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DOE OSTI2014

Analog neutron transport for nuclear instrumentation applications with the Monte-Carlo code TRIPOLI-4 - Paper 3

Monte-Carlo codes simulating particle transport usually make use of well-known variance reduction techniques, leading to more efficient simulations. Even in configurations which do not deal with deep-penetration, some of these techniques, such as the implicit capture, are actually often used by default. The statistical weight of the simulated particles is then modified in order to take into account the modification of the sampling without introducing any bias in the estimation process. For instance, the implicit capture decreases the statistical weight of the particles at each collision, whereas the sampling of a single outgoing particle increases this statistical weight in case of scattering reactions producing multiple neutrons. In addition, Russian roulette and splitting operations, which are usually also played at each collision, lead to other modifications of the statistical weights. As a result, Monte-Carlo codes provide particle histories that are valid on average, but collect statistical weights which differ from 1, when tallying the estimation of the quantities of interest. Yet, some nuclear instrumentation calculations inherently need to work with statistical weights equal to 1, since either individual events have to be collected (such as the arrival of a particle at a detector, the measurement of a time of flight, etc), or the coincidence of events or also correlations between individual events have to be taken into account. To address these specific applications with Monte-Carlo simulation codes, users need a full analog simulation mode. The continuous-energy Monte-Carlo code TRIPOLI4{sup R} is mainly dedicated to shielding, reactor physics and criticality safety calculations. Its capabilities have recently been extended to widen its use to nuclear instrumentation with the possibility of γ-spectrometry calculations and the simulation of photonuclear reactions for instance. Other features needed by nuclear instrumentation applications have been available in TRIPOLI-4 for a long time, such as the sampling of the time variable and the possibility of tallying in a time grid for most of the available tallies (fluxes, reaction rates, dose equivalent rates, energy depositions, etc). In TRIPOLI-4, the transport between two consecutive collisions is analog if no variance reduction technique is requested by the user to handle deep-penetration configurations. However, up to the version 9 of the code, the collision itself is sampled in a non-analog way. This paper presents a first implementation of a full analog neutron transport mode in TRIPOLI-4, preliminary verification results and an application of this functionality to the simulation of the Nucifer detector located in Saclay, France. It should be noted that only neutron transport is dealt with in this work and that all nuclear data come from JEFF-3.1.1 or ENDF/B-VII.0 evaluations, including fission data when needed. (authors)

61 RADIATION PROTECTION AND DOSIMETRY↗
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Applying Nonlinear Diffusion Acceleration to Fixed-Source Problems with Anisotropic Scattering - Paper 15

Work over the last several years has demonstrated the impressive performance gains that can be attained by using Moment-Based Acceleration (MBA), or High-Order/Low-Order (HOLO), methods in a variety of transport applications. In the field of neutronics, Nonlinear Diffusion Acceleration (NDA) has been used to accelerate the solution to the fixed-source problem and the k-eigenvalue problem. These methods efficiently move a major portion of the computational work out of the full phase-space (high-order (HO)) domain into a reduced phase-space (low-order (LO)) setting. In previous implementations, NDA has only been applied to isotropic scattering problems. In this paper, we will extend this methodology to anisotropic scattering. This initial work on anisotropic scattering will focus on the 1-D multi-group neutron transport equation. This summary focuses on a fixed-source problem, but the extension to criticality problems is straight-forward via NDA-NCA. Within this paper, we have described a method for extending Nonlinear Diffusion Acceleration to solve problems with anisotropic scattering. We demonstrated that a consistent low-order system, resembling the low-order system from, could be constructed. We demonstrated that the NDA method performed exceptionally in the presence of anisotropic scattering. NDA reduced the number of iterations required to meet the convergence tolerance by a factor of 500 over source iteration for a one fixed-source problem and by 3.5-50 x for the Henyey-Greenstein problems. This method is equally well-suited for eigenvalue problems. The NDA method allows us to transition the eigenvalue iteration to the low-order domain.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2014

