of components. The characterization of the source-shielding system was necessary to evaluate the radiation environment for future experiments. Characterization of the sourcemore » was also necessary because the documentation for the three component sources and their relative alignment within the Special Form Capsule (SFC) was inadequate. The system consisting of the source and shielding was modeled using Monte Carlo N-Particle transport code (MCNP). The model was validated by benchmarking it against measurements using multiple techniques. To characterize the radiation fields over the full spatial geometry of the irradiation system, it was necessary to use a number of instruments of varying sensitivities. First, the computed photon radiography assisted in determining orientation of the component sources. With the capsule properly oriented inside the shielding, the neutron spectra were measured using a variety of techniques. A N-probe Microspec and a neutron Bubble Dosimeter Spectrometer (BDS) set were used to characterize the neutron spectra/field in several locations. In the third technique, neutron foil activation was used to ascertain the neutron spectra. A high purity germanium (HPGe) detector was used to characterize the photon spectrum. The experimentally measured spectra and the MCNP results compared well. Once the MCNP model was validated to an adequate level of confidence, parametric analyses was performed on the model to optimize for potential experimental configurations and neutron spectra for component irradiation. The final product of this work is a MCNP model validated by measurements, an overall understanding of neutron irradiation system including photon/neutron transport and effective dose rates throughout the system, and possible experimental configurations for future irradiation of components.« less

Tomographic Neutron Imaging using SIRT

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gregor, Jens; FINNEY, Charles E A; Toops, Todd J

2013-01-01

Neutron imaging is complementary to x-ray imaging in that materials such as water and plastic are highly attenuating while material such as metal is nearly transparent. We showcase tomographic imaging of a diesel particulate filter. Reconstruction is done using a modified version of SIRT called PSIRT. We expand on previous work and introduce Tikhonov regularization. We show that near-optimal relaxation can still be achieved. The algorithmic ideas apply to cone beam x-ray CT and other inverse problems.

On the Compton scattering redistribution function in plasma

NASA Astrophysics Data System (ADS)

Madej, J.; Różańska, A.; Majczyna, A.; Należyty, M.

2017-08-01

Compton scattering is the dominant opacity source in hot neutron stars, accretion discs around black holes and hot coronae. We collected here a set of numerical expressions of the Compton scattering redistribution functions (RFs) for unpolarized radiation, which are more exact than the widely used Kompaneets equation. The principal aim of this paper is the presentation of the RF by Guilbert, which is corrected for the computational errors in the original paper. This corrected RF was used in the series of papers on model atmosphere computations of hot neutron stars. We have also organized four existing algorithms for the RF computations into a unified form ready to use in radiative transfer and model atmosphere codes. The exact method by Nagirner & Poutanen was numerically compared to all other algorithms in a very wide spectral range from hard X-rays to radio waves. Sample computations of the Compton scattering RFs in thermal plasma were done for temperatures corresponding to the atmospheres of bursting neutron stars and hot intergalactic medium. Our formulae are also useful to study the Compton scattering of unpolarized microwave background radiation in hot intracluster gas and the Sunyaev-Zeldovich effect. We conclude that the formulae by Guilbert and the exact quantum mechanical formulae yield practically the same RFs for gas temperatures relevant to the atmospheres of X-ray bursting neutron stars, T ≤ 108 K.

The Use of Fast Neutron Detection for Materials Accountability

NASA Astrophysics Data System (ADS)

Nakae, L. F.; Chapline, G. F.; Glenn, A. M.; Kerr, P. L.; Kim, K. S.; Ouedraogo, S. A.; Prasad, M. K.; Sheets, S. A.; Snyderman, N. J.; Verbeke, J. M.; Wurtz, R. E.

2014-02-01

For many years at LLNL, we have been developing time-correlated neutron detection techniques and algorithms for applications such as Arms Control, Threat Detection and Nuclear Material Assay. Many of our techniques have been developed specifically for the relatively low efficiency (a few percent) inherent in man-portable systems. Historically, thermal neutron detectors (mainly 3He) were used, taking advantage of the high thermal neutron interaction cross-sections, but more recently we have been investigating the use of fast neutron detection with liquid scintillators, inorganic crystals, and in the near future, pulse-shape discriminating plastics that respond over 1000 times faster (nanoseconds versus tens of microseconds) than thermal neutron detectors. Fast neutron detection offers considerable advantages, since the inherent nanosecond production timescales of fission and neutron-induced fission are preserved and measured instead of being lost in the thermalization of thermal neutron detectors. We are now applying fast neutron technology to the safeguards regime in the form of high efficiency counters. Faster detector response times and sensitivity to neutron momentum show promise in measuring, differentiating, and assaying samples that have modest to very high count rates, as well as mixed neutron sources (e.g., Pu oxide or Mixed Cm and Pu). Here we report on measured results with our existing liquid scintillator array and promote the design of a nuclear material assay system that incorporates fast neutron detection, including the surprising result that fast liquid scintillator becomes competitive and even surpasses the precision of 3He counters measuring correlated pairs in modest (kg) samples of plutonium.

Retrospective dosimetry analyses of reactor vessel cladding samples

DOE Office of Scientific and Technical Information (OSTI.GOV)

Greenwood, L. R.; Soderquist, C. Z.; Fero, A. H.

2011-07-01

Reactor pressure vessel cladding samples for Ringhals Units 3 and 4 in Sweden were analyzed using retrospective reactor dosimetry techniques. The objective was to provide the best estimates of the neutron fluence for comparison with neutron transport calculations. A total of 51 stainless steel samples consisting of chips weighing approximately 100 to 200 mg were removed from selected locations around the pressure vessel and were sent to Pacific Northwest National Laboratory for analysis. The samples were fully characterized and analyzed for radioactive isotopes, with special interest in the presence of Nb-93m. The RPV cladding retrospective dosimetry results will be combinedmore » with a re-evaluation of the surveillance capsule dosimetry and with ex-vessel neutron dosimetry results to form a comprehensive 3D comparison of measurements to calculations performed with 3D deterministic transport code. (authors)« less

Road screening and distribution route multi-objective robust optimization for hazardous materials based on neural network and genetic algorithm.

PubMed

Ma, Changxi; Hao, Wei; Pan, Fuquan; Xiang, Wang

2018-01-01

Route optimization of hazardous materials transportation is one of the basic steps in ensuring the safety of hazardous materials transportation. The optimization scheme may be a security risk if road screening is not completed before the distribution route is optimized. For road screening issues of hazardous materials transportation, a road screening algorithm of hazardous materials transportation is built based on genetic algorithm and Levenberg-Marquardt neural network (GA-LM-NN) by analyzing 15 attributes data of each road network section. A multi-objective robust optimization model with adjustable robustness is constructed for the hazardous materials transportation problem of single distribution center to minimize transportation risk and time. A multi-objective genetic algorithm is designed to solve the problem according to the characteristics of the model. The algorithm uses an improved strategy to complete the selection operation, applies partial matching cross shift and single ortho swap methods to complete the crossover and mutation operation, and employs an exclusive method to construct Pareto optimal solutions. Studies show that the sets of hazardous materials transportation road can be found quickly through the proposed road screening algorithm based on GA-LM-NN, whereas the distribution route Pareto solutions with different levels of robustness can be found rapidly through the proposed multi-objective robust optimization model and algorithm.

MARS15

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mokhov, Nikolai

MARS is a Monte Carlo code for inclusive and exclusive simulation of three-dimensional hadronic and electromagnetic cascades, muon, heavy-ion and low-energy neutron transport in accelerator, detector, spacecraft and shielding components in the energy range from a fraction of an electronvolt up to 100 TeV. Recent developments in the MARS15 physical models of hadron, heavy-ion and lepton interactions with nuclei and atoms include a new nuclear cross section library, a model for soft pion production, the cascade-exciton model, the quark gluon string models, deuteron-nucleus and neutrino-nucleus interaction models, detailed description of negative hadron and muon absorption and a unified treatment ofmore » muon, charged hadron and heavy-ion electromagnetic interactions with matter. New algorithms are implemented into the code and thoroughly benchmarked against experimental data. The code capabilities to simulate cascades and generate a variety of results in complex media have been also enhanced. Other changes in the current version concern the improved photo- and electro-production of hadrons and muons, improved algorithms for the 3-body decays, particle tracking in magnetic fields, synchrotron radiation by electrons and muons, significantly extended histograming capabilities and material description, and improved computational performance. In addition to direct energy deposition calculations, a new set of fluence-to-dose conversion factors for all particles including neutrino are built into the code. The code includes new modules for calculation of Displacement-per-Atom and nuclide inventory. The powerful ROOT geometry and visualization model implemented in MARS15 provides a large set of geometrical elements with a possibility of producing composite shapes and assemblies and their 3D visualization along with a possible import/export of geometry descriptions created by other codes (via the GDML format) and CAD systems (via the STEP format). The built-in MARS-MAD Beamline Builder (MMBLB) was redesigned for use with the ROOT geometry package that allows a very efficient and highly-accurate description, modeling and visualization of beam loss induced effects in arbitrary beamlines and accelerator lattices. The MARS15 code includes links to the MCNP-family codes for neutron and photon production and transport below 20 MeV, to the ANSYS code for thermal and stress analyses and to the STRUCT code for multi-turn particle tracking in large synchrotrons and collider rings.« less

Benchmarking of neutron production of heavy-ion transport codes

DOE Office of Scientific and Technical Information (OSTI.GOV)

Remec, I.; Ronningen, R. M.; Heilbronn, L.

Document available in abstract form only, full text of document follows: Accurate prediction of radiation fields generated by heavy ion interactions is important in medical applications, space missions, and in design and operation of rare isotope research facilities. In recent years, several well-established computer codes in widespread use for particle and radiation transport calculations have been equipped with the capability to simulate heavy ion transport and interactions. To assess and validate these capabilities, we performed simulations of a series of benchmark-quality heavy ion experiments with the computer codes FLUKA, MARS15, MCNPX, and PHITS. We focus on the comparisons of secondarymore » neutron production. Results are encouraging; however, further improvements in models and codes and additional benchmarking are required. (authors)« less

Optimization of 6LiF:ZnS(Ag) scintillator light yield using GEANT4

NASA Astrophysics Data System (ADS)

Yehuda-Zada, Y.; Pritchard, K.; Ziegler, J. B.; Cooksey, C.; Siebein, K.; Jackson, M.; Hurlbut, C.; Kadmon, Y.; Cohen, Y.; Ibberson, R. M.; Majkrzak, C. F.; Maliszewskyj, N. C.; Orion, I.; Osovizky, A.

2018-06-01

A new cold neutron detector has been developed at the NIST Center for Neutron Research (NCNR) for the CANDoR (Chromatic Analysis Neutron Diffractometer or Reflectometer) project. Geometric and performance constraints dictate that this detector be exceptionally thin (∼ 2 mm). For this reason, the design of the detector consists of a 6LiF:ZnS(Ag) scintillator with embedded wavelength shifting (WLS) fibers. We used the GEANT4 package to simulate neutron capture and light transport in the detector to optimize the composition and arrangement of materials to satisfy the competing requirements of high neutron capture probability and light production and transport. In the process, we have developed a method for predicting light collection and total neutron detection efficiency for different detector configurations. The simulation was performed by adjusting crucial parameters such as the scintillator stoichiometry, light yield, component grain size, WLS fiber geometry, and reflectors at the outside edges of the scintillator volume. Three different detector configurations were fabricated and their test results were correlated with the simulations. Through this correlation we have managed to find a common photon threshold for the different detector configurations which was then used to simulate and predict the efficiencies for many other detector configurations. New detectors that have been fabricated based on simulation results yielding the desired sensitivity of 90% for 3.27 meV (5 Å) cold neutrons. The simulation has proven to be a useful tool by dramatically reducing the development period and the required number of detector prototypes. It can be used to test new designs with different thicknesses and different target neutron energies.

A One-group, One-dimensional Transport Benchmark in Cylindrical Geometry

DOE Office of Scientific and Technical Information (OSTI.GOV)

Barry Ganapol; Abderrafi M. Ougouag

A 1-D, 1-group computational benchmark in cylndrical geometry is described. This neutron transport benchmark is useful for evaluating reactor concepts that possess azimuthal symmetry such as a pebble-bed reactor.

Neutronics calculation of RTP core

NASA Astrophysics Data System (ADS)

Rabir, Mohamad Hairie B.; Zin, Muhammad Rawi B. Mohamed; Karim, Julia Bt. Abdul; Bayar, Abi Muttaqin B. Jalal; Usang, Mark Dennis Anak; Mustafa, Muhammad Khairul Ariff B.; Hamzah, Na'im Syauqi B.; Said, Norfarizan Bt. Mohd; Jalil, Muhammad Husamuddin B.

2017-01-01

Reactor calculation and simulation are significantly important to ensure safety and better utilization of a research reactor. The Malaysian's PUSPATI TRIGA Reactor (RTP) achieved initial criticality on June 28, 1982. The reactor is designed to effectively implement the various fields of basic nuclear research, manpower training, and production of radioisotopes. Since early 90s, neutronics modelling were used as part of its routine in-core fuel management activities. The are several computer codes have been used in RTP since then, based on 1D neutron diffusion, 2D neutron diffusion and 3D Monte Carlo neutron transport method. This paper describes current progress and overview on neutronics modelling development in RTP. Several important parameters were analysed such as keff, reactivity, neutron flux, power distribution and fission product build-up for the latest core configuration. The developed core neutronics model was validated by means of comparison with experimental and measurement data. Along with the RTP core model, the calculation procedure also developed to establish better prediction capability of RTP's behaviour.

CHARACTERIZATION OF A THIN SILICON SENSOR FOR ACTIVE NEUTRON PERSONAL DOSEMETERS.

PubMed

Takada, M; Nunomiya, T; Nakamura, T; Matsumoto, T; Masuda, A

2016-09-01

A thin silicon sensor has been developed for active neutron personal dosemeters for use by aircrews and first responders. This thin silicon sensor is not affected by the funneling effect, which causes detection of cosmic protons and over-response to cosmic neutrons. There are several advantages to the thin silicon sensor: a decrease in sensitivity to gamma rays, an improvement of the energy detection limit for neutrons down to 0.8 MeV and an increase in the sensitivity to fast neutrons. Neutron response functions were experimentally obtained using 2.5 and 5 MeV monoenergy neutron beams and a (252)Cf neutron source. Simulation results using the Monte Carlo N-Particle transport code agree quite well with the experimental ones when an energy deposition region shaped like a circular truncated cone is used in place of a cylindrical region. © The Author 2016. Published by Oxford University Press. All rights reserved. For Permissions, please email: journals.permissions@oup.com.

A Study of Neutron Leakage in Finite Objects

NASA Technical Reports Server (NTRS)

Wilson, John W.; Slaba, Tony C.; Badavi, Francis F.; Reddell, Brandon D.; Bahadori, Amir A.

2015-01-01

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 simple shielded objects. Monte Carlo (MC) benchmarks were used to verify the 3DHZETRN methodology in slab and spherical geometry, and it was shown that 3DHZETRN agrees with MC codes to the degree that various MC codes agree among themselves. One limitation in the verification process is that all of the codes (3DHZETRN and three MC codes) utilize different nuclear models/databases. In the present report, the new algorithm, with well-defined convergence criteria, is used to quantify the neutron leakage from simple geometries to provide means of verifying 3D effects and to provide guidance for further code development.

On neutron-gamma mixed field dosimetry with LiF:Mg,Ti at radiation protection dose levels.

PubMed

Weinstein, M; German, U; Alfassi, Z B

2006-01-01

The possibility of using the specific responses of the high temperature Peaks 6 and 7 and Peaks 4 and 5 to different LET radiations was mentioned in the past mainly for very high doses. The applicability of the two regions method for thermal neutrons--gamma ray mixed field dosimetry was investigated by analysing the response of LiF:Mg,Ti dosemeters irradiated to different ratios of thermal neutrons and gamma rays at radiation protection dose levels encountered in routine work conditions, up to approximately 50 mSv. The Region of Interest method was used to define the areas of the Peaks 4 + 5 and 6 + 7. We found that a simple algorithm can be used to determine with good accuracy the separate contributions of neutron and gamma doses.

Method for non-intrusively identifying a contained material utilizing uncollided nuclear transmission measurements

DOEpatents

Morrison, John L.; Stephens, Alan G.; Grover, S. Blaine

2001-11-20

An improved nuclear diagnostic method identifies a contained target material by measuring on-axis, mono-energetic uncollided particle radiation transmitted through a target material for two penetrating radiation beam energies, and applying specially developed algorithms to estimate a ratio of macroscopic neutron cross-sections for the uncollided particle radiation at the two energies, where the penetrating radiation is a neutron beam, or a ratio of linear attenuation coefficients for the uncollided particle radiation at the two energies, where the penetrating radiation is a gamma-ray beam. Alternatively, the measurements are used to derive a minimization formula based on the macroscopic neutron cross-sections for the uncollided particle radiation at the two neutron beam energies, or the linear attenuation coefficients for the uncollided particle radiation at the two gamma-ray beam energies. A candidate target material database, including known macroscopic neutron cross-sections or linear attenuation coefficients for target materials at the selected neutron or gamma-ray beam energies, is used to approximate the estimated ratio or to solve the minimization formula, such that the identity of the contained target material is discovered.

Paving the Road for Modern Particle Therapy - What Can We Learn from the Experience Gained with Fast Neutron Therapy in Munich?

PubMed

Specht, Hanno M; Neff, Teresa; Reuschel, Waltraud; Wagner, Franz M; Kampfer, Severin; Wilkens, Jan J; Petry, Winfried; Combs, Stephanie E

2015-01-01

While neutron therapy was a highly topical subject in the 70s and 80s, today there are only a few remaining facilities offering fast neutron therapy (FNT). Nevertheless, up to today more than 30,000 patients were treated with neutron therapy. For some indications like salivary gland tumors and malignant melanoma, there is clinical evidence that the addition of FNT leads to superior local control compared to photon treatment alone. FNT was available in Munich from 1985 until 2000 at the Reactor Neutron Therapy (RENT) facility. Patient treatment continued at the new research reactor FRM II in 2007 under improved treatment conditions, and today it can still be offered to selected patients as an individual treatment option. As there is a growing interest in high-linear energy transfer (LET) therapy with new hadron therapy centers emerging around the globe, the clinical data generated by neutron therapy might help to develop biologically driven treatment planning algorithms. Also FNT might experience its resurgence as a combinational partner of modern immunotherapies.

Paving the Road for Modern Particle Therapy – What Can We Learn from the Experience Gained with Fast Neutron Therapy in Munich?

PubMed Central

Specht, Hanno M.; Neff, Teresa; Reuschel, Waltraud; Wagner, Franz M.; Kampfer, Severin; Wilkens, Jan J.; Petry, Winfried; Combs, Stephanie E.

2015-01-01

While neutron therapy was a highly topical subject in the 70s and 80s, today there are only a few remaining facilities offering fast neutron therapy (FNT). Nevertheless, up to today more than 30,000 patients were treated with neutron therapy. For some indications like salivary gland tumors and malignant melanoma, there is clinical evidence that the addition of FNT leads to superior local control compared to photon treatment alone. FNT was available in Munich from 1985 until 2000 at the Reactor Neutron Therapy (RENT) facility. Patient treatment continued at the new research reactor FRM II in 2007 under improved treatment conditions, and today it can still be offered to selected patients as an individual treatment option. As there is a growing interest in high-linear energy transfer (LET) therapy with new hadron therapy centers emerging around the globe, the clinical data generated by neutron therapy might help to develop biologically driven treatment planning algorithms. Also FNT might experience its resurgence as a combinational partner of modern immunotherapies. PMID:26640777

MCNP simulation of the dose distribution in liver cancer treatment for BNC therapy

NASA Astrophysics Data System (ADS)

Krstic, Dragana; Jovanovic, Zoran; Markovic, Vladimir; Nikezic, Dragoslav; Urosevic, Vlade

2014-10-01

The Boron Neutron Capture Therapy ( BNCT) is based on selective uptake of boron in tumour tissue compared to the surrounding normal tissue. Infusion of compounds with boron is followed by irradiation with neutrons. Neutron capture on 10B, which gives rise to an alpha particle and recoiled 7Li ion, enables the therapeutic dose to be delivered to tumour tissue while healthy tissue can be spared. Here, therapeutic abilities of BNCT were studied for possible treatment of liver cancer using thermal and epithermal neutron beam. For neutron transport MCNP software was used and doses in organs of interest in ORNL phantom were evaluated. Phantom organs were filled with voxels in order to obtain depth-dose distributions in them. The result suggests that BNCT using an epithermal neutron beam could be applied for liver cancer treatment.

A Novel Segment-Based Approach for Improving Classification Performance of Transport Mode Detection.

PubMed

Guvensan, M Amac; Dusun, Burak; Can, Baris; Turkmen, H Irem

2017-12-30

Transportation planning and solutions have an enormous impact on city life. To minimize the transport duration, urban planners should understand and elaborate the mobility of a city. Thus, researchers look toward monitoring people's daily activities including transportation types and duration by taking advantage of individual's smartphones. This paper introduces a novel segment-based transport mode detection architecture in order to improve the results of traditional classification algorithms in the literature. The proposed post-processing algorithm, namely the Healing algorithm, aims to correct the misclassification results of machine learning-based solutions. Our real-life test results show that the Healing algorithm could achieve up to 40% improvement of the classification results. As a result, the implemented mobile application could predict eight classes including stationary, walking, car, bus, tram, train, metro and ferry with a success rate of 95% thanks to the proposed multi-tier architecture and Healing algorithm.

Novel neutron sources at the Radiological Research Accelerator Facility

PubMed Central

Xu, Yanping; Garty, Guy; Marino, Stephen A.; Massey, Thomas N.; Randers-Pehrson, Gerhard; Johnson, Gary W.; Brenner, David J.

2012-01-01

Since the 1960s, the Radiological Research Accelerator Facility (RARAF) has been providing researchers in biology, chemistry and physics with advanced irradiation techniques, using charged particles, photons and neutrons. We are currently developing a unique facility at RARAF, to simulate neutron spectra from an improvised nuclear device (IND), based on calculations of the neutron spectrum at 1.5 km from the epicenter of the Hiroshima atom bomb. This is significantly different from a standard fission spectrum, because the spectrum changes as the neutrons are transported through air, and is dominated by neutron energies between 0.05 and 8 MeV. This facility will be based on a mixed proton/deuteron beam impinging on a thick beryllium target. A second, novel facility under development is our new neutron microbeam. The neutron microbeam will, for the first time, provide a kinematically collimated neutron beam, 10–20 micron in diameter. This facility is based on a Proton Microbeam, impinging on a thin lithium target near the threshold of the 7Li(p,n)7Be reaction. This novel neutron microbeam will enable studies of neutron damage to small targets, such as single cells, individual organs within small animals or microelectronic components. PMID:22545061

Novel neutron sources at the Radiological Research Accelerator Facility.

PubMed

Xu, Yanping; Garty, Guy; Marino, Stephen A; Massey, Thomas N; Randers-Pehrson, Gerhard; Johnson, Gary W; Brenner, David J

2012-03-16

Since the 1960s, the Radiological Research Accelerator Facility (RARAF) has been providing researchers in biology, chemistry and physics with advanced irradiation techniques, using charged particles, photons and neutrons.We are currently developing a unique facility at RARAF, to simulate neutron spectra from an improvised nuclear device (IND), based on calculations of the neutron spectrum at 1.5 km from the epicenter of the Hiroshima atom bomb. This is significantly different from a standard fission spectrum, because the spectrum changes as the neutrons are transported through air, and is dominated by neutron energies between 0.05 and 8 MeV. This facility will be based on a mixed proton/deuteron beam impinging on a thick beryllium target.A second, novel facility under development is our new neutron microbeam. The neutron microbeam will, for the first time, provide a kinematically collimated neutron beam, 10-20 micron in diameter. This facility is based on a Proton Microbeam, impinging on a thin lithium target near the threshold of the (7)Li(p,n)(7)Be reaction. This novel neutron microbeam will enable studies of neutron damage to small targets, such as single cells, individual organs within small animals or microelectronic components.

Novel neutron sources at the Radiological Research Accelerator Facility

DOE PAGES

Xu, Yanping; Garty, G.; Marino, S. A.; ...

2012-03-16

Since the 1960s, the Radiological Research Accelerator Facility (RARAF) has been providing researchers in biology, chemistry and physics with advanced irradiation techniques, using charged particles, photons and neutrons. We are currently developing a unique facility at RARAF, to simulate neutron spectra from an improvised nuclear device (IND), based on calculations of the neutron spectrum at 1.5 km from the epicenter of the Hiroshima atom bomb. This is significantly different from a standard fission spectrum, because the spectrum changes as the neutrons are transported through air, and is dominated by neutron energies between 0.05 and 8 MeV. This facility will bemore » based on a mixed proton/deuteron beam impinging on a thick beryllium target. A second, novel facility under development is our new neutron microbeam. The neutron microbeam will, for the first time, provide a kinematically collimated neutron beam, 10-20 micron in diameter. This facility is based on a proton microbeam, impinging on a thin lithium target near the threshold of the Li-7(p,n)Be-7 reaction. Lastly, this novel neutron microbeam will enable studies of neutron damage to small targets, such as single cells, individual organs within small animals or microelectronic components.« less

Novel neutron sources at the Radiological Research Accelerator Facility

NASA Astrophysics Data System (ADS)

Xu, Y.; Garty, G.; Marino, S. A.; Massey, T. N.; Randers-Pehrson, G.; Johnson, G. W.; Brenner, D. J.

2012-03-01

Since the 1960s, the Radiological Research Accelerator Facility (RARAF) has been providing researchers in biology, chemistry and physics with advanced irradiation techniques, using charged particles, photons and neutrons. We are currently developing a unique facility at RARAF, to simulate neutron spectra from an improvised nuclear device (IND), based on calculations of the neutron spectrum at 1.5 km from the epicenter of the Hiroshima atom bomb. This is significantly different from a standard fission spectrum, because the spectrum changes as the neutrons are transported through air, and is dominated by neutron energies between 0.05 and 8 MeV. This facility will be based on a mixed proton/deuteron beam impinging on a thick beryllium target. A second, novel facility under development is our new neutron microbeam. The neutron microbeam will, for the first time, provide a kinematically collimated neutron beam, 10-20 micron in diameter. This facility is based on a proton microbeam, impinging on a thin lithium target near the threshold of the 7Li(p,n)7Be reaction. This novel neutron microbeam will enable studies of neutron damage to small targets, such as single cells, individual organs within small animals or microelectronic components.

Associated-particle sealed-tube neutron probe for nonintrusive inspection

NASA Astrophysics Data System (ADS)

Rhodes, E.; Dickerman, C. E.

1997-02-01

The development and investigation of a small associated-particle sealed-tube neutron generator (APSTNG) show potential for the associated-particle method to move out of the laboratory into field applications. This paper is a review of ANL investigations of this technology. Alpha particles associated with 14-MeV neutrons generated from the D-T reaction travel in the opposite direction and are detected inside the sealed tube. Gamma-ray spectra of resulting neutron reactions in the inspected volume encompassed by the alpha-detector solid angle identify many nuclides. Flight-times determined from detection times of the gamma rays and alpha particles separate the prompt and delayed gamma-rays and can yield a separate coarse tomographic image of each identified nuclide, from a single orientation without collimation. A continuous ion beam allows data acquisition by relatively low-bandwidth electronics. When a compact sealed-tube neutron generator is used, a relatively small and easily maintainable inspection system can be developed, that is rugged enough to be transportable. Proof-of-concept laboratory experiments have been performed for simulated explosives, drugs, special nuclear materials, and chemical warfare agents. Efficient collection of maximum information from each detected neutron with low background rates can allow a much lower source intensity than pulsed accelerator methods and yield a preference for an APSTNG system, when it can provide adequate usable source intensity. Based on lessons learned with the present system, an advanced APSTNG system is being designed and built that will be transportable and yield substantial increases in neutron output and target lifetime.

Measurements and analysis of leakage neutron spectra from multiple-slab sample assemblies comprising W,U,C, and CH2 with D-T neutron irradiation.

PubMed

Luo, F; Han, R; Chen, Z; Nie, Y; Sun, Q; Shi, F; Zhang, S; Tian, G; Song, L; Ruan, X; Ye, M Y

2018-07-01

The accelerator driven subcritical system (ADS) is regarded as a safe and clean nuclear power system, which can be used for the transmutation of nuclear waste and the breeding of nuclear fuel. In this study, in order to validate nuclear data and the neutron transportation performance of the materials related to ADS, we measured the leakage neutron spectra from multiple-slab sample assemblies using 14.8 MeV D-T neutrons. Two types of assemblies comprising A-1 (W+U+C+CH 2 ) and A-2 (U+C+CH 2 ) were both built up gradually starting with the first wall. The measured spectra were compared with those calculated using the Monte Carlo code neutron transport coed (MCNP)-4C. A comparison of the results showed that the experimental leakage neutron spectra for both A-1 or A-2 were reproduced well by the three evaluated nuclear data libraries with discrepancies of less than 15% (A-1) and 12% (A-2), except when below 3 MeV. For 2-cm and 5-cm uranium samples, the CENDL-3.1 calculations exhibited large discrepancies in the energy range of 2-8 MeV and above 13 MeV. Thus, the CENDL-3.1 library for uranium should be reevaluated, especially around this energy range. It was significant that the leakage neuron spectra changed clearly when the latest material layer was added during the building of assemblies A-1 and A-2. Copyright © 2018 Elsevier Ltd. All rights reserved.

A United Effort for Crystal Growth, Neutron Scattering, and X-ray Scattering Studies of Novel Correlated Electron Materials

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lee, Young S.

2015-02-12

The research accomplishments during the award involved experimental studies of correlated electron systems and quantum magnetism. The techniques of crystal growth, neutron scattering, x-ray scattering, and thermodynamic & transport measurements were employed, and graduate students and postdoctoral research associates were trained in these techniques.

Intercomparison of Monte Carlo radiation transport codes to model TEPC response in low-energy neutron and gamma-ray fields.

PubMed

Ali, F; Waker, A J; Waller, E J

2014-10-01

Tissue-equivalent proportional counters (TEPC) can potentially be used as a portable and personal dosemeter in mixed neutron and gamma-ray fields, but what hinders this use is their typically large physical size. To formulate compact TEPC designs, the use of a Monte Carlo transport code is necessary to predict the performance of compact designs in these fields. To perform this modelling, three candidate codes were assessed: MCNPX 2.7.E, FLUKA 2011.2 and PHITS 2.24. In each code, benchmark simulations were performed involving the irradiation of a 5-in. TEPC with monoenergetic neutron fields and a 4-in. wall-less TEPC with monoenergetic gamma-ray fields. The frequency and dose mean lineal energies and dose distributions calculated from each code were compared with experimentally determined data. For the neutron benchmark simulations, PHITS produces data closest to the experimental values and for the gamma-ray benchmark simulations, FLUKA yields data closest to the experimentally determined quantities. © The Author 2013. Published by Oxford University Press. All rights reserved. For Permissions, please email: journals.permissions@oup.com.

Transport analysis of measured neutron leakage spectra from spheres as tests of evaluated high energy cross sections

NASA Technical Reports Server (NTRS)

Bogart, D. D.; Shook, D. F.; Fieno, D.

