Richard S. Wittman

Multiple-estimate Monte Carlo calculation of the dose rate constant for a cesium-131 interstitial brachytherapy seed

The purpose of this study was to calculate a more accurate dose rate constant for the 131Cs (model CS-1, IsoRay Medical, Inc., Richland, WA) interstitial brachytherapy seed. Previous measurements of the dose rate constant for this seed have been reported by others with incongruity. Recent direct measurements by thermoluminescence dosimetry and by gamma-ray spectroscopy were about 15% greater than earlier thermoluminescence dosimetry measurements. Therefore, we set about to calculate independent values by a Monte Carlo approach that combined three estimates as a consistency check, and to quantify the computational uncertainty. The calculated dose rate constant for the 131Cs seed was 1.040 cGy h(-1) U(-1) for an ionization chamber model and 1.032 cGy h(-1) U(-1) for a circular ring model. A formal value of 2.2% uncertainty was calculated for both values. The range of our multiestimate values were from 1.032 to 1.061 cGy h(-1) U(-1). We also modeled three 125I seeds with known dose rate constants to test the accuracy of this studys approach.

Verifying the presence of low levels of neptunium in a uranium matrix with electron energy-loss spectroscopy

This paper examines the problems associated with analysis of low levels of neptunium in a uranium matrix with electron energy-loss spectroscopy (EELS) on the transmission electron microscope (TEM). The detection of neptunium in a matrix of uranium can be impeded by the occurrence of a plural scattering event from uranium (U-M(5)+U-O(4,5)) that results in severe overlap on the Np-M(5) edge at 3665 eV. Low levels of Np (1600-6300 ppm) can be detected in a uranium solid, uranophane [Ca(UO(2))(2)(SiO(3)OH)(2)(H(2)O)(5)], by confirming that the energy gap between the Np-M(5) and Np-M(4) edges is at 184 eV and showing that the M(4)/M(5) ratio for the neptunium is smaller than that for uranium. The Richardson-Lucy deconvolution method was applied to energy-loss spectral images and was shown to increase the signal to noise ratio.

THE COUPLING OF A DETERMINISTIC TRANSPORT FIELD SOLUTION TO A MONTE CARLO BOUNDARY CONDITION FOR THE SIMULATION OF LARGE GAMMA-RAY SPECTROMETERS

Abstract Monte Carlo methods are typically used for simulating radiation fields around gamma-ray spectrometers and pulse-height tallies within those spectrometers. Deterministic codes that discretize the linear Boltzmann transport equation can offer significant advantages in computational efficiency for calculating radiation fields, but stochastic codes remain the most dependable tools for calculating the response within spectrometers. For a deterministic field solution to become useful to radiation detection analysts, it must be coupled to a method for calculating spectrometer response functions. This coupling is done in the RADSAT toolbox. Previous work has been successful using a Monte Carlo boundary sphere around a handheld detector. It is desirable to extend this coupling to larger detector systems such as the portal monitors now being used to screen vehicles crossing borders. Challenges to providing an accurate Monte Carlo boundary condition from the deterministic field solution include the greater possibility of large radiation gradients along the detector and the detector itself perturbing the field solution, unlike smaller detector systems. The method of coupling the deterministic results to a stochastic code for large detector systems can be described as spatially defined rectangular patches that minimize gradients. The coupled method was compared to purely stochastic simulation data of identical problems, showing the methods produce consistent detector responses while the purely stochastic run times are substantially longer in some cases, such as highly shielded geometries. For certain cases, this method has the ability to faithfully emulate large sensors in a more reasonable amount of time than other methods.

Bayesian Radiation Source Localization

Abstract Locating illicit radiological sources using gamma-ray or neutron detection is a key challenge for both homeland security and nuclear nonproliferation. Localization methods using an array of detectors or a sequence of observations in time and space must provide rapid results while accounting for a dynamic attenuating environment. In the presence of significant attenuation and scatter, more extensive numerical transport calculations in place of the standard analytical approximations may be required to achieve accurate results. Numerical adjoints based on deterministic transport codes provide relatively efficient detector response calculations needed to determine the most likely location of a true source given a set of observed count rates. Probabilistic representations account for uncertainty in the source location resulting from uncertainties in detector responses and the potential for nonunique solutions. A Bayesian approach improves on previous likelihood methods for source localization by allowing the incorporation of all available information to help constrain solutions. We present an approach to localizing radiological sources that uses numerical adjoints and a Bayesian formulation and demonstrate the approach on two simple example scenarios. Results indicate accurate estimates of source locations. We briefly study the effect of neglecting the contribution of all scattered radiation in the adjoints, as analytical transport approximations do, for a case with moderately attenuating material between detectors and sources. The source location accuracy of the uncollided-only solutions appears to be significantly worse at the source strength considered here, suggesting that the higher physical fidelity that is provided by full numerical adjoint-based solutions may provide an advantage in operational settings.