MCNP6 Elevation Scaling of Cosmic Ray Backgrounds - Paper 29

The galactic cosmic-ray (GCR) source option was implemented in the all-energy, all-particle transport code MCNP62 in 2010. This source option along with other newly added features to MCNP6 have been used to produce increasingly accurate global cosmic and terrestrial background flux data files (background.dat).3 This data file is produced on a grid and is read and sampled by MCNP6 whenever a user invokes the background source option. The most recent adaptation of this data file, Version 4, will be completed in 2014. This paper reports on current progress toward enhancements that have been made to the modeling and simulation of these background spectra, in particular the elevation scaling used to produce more accurate spectra at off-grid locations. Cosmic radiation continuously bombards Earth with various particles, such as protons, alpha particles, and heavier nuclei, some of which are deflected by the Earth's shielding magnetic field. Particles that carry sufficient momentum can overcome the deflection and penetrate into the atmosphere. The sufficient momentum is dependent on the terrestrial coordinates due to the shape of the Earth's magnetic field and the Lorentz force's proportionality to the sine of the angle between the velocity vector of the incoming particle and the magnetic field direction. As the particles propagate through the atmosphere, collisions with atmospheric molecules generate secondary particles such as neutrons, protons, photons, muons, pions, and other exotic particles. These secondary particles often have sufficient energy to undergo additional nuclear interactions, and so on, forming what is known as a cascade shower. The particles in the cascade shower are constantly undergoing energy losses due to interactions, especially charged particles whose paths are bent into helical trajectories around magnetic field lines. These energy losses make the integral flux of the cosmic background at the ground level highly dependent on the atmospheric depth traveled by the incident particle and its progeny, and therefore dependent on the elevation of the ground where the cascade deposits its energy. For relatively small depth changes of approximately 100 g/cm{sup 2} or less, the ground-level cosmic flux follows an exponential model. The tabulation of background particle fluxes on the surface of the earth is important for a variety of reasons, one of which is the design of nuclear material detection systems. This method is also useful in producing user-specified effective attenuation lengths for particles. Attenuation lengths are an essential parameter used to compare particle spectra from different locations, elevations, and energy ranges. Traditionally these attenuation lengths are inferred from measured data. Now with the ability to simulate an accurate galactic cosmic ray source and transport it and its shower through a modeled atmosphere, these attenuation lengths can be calculated for any set of conditions. (authors)

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2014

Comparing Measured (a,n) Thick Target Yields to MCNP6 using TENDL 2012 Libraries - Paper 31

MCNP6, due to the heritage of MCNPX, has always possessed the capability to transport the light ions, such as alpha particles. Two types of competing physics take place during the particle transport: atomic and nuclear physics. Charged particles lose energy to media due to Coulombic interactions. MCNP6 models Coulombic interactions through the use of the continuous slowing down approximation. Stopping powers are calculated at a variety of energies using the Bethe-Bloch formula with Sternheimer and Peierls density effect correction. For energies less than 1.31 MeV/AMU, a low-energy SPAR model is used. When the energy of the particle is high enough, there is a transition from nuclear interactions to nucleon interactions. MCNP6 possess various physics models to account for these types of intra-nuclear cascades. These models are less accurate at low energies. Even though MCNP6 will transport these light ions to low energies (i.e. less than 20 MeV), most of the available model physics in MCNP6, for nuclear interactions, has a suggested minimum energy of ∼20-200 MeV. The result of the poorer prediction at low energies causes important light ion events, such as F-19(a,n)Na-22 from U-234 decay in UF{sub 6}, to produce orders of magnitude fewer neutrons in MCNP6 than should be emitted. As with neutrons, ACE (A Compact ENDF) formatted light ion libraries have been developed from several different evaluations. These libraries possess better physics at low energies where the physics models break down. The TENDL 2012 evaluations possess alpha transport libraries, in ACE format, such that they can easily be used with codes like MCNP6 that read ACE formatted cross sections. As a result, a capability was developed for MCNP6 to read and implement these library cross sections when the light ion reached energies below the library upper energy cutoff. Perfectly modelling an experiment for benchmark comparisons to a code calculation can be tricky. For complex setups, systematic error is always possible; and usually a hefty uncertainty analysis is required to quantify just the geometry setup and material combination choices. For an (a,n) interaction, a thick target yield is defined as the amount of neutrons emitted given that the alpha ranges out in the material. For alpha transport at low energy <10 MeV, in reasonably dense materials (>1-2 g/cc), with a material dimension length larger than a few hundred microns, alpha particles will tend to completely stop and range out in the material. For Li, Be, B, C, N, O, F, Na, Mg and Al thick target yields exist in 200 keV intervals from 1 to 9.8 MeV (most of the elements do have (a,n) thresholds above 1 MeV). Because modelling a thick target yield in MCNP6 is much easier than attempting to account for the complexity of systematic error from a full out experiment, we present here comparisons of (a,n) thick target yields to those calculated in MCNP6 using the TENDL 2012 alpha library evaluations.