1973-01-01

Integral tests of evaluated ENDF/B high-energy cross sections have been made by comparing measured and calculated neutron leakage flux spectra from spheres of various materials. An Am-Be (alpha,n) source was used to provide fast neutrons at the center of the test spheres of Be, CH2, Pb, Nb, Mo, Ta, and W. The absolute leakage flux spectra were measured in the energy range 0.5 to 12 MeV using a calibrated NE213 liquid scintillator neutron spectrometer. Absolute calculations of the spectra were made using version 3 ENDF/B cross sections and an S sub n discrete ordinates multigroup transport code. Generally excellent agreement was obtained for Be, CH2, Pb, and Mo, and good agreement was observed for Nb although discrepancies were observed for some energy ranges. Poor comparative results, obtained for Ta and W, are attributed to unsatisfactory nonelastic cross sections. The experimental sphere leakage flux spectra are tabulated and serve as possible benchmarks for these elements against which reevaluated cross sections may be tested.

A method to optimize the shield compact and lightweight combining the structure with components together by genetic algorithm and MCNP code.

PubMed

Cai, Yao; Hu, Huasi; Pan, Ziheng; Hu, Guang; Zhang, Tao

2018-05-17

To optimize the shield for neutrons and gamma rays compact and lightweight, a method combining the structure and components together was established employing genetic algorithms and MCNP code. As a typical case, the fission energy spectrum of 235 U which mixed neutrons and gamma rays was adopted in this study. Six types of materials were presented and optimized by the method. Spherical geometry was adopted in the optimization after checking the geometry effect. Simulations have made to verify the reliability of the optimization method and the efficiency of the optimized materials. To compare the materials visually and conveniently, the volume and weight needed to build a shield are employed. The results showed that, the composite multilayer material has the best performance. Copyright © 2018 Elsevier Ltd. All rights reserved.

Neutron stars in screened modified gravity: Chameleon versus dilaton

NASA Astrophysics Data System (ADS)

Brax, Philippe; Davis, Anne-Christine; Jha, Rahul

2017-04-01

We consider the scalar field profile around relativistic compact objects such as neutron stars for a range of modified gravity models with screening mechanisms of the chameleon and Damour-Polyakov types. We focus primarily on inverse power law chameleons and the environmentally dependent dilaton as examples of both mechanisms. We discuss the modified Tolman-Oppenheimer-Volkoff equation and then implement a relaxation algorithm to solve for the scalar profiles numerically. We find that chameleons and dilatons behave in a similar manner and that there is a large degeneracy between the modified gravity parameters and the neutron star equation of state. This is exemplified by the modifications to the mass-radius relationship for a variety of model parameters.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bushuev, A. V.; Kozhin, A. F., E-mail: alexfkozhin@yandex.ru; Aleeva, T. B.

An active neutron method for measuring the residual mass of {sup 235}U in spent fuel assemblies (FAs) of the IRT MEPhI research reactor is presented. The special measuring stand design and uniform irradiation of the fuel with neutrons along the entire length of the active part of the FA provide high accuracy of determination of the residual {sup 235}U content. AmLi neutron sources yield a higher effect/background ratio than other types of sources and do not induce the fission of {sup 238}U. The proposed method of transfer of the isotope source in accordance with a given algorithm may be usedmore » in experiments where the studied object needs to be irradiated with a uniform fluence.« less

Hybrid Skyshine Calculations for Complex Neutron and Gamma-Ray Sources

DOE Office of Scientific and Technical Information (OSTI.GOV)

Shultis, J. Kenneth

2000-10-15

A two-step hybrid method is described for computationally efficient estimation of neutron and gamma-ray skyshine doses far from a shielded source. First, the energy and angular dependence of radiation escaping into the atmosphere from a source containment is determined by a detailed transport model such as MCNP. Then, an effective point source with this energy and angular dependence is used in the integral line-beam method to transport the radiation through the atmosphere up to 2500 m from the source. An example spent-fuel storage cask is analyzed with this hybrid method and compared to detailed MCNP skyshine calculations.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kelley, R.P., E-mail: rpkelley@ufl.edu; Ray, H.; Jordan, K.A.

An empirical investigation of the scintillation mechanism in a pressurized {sup 4}He gas fast neutron detector was conducted using pulse shape fitting. Scintillation signals from neutron interactions were measured and averaged to produce a single generic neutron pulse shape from both a {sup 252}Cf spontaneous fission source and a (d,d) neutron generator. An expression for light output over time was then developed by treating the decay of helium excited states in the same manner as the decay of radioactive isotopes. This pulse shape expression was fitted to the measured neutron pulse shape using a least-squares optimization algorithm, allowing an empiricalmore » analysis of the mechanism of scintillation inside the {sup 4}He detector. A further understanding of this mechanism in the {sup 4}He detector will advance the use of this system as a neutron spectrometer. For {sup 252}Cf neutrons, the triplet and singlet time constants were found to be 970 ns and 686 ns, respectively. For neutrons from the (d,d) generator, the time constants were found to be 884 ns and 636 ns. Differences were noted in the magnitude of these parameters compared to previously published data, however the general relationships were noted to be the same and checked with expected trends from theory. Of the excited helium states produced from a {sup 252}Cf neutron interaction, 76% were found to be born as triplet states, similar to the result from the neutron generator of 71%. The two sources yielded similar pulse shapes despite having very different neutron energy spectra, validating the robustness of the fits across various neutron energies.« less

A prestorage method to measure neutron transmission of ultracold neutron guides

NASA Astrophysics Data System (ADS)

Blau, B.; Daum, M.; Fertl, M.; Geltenbort, P.; Göltl, L.; Henneck, R.; Kirch, K.; Knecht, A.; Lauss, B.; Schmidt-Wellenburg, P.; Zsigmond, G.

2016-01-01

There are worldwide efforts to search for physics beyond the Standard Model of particle physics. Precision experiments using ultracold neutrons (UCN) require very high intensities of UCN. Efficient transport of UCN from the production volume to the experiment is therefore of great importance. We have developed a method using prestored UCN in order to quantify UCN transmission in tubular guides. This method simulates the final installation at the Paul Scherrer Institute's UCN source where neutrons are stored in an intermediate storage vessel serving three experimental ports. This method allowed us to qualify UCN guides for their intended use and compare their properties.

On-line analysis of bulk materials using pulsed neutron interrogation

NASA Astrophysics Data System (ADS)

Lebrun, P.; Tourneur, P. Le; Poumarede, B.; Möller, H.; Bach, P.

1999-06-01

On the basis of our joint experience in neutronics for SODERN and in cement plant engineering for KRUPP POLYSIUS, we have developed a new on-line bulk materials analyser for the cement industry. This equipment includes a pulsed neutron generator GENIE 16, some gamma ray and neutron detectors, specially designed electronics with high counting rate, software delivering the mean elemental composition of raw material, and adequate shielding. This material is transported through the equipment on a conveyor belt, the size of which is adapted to the requirements. This paper briefly describes the equipment and some results, as obtained in dynamic test from a demonstrator installed in Germany.

Realization of highly efficient hexagonal boron nitride neutron detectors

DOE PAGES

Maity, A.; Doan, T. C.; Li, J.; ...

2016-08-16

Here, we report the achievement of highly efficient 10B enriched hexagonal boron nitride (h- 10BN) direct conversion neutron detectors. These detectors were realized from freestanding 4-in. diameter h- 10BN wafers 43 μm in thickness obtained from epitaxy growth and subsequent mechanical separation from sapphire substrates. Both sides of the film were subjected to ohmic contact deposition to form a simple vertical “photoconductor-type” detector. Transport measurements revealed excellent vertical transport properties including high electrical resistivity (>10 13 Ω cm) and mobility-lifetime (μτ) products. A much larger μτ product for holes compared to that of electrons along the c-axis of h- BNmore » was observed, implying that holes (electrons) behave like majority (minority) carriers in undoped h- BN. Exposure to thermal neutrons from a californium-252 ( 252Cf) source moderated by a high density polyethylene moderator reveals that 43 μm h- 10BN detectors possess 51.4% detection efficiency at a bias voltage of 400 V, which is the highest reported efficiency for any semiconductor-based neutron detector. The results point to the possibility of obtaining highly efficient, compact solid-state neutron detectors with high gamma rejection and low manufacturing and maintenance costs.« less

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

PubMed

Imanaka, T

2001-09-01

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

The diversity and unit of reactor noise theory

NASA Astrophysics Data System (ADS)

Kuang, Zhifeng

The study of reactor noise theory concerns questions about cause and effect relationships, and utilisation of random noise in nuclear reactor systems. The diversity of reactor noise theory arises from the variety of noise sources, the various mathematical treatments applied and various practical purposes. The neutron noise in zero- energy systems arises from the fluctuations in the number of neutrons per fission, the time between nuclear events, and the type of reactions. It can be used to evaluate system parameters. The mathematical treatment is based on the master equation of stochastic branching processes. The noise in power reactor systems is given rise by random processes of technological origin such as vibration of mechanical parts, boiling of the coolant, fluctuations of temperature and pressure. It can be used to monitor reactor behaviour with the possibility of detecting malfunctions at an early stage. The mathematical treatment is based on the Langevin equation. The unity of reactor noise theory arises from the fact that useful information from noise is embedded in the second moments of random variables, which lends the possibility of building up a unified mathematical description and analysis of the various reactor noise sources. Exploring such possibilities is the main subject among the three major topics reported in this thesis. The first subject is within the zero power noise in steady media, and we reported on the extension of the existing theory to more general cases. In Paper I, by use of the master equation approach, we have derived the most general Feynman- and Rossi-alpha formulae so far by taking the full joint statistics of the prompt and all the six groups of delayed neutron precursors, and a multiple emission source into account. The involved problems are solved with a combination of effective analytical techniques and symbolic algebra codes (Mathematica). Paper II gives a numerical evaluation of these formulae. An assessment of the contribution of the terms that are novel as compared to the traditional formulae has been made. The second subject treats a problem in power reactor noise with the Langevin formalism. With a very few exceptions, in all previous work the diffusion approximation was used. In order to extend the treatment to transport theory, in Paper III, we introduced a novel method, i.e. Padé approximation via Lanczos algorithm to calculate the transfer function of a finite slab reactor described by one-group transport equation. It was found that the local-global decomposition of the neutron noise, formerly only reproduced in at least 2- group theory, can be reconstructed. We have also showed the existence of a boundary layer of the neutron noise close to the boundary. Finally, we have explored the possibility of building up a unified theory to account for the coexistence of zero power and power reactor noise in a system. In Paper IV, a unified description of the neutron noise is given by the use of backward master equations in a model where the cross section fluctuations are given as a simple binary pseudorandom process. The general solution contains both the zero power and power reactor noise concurrently, and they can be extracted individually as limiting cases of the general solution. It justified the separate treatments of zero power and power reactor noise. The result was extended to the case including one group of delayed neutron precursors in Paper V.

Neutron flux and gamma dose measurement in the BNCT irradiation facility at the TRIGA reactor of the University of Pavia

NASA Astrophysics Data System (ADS)

Bortolussi, S.; Protti, N.; Ferrari, M.; Postuma, I.; Fatemi, S.; Prata, M.; Ballarini, F.; Carante, M. P.; Farias, R.; González, S. J.; Marrale, M.; Gallo, S.; Bartolotta, A.; Iacoviello, G.; Nigg, D.; Altieri, S.

2018-01-01

University of Pavia is equipped with a TRIGA Mark II research nuclear reactor, operating at a maximum steady state power of 250 kW. It has been used for many years to support Boron Neutron Capture Therapy (BNCT) research. An irradiation facility was constructed inside the thermal column of the reactor to produce a sufficient thermal neutron flux with low epithermal and fast neutron components, and low gamma dose. In this irradiation position, the liver of two patients affected by hepatic metastases from colon carcinoma were irradiated after borated drug administration. The facility is currently used for cell cultures and small animal irradiation. Measurements campaigns have been carried out, aimed at characterizing the neutron spectrum and the gamma dose component. The neutron spectrum has been measured by means of multifoil neutron activation spectrometry and a least squares unfolding algorithm; gamma dose was measured using alanine dosimeters. Results show that in a reference position the thermal neutron flux is (1.20 ± 0.03) ×1010 cm-2 s-1 when the reactor is working at the maximum power of 250 kW, with the epithermal and fast components, respectively, 2 and 3 orders of magnitude lower than the thermal component. The ratio of the gamma dose with respect to the thermal neutron fluence is 1.2 ×10-13 Gy/(n/cm2).

Convergence analysis of two-node CMFD method for two-group neutron diffusion eigenvalue problem

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jeong, Yongjin; Park, Jinsu; Lee, Hyun Chul

2015-12-01

In this paper, the nonlinear coarse-mesh finite difference method with two-node local problem (CMFD2N) is proven to be unconditionally stable for neutron diffusion eigenvalue problems. The explicit current correction factor (CCF) is derived based on the two-node analytic nodal method (ANM2N), and a Fourier stability analysis is applied to the linearized algorithm. It is shown that the analytic convergence rate obtained by the Fourier analysis compares very well with the numerically measured convergence rate. It is also shown that the theoretical convergence rate is only governed by the converged second harmonic buckling and the mesh size. It is also notedmore » that the convergence rate of the CCF of the CMFD2N algorithm is dependent on the mesh size, but not on the total problem size. This is contrary to expectation for eigenvalue problem. The novel points of this paper are the analytical derivation of the convergence rate of the CMFD2N algorithm for eigenvalue problem, and the convergence analysis based on the analytic derivations.« less

The 14 MeV Neutron Irradiation Facility in MARIA Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Prokopowicz, R.; Pytel, K.; Dorosz, M.

2015-07-01

The MARIA reactor with thermal neutron flux density up to 3x10{sup 14} cm{sup -2} s{sup -1} and a number of vertical channels is well suited to material testing by thermal neutron treatment. Beside of that some fast neutron irradiation facilities are operated in MARIA reactor as well. One of them is thermal to 14 MeV neutron converter launched in 2014. It is especially devoted to fusion devices material testing irradiation. The ITER and DEMO research thermonuclear facilities are to be run using the deuterium - tritium fusion reaction. Fast neutrons (of energy approximately 14 MeV) resulting from the reaction aremore » essential to carry away the released thermonuclear energy and to breed tritium. However, constructional materials of which thermonuclear reactors are to be built must be specially selected to survive intense fluxes of fast neutrons. Strong sources of 14 MeV neutrons are needed if research on resistance of candidate materials to such fluxes is to be carried out effectively. Nuclear reactor-based converter capable to convert thermal neutrons into 14 MeV fast neutrons may be used to that purpose. The converter based on two stage nuclear reaction on lithium-6 and deuterium compounds leading to 14 MeV neutron production. The reaction chain is begun by thermal neutron capture by lithium-6 nucleus resulted in triton release. The neutron and triton transport calculations have been therefore carried-out to estimate the thermal to 14 MeV neutron conversion efficiency and optimize converter construction. The usable irradiation space of ca. 60 cm{sup 3} has been obtained. The released energy have been calculated. Heat transport has been asses to ensure proper device cooling. A set of thermocouples has been installed in converter to monitor its temperature distribution on-line. Influence of converter on reactor operation has been studied. Safety analyses of steady states and transients have been done. Performed calculations and analyses allow designing the converter and formulate its operation limits and conditions. During first tested operation of the converter the 14 MeV neutron flux density was estimated to 10{sup 9} cm{sup -2} s{sup -1}, whereas fast fission neutrons inside converter achieved 10{sup 12} cm{sup -2} s{sup -1}, and thermal neutrons were reduced down to 109 cm-2 s-1. Taking into account the feasibility of almost incessant converter operation for a number of months, its arisen as one of the most powerful (in terms of fluence), currently available 14 MeV neutron source. Such a converter currently under operation in the MARIA reactor core will be presented. (authors)« less

Post-merger evolution of a neutron star-black hole binary with neutrino transport

NASA Astrophysics Data System (ADS)

Foucart, Francois; O'Connor, Evan; Roberts, Luke; Duez, Matthew; Kidder, Lawrence; Ott, Christian; Pfeiffer, Harald; Scheel, Mark; Szilagyi, Bela; SXS Collaboration

2015-04-01

We present a first simulation of the post-merger evolution of a black hole-neutron star binary in full general relativity using an energy-integrated truncated moment formalism for neutrino transport. The moment formalism is included as a new module in the SpEC code. We describe the implementation and tests of this new module, and its use to study the formation phase of an accretion disk after a black hole-neutron star merger. We discuss differences with simpler treatments of the neutrinos, the importance of relativistic effects, and the impact of the formation phase of the disk on its expected long-term evolution. We also show that a small amount of material is ejected in the polar region during the circularization of the disk and its interactions with fallback material, and discuss its effects on potential electromagnetic counterparts to the merger.

Observation of enhanced radial transport of energetic ion due to energetic particle mode destabilized by helically-trapped energetic ion in the Large Helical Device

NASA Astrophysics Data System (ADS)

Ogawa, K.; Isobe, M.; Kawase, H.; Nishitani, T.; Seki, R.; Osakabe, M.; LHD Experiment Group

2018-04-01

A deuterium experiment was initiated to achieve higher-temperature and higher-density plasmas in March 2017 in the Large Helical Device (LHD). The central ion temperature notably increases compared with that in hydrogen experiments. However, an energetic particle mode called the helically-trapped energetic-ion-driven resistive interchange (EIC) mode is often excited by intensive perpendicular neutral beam injections on high ion-temperature discharges. The mode leads to significant decrease of the ion temperature or to limiting the sustainment of the high ion-temperature state. To understand the effect of EIC on the energetic ion confinement, the radial transport of energetic ions is studied by means of the neutron flux monitor and vertical neutron camera newly installed on the LHD. Decreases of the line-integrated neutron profile in core channels show that helically-trapped energetic ions are lost from the plasma.

Status Report on NEAMS PROTEUS/ORIGEN Integration

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wieselquist, William A

2016-02-18

The US Department of Energy’s Nuclear Energy Advanced Modeling and Simulation (NEAMS) Program has contributed significantly to the development of the PROTEUS neutron transport code at Argonne National Laboratory and to the Oak Ridge Isotope Generation and Depletion Code (ORIGEN) depletion/decay code at Oak Ridge National Laboratory. PROTEUS’s key capability is the efficient and scalable (up to hundreds of thousands of cores) neutron transport solver on general, unstructured, three-dimensional finite-element-type meshes. The scalability and mesh generality enable the transfer of neutron and power distributions to other codes in the NEAMS toolkit for advanced multiphysics analysis. Recently, ORIGEN has received considerablemore » modernization to provide the high-performance depletion/decay capability within the NEAMS toolkit. This work presents a description of the initial integration of ORIGEN in PROTEUS, mainly performed during FY 2015, with minor updates in FY 2016.« less

An improved numerical method to compute neutron/gamma deexcitation cascades starting from a high spin state

DOE PAGES

Regnier, D.; Litaize, O.; Serot, O.

2015-12-23

Numerous nuclear processes involve the deexcitation of a compound nucleus through the emission of several neutrons, gamma-rays and/or conversion electrons. The characteristics of such a deexcitation are commonly derived from a total statistical framework often called “Hauser–Feshbach” method. In this work, we highlight a numerical limitation of this kind of method in the case of the deexcitation of a high spin initial state. To circumvent this issue, an improved technique called the Fluctuating Structure Properties (FSP) method is presented. Two FSP algorithms are derived and benchmarked on the calculation of the total radiative width for a thermal neutron capture onmore » 238U. We compare the standard method with these FSP algorithms for the prediction of particle multiplicities in the deexcitation of a high spin level of 143Ba. The gamma multiplicity turns out to be very sensitive to the numerical method. The bias between the two techniques can reach 1.5 γγ/cascade. Lastly, the uncertainty of these calculations coming from the lack of knowledge on nuclear structure is estimated via the FSP method.« less

Advanced algorithms for radiographic material discrimination and inspection system design

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gilbert, Andrew J.; McDonald, Benjamin S.; Deinert, Mark R.

X-ray and neutron radiography are powerful tools for non-invasively inspecting the interior of objects. Materials can be discriminated by noting how the radiographic signal changes with variations in the input spectrum or inspection mode. However, current methods are limited in their ability to differentiate when multiple materials are present, especially within large and complex objects. With X-ray radiography, the inability to distinguish materials of a similar atomic number is especially problematic. To overcome these critical limitations, we augmented our existing inverse problem framework with two important expansions: 1) adapting the previous methodology for use with multi-modal radiography and energy-integrating detectors,more » and 2) applying the Cramer-Rao lower bound to select an optimal set of inspection modes for a given application a priori. Adding these expanded capabilities to our algorithmic framework with adaptive regularization, we observed improved discrimination between high-Z materials, specifically plutonium and tungsten. The combined system can estimate plutonium mass within our simulated system to within 1%. Three types of inspection modes were modeled: multi-endpoint X-ray radiography alone; in combination with neutron radiography using deuterium-deuterium (DD); or in combination with neutron radiography using deuterium-tritium (DT) sources.« less

Comparison and Validation of FLUKA and HZETRN as Tools for Investigating the Secondary Neutron Production in Large Space Vehicles

NASA Technical Reports Server (NTRS)

Rojdev, Kristina; Koontz, Steve; Reddell, Brandon; Atwell, William; Boeder, Paul

2015-01-01

NASA's exploration goals are focused on deep space travel and Mars surface operations. To accomplish these goals, large structures will be necessary to transport crew and logistics in the initial stages, and NASA will need to keep the crew and the vehicle safe during transport and any surface activities. One of the major challenges of deep space travel is the space radiation environment and its impacts on the crew, the electronics, and the vehicle materials. The primary radiation from the sun (solar particle events) and from outside the solar system (galactic cosmic rays) interact with materials of the vehicle. These interactions lead to some of the primary radiation being absorbed, being modified, or producing secondary radiation (primarily neutrons). With all vehicles, the high energy primary radiation is of most concern. However, with larger vehicles that have large shielding masses, there is more opportunity for secondary radiation production, and this secondary radiation can be significant enough to cause concern. When considering surface operations, there is also a secondary radiation source from the surface of the planet, known as albedo, with neutrons being one of the most significant species. Given new vehicle designs for deep space and Mars missions, the secondary radiation environment and the implications of that environment is currently not well understood. Thus, several studies are necessary to fill the knowledge gaps of this secondary radiation environment. In this paper, we put forth the initial steps to increasing our understanding of neutron production from large vehicles by comparing the neutron production resulting from our radiation transport codes and providing a preliminary validation of our results against flight data. This paper will review the details of these results and discuss the finer points of the analysis.

A 10(9) neutrons/pulse transportable pulsed D-D neutron source based on flexible head plasma focus unit.

PubMed

Niranjan, Ram; Rout, R K; Srivastava, R; Kaushik, T C; Gupta, Satish C

2016-03-01

A 17 kJ transportable plasma focus (PF) device with flexible transmission lines is developed and is characterized. Six custom made capacitors are used for the capacitor bank (CB). The common high voltage plate of the CB is fixed to a centrally triggered spark gap switch. The output of the switch is coupled to the PF head through forty-eight 5 m long RG213 cables. The CB has a quarter time-period of 4 μs and an estimated current of 506 kA is delivered to the PF device at 17 kJ (60 μF, 24 kV) energy. The average neutron yield measured using silver activation detector in the radial direction is (7.1 ± 1.4) × 10(8) neutrons/shot over 4π sr at 5 mbar optimum D2 pressure. The average neutron yield is more in the axial direction with an anisotropy factor of 1.33 ± 0.18. The average neutron energies estimated in the axial as well as in the radial directions are (2.90 ± 0.20) MeV and (2.58 ± 0.20) MeV, respectively. The flexibility of the PF head makes it useful for many applications where the source orientation and the location are important factors. The influence of electromagnetic interferences from the CB as well as from the spark gap on applications area can be avoided by putting a suitable barrier between the bank and the PF head.

Adaptive Nodal Transport Methods for Reactor Transient Analysis

DOE Office of Scientific and Technical Information (OSTI.GOV)

Thomas Downar; E. Lewis

2005-08-31

Develop methods for adaptively treating the angular, spatial, and time dependence of the neutron flux in reactor transient analysis. These methods were demonstrated in the DOE transport nodal code VARIANT and the US NRC spatial kinetics code, PARCS.

Novel concept for neutron detection: proportional counter filled with 10B nanoparticle aerosol

NASA Astrophysics Data System (ADS)

Amaro, F. D.; Monteiro, C. M. B.; Dos Santos, J. M. F.; Antognini, A.

2017-02-01

The high neutron detection efficiency, good gamma-ray discrimination and non-toxicity of 3He made of proportional counters filled with this gas the obvious choice for neutron detection, particularly in radiation portal monitors (RPM), used to control the illicit transport of nuclear material, of which neutron detectors are key components. 3He is very rare and during the last decade this gas has become increasingly difficult to acquire. With the exception of BF3, which is toxic, no other gas can be used for neutron detection in proportional counters. We present an alternative where the 3He atoms are replaced by nanoparticles made of another neutron sensitive material, 10B. The particles are dispersed in a gaseous volume, forming an aerosol with neutron sensitive properties. A proportional counter filled with such aerosol was exposed to a thermal neutron beam and the recorded response indicates that the neutrons have interacted with the particles in the aerosol. This original technique, which transforms a standard proportional gas mixture into a neutron sensitive aerosol, is a breakthrough in the field of radiation detection and has the potential to become an alternative to the use of 3He in proportional counters.

Characterization of the high-energy neutron beam of the PRISMA beamline using a diamond detector

NASA Astrophysics Data System (ADS)

Cazzaniga, C.; Frost, C. D.; Minniti, T.; Schooneveld, E.; Perelli Cippo, E.; Tardocchi, M.; Rebai, M.; Gorini, G.

2016-07-01

The high-energy neutron component (En > 10 MeV) of the neutron spectrum of PRISMA, a beam-line at the ISIS spallation source, has been characterized for the first time. Neutron measurements using a Single-crystal Diamond Detector at a short-pulse source are obtained by a combination of pulse height and time of flight analysis. An XY scan provides a 2D map of the high-energy neutron beam which has a diameter of about 40 mm. The high neutron flux, that has been found to be (3.8 ± 0.7) · 105 cm-2s-1 for En > 10 MeV in the centre, opens up for a possible application of the beam-line as a high-energy neutron irradiation position. Results are of interest for the development of the ChipIR beam-line, which will feature an atmospheric-like neutron spectrum for chip irradiation experiment. Furthermore, these results demonstrate that diamond detectors can be used at spallation sources to investigate the transport of high-energy neutrons down instruments which is of interest in general to designers as high-energy neutrons are a source of background in thermal beamlines.

Li + transport in poly(ethylene oxide) based electrolyte: A combined study of neutron scattering, dielectric spectroscopy, and MD simulation

NASA Astrophysics Data System (ADS)

Do, Changwoo; Lunkenheimer, Peter; Diddens, Diddo; Götz, Marion; Weiß, Matthias; Loidl, Alois; Sun, Xiao-Guang; Allgaier, Jürgen; Ohl, Michael

2013-03-01

Dynamics of Li + transport in polyethylene oxide (PEO) and lithium bis(trifluoromethanesulfonyl)imde (LiTFSI) mixtures are investigated by combining various experimental techniques (neutron spin-echo and dielectric spectroscopy) with molecular dynamics (MD) simulations. Our results suggest that the characteristic live times within the cages formed by oxygens are mainly determined by the alpha-relaxation which corresponds to local segmental motions of polymers, to a much lesser extent by the main chain relaxation, and not at all by the beta-relaxation or any other faster processes. The significant contribution of Li + hopping process to the ion conductivity is also identified. Subsequently, detailed characteristic length and time scales of various Li + transport processes in solid polymer electrolytes are presented and interpreted.

Calculations vs. measurements of remnant dose rates for SNS spent structures

NASA Astrophysics Data System (ADS)

Popova, I. I.; Gallmeier, F. X.; Trotter, S.; Dayton, M.

2018-06-01

Residual dose rate measurements were conducted on target vessel #13 and proton beam window #5 after extraction from their service locations. These measurements were used to verify calculation methods of radionuclide inventory assessment that are typically performed for nuclear waste characterization and transportation of these structures. Neutronics analyses for predicting residual dose rates were carried out using the transport code MCNPX and the transmutation code CINDER90. For transport analyses complex and rigorous geometry model of the structures and their surrounding are applied. The neutronics analyses were carried out using Bertini and CEM high energy physics models for simulating particles interaction. Obtained preliminary calculational results were analysed and compared to the measured dose rates and overall are showing good agreement with in 40% in average.

Calculations vs. measurements of remnant dose rates for SNS spent structures

DOE Office of Scientific and Technical Information (OSTI.GOV)

Popova, Irina I.; Gallmeier, Franz X.; Trotter, Steven M.

Residual dose rate measurements were conducted on target vessel #13 and proton beam window #5 after extraction from their service locations. These measurements were used to verify calculation methods of radionuclide inventory assessment that are typically performed for nuclear waste characterization and transportation of these structures. Neutronics analyses for predicting residual dose rates were carried out using the transport code MCNPX and the transmutation code CINDER90. For transport analyses complex and rigorous geometry model of the structures and their surrounding are applied. The neutronics analyses were carried out using Bertini and CEM high energy physics models for simulating particles interaction.more » Obtained preliminary calculational results were analysed and compared to the measured dose rates and overall are showing good agreement with in 40% in average.« less

Out-of-field in vivo dosimetry using TLD in SABR for primary kidney cancer involving mixed photon fields.

PubMed

Lonski, P; Keehan, S; Siva, S; Pham, D; Franich, R D; Taylor, M L; Kron, T

2017-05-01

To assess out-of-field dose using three different variants of LiF thermoluminescence dosimeters (TLD) for ten patients who underwent stereotactic ablative body radiotherapy (SABR) for primary renal cell carcinoma (RCC) and compare with treatment planning system (TPS) dose calculations. Thermoluminescent dosimeter (TLD) measurements were conducted at 20, 30, 40 and 50cm from isocentre on ten patients undergoing SABR for primary RCC. Three types of high-sensitivity LiF:Mg,Cu,P TLD material with different 6 Li/ 7 Li isotope ratios were used. Patient plans were calculated using Eclipse Anisotropic Analytical Algorithm (AAA) for clinical evaluation and recalculated using Pencil Beam Convolution (PBC) algorithm for comparison. Both AAA and PBC showed diminished accuracy for photon doses at increasing distance out-of-field. At 50cm, measured photon dose was 0.3cGy normalised to a 10Gy prescription on average with only small variation across all patients. This is likely due to the leakage component of the out-of-field dose. The 6 Li-enriched TLD materials showed increased signal attributable to additional neutron contribution. LiF:Mg,Cu,P TLD containing 6 Li is sensitive enough to measure out-of-field dose 50cm from isocentre however will over-estimate the photon component of out-of-field dose in high energy treatments due to the presence of thermal neutrons. 7 Li enriched materials which are insensitive to neutrons are therefore required for accurate photon dosimetry. Neutron signal has been shown here to increase with MUs and is higher for patients treated using certain non coplanar beam arrangements. Further work is required to convert this additional neutron signal to dose. Copyright © 2017 Associazione Italiana di Fisica Medica. Published by Elsevier Ltd. All rights reserved.

