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9-Zoom : A One-Dimensional, Multigroup, Neutron Diffusion Theory Reactor Code for the IBM 709

Description: The following document describes the usage and purpose of the neutron diffusion theory reactor program 9-Zoom, a memory-contained program that takes advantage of 709 features such as, for example, preferential order of multiply by zero, and for small problems approaches input-output limitations with excellent convergence properties.
Date: August 25, 1959
Creator: Stone, S. P.; Collins, E. T. & Lenihan, S. R..
Partner: UNT Libraries Government Documents Department

Multigroup Methods for Neutron Diffusion Problems

Description: Abstract: "The age-diffusion is adequate to describe the neutron behavior of a very large class of nonthermal reactors (all except those whose dimensions are comparable to the neutron mean free path). Thus, a convenient means of obtaining an accurate solution to this equation is very useful for general reactor calculations. Methods for reducing the age-diffusion equation to a finite set of coupled ordinary differential equations, called multigroup equations, are described. The relative merits of several alternate schemes are discussed. The multigroup equations may be solved by iterative procedures based on an assumed spatial distribution of the fission source neutrons. In practice the initially assumed source shape is accurate enough so that additional iterations are unnecessary. Analytical and numerical methods for solving the multigroup equations with the assumed source are discussed. The adjoint equations are also reduced to multigroup form, and examples of the adjoint function in obtaining improved reactivity values are given."
Date: August 20, 1953
Creator: Hurwitz, H. & Ehrlich, R.
Partner: UNT Libraries Government Documents Department

Singular perturbation analysis of the neutron transport equation

Description: A singular perturbation technique is applied to the one-speed, one- dimensional neutron transport equation with isotropic scattering. Our technique extends previous singular perturbation applications to higher-order and reduces the transport problem to a series of diffusion theory problems in the interior medium and a series of analytically solvable transport problems in the boundary layers. Asymptotic matching links the two solutions, yielding boundary conditions and a composite expansion valid throughout the media. Our formulation generates an accurate correction for the material interface condition used in global diffusion theory calculations.
Date: July 1, 1996
Creator: Losey, D.C. & Lee, J.C.
Partner: UNT Libraries Government Documents Department

Investigating the Use of 3-D Deterministic Transport for Core Safety Analysis

Description: An LDRD (Laboratory Directed Research and Development) project is underway at the Idaho National Laboratory (INL) to demonstrate the feasibility of using a three-dimensional multi-group deterministic neutron transport code (Attila®) to perform global (core-wide) criticality, flux and depletion calculations for safety analysis of the Advanced Test Reactor (ATR). This paper discusses the ATR, model development, capabilities of Attila, generation of the cross-section libraries, comparisons to experimental results for Advanced Fuel Cycle (AFC) concepts, and future work planned with Attila.
Date: April 1, 2004
Creator: Gougar, H. D. & Scott, D.
Partner: UNT Libraries Government Documents Department

A Complex-Geometry Validation Experiment for Advanced Neutron Transport Codes

Description: The Idaho National Laboratory (INL) has initiated a focused effort to upgrade legacy computational reactor physics software tools and protocols used for support of core fuel management and experiment management in the Advanced Test Reactor (ATR) and its companion critical facility (ATRC) at the INL.. This will be accomplished through the introduction of modern high-fidelity computational software and protocols, with appropriate new Verification and Validation (V&V) protocols, over the next 12-18 months. Stochastic and deterministic transport theory based reactor physics codes and nuclear data packages that support this effort include MCNP5[1], SCALE/KENO6[2], HELIOS[3], SCALE/NEWT[2], and ATTILA[4]. Furthermore, a capability for sensitivity analysis and uncertainty quantification based on the TSUNAMI[5] system has also been implemented. Finally, we are also evaluating the Serpent[6] and MC21[7] codes, as additional verification tools in the near term as well as for possible applications to full three-dimensional Monte Carlo based fuel management modeling in the longer term. On the experimental side, several new benchmark-quality code validation measurements based on neutron activation spectrometry have been conducted using the ATRC. Results for the first four experiments, focused on neutron spectrum measurements within the Northwest Large In-Pile Tube (NW LIPT) and in the core fuel elements surrounding the NW LIPT and the diametrically opposite Southeast IPT have been reported [8,9]. A fifth, very recent, experiment focused on detailed measurements of the element-to-element core power distribution is summarized here and examples of the use of the measured data for validation of corresponding MCNP5, HELIOS, NEWT, and Serpent computational models using modern least-square adjustment methods are provided.
Date: November 1, 2013
Creator: Nigg, David W.; LaPorta, Anthony W.; Nielsen, Joseph W.; Parry, James; DeHart, Mark D.; Bays, Samuel E. et al.
Partner: UNT Libraries Government Documents Department

Corn Pone: A Multigroup, Multiregion Reactor Code

Description: Report covering the development of a set of difference equations for machine solution. In addition, this report serves as user's manual for Corn Pone, which is a code for the Oracle, the Oak Ridge National Laboratory computer.
Date: 1961?
Creator: Kinney, W. E. & Coveyou, R. R.
Partner: UNT Libraries Government Documents Department