61 RADIATION PROTECTION AND DOSIMETRY↗
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The Preliminary Verification of the 2-D Transport Module of Radiation Shielding Code ARES - Paper 33

ARES is a multi-group SN particle transport code system with arbitrary order anisotropic scattering. It is primarily used for the nuclear systems' shielding calculation and analysis in the case of one dimensional, two dimensional and three dimensional systems. ARES2D is the 2-D transport module of ARES code system. It is used to solve the first-order form of the steady-state Boltzmann transport equation in the two dimensional geometry with uniform rectangular mesh. The equation's energy dependence is modeled using multiple energy groups. The angular dependence is modeled using discrete ordinates method. The spatial variable is modeled using finite differencing scheme. The overall solution technique is source iteration with Partial Current Rebalance acceleration of the scattering source. ARES2D is capable of calculating k eigenvalue and fixed source problem with arbitrary order anisotropic scattering in two dimensional geometry. (authors)

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2014

MCNP6 Gets Correlated with CGM 3.4 - Paper 41

Version 3.4 of the Cascading Gamma-ray Multiplicity (CGM) code has been integrated into version 6 of the Monte Carlo radiation transport code MCNP. The current release is MCNP6 which is the merger of MCNPX 2.7.0 and MCNP5. MCNP6 is a general purpose Monte Carlo radiation transport code that can transport 34 different particles and heavy ions. CGM was originally integrated into MCNPX 2.7.0 as an undocumented feature, but was never released to the public. With the integration of CGM into MCNP6 the code now has the capability of producing correlated secondary particle/gamma emissions and should be available in future release of MCNP. Before the integration of CGM, MCNP only used libraries that contain the total multiplicity but not the multiplicity distribution or the correlations. Lacking this information requires the code to use integer sampling to determine the multiplicity. In the example of thermal neutron capture on Cl-35, presented in the results section, MCNP does integer sampling. This means MCNP would produce one gamma 17% of the time and two gammas 82% giving a mean multiplicity of 1.83. Using CGM, MCNP can now produce a spectrum of multiplicities. The use of CGM has several advantages over the use of ACE data libraries in MCNP6: 1) it has correlated gamma production, 2) it produces a distribution of gamma multiplicities per interaction, whereas ACE produces an average between two integer multiplicities, 3) it is a theoretical model that may predict experimentally unmeasured physical quantities that may not be in experimental based data libraries. CGM conserves energy and angular momentum internally to CGM; however, angular emission information is not currently sent back to MCNP6. Future work will include the integration of correlated neutrons from CGM. Also, the reaction type will be passed to CGM to set the spin and parity for a particular reaction when possible. (authors)

61 RADIATION PROTECTION AND DOSIMETRY↗
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Computational Assessment of Naturally Occurring Background Radiation Produced by Extraterrestrial Sources - Paper 43

In recent years there has been increased interest in developing the ability to account for background radiation in real (or near-real) time in support of end-user field experiments or operational search missions. More specifically, the focus has been on creating an array of tools that can provide an estimate of the natural occurring radiation that an experimentalist or search operator can use in the field in real time while making a measurement. Conceptually, these tools would allow the experimentalist or search operator to be able to dynamically account for the background associated with the measurement resulting in significant improvement in the probability of detection and minimization of the false alarm rate. The purpose of this work is to assess the extraterrestrial sources that produce naturally occurring background, the variation of these sources, and the variation in the background due to the sources and the potential measurement environment. This work is part of a larger project that will also assess terrestrial sources of background radiation. The focus of this paper is only the neutron background produced by extraterrestrial sources. The computational assessment will be performed using MCNP6. (authors)

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2014

Study of {sup 31}P(n,α){sup 28}Al reaction using d(d,n){sup 3}He for the possible interference with {sup 27}Al(n,γ){sup 28}Al - Paper 44