Accretion-induced spin-wandering effects on the neutron star in Scorpius X-1: Implications for continuous gravitational wave searches

NASA Astrophysics Data System (ADS)

Mukherjee, Arunava; Messenger, Chris; Riles, Keith

2018-02-01

The LIGO's discovery of binary black hole mergers has opened up a new era of transient gravitational wave astronomy. The potential detection of gravitational radiation from another class of astronomical objects, rapidly spinning nonaxisymmetric neutron stars, would constitute a new area of gravitational wave astronomy. Scorpius X-1 (Sco X-1) is one of the most promising sources of continuous gravitational radiation to be detected with present-generation ground-based gravitational wave detectors, such as Advanced LIGO and Advanced Virgo. As the sensitivity of these detectors improve in the coming years, so will power of the search algorithms being used to find gravitational wave signals. Those searches will still require integration over nearly year long observational spans to detect the incredibly weak signals from rotating neutron stars. For low mass X-ray binaries such as Sco X-1 this difficult task is compounded by neutron star "spin wandering" caused by stochastic accretion fluctuations. In this paper, we analyze X-ray data from the R X T E satellite to infer the fluctuating torque on the neutron star in Sco X-1. We then perform a large-scale simulation to quantify the statistical properties of spin-wandering effects on the gravitational wave signal frequency and phase evolution. We find that there are a broad range of expected maximum levels of frequency wandering corresponding to maximum drifts of between 0.3 - 50 μ Hz /sec over a year at 99% confidence. These results can be cast in terms of the maximum allowed length of a coherent signal model neglecting spin-wandering effects as ranging between 5-80 days. This study is designed to guide the development and evaluation of Sco X-1 search algorithms.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Tuve, C.; Albergo, S.; Boemi, D.

In the frame of the Transport Collaboration neutrons in coincidence with charged fragments produced in the {sup 40}Ca + H reaction at E{sub lab} = 357 and 565 AMeV have been measured at the Heavy Ion Spectrometer System (HISS) facility of the Lawrence Berkeley National Laboratory, using the multifunctional neutron spectrometer MUFFINS. The detector covered a narrow angular range about the beam in the forward direction (0? - 3.2?). In this contribution we report absolute neutron production cross sections in coincidence with charged fragments (10 {<=} Z {<=} 20). The neutron multiplicities have been estimated from the comparison between themore » neutron cross sections, in coincidence with the fragments, and the elemental cross sections. We have found evidence for a pre-equilibrium emission of prompt neutrons in superposition to a 'slower' deexcitation of the equilibrated remnant by emission of nucleons and fragments, as already seen in the inclusive rapidity distributions.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dickens, Peter T.; Marcial, Jose; McCloy, John

In this study, LiAlO 2 crystals doped with rare-earth elements and Ti were produced by the CZ method and spectroscopic and neutron detection properties were investigated. Photoluminescence revealed no clear luminescent activation of LiAlO 2 by the rare-earth dopants though some interesting luminescence was observed from secondary phases within the crystal. Gamma-ray pulse height spectra collected using a 137Cs source exhibited only a Compton edge for the crystals. Neutron modeling using Monte Carlo N-Particle Transport Code revealed most neutrons used in the detection setup are thermalized, and while using natural lithium in the crystal growth, which contains 7.6% 6Li, amore » 10 mm Ø by 10 mm sample of LiAlO 2 has a 70.7% intrinsic thermal neutron capture efficiency. Furthermore, the pulse height spectra collected using a 241Am-Be neutron source demonstrated a distinct neutron peak.« less

FY16 Status Report on NEAMS Neutronics Activities

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lee, C. H.; Shemon, E. R.; Smith, M. A.

2016-09-30

The goal of the NEAMS neutronics effort is to develop a neutronics toolkit for use on sodium-cooled fast reactors (SFRs) which can be extended to other reactor types. The neutronics toolkit includes the high-fidelity deterministic neutron transport code PROTEUS and many supporting tools such as a cross section generation code MC 2-3, a cross section library generation code, alternative cross section generation tools, mesh generation and conversion utilities, and an automated regression test tool. The FY16 effort for NEAMS neutronics focused on supporting the release of the SHARP toolkit and existing and new users, continuing to develop PROTEUS functions necessarymore » for performance improvement as well as the SHARP release, verifying PROTEUS against available existing benchmark problems, and developing new benchmark problems as needed. The FY16 research effort was focused on further updates of PROTEUS-SN and PROTEUS-MOCEX and cross section generation capabilities as needed.« less

The exact solution of the monoenergetic transport equation for critical cylinders

NASA Technical Reports Server (NTRS)

Westfall, R. M.; Metcalf, D. R.

1972-01-01

An analytic solution for the critical, monoenergetic, bare, infinite cylinder is presented. The solution is obtained by modifying a previous development based on a neutron density transform and Case's singular eigenfunction method. Numerical results for critical radii and the neutron density as a function of position are included and compared with the results of other methods.

Neutron skyshine calculations with the integral line-beam method

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gui, A.A.; Shultis, J.K.; Faw, R.E.

1997-10-01

Recently developed line- and conical-beam response functions are used to calculate neutron skyshine doses for four idealized source geometries. These calculations, which can serve as benchmarks, are compared with MCNP calculations, and the excellent agreement indicates that the integral conical- and line-beam method is an effective alternative to more computationally expensive transport calculations.

Advances in neutron based bulk explosive detection

NASA Astrophysics Data System (ADS)

Gozani, Tsahi; Strellis, Dan

2007-08-01

Neutron based explosive inspection systems can detect a wide variety of national security threats. The inspection is founded on the detection of characteristic gamma rays emitted as the result of neutron interactions with materials. Generally these are gamma rays resulting from thermal neutron capture and inelastic scattering reactions in most materials and fast and thermal neutron fission in fissile (e.g.235U and 239Pu) and fertile (e.g.238U) materials. Cars or trucks laden with explosives, drugs, chemical agents and hazardous materials can be detected. Cargo material classification via its main elements and nuclear materials detection can also be accomplished with such neutron based platforms, when appropriate neutron sources, gamma ray spectroscopy, neutron detectors and suitable decision algorithms are employed. Neutron based techniques can be used in a variety of scenarios and operational modes. They can be used as stand alones for complete scan of objects such as vehicles, or for spot-checks to clear (or validate) alarms indicated by another inspection system such as X-ray radiography. The technologies developed over the last two decades are now being implemented with good results. Further advances have been made over the last few years that increase the sensitivity, applicability and robustness of these systems. The advances range from the synchronous inspection of two sides of vehicles, increasing throughput and sensitivity and reducing imparted dose to the inspected object and its occupants (if any), to taking advantage of the neutron kinetic behavior of cargo to remove systematic errors, reducing background effects and improving fast neutron signals.

Neutron production in coincidence with fragments from the 4Ca+H reactions at Elab=357 and 565 A MeV

NASA Astrophysics Data System (ADS)

Tuvà, C.; Albergo, S.; Boemi, D.; Caccia, Z.; Chen, C.-X.; Costa, S.; Crawford, H. J.; Cronqvist, M.; Engelage, J.; Greiner, L.; Guzik, T. G.; Insolia, A.; Knott, C. N.; Lindstrom, P. J.; Mitchell, J. W.; Potenza, R.; Russo, G. V.; Soutoul, A.; Testard, O.; Tricomi, A.; Tull, C. E.; Waddington, C. J.; Webber, W. R.; Wefel, J. P.

2000-04-01

In the frame of the Transport Collaboration neutrons in coincidence with charged fragments produced in the 40Ca+H reaction at Elab=357 and 565 A MeV have been measured at the Heavy Ion Spectrometer System (HISS) facility of the Lawrence Berkeley National Laboratory, using the multifunctional neutron spectrometer MUFFINS. The detector covered a narrow angular range about the beam in the forward direction (0°-3.2°). In this contribution we report absolute neutron production cross sections in coincidence with charged fragments (10⩽Z⩽20). The neutron multiplicities have been estimated from the comparison between the neutron cross sections, in coincidence with the fragments, and the elemental cross sections. We have found evidence for a pre-equilibrium emission of prompt neutrons in superposition to a `slower' deexcitation of the equilibrated remnant by emission of nucleons and fragments, as already seen in the inclusive rapidity distributions.

Spectroscopic and neutron detection properties of rare earth and titanium doped LiAlO 2 single crystals

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dickens, Peter T.; Marcial, José; McCloy, John

In this study, LiAlO2 crystals doped with rare-earth elements and Ti were produced by the CZ method and spectroscopic and neutron detection properties were investigated. Photoluminescence revealed no clear luminescent activation of LiAlO2 by the rare-earth dopants though some interesting luminescence was observed from secondary phases within the crystal. Gamma-ray pulse height spectra collected using a 137Cs source exhibited only a Compton edge for the crystals. Neutron modeling using Monte Carlo N-Particle Transport Code revealed most neutrons used in the detection setup are thermalized, and while using natural lithium in the crystal growth, which contains 7.6 % 6Li, a 10more » mm Ø by 10 mm sample of LiAlO2 has a 70.7 % intrinsic thermal neutron capture efficiency. Furthermore, the pulse height spectra collected using a 241Am-Be neutron source demonstrated a distinct neutron peak.« less

Spectroscopic and neutron detection properties of rare earth and titanium doped LiAlO 2 single crystals

DOE PAGES

Dickens, Peter T.; Marcial, Jose; McCloy, John; ...

2017-05-17

In this study, LiAlO 2 crystals doped with rare-earth elements and Ti were produced by the CZ method and spectroscopic and neutron detection properties were investigated. Photoluminescence revealed no clear luminescent activation of LiAlO 2 by the rare-earth dopants though some interesting luminescence was observed from secondary phases within the crystal. Gamma-ray pulse height spectra collected using a 137Cs source exhibited only a Compton edge for the crystals. Neutron modeling using Monte Carlo N-Particle Transport Code revealed most neutrons used in the detection setup are thermalized, and while using natural lithium in the crystal growth, which contains 7.6% 6Li, amore » 10 mm Ø by 10 mm sample of LiAlO 2 has a 70.7% intrinsic thermal neutron capture efficiency. Furthermore, the pulse height spectra collected using a 241Am-Be neutron source demonstrated a distinct neutron peak.« less

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

DOE PAGES

Ibrahim, Ahmad M.; Wilson, Paul P.H.; Sawan, Mohamed E.; ...

2015-06-30

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 muchmore » 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.« less

Numerical relativity simulations of neutron star merger remnants using conservative mesh refinement

NASA Astrophysics Data System (ADS)

Dietrich, Tim; Bernuzzi, Sebastiano; Ujevic, Maximiliano; Brügmann, Bernd

2015-06-01

We study equal- and unequal-mass neutron star mergers by means of new numerical relativity simulations in which the general relativistic hydrodynamics solver employs an algorithm that guarantees mass conservation across the refinement levels of the computational mesh. We consider eight binary configurations with total mass M =2.7 M⊙, mass ratios q =1 and q =1.16 , four different equations of state (EOSs) and one configuration with a stiff EOS, M =2.5 M⊙ and q =1.5 , which is one of the largest mass ratios simulated in numerical relativity to date. We focus on the postmerger dynamics and study the merger remnant, the dynamical ejecta, and the postmerger gravitational wave spectrum. Although most of the merger remnants are a hypermassive neutron star collapsing to a black hole+disk system on dynamical time scales, stiff EOSs can eventually produce a stable massive neutron star. During the merger process and on very short time scales, about ˜10-3- 10-2M⊙ of material become unbound with kinetic energies ˜1050 erg . Ejecta are mostly emitted around the orbital plane and favored by large mass ratios and softer EOS. The postmerger wave spectrum is mainly characterized by the nonaxisymmetric oscillations of the remnant neutron star. The stiff EOS configuration consisting of a 1.5 M⊙ and a 1.0 M⊙ neutron star, simulated here for the first time, shows a rather peculiar dynamics. During merger the companion star is very deformed; about ˜0.03 M⊙ of the rest mass becomes unbound from the tidal tail due to the torque generated by the two-core inner structure. The merger remnant is a stable neutron star surrounded by a massive accretion disk of rest mass ˜0.3 M⊙. This and similar configurations might be particularly interesting for electromagnetic counterparts. Comparing results obtained with and without the conservative mesh refinement algorithm, we find that postmerger simulations can be affected by systematic errors if mass conservation is not enforced in the mesh refinement strategy. However, mass conservation also depends on grid details and on the artificial atmosphere setup; the latter are particularly significant in the computation of the dynamical ejecta.

LDRD Final Review: Radiation Transport Calculations

DOE Office of Scientific and Technical Information (OSTI.GOV)

Goorley, John Timothy; Morgan, George Lake; Lestone, John Paul

2017-06-22

Both high-fidelity & toy simulations are being used to understand measured signals and improve the Area 11 NDSE diagnostic. We continue to gain more and more confidence in the ability for MCNP to simulate neutron and photon transport from source to radiation detector.

Accelerator-Based Biological Irradiation Facility Simulating Neutron Exposure from an Improvised Nuclear Device

PubMed Central

Xu, Yanping; Randers-Pehrson, Gerhard; Turner, Helen C.; Marino, Stephen A.; Geard, Charles R.; Brenner, David J.; Garty, Guy

2015-01-01

We describe here an accelerator-based neutron irradiation facility, intended to expose blood or small animals to neutron fields mimicking those from an improvised nuclear device at relevant distances from the epicenter. Neutrons are generated by a mixed proton/deuteron beam on a thick beryllium target, generating a broad spectrum of neutron energies that match those estimated for the Hiroshima bomb at 1.5 km from ground zero. This spectrum, dominated by neutron energies between 0.2 and 9 MeV, is significantly different from the standard reactor fission spectrum, as the initial bomb spectrum changes when the neutrons are transported through air. The neutron and gamma dose rates were measured using a custom tissue-equivalent gas ionization chamber and a compensated Geiger-Mueller dosimeter, respectively. Neutron spectra were evaluated by unfolding measurements using a proton-recoil proportional counter and a liquid scintillator detector. As an illustration of the potential use of this facility we present micronucleus yields in single divided, cytokinesis-blocked human peripheral lymphocytes up to 1.5 Gy demonstrating 3- to 5-fold enhancement over equivalent X-ray doses. This facility is currently in routine use, irradiating both mice and human blood samples for evaluation of neutron-specific biodosimetry assays. Future studies will focus on dose reconstruction in realistic mixed neutron/photon fields. PMID:26414507

Accelerator-Based Biological Irradiation Facility Simulating Neutron Exposure from an Improvised Nuclear Device.

PubMed

Xu, Yanping; Randers-Pehrson, Gerhard; Turner, Helen C; Marino, Stephen A; Geard, Charles R; Brenner, David J; Garty, Guy

2015-10-01

We describe here an accelerator-based neutron irradiation facility, intended to expose blood or small animals to neutron fields mimicking those from an improvised nuclear device at relevant distances from the epicenter. Neutrons are generated by a mixed proton/deuteron beam on a thick beryllium target, generating a broad spectrum of neutron energies that match those estimated for the Hiroshima bomb at 1.5 km from ground zero. This spectrum, dominated by neutron energies between 0.2 and 9 MeV, is significantly different from the standard reactor fission spectrum, as the initial bomb spectrum changes when the neutrons are transported through air. The neutron and gamma dose rates were measured using a custom tissue-equivalent gas ionization chamber and a compensated Geiger-Mueller dosimeter, respectively. Neutron spectra were evaluated by unfolding measurements using a proton-recoil proportional counter and a liquid scintillator detector. As an illustration of the potential use of this facility we present micronucleus yields in single divided, cytokinesis-blocked human peripheral lymphocytes up to 1.5 Gy demonstrating 3- to 5-fold enhancement over equivalent X-ray doses. This facility is currently in routine use, irradiating both mice and human blood samples for evaluation of neutron-specific biodosimetry assays. Future studies will focus on dose reconstruction in realistic mixed neutron/photon fields.

Analysis of neutron and gamma-ray streaming along the maze of NRCAM thallium production target room.

PubMed

Raisali, G; Hajiloo, N; Hamidi, S; Aslani, G

2006-08-01

Study of the shield performance of a thallium-203 production target room has been investigated in this work. Neutron and gamma-ray equivalent dose rates at various points of the maze are calculated by simulating the transport of streaming neutrons, and photons using Monte Carlo method. For determination of neutron and gamma-ray source intensities and their energy spectrum, we have applied SRIM 2003 and ALICE91 computer codes to Tl target and its Cu substrate for a 145 microA of 28.5 MeV protons beam. The MCNP/4C code has been applied with neutron source term in mode n p to consider both prompt neutrons and secondary gamma-rays. Then the code is applied for the prompt gamma-rays as the source term. The neutron-flux energy spectrum and equivalent dose rates for neutron and gamma-rays in various positions in the maze have been calculated. It has been found that the deviation between calculated and measured dose values along the maze is less than 20%.

The EURITRACK project: development of a tagged neutron inspection system for cargo containers

NASA Astrophysics Data System (ADS)

Perot, Bertrand; Perret, Gregory; Mariani, Alain; Ma, Jean-Luc; Szabo, Jean-Louis; Mercier, Emmanuel; Sannie, Guillaume; Viesti, Giuseppe; Nebbia, Giancarlo; Pesente, Silvia; Lunardon, Marcello; Formisano, Paola; Moretto, Sandra; Fabris, Daniela; Zenoni, Aldo; Bonomi, Germano; Donzella, Antonietta; Fontana, Andrea; Boghen, Gaia; Valkovic, Vladivoj; Sudac, Darovin; Moszynski, Marek; Batsch, Tadeusz; Gierlik, Michal; Wolski, Dariusz; Klamra, Wlodzimierz; Isaksson, Patrick; Le Tourneur, Philippe; Lhuissier, Miguel; Colonna, Annamaria; Tintori, Carlo; Peerani, Paolo; Sequeira, Vitor; Salvato, Martino

2006-05-01

The EURopean Illicit TRAfficing Countermeasures Kit project is part of the 6th European Union Framework Program, and aims at developing a neutron inspection system for detecting threat materials (explosives, drugs, etc.) in cargo containers. Neutron interaction in the container produces specific gamma-rays used to determine the chemical composition of the inspected material. An associated particle sealed tube neutron generator is developed to allow precise location of the interaction point by direction and time-of-flight measurements of the neutrons tagged by alpha-particles. The EURITRACK project consists in developing: a transportable deuterium-tritium neutron generator including a position sensitive alpha detector (8×8 matrix of YAP:Ce crystals coupled to a multi-anode photomultiplier), fast neutron and gamma-ray detectors, front-end electronics to perform coincidence and spectroscopic measurements, and an integrated software which manages neutron generator and detectors positioning, data acquisition and analysis. Hardware components have been developed and tested by the consortium partners. Current status of this work and provisional performances of the system assessed by Monte Carlo calculations are presented.

Experimental determination of spin-dependent electron density by joint refinement of X-ray and polarized neutron diffraction data.

PubMed

Deutsch, Maxime; Claiser, Nicolas; Pillet, Sébastien; Chumakov, Yurii; Becker, Pierre; Gillet, Jean Michel; Gillon, Béatrice; Lecomte, Claude; Souhassou, Mohamed

2012-11-01

New crystallographic tools were developed to access a more precise description of the spin-dependent electron density of magnetic crystals. The method combines experimental information coming from high-resolution X-ray diffraction (XRD) and polarized neutron diffraction (PND) in a unified model. A new algorithm that allows for a simultaneous refinement of the charge- and spin-density parameters against XRD and PND data is described. The resulting software MOLLYNX is based on the well known Hansen-Coppens multipolar model, and makes it possible to differentiate the electron spins. This algorithm is validated and demonstrated with a molecular crystal formed by a bimetallic chain, MnCu(pba)(H(2)O)(3)·2H(2)O, for which XRD and PND data are available. The joint refinement provides a more detailed description of the spin density than the refinement from PND data alone.

Determination of Diffusion Parameters of Mean Moderation by Means of a Pulsed Neutron Source. I. Dowtherm A at 20 C; DETERMINAZIONE DEI PARAMETRI DI DIFFUSIONE DEI MEZZI MODERANTI CONIL METODO DELLA SORGENTE DI NEUTRONI PULSATA. I.DOWTHERM A (TEMPERATURE 20 C)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Demanins, F.; Rado, V.; Vinci, F.

1963-04-01

The macroscopic absorption cross section, diffusion constant, diffusion cooling constant, transport mean free patu, extrapolated distance, diffusion length, and mean life for thermal neutrons were determined for Dowtherm A at 20 deg C, using a pulsed neutron source. The experimental assembly and data analysis method are described, and the results are compared with other determinations. (auth)

Application of nuclear analytical techniques using long-life sealed-tube neutron generators.

PubMed

Bach, P; Cluzeau, S; Lambermont, C

1994-01-01

The new range of sealed-tube neutron generators developed by SODERN appears to be appropriate for the industrial environment. The main characteristics are the high emission stability during the very long lifetime of the tube, flexible pulsed mode capability, safety in operation with no radiation in "off" state, and the easy transportation of equipment. Some applications of the neutron generators, called GENIE, are considered: high-sensitivity measurement of transuranic elements in nuclear waste drums, bulk material analysis for process control, and determination of the airborne pollutants for environmental monitoring.

Depth dependency of neutron density produced by cosmic rays in the lunar subsurface

NASA Astrophysics Data System (ADS)

Ota, S.; Sihver, L.; Kobayashi, S.; Hasebe, N.

2014-11-01

Depth dependency of neutrons produced by cosmic rays (CRs) in the lunar subsurface was estimated using the three-dimensional Monte Carlo particle and heavy ion transport simulation code, PHITS, incorporating the latest high energy nuclear data, JENDL/HE-2007. The PHITS simulations of equilibrium neutron density profiles in the lunar subsurface were compared with the measurement by Apollo 17 Lunar Neutron Probe Experiment (LNPE). Our calculations reproduced the LNPE data except for the 350-400 mg/cm2 region under the improved condition using the CR spectra model based on the latest observations, well-tested nuclear interaction models with systematic cross section data, and JENDL/HE-2007.

New techniques in neutron data measurements above 30 MeV

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lisowski, P.W.; Haight, R.C.

1991-01-01

Recent developments in experimental facilities have enabled new techniques for measurements of neutron interactions above 30 MeV. Foremost is the development of both monoenergetic and continuous neutron sources using accelerators in the medium energy region between 100 and 800 MeV. Measurements of the reaction products have been advanced by the continuous improvement in detector systems, electronics and computers. Corresponding developments in particle transport codes and in the theory of nuclear reactions at these energies have allowed more precise design of neutron sources, experimental shielding and detector response. As a result of these improvements, many new measurements are possible and themore » data base in this energy range is expanding quickly.« less

Progress toward the development and testing of source reconstruction methods for NIF neutron imaging.

PubMed

Loomis, E N; Grim, G P; Wilde, C; Wilson, D C; Morgan, G; Wilke, M; Tregillis, I; Merrill, F; Clark, D; Finch, J; Fittinghoff, D; Bower, D

2010-10-01

Development of analysis techniques for neutron imaging at the National Ignition Facility is an important and difficult task for the detailed understanding of high-neutron yield inertial confinement fusion implosions. Once developed, these methods must provide accurate images of the hot and cold fuels so that information about the implosion, such as symmetry and areal density, can be extracted. One method under development involves the numerical inversion of the pinhole image using knowledge of neutron transport through the pinhole aperture from Monte Carlo simulations. In this article we present results of source reconstructions based on simulated images that test the methods effectiveness with regard to pinhole misalignment.

Simulations of GCR interactions within planetary bodies using GEANT4

NASA Astrophysics Data System (ADS)

Mesick, K.; Feldman, W. C.; Stonehill, L. C.; Coupland, D. D. S.

2017-12-01

On planetary bodies with little to no atmosphere, Galactic Cosmic Rays (GCRs) can hit the body and produce neutrons primarily through nuclear spallation within the top few meters of the surfaces. These neutrons undergo further nuclear interactions with elements near the planetary surface and some will escape the surface and can be detected by landed or orbiting neutron radiation detector instruments. The neutron leakage signal at fast neutron energies provides a measure of average atomic mass of the near-surface material and in the epithermal and thermal energy ranges is highly sensitive to the presence of hydrogen. Gamma-rays can also escape the surface, produced at characteristic energies depending on surface composition, and can be detected by gamma-ray instruments. The intra-nuclear cascade (INC) that occurs when high-energy GCRs interact with elements within a planetary surface to produce the leakage neutron and gamma-ray signals is highly complex, and therefore Monte Carlo based radiation transport simulations are commonly used for predicting and interpreting measurements from planetary neutron and gamma-ray spectroscopy instruments. In the past, the simulation code that has been widely used for this type of analysis is MCNPX [1], which was benchmarked against data from the Lunar Neutron Probe Experiment (LPNE) on Apollo 17 [2]. In this work, we consider the validity of the radiation transport code GEANT4 [3], another widely used but open-source code, by benchmarking simulated predictions of the LPNE experiment to the Apollo 17 data. We consider the impact of different physics model options on the results, and show which models best describe the INC based on agreement with the Apollo 17 data. The success of this validation then gives us confidence in using GEANT4 to simulate GCR-induced neutron leakage signals on Mars in relevance to a re-analysis of Mars Odyssey Neutron Spectrometer data. References [1] D.B. Pelowitz, Los Alamos National Laboratory, LA-CP-05-0369, 2005. [2] G.W. McKinney et al, Journal of Geophysics Research, 111, E06004, 2006. [3] S. Agostinelli et al, Nuclear Instrumentation and Methods A, 506, 2003.

Hydrodynamical simulation of detonations in superbursts. I. The hydrodynamical algorithm and some preliminary one-dimensional results

NASA Astrophysics Data System (ADS)

Noël, C.; Busegnies, Y.; Papalexandris, M. V.; Deledicque, V.; El Messoudi, A.

2007-08-01

Aims:This work presents a new hydrodynamical algorithm to study astrophysical detonations. A prime motivation of this development is the description of a carbon detonation in conditions relevant to superbursts, which are thought to result from the propagation of a detonation front around the surface of a neutron star in the carbon layer underlying the atmosphere. Methods: The algorithm we have developed is a finite-volume method inspired by the original MUSCL scheme of van Leer (1979). The algorithm is of second-order in the smooth part of the flow and avoids dimensional splitting. It is applied to some test cases, and the time-dependent results are compared to the corresponding steady state solution. Results: Our algorithm proves to be robust to test cases, and is considered to be reliably applicable to astrophysical detonations. The preliminary one-dimensional calculations we have performed demonstrate that the carbon detonation at the surface of a neutron star is a multiscale phenomenon. The length scale of liberation of energy is 106 times smaller than the total reaction length. We show that a multi-resolution approach can be used to solve all the reaction lengths. This result will be very useful in future multi-dimensional simulations. We present also thermodynamical and composition profiles after the passage of a detonation in a pure carbon or mixed carbon-iron layer, in thermodynamical conditions relevant to superbursts in pure helium accretor systems.

Simple algorithms for digital pulse-shape discrimination with liquid scintillation detectors

NASA Astrophysics Data System (ADS)

Alharbi, T.

2015-01-01

The development of compact, battery-powered digital liquid scintillation neutron detection systems for field applications requires digital pulse processing (DPP) algorithms with minimum computational overhead. To meet this demand, two DPP algorithms for the discrimination of neutron and γ-rays with liquid scintillation detectors were developed and examined by using a NE213 liquid scintillation detector in a mixed radiation field. The first algorithm is based on the relation between the amplitude of a current pulse at the output of a photomultiplier tube and the amount of charge contained in the pulse. A figure-of-merit (FOM) value of 0.98 with 450 keVee (electron equivalent energy) energy threshold was achieved with this method when pulses were sampled at 250 MSample/s and with 8-bit resolution. Compared to the similar method of charge-comparison this method requires only a single integration window, thereby reducing the amount of computations by approximately 40%. The second approach is a digital version of the trailing-edge constant-fraction discrimination method. A FOM value of 0.84 with an energy threshold of 450 keVee was achieved with this method. In comparison with the similar method of rise-time discrimination this method requires a single time pick-off, thereby reducing the amount of computations by approximately 50%. The algorithms described in this work are useful for developing portable detection systems for applications such as homeland security, radiation dosimetry and environmental monitoring.

A real-time n/γ digital pulse shape discriminator based on FPGA.

PubMed

Li, Shiping; Xu, Xiufeng; Cao, Hongrui; Yuan, Guoliang; Yang, Qingwei; Yin, Zejie

2013-02-01

A FPGA-based real-time digital pulse shape discriminator has been employed to distinguish between neutrons (n) and gammas (γ) in the Neutron Flux Monitor (NFM) for International Thermonuclear Experimental Reactor (ITER). The discriminator takes advantages of the Field Programmable Gate Array (FPGA) parallel and pipeline process capabilities to carry out the real-time sifting of neutrons in n/γ mixed radiation fields, and uses the rise time and amplitude inspection techniques simultaneously as the discrimination algorithm to observe good n/γ separation. Some experimental results have been presented which show that this discriminator can realize the anticipated goals of NFM perfectly with its excellent discrimination quality and zero dead time. Copyright © 2012 Elsevier Ltd. All rights reserved.

Spent nuclear fuel assembly inspection using neutron computed tomography

NASA Astrophysics Data System (ADS)

Pope, Chad Lee

The research presented here focuses on spent nuclear fuel assembly inspection using neutron computed tomography. Experimental measurements involving neutron beam transmission through a spent nuclear fuel assembly serve as benchmark measurements for an MCNP simulation model. Comparison of measured results to simulation results shows good agreement. Generation of tomography images from MCNP tally results was accomplished using adapted versions of built in MATLAB algorithms. Multiple fuel assembly models were examined to provide a broad set of conclusions. Tomography images revealing assembly geometric information including the fuel element lattice structure and missing elements can be obtained using high energy neutrons. A projection difference technique was developed which reveals the substitution of unirradiated fuel elements for irradiated fuel elements, using high energy neutrons. More subtle material differences such as altering the burnup of individual elements can be identified with lower energy neutrons provided the scattered neutron contribution to the image is limited. The research results show that neutron computed tomography can be used to inspect spent nuclear fuel assemblies for the purpose of identifying anomalies such as missing elements or substituted elements. The ability to identify anomalies in spent fuel assemblies can be used to deter diversion of material by increasing the risk of early detection as well as improve reprocessing facility operations by confirming the spent fuel configuration is as expected or allowing segregation if anomalies are detected.

FASTER 3: A generalized-geometry Monte Carlo computer program for the transport of neutrons and gamma rays. Volume 2: Users manual

NASA Technical Reports Server (NTRS)

Jordan, T. M.

1970-01-01

A description of the FASTER-III program for Monte Carlo Carlo calculation of photon and neutron transport in complex geometries is presented. Major revisions include the capability of calculating minimum weight shield configurations for primary and secondary radiation and optimal importance sampling parameters. The program description includes a users manual describing the preparation of input data cards, the printout from a sample problem including the data card images, definitions of Fortran variables, the program logic, and the control cards required to run on the IBM 7094, IBM 360, UNIVAC 1108 and CDC 6600 computers.

Neutron powder diffraction study on the iron-based nitride superconductor ThFeAsN

NASA Astrophysics Data System (ADS)

Mao, Huican; Wang, Cao; Maynard-Casely, Helen E.; Huang, Qingzhen; Wang, Zhicheng; Cao, Guanghan; Li, Shiliang; Luo, Huiqian

2017-03-01

We report neutron diffraction and transport results on the newly discovered superconducting nitride ThFeAsN with T_c= 30 \\text{K} . No magnetic transition, but a weak structural distortion around 160 K, is observed by cooling from 300 K to 6 K. Analysis on the resistivity, Hall transport and crystal structure suggests that this material behaves as an electron optimally doped pnictide superconductor due to extra electrons from nitrogen deficiency or oxygen occupancy at the nitrogen site, which, together with the low arsenic height, may enhance the electron itinerancy and reduce the electron correlations, thus suppressing the static magnetic order.

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

White, Morgan C.