Recent advances in neutral particle transport methods and codes

Description: An overview of ORNL`s three-dimensional neutral particle transport code, TORT, is presented. Special features of the code that make it invaluable for large applications are summarized for the prospective user. Advanced capabilities currently under development and installation in the production release of TORT are discussed; they include: multitasking on Cray platforms running the UNICOS operating system; Adjacent cell Preconditioning acceleration scheme; and graphics codes for displaying computed quantities such as the flux. Further developments for TORT and its companion codes to enhance its present capabilities, as well as expand its range of applications are disucssed. Speculation on the next generation of neutron particle transport codes at ORNL, especially regarding unstructured grids and high order spatial approximations, are also mentioned.
Date: June 1, 1996
Creator: Azmy, Y.Y.
Partner: UNT Libraries Government Documents Department

Automated Approach to Quantitative Error Analysis in Neutron Transport Calculations

Description: A method is described how a quantitative measure for the robustness of a given transport theory code for coarse network calculations can be obtained. A code, that performs this task automatically and at only nominal cost, is described and has been implemented for slab geometry. This code generates also user oriented benchmark problems which exhibit the analytic behavior at interfaces.
Date: September 1976
Creator: Bareiss, Erwin H. & Derstine, Keith L.
Partner: UNT Libraries Government Documents Department

Slow Neutron Leakage Spectra from Spallation Neutron Sources

Description: An efficient technique is described for Monte Carlo simulation of neutron beam spectra from target-moderator-reflector assemblies typical of pulsed spallation neutron sources. The technique involves the scoring of the transport-theoretical probability that a neutron will emerge from the moderator surface in the direction of interest, at each collision. An angle-biasing probability is also introduced which further enhances efficiency in simple problems.
Date: February 1980
Creator: Das, Shashikala G.; Carpenter, J. M. & Prael, R. E.
Partner: UNT Libraries Government Documents Department

Time-Independent One-Speed Neutron Transport Equation with Anisotropic Scattering in Absorbing Media

Description: This report treats the time-independent, one-speed neutron transport equation with anisotropic scattering in absorbing media. For nuclear gain operators existence and uniqueness of solutions to the half-space and finite-slab problems are proved in L₂-space. The formulas needed for explicit calculations are derived by the use of perturbation theory techniques.
Date: June 1980
Creator: Hangelbroek, Rutger Jan
Partner: UNT Libraries Government Documents Department

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

Description: The Monte Carlo ~djoint ~ielding Code System, MASH, calculates neutron and gamma- ray environments and radiation protection factors for armored military vehicles, structures, trenches, and other shielding configurations by coupling a forward discrete ordinates air- over-ground transport calculation with an adjoint Monte Carlo treatment of the shielding geometry. Efficiency and optimum use of computer time are emphasized. The code system includes the GRTUNCL and DORT codes for air-over-ground transport calculations, the MORSE code with the GIFT5 combinatorial geometry package for adjoint shielding calculations, and several peripheral codes that perform the required data preparations, transformations, and coupling functions. The current version, MASH v 1.5, is the successor to the original MASH v 1.0 code system initially developed at Oak Ridge National Laboratory (ORNL). The discrete ordinates calculation determines the fluence on a coupling surface surrounding the shielding geometry due to an external neutron/gamma-ray source. The Monte Carlo calculation determines the effectiveness of the fluence at that surface in causing a response in a detector within the shielding geometry, i.e., the "dose importance" of the coupling surface fluence. A coupling code folds the fluence together with the dose importance, giving the desired dose response. The coupling code can determine the dose response as a function of the shielding geometry orientation relative to the source, distance from the source, and energy response of the detector. This user's manual includes a short description of each code, the input required to execute the code along with some helpful input data notes, and a representative sample problem.
Date: October 1998
Creator: Slater, C. O.; Barnes, J. M.; Johnson, J. O. & Drischler, J. D.
Partner: UNT Libraries Government Documents Department

IBM Multigroup Numerical Procedures

Description: Introduction: "This paper presents in formula form IBM computational procedure for averaging cross-section data and for calculating reflected spherical reactors by the linear by the linear multigroup numerical method developed by the Physics Section of the ANP Central Design Group. The adjoint reactor numerical procedural is included."
Date: November 28, 1950
Creator: Holmes, David K. & Schulze, O. A.
Partner: UNT Libraries Government Documents Department

Applications of the 3-D Deterministic Transport Code Attlla for Core Safety Analysis

Description: An LDRD (Laboratory Directed Research and Development) project is ongoing at the Idaho National Engineering and Environmental Laboratory (INEEL) for applying the three-dimensional multi-group deterministic neutron transport code (Attila®) to criticality, flux and depletion calculations of the Advanced Test Reactor (ATR). This paper discusses the model development, capabilities of Attila, generation of the cross-section libraries, and comparisons to an ATR MCNP model and future.
Date: October 1, 2004
Creator: Lucas, D. S.
Partner: UNT Libraries Government Documents Department

Singular perturbation applications in neutron transport

Description: This is a paper on singular perturbation applications in neutron transport for submission at the next ANS conference. A singular perturbation technique was developed for neutron transport analysis by postulating expansion in terms of a small ordering parameter {eta}. Our perturbation analysis is carried, without approximation, through {Omicron}({eta}{sup 2}) to derive a material interface correction for diffusion theory. Here we present results from an analytical application of the perturbation technique to a fixed source problem and then describe and implementation of the technique in a computational scheme.
Date: September 1, 1996
Creator: Losey, D.C. & Lee, J.C.
Partner: UNT Libraries Government Documents Department