Novamene: A new class of carbon allotropes

We announce a new class of carbon allotropes. The basis of this new classification resides on the concept of combining hexagonal diamond (sp3 bonded carbon − lonsdaleite) and ring carbon (sp2 bonded carbon − graphene). Since hexagonal diamond acts as an insulator and sp2 bonded rings act as conductors, these predicted materials have potential applications for transistors and other electronic components. We describe the structure of a proposed series of carbon allotropes, novamene, and carry out a detailed computational analysis of the structural and electronic properties of the simplest compound in this class: the single-ring novamene. In addition, we suggest how hundreds of different allotropes of carbon could be constructed within this class.

Design and Testing of a Boron Carbide Capsule for Spectral Tailoring in Mixed-Spectrum Reactors

A boron carbide capsule has been designed and used for spectral-tailoring experiments at the TRIGA reactor at Washington State University. Irradiations were conducted in pulsed mode and in continuous operation for up to 4 h. A cadmium cover was used to reduce thermal heating. The neutron spectrum calculated with the Monte Carlo N-particle transport code was found to be in good agreement with reactor dosimetry measurements using the STAY’SL computer code. The neutron spectrum resembles that of a fast reactor. The design of a capsule using boron carbide fully enriched in 10B shows that it is possible to produce a neutron spectrum similar to that of 235U fission.

Development of a Chemistry-Based, Predictive Method for Determining the Amount of Non-Pertechnetate Technetium in the Hanford Tanks: FY 2012 Progress Report

This report describes investigations directed toward understanding the extent of the presence of highly alkaline soluble, non-pertechnetate technetium (n-Tc) in the Hanford Tank supernatants. The goals of this report are to: a) present a review of the available literature relevant to the speciation of technetium in the Hanford tank supernatants, b) attempt to establish a chemically logical correlation between available Hanford tank measurements and the presence of supernatant soluble n-Tc, c) use existing measurement data to estimate the amount of n-Tc in the Hanford tank supernatants, and d) report on any likely, process-friendly methods to eventually sequester soluble n-Tc from Hanford tank supernatants.

Coupling the Mixed Potential and Radiolysis Models for Used Fuel Degradation

The primary purpose of this report is to describe the strategy for coupling three process level models to produce an integrated Used Fuel Degradation Model (FDM). The FDM, which is based on fundamental chemical and physical principals, provides direct calculation of radionuclide source terms for use in repository performance assessments. The G-value for H2O2 production (Gcond) to be used in the Mixed Potential Model (MPM) (H2O2 is the only radiolytic product presently included but others will be added as appropriate) needs to account for intermediate spur reactions. The effects of these intermediate reactions on [H2O2] are accounted for in the Radiolysis Model (RM). This report details methods for applying RM calculations that encompass the effects of these fast interactions on [H2O2] as the solution composition evolves during successive MPM iterations and then represent the steady-state [H2O2] in terms of an “effective instantaneous or conditional” generation value (Gcond). It is anticipated that the value of Gcond will change slowly as the reaction progresses through several iterations of the MPM as changes in the nature of fuel surface occur. The Gcond values will be calculated with the RM either after several iterations or when concentrations of key reactants reach threshold values determined from previous sensitivity runs. Sensitivity runs with RM indicate significant changes in G-value can occur over narrow composition ranges. The objective of the mixed potential model (MPM) is to calculate the used fuel degradation rates for a wide range of disposal environments to provide the source term radionuclide release rates for generic repository concepts. The fuel degradation rate is calculated for chemical and oxidative dissolution mechanisms using mixed potential theory to account for all relevant redox reactions at the fuel surface, including those involving oxidants produced by solution radiolysis and provided by the radiolysis model (RM). The RM calculates the concentration of species generated at any specific time and location from the surface of the fuel. Several options being considered for coupling the RM and MPM are described in the report. Different options have advantages and disadvantages based on the extent of coding that would be required and the ease of use of the final product.

Transport Test Problems for Hybrid Methods Development

This report presents 9 test problems to guide testing and development of hybrid calculations for the ADVANTG code at ORNL. These test cases can be used for comparing different types of radiation transport calculations, as well as for guiding the development of variance reduction methods. Cases are drawn primarily from existing or previous calculations with a preference for cases which include experimental data, or otherwise have results with a high level of confidence, are non-sensitive, and represent problem sets of interest to NA-22.

The Formation and Modeling of Colloids from the Corrosion of Nuclear Waste Forms

This paper describes a model for determining the stability and associated radionuclide concentrations of colloids that might be present in the nuclear waste package environment from degradation of the nuclear waste forms. The model simplifies radionuclide–colloid behavior by assuming that all colloids can be defined as either smectite clay, a mixed actinide-bearing rare earth-zirconium oxide, iron oxyhydroxide (ferrihydrite {FeOOH}, or uranophane {Ca(UO2)2(SiO3OH)2(H2O)5}. However, for the purposes of predictive stability modeling, the colloids are conceptually represented as montmorillonite, ZrO2, hematite, and meta autunite, respectively. The model uses theoretical calculations and laboratory data to determine the stability of modeled colloids with ionic strength and pH. The true nature of colloid composition and heterogeneity, generation, and flocculation will be extremely complex, involving the formation of numerous types of phases, often depending on the composition of the various waste forms and waste package materials. This model strives to capture the uncertainty of the real system using theoretical models. In this paper, one of the four representative colloids designed to capture the behavior of the spent fuel derived colloids is described in detail.