Avneet Sood

MCNP™ Version 5

Abstract The Monte Carlo transport workhorse, MCNP [Los Alamos National Laboratory report LA-13709-M, 2000], is undergoing a massive renovation at Los Alamos National Laboratory (LANL) in support of the Eolus Project of the Advanced Simulation and Computing (ASCI) Program. MCNP 1 Version 5 (V5) (expected to be released to RSICC in Fall 2002) will consist of a major restructuring from FORTRAN-77 (with extensions) to ANSI-standard FORTRAN-90 [American National Standard for Programming Language – Fortran-Extended, ANSI X3. 198-1992, 1992] with support for all of the features available in the present release (MCNP-4C2/4C3). To most users, the look-and-feel of MCNP will not change much except for the improvements (improved graphics, easier installation, better online documentation). For example, even with the major format change, full support for incremental patching will still be provided. In addition to the language and style updates, MCNP V5 will have various new user features. These include improved photon physics, neutral particle radiography, enhancements and additions to variance reduction methods, new source options, improved parallelism support (PVM, MPI, OpenMP), and new nuclear and atomic data libraries.

Analytical Benchmark Test Set for Criticality Code Verification

A number of published numerical solutions to analytic eigenvalue (k{sub eff}) and eigenfunction equations are summarized for the purpose of creating a criticality verification benchmark test set. The 75-problem test set allows the user to verify the correctness of a criticality code for infinite medium and simple geometries in one- and two-energy groups, one- and two-media, and both isotropic and linearly anisotropic neutron scattering. A three- and six-energy group infinite medium problem are also included in the test set. The problem specifications will produce both k{sub eff}=1 and the quoted k{sub {infinity}} to at least five decimal places. Additional uses of the test set for code verification are also discussed. Los Alamos report LA-13511 contains the details of all 75 test problems.

A new Monte Carlo assisted approach to detector response functions

The physical mechanisms that describe the components of NaI, Ge and Si(Li) detector responses have been investigated using Monte Carlo simulation. The mechanisms described focus on the shape of the Compton edge, the magnitude of the flat continuum, and the shape of the exponential tails features. These features are not accurately predicted by previous Monte Carlo simulations. Probable interaction mechanisms for each detector response component are given based on this Monte Carlo simulation.

Single peak versus library least-squares analysis methods for the PGNAA analysis of vitrified waste

Abstract A feasibility study (Gardner et al., 1995) for the neutron capture prompt gamma-ray analysis (PGNAA) of simulated vitrified nuclear waste containing SiO2/Al2O3 and SiO2/Fe2O3 mixtures has been carried out. Analyses for elements Al, Si, and Fe were made by both the single peak area and the library least-squares methods using measured and simulated data. It is shown that the library least-squares method is about 2.5 times better than the single peak analysis method.

Verification of the pulse height tally in MCNP 5

Abstract Pulse height tallies are commonly used in Monte Carlo codes to predict detailed measured photon spectra for spectrometry purposes. The pulse height tally is unique among the various tallies in MCNP. Unlike flux or current tallies, which are calculated as soon as the particle exits or collides in the cell, the entire set of tracks for a history must be completed before the pulse height tally can be made. The objective of this work was to verify the pulse height tally and prepare to verify the new MCNP 5 variance reduction features with the pulse height tally. In this paper, we give details to the analytic solution of the pulse height distribution using a modification to Shuttleworth’s fictitious elements, report MCNP 5 results for the pulse height tally, energy deposited and current tallies for the problem.

Steady neutron source measurement method for Σa and Σs in geological samples

An improved experimental method, using a steady neutron source in a moderating medium, has been developed to determine thermal neutron absorption and scattering cross sections for bulk geological media using discrete samples. The system design has been optimized and experimental results have been benchmarked using Monte Carlo simulation. Studies have been performed to improve the measurement sensitivity and reproducibility over previous designs. A semi-empirical model has been developed for determining both absorption and scattering cross sections of the sample.

Correlated Prompt Fission Data in Transport Simulations

Abstract.Detailed information on the fission process can be inferred from the observation, modeling and theoretical understanding of prompt fission neutron and

On the future of Monte Carlo simulation for nuclear logs

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Monte Carlo aided treatments of the nonlinear inverse PGNAA measurement problem for various continuous on-line applications

γ-ray observables. Beyond simple average quantities, the study of distributions and correlations in prompt data, e.g., multiplicity-dependent neutron and

Feasibility of neutron activation methods for measurement of sodium and aluminium in green liquor

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