At McMaster we use neutrons from the {sup 7}Li(p,n){sup 7}Be reaction to measure aluminum in human subjects via the {sup 27}Al(n,γ){sup 28}Al reaction. We use proton energies of between 1.9 MeV and 2.3 MeV, yielding maximum neutron energy of 0.575 MeV. In principle, {sup 28}Al can also be produced via the {sup 31}P(n,α){sup 28}Al reaction, which has a neutron energy threshold of 2.008 MeV. When aluminum is measured in bone, there can be 10{sup 5} times the mass of phosphorus present as there is aluminum. However, this potential interference cannot arise using the {sup 7}Li(p,n){sup 7}Be reaction in the proton energy range specified. Collaborators at Atomic Energy of Canada Limited (AECL), Chalk River, Purdue University in the USA and Shanxi Medical University in Taiyuan, China are also interested in being able to measure aluminum in bone in living human subjects. Both AECL and Purdue have acquired D-D sets (accelerators producing neutrons via the d(d,n){sup 3}He reaction). These operate at high current (∼mA) and relatively low accelerating voltage (150 kV), relying on quantum mechanical tunneling through the Coulomb barrier. These D-D sets produce neutrons of energy 2.565 MeV. This is above the threshold for the {sup 31}P(n,α){sup 28}Al reaction, but we have not been able to find any report in which this reaction has had a cross section measurably greater than zero for neutron energies of less than 3 MeV. So, it is an open question as to whether the potential interfering reaction with phosphorus is a concern when using neutrons from a D-D set to measure aluminum in bone. (authors)

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2014

On the Ill-Posed Nature of Inverse Problems: An Exactly Solvable Paradigm Inverse Neutron Diffusion Problem Illustrating the Solution's Non-Computability - Paper 48

In general, a physical system and/or the result of an indirect experimental measurement are modeled mathematically in terms of: (a) A system of linear and/or nonlinear equations that relate the system's independent variables and parameters to the system's state (i.e., dependent) variables; (b) Probability distributions, moments thereof, inequality and/or equality constraints that define the range of variations of the system's parameters; (c) One or several quantities, customarily referred to as system responses (or objective functions, or indices of performance), which are computed using the mathematical model. The 'direct problem' solves the 'parameter-to-output' mapping that describes the 'cause-to-effect' relationship in the respective physical process. In practice, the model parameters and/or responses are experimentally measured quantities and are therefore subject to uncertainties. Many measurement problems are 'inverse' to the direct problem in that they seek to determine the properties of the medium (i.e., various cross sections), or the properties of the source, and/or the size of the medium on its boundaries from measurements of quantities that depend on the neutron flux (or from direct measurements of the neutron flux). In general, two problems are called inverses of one another if the formulation of each involves all or part of the solution of the other. Some authors further group such inverse problems into 'invasive', when the interior flux is accessible for measurements as opposed to 'non-invasive', in which only fluxes on the boundary of (or exterior to) the medium can be measured. Examples of inverse radiative transfer problems have been reviewed by McCormick, while examples of inverse source problems for neutron transport have been provided by Sanchez and McCormick. More recently, Bledsoe et al., have presented two methods (the Levenberg-Marquard, and 'differential evolution method') for addressing numerically inverse transport problems. However, although these works (as well as other works cited therein) mentioned the detrimental effects of 'small perturbations in the input data', none of the works published thus far on inverse problems for particle transport or diffusion actually showed explicitly the effects that uncertainties in the measured responses (e.g., detector measurements) would actually have on the 'inverse' deduction (from such measurements) of the underlying medium's properties, sources, etc. This work highlights explicitly the nature of the difficulties encountered in so-called 'inverse problems' by presenting a paradigm inverse neutron diffusion problem admitting an exact solution, in which the 'inverse problem' is to determine the medium's diffusion coefficient from flux measurements. It will be shown explicitly (analytically) that this is not possible to achieve in a unique manner, even in the absence of measurement uncertainties, unless additional error-free information is provided about the quantity to be determined (in this case, the medium's diffusion coefficient). It if further shown that, even if this additional error-free information about the diffusion coefficient to be determined is available, the unavoidable experimental measurements errors, albeit it vanishingly small, in the measured flux render the 'exact solution' useless for computational purposes, since the exact solution turns out to be a divergent series. As discussed in this paper's ENDNOTES section, the paradigm problem presented in this work also underscores the need for 'regularization', which is a qualifier indicating that the inverse problem is to be solved in some approximate manner. (authors)