2000-07-01

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

Quantitative non-destructive assay of PuBe neutron sources

NASA Astrophysics Data System (ADS)

Lakosi, László; Bagi, János; Nguyen, Cong Tam

2006-02-01

PuBe neutron sources were assayed, using a combination of high resolution γ-spectrometry (HRGS) and neutron correlation technique. In a previous publication [J. Bagi, C. Tam Nguyen, L. Lakosi, Nucl. Instr. and Meth. B 222 (2004) 242] a passive neutron well-counter was reported with 3He tubes embedded in a polyamide (TERRAMID) moderator (lined inside with Cd) surrounding the sources to be measured. Gross and coincidence neutron counting was performed, and the Pu content of the sources was found out from isotope analysis and by adopting specific (α, n) reaction yields of the Pu isotopes and 241Am in Be, based on supplier's information and literature data. The method was further developed and refined. Evaluation algorithm was more precisely worked out. The contribution of secondary (correlated) neutrons to the total neutron output was derived from the coincidence (doubles) count rate and taken into account in assessing the Pu content. A new evaluation of former results was performed. Assay was extended to other PuBe sources, and new results were added. In order to attain higher detection efficiency, a more efficient moderator was also applied, with and without Cd shielding around the assay chamber. Calibration seems possible using neutron measurements only (without γ-spectrometry), based on a correlation between the Pu amount and the coincidence-to-total ratio. It is expected that the method could be used for Pu accountancy and safeguards verification as well as identification and assay of seized, found, or not documented PuBe neutron sources.

MCViNE- An object oriented Monte Carlo neutron ray tracing simulation package

DOE PAGES

Lin, J. Y. Y.; Smith, Hillary L.; Granroth, Garrett E.; ...

2015-11-28

MCViNE (Monte-Carlo VIrtual Neutron Experiment) is an open-source Monte Carlo (MC) neutron ray-tracing software for performing computer modeling and simulations that mirror real neutron scattering experiments. We exploited the close similarity between how instrument components are designed and operated and how such components can be modeled in software. For example we used object oriented programming concepts for representing neutron scatterers and detector systems, and recursive algorithms for implementing multiple scattering. Combining these features together in MCViNE allows one to handle sophisticated neutron scattering problems in modern instruments, including, for example, neutron detection by complex detector systems, and single and multiplemore » scattering events in a variety of samples and sample environments. In addition, MCViNE can use simulation components from linear-chain-based MC ray tracing packages which facilitates porting instrument models from those codes. Furthermore it allows for components written solely in Python, which expedites prototyping of new components. These developments have enabled detailed simulations of neutron scattering experiments, with non-trivial samples, for time-of-flight inelastic instruments at the Spallation Neutron Source. Examples of such simulations for powder and single-crystal samples with various scattering kernels, including kernels for phonon and magnon scattering, are presented. As a result, with simulations that closely reproduce experimental results, scattering mechanisms can be turned on and off to determine how they contribute to the measured scattering intensities, improving our understanding of the underlying physics.« less

Light transport feature for SCINFUL.

PubMed

Etaati, G R; Ghal-Eh, N

2008-03-01

An extended version of the scintillator response function prediction code SCINFUL has been developed by incorporating PHOTRACK, a Monte Carlo light transport code. Comparisons of calculated and experimental results for organic scintillators exposed to neutrons show that the extended code improves the predictive capability of SCINFUL.

Indoor Fast Neutron Generator for Biophysical and Electronic Applications

NASA Astrophysics Data System (ADS)

Cannuli, A.; Caccamo, M. T.; Marchese, N.; Tomarchio, E. A.; Pace, C.; Magazù, S.

2018-05-01

This study focuses the attention on an indoor fast neutron generator for biophysical and electronic applications. More specifically, the findings obtained by several simulations with the MCNP Monte Carlo code, necessary for the realization of a shield for indoor measurements, are presented. Furthermore, an evaluation of the neutron spectrum modification caused by the shielding is reported. Fast neutron generators are a valid and interesting available source of neutrons, increasingly employed in a wide range of research fields, such as science and engineering. The employed portable pulsed neutron source is a MP320 Thermo Scientific neutron generator, able to generate 2.5 MeV neutrons with a neutron yield of 2.0 x 106 n/s, a pulse rate of 250 Hz to 20 KHz and a duty factor varying from 5% to 100%. The neutron generator, based on Deuterium-Deuterium nuclear fusion reactions, is employed in conjunction with a solid-state photon detector, made of n-type high-purity germanium (PINS-GMX by ORTEC) and it is mainly addressed to biophysical and electronic studies. The present study showed a proposal for the realization of a shield necessary for indoor applications for MP320 neutron generator, with a particular analysis of the transport of neutrons simulated with Monte Carlo code and described the two main lines of research in which the source will be used.

The effects of resonant magnetic perturbations on fast ion confinement in the Mega Amp Spherical Tokamak

NASA Astrophysics Data System (ADS)

McClements, K. G.; Akers, R. J.; Boeglin, W. U.; Cecconello, M.; Keeling, D.; Jones, O. M.; Kirk, A.; Klimek, I.; Perez, R. V.; Shinohara, K.; Tani, K.

2015-07-01

The effects of resonant magnetic perturbations (RMPs) on the confinement of energetic (neutral beam) ions in the Mega Amp Spherical Tokamak (MAST) are assessed experimentally using measurements of neutrons, fusion protons and fast ion Dα (FIDA) light emission. In single null-diverted (SND) MAST pulses with relatively low plasma current (400 kA), the total neutron emission dropped by approximately a factor of two when RMPs with toroidal mode number n = 3 were applied. The measured neutron rate during RMPs was much lower than that calculated using the TRANSP plasma simulation code, even when non-classical (but axisymmetric) ad hoc fast ion transport was taken into account in the latter. Sharp drops in spatially-resolved neutron rates, fusion proton rates and FIDA emission were also observed. First principles-based simulations of RMP-induced fast ion transport in MAST, using the F3D-OFMC code, show similar losses for two alternative representations of the MAST first wall, with and without full orbit effects taken into account; for n = 6 RMPs in a 600 kA plasma, the additional loss of beam power due to the RMPs was found in the simulations to be approximately 11%.

CAD-Based Shielding Analysis for ITER Port Diagnostics

NASA Astrophysics Data System (ADS)

Serikov, Arkady; Fischer, Ulrich; Anthoine, David; Bertalot, Luciano; De Bock, Maartin; O'Connor, Richard; Juarez, Rafael; Krasilnikov, Vitaly

2017-09-01

Radiation shielding analysis conducted in support of design development of the contemporary diagnostic systems integrated inside the ITER ports is relied on the use of CAD models. This paper presents the CAD-based MCNP Monte Carlo radiation transport and activation analyses for the Diagnostic Upper and Equatorial Port Plugs (UPP #3 and EPP #8, #17). The creation process of the complicated 3D MCNP models of the diagnostics systems was substantially accelerated by application of the CAD-to-MCNP converter programs MCAM and McCad. High performance computing resources of the Helios supercomputer allowed to speed-up the MCNP parallel transport calculations with the MPI/OpenMP interface. The found shielding solutions could be universal, reducing ports R&D costs. The shield block behind the Tritium and Deposit Monitor (TDM) optical box was added to study its influence on Shut-Down Dose Rate (SDDR) in Port Interspace (PI) of EPP#17. Influence of neutron streaming along the Lost Alpha Monitor (LAM) on the neutron energy spectra calculated in the Tangential Neutron Spectrometer (TNS) of EPP#8. For the UPP#3 with Charge eXchange Recombination Spectroscopy (CXRS-core), an excessive neutron streaming along the CXRS shutter, which should be prevented in further design iteration.

Cosmic Ray Interactions in Shielding Materials

DOE Office of Scientific and Technical Information (OSTI.GOV)

Aguayo Navarrete, Estanislao; Kouzes, Richard T.; Ankney, Austin S.

2011-09-08

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-electronmore » 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.« less

On the origin of the March 5, 1979 gamma ray transient: A vibrating neutron star in the Large Magellanic Cloud

NASA Technical Reports Server (NTRS)

Ramaty, R.; Bonazzola, S.; Cline, T. L.; Kazanas, D.; Meszaros, P.; Lingenfelter, R. E.

1980-01-01

It is proposed that a vibrating neutron star in the Large Magellanic Cloud is the source of the March 5 transient. Neutron star vibrations transport energy rapidly to the surface, heat the atmosphere by wave dissipation, and decay by gravitational radiation reaction. The electromagnetic emission arises from e(+)-e(-) pairs which cool and annihilate in the strong magnetic field of the neutron star. The field also confines the pairs, and this allows the production of the redshifted annihilation feature observed in the data. The redshift implies a gravitational radiation damping time which agrees with the 0.15 second duration of the impulsive phase of the event. Thus, the March 5 transient may be both the first detection of a vibrating neutron star and indirect evidence for gravitational radiation.

MCNP/X TRANSPORT IN THE TABULAR REGIME

DOE Office of Scientific and Technical Information (OSTI.GOV)

HUGHES, H. GRADY

2007-01-08

The authors review the transport capabilities of the MCNP and MCNPX Monte Carlo codes in the energy regimes in which tabular transport data are available. Giving special attention to neutron tables, they emphasize the measures taken to improve the treatment of a variety of difficult aspects of the transport problem, including unresolved resonances, thermal issues, and the availability of suitable cross sections sets. They also briefly touch on the current situation in regard to photon, electron, and proton transport tables.

Improved Convergence Rate of Multi-Group Scattering Moment Tallies for Monte Carlo Neutron Transport Codes

NASA Astrophysics Data System (ADS)

Nelson, Adam

Multi-group scattering moment matrices are critical to the solution of the multi-group form of the neutron transport equation, as they are responsible for describing the change in direction and energy of neutrons. These matrices, however, are difficult to correctly calculate from the measured nuclear data with both deterministic and stochastic methods. Calculating these parameters when using deterministic methods requires a set of assumptions which do not hold true in all conditions. These quantities can be calculated accurately with stochastic methods, however doing so is computationally expensive due to the poor efficiency of tallying scattering moment matrices. This work presents an improved method of obtaining multi-group scattering moment matrices from a Monte Carlo neutron transport code. This improved method of tallying the scattering moment matrices is based on recognizing that all of the outgoing particle information is known a priori and can be taken advantage of to increase the tallying efficiency (therefore reducing the uncertainty) of the stochastically integrated tallies. In this scheme, the complete outgoing probability distribution is tallied, supplying every one of the scattering moment matrices elements with its share of data. In addition to reducing the uncertainty, this method allows for the use of a track-length estimation process potentially offering even further improvement to the tallying efficiency. Unfortunately, to produce the needed distributions, the probability functions themselves must undergo an integration over the outgoing energy and scattering angle dimensions. This integration is too costly to perform during the Monte Carlo simulation itself and therefore must be performed in advance by way of a pre-processing code. The new method increases the information obtained from tally events and therefore has a significantly higher efficiency than the currently used techniques. The improved method has been implemented in a code system containing a new pre-processor code, NDPP, and a Monte Carlo neutron transport code, OpenMC. This method is then tested in a pin cell problem and a larger problem designed to accentuate the importance of scattering moment matrices. These tests show that accuracy was retained while the figure-of-merit for generating scattering moment matrices and fission energy spectra was significantly improved.

Scintillator and solid-state neutron detectors and their applications

NASA Astrophysics Data System (ADS)

Carturan, Sara Maria; Marchi, Tommaso; Fanchini, Erica; De Vita, Raffaella; Finocchiaro, Paolo; Pappalardo, Alfio

2014-10-01

The application range of neutron detectors covers many topics, not only involving experimental research, but spanning tens of industrial, health, transport, cultural heritage fields of interest. Several studies focus on new scintillating materials where the light response, under fast and slow neutrons exposure, is triggered by proton recoil or by the presence of neutron capture materials as 10B, 6Li or 157Gd. Neutron monitors, where the robustness of silicon-based detectors can be fully exploited by coupling with suitable neutron absorber/converter materials, have recently proved their outstanding performances. Discrimination between neutron signals from other radiations, such as - or cosmic rays, is achieved through timing techniques or with pulse shape analysis. Furthermore, the choice of the detection/discrimination techniques depends on the type of application the detector will be used for. An example is Radiation Portal Monitors (RPM) for cargo inspection or luggage control that are required to satisfy specific international standards for and neutron detection efficiencies. This paper is an overview of some of the National Institute of Nuclear Physics (INFN) activities in the field of neutron detection, involving novel technologies. We will describe the most recent advances related to scintillators and silicon-based detectors coupled with thin films of suitable converters for neutron detection and we will discuss applications in the field of nuclear security.

Study of neutron shielding collimators for curved beamlines at the European Spallation Source

NASA Astrophysics Data System (ADS)

Santoro, V.; DiJulio, D. D.; Ansell, S.; Cherkashyna, N.; Muhrer, G.; Bentley, P. M.

2018-06-01

The European Spallation Source is being constructed in Lund, Sweden and is planned to be the world’s brightest pulsed spallation neutron source for cold and thermal neutron beams (≤ 1 eV). The facility uses a 2 GeV proton beam to produce neutrons from a tungsten target. The neutrons are then moderated in a moderator assembly consisting of both liquid hydrogen and water compartments. Surrounding the moderator are 22 beamports, which view the moderator’s outside surfaces. The beamports are connected to long neutron guides that transport the moderated neutrons to the sample position via reflections. As well as the desired moderated neutrons, fast neutrons coming directly from the target can find their way down the beamlines. These can create unwanted sources of background for the instruments. To mitigate such a kind of background, several instruments will use curved guides to lose direct line-of-sight (LoS) to the moderator and the target. In addition instruments can also use shielding collimators to reduce the amount of fast neutrons further traveling down the guide due to albedo reflections or streaming. Several different materials have been proposed for this purpose. We present the results of a study of different options for collimators and identify the optimal choices that balance cost, background and activation levels.

A 10{sup 9} neutrons/pulse transportable pulsed D-D neutron source based on flexible head plasma focus unit

DOE Office of Scientific and Technical Information (OSTI.GOV)

Niranjan, Ram, E-mail: niranjan@barc.gov.in; Rout, R. K.; Srivastava, R.

2016-03-15

A 17 kJ transportable plasma focus (PF) device with flexible transmission lines is developed and is characterized. Six custom made capacitors are used for the capacitor bank (CB). The common high voltage plate of the CB is fixed to a centrally triggered spark gap switch. The output of the switch is coupled to the PF head through forty-eight 5 m long RG213 cables. The CB has a quarter time-period of 4 μs and an estimated current of 506 kA is delivered to the PF device at 17 kJ (60 μF, 24 kV) energy. The average neutron yield measured using silvermore » activation detector in the radial direction is (7.1 ± 1.4) × 10{sup 8} neutrons/shot over 4π sr at 5 mbar optimum D{sub 2} pressure. The average neutron yield is more in the axial direction with an anisotropy factor of 1.33 ± 0.18. The average neutron energies estimated in the axial as well as in the radial directions are (2.90 ± 0.20) MeV and (2.58 ± 0.20) MeV, respectively. The flexibility of the PF head makes it useful for many applications where the source orientation and the location are important factors. The influence of electromagnetic interferences from the CB as well as from the spark gap on applications area can be avoided by putting a suitable barrier between the bank and the PF head.« less

Thermoluminescence measurements of neutron dose around a medical linac.

PubMed

Barquero, R; Méndez, R; Iñiguez, M P; Vega, H R; Voytchev, M

2002-01-01

The photoncutron ambient dose around a 18 MV medical electron lineal accelerator has been measured with LiF:Mg,Ti chips of 3 x 3 x 1 mm inside moderating spheres. During the measurements a water phantom was irradiated in a field of 40 x 40 cm2. Two methods have been considered for comparison. In the first, a TLD-600/TLD-700 pair at the centre of a 25 cm diameter paraffine sphere was used, with the system behaving as a rem meter. In the second method, TLD-600/TLD-700 pairs, bare and at the centre of 7.6, 12.7, 20.3, 25.4, and 30.5 cm diameter polyethylene Bonner spheres were used to obtain the neutron spectrum. This was unfolded using the BUNKIUT code with the SPUNIT algorithm and the UTA4 and ARKI response functions. The neutron dose was followed by multiplying the unfolded neutron spectrum by the ambient dose equivalent to neutron fluence conversion factors. Both methods result in 0.5 mSv x Gy(-1) m away from the isocentre.

Self characterization of a coded aperture array for neutron source imaging

NASA Astrophysics Data System (ADS)

Volegov, P. L.; Danly, C. R.; Fittinghoff, D. N.; Guler, N.; Merrill, F. E.; Wilde, C. H.

2014-12-01

The neutron imaging system at the National Ignition Facility (NIF) is an important diagnostic tool for measuring the two-dimensional size and shape of the neutrons produced in the burning deuterium-tritium plasma during the stagnation stage of inertial confinement fusion implosions. Since the neutron source is small (˜100 μm) and neutrons are deeply penetrating (>3 cm) in all materials, the apertures used to achieve the desired 10-μm resolution are 20-cm long, triangular tapers machined in gold foils. These gold foils are stacked to form an array of 20 apertures for pinhole imaging and three apertures for penumbral imaging. These apertures must be precisely aligned to accurately place the field of view of each aperture at the design location, or the location of the field of view for each aperture must be measured. In this paper we present a new technique that has been developed for the measurement and characterization of the precise location of each aperture in the array. We present the detailed algorithms used for this characterization and the results of reconstructed sources from inertial confinement fusion implosion experiments at NIF.

Computational attributes of the integral form of the equation of transfer

NASA Technical Reports Server (NTRS)

Frankel, J. I.

1991-01-01

Difficulties can arise in radiative and neutron transport calculations when a highly anisotropic scattering phase function is present. In the presence of anisotropy, currently used numerical solutions are based on the integro-differential form of the linearized Boltzmann transport equation. This paper, departs from classical thought and presents an alternative numerical approach based on application of the integral form of the transport equation. Use of the integral formalism facilitates the following steps: a reduction in dimensionality of the system prior to discretization, the use of symbolic manipulation to augment the computational procedure, and the direct determination of key physical quantities which are derivable through the various Legendre moments of the intensity. The approach is developed in the context of radiative heat transfer in a plane-parallel geometry, and results are presented and compared with existing benchmark solutions. Encouraging results are presented to illustrate the potential of the integral formalism for computation. The integral formalism appears to possess several computational attributes which are well-suited to radiative and neutron transport calculations.

Review of current neutron detection systems for emergency response

DOE PAGES

Mukhopadhyay, Sanjoy; Maurer, Richard; Guss, Paul; ...

2014-09-05

Neutron detectors are utilized in a myriad of applications—from safeguarding special nuclear materials (SNM) to determining lattice spacing in soft materials. The transformational changes taking place in neutron detection and imaging techniques in the last few years are largely being driven by the global shortage of helium-3 ( 3He). This article reviews the status of neutron sensors used specifically for SNM detection in radiological emergency response. These neutron detectors must be highly efficient, be rugged, have fast electronics to measure neutron multiplicity, and be capable of measuring direction of the neutron sources and possibly image them with high spatial resolution.more » Neutron detection is an indirect physical process: neutrons react with nuclei in materials to initiate the release of one or more charged particles that produce electric signals that can be processed by the detection system. Therefore, neutron detection requires conversion materials as active elements of the detection system; these materials may include boron-10 ( 10B), lithium-6 ( 6Li), and gadollinium-157 ( 157Gd), to name a few, but the number of materials available for neutron detection is limited. However, in recent years, pulse-shape-discriminating plastic scintillators, scintillators made of helium-4 ( 4He) under high pressure, pillar and trench semiconductor diodes, and exotic semiconductor neutron detectors made from uranium oxide and other materials have widely expanded the parameter space in neutron detection methodology. In this article we will pay special attention to semiconductor-based neutron sensors. Finally, modern microfabricated nanotubes covered inside with neutron converter materials and with very high aspect ratios for better charge transport will be discussed.« less

Review of current neutron detection systems for emergency response

NASA Astrophysics Data System (ADS)

Mukhopadhyay, Sanjoy; Maurer, Richard; Guss, Paul; Kruschwitz, Craig

2014-09-01

Neutron detectors are used in a myriad of applications—from safeguarding special nuclear materials (SNM) to determining lattice spacing in soft materials. The transformational changes taking place in neutron detection and imaging techniques in the last few years are largely being driven by the global shortage of helium-3 (3He). This article reviews the status of neutron sensors used specifically for SNM detection in radiological emergency response. These neutron detectors must be highly efficient, be rugged, have fast electronics to measure neutron multiplicity, and be capable of measuring direction of the neutron sources and possibly image them with high spatial resolution. Neutron detection is an indirect physical process: neutrons react with nuclei in materials to initiate the release of one or more charged particles that produce electric signals that can be processed by the detection system. Therefore, neutron detection requires conversion materials as active elements of the detection system; these materials may include boron-10 (10B), lithium-6 (6Li), and gadollinium-157 (157Gd), to name a few, but the number of materials available for neutron detection is limited. However, in recent years, pulse-shape-discriminating plastic scintillators, scintillators made of helium-4 (4He) under high pressure, pillar and trench semiconductor diodes, and exotic semiconductor neutron detectors made from uranium oxide and other materials have widely expanded the parameter space in neutron detection methodology. In this article we will pay special attention to semiconductor-based neutron sensors. Modern microfabricated nanotubes covered inside with neutron converter materials and with very high aspect ratios for better charge transport will be discussed.

Methodes d'optimisation des parametres 2D du reflecteur dans un reacteur a eau pressurisee

NASA Astrophysics Data System (ADS)

Clerc, Thomas

With a third of the reactors in activity, the Pressurized Water Reactor (PWR) is today the most used reactor design in the world. This technology equips all the 19 EDF power plants. PWRs fit into the category of thermal reactors, because it is mainly the thermal neutrons that contribute to the fission reaction. The pressurized light water is both used as the moderator of the reaction and as the coolant. The active part of the core is composed of uranium, slightly enriched in uranium 235. The reflector is a region surrounding the active core, and containing mostly water and stainless steel. The purpose of the reflector is to protect the vessel from radiations, and also to slow down the neutrons and reflect them into the core. Given that the neutrons participate to the reaction of fission, the study of their behavior within the core is capital to understand the general functioning of how the reactor works. The neutrons behavior is ruled by the transport equation, which is very complex to solve numerically, and requires very long calculation. This is the reason why the core codes that will be used in this study solve simplified equations to approach the neutrons behavior in the core, in an acceptable calculation time. In particular, we will focus our study on the diffusion equation and approximated transport equations, such as SPN or S N equations. The physical properties of the reflector are radically different from those of the fissile core, and this structural change causes important tilt in the neutron flux at the core/reflector interface. This is why it is very important to accurately design the reflector, in order to precisely recover the neutrons behavior over the whole core. Existing reflector calculation techniques are based on the Lefebvre-Lebigot method. This method is only valid if the energy continuum of the neutrons is discretized in two energy groups, and if the diffusion equation is used. The method leads to the calculation of a homogeneous reflector. The aim of this study is to create a computational scheme able to compute the parameters of heterogeneous, multi-group reflectors, with both diffusion and SPN/SN operators. For this purpose, two computational schemes are designed to perform such a reflector calculation. The strategy used in both schemes is to minimize the discrepancies between a power distribution computed with a core code and a reference distribution, which will be obtained with an APOLLO2 calculation based on the method Method Of Characteristics (MOC). In both computational schemes, the optimization parameters, also called control variables, are the diffusion coefficients in each zone of the reflector, for diffusion calculations, and the P-1 corrected macroscopic total cross-sections in each zone of the reflector, for SPN/SN calculations (or correction factors on these parameters). After a first validation of our computational schemes, the results are computed, always by optimizing the fast diffusion coefficient for each zone of the reflector. All the tools of the data assimilation have been used to reflect the different behavior of the solvers in the different parts of the core. Moreover, the reflector is refined in six separated zones, corresponding to the physical structure of the reflector. There will be then six control variables for the optimization algorithms. [special characters omitted]. Our computational schemes are then able to compute heterogeneous, 2-group or multi-group reflectors, using diffusion or SPN/SN operators. The optimization performed reduces the discrepancies distribution between the power computed with the core codes and the reference power. However, there are two main limitations to this study: first the homogeneous modeling of the reflector assemblies doesn't allow to properly describe its physical structure near the core/reflector interface. Moreover, the fissile assemblies are modeled in infinite medium, and this model reaches its limit at the core/reflector interface. These two problems should be tackled in future studies. (Abstract shortened by UMI.).

United States Air Force Summer Research Program -- 1993 Summer Research Program Final Reports. Volume 12. Armstrong Laboratory

DTIC Science & Technology

1993-01-01

Panasonic TLD . Panasonic Industrial Company; Secaucus, New Jersey. 5. Thurlow, Ronald M. "Neutron Dosimetry Using a Panasonic Thermoluminescent Dosimeter." A...steps 8-12. 29-15 THE BUILDING OF THE USAF PANASONIC UD-809AS ALGORITHM Katherine M. Arnold Research Associate Radiation Dosimetry Branch Brooks Air...Research August 1993 30-1 THE BUILDING OF THE USAF PANASONIC UD-809AS ALGORITHM Katherine M. Arnold Research Associate Radiation Dosimetry Branch

Implementation of the ground level enhancement alert software at NMDB database

NASA Astrophysics Data System (ADS)

Mavromichalaki, Helen; Souvatzoglou, George; Sarlanis, Christos; Mariatos, George; Papaioannou, Athanasios; Belov, Anatoly; Eroshenko, Eugenia; Yanke, Victor; NMDB Team

2010-11-01

The European Commission is supporting the real-time database for high-resolution neutron monitor measurements (NMDB) as an e-Infrastructures project in the Seventh Framework Programme in the Capacities section. The realization of the NMDB will provide the opportunity for several applications most of which will be implemented in real-time. An important application will be the establishment of an Alert signal when dangerous solar particle events are heading to the Earth, resulting into a ground level enhancement (GLE) registered by neutron monitors (NMs). The cosmic ray community has been occupied with the question of establishing such an Alert for many years and recently several groups succeeded in creating a proper algorithm capable of detecting space weather threats in an off-line mode. A lot of original work has been done to this direction and every group working in this field performed routine runs for all GLE cases, resulting into statistical analyses of GLE events. The next step was to make this algorithm as accurate as possible and most importantly, working in real-time. This was achieved when, during the last GLE observed so far, a real-time GLE Alert signal was produced. In this work, the steps of this procedure as well as the functionality of this algorithm for both the scientific community and users are being discussed. Nevertheless, the transition of the Alert algorithm to the NMDB is also being discussed.

A stochastic method for Brownian-like optical transport calculations in anisotropic biosuspensions and blood

NASA Astrophysics Data System (ADS)

Miller, Steven

1998-03-01

A generic stochastic method is presented that rapidly evaluates numerical bulk flux solutions to the one-dimensional integrodifferential radiative transport equation, for coherent irradiance of optically anisotropic suspensions of nonspheroidal bioparticles, such as blood. As Fermat rays or geodesics enter the suspension, they evolve into a bundle of random paths or trajectories due to scattering by the suspended bioparticles. Overall, this can be interpreted as a bundle of Markov trajectories traced out by a "gas" of Brownian-like point photons being scattered and absorbed by the homogeneous distribution of uncorrelated cells in suspension. By considering the cumulative vectorial intersections of a statistical bundle of random trajectories through sets of interior data planes in the space containing the medium, the effective equivalent information content and behavior of the (generally unknown) analytical flux solutions of the radiative transfer equation rapidly emerges. The fluxes match the analytical diffuse flux solutions in the diffusion limit, which verifies the accuracy of the algorithm. The method is not constrained by the diffusion limit and gives correct solutions for conditions where diffuse solutions are not viable. Unlike conventional Monte Carlo and numerical techniques adapted from neutron transport or nuclear reactor problems that compute scalar quantities, this vectorial technique is fast, easily implemented, adaptable, and viable for a wide class of biophotonic scenarios. By comparison, other analytical or numerical techniques generally become unwieldy, lack viability, or are more difficult to utilize and adapt. Illustrative calculations are presented for blood medias at monochromatic wavelengths in the visible spectrum.

A study of gamma-ray and neutron radiation in the interaction of a 2 MeV proton beam with various materials.

PubMed

Kasatov, D; Makarov, A; Shchudlo, I; Taskaev, S

2015-12-01

Epithermal neutron source based on a tandem accelerator with vacuum insulation and lithium target has been proposed, developed and operated in Budker Institute of Nuclear Physics. The source is regarded as a prototype of a future compact device suitable for carrying out BNCT in oncology centers. In this work the measurements of gamma-ray and neutron radiation are presented for the interaction of a 2 MeV proton beam with various materials (Li, C, F, Al, V, Ti, Cu, Mo, stainless steel, and Ta). The obtained results enabled the optimization of the neutron-generating target and the high energy beam transportation path. Copyright © 2015 Elsevier Ltd. All rights reserved.

Conceptual design of a neutron camera for MAST Upgrade

DOE Office of Scientific and Technical Information (OSTI.GOV)

Weiszflog, M., E-mail: matthias.weiszflog@physics.uu.se; Sangaroon, S.; Cecconello, M.

2014-11-15

This paper presents two different conceptual designs of neutron cameras for Mega Ampere Spherical Tokamak (MAST) Upgrade. The first one consists of two horizontal cameras, one equatorial and one vertically down-shifted by 65 cm. The second design, viewing the plasma in a poloidal section, also consists of two cameras, one radial and the other one with a diagonal view. Design parameters for the different cameras were selected on the basis of neutron transport calculations and on a set of target measurement requirements taking into account the predicted neutron emissivities in the different MAST Upgrade operating scenarios. Based on a comparisonmore » of the cameras’ profile resolving power, the horizontal cameras are suggested as the best option.« less

Quasi solution of radiation transport equation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pogosbekyan, L.R.; Lysov, D.A.

There is uncertainty with experimental data as well as with input data of theoretical calculations. The neutron distribution from the variational principle, which takes into account both theoretical and experimental data, is obtained to increase the accuracy and speed of neutronic calculations. The neutron imbalance in mesh cells and the discrepancy between experimentally measured and calculated functional of the neutron distribution are simultaneously minimized. A fast-working and simple-programming iteration method is developed to minimize the objective functional. The method can be used in the core monitoring and control system for (a) power distribution calculations, (b) in- and ex-core detector calibration,more » (c) macro-cross sections or isotope distribution correction by experimental data, and (d) core and detector diagnostics.« less

Kinetic: A system code for analyzing nuclear thermal propulsion rocket engine transients

NASA Astrophysics Data System (ADS)

Schmidt, Eldon; Lazareth, Otto; Ludewig, Hans

The topics are presented in viewgraph form and include the following: outline of kinetic code; a kinetic information flow diagram; kinetic neutronic equations; turbopump/nozzle algorithm; kinetic heat transfer equations per node; and test problem diagram.

Neutron optics concept for the materials engineering diffractometer at the ESS

NASA Astrophysics Data System (ADS)

Šaroun, J.; Fenske, J.; Rouijaa, M.; Beran, P.; Navrátil, J.; Lukáš, P.; Schreyer, A.; Strobl, M.

2016-09-01

The Beamline for European Materials Engineering Research (BEER) has been recently proposed to be built at the European Spallation Source (ESS). The presented concept of neutron delivery optics for this instrument addresses the problems of bi-spectral beam extraction from a small moderator, optimization of neutron guides profile for long-range neutron transport and focusing at the sample under various constraints. They include free space before and after the guides, a narrow guide section with gaps for choppers, closing of direct line of sight and cost reduction by optimization of the guides cross-section and coating. A system of slits and exchangeable focusing optics is proposed in order to match various wavelength resolution options provided by the pulse shaping and modulation choppers, which permits to efficiently trade resolution for intensity in a wide range. Simulated performance characteristics such as brilliance transfer ratio are complemented by the analysis of the histories of “useful” neutrons obtained by back tracing neutrons hitting the sample, which helps to optimize some of the neutron guide parameters such as supermirror coating.

Opportunities for Undergraduate Research in Nuclear Physics

DOE PAGES

Hicks, S. F.; Nguyen, T. D.; Jackson, D. T.; ...

