Jeffrey O. Johnson

ADVANTG An Automated Variance Reduction Parameter Generator

The primary objective of ADVANTG is to reduce both the user effort and the computational time required to obtain accurate and precise tally estimates across a broad range of challenging transport applications. ADVANTG has been applied to simulations of real-world radiation shielding, detection, and neutron activation problems. Examples of shielding applications include material damage and dose rate analyses of the Oak Ridge National Laboratory (ORNL) Spallation Neutron Source and High Flux Isotope Reactor (Risner and Blakeman 2013) and the ITER Tokamak (Ibrahim et al. 2011). ADVANTG has been applied to a suite of radiation detection, safeguards, and special nuclear material movement detection test problems (Shaver et al. 2011). ADVANTG has also been used in the prediction of activation rates within light water reactor facilities (Pantelias and Mosher 2013). In these projects, ADVANTG was demonstrated to significantly increase the tally figure of merit (FOM) relative to an analog MCNP simulation. The ADVANTG-generated parameters were also shown to be more effective than manually generated geometry splitting parameters.

Proton Induced Activation in Mercury: Comparison of Measurements and Calculations

Measurements and simulations of the proton beam interaction with the mercury target were performed to support Spallation Neutron Source design. Due to the abundance of isotopes produced in mercury, the long delay between the irradiation and the measurements, and the self-shielding of the mercury sample, the measurements were difficult to perform and the activities of several isotopes have large uncertainties. Calculations predicted the activities of the most reliably measured isotopes within 20%–40%; however, some large discrepancies were observed for some isotopes for which the measurements were considered less reliable. Predicted dose rates were in very good agreement with the measurements.

Target Systems Overview for the Spallation Neutron Source

The purpose and requirements of target systems as well as the technologies that are being utilized to design and build a state-of-the-art neutron spallation source, the Spallation Neutron Source, are discussed. Emphasis is given to the technology issues that present the greatest scientific challenges. The present facility configuration, ongoing analysis, and planned hardware research and development program are also described.

Initial neutronic target station studies for the national spallation neutron source (NSNS)

Abstract Results found during initial NSNS target station neutronic design efforts are reported including the success of comparing neutron sources at 1 eV and moderator performance normalized to 1 eV. The usefulness of an analytic form is demonstrated. The angular dependence of the neutron current from a moderator face is presented together with the changes in neutron current with variation of moderator width, poison plate location and moderator material. The formation of an equilibrium state at low neutron energy is also discussed.

Computing the three-dimensional distribution of the gamma dose rate in the target service cell of the spallation neutron source using the doors package

Design of the Target Service Cell (TSC) of the Spallation Neutron Source (SNS) requires computing detailed profiles of the gamma-ray dose rate outside its structure in order to abide by exposure limits and plan access control for the facilitys personnel. Three-dimensional spatial distributions of the dose rate inside the TSC are also necessary to optimize the locations of electronic instruments and verify their design criteria at selected sites. For these reasons, in addition to the deep penetration feature typical of shielding calculations, a deterministic transport method solution, namely discrete ordinates, is preferred over Monte Carlo. Even then, the computation is complicated by the large size of the structure, large volume of air (internal void), optical thickness of the enclosing walls, and multiplicity of radiation sources. Furthermore, severe ray effects observed in preliminary calculations throughout the TSC internal cavity that persist in the transport through the concrete walls require special treatment. The computational model for conducting this complex calculation using Oak Ridge National Laboratorys TORT code, with support from peripheral codes in the Discrete Ordinates Oak Ridge System (DOORS), is presented. Successful elimination of primary ray effects via the newly developed three-dimensional uncollided flux and first collision source code GRTUNCL3D is also illustrated.

Systematic Assessment of Neutron and Gamma Backgrounds Relevant to Operational Modeling and Detection Technology Implementation

This report summarizes the findings of a two year effort to systematically assess neutron and gamma backgrounds relevant to operational modeling and detection technology implementation. The first year effort focused on reviewing the origins of background sources and their impact on measured rates in operational scenarios of interest. The second year has focused on the assessment of detector and algorithm performance as they pertain to operational requirements against the various background sources and background levels.

The National Spallation Neutron Source target station: a general overview

The technologies that are being utilized to design and build a state-of-the-art neutron spallation source, the National Spallation Neutron Source (NSNS), are discussed. Emphasis is given to the technology issues that present the greatest scientific challenges. The present facility configuration, ongoing analysis and the planned hardware research and development program are also described.

Sensitivities and Uncertainties Related to Numerics and Building Features in Urban Modeling

Abstract Oak Ridge National Laboratory (ORNL) has been engaged in the development and testing of a computational system that would use a grid of activation foil detectors to provide postdetonation forensic information from a nuclear device detonation. ORNL has developed a high-performance, three-dimensional (3-D) deterministic radiation transport code called Denovo. Denovo solves the multigroup discrete ordinates (SN) equations and can output 3-D data in a platform-independent format that can be efficiently analyzed using parallel, high-performance visualization tools. To evaluate the sensitivities and uncertainties associated with the deterministic computational method numerics, a numerical study on the New York City Times Square model was conducted using Denovo. In particular, the sensitivities and uncertainties associated with various components of the calculational method were systematically investigated, including (a) the Legendre polynomial expansion order of the scattering cross sections, (b) the angular quadrature, (c) multigroup energy binning, (d) spatial mesh sizes, (e) the material compositions of the building models, (f) the composition of the foundations upon which the buildings rest (e.g., ground, concrete, or asphalt), and (g) the amount of detail included in the building models. Although Denovo may calculate the idealized model well, there may be uncertainty in the results because of slight departures of the above-named parameters from those used in the idealized calculations. Fluxes and activities at selected locations from perturbed calculations are compared with corresponding values from the idealized or base case to determine the sensitivities associated with specified parameter changes. Results indicate that uncertainties related to numerics can be controlled by using higher fidelity models, but more work is needed to control the uncertainties related to the model.

Shielding Calculations in Support of the Spallation Neutron Source (SNS) Proton Beam Transport System

Determining the bulk shielding requirements for accelerator environments is generally an easy task compared to analyzing the radiation transport through the complex shield configurations and penetrations typically associated with the detailed Title II design efforts of a facility. Shielding calculations for penetrations in the SNS accelerator environment are presented based on hybrid Monte Carlo and discrete ordinates particle transport methods. This methodology relies on coupling tools that map boundary surface leakage information from the Monte Carlo calculations to boundary sources for one‐, two‐, and three‐dimensional discrete ordinates calculations. The paper will briefly introduce the coupling tools for coupling MCNPX to the one‐, two‐, and three‐dimensional discrete ordinates codes in the DOORS code suite. The paper will briefly present typical applications of these tools in the design of complex shield configurations and penetrations in the SNS proton beam transport system.

Neutronic Design Studies for the Spallation Neutron Source

Neutronics analyses are now in progress to support initial selection of target system design features, materials, geometry, and component sizes for the proposed Spallation Neutron Source. Calculations have been performed to determine the neutron, proton, heavy-ion, and gamma-ray flux spectra as a function of time, energy, and space for the major components of the target station (target, moderators, reflectors, etc.). These analyses have also been performed to establish an initial set of performance characteristics for the neutron source. The methodology, reference performance characteristics, and results of initial optimization studies involving moderator poison plate location, target material performance, reflector performance, moderator position, and premoderator performance for the target system are presented.