61 RADIATION PROTECTION AND DOSIMETRY↗
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Parallel Monte Carlo Methods for Heterogeneous Hardware Computer Systems Using GPUs and Coprocessors: Recent Development of ARCHER Code - Paper 54

Traditional methods of accelerating Monte Carlo (MC) calculations typically use CPUs as the computing unit. The parallelism is based upon multithreading on a multi-core processor, and message passing interface (MPI) across many processors and nodes. In recent years, the hardware accelerators have been introduced by the high performance computing industry as an alternative approach to achieve massive parallelism. Hardware accelerators have the advantage of high Floating-point Operations Per Second (FLOPS) and high energy efficiency. Examples include graphics processing units (GPUs) and coprocessors. Compute-intensive programs can be off-loaded to them to be concurrently executed by hundreds or thousands of active threads. Existing production MC codes, however, cannot be directly compiled or executed on the new platforms. It is therefore necessary to rewrite or redevelop the code using the hardware accelerator specific programming models and APIs. Several studies have been conducted on using GPUs to accelerate MC calculations in reactor analysis and medical physics. In this paper, we present the recent progress at Rensselaer Polytechnic Institute (RPI) in developing a new parallel MC code called ARCHER (Accelerated Radiation-transport Computations in Heterogeneous Environments) for coupled photon-electron transport simulations. The code is designed as a test bed to evaluate the efficacy of heterogeneous computing systems for low-cost MC-based radiation dosimetric calculations. The code has three variants corresponding to three parallel hardware platforms: ARCHER-CPU (multi-core CPU), ARCHER-GPU (Nvidia GPU) and ARCHER-COP (Intel Xeon Phi coprocessor). The code has been used for several applications, including CT imaging dosimetry, radiotherapy dosimetry and radiation shielding design. (authors)

61 RADIATION PROTECTION AND DOSIMETRY↗
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SNS Sample Activation Calculator Experimental Validation - Paper 79

The Spallation Neutron Source (SNS) at Oak Ridge National Laboratory (ORNL) uses the Sample Activation Calculator (SAC) to calculate the activation of all samples after they have been exposed to the neutron beam in any of the neutron scattering instruments at the SNS beamlines. The SAC web page takes user inputs (choice of beamline, the mass, composition and area of the sample, irradiation time, decay time, etc.) and then calculates the activation for the sample. In recent years, the SAC has been incorporated into the user proposal and sample handling process, and beamline scientists and users have noticed discrepancies in the predicted activation of their samples. As a part of the Neutronics Analysis Group at SNS, we were tasked with validating the SAC tool and diagnosing the origin of the discrepancies seen by the beamline scientists and the users. In the process of performing these measurements, we additionally found that several aspects of realistic neutron scattering operations, and thus sample exposure, cannot be captured with current configuration of the SAC tool. Typical neutron scattering samples are irradiated for multiple periods with different neutron energy ranges; the realities of accelerator operation indicate that not all of those periods will be at the same accelerator power, and scattering samples are frequently not fully illuminated by the well-collimated neutron beam. Finally, we will explore whether neutron beams, as modified by neutron guide systems with energy-dependent intensity modifications and by precise energy-band defining mechanical choppers, are suitably analyzed with multigroup cross section libraries in which there are typically very few energy groups below one eV. (authors)

43 PARTICLE ACCELERATORS↗
DOE OSTI2014

A new solution for the 1D neutral particle transport equation in heterogeneous media - Paper 89

The equation of photon transport in directionally discretized 2N streams has been the subject of various numerical algorithms over the past 65 years since Chandrasekhar first explored the concept. The importance of this equation cannot be overemphasized, as it characterizes neutral particles (in this case photons) undergoing collisions as characterized by the scattering phase function in a 1D slab medium. More precisely, particles enter a homogeneous slab medium with the particle intensity distribution to be numerically determined. We find applications in light scattering for graphical rendering as well as deep penetration photon transport in a scattering medium as considered here. Noting the algebraic nature of the solution, numerous numerical attempts at an efficient algorithm have appeared in the literature-- many involving numerical linear algebra. Most notable are the Discrete Ordinates Method [DOM] of Stamnes and the Analytical Discrete Ordinates (ADO) method of Siewert as well as the Laplace Transform (LTSN) method of Vilhena. The methods are all centered on the determination of eigenvalues (or zeros of a polynomial in the case of the LTSN method) of a linear system enabling the solution to be represented as a sum of exponential eigenvectors. Because of the unstable nature of the solution, unless one takes care, such a representation leads to numerical catastrophe. Here, we derive an alternative form of solution to overcome this difficulty for application to deep penetration.