2017-10-26

University of Dallas (UD) physics majors are offered a variety of undergraduate research opportunities in nuclear physics through an established program at the University of Kentucky Accelerator Laboratory (UKAL). The 7-MV Model CN Van de Graaff accelerator and the neutron production and detection facilities located there are used by UD students to investigate how neutrons scatter from materials that are important in nuclear energy production and for our basic understanding of how neutrons interact with matter. Recent student projects include modeling of the laboratory using the neutron transport code MCNP to investigate the effectiveness of laboratory shielding, testing the long-termmore » gain stability of C 6D 6 liquid scintillation detectors, and deducing neutron elastic and inelastic scattering cross sections for 12C. Finally, results of these student projects are presented that indicate the pit below the scattering area reduces background by as much as 30%; the detectors show no significant gain instabilities; and new insights into existing 12C neutron inelastic scattering cross-section discrepancies near a neutron energy of 6.0 MeV are obtained.« less

Opportunities for Undergraduate Research in Nuclear Physics

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hicks, S. F.; Nguyen, T. D.; Jackson, D. T.

University of Dallas (UD) physics majors are offered a variety of undergraduate research opportunities in nuclear physics through an established program at the University of Kentucky Accelerator Laboratory (UKAL). The 7-MV Model CN Van de Graaff accelerator and the neutron production and detection facilities located there are used by UD students to investigate how neutrons scatter from materials that are important in nuclear energy production and for our basic understanding of how neutrons interact with matter. Recent student projects include modeling of the laboratory using the neutron transport code MCNP to investigate the effectiveness of laboratory shielding, testing the long-termmore » gain stability of C 6D 6 liquid scintillation detectors, and deducing neutron elastic and inelastic scattering cross sections for 12C. Finally, results of these student projects are presented that indicate the pit below the scattering area reduces background by as much as 30%; the detectors show no significant gain instabilities; and new insights into existing 12C neutron inelastic scattering cross-section discrepancies near a neutron energy of 6.0 MeV are obtained.« less

Radioactivity measurements of ITER materials using the TFTR D-T neutron field

NASA Astrophysics Data System (ADS)

Kumar, A.; Abdou, M. A.; Barnes, C. W.; Kugel, H. W.

1994-06-01

The availability of high D-T fusion neutron yields at TFTR has provided a useful opportunity to directly measure D-T neutron-induced radioactivity in a realistic tokamak fusion reactor environment for materials of vital interest to ITER. These measurements are valuable for characterizing radioactivity in various ITER candidate materials, for validating complex neutron transport calculations, and for meeting fusion reactor licensing requirements. The radioactivity measurements at TFTR involve potential ITER materials including stainless steel 316, vanadium, titanium, chromium, silicon, iron, cobalt, nickel, molybdenum, aluminum, copper, zinc, zirconium, niobium, and tungsten. Small samples of these materials were irradiated close to the plasma and just outside the vacuum vessel wall of TFTR, locations of different neutron energy spectra. Saturation activities for both threshold and capture reactions were measured. Data from dosimetric reactions have been used to obtain preliminary neutron energy spectra. Spectra from the first wall were compared to calculations from ITER and to measurements from accelerator-based tests.

Comparison of Analysis Results Between 2D/1D Synthesis and RAPTOR-M3G in the Korea Standard Nuclear Plant (KSNP)

NASA Astrophysics Data System (ADS)

Joung Lim, Mi; Maeng, Young Jae; Fero, Arnold H.; Anderson, Stanwood L.

2016-02-01

The 2D/1D synthesis methodology has been used to calculate the fast neutron (E > 1.0 MeV) exposure to the beltline region of the reactor pressure vessel. This method uses the DORT 3.1 discrete ordinates code and the BUGLE-96 cross-section library based on ENDF/B-VI. RAPTOR-M3G (RApid Parallel Transport Of Radiation-Multiple 3D Geometries) which performs full 3D calculations was developed and is based on domain decomposition algorithms, where the spatial and angular domains are allocated and processed on multi-processor computer architecture. As compared to traditional single-processor applications, this approach reduces the computational load as well as the memory requirement per processor. Both methods are applied to surveillance test results for the Korea Standard Nuclear Plant (KSNP)-OPR (Optimized Power Reactor) 1000 MW. The objective of this paper is to compare the results of the KSNP surveillance program between 2D/1D synthesis and RAPTOR-M3G. Each operating KSNP has a reactor vessel surveillance program consisting of six surveillance capsules located between the core and the reactor vessel in the downcomer region near the reactor vessel wall. In addition to the In-Vessel surveillance program, an Ex-Vessel Neutron Dosimetry (EVND) program has been implemented. In order to estimate surveillance test results, cycle-specific forward transport calculations were performed by 2D/1D synthesis and by RAPTOR-M3G. The ratio between measured and calculated (M/C) reaction rates will be discussed. The current plan is to install an EVND system in all of the Korea PWRs including the new reactor type, APR (Advanced Power Reactor) 1400 MW. This work will play an important role in establishing a KSNP-specific database of surveillance test results and will employ RAPTOR-M3G for surveillance dosimetry location as well as positions in the KSNP reactor vessel.

Simulation of Charge Collection in Diamond Detectors Irradiated with Deuteron-Triton Neutron Sources

DOE Office of Scientific and Technical Information (OSTI.GOV)

Milocco, Alberto; Trkov, Andrej; Pillon, Mario

2011-12-13

Diamond-based neutron spectrometers exhibit outstanding properties such as radiation hardness, low sensitivity to gamma rays, fast response and high-energy resolution. They represent a very promising application of diamonds for plasma diagnostics in fusion devices. The measured pulse height spectrum is obtained from the collection of helium and beryllium ions produced by the reactions on {sup 12}C. An original code is developed to simulate the production and the transport of charged particles inside the diamond detector. The ion transport methodology is based on the well-known TRIM code. The reactions of interest are triggered using the ENDF/B-VII.0 nuclear data for the neutronmore » interactions on carbon. The model is implemented in the TALLYX subroutine of the MCNP5 and MCNPX codes. Measurements with diamond detectors in a {approx}14 MeV neutron field have been performed at the FNG (Rome, Italy) and IRMM (Geel, Belgium) facilities. The comparison of the experimental data with the simulations validates the proposed model.« less

Dosimetry analyses of the Ringhals 3 and 4 reactor pressure vessels

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kulesza, J.A.; Fero, A.H.; Rouden, J.

2011-07-01

A comprehensive series of neutron dosimetry measurements consisting of surveillance capsules, reactor pressure vessel cladding samples, and ex-vessel neutron dosimetry has been analyzed and compared to the results of three-dimensional, cycle-specific neutron transport calculations for the Ringhals Unit 3 and Unit 4 reactors in Sweden. The comparisons show excellent agreement between calculations and measurements. The measurements also demonstrate that it is possible to perform retrospective dosimetry measurements using the {sup 93}Nb (n,n') {sup 93m}Nb reaction on samples of 18-8 austenitic stainless steel with only trace amounts of elemental niobium. (authors)

The LBM program at the EPFL/LOTUS Facility

DOE Office of Scientific and Technical Information (OSTI.GOV)

File, J.; Jassby, D.L.; Tsang, F.Y.

1986-11-01

An experimental program of neutron transport studies of the Lithium Blanket Module (LBM) is being carried out with the LOTUS point-neutron source facility at Ecole Polytechnique Federale de Lausanne (EPFL), Switzerland. Preliminary experiments use passive neutron dosimetry within the fuel rods in the LBM central zone, as well as, both thermal extraction and dissolution methods to assay tritium bred in Li/sub 2/O diagnostic wafers and LBM pellets. These measurements are being compared and reconciled with each other and with the predictions of two-dimensional discrete-ordinates and continuous-energy Monte-Carlo analyses of the Lotus/LBM system.

New Methodologies for Generation of Multigroup Cross Sections for Shielding Applications

NASA Astrophysics Data System (ADS)

Arzu Alpan, F.; Haghighat, Alireza

2003-06-01

Coupled neutron and gamma multigroup (broad-group) libraries used for Light Water Reactor shielding and dosimetry commonly include 47-neutron and 20-gamma groups. These libraries are derived from the 199-neutron, 42-gamma fine-group VITAMIN-B6 library. In this paper, we introduce modifications to the generation procedure of the broad-group libraries. Among these modifications, we show that the fine-group structure and collapsing technique have the largest impact. We demonstrate that a more refined fine-group library and the bi-linear adjoint weighting collapsing technique can improve the accuracy of transport calculation results.

Fast non-overlapping Schwarz domain decomposition methods for solving the neutron diffusion equation

NASA Astrophysics Data System (ADS)

Jamelot, Erell; Ciarlet, Patrick

2013-05-01

Studying numerically the steady state of a nuclear core reactor is expensive, in terms of memory storage and computational time. In order to address both requirements, one can use a domain decomposition method, implemented on a parallel computer. We present here such a method for the mixed neutron diffusion equations, discretized with Raviart-Thomas-Nédélec finite elements. This method is based on the Schwarz iterative algorithm with Robin interface conditions to handle communications. We analyse this method from the continuous point of view to the discrete point of view, and we give some numerical results in a realistic highly heterogeneous 3D configuration. Computations are carried out with the MINOS solver of the APOLLO3® neutronics code. APOLLO3 is a registered trademark in France.

MC ray-tracing optimization of lobster-eye focusing devices with RESTRAX

NASA Astrophysics Data System (ADS)

Šaroun, Jan; Kulda, Jiří

2006-11-01

The enhanced functionalities of the latest version of the RESTRAX software, providing a high-speed Monte Carlo (MC) ray-tracing code to represent a virtual three-axis neutron spectrometer, include representation of parabolic and elliptic guide profiles and facilities for numerical optimization of parameter values, characterizing the instrument components. As examples, we present simulations of a doubly focusing monochromator in combination with cold neutron guides and lobster-eye supermirror devices, concentrating a monochromatic beam to small sample volumes. A Levenberg-Marquardt minimization algorithm is used to optimize simultaneously several parameters of the monochromator and lobster-eye guides. We compare the performance of optimized configurations in terms of monochromatic neutron flux and energy spread and demonstrate the effect of lobster-eye optics on beam transformations in real and momentum subspaces.

Scientific Design of the New Neutron Radiography Facility (SANRAD) at SAFARI-1 for South Africa

NASA Astrophysics Data System (ADS)

de Beer, F. C.; Gruenauer, F.; Radebe, J. M.; Modise, T.; Schillinger, B.

The final scientific design for an upgraded neutron radiography/tomography facility at beam port no.2 of the SAFARI-1 nuclear research reactor has been performed through expert advice from Physics Consulting, FRMII in Germany and IPEN, Brazil. A need to upgrade the facility became apparent due to the identification of various deficiencies of the current SANRAD facility during an IAEA-sponsored expert mission of international scientists to Necsa, South Africa. A lack of adequate shielding that results in high neutron background on the beam port floor, a mismatch in the collimator aperture to the core that results in a high gradient in neutron flux on the imaging plane and due to a relative low L/D the quality of the radiographs are poor, are a number of deficiencies to name a few.The new design, based on results of Monte Carlo (MCNP-X) simulations of neutron- and gamma transport from the reactor core and through the new facility, is being outlined. The scientific design philosophy, neutron optics and imaging capabilities that include the utilization of fission neutrons, thermal neutrons, and gamma-rays emerging from the core of SAFARI-1 are discussed.

Monte Carlo simulation of explosive detection system based on a Deuterium-Deuterium (D-D) neutron generator.

PubMed

Bergaoui, K; Reguigui, N; Gary, C K; Brown, C; Cremer, J T; Vainionpaa, J H; Piestrup, M A

2014-12-01

An explosive detection system based on a Deuterium-Deuterium (D-D) neutron generator has been simulated using the Monte Carlo N-Particle Transport Code (MCNP5). Nuclear-based explosive detection methods can detect explosives by identifying their elemental components, especially nitrogen. Thermal neutron capture reactions have been used for detecting prompt gamma emission (10.82MeV) following radiative neutron capture by (14)N nuclei. The explosive detection system was built based on a fully high-voltage-shielded, axial D-D neutron generator with a radio frequency (RF) driven ion source and nominal yield of about 10(10) fast neutrons per second (E=2.5MeV). Polyethylene and paraffin were used as moderators with borated polyethylene and lead as neutron and gamma ray shielding, respectively. The shape and the thickness of the moderators and shields are optimized to produce the highest thermal neutron flux at the position of the explosive and the minimum total dose at the outer surfaces of the explosive detection system walls. In addition, simulation of the response functions of NaI, BGO, and LaBr3-based γ-ray detectors to different explosives is described. Copyright © 2014 Elsevier Ltd. All rights reserved.

Evaluation of neutron skyshine from a cyclotron

DOE Office of Scientific and Technical Information (OSTI.GOV)

Huyashi, K.; Nakamura, T.

1984-06-01

The dose distribution and the spectrum variation of neutrons due to the skyshine effect have been measured with various detectors in the environment surrounding the cyclotron of the Institute for Nuclear Study, University of Tokyo. The source neutrons were produced by stopping a 52-MeV proton beam into a carbon beam stopper and were extracted upward from the opening in the concrete shield surrounding the cyclotron and then leaked into the atmosphere through the cyclotron building. The dose distribution and the spectrum of neutrons near the beam stopper were also measured in order to get information on the skyshine source. Themore » measured skyshine neutron spectra and dose distribution were analyzed with two codes, MMCR2 and SKYSHINE-II, with the result that the calculated results are in good agreement with the experiment. Valuable characteristics of this experiment are the determination of the energy spectrum and dose distribution of source neutron and the measurement of skyshine neutrons from an actual large-scale accelerator building to the exclusion of direct neutrons transported through the air. This experiment must be useful as a kind of benchmark experiment on the skyshine phenomenon.« less

Active-Interrogation Measurements of Induced-Fission Neutrons from Low-Enriched Uranium

DOE Office of Scientific and Technical Information (OSTI.GOV)

J. L. Dolan; M. J. Marcath; M. Flaska

2012-07-01

Protection and control of nuclear fuels is paramount for nuclear security and safeguards; therefore, it is important to develop fast and robust controlling mechanisms to ensure the safety of nuclear fuels. Through both passive- and active-interrogation methods we can use fast-neutron detection to perform real-time measurements of fission neutrons for process monitoring. Active interrogation allows us to use different ranges of incident neutron energy to probe for different isotopes of uranium. With fast-neutron detectors, such as organic liquid scintillation detectors, we can detect the induced-fission neutrons and photons and work towards quantifying a sample’s mass and enrichment. Using MCNPX-PoliMi, amore » system was designed to measure induced-fission neutrons from U-235 and U-238. Measurements were then performed in the summer of 2010 at the Joint Research Centre in Ispra, Italy. Fissions were induced with an associated particle D-T generator and an isotopic Am-Li source. The fission neutrons, as well as neutrons from (n, 2n) and (n, 3n) reactions, were measured with five 5” by 5” EJ-309 organic liquid scintillators. The D-T neutron generator was available as part of a measurement campaign in place by Padova University. The measurement and data-acquisition systems were developed at the University of Michigan utilizing a CAEN V1720 digitizer and pulse-shape discrimination algorithms to differentiate neutron and photon detections. Low-enriched uranium samples of varying mass and enrichment were interrogated. Acquired time-of-flight curves and cross-correlation curves are currently analyzed to draw relationships between detected neutrons and sample mass and enrichment. In the full paper, the promise of active-interrogation measurements and fast-neutron detection will be assessed through the example of this proof-of-concept measurement campaign. Additionally, MCNPX-PoliMi simulation results will be compared to the measured data to validate the MCNPX-PoliMi code when used for active-interrogation simulations.« less

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

PubMed

Schneider, Uwe; Hälg, Roger A; Lomax, Tony

2017-06-01

One of the essential elements of an epidemiological study to decide if proton therapy may be associated with increased or decreased subsequent malignancies compared to photon therapy is an ability to estimate all doses to non-target tissues, including neutron dose. This work therefore aims to predict for patients using proton pencil beam scanning the spatially localized neutron doses and dose equivalents. The proton pencil beam of Gantry 1 at the Paul Scherrer Institute (PSI) was Monte Carlo simulated using GEANT. Based on the simulated neutron dose and neutron spectra an analytical mechanistic dose model was developed. The pencil beam algorithm used for treatment planning at PSI has been extended using the developed model in order to calculate the neutron component of the delivered dose distribution for each treated patient. The neutron dose was estimated for two patient example cases. The analytical neutron dose model represents the three-dimensional Monte Carlo simulated dose distribution up to 85cm from the proton pencil beam with a satisfying precision. The root mean square error between Monte Carlo simulation and model is largest for 138MeV protons and is 19% and 20% for dose and dose equivalent, respectively. The model was successfully integrated into the PSI treatment planning system. In average the neutron dose is increased by 10% or 65% when using 160MeV or 177MeV instead of 138MeV. For the neutron dose equivalent the increase is 8% and 57%. The presented neutron dose calculations allow for estimates of dose that can be used in subsequent epidemiological studies or, should the need arise, to estimate the neutron dose at any point where a subsequent secondary tumour may occur. It was found that the neutron dose to the patient is heavily increased with proton energy. Copyright © 2016. Published by Elsevier GmbH.

Dose conversion coefficients for neutron exposure to the lens of the human eye.

PubMed

Manger, R P; Bellamy, M B; Eckerman, K F

2012-03-01

Dose conversion coefficients for the lens of the human eye have been calculated for neutron exposure at energies from 1 × 10(-9) to 20 MeV and several standard orientations: anterior-to-posterior, rotational and right lateral. MCNPX version 2.6.0, a Monte Carlo-based particle transport package, was used to determine the energy deposited in the lens of the eye. The human eyeball model was updated by partitioning the lens into sensitive and insensitive volumes as the anterior portion (sensitive volume) of the lens being more radiosensitive and prone to cataract formation. The updated eye model was used with the adult UF-ORNL mathematical phantom in the MCNPX transport calculations.

Calculation of dose contributions of electron and charged heavy particles inside phantoms irradiated by monoenergetic neutron.

PubMed

Satoh, Daiki; Takahashi, Fumiaki; Endo, Akira; Ohmachi, Yasushi; Miyahara, Nobuyuki

2008-09-01

The radiation-transport code PHITS with an event generator mode has been applied to analyze energy depositions of electrons and charged heavy particles in two spherical phantoms and a voxel-based mouse phantom upon neutron irradiation. The calculations using the spherical phantoms quantitatively clarified the type and energy of charged particles which are released through interactions of neutrons with the phantom elements and contribute to the radiation dose. The relative contribution of electrons increased with an increase in the size of the phantom and with a decrease in the energy of the incident neutrons. Calculations with the voxel-based mouse phantom for 2.0-MeV neutron irradiation revealed that the doses to different locations inside the body are uniform, and that the energy is mainly deposited by recoil protons. The present study has demonstrated that analysis using PHITS can yield dose distributions that are accurate enough for RBE evaluation.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Corre, G.; Boudergui, K.; Sannie, G.

Homeland security requests the use Radiation Portal Monitor (RPM). They must be able to detect and differentiate gamma and neutron radiation. Gamma detection is required for illicit transportation of radioactive matter detection. Neutron detection is important to control nonproliferation of enriched material. Manufacturers worldwide propose sensors based on {sup 3}He which give the actual state of art in term of neutron detection. The imminent shortage of {sup 3}He forces manufacturers to find viable alternative. From 10 years sensors providers have the challenge to replace previous {sup 3}He detectors that are known to be the most commonly deployed neutron sensor. Asmore » {sup 3}He detectors can only detect neutron, they must be completed with gamma detector. The proposed approach is based on pulse time correlation between adjacent sensors from signal collected by EJ200 plastic scintillators. Results obtained during FP7 Scintilla project test campaigns show the system relevance for replacement of today's {sup 3}He detectors. (authors)« less

Neutronics Analysis of Water-Cooled Ceramic Breeder Blanket for CFETR

NASA Astrophysics Data System (ADS)

Zhu, Qingjun; Li, Jia; Liu, Songlin

2016-07-01

In order to investigate the nuclear response to the water-cooled ceramic breeder blanket models for CFETR, a detailed 3D neutronics model with 22.5° torus sector was developed based on the integrated geometry of CFETR, including heterogeneous WCCB blanket models, shield, divertor, vacuum vessel, toroidal and poloidal magnets, and ports. Using the Monte Carlo N-Particle Transport Code MCNP5 and IAEA Fusion Evaluated Nuclear Data Library FENDL2.1, the neutronics analyses were performed. The neutron wall loading, tritium breeding ratio, the nuclear heating, neutron-induced atomic displacement damage, and gas production were determined. The results indicate that the global TBR of no less than 1.2 will be a big challenge for the water-cooled ceramic breeder blanket for CFETR. supported by the National Magnetic Confinement Fusion Science Program of China (Nos. 2013GB108004, 2014GB122000, and 2014GB119000), and National Natural Science Foundation of China (No. 11175207)

In-vivo assessment of total body protein in rats by prompt-γ neutron activation analysis

NASA Astrophysics Data System (ADS)

Stamatelatos, Ion E.; Boozer, Carol N.; Ma, Ruimei; Yasumura, Seiichi

1997-02-01

A prompt-(gamma) neutron activation analysis facility for in vivo determination of total body protein (TBP) in rats has been designed. TBP is determined in vivo by assessment of total body nitrogen. The facility is based on a 252Cf radionuclide neutron source within a heavy water moderator assembly and two NaI(Tl) scintillation detectors. The in vivo precision of the technique, as estimated by three repeated measurements of 15 rats is 6 percent, for a radiation dose equivalent of 60 mSv. The radiation dose per measurement is sufficiently low to enable serial measurements on the same animal. MCNP-4A Monte Carlo transport code was utilized to calculate thermal neutron flux correction factors to account for differences in size and shape of the rats and calibration phantoms. Good agrement was observed in comparing body nitrogen assessment by prompt-(gamma) neutron activation and chemical carcass analysis.

Towards wide-angle neutron polarization analysis with a 3He spin filter for TOPAS and NEAT: Testing magic PASTIS on V20 at HZB

NASA Astrophysics Data System (ADS)

Babcock, Earl; Salhi, Zahir; Gainov, Ramil; Woracek, Robin; Soltner, Helmut; Pistel, Patrick; Beule, Fabian; Bussmann, Klaus; Heynen, Achim; Kämmerling, Hans; Suxdorf, Frank; Strobl, Marcus; Russina, Margarita; Voigt, Jörg; Ioffe, Alexander

2018-05-01

An XYZ polarization analysis solution has been developed for the new thermal time-of-flight spectrometer TOPAS [1], to be operated in the coming east neutron guide hall at the MLZ. This prototype is currently being prepared to use on NEAT at HZB [2]. Polarization Analysis Studies on a Thermal Inelastic Spectrometer, commonly called PASTIS [3], is based on polarized 3He neutron spin filters and an XYZ field configuration for the sample environment and a polarization-preserving neutron guide field. The complete system was designed to provide adiabatic transport of the neutron polarization to the sample position on TOPAS while maintaining the homogeneity of the XYZ field. This complete system has now been tested on the polarized time-of-flight ESS test beam line V20 at HZB [4]. We present results of this test and the next steps forward.

Experimental evaluation of the extended Dytlewski-style dead time correction formalism for neutron multiplicity counting

NASA Astrophysics Data System (ADS)

Lockhart, M.; Henzlova, D.; Croft, S.; Cutler, T.; Favalli, A.; McGahee, Ch.; Parker, R.

2018-01-01

Over the past few decades, neutron multiplicity counting has played an integral role in Special Nuclear Material (SNM) characterization pertaining to nuclear safeguards. Current neutron multiplicity analysis techniques use singles, doubles, and triples count rates because a methodology to extract and dead time correct higher order count rates (i.e. quads and pents) was not fully developed. This limitation is overcome by the recent extension of a popular dead time correction method developed by Dytlewski. This extended dead time correction algorithm, named Dytlewski-Croft-Favalli(DCF), is detailed in reference Croft and Favalli (2017), which gives an extensive explanation of the theory and implications of this new development. Dead time corrected results can then be used to assay SNM by inverting a set of extended point model equations which as well have only recently been formulated. The current paper discusses and presents the experimental evaluation of practical feasibility of the DCF dead time correction algorithm to demonstrate its performance and applicability in nuclear safeguards applications. In order to test the validity and effectiveness of the dead time correction for quads and pents, 252Cf and SNM sources were measured in high efficiency neutron multiplicity counters at the Los Alamos National Laboratory (LANL) and the count rates were extracted up to the fifth order and corrected for dead time. In order to assess the DCF dead time correction, the corrected data is compared to traditional dead time correction treatment within INCC. The DCF dead time correction is found to provide adequate dead time treatment for broad range of count rates available in practical applications.

Simulation of a beam rotation system for a spallation source

NASA Astrophysics Data System (ADS)

Reiss, Tibor; Reggiani, Davide; Seidel, Mike; Talanov, Vadim; Wohlmuther, Michael

2015-04-01

With a nominal beam power of nearly 1 MW on target, the Swiss Spallation Neutron Source (SINQ), ranks among the world's most powerful spallation neutron sources. The proton beam transport to the SINQ target is carried out exclusively by means of linear magnetic elements. In the transport line to SINQ the beam is scattered in two meson production targets and as a consequence, at the SINQ target entrance the beam shape can be described by Gaussian distributions in transverse x and y directions with tails cut short by collimators. This leads to a highly nonuniform power distribution inside the SINQ target, giving rise to thermal and mechanical stresses. In view of a future proton beam intensity upgrade, the possibility of homogenizing the beam distribution by means of a fast beam rotation system is currently under investigation. Important aspects which need to be studied are the impact of a rotating proton beam on the resulting neutron spectra, spatial flux distributions and additional—previously not present—proton losses causing unwanted activation of accelerator components. Hence a new source description method was developed for the radiation transport code MCNPX. This new feature makes direct use of the results from the proton beam optics code TURTLE. Its advantage to existing MCNPX source options is that all phase space information and correlations of each primary beam particle computed with TURTLE are preserved and transferred to MCNPX. Simulations of the different beam distributions together with their consequences in terms of neutron production are presented in this publication. Additionally, a detailed description of the coupling method between TURTLE and MCNPX is provided.

Evidence of neutron leakage at the Fukushima nuclear plant from measurements of radioactive 35S in California

PubMed Central

Priyadarshi, Antra; Dominguez, Gerardo; Thiemens, Mark H.

2011-01-01

A recent earthquake and the subsequent tsunami have extensively damaged the Fukushima nuclear power plant, releasing harmful radiation into the environment. Despite the obvious implication for human health and the surrounding ecology, there are no quantitative estimates of the neutron flux leakage during the weeks following the earthquake. Here, using measurements of radioactive 35S contained in sulfate aerosols and SO2 gas at a coastal site in La Jolla, California, we show that nearly 4 × 1011 neutrons per m2 leaked at the Fukushima nuclear power plant before March 20, 2011. A significantly higher activity as measured on March 28 is in accord with neutrons escaping the reactor core and being absorbed by the coolant seawater 35Cl to produce 35S by a (n, p) reaction. Once produced, 35S oxidizes to and and was then transported to Southern California due to the presence of strong prevailing westerly winds at this time. Based on a moving box model, we show that the observed activity enhancement in is compatible with long-range transport of the radiation plume from Fukushima. Our model predicts that , the concentration in the marine boundary layer at Fukushima, was approximately 2 × 105 atoms per m3, which is approximately 365 times above expected natural concentrations. These measurements and model calculations imply that approximately 0.7% of the total radioactive sulfate present at the marine boundary layer at Fukushima reached Southern California as a result of the trans-Pacific transport. PMID:21844372

Phillips-Tikhonov regularization with a priori information for neutron emission tomographic reconstruction on Joint European Torus

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bielecki, J.; Scholz, M.; Drozdowicz, K.

A method of tomographic reconstruction of the neutron emissivity in the poloidal cross section of the Joint European Torus (JET, Culham, UK) tokamak was developed. Due to very limited data set (two projection angles, 19 lines of sight only) provided by the neutron emission profile monitor (KN3 neutron camera), the reconstruction is an ill-posed inverse problem. The aim of this work consists in making a contribution to the development of reliable plasma tomography reconstruction methods that could be routinely used at JET tokamak. The proposed method is based on Phillips-Tikhonov regularization and incorporates a priori knowledge of the shape ofmore » normalized neutron emissivity profile. For the purpose of the optimal selection of the regularization parameters, the shape of normalized neutron emissivity profile is approximated by the shape of normalized electron density profile measured by LIDAR or high resolution Thomson scattering JET diagnostics. In contrast with some previously developed methods of ill-posed plasma tomography reconstruction problem, the developed algorithms do not include any post-processing of the obtained solution and the physical constrains on the solution are imposed during the regularization process. The accuracy of the method is at first evaluated by several tests with synthetic data based on various plasma neutron emissivity models (phantoms). Then, the method is applied to the neutron emissivity reconstruction for JET D plasma discharge #85100. It is demonstrated that this method shows good performance and reliability and it can be routinely used for plasma neutron emissivity reconstruction on JET.« less

Feasibility study of extremity dosemeter based on polyallyldiglycolcarbonate (CR-39) for neutron exposure.

PubMed

Chau, Q; Bruguier, P

2007-01-01

In nuclear facilities, some activities such as reprocessing, recycling and production of bare fuel rods expose the workers to mixed neutron-photon fields. For several workplaces, particularly in glove boxes, some workers expose their hands to mixed fields. The mastery of the photon extremity dosimetry is relatively good, whereas the neutron dosimetry still raises difficulties. In this context, the Institute for Radiological Protection and Nuclear Safety (IRSN) has proposed a study on a passive neutron extremity dosemeter based on chemically etched CR-39 (PADC: polyallyldiglycolcarbonate), named PN-3, already used in routine practice for whole body dosimetry. This dosemeter is a chip of plastic sensitive to recoil protons. The chemical etching process amplifies the size of the impact. The reading system for tracks counting is composed of a microscope, a video camera and an image analyser. This system is combined with the dose evaluation algorithm. The performance of the dosemeter PN-3 has been largely studied and proved by several laboratories in terms of passive individual neutron dosemeter which is used in routine production by different companies. This study focuses on the sensitivity of the extremity dosemeter, as well as its performance in the function of the level of the neutron energy. The dosemeter was exposed to monoenergetic neutron fields in laboratory conditions and to mixed fields in glove boxes at workplaces.

Sodium ion transport mechanisms in antiperovskite electrolytes Na 3OBr and Na 4OI 2: An in Situ neutron diffraction study

DOE Office of Scientific and Technical Information (OSTI.GOV)

Zhu, Jinlong; Wang, Yonggang; Li, Shuai

Na-rich antiperovskites are recently developed solid electrolytes with enhanced sodium ionic conductivity and show promising functionality as a novel solid electrolyte in an all solid-stat battery. In this work, the sodium ionic transport pathways of the parent compound Na 3OBr, as well as the modified layered antiperovskite Na 4OI 2, were studied and compared through temperature dependent neutron diffraction combined with the maximum entropy method. In the cubic Na 3OBr antiperovskite, the nuclear density distribution maps at 500 K indicate that sodium ions ho within and among oxygen octahedra, and Br - ions are not involved in the tetragonal Namore » 4OI 2 antiperovskite, Na ions, which connect octahedra in the ab plane, have the lowest activation energy barrier. In conclusion, the transport of sodium ions along the c axis is assisted by I - ions.« less

Sodium ion transport mechanisms in antiperovskite electrolytes Na 3OBr and Na 4OI 2: An in Situ neutron diffraction study

DOE PAGES

Zhu, Jinlong; Wang, Yonggang; Li, Shuai; ...

2016-06-02

Na-rich antiperovskites are recently developed solid electrolytes with enhanced sodium ionic conductivity and show promising functionality as a novel solid electrolyte in an all solid-stat battery. In this work, the sodium ionic transport pathways of the parent compound Na 3OBr, as well as the modified layered antiperovskite Na 4OI 2, were studied and compared through temperature dependent neutron diffraction combined with the maximum entropy method. In the cubic Na 3OBr antiperovskite, the nuclear density distribution maps at 500 K indicate that sodium ions ho within and among oxygen octahedra, and Br - ions are not involved in the tetragonal Namore » 4OI 2 antiperovskite, Na ions, which connect octahedra in the ab plane, have the lowest activation energy barrier. In conclusion, the transport of sodium ions along the c axis is assisted by I - ions.« less

Broad Energy Range Neutron Spectroscopy using a Liquid Scintillator and a Proportional Counter: Application to a Neutron Spectrum Similar to that from an Improvised Nuclear Device

PubMed Central

Randers-Pehrson, Gerhard; Marino, Stephen A.; Garty, Guy; Harken, Andrew; Brenner, David J.

2015-01-01

A novel neutron irradiation facility at the Radiological Research Accelerator Facility (RARAF) has been developed to mimic the neutron radiation from an Improvised Nuclear Device (IND) at relevant distances (e.g. 1.5 km) from the epicenter. The neutron spectrum of this IND-like neutron irradiator was designed according to estimations of the Hiroshima neutron spectrum at 1.5 km. It is significantly different from a standard reactor fission spectrum, because the spectrum changes as the neutrons are transported through air, and it is dominated by neutron energies from 100 keV up to 9 MeV. To verify such wide energy range neutron spectrum, detailed here is the development of a combined spectroscopy system. Both a liquid scintillator detector and a gas proportional counter were used for the recoil spectra measurements, with the individual response functions estimated from a series of Monte Carlo simulations. These normalized individual response functions were formed into a single response matrix for the unfolding process. Several accelerator-based quasi-monoenergetic neutron source spectra were measured and unfolded to test this spectroscopy system. These reference neutrons were produced from two reactions: T(p,n)3He and D(d,n)3He, generating neutron energies in the range between 0.2 and 8 MeV. The unfolded quasi-monoenergetic neutron spectra indicated that the detection system can provide good neutron spectroscopy results in this energy range. A broad-energy neutron spectrum from the 9Be(d,n) reaction using a 5 MeV deuteron beam, measured at 60 degrees to the incident beam was measured and unfolded with the evaluated response matrix. The unfolded broad neutron spectrum is comparable with published time-of-flight results. Finally, the pair of detectors were used to measure the neutron spectrum generated at the RARAF IND-like neutron facility and a comparison is made to the neutron spectrum of Hiroshima. PMID:26273118

Broad Energy Range Neutron Spectroscopy using a Liquid Scintillator and a Proportional Counter: Application to a Neutron Spectrum Similar to that from an Improvised Nuclear Device.

PubMed

Xu, Yanping; Randers-Pehrson, Gerhard; Marino, Stephen A; Garty, Guy; Harken, Andrew; Brenner, David J

2015-09-11

A novel neutron irradiation facility at the Radiological Research Accelerator Facility (RARAF) has been developed to mimic the neutron radiation from an Improvised Nuclear Device (IND) at relevant distances (e.g. 1.5 km) from the epicenter. The neutron spectrum of this IND-like neutron irradiator was designed according to estimations of the Hiroshima neutron spectrum at 1.5 km. It is significantly different from a standard reactor fission spectrum, because the spectrum changes as the neutrons are transported through air, and it is dominated by neutron energies from 100 keV up to 9 MeV. To verify such wide energy range neutron spectrum, detailed here is the development of a combined spectroscopy system. Both a liquid scintillator detector and a gas proportional counter were used for the recoil spectra measurements, with the individual response functions estimated from a series of Monte Carlo simulations. These normalized individual response functions were formed into a single response matrix for the unfolding process. Several accelerator-based quasi-monoenergetic neutron source spectra were measured and unfolded to test this spectroscopy system. These reference neutrons were produced from two reactions: T(p,n) 3 He and D(d,n) 3 He, generating neutron energies in the range between 0.2 and 8 MeV. The unfolded quasi-monoenergetic neutron spectra indicated that the detection system can provide good neutron spectroscopy results in this energy range. A broad-energy neutron spectrum from the 9 Be(d,n) reaction using a 5 MeV deuteron beam, measured at 60 degrees to the incident beam was measured and unfolded with the evaluated response matrix. The unfolded broad neutron spectrum is comparable with published time-of-flight results. Finally, the pair of detectors were used to measure the neutron spectrum generated at the RARAF IND-like neutron facility and a comparison is made to the neutron spectrum of Hiroshima.

Fuzzy multi objective transportation problem – evolutionary algorithm approach

NASA Astrophysics Data System (ADS)

Karthy, T.; Ganesan, K.

2018-04-01

This paper deals with fuzzy multi objective transportation problem. An fuzzy optimal compromise solution is obtained by using Fuzzy Genetic Algorithm. A numerical example is provided to illustrate the methodology.

Optimization of a mirror-based neutron source using differential evolution algorithm

NASA Astrophysics Data System (ADS)

Yurov, D. V.; Prikhodko, V. V.

2016-12-01

This study is dedicated to the assessment of capabilities of gas-dynamic trap (GDT) and gas-dynamic multiple-mirror trap (GDMT) as potential neutron sources for subcritical hybrids. In mathematical terms the problem of the study has been formulated as determining the global maximum of fusion gain (Q pl), the latter represented as a function of trap parameters. A differential evolution method has been applied to perform the search. Considered in all calculations has been a configuration of the neutron source with 20 m long distance between the mirrors and 100 MW heating power. It is important to mention that the numerical study has also taken into account a number of constraints on plasma characteristics so as to provide physical credibility of searched-for trap configurations. According to the results obtained the traps considered have demonstrated fusion gain up to 0.2, depending on the constraints applied. This enables them to be used either as neutron sources within subcritical reactors for minor actinides incineration or as material-testing facilities.

Monte Carlo simulation of moderator and reflector in coal analyzer based on a D-T neutron generator.

PubMed

Shan, Qing; Chu, Shengnan; Jia, Wenbao

2015-11-01

Coal is one of the most popular fuels in the world. The use of coal not only produces carbon dioxide, but also contributes to the environmental pollution by heavy metals. In prompt gamma-ray neutron activation analysis (PGNAA)-based coal analyzer, the characteristic gamma rays of C and O are mainly induced by fast neutrons, whereas thermal neutrons can be used to induce the characteristic gamma rays of H, Si, and heavy metals. Therefore, appropriate thermal and fast neutrons are beneficial in improving the measurement accuracy of heavy metals, and ensure that the measurement accuracy of main elements meets the requirements of the industry. Once the required yield of the deuterium-tritium (d-T) neutron generator is determined, appropriate thermal and fast neutrons can be obtained by optimizing the neutron source term. In this article, the Monte Carlo N-Particle (MCNP) Transport Code and Evaluated Nuclear Data File (ENDF) database are used to optimize the neutron source term in PGNAA-based coal analyzer, including the material and shape of the moderator and neutron reflector. The optimized targets include two points: (1) the ratio of the thermal to fast neutron is 1:1 and (2) the total neutron flux from the optimized neutron source in the sample increases at least 100% when compared with the initial one. The simulation results show that, the total neutron flux in the sample increases 102%, 102%, 85%, 72%, and 62% with Pb, Bi, Nb, W, and Be reflectors, respectively. Maximum optimization of the targets is achieved when the moderator is a 3-cm-thick lead layer coupled with a 3-cm-thick high-density polyethylene (HDPE) layer, and the neutron reflector is a 27-cm-thick hemispherical lead layer. Copyright © 2015 Elsevier Ltd. All rights reserved.

PHITS simulations of absorbed dose out-of-field and neutron energy spectra for ELEKTA SL25 medical linear accelerator.

PubMed

Puchalska, Monika; Sihver, Lembit

2015-06-21

Monte Carlo (MC) based calculation methods for modeling photon and particle transport, have several potential applications in radiotherapy. An essential requirement for successful radiation therapy is that the discrepancies between dose distributions calculated at the treatment planning stage and those delivered to the patient are minimized. It is also essential to minimize the dose to radiosensitive and critical organs. With MC technique, the dose distributions from both the primary and scattered photons can be calculated. The out-of-field radiation doses are of particular concern when high energy photons are used, since then neutrons are produced both in the accelerator head and inside the patients. Using MC technique, the created photons and particles can be followed and the transport and energy deposition in all the tissues of the patient can be estimated. This is of great importance during pediatric treatments when minimizing the risk for normal healthy tissue, e.g. secondary cancer. The purpose of this work was to evaluate 3D general purpose PHITS MC code efficiency as an alternative approach for photon beam specification. In this study, we developed a model of an ELEKTA SL25 accelerator and used the transport code PHITS for calculating the total absorbed dose and the neutron energy spectra infield and outside the treatment field. This model was validated against measurements performed with bubble detector spectrometers and Boner sphere for 18 MV linacs, including both photons and neutrons. The average absolute difference between the calculated and measured absorbed dose for the out-of-field region was around 11%. Taking into account a simplification for simulated geometry, which does not include any potential scattering materials around, the obtained result is very satisfactorily. A good agreement between the simulated and measured neutron energy spectra was observed while comparing to data found in the literature.

PHITS simulations of absorbed dose out-of-field and neutron energy spectra for ELEKTA SL25 medical linear accelerator

NASA Astrophysics Data System (ADS)

Puchalska, Monika; Sihver, Lembit

2015-06-01

Monte Carlo (MC) based calculation methods for modeling photon and particle transport, have several potential applications in radiotherapy. An essential requirement for successful radiation therapy is that the discrepancies between dose distributions calculated at the treatment planning stage and those delivered to the patient are minimized. It is also essential to minimize the dose to radiosensitive and critical organs. With MC technique, the dose distributions from both the primary and scattered photons can be calculated. The out-of-field radiation doses are of particular concern when high energy photons are used, since then neutrons are produced both in the accelerator head and inside the patients. Using MC technique, the created photons and particles can be followed and the transport and energy deposition in all the tissues of the patient can be estimated. This is of great importance during pediatric treatments when minimizing the risk for normal healthy tissue, e.g. secondary cancer. The purpose of this work was to evaluate 3D general purpose PHITS MC code efficiency as an alternative approach for photon beam specification. In this study, we developed a model of an ELEKTA SL25 accelerator and used the transport code PHITS for calculating the total absorbed dose and the neutron energy spectra infield and outside the treatment field. This model was validated against measurements performed with bubble detector spectrometers and Boner sphere for 18 MV linacs, including both photons and neutrons. The average absolute difference between the calculated and measured absorbed dose for the out-of-field region was around 11%. Taking into account a simplification for simulated geometry, which does not include any potential scattering materials around, the obtained result is very satisfactorily. A good agreement between the simulated and measured neutron energy spectra was observed while comparing to data found in the literature.

The coupling of the neutron transport application RATTLESNAKE to the nuclear fuels performance application BISON under the MOOSE framework

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gleicher, Frederick N.; Williamson, Richard L.; Ortensi, Javier

The MOOSE neutron transport application RATTLESNAKE was coupled to the fuels performance application BISON to provide a higher fidelity tool for fuel performance simulation. This project is motivated by the desire to couple a high fidelity core analysis program (based on the self-adjoint angular flux equations) to a high fidelity fuel performance program, both of which can simulate on unstructured meshes. RATTLESNAKE solves self-adjoint angular flux transport equation and provides a sub-pin level resolution of the multigroup neutron flux with resonance treatment during burnup or a fast transient. BISON solves the coupled thermomechanical equations for the fuel on a sub-millimetermore » scale. Both applications are able to solve their respective systems on aligned and unaligned unstructured finite element meshes. The power density and local burnup was transferred from RATTLESNAKE to BISON with the MOOSE Multiapp transfer system. Multiple depletion cases were run with one-way data transfer from RATTLESNAKE to BISON. The eigenvalues are shown to agree well with values obtained from the lattice physics code DRAGON. The one-way data transfer of power density is shown to agree with the power density obtained from an internal Lassman-style model in BISON.« less

Fast neutron background characterization with the Radiological Multi-sensor Analysis Platform (RadMAP)

DOE PAGES

Davis, John R.; Brubaker, Erik; Vetter, Kai

2017-03-29

In an effort to characterize the fast neutron radiation background, 16 EJ-309 liquid scintillator cells were installed in the Radiological Multi-sensor Analysis Platform (RadMAP) to collect data in the San Francisco Bay Area. Each fast neutron event was associated with specific weather metrics (pressure, temperature, absolute humidity) and GPS coordinates. Furthermore, the expected exponential dependence of the fast neutron count rate on atmospheric pressure was demonstrated and event rates were subsequently adjusted given the measured pressure at the time of detection. Pressure adjusted data was also used to investigate the influence of other environmental conditions on the neutron background rate.more » Using National Oceanic and Atmospheric Administration (NOAA) coastal area lidar data, an algorithm was implemented to approximate sky-view factors (the total fraction of visible sky) for points along RadMAPs route. In the three areas we analyzed, San Francisco, Downtown Oakland, and Berkeley, all demonstrated a suppression in the background rate of over 50% for the range of sky-view factors measured. This effect, which is due to the shielding of cosmic-ray produced neutrons by surrounding buildings, was comparable to the pressure influence which yielded a 32% suppression in the count rate over the range of pressures measured.« less

A search for rapidly modulated emission in bright X-ray sources using the HEAO A-1 data base

NASA Technical Reports Server (NTRS)

Fairbank, William M.

1987-01-01

A search was performed in the HEAO A-1 Data Base (located at the Naval Research Laboratory in Washington, D.C.) for evidence of rapidly-rotating neutron stars that could be sources of coherent gravitational radiation. A new data analysis algorithm, which was developed, is described. The algorithm was applied to data from observations of Cyg X-2, Cyg X-3, and 1820-30. Upper limits on pulse fraction were derived and reported.

Simulation of Automatic Incidents Detection Algorithm on the Transport Network

ERIC Educational Resources Information Center

Nikolaev, Andrey B.; Sapego, Yuliya S.; Jakubovich, Anatolij N.; Berner, Leonid I.; Ivakhnenko, Andrey M.

2016-01-01

Management of traffic incident is a functional part of the whole approach to solving traffic problems in the framework of intelligent transport systems. Development of an effective process of traffic incident management is an important part of the transport system. In this research, it's suggested algorithm based on fuzzy logic to detect traffic…

Sensitivity of CO2 Simulation in a GCM to the Convective Transport Algorithms

NASA Technical Reports Server (NTRS)

Zhu, Z.; Pawson, S.; Collatz, G. J.; Gregg, W. W.; Kawa, S. R.; Baker, D.; Ott, L.

2014-01-01

Convection plays an important role in the transport of heat, moisture and trace gases. In this study, we simulated CO2 concentrations with an atmospheric general circulation model (GCM). Three different convective transport algorithms were used. One is a modified Arakawa-Shubert scheme that was native to the GCM; two others used in two off-line chemical transport models (CTMs) were added to the GCM here for comparison purposes. Advanced CO2 surfaced fluxes were used for the simulations. The results were compared to a large quantity of CO2 observation data. We find that the simulation results are sensitive to the convective transport algorithms. Overall, the three simulations are quite realistic and similar to each other in the remote marine regions, but are significantly different in some land regions with strong fluxes such as Amazon and Siberia during the convection seasons. Large biases against CO2 measurements are found in these regions in the control run, which uses the original GCM. The simulation with the simple diffusive algorithm is better. The difference of the two simulations is related to the very different convective transport speed.

New Multi-group Transport Neutronics (PHISICS) Capabilities for RELAP5-3D and its Application to Phase I of the OECD/NEA MHTGR-350 MW Benchmark

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gerhard Strydom; Cristian Rabiti; Andrea Alfonsi

2012-10-01

PHISICS is a neutronics code system currently under development at the Idaho National Laboratory (INL). Its goal is to provide state of the art simulation capability to reactor designers. The different modules for PHISICS currently under development are a nodal and semi-structured transport core solver (INSTANT), a depletion module (MRTAU) and a cross section interpolation (MIXER) module. The INSTANT module is the most developed of the mentioned above. Basic functionalities are ready to use, but the code is still in continuous development to extend its capabilities. This paper reports on the effort of coupling the nodal kinetics code package PHISICSmore » (INSTANT/MRTAU/MIXER) to the thermal hydraulics system code RELAP5-3D, to enable full core and system modeling. This will enable the possibility to model coupled (thermal-hydraulics and neutronics) problems with more options for 3D neutron kinetics, compared to the existing diffusion theory neutron kinetics module in RELAP5-3D (NESTLE). In the second part of the paper, an overview of the OECD/NEA MHTGR-350 MW benchmark is given. This benchmark has been approved by the OECD, and is based on the General Atomics 350 MW Modular High Temperature Gas Reactor (MHTGR) design. The benchmark includes coupled neutronics thermal hydraulics exercises that require more capabilities than RELAP5-3D with NESTLE offers. Therefore, the MHTGR benchmark makes extensive use of the new PHISICS/RELAP5-3D coupling capabilities. The paper presents the preliminary results of the three steady state exercises specified in Phase I of the benchmark using PHISICS/RELAP5-3D.« less

Geant4 Modifications for Accurate Fission Simulations

NASA Astrophysics Data System (ADS)

Tan, Jiawei; Bendahan, Joseph

Monte Carlo is one of the methods to simulate the generation and transport of radiation through matter. The most widely used radiation simulation codes are MCNP and Geant4. The simulation of fission production and transport by MCNP has been thoroughly benchmarked. There is an increasing number of users that prefer using Geant4 due to the flexibility of adding features. However, it has been found that Geant4 does not have the proper fission-production cross sections and does not produce the correct fission products. To achieve accurate results for studies in fissionable material applications, Geant4 was modified to correct these inaccuracies and to add new capabilities. The fission model developed by the Lawrence Livermore National Laboratory was integrated into the neutron-fission modeling package. The photofission simulation capability was enabled using the same neutron-fission library under the assumption that nuclei fission in the same way, independent of the excitation source. The modified fission code provides the correct multiplicity of prompt neutrons and gamma rays, and produces delayed gamma rays and neutrons with time and energy dependencies that are consistent with ENDF/B-VII. The delayed neutrons are now directly produced by a custom package that bypasses the fragment cascade model. The modifications were made for U-235, U-238 and Pu-239 isotopes; however, the new framework allows adding new isotopes easily. The SLAC nuclear data library is used for simulation of isotopes with an atomic number above 92 because it is not available in Geant4. Results of the modified Geant4.10.1 package of neutron-fission and photofission for prompt and delayed radiation are compared with ENDFB-VII and with results produced with the original package.

Rates for neutron-capture reactions on tungsten isotopes in iron meteorites. [Abstract only

NASA Technical Reports Server (NTRS)

Masarik, J.; Reedy, R. C.

1994-01-01

High-precision W isotopic analyses by Harper and Jacobsen indicate the W-182/W-183 ratio in the Toluca iron meteorite is shifted by -(3.0 +/- 0.9) x 10(exp -4) relative to a terrestrial standard. Possible causes of this shift are neutron-capture reactions on W during Toluca's approximately 600-Ma exposure to cosmic ray particles or radiogenic growth of W-182 from 9-Ma Hf-182 in the silicate portion of the Earth after removal of W to the Earth's core. Calculations for the rates of neutron-capture reactions on W isotopes were done to study the first possibility. The LAHET Code System (LCS) which consists of the Los Alamos High Energy Transport (LAHET) code and the Monte Carlo N-Particle(MCNP) transport code was used to numerically simulate the irradiation of the Toluca iron meteorite by galactic-cosmic-ray (GCR) particles and to calculate the rates of W(n, gamma) reactions. Toluca was modeled as a 3.9-m-radius sphere with the composition of a typical IA iron meteorite. The incident GCR protons and their interactions were modeled with LAHET, which also handled the interactions of neutrons with energies above 20 MeV. The rates for the capture of neutrons by W-182, W-183, and W-186 were calculated using the detailed library of (n, gamma) cross sections in MCNP. For this study of the possible effect of W(n, gamma) reactions on W isotope systematics, we consider the peak rates. The calculated maximum change in the normalized W-182/W-183 ratio due to neutron-capture reactions cannot account for more than 25% of the mass 182 deficit observed in Toluca W.

Compact, inexpensive, epithermal neutron source for BNCT

DOE Office of Scientific and Technical Information (OSTI.GOV)

Swenson, D. A.

1999-06-10

A new rf-focused linac structure, designed specifically to increase the acceleration efficiency and reduce the cost of linac structures in the few-MeV range, may win the role as the optimum accelerator-based epithermal neutron source for the BNCT application. This new linac structure resembles a drift tube linac (DTL) with radio frequency quadrupole (RFQ) focusing incorporated into each 'drift tube,' hence the name R lowbar f F lowbar ocused D lowbar TL, or RFD. It promises superior acceleration properties, focusing properties, and CW capabilities. We have a proposal under consideration for the development of an epithermal neutron source, based on themore » 2.5-MeV RFD linac system with an average current of 10 mA, having the following components: an ion source, a short low-energy transport system, a short RFQ linac section, an RFD linac section, an rf power system, a high-energy beam transport system, a proton beam target, and a neutron beam moderator system. We propose to develop a solid lithium target for this application in the form of a thin lithium layer on the inner surface of a truncated aluminum cone, cooled by the heavy water moderator, where the proton beam is expanded to a diameter of 3 cm and scanned along a circular path, striking the lithium layer at the cone's half-angle of 30 degrees. We propose to develop a moderator assembly designed to transmit a large fraction of the source neutrons from the target to the patient treatment port, while shifting the neutron energies to an appropriate epithermal energy spectrum and minimizing the gamma-ray dose. The status of this proposal and these plans are presented.« less

Counting neutrons with a commercial S-CMOS camera

NASA Astrophysics Data System (ADS)

Patrick, Van Esch; Paolo, Mutti; Emilio, Ruiz-Martinez; Estefania, Abad Garcia; Marita, Mosconi; Jon, Ortega

2018-01-01

It is possible to detect individual flashes from thermal neutron impacts in a ZnS scintillator using a CMOS camera looking at the scintillator screen, and off line image processing. Some preliminary results indicated that the efficiency of recognition could be improved by optimizing the light collection and the image processing. We will report on this ongoing work which is a result from the collaboration between ESS Bilbao and the ILL. The main progress to be reported is situated on the level of the on-line treatment of the imaging data. If this technology is to work on a genuine scientific instrument, it is necessary that all the processing happens on line, to avoid the accumulation of large amounts of image data to be analyzed off line. An FPGA-based real-time full-deca mode VME-compatible CameraLink board has been developed at the SCI of the ILL, which is able to manage the data flow from the camera and convert it in a reasonable "neutron impact" data flow like from a usual neutron counting detector. The main challenge of the endeavor is the optical light collection from the scintillator. While the light yield of a ZnS scintillator is a priori rather important, the amount of light collected with a photographic objective is small. Different scintillators and different light collection techniques have been experimented with and results will be shown for different setups improving upon the light recuperation on the camera sensor. Improvements on the algorithm side will also be presented. The algorithms have to be at the same time efficient in their recognition of neutron signals, in their rejection of noise signals (internal and external to the camera) but also have to be simple enough to be easily implemented in the FPGA. The path from the idea of detecting individual neutron impacts with a CMOS camera to a practical working instrument detector is challenging, and in this paper we will give an overview of the part of the road that has already been walked.

Scalable Domain Decomposed Monte Carlo Particle Transport

NASA Astrophysics Data System (ADS)

O'Brien, Matthew Joseph

In this dissertation, we present the parallel algorithms necessary to run domain decomposed Monte Carlo particle transport on large numbers of processors (millions of processors). Previous algorithms were not scalable, and the parallel overhead became more computationally costly than the numerical simulation. The main algorithms we consider are: • Domain decomposition of constructive solid geometry: enables extremely large calculations in which the background geometry is too large to fit in the memory of a single computational node. • Load Balancing: keeps the workload per processor as even as possible so the calculation runs efficiently. • Global Particle Find: if particles are on the wrong processor, globally resolve their locations to the correct processor based on particle coordinate and background domain. • Visualizing constructive solid geometry, sourcing particles, deciding that particle streaming communication is completed and spatial redecomposition. These algorithms are some of the most important parallel algorithms required for domain decomposed Monte Carlo particle transport. We demonstrate that our previous algorithms were not scalable, prove that our new algorithms are scalable, and run some of the algorithms up to 2 million MPI processes on the Sequoia supercomputer.

Estimating the solute transport parameters of the spatial fractional advection-dispersion equation using Bees Algorithm

NASA Astrophysics Data System (ADS)

Mehdinejadiani, Behrouz

2017-08-01

This study represents the first attempt to estimate the solute transport parameters of the spatial fractional advection-dispersion equation using Bees Algorithm. The numerical studies as well as the experimental studies were performed to certify the integrity of Bees Algorithm. The experimental ones were conducted in a sandbox for homogeneous and heterogeneous soils. A detailed comparative study was carried out between the results obtained from Bees Algorithm and those from Genetic Algorithm and LSQNONLIN routines in FracFit toolbox. The results indicated that, in general, the Bees Algorithm much more accurately appraised the sFADE parameters in comparison with Genetic Algorithm and LSQNONLIN, especially in the heterogeneous soil and for α values near to 1 in the numerical study. Also, the results obtained from Bees Algorithm were more reliable than those from Genetic Algorithm. The Bees Algorithm showed the relative similar performances for all cases, while the Genetic Algorithm and the LSQNONLIN yielded different performances for various cases. The performance of LSQNONLIN strongly depends on the initial guess values so that, compared to the Genetic Algorithm, it can more accurately estimate the sFADE parameters by taking into consideration the suitable initial guess values. To sum up, the Bees Algorithm was found to be very simple, robust and accurate approach to estimate the transport parameters of the spatial fractional advection-dispersion equation.

Estimating the solute transport parameters of the spatial fractional advection-dispersion equation using Bees Algorithm.

PubMed

Mehdinejadiani, Behrouz

2017-08-01

This study represents the first attempt to estimate the solute transport parameters of the spatial fractional advection-dispersion equation using Bees Algorithm. The numerical studies as well as the experimental studies were performed to certify the integrity of Bees Algorithm. The experimental ones were conducted in a sandbox for homogeneous and heterogeneous soils. A detailed comparative study was carried out between the results obtained from Bees Algorithm and those from Genetic Algorithm and LSQNONLIN routines in FracFit toolbox. The results indicated that, in general, the Bees Algorithm much more accurately appraised the sFADE parameters in comparison with Genetic Algorithm and LSQNONLIN, especially in the heterogeneous soil and for α values near to 1 in the numerical study. Also, the results obtained from Bees Algorithm were more reliable than those from Genetic Algorithm. The Bees Algorithm showed the relative similar performances for all cases, while the Genetic Algorithm and the LSQNONLIN yielded different performances for various cases. The performance of LSQNONLIN strongly depends on the initial guess values so that, compared to the Genetic Algorithm, it can more accurately estimate the sFADE parameters by taking into consideration the suitable initial guess values. To sum up, the Bees Algorithm was found to be very simple, robust and accurate approach to estimate the transport parameters of the spatial fractional advection-dispersion equation. Copyright © 2017 Elsevier B.V. All rights reserved.

Measuring neutron spectra in radiotherapy using the nested neutron spectrometer.

PubMed

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

2015-11-01

Out-of-field neutron doses resulting from photonuclear interactions in the head of a linear accelerator pose an iatrogenic risk to patients and an occupational risk to personnel during radiotherapy. To quantify neutron production, in-room measurements have traditionally been carried out using Bonner sphere systems (BSS) with activation foils and TLDs. In this work, a recently developed active detector, the nested neutron spectrometer (NNS), was tested in radiotherapy bunkers. The NNS is designed for easy handling and is more practical than the traditional BSS. Operated in current-mode, the problem of pulse pileup due to high dose-rates is overcome by measuring current, similar to an ionization chamber. In a bunker housing a Varian Clinac 21EX, the performance of the NNS was evaluated in terms of reproducibility, linearity, and dose-rate effects. Using a custom maximum-likelihood expectation-maximization algorithm, measured neutron spectra at various locations inside the bunker were then compared to Monte Carlo simulations of an identical setup. In terms of dose, neutron ambient dose equivalents were calculated from the measured spectra and compared to bubble detector neutron dose equivalent measurements. The NNS-measured spectra for neutrons at various locations in a treatment room were found to be consistent with expectations for both relative shape and absolute magnitude. Neutron fluence-rate decreased with distance from the source and the shape of the spectrum changed from a dominant fast neutron peak near the Linac head to a dominant thermal neutron peak in the moderating conditions of the maze. Monte Carlo data and NNS-measured spectra agreed within 30% at all locations except in the maze where the deviation was a maximum of 40%. Neutron ambient dose equivalents calculated from the authors' measured spectra were consistent (one standard deviation) with bubble detector measurements in the treatment room. The NNS may be used to reliably measure the neutron spectrum of a radiotherapy beam in less than 1 h, including setup and data unfolding. This work thus represents a new, fast, and practical method for neutron spectral measurements in radiotherapy.

Effect of high fluence neutron irradiation on transport properties of thermoelectrics

NASA Astrophysics Data System (ADS)

Wang, H.; Leonard, K. J.

2017-07-01

Thermoelectric materials were subjected to high fluence neutron irradiation in order to understand the effect of radiation damage on transport properties. This study is relevant to the NASA Radioisotope Thermoelectric Generator (RTG) program in which thermoelectric elements are exposed to radiation over a long period of time in space missions. Selected n-type and p-type bismuth telluride materials were irradiated at the High Flux Isotope Reactor with a neutron fluence of 1.3 × 1018 n/cm2 (E > 0.1 MeV). The increase in the Seebeck coefficient in the n-type material was partially off-set by an increase in electrical resistivity, making the power factor higher at lower temperatures. For the p-type materials, although the Seebeck coefficient was not affected by irradiation, electrical resistivity decreased slightly. The figure of merit, zT, showed a clear drop in the 300-400 K range for the p-type material and an increase for the n-type material. Considering that the p-type and n-type materials are connected in series in a module, the overall irradiation damages at the device level were limited. These results, at neutron fluences exceeding a typical space mission, are significant to ensure that the radiation damage to thermoelectrics does not affect the performance of RTGs.

Neutron Diffraction and Electrical Transport Studies on Magnetic Transition in Terbium at High Pressures and Low Temperatures

NASA Astrophysics Data System (ADS)

Thomas, Sarah; Montgomery, Jeffrey; Tsoi, Georgiy; Vohra, Yogesh; Weir, Samuel; Tulk, Christopher; Moreira Dos Santos, Antonio

2013-06-01

Neutron diffraction and electrical transport measurements have been carried out on the heavy rare earth metal terbium at high pressures and low temperatures in order to elucidate its transition from a helical antiferromagnetic to a ferromagnetic ordered phase as a function of pressure. The electrical resistance measurements using designer diamonds show a change in slope as the temperature is lowered through the ferromagnetic Curie temperature. The temperature of the ferromagnetic transition decreases at a rate of -16.7 K/GPa till 3.6 GPa, where terbium undergoes a structural transition from hexagonal close packed (hcp) to an α-Sm phase. Above this pressure, the electrical resistance measurements no longer exhibit a change in slope. In order to confirm the change in magnetic phase suggested by the electrical resistance measurements, neutron diffraction measurements were conducted at the SNAP beamline at the Oak Ridge National Laboratory. Measurements were made at pressures to 5.3 GPa and temperatures as low as 90 K. An abrupt increase in peak intensity in the neutron diffraction spectra signaled the onset of magnetic order below the Curie temperature. A magnetic phase diagram of rare earth metal terbium will be presented to 5.3 GPa and 90 K based on these studies.