61 RADIATION PROTECTION AND DOSIMETRY↗
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Identification of Leakage Spectra from Neutron Activation in Glass/Cherenkov Detectors - Paper 96

Oak Ridge National Laboratory has developed a new detector system concept that can be deployed in major US cities to identify the leakage spectrum of a nuclear weapon detonation. The detector is based on a small piece of glass that has been doped with an element that has a high activation cross section and creates a radioactive activation product with a reasonable length half-life. The decay of the activation product(s) after exposure creates Cherenkov light which can propagate to a photomultiplier and can be turned into a count-rate signal. The signal from each glass in each detector in a city can be individually analyzed to determine the amount of activation products produced in the initial exposure. Several activation products can be determined in a single glass if the half-lives are not the same. With a known source location and a detailed 3D model of a city, neutron transport simulations can be made by using a library of known sources to predict the activation amounts each source would have made in the glass detectors. A simple comparison of the measured activation amounts and the predicted activation amounts can identify the source spectrum contained in the library that is the best match. Choosing dopants with a large range in the threshold energy for the activation reactions can help to differentiate the spectra in the library. The method of matching the measured activation amounts to the predicted activation amounts to identify the source spectrum is similar to the analysis used for the passive Activation Foil Integrated Detector System (AFIDS), also developed at Oak Ridge National Laboratory. In AFIDS, wires or foils of known mass and isotopic composition are deployed before and then collected after exposure. Counting gammas in a nearby laboratory determines the activation amounts. The glass/Cherenkov system is active - it uses a battery-powered detector to count flashes of Cherenkov light and to communicate its count rates to a remote location for further analysis. Work has been performed in several areas, including glass making, automated detection and communication systems and software tools for analysis of the signal and source spectrum identification. Bell and Boatner have developed phosphate glasses that incorporate elements with high capture cross sections, and Hayward has simulated the response with Geant4 and measured the response with isotopic gamma sources. This paper will focus on the software tools for determining the activation amounts from the Cherenkov count-rate signal and finding the best matching source spectrum. A test problem in an urban environment is used to demonstrate the potential of a spectrum identification system using neutron activation. The analysis methods developed to identify the neutron leakage spectra from Cherenkov light given off by neutron activation in doped glasses work well. The calculation used to unfold the count-rate information into the number of actual activation products uses robust algorithms that can also compute associated uncertainties. These uncertainties and the known uncertainties of the glass tube material compositions are incorporated into the matching analysis to determine which of the source spectra in the library best matches the measured activation data. Any measurement method used to identify a neutron source spectrum relies at some point on a comparison against known sources; no identification method is definitive. Detection systems measure some physical aspect of the source, not its serial number. Many systems measure the physical aspect to determine that the source has some property, and then leave it to experts to determine which potential sources share that property. The analysis system presented here incorporates the comparison into the analysis itself. The statistical tests show which of the known library sources match the measured data the best and whether that match is a good match. If the best match still does not fit the measured data well (a very high χ{sup 2} value), that indicates the true source is not one of the sources in the library. This is still valuable information for decision makers. (authors)

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2014

Geant4 and beyond: recent progress in precision physics modeling - Paper 100

This extended abstract briefly summarizes ongoing research activity on the evaluation and experimental validation of physics methods for photon and electron transport. The analysis includes physics models currently implemented in Geant4 as well as modeling methods used in other Monte Carlo codes, or not yet considered in general purpose Monte Carlo simulation systems. The validation of simulation models is performed with the support of rigorous statistical methods, which involve goodness-of-fit tests followed by categorical analysis. All results are quantitative, and are fully documented. (authors)

61 RADIATION PROTECTION AND DOSIMETRY↗
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CINDER08 and Scripting Environment for Accelerator Activation Problems - Paper 101