Phase diagram and neutron spin resonance of superconducting NaFe 1 - x Cu x As

DOE PAGES

Tan, Guotai; Song, Yu; Zhang, Rui; ...

2017-02-03

In this paper, we use transport and neutron scattering to study the electronic phase diagram and spin excitations of NaFe 1-xCu xAs single crystals. Similar to Co- and Ni-doped NaFeAs, a bulk superconducting phase appears near x≈2% with the suppression of stripe-type magnetic order in NaFeAs. Upon further increasing Cu concentration the system becomes insulating, culminating in an antiferromagnetically ordered insulating phase near x≈50%. Using transport measurements, we demonstrate that the resistivity in NaFe 1-xCu xAs exhibits non-Fermi-liquid behavior near x≈1.8%. Our inelastic neutron scattering experiments reveal a single neutron spin resonance mode exhibiting weak dispersion along c axis inmore » NaFe 0.98Cu 0.02As. The resonance is high in energy relative to the superconducting transition temperature T c but weak in intensity, likely resulting from impurity effects. These results are similar to other iron pnictides superconductors despite that the superconducting phase in NaFe 1-xCu xAs is continuously connected to an antiferromagnetically ordered insulating phase near x≈50% with significant electronic correlations. Finally, therefore, electron correlations is an important ingredient of superconductivity in NaFe 1-xCu xAs and other iron pnictides.« less

Comparing genetic algorithm and particle swarm optimization for solving capacitated vehicle routing problem

NASA Astrophysics Data System (ADS)

Iswari, T.; Asih, A. M. S.

2018-04-01

In the logistics system, transportation plays an important role to connect every element in the supply chain, but it can produces the greatest cost. Therefore, it is important to make the transportation costs as minimum as possible. Reducing the transportation cost can be done in several ways. One of the ways to minimizing the transportation cost is by optimizing the routing of its vehicles. It refers to Vehicle Routing Problem (VRP). The most common type of VRP is Capacitated Vehicle Routing Problem (CVRP). In CVRP, the vehicles have their own capacity and the total demands from the customer should not exceed the capacity of the vehicle. CVRP belongs to the class of NP-hard problems. These NP-hard problems make it more complex to solve such that exact algorithms become highly time-consuming with the increases in problem sizes. Thus, for large-scale problem instances, as typically found in industrial applications, finding an optimal solution is not practicable. Therefore, this paper uses two kinds of metaheuristics approach to solving CVRP. Those are Genetic Algorithm and Particle Swarm Optimization. This paper compares the results of both algorithms and see the performance of each algorithm. The results show that both algorithms perform well in solving CVRP but still needs to be improved. From algorithm testing and numerical example, Genetic Algorithm yields a better solution than Particle Swarm Optimization in total distance travelled.

a Dosimetry Assessment for the Core Restraint of AN Advanced Gas Cooled Reactor

NASA Astrophysics Data System (ADS)

Thornton, D. A.; Allen, D. A.; Tyrrell, R. J.; Meese, T. C.; Huggon, A. P.; Whiley, G. S.; Mossop, J. R.

2009-08-01

This paper describes calculations of neutron damage rates within the core restraint structures of Advanced Gas Cooled Reactors (AGRs). Using advanced features of the Monte Carlo radiation transport code MCBEND, and neutron source data from core follow calculations performed with the reactor physics code PANTHER, a detailed model of the reactor cores of two of British Energy's AGR power plants has been developed for this purpose. Because there are no relevant neutron fluence measurements directly supporting this assessment, results of benchmark comparisons and successful validation of MCBEND for Magnox reactors have been used to estimate systematic and random uncertainties on the predictions. In particular, it has been necessary to address the known under-prediction of lower energy fast neutron responses associated with the penetration of large thicknesses of graphite.

Anomalous hydrodynamics kicks neutron stars

DOE PAGES

Kaminski, Matthias; Uhlemann, Christoph F.; Bleicher, Marcus; ...

2016-06-28

Observations show that, at the beginning of their existence, neutron stars are accelerated briskly to velocities of up to a thousand kilometers per second. We argue that this remarkable effect can be explained as a manifestation of quantum anomalies on astrophysical scales. To theoretically describe the early stage in the life of neutron stars we use hydrodynamics as a systematic effective-field-theory framework. Within this framework, anomalies of the Standard Model of particle physics as underlying microscopic theory imply the presence of a particular set of transport terms, whose form is completely fixed by theoretical consistency. Here, the resulting chiral transportmore » effects in proto-neutron stars enhance neutrino emission along the internal magnetic field, and the recoil can explain the order of magnitude of the observed kick velocities.« less

Monte-Carlo Application for Nondestructive Nuclear Waste Analysis

NASA Astrophysics Data System (ADS)

Carasco, C.; Engels, R.; Frank, M.; Furletov, S.; Furletova, J.; Genreith, C.; Havenith, A.; Kemmerling, G.; Kettler, J.; Krings, T.; Ma, J.-L.; Mauerhofer, E.; Neike, D.; Payan, E.; Perot, B.; Rossbach, M.; Schitthelm, O.; Schumann, M.; Vasquez, R.

2014-06-01

Radioactive waste has to undergo a process of quality checking in order to check its conformance with national regulations prior to its transport, intermediate storage and final disposal. Within the quality checking of radioactive waste packages non-destructive assays are required to characterize their radio-toxic and chemo-toxic contents. The Institute of Energy and Climate Research - Nuclear Waste Management and Reactor Safety of the Forschungszentrum Jülich develops in the framework of cooperation nondestructive analytical techniques for the routine characterization of radioactive waste packages at industrial-scale. During the phase of research and development Monte Carlo techniques are used to simulate the transport of particle, especially photons, electrons and neutrons, through matter and to obtain the response of detection systems. The radiological characterization of low and intermediate level radioactive waste drums is performed by segmented γ-scanning (SGS). To precisely and accurately reconstruct the isotope specific activity content in waste drums by SGS measurement, an innovative method called SGSreco was developed. The Geant4 code was used to simulate the response of the collimated detection system for waste drums with different activity and matrix configurations. These simulations allow a far more detailed optimization, validation and benchmark of SGSreco, since the construction of test drums covering a broad range of activity and matrix properties is time consuming and cost intensive. The MEDINA (Multi Element Detection based on Instrumental Neutron Activation) test facility was developed to identify and quantify non-radioactive elements and substances in radioactive waste drums. MEDINA is based on prompt and delayed gamma neutron activation analysis (P&DGNAA) using a 14 MeV neutron generator. MCNP simulations were carried out to study the response of the MEDINA facility in terms of gamma spectra, time dependence of the neutron energy spectrum, neutron flux distribution. The validation of the measurements simulations with Mont-Carlo transport codes for the design, optimization and data analysis of further P&DGNAA facilities is performed in collaboration with LMN CEA Cadarache. The performance of the prompt gamma neutron activation analysis (PGNAA) for the nondestructive determination of actinides in small samples is investigated. The quantitative determination of actinides relies on the precise knowledge of partial neutron capture cross sections. Up to today these cross sections are not very accurate for analytical purpose. The goal of the TANDEM (Trans-uranium Actinides' Nuclear Data - Evaluation and Measurement) Collaboration is the evaluation of these cross sections. Cross sections are measured using prompt gamma activation analysis facilities in Budapest and Munich. Geant4 is used to optimally design the detection system with Compton suppression. Furthermore, for the evaluation of the cross sections it is strongly needed to correct the results to the self-attenuation of the prompt gammas within the sample. In the framework of cooperation RWTH Aachen University, Forschungszentrum Jülich and the Siemens AG will study the feasibility of a compact Neutron Imaging System for Radioactive waste Analysis (NISRA). The system is based on a 14 MeV neutron source and an advanced detector system (a-Si flat panel) linked to an exclusive converter/scintillator for fast neutrons. For shielding and radioprotection studies the codes MCNPX and Geant4 were used. The two codes were benchmarked in processing time and accuracy in the neutron and gamma fluxes. Also the detector response was simulated with Geant4 to optimize components of the system.

Dose Calibration of the ISS-RAD Fast Neutron Detector

NASA Technical Reports Server (NTRS)

Zeitlin, C.

2015-01-01

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

Reducing the spatial resolution range of neutron radiographs cast by thick objects

NASA Astrophysics Data System (ADS)

Almeida, G. L.; Silvani, M. I.; Souza, E. S.; Lopes, R. T.

2017-11-01

The quality of a neutron radiograph is strongly dependent upon the features of the acquisition system. Most of them, such as detector resolution, electronic noise and statistical fluctuation can hardly be improved. Yet, a main parameter ruling the image spatial resolution, namely the L/D ratio of the system can be increased simply by lengthening the source-detector clearance. Such an option eventually may not be feasible due to neutron flux decreasing or engineering constraints. Under this circumstance, a radiograph improvement is only possible by some kind of after-acquisition procedure capable to retrieve, at least partially, the information concealed by the degradation process. Since the spoiling agent tied to the L/D has a systematic character, its impact can be reduced by an unfolding procedure such as Richardson-Lucy algorithm. However, that agent should be fully characterized and furnished to the algorithm as a Point Spread Function - PSF unfolding function. A main drawback of unfolding algorithms like Richardson-Lucy is that the PSF should be fixed, i.e., it assumes a certain constant image spatial resolution, rather than a variable one as actually occurs for thick objects. This work presents a methodology to minimize this difficulty by making all planes of the inspected object to cast a resolution within the shorter gap comprised between the object central plane and the detector. The image can then be unfolded with a lower resolution within a tighter range, yielding a better quality. The process is performed with two radiographs, where one of them is acquired with the object turned by 180° on its vertical axis with regard to the other. After a mirroring of one of them about its vertical axis, the images are added. As the resolution increases linearly with the object-detector gap, it would remain always lower than that of the central one. Therefore, the overall resolution of the composite radiograph is enhanced. A further improvement can then be achieved through an efficient unfolding since the object has been virtually shrunk along the neutron path.

Self characterization of a coded aperture array for neutron source imaging

DOE PAGES

Volegov, P. L.; Danly, C. R.; Fittinghoff, D. N.; ...

2014-12-15

The neutron imaging system at the National Ignition Facility (NIF) is an important diagnostic tool for measuring the two-dimensional size and shape of the neutrons produced in the burning DT plasma during the stagnation stage of ICF implosions. Since the neutron source is small (~100 μm) and neutrons are deeply penetrating (>3 cm) in all materials, the apertures used to achieve the desired 10-μm resolution are 20-cm long, triangular tapers machined in gold foils. These gold foils are stacked to form an array of 20 apertures for pinhole imaging and three apertures for penumbral imaging. These apertures must be preciselymore » aligned to accurately place the field of view of each aperture at the design location, or the location of the field of view for each aperture must be measured. In this paper we present a new technique that has been developed for the measurement and characterization of the precise location of each aperture in the array. We present the detailed algorithms used for this characterization and the results of reconstructed sources from inertial confinement fusion implosion experiments at NIF.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Davis, John R.; Brubaker, Erik; Vetter, Kai

In an effort to characterize the fast neutron radiation background, 16 EJ-309 liquid scintillator cells were installed in the Radiological Multi-sensor Analysis Platform (RadMAP) to collect data in the San Francisco Bay Area. Each fast neutron event was associated with specific weather metrics (pressure, temperature, absolute humidity) and GPS coordinates. Furthermore, the expected exponential dependence of the fast neutron count rate on atmospheric pressure was demonstrated and event rates were subsequently adjusted given the measured pressure at the time of detection. Pressure adjusted data was also used to investigate the influence of other environmental conditions on the neutron background rate.more » Using National Oceanic and Atmospheric Administration (NOAA) coastal area lidar data, an algorithm was implemented to approximate sky-view factors (the total fraction of visible sky) for points along RadMAPs route. In the three areas we analyzed, San Francisco, Downtown Oakland, and Berkeley, all demonstrated a suppression in the background rate of over 50% for the range of sky-view factors measured. This effect, which is due to the shielding of cosmic-ray produced neutrons by surrounding buildings, was comparable to the pressure influence which yielded a 32% suppression in the count rate over the range of pressures measured.« less

Neutron-neutron angular correlations in spontaneous fission of 252Cf and 240Pu

NASA Astrophysics Data System (ADS)

Verbeke, J. M.; Nakae, L. F.; Vogt, R.

2018-04-01

Background: Angular anisotropy has been observed between prompt neutrons emitted during the fission process. Such an anisotropy arises because the emitted neutrons are boosted along the direction of the parent fragment. Purpose: To measure the neutron-neutron angular correlations from the spontaneous fission of 252Cf and 240Pu oxide samples using a liquid scintillator array capable of pulse-shape discrimination. To compare these correlations to simulations combining the Monte Carlo radiation transport code MCNPX with the fission event generator FREYA. Method: Two different analysis methods were used to study the neutron-neutron correlations with varying energy thresholds. The first is based on setting a light output threshold while the second imposes a time-of-flight cutoff. The second method has the advantage of being truly detector independent. Results: The neutron-neutron correlation modeled by FREYA depends strongly on the sharing of the excitation energy between the two fragments. The measured asymmetry enabled us to adjust the FREYA parameter x in 240Pu, which controls the energy partition between the fragments and is so far inaccessible in other measurements. The 240Pu data in this analysis was the first available to quantify the energy partition for this isotope. The agreement between data and simulation is overall very good for 252Cf(sf ) and 240Pu(sf ) . Conclusions: The asymmetry in the measured neutron-neutron angular distributions can be predicted by FREYA. The shape of the correlation function depends on how the excitation energy is partitioned between the two fission fragments. Experimental data suggest that the lighter fragment is disproportionately excited.

Neutron Radiography of Fluid Flow for Geothermal Energy Research

NASA Astrophysics Data System (ADS)

Bingham, P.; Polsky, Y.; Anovitz, L.; Carmichael, J.; Bilheux, H.; Jacobsen, D.; Hussey, D.

Enhanced geothermal systems seek to expand the potential for geothermal energy by engineering heat exchange systems within the earth. A neutron radiography imaging method has been developed for the study of fluid flow through rock under environmental conditions found in enhanced geothermal energy systems. For this method, a pressure vessel suitable for neutron radiography was designed and fabricated, modifications to imaging instrument setups were tested, multiple contrast agents were tested, and algorithms developed for tracking of flow. The method has shown success for tracking of single phase flow through a manufactured crack in a 3.81 cm (1.5 inch) diameter core within a pressure vessel capable of confinement up to 69 MPa (10,000 psi) using a particle tracking approach with bubbles of fluorocarbon-based fluid as the ;particles; and imaging with 10 ms exposures.

Signatures of cosmic-ray interactions on the solar surface

NASA Technical Reports Server (NTRS)

Seckel, D.; Stanev, Todor; Gaisser, T. K.

1991-01-01

The fluxes of neutrinos, gamma rays, antiprotons, neutrons, and antineutrons that result from collisions of high-energy Galactic cosmic rays with the solar atmosphere are estimated. The results are sensitive to assumptions about cosmic-ray transport in the magnetic fields of the inner solar system. The high-energy photon flux should be observable by the Gamma Ray Observatory. The neutrino flux should produce less than one event per year in the next generation of neutrino telescopes. The antiproton flux is unobservable against the Galactic background. The neutron and antineutron fluxes are detectable only if neutrons produced in terrestrial cosmic-ray events may be discriminated against.

Impact of Including Higher Actinides in Fast Reactor Transmutation Analyses

DOE Office of Scientific and Technical Information (OSTI.GOV)

B. Forget; M. Asgari; R. Ferrer

2007-09-01

Previous fast reactor transmutation studies generally disregarded higher mass minor actinides beyond Cm-246 due to various considerations including deficiencies in nuclear cross-section data. Although omission of these higher mass actinides does not significantly impact the neutronic calculations and fuel cycle performance parameters follow-on neutron dose calculations related to fuel recycling, transportation and handling are significantly impacted. This report shows that including the minor actinides in the equilibrium fast reactor calculations will increase the predicted neutron emission by about 30%. In addition a sensitivity study was initiated by comparing the impact of different cross-section evaluation file for representing these minor actinides.

Using a derivative-free optimization method for multiple solutions of inverse transport problems

DOE PAGES

Armstrong, Jerawan C.; Favorite, Jeffrey A.

2016-01-14

Identifying unknown components of an object that emits radiation is an important problem for national and global security. Radiation signatures measured from an object of interest can be used to infer object parameter values that are not known. This problem is called an inverse transport problem. An inverse transport problem may have multiple solutions and the most widely used approach for its solution is an iterative optimization method. This paper proposes a stochastic derivative-free global optimization algorithm to find multiple solutions of inverse transport problems. The algorithm is an extension of a multilevel single linkage (MLSL) method where a meshmore » adaptive direct search (MADS) algorithm is incorporated into the local phase. Furthermore, numerical test cases using uncollided fluxes of discrete gamma-ray lines are presented to show the performance of this new algorithm.« less

Fuzzy multi-objective chance-constrained programming model for hazardous materials transportation

NASA Astrophysics Data System (ADS)

Du, Jiaoman; Yu, Lean; Li, Xiang

2016-04-01

Hazardous materials transportation is an important and hot issue of public safety. Based on the shortest path model, this paper presents a fuzzy multi-objective programming model that minimizes the transportation risk to life, travel time and fuel consumption. First, we present the risk model, travel time model and fuel consumption model. Furthermore, we formulate a chance-constrained programming model within the framework of credibility theory, in which the lengths of arcs in the transportation network are assumed to be fuzzy variables. A hybrid intelligent algorithm integrating fuzzy simulation and genetic algorithm is designed for finding a satisfactory solution. Finally, some numerical examples are given to demonstrate the efficiency of the proposed model and algorithm.

Dose conversion coefficients for neutron exposure to the lens of the human eye

DOE Office of Scientific and Technical Information (OSTI.GOV)

Manger, Ryan P; Bellamy, Michael B; Eckerman, Keith F

Dose conversion coefficients for the lens of the human eye have been calculated for neutron exposure at energies from 1 x 10{sup -9} to 20 MeV and several standard orientations: anterior-to-posterior, rotational and right lateral. MCNPX version 2.6.0, a Monte Carlo-based particle transport package, was used to determine the energy deposited in the lens of the eye. The human eyeball model was updated by partitioning the lens into sensitive and insensitive volumes as the anterior portion (sensitive volume) of the lens being more radiosensitive and prone to cataract formation. The updated eye model was used with the adult UF-ORNL mathematicalmore » phantom in the MCNPX transport calculations.« less

Simulations of the Neutron Star Crust

NASA Astrophysics Data System (ADS)

Schramm, Stefan; Nandi, Rana

The properties of the neutron star crust are crucially important for many physical processes occurring in the star. For instance, the crustal transport coefficients define the temperature evolution of accreting stars after bursts, which can be compared to observation. Furthermore, the structure of the inner crust can modify the neutrino transport through the matter considerably, significantly impacting the dynamics of supernova explosions. Therefore, we perform numerical studies of the inner crust, and among other aspects, investigate the dependence of the pasta phase on the isospin properties of the nuclear interactions. To this end we developed an efficient computer code to simulate the inner and outer crust using molecular dynamics techniques. First results of the simulations and insights into the crust-core transition are presented.

Reasons for 2011 Release of the Evaluated Nuclear Data Library (ENDL2011.0)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Brown, D.; Escher, J.; Hoffman, R.

LLNL's Computational Nuclear Physics Group and Nuclear Theory and Modeling Group have collaborated to create the 2011 release of the Evaluated Nuclear Data Library (ENDL2011). ENDL2011 is designed to sup- port LLNL's current and future nuclear data needs. This database is currently the most complete nuclear database for Monte Carlo and deterministic transport of neutrons and charged particles, surpassing ENDL2009.0 [1]. The ENDL2011 release [2] contains 918 transport-ready eval- uations in the neutron sub-library alone. ENDL2011 was assembled with strong support from the ASC program, leveraged with support from NNSA science campaigns and the DOE/Offce of Science US Nuclear Datamore » Pro- gram.« less

New Result for the β-decay Asymmetry Parameter A0 from the UCNA Experiment

NASA Astrophysics Data System (ADS)

Brown, M. A.-P.; UCNA Collaboration

2017-09-01

The UCNA Experiment at the Ultracold Neutron facility at LANL uses polarized ultracold neutrons (UCN) to determine the neutron β-decay asymmetry parameter A0, the angular correlation between the neutron spin and the decay electron's momentum. A0 further determines λ =gA /gV , which, when combined with the neutron lifetime, permits extraction of the CKM matrix element Vud solely from neutron decay. In the UCNA experiment, UCN are produced in a pulsed, spallation driven solid deuterium source, polarized using a 7 T magnetic field, and transported through an Adiabatic Fast Passage (AFP) spin flipper prior to storage within a 1 T solenoidal spectrometer housing electron detectors at each end. The spin-flipper allows one to form a super-ratio of decay rates for neutron spins aligned parallel and anti-parallel to the 1 T magnetic field, eliminating to first order errors due to variations in the decay rate and detector efficiencies. Leading systematics and analysis techniques from the most recent analysis of data collected from 2011-2013 will be presented. This material is based upon work supported by the U.S. Department of Energy, Office of Science, Office of Nuclear Physics, under Award Number DE-SC-0014622.

Activation Inventories after Exposure to DD/DT Neutrons in Safety Analysis of Nuclear Fusion Installations.

PubMed

Stankunas, Gediminas; Cufar, Aljaz; Tidikas, Andrius; Batistoni, Paola

2017-11-23

Irradiations with 14 MeV fusion neutrons are planned at Joint European Torus (JET) in DT operations with the objective to validate the calculation of the activation of structural materials in functional materials expected in ITER and fusion plants. This study describes the activation and dose rate calculations performed for materials irradiated throughout the DT plasma operation during which the samples of real fusion materials are exposed to 14 MeV neutrons inside the JET vacuum vessel. Preparatory activities are in progress during the current DD operations with dosimetry foils to measure the local neutron fluence and spectrum at the sample irradiation position. The materials included those used in the manufacturing of the main in-vessel components, such as ITER-grade W, Be, CuCrZr, 316 L(N) and the functional materials used in diagnostics and heating systems. The neutron-induced activities and dose rates at shutdown were calculated by the FISPACT code, using the neutron fluxes and spectra that were provided by the preceding MCNP neutron transport calculations. © The Author 2017. Published by Oxford University Press. All rights reserved. For Permissions, please email: journals.permissions@oup.com.

A Monte-Carlo Benchmark of TRIPOLI-4® and MCNP on ITER neutronics

NASA Astrophysics Data System (ADS)

Blanchet, David; Pénéliau, Yannick; Eschbach, Romain; Fontaine, Bruno; Cantone, Bruno; Ferlet, Marc; Gauthier, Eric; Guillon, Christophe; Letellier, Laurent; Proust, Maxime; Mota, Fernando; Palermo, Iole; Rios, Luis; Guern, Frédéric Le; Kocan, Martin; Reichle, Roger

2017-09-01

Radiation protection and shielding studies are often based on the extensive use of 3D Monte-Carlo neutron and photon transport simulations. ITER organization hence recommends the use of MCNP-5 code (version 1.60), in association with the FENDL-2.1 neutron cross section data library, specifically dedicated to fusion applications. The MCNP reference model of the ITER tokamak, the `C-lite', is being continuously developed and improved. This article proposes to develop an alternative model, equivalent to the 'C-lite', but for the Monte-Carlo code TRIPOLI-4®. A benchmark study is defined to test this new model. Since one of the most critical areas for ITER neutronics analysis concerns the assessment of radiation levels and Shutdown Dose Rates (SDDR) behind the Equatorial Port Plugs (EPP), the benchmark is conducted to compare the neutron flux through the EPP. This problem is quite challenging with regard to the complex geometry and considering the important neutron flux attenuation ranging from 1014 down to 108 n•cm-2•s-1. Such code-to-code comparison provides independent validation of the Monte-Carlo simulations, improving the confidence in neutronic results.

Ex-vessel neutron dosimetry analysis for westinghouse 4-loop XL pressurized water reactor plant using the RadTrack{sup TM} Code System with the 3D parallel discrete ordinates code RAPTOR-M3G

DOE Office of Scientific and Technical Information (OSTI.GOV)

Chen, J.; Alpan, F. A.; Fischer, G.A.

2011-07-01

Traditional two-dimensional (2D)/one-dimensional (1D) SYNTHESIS methodology has been widely used to calculate fast neutron (>1.0 MeV) fluence exposure to reactor pressure vessel in the belt-line region. However, it is expected that this methodology cannot provide accurate fast neutron fluence calculation at elevations far above or below the active core region. A three-dimensional (3D) parallel discrete ordinates calculation for ex-vessel neutron dosimetry on a Westinghouse 4-Loop XL Pressurized Water Reactor has been done. It shows good agreement between the calculated results and measured results. Furthermore, the results show very different fast neutron flux values at some of the former plate locationsmore » and elevations above and below an active core than those calculated by a 2D/1D SYNTHESIS method. This indicates that for certain irregular reactor internal structures, where the fast neutron flux has a very strong local effect, it is required to use a 3D transport method to calculate accurate fast neutron exposure. (authors)« less

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

NASA Astrophysics Data System (ADS)

Brovchenko, Mariya; Duhamel, Isabelle; Dechenaux, Benjamin

2017-09-01

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

Radiation damage and waste management options for the SOMBRERO final focus system and neutron dumps

DOE Office of Scientific and Technical Information (OSTI.GOV)

Latkowski, J F; Meier, W R; Reyes, S

1999-08-09

Previous studies of the safety and environmental aspects of the SOMBRERO inertial fusion energy (IFE) power plant design did not completely address the issues associated with the final focus system. While past work calculated neutron fluences for a grazing incidence metal mirror (GIMM) and a final focus mirror, scattering off of the final optical component was not included, and thus, fluences in the final focus mirror were significantly underestimated. In addition, past work did not consider neutron-induced gamma-rays. Finally, power plant lifetime waste volumes may have been underestimated as neutron activation of the neutron dumps and building structure were notmore » addressed. In the present work, a modified version of the SOMBRERO target building is presented where a significantly larger open solid-angle fraction (5%) is used to enhance beam smoothing of a diode-pumped solid-state laser (DPSSL). The GIMMs are replaced with transmissive fused silica wedges and have been included in three -dimensional neutron and photon transport calculations. This work shows that a power plant with a large open solid-angle fraction, needed for beam smoothing with a DPSSL, is acceptable from tritium breeding, and neutron activation points-of-view.« less

A study of heavy-heavy nuclear reactions. [nuclear research/nuclear particles

NASA Technical Reports Server (NTRS)

Khandelwal, G. S.

1975-01-01

Calculations are presented for the reaction products in high energy collisions and of the atmospheric transport of particles such as protons, neutrons and other nucleons. The magnetic moments of charmed baryons are examined. Total cross sections which are required for cosmic heavy ion transport and shielding studies are also examined.

Monte carlo simulations of the n_TOF lead spallation target with the Geant4 toolkit: A benchmark study

NASA Astrophysics Data System (ADS)

Lerendegui-Marco, J.; Cortés-Giraldo, M. A.; Guerrero, C.; Quesada, J. M.; Meo, S. Lo; Massimi, C.; Barbagallo, M.; Colonna, N.; Mancussi, D.; Mingrone, F.; Sabaté-Gilarte, M.; Vannini, G.; Vlachoudis, V.; Aberle, O.; Andrzejewski, J.; Audouin, L.; Bacak, M.; Balibrea, J.; Bečvář, F.; Berthoumieux, E.; Billowes, J.; Bosnar, D.; Brown, A.; Caamaño, M.; Calviño, F.; Calviani, M.; Cano-Ott, D.; Cardella, R.; Casanovas, A.; Cerutti, F.; Chen, Y. H.; Chiaveri, E.; Cortés, G.; Cosentino, L.; Damone, L. A.; Diakaki, M.; Domingo-Pardo, C.; Dressler, R.; Dupont, E.; Durán, I.; Fernández-Domínguez, B.; Ferrari, A.; Ferreira, P.; Finocchiaro, P.; Göbel, K.; Gómez-Hornillos, M. B.; García, A. R.; Gawlik, A.; Gilardoni, S.; Glodariu, T.; Gonçalves, I. F.; González, E.; Griesmayer, E.; Gunsing, F.; Harada, H.; Heinitz, S.; Heyse, J.; Jenkins, D. G.; Jericha, E.; Käppeler, F.; Kadi, Y.; Kalamara, A.; Kavrigin, P.; Kimura, A.; Kivel, N.; Kokkoris, M.; Krtička, M.; Kurtulgil, D.; Leal-Cidoncha, E.; Lederer, C.; Leeb, H.; Lonsdale, S. J.; Macina, D.; Marganiec, J.; Martínez, T.; Masi, A.; Mastinu, P.; Mastromarco, M.; Maugeri, E. A.; Mazzone, A.; Mendoza, E.; Mengoni, A.; Milazzo, P. M.; Musumarra, A.; Negret, A.; Nolte, R.; Oprea, A.; Patronis, N.; Pavlik, A.; Perkowski, J.; Porras, I.; Praena, J.; Radeck, D.; Rauscher, T.; Reifarth, R.; Rout, P. C.; Rubbia, C.; Ryan, J. A.; Saxena, A.; Schillebeeckx, P.; Schumann, D.; Smith, A. G.; Sosnin, N. V.; Stamatopoulos, A.; Tagliente, G.; Tain, J. L.; Tarifeño-Saldivia, A.; Tassan-Got, L.; Valenta, S.; Variale, V.; Vaz, P.; Ventura, A.; Vlastou, R.; Wallner, A.; Warren, S.; Woods, P. J.; Wright, T.; Žugec, P.

2017-09-01

Monte Carlo (MC) simulations are an essential tool to determine fundamental features of a neutron beam, such as the neutron flux or the γ-ray background, that sometimes can not be measured or at least not in every position or energy range. Until recently, the most widely used MC codes in this field had been MCNPX and FLUKA. However, the Geant4 toolkit has also become a competitive code for the transport of neutrons after the development of the native Geant4 format for neutron data libraries, G4NDL. In this context, we present the Geant4 simulations of the neutron spallation target of the n_TOF facility at CERN, done with version 10.1.1 of the toolkit. The first goal was the validation of the intra-nuclear cascade models implemented in the code using, as benchmark, the characteristics of the neutron beam measured at the first experimental area (EAR1), especially the neutron flux and energy distribution, and the time distribution of neutrons of equal kinetic energy, the so-called Resolution Function. The second goal was the development of a Monte Carlo tool aimed to provide useful calculations for both the analysis and planning of the upcoming measurements at the new experimental area (EAR2) of the facility.

Footprint Characteristics of Cosmic-Ray Neutron Sensors for Soil Moisture Monitoring

NASA Astrophysics Data System (ADS)

Schrön, Martin; Köhli, Markus; Zreda, Marek; Dietrich, Peter; Zacharias, Steffen

2015-04-01

Cosmic-ray neutron sensing is a unique and an increasingly accepted method to monitor the effective soil water content at the field scale. The technology is famous for its low maintenance, non-invasiveness, continuous measurement, and most importantly, for its large footprint. Being more representative than point data and finer resolved than remote-sensing products, cosmic-ray neutron derived soil moisture products provide unrivaled advantage for mesoscale hydrologic and land surface models. The method takes advantage of neutrons induced by cosmic radiation which are extraordinarily sensitive to hydrogen and behave like a hot gas. Information about nearby water sources are quickly mixed in a domain of tens of hectares in air. Since experimental determination of the actual spatial extent is hardly possible, scientists have applied numerical models to address the footprint characteristics. We have revisited previous neutron transport simulations and present a modified conceptual design and refined physical assumptions. Our revised study reveals new insights into probing distance and water sensitivity of detected neutrons under various environmental conditions. These results sharpen the range of interpretation concerning the spatial extent of integral soil moisture products derived from cosmic-ray neutron counts. Our findings will have important impact on calibration strategies, on scales for data assimilation and on the interpolation of soil moisture data derived from mobile cosmic-ray neutron surveys.