CINDER is a transmutation code, developed originally at the Bettis Atomic Power Laboratory, which predicts induced radioactivity due to either reactor or accelerator-driven radiation sources. The code and associated data libraries have been expanded and improved over the years at Los Alamos National Laboratory. CINDER08 is the latest update in the series and represents a complete rewrite of CINDER, improving it in many areas. Beginning with CINDER90, several Perl scripts were developed to treat multi-cell problems in combination with the radiation transport code MCNPX. These scripts simplified the preparation of CINDER input and made the code much more user-friendly. In concert with the development of CINDER08, these scripting tools have also been updated. The script developers are intensive users of CINDER and had significant input into the development of CINDER08. CINDER08, through use of these scripts, also works with output from MCNP5, which is useful for problems that do not require the extended particle capabilities of MCNPX and do not require particle energies outside of the tabular physics regime, and with MCNP6. A new version of CINDER, CINDER08, incorporates many improvements to the coding and the data libraries. Updated versions of scripting tools improve the code's usability and simplify the use of CINDER output as the basis of gamma source terms for subsequent calculations. (authors)

61 RADIATION PROTECTION AND DOSIMETRY↗
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Optimal Shielding Design For Minimum Materials Cost or Mass - Paper 128

Material costs dominate some shielding design problems. This is certainly the case for manned nuclear power space applications for which shielding is essential and the cost of launching by rocket from Earth is high. In such situations or in those where shielding volume or mass is constrained, it is important to optimize the design. Although trial and error synthesis methods may succeed, a more systematic approach is warranted. Design automation may also potentially reduce engineering costs. The objective is to automate the design of optimal radiation shielding to meet specified attenuation requirements given a menu of allowed shielding materials, their properties and their delivered costs per unit volume. The cost-optimal design is one which, at each location in the shield, chooses the menu material minimizing a particular functional of forward and adjoint angular fluxes plus material cost. Thus a Boltzmann forward and adjoint angular flux solver with additional calculations can evaluate whether a shield design is optimal or whether improvements are possible. An iterative algorithm based on approximated optimality is proposed to solve for the optimal shield design and numerical experiments with it are discussed. The future use of 3D Boltzmann solvers will allow extending results to shaped shield designs in which material is not wasted to maintain unneeded spherical symmetry. The largest cost savings of automated optimal design of shielding is expected to be for applications involving nuclear powered manned space missions where the required shielding attenuation of radiation and costs are both large. However, less exotic applications such as the shielding of radiation facilities on Earth could also benefit from optimal shielding design, and automation of the design process may reduce engineering costs. (author)

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2014

Efficiency Calibration of 4{pi} NaI(Tl) Gamma-ray Detectors for Plane Sources - Paper 130

4{pi} NaI(Tl) detectors are useful for low level gamma activity measurements due to the near 4{pi} solid angle that can be obtained with them. An analytical theoretical approach is presented here to calculate the absolute efficiencies for this type of detectors. The approach depends on the accurate calculation of two important factors; the path length, d, the photon traverses within the active volume of a gamma detector, and the geometrical solid angle,{Omega}, subtended by the source to the detector at the point of entrance. These two factors are theoretically derived through a straightforward analytical approach. In this work, this model is extended to an isotropic radiating point source, and a thin circular disk source. The results of the present absolute efficiency values were compared with the experimental and M.C. method works reported in the literature. The present theoretical and the experimental efficiency values are in good agreement. The present approach offers a good method to calibrate 4{pi} NaI(Tl) detectors over a large energy range without the need for standard sources, as is the case for experimental methods, nor optimization of detector parameters as for Monte Carlo simulations. Discrepancies with experimental results are reduced to lower than 4%. (author)

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2014

Fragmentation Modeling for Space Applications: Charge Dispersion Effects - Paper 136

Fragmenting of high-energy heavy ions in nuclear collisions is an important process in space radiation transport, shielding and risk assessment since the secondary particles produced by these collisions have ranges greater than their parent nucleus, and are damaging to humans and spacecraft components. This work uses a quantum-mechanical optical potential abrasion-ablation collision-modeling concept in which nucleons in both the projectile and target are removed by a combination of knockout and evaporation processes in these collisions. In the knockout stage, the collision results in some of the struck projectile nucleons exiting the nucleus and leaving an excited pre-fragment, which then deexcites by a emitting photons, nucleons and light ions by a cascade-evaporation process. The main thrust of this work is to investigate differences in isotope and element production cross sections as a result of utilizing two very different models of charge-to mass ratios for the projectile pre-fragments produced by the abrasion/knockout process. (authors)

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