A collimated neutron detector for RFP plasmas in MST

DOE Office of Scientific and Technical Information (OSTI.GOV)

Capecchi, W. J., E-mail: capecchi@wisc.edu; Anderson, J. K.; Bonofiglo, P. J.

The neutron emissivity profile in the Madison Symmetric Torus is being reconstructed through the use of a collimated neutron detector. A scintillator-photomultiplier tube (PMT) system is employed to detect the fusion neutrons with the plasma viewing volume defined by a 55 cm deep, 5 cm diameter aperture. Effective detection of neutrons from the viewing volume is achieved through neutron moderation using 1300 lbs of high density polyethylene shielding, which modeling predicts attenuates the penetrating flux by a factor of 10{sup 4} or more. A broad spectrum of gamma radiation is also present due to the unconfined fusion proton bombardment ofmore » the thick aluminum vacuum vessel. A 15 cm cylindrical liquid scintillator of 3.8 cm diameter is used to further increase directional sensitivity. A fast (5 ns rise time) preamplifier and digitization at 500 MHz prevent pulse pile-up even at high count rates (∼10{sup 4}/s). The entire neutron camera system is situated on an adjustable inclining base which provides the differing plasma viewing volumes necessary for reconstruction of the neutron emissivity profile. This profile, directly related to the fast-ion population, allows for an investigation of the critical fast-ion pressure gradient required to destabilize a neutral beam driven Alfvénic mode which has been shown to transport fast ions.« less

Characterization of the graphite pile as a source of thermal neutrons

NASA Astrophysics Data System (ADS)

Vykydal, Zdenek; Králík, Miloslav; Jančář, Aleš; Kopecký, Zdeněk; Dressler, Jan; Veškrna, Martin

2015-11-01

A new graphite pile designed to serve as a standard source of thermal neutrons has been built at the Czech Metrology Institute. Actual dimensions of the pile are 1.95 m (W)×1.95 m (L)×2.0 m (H). At its center, there is a measurement channel whose dimensions are 0.4 m×0.4 m×1.25 m (depth). The channel is equipped with a calibration bench, which allows reproducible placement of the tested/calibrated device. At a distance of 80 cm from the channel axis, six holes are symmetrically located allowing the placement of radionuclide neutron sources of Pu-Be and/or Am-Be type. Spatial distribution of thermal neutron fluence in the cavity was calculated in detail with the MCNP neutron transport code. Experimentally, it was measured with two active detectors: a small 3He proportional detector by the French company LMT, type 0.5 NH 1/1 KF, and a silicon pixel detector Timepix with 10B converter foil. The relative values of thermal neutron fluence rate obtained with active detectors were converted to absolute ones using thermal neutron fluence rates measured by means of gold foil activation. The quality of thermal neutron field was characterized by the cadmium ratio.

General relativistic viscous hydrodynamics of differentially rotating neutron stars

NASA Astrophysics Data System (ADS)

Shibata, Masaru; Kiuchi, Kenta; Sekiguchi, Yu-ichiro

2017-04-01

Employing a simplified version of the Israel-Stewart formalism for general-relativistic shear-viscous hydrodynamics, we perform axisymmetric general-relativistic simulations for a rotating neutron star surrounded by a massive torus, which can be formed from differentially rotating stars. We show that with our choice of a shear-viscous hydrodynamics formalism, the simulations can be stably performed for a long time scale. We also demonstrate that with a possibly high shear-viscous coefficient, not only viscous angular momentum transport works but also an outflow could be driven from a hot envelope around the neutron star for a time scale ≳100 ms with the ejecta mass ≳10-2 M⊙ , which is comparable to the typical mass for dynamical ejecta of binary neutron-star mergers. This suggests that massive neutron stars surrounded by a massive torus, which are typical outcomes formed after the merger of binary neutron stars, could be the dominant source for providing neutron-rich ejecta, if the effective shear viscosity is sufficiently high, i.e., if the viscous α parameter is ≳10-2. The present numerical result indicates the importance of a future high-resolution magnetohydrodynamics simulation that is the unique approach to clarify the viscous effect in the merger remnants of binary neutron stars by the first-principle manner.

Transport Corrections in Nodal Diffusion Codes for HTR Modeling

DOE Office of Scientific and Technical Information (OSTI.GOV)

Abderrafi M. Ougouag; Frederick N. Gleicher

2010-08-01

The cores and reflectors of High Temperature Reactors (HTRs) of the Next Generation Nuclear Plant (NGNP) type are dominantly diffusive media from the point of view of behavior of the neutrons and their migration between the various structures of the reactor. This means that neutron diffusion theory is sufficient for modeling most features of such reactors and transport theory may not be needed for most applications. Of course, the above statement assumes the availability of homogenized diffusion theory data. The statement is true for most situations but not all. Two features of NGNP-type HTRs require that the diffusion theory-based solutionmore » be corrected for local transport effects. These two cases are the treatment of burnable poisons (BP) in the case of the prismatic block reactors and, for both pebble bed reactor (PBR) and prismatic block reactor (PMR) designs, that of control rods (CR) embedded in non-multiplying regions near the interface between fueled zones and said non-multiplying zones. The need for transport correction arises because diffusion theory-based solutions appear not to provide sufficient fidelity in these situations.« less

Progress on Establishing the Feasibility of Lead Slowing Down Spectroscopy for Direct Measurement of Plutonium in Used Fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kulisek, Jonathan A.; Anderson, Kevin K.; Bowyer, Sonya M.

2012-07-19

Developing a method for the accurate, direct, and independent assay of the fissile isotopes in bulk materials (such as used fuel) of next-generation domestic nuclear fuel cycles is a goal of the Office of Nuclear Energy, Fuel Cycle R&D, Material Protection and Control Technology (MPACT) Campaign. To meet this goal, MPACT continues to support a multi-institutional collaboration to address the feasibility of Lead Slowing Down Spectroscopy (LSDS) as an active nondestructive assay method that has the potential to provide independent, direct measurement of Pu and U isotopic masses in used fuel with an uncertainty considerably lower than the approximately 10%more » typical of today’s confirmatory assay methods. An LSDS is comprised of a stack of lead (typically 1-6 m3) in which materials to be measured are placed in the lead and a pulse of neutrons is injected. The neutrons in this pulse lose energy due to inelastic and (subsequently) elastic scattering and the average energy of the neutrons decreases as the time increases by a well-defined relationship. In the interrogation energy region (~0.1-1000 eV) the neutrons have little energy spread (~30%) about the average neutron energy. Due to this characteristic, the energy of the (assay) neutrons can then be determined by measuring the time elapsed since the neutron pulse. By measuring the induced fission neutrons emitted from the used fuel, it is possible to determine isotopic-mass content by unfolding the unique structure of isotopic resonances across the interrogation energy region. This paper will present efforts on the development of time-spectral analysis algorithms, fast neutron detector advances, and validation and testing measurements.« less

Adjoint-Based Implicit Uncertainty Analysis for Figures of Merit in a Laser Inertial Fusion Engine

DOE Office of Scientific and Technical Information (OSTI.GOV)

Seifried, J E; Fratoni, M; Kramer, K J

A primary purpose of computational models is to inform design decisions and, in order to make those decisions reliably, the confidence in the results of such models must be estimated. Monte Carlo neutron transport models are common tools for reactor designers. These types of models contain several sources of uncertainty that propagate onto the model predictions. Two uncertainties worthy of note are (1) experimental and evaluation uncertainties of nuclear data that inform all neutron transport models and (2) statistical counting precision, which all results of a Monte Carlo codes contain. Adjoint-based implicit uncertainty analyses allow for the consideration of anymore » number of uncertain input quantities and their effects upon the confidence of figures of merit with only a handful of forward and adjoint transport calculations. When considering a rich set of uncertain inputs, adjoint-based methods remain hundreds of times more computationally efficient than Direct Monte-Carlo methods. The LIFE (Laser Inertial Fusion Energy) engine is a concept being developed at Lawrence Livermore National Laboratory. Various options exist for the LIFE blanket, depending on the mission of the design. The depleted uranium hybrid LIFE blanket design strives to close the fission fuel cycle without enrichment or reprocessing, while simultaneously achieving high discharge burnups with reduced proliferation concerns. Neutron transport results that are central to the operation of the design are tritium production for fusion fuel, fission of fissile isotopes for energy multiplication, and production of fissile isotopes for sustained power. In previous work, explicit cross-sectional uncertainty analyses were performed for reaction rates related to the figures of merit for the depleted uranium hybrid LIFE blanket. Counting precision was also quantified for both the figures of merit themselves and the cross-sectional uncertainty estimates to gauge the validity of the analysis. All cross-sectional uncertainties were small (0.1-0.8%), bounded counting uncertainties, and were precise with regard to counting precision. Adjoint/importance distributions were generated for the same reaction rates. The current work leverages those adjoint distributions to transition from explicit sensitivities, in which the neutron flux is constrained, to implicit sensitivities, in which the neutron flux responds to input perturbations. This treatment vastly expands the set of data that contribute to uncertainties to produce larger, more physically accurate uncertainty estimates.« less

Modeling the tagged-neutron UXO identification technique using the Geant4 toolkit

DOE Office of Scientific and Technical Information (OSTI.GOV)

Zhou Y.; Mitra S.; Zhu X.

2011-10-16

It is proposed to use 14 MeV neutrons tagged by the associated particle neutron time-of-flight technique (APnTOF) to identify the fillers of unexploded ordnances (UXO) by characterizing their carbon, nitrogen and oxygen contents. To facilitate the design and construction of a prototype system, a preliminary simulation model was developed, using the Geant4 toolkit. This work established the toolkit environment for (a) generating tagged neutrons, (b) their transport and interactions within a sample to induce emission and detection of characteristic gamma-rays, and (c) 2D and 3D-image reconstruction of the interrogated object using the neutron and gamma-ray time-of-flight information. Using the modeling,more » this article demonstrates the novelty of the tagged-neutron approach for extracting useful signals with high signal-to-background discrimination of an object-of-interest from that of its environment. Simulations indicated that an UXO filled with the RDX explosive, hexogen (C{sub 3}H{sub 6}O{sub 6}N{sub 6}), can be identified to a depth of 20 cm when buried in soil.« less

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

NASA Astrophysics Data System (ADS)

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

2017-01-01

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

Report of the M16 Rifle Review Panel. Volume 10, Appendix 9. Audit Trail of Chief of Staff, Army Actions and Decisions Concerning the M16

DTIC Science & Technology

1968-06-01

CHEMORECEPTORS CHEMOTAXIS CHEMOTHERAPEUTIC AGENTS CHEMOTHERAPY CHERRIES CHESAPEAKE BAY CHI SQUARE TEST CHICKENS CHIKUNGUNYA VIRUS CHILDREN CHILE...METHYL SULFOXIDE METHYLAL METHYLAMINE METHYLATION METHYLENE BLUE METHYLENES METRIC SYSTEM METROLOGY MEXICO MEXICO GULF MICA MICA CAPACITORS...NEUTRON TRANSPORT THEORY NEUTRONS NEVADA NEW BRUNSWICK NEW ENGLAND NEW GUINEA NEW HAMPSHIRE NEW JERSEY NEW MEXICO NEW YORK NEW YORK CITY

Neutron Protection Factor Determination and Validation for a Vehicle Surrogate Using a Californium Fission Source

DTIC Science & Technology

2017-06-01

protection factors . The success of this research is a direct result of the immense collaboration across a number of institutions that all shared a...at post detonation neutron transport, an exact solution is not needed. Instead, the RPF research campaign uses a statistical-based method through a... factors of selected light vehicles against residual radiation,” United States Army Ballistic Research Laboratory, Aberdeen Proving Ground, MD, 1988

Modeling of Radioxenon Production and Release Pathways

DTIC Science & Technology

2010-09-01

MCNP is utilized to model the neutron transport, while ORIGEN 2.2 is utilized to calculate the production and decay of fission products, activation...products, and transuranics resulting from the calculated neutron flux profile. MONTEBURNS is a pearl script that couples the MCNP and ORIGEN 2.2...core enriched to 93.80% 239Pu was used. MCNP was used to determine the thickness of soil or rock necessary to accurately model the attenuation and

The structural and magnetic phase transitions in a ``parent'' Fe pnictide compound

NASA Astrophysics Data System (ADS)

Ni, Ni; Allred, Jared; Cao, Huibo; Tian, Wei; Liu, Lian; Cho, Kyuil; Krogstad, Matthew; Ma, Jie; Taddei, Keith; Tanatar, Makariy; Prozorov, Ruslan; Matsuda, Masaaki; Rosenkranz, Stephan; Uemura, Yasutomo; Jiang, Shan

2015-03-01

We will present transport, thermodynamic, synchrotron X-ray, neutron diffraction, μSR, ARPES and polarized optical image measurements on the ``parent'' compound of the 112 high Tc superconducting Fe pnictide family. Structural and magnetic phase transitions are revealed. Detailed magnetic structure was solved by single crystal neutron diffraction. We will discuss the similarity and difference of these transitions comparing to the parent compounds of other Fe pnictide superconductors.

Optimizing the real-time ground level enhancement alert system based on neutron monitor measurements: Introducing GLE Alert Plus

NASA Astrophysics Data System (ADS)

Souvatzoglou, G.; Papaioannou, A.; Mavromichalaki, H.; Dimitroulakos, J.; Sarlanis, C.

2014-11-01

Whenever a significant intensity increase is being recorded by at least three neutron monitor stations in real-time mode, a ground level enhancement (GLE) event is marked and an automated alert is issued. Although, the physical concept of the algorithm is solid and has efficiently worked in a number of cases, the availability of real-time data is still an open issue and makes timely GLE alerts quite challenging. In this work we present the optimization of the GLE alert that has been set into operation since 2006 at the Athens Neutron Monitor Station. This upgrade has led to GLE Alert Plus, which is currently based upon the Neutron Monitor Database (NMDB). We have determined the critical values per station allowing us to issue reliable GLE alerts close to the initiation of the event while at the same time we keep the false alert rate at low levels. Furthermore, we have managed to treat the problem of data availability, introducing the Go-Back-N algorithm. A total of 13 GLE events have been marked from January 2000 to December 2012. GLE Alert Plus issued an alert for 12 events. These alert times are compared to the alert times of GOES Space Weather Prediction Center and Solar Energetic Particle forecaster of the University of Málaga (UMASEP). In all cases GLE Alert Plus precedes the GOES alert by ≈8-52 min. The comparison with UMASEP demonstrated a remarkably good agreement. Real-time GLE alerts by GLE Alert Plus may be retrieved by http://cosray.phys.uoa.gr/gle_alert_plus.html, http://www.nmdb.eu, and http://swe.ssa.esa.int/web/guest/space-radiation. An automated GLE alert email notification system is also available to interested users.

Experimental evaluation of the extended Dytlewski-style dead time correction formalism for neutron multiplicity counting

DOE PAGES

Lockhart, M.; Henzlova, D.; Croft, S.; ...

2017-09-20

Over the past few decades, neutron multiplicity counting has played an integral role in Special Nuclear Material (SNM) characterization pertaining to nuclear safeguards. Current neutron multiplicity analysis techniques use singles, doubles, and triples count rates because a methodology to extract and dead time correct higher order count rates (i.e. quads and pents) was not fully developed. This limitation is overcome by the recent extension of a popular dead time correction method developed by Dytlewski. This extended dead time correction algorithm, named Dytlewski-Croft-Favalli (DCF), is detailed in reference Croft and Favalli (2017), which gives an extensive explanation of the theory andmore » implications of this new development. Dead time corrected results can then be used to assay SNM by inverting a set of extended point model equations which as well have only recently been formulated. Here, we discuss and present the experimental evaluation of practical feasibility of the DCF dead time correction algorithm to demonstrate its performance and applicability in nuclear safeguards applications. In order to test the validity and effectiveness of the dead time correction for quads and pents, 252Cf and SNM sources were measured in high efficiency neutron multiplicity counters at the Los Alamos National Laboratory (LANL) and the count rates were extracted up to the fifth order and corrected for dead time. To assess the DCF dead time correction, the corrected data is compared to traditional dead time correction treatment within INCC. In conclusion, the DCF dead time correction is found to provide adequate dead time treatment for broad range of count rates available in practical applications.« less

Experimental evaluation of the extended Dytlewski-style dead time correction formalism for neutron multiplicity counting

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lockhart, M.; Henzlova, D.; Croft, S.

Over the past few decades, neutron multiplicity counting has played an integral role in Special Nuclear Material (SNM) characterization pertaining to nuclear safeguards. Current neutron multiplicity analysis techniques use singles, doubles, and triples count rates because a methodology to extract and dead time correct higher order count rates (i.e. quads and pents) was not fully developed. This limitation is overcome by the recent extension of a popular dead time correction method developed by Dytlewski. This extended dead time correction algorithm, named Dytlewski-Croft-Favalli (DCF), is detailed in reference Croft and Favalli (2017), which gives an extensive explanation of the theory andmore » implications of this new development. Dead time corrected results can then be used to assay SNM by inverting a set of extended point model equations which as well have only recently been formulated. Here, we discuss and present the experimental evaluation of practical feasibility of the DCF dead time correction algorithm to demonstrate its performance and applicability in nuclear safeguards applications. In order to test the validity and effectiveness of the dead time correction for quads and pents, 252Cf and SNM sources were measured in high efficiency neutron multiplicity counters at the Los Alamos National Laboratory (LANL) and the count rates were extracted up to the fifth order and corrected for dead time. To assess the DCF dead time correction, the corrected data is compared to traditional dead time correction treatment within INCC. In conclusion, the DCF dead time correction is found to provide adequate dead time treatment for broad range of count rates available in practical applications.« less

Hybrid geometric-random template-placement algorithm for gravitational wave searches from compact binary coalescences

NASA Astrophysics Data System (ADS)

Roy, Soumen; Sengupta, Anand S.; Thakor, Nilay

2017-05-01

Astrophysical compact binary systems consisting of neutron stars and black holes are an important class of gravitational wave (GW) sources for advanced LIGO detectors. Accurate theoretical waveform models from the inspiral, merger, and ringdown phases of such systems are used to filter detector data under the template-based matched-filtering paradigm. An efficient grid over the parameter space at a fixed minimal match has a direct impact on the overall time taken by these searches. We present a new hybrid geometric-random template placement algorithm for signals described by parameters of two masses and one spin magnitude. Such template banks could potentially be used in GW searches from binary neutron stars and neutron star-black hole systems. The template placement is robust and is able to automatically accommodate curvature and boundary effects with no fine-tuning. We also compare these banks against vanilla stochastic template banks and show that while both are equally efficient in the fitting-factor sense, the bank sizes are ˜25 % larger in the stochastic method. Further, we show that the generation of the proposed hybrid banks can be sped up by nearly an order of magnitude over the stochastic bank. Generic issues related to optimal implementation are discussed in detail. These improvements are expected to directly reduce the computational cost of gravitational wave searches.

SU-E-T-168: Characterization of Neutrons From the TrueBeam Treatment Head

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sawkey, D; Svatos, M

2015-06-15

Purpose: Calculate neutron production and transport in the TrueBeam treatment head, as input for vault design and phantom dose calculations. Methods: A detailed model of the treatment head, including shielding components off the beam axis, was created from manufacturer’s engineering drawings. Simulations were done with Geant4 for the 18X, 15X, 10X and 10FFF beams, tuned to match measured dose distributions inside the treatment field. Particles were recorded on a 70 cm radius sphere surrounding the treatment head enabling input into simulations of vaults. Results: For the 18X beam, 11×10{sup 9} neutrons/MU were observed. The energy spectrum was a broad peakmore » with average energy 0.37 MeV. With jaws closed, 48% of the neutrons were generated in the primary collimator, 18% in the jaws, 12% in the target, and 10% in the flattening filter. With wide open jaws, few neutrons were produced in the jaws and consequently total neutron production dropped to 8.5×10{sup 9} neutrons/MU. Angular distributions were greatest along the beam axis (12×10{sup 9} neutrons/MU/sr, within 2 deg of the beam axis) and antiparallel to the beam axis (7×10{sup 9} neutrons/MU/sr). Peaks were observed in the neutron energy spectrum, corresponding to elastic scattering resonances in the shielding materials. Neutron production was lower for the other beams studied: 4.1×10{sup 9} neutrons/MU for 15X, 0.38×10{sup 9} neutrons/MU for 10X, and 0.22×10{sup 9} neutrons/MU for 10FFF. Despite dissimilar treatment head geometries and materials, the neutron production and energy spectrum were similar to those reported for Clinac accelerators. Conclusion: Detailed neutron production and leakage calculations for the TrueBeam treatment head were done. Unlike other studies, results are independent of the surrounding vault, enabling vault design calculations.« less

New approach to description of (d,xn) spectra at energies below 50 MeV in Monte Carlo simulation by intra-nuclear cascade code with Distorted Wave Born Approximation

NASA Astrophysics Data System (ADS)

Hashimoto, S.; Iwamoto, Y.; Sato, T.; Niita, K.; Boudard, A.; Cugnon, J.; David, J.-C.; Leray, S.; Mancusi, D.

2014-08-01

A new approach to describing neutron spectra of deuteron-induced reactions in the Monte Carlo simulation for particle transport has been developed by combining the Intra-Nuclear Cascade of Liège (INCL) and the Distorted Wave Born Approximation (DWBA) calculation. We incorporated this combined method into the Particle and Heavy Ion Transport code System (PHITS) and applied it to estimate (d,xn) spectra on natLi, 9Be, and natC targets at incident energies ranging from 10 to 40 MeV. Double differential cross sections obtained by INCL and DWBA successfully reproduced broad peaks and discrete peaks, respectively, at the same energies as those observed in experimental data. Furthermore, an excellent agreement was observed between experimental data and PHITS-derived results using the combined method in thick target neutron yields over a wide range of neutron emission angles in the reactions. We also applied the new method to estimate (d,xp) spectra in the reactions, and discussed the validity for the proton emission spectra.

Viscous Dissipation and Heat Conduction in Binary Neutron-Star Mergers.

PubMed

Alford, Mark G; Bovard, Luke; Hanauske, Matthias; Rezzolla, Luciano; Schwenzer, Kai

2018-01-26

Inferring the properties of dense matter is one of the most exciting prospects from the measurement of gravitational waves from neutron star mergers. However, it requires reliable numerical simulations that incorporate viscous dissipation and energy transport as these can play a significant role in the survival time of the post-merger object. We calculate time scales for typical forms of dissipation and find that thermal transport and shear viscosity will not be important unless neutrino trapping occurs, which requires temperatures above 10 MeV and gradients over length scales of 0.1 km or less. On the other hand, if direct-Urca processes remain suppressed, leaving modified-Urca processes to establish flavor equilibrium, then bulk viscous dissipation could provide significant damping to density oscillations right after merger. When comparing with data from state-of-the-art merger simulations, we find that the bulk viscosity takes values close to its resonant maximum in a typical merger, motivating a more careful assessment of the role of bulk viscous dissipation in the gravitational-wave signal from merging neutron stars.

Viscous Dissipation and Heat Conduction in Binary Neutron-Star Mergers

NASA Astrophysics Data System (ADS)

Alford, Mark G.; Bovard, Luke; Hanauske, Matthias; Rezzolla, Luciano; Schwenzer, Kai

2018-01-01

Inferring the properties of dense matter is one of the most exciting prospects from the measurement of gravitational waves from neutron star mergers. However, it requires reliable numerical simulations that incorporate viscous dissipation and energy transport as these can play a significant role in the survival time of the post-merger object. We calculate time scales for typical forms of dissipation and find that thermal transport and shear viscosity will not be important unless neutrino trapping occurs, which requires temperatures above 10 MeV and gradients over length scales of 0.1 km or less. On the other hand, if direct-Urca processes remain suppressed, leaving modified-Urca processes to establish flavor equilibrium, then bulk viscous dissipation could provide significant damping to density oscillations right after merger. When comparing with data from state-of-the-art merger simulations, we find that the bulk viscosity takes values close to its resonant maximum in a typical merger, motivating a more careful assessment of the role of bulk viscous dissipation in the gravitational-wave signal from merging neutron stars.

Algorithms for optimization of the transport system in living and artificial cells.

PubMed

Melkikh, A V; Sutormina, M I

2011-06-01

An optimization of the transport system in a cell has been considered from the viewpoint of the operations research. Algorithms for an optimization of the transport system of a cell in terms of both the efficiency and a weak sensitivity of a cell to environmental changes have been proposed. The switching of various systems of transport is considered as the mechanism of weak sensitivity of a cell to changes in environment. The use of the algorithms for an optimization of a cardiac cell has been considered by way of example. We received theoretically for a cell of a cardiac muscle that at the increase of potassium concentration in the environment switching of transport systems for this ion takes place. This conclusion qualitatively coincides with experiments. The problem of synthesizing an optimal system in an artificial cell has been stated.

Scalable Domain Decomposed Monte Carlo Particle Transport

DOE Office of Scientific and Technical Information (OSTI.GOV)

O'Brien, Matthew Joseph

2013-12-05

In this dissertation, we present the parallel algorithms necessary to run domain decomposed Monte Carlo particle transport on large numbers of processors (millions of processors). Previous algorithms were not scalable, and the parallel overhead became more computationally costly than the numerical simulation.

Particle Filter-Based Recursive Data Fusion With Sensor Indexing for Large Core Neutron Flux Estimation

NASA Astrophysics Data System (ADS)

Tamboli, Prakash Kumar; Duttagupta, Siddhartha P.; Roy, Kallol

2017-06-01

We introduce a sequential importance sampling particle filter (PF)-based multisensor multivariate nonlinear estimator for estimating the in-core neutron flux distribution for pressurized heavy water reactor core. Many critical applications such as reactor protection and control rely upon neutron flux information, and thus their reliability is of utmost importance. The point kinetic model based on neutron transport conveniently explains the dynamics of nuclear reactor. The neutron flux in the large core loosely coupled reactor is sensed by multiple sensors measuring point fluxes located at various locations inside the reactor core. The flux values are coupled to each other through diffusion equation. The coupling facilitates redundancy in the information. It is shown that multiple independent data about the localized flux can be fused together to enhance the estimation accuracy to a great extent. We also propose the sensor anomaly handling feature in multisensor PF to maintain the estimation process even when the sensor is faulty or generates data anomaly.

Skyshine study for next generation of fusion devices

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gohar, Y.; Yang, S.

1987-02-01

A shielding analysis for next generation of fusion devices (ETR/INTOR) was performed to study the dose equivalent outside the reactor building during operation including the contribution from neutrons and photons scattered back by collisions with air nuclei (skyshine component). Two different three-dimensional geometrical models for a tokamak fusion reactor based on INTOR design parameters were developed for this study. In the first geometrical model, the reactor geometry and the spatial distribution of the deuterium-tritium neutron source were simplified for a parametric survey. The second geometrical model employed an explicit representation of the toroidal geometry of the reactor chamber and themore » spatial distribution of the neutron source. The MCNP general Monte Carlo code for neutron and photon transport was used to perform all the calculations. The energy distribution of the neutron source was used explicitly in the calculations with ENDF/B-V data. The dose equivalent results were analyzed as a function of the concrete roof thickness of the reactor building and the location outside the reactor building.« less

Materials for Low-Energy Neutron Radiation Shielding

NASA Technical Reports Server (NTRS)

Singleterry, Robert C., Jr.; Thibeault, Sheila A.

2000-01-01

Various candidate aircraft and spacecraft materials were analyzed and compared in a low-energy neutron environment using the Monte Carlo N-Particle (MCNP) transport code with an energy range up to 20 MeV. Some candidate materials have been tested in particle beams, and others seemed reasonable to analyze in this manner before deciding to test them. The two metal alloys analyzed are actual materials being designed into or used in aircraft and spacecraft today. This analysis shows that hydrogen-bearing materials have the best shielding characteristics over the metal alloys. It also shows that neutrons above 1 MeV are reflected out of the face of the slab better by larger quantities of carbon in the material. If a low-energy absorber is added to the material, fewer neutrons are transmitted through the material. Future analyses should focus on combinations of scatterers and absorbers to optimize these reaction channels and on the higher energy neutron component (above 50 MeV).

Neutron interrogation of high-enriched uranium by a 4 MeV linac

NASA Astrophysics Data System (ADS)

Lakosi, László; Nguyen, Cong Tam

2008-07-01

For revealing unauthorized transport (illicit trafficking) of nuclear materials, a non-destructive method reported earlier, utilizing a 4 MeV linear accelerator for photoneutron interrogation, was further developed. The linac served as a pulsed neutron source for assay of highly enriched uranium. Produced in beryllium or heavy water by bremsstrahlung, neutrons subsequently induced fission in the samples. Delayed neutrons were detected by a newly designed neutron collar built up of 14 3He counters embedded in a polyethylene moderator. A PC controlled multiscaler served as a time analyzer, triggering the detector startup by the beam pulse. Significant progress was achieved in enhancing the detector response, hence the sensitivity for revealing illicit material. A lower sensitivity limit of the order of 10 mg 235U was determined in a 20 s measurement time with a reasonable amount of beryllium (170 g) or of heavy water (100 g) and a mean electron current of 10 μA. Sensitivity can be further enhanced by increasing the measurement time.

Spatial heterogeneity of tungsten transmutation in a fusion device

NASA Astrophysics Data System (ADS)

Gilbert, M. R.; Sublet, J.-Ch.; Dudarev, S. L.

2017-04-01

Accurately quantifying the transmutation rate of tungsten (W) under neutron irradiation is a necessary requirement in the assessment of its performance as an armour material in a fusion power plant. The usual approach of calculating average responses, assuming large, homogenised material volumes, is insufficient to capture the full complexity of the transmutation picture in the context of a realistic fusion power plant design, particularly for rhenium (Re) production from W. Combined neutron transport and inventory simulations for representative spatially heterogeneous high-resolution models of a fusion power plant show that the production rate of Re is strongly influenced by the surrounding local spatial environment. Localised variation in neutron moderation (slowing down) due to structural steel and coolant, particularly water, can dramatically increase Re production because of the huge cross sections of giant resolved resonances in the neutron-capture reaction of 186W at low neutron energies. Calculations using cross section data corrected for temperature (Doppler) effects suggest that temperature may have a relatively lesser influence on transmutation rates.

Neutron Capture gamma ENDF libraries for modeling and identification of neutron sources

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sleaford, B

2007-10-29

There are a number of inaccuracies and data omissions with respect to gammas from neutron capture in the ENDF libraries used as field reference information and by modeling codes used in JTOT. As the use of Active Neutron interrogation methods is expanded, these shortfalls become more acute. A new, more accurate and complete evaluated experimental database of gamma rays (over 35,000 lines for 262 isotopes up to U so far) from thermal neutron capture has recently become available from the IAEA. To my knowledge, none of this new data has been installed in ENDF libraries and disseminated. I propose tomore » upgrade libraries of {sup 184,186}W, {sup 56}Fe, {sup 204,206,207}Pb, {sup 104}Pd, and {sup 19}F the 1st year. This will involve collaboration with Richard Firestone at LBL in evaluating the data and installing it in the libraries. I will test them with the transport code MCNP5.« less