ENDF/B-VII.1 Nuclear Data for Science and Technology: Cross Sections, Covariances, Fission Product Yields and Decay Data

https://digital.library.unt.edu/ark:/67531/metadc888955/m2/1/high_res_d/900147.pdf

ENDF/B-VII.0: Next Generation Evaluated Nuclear Data Library for Nuclear Science and Technology

https://escholarship.org/content/qt2dj9p3h4/qt2dj9p3h4.pdf?t=pmi3bw

ENDF/B-VIII.0: The 8th Major Release of the Nuclear Reaction Data Library with CIELO-project Cross Sections, New Standards and Thermal Scattering Data

Modern Nuclear Data Evaluation with the TALYS Code System

Ordinary differential equations of the first order linear differential equations complex numbers and linear algebra simultaneous linear differential equations numerical methods the descriptive theory of nonlinear differential equations mechanical systems and electric circuits Fourier series Fourier integrals and Fourier transforms the Laplace transformation partial differential equations Bessel functions and Legendre polynomials applications and further properties of matrices vector analysis the calculus of variations analytic functions of a complex variable infinite series in the complex plane the theory of residues conformal mapping.

Advanced Engineering Mathematics. By C. R. Wylie. Pp. xiv, 813. 1966. (McGraw-Hill.)

MCNP6 is simply and accurately described as the merger of MCNP5 and MCNPX capabilities, but it is much more than the sum of those two computer codes. MCNP6 is the result of five years of effort by ...

Initial MCNP6 Release Overview

New accelerator-driven technologies that utilize spallation neutrons, such as the production of tritium and the transmutation of radioactive waste, require accurate nuclear data to model the performance of the target/blanket assembly and to predict neutron production, activation, heating, shielding requirements, and material damage. To meet these needs, nuclear-data evaluations and libraries up to 150 MeV have been developed for use in transport calculations to guide engineering design. By using advanced nuclear models that account for details of nuclear structure and the quantum nature of the nuclear scattering, significant gains in accuracy can be achieved below 150 MeV, where intranuclear cascade calculations become less accurate. Evaluations are in ENDF-6 format for important target/blanket and shielding materials (isotopes of H, C, N, O, Al, Si, P, Ca, Cr, Fe, Ni, Cu, Nb, W, Hg, and Pb) for both incident neutrons and incident protons. The evaluations are based on measured data as well as predictions from the GNASH nuclear model code, which calculates cross sections using Hauser-Feshbach, exciton, and Feshbach-Kerman-Koonin preequilibrium models. Elastic scattering distributions and direct reactions are calculated from the optical model. All evaluations specify production cross sections and energy-angle correlated spectra of secondary light particles as well as production crossmore » sections and energy distributions of heavy recoils and gamma rays. A formalism developed to calculate recoil energy distributions is presented. The use of these nuclear data in the MCNPX radiation transport code is also briefly described. This code merges essential elements of the LAHET and MCNP codes and uses these new data below 150 MeV and intranuclear cascade collision physics at higher energies. Extensive comparisons are shown between the evaluated results and experimental cross-section data to benchmark and validate the evaluated library. In addition, integral benchmarks of calculated and measured kerma coefficients for neutron energy deposition and neutron transmission through an iron slab compared with MCNPX calculations are provided. These evaluations have been accepted into the ENDF/B-VI library as Release 6.« less

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Cross-Section Evaluations to 150 MeV for Accelerator-Driven Systems and Implementation in MCNPX

/pdf/initial-mcnp6-release-overview-mcnp6-version-1-0-2oyelp09z4.pdf

Initial MCNP6 Release Overview - MCNP6 version 1.0

MCNPX (Monte Carlo N‐Particle eXtended) is a general‐purpose Monte Carlo radiation transport code with three‐dimensional geometry and continuous‐energy transport of 34 particles and light ions It contains flexible source and tally options, interactive graphics, and support for both sequential and multi‐processing computer platforms MCNPX is based on MCNP4c and has been upgraded to most MCNP5 capabilities MCNP is a highly stable code tracking neutrons, photons and electrons, and using evaluated nuclear data libraries for low‐energy interaction probabilities MCNPX has extended this base to a comprehensive set of particles and light ions, with heavy ion transport in development Models have been included to calculate interaction probabilities when libraries are not available Recent additions focus on the time evolution of residual nuclei decay, allowing calculation of transmutation and delayed particle emission MCNPX is now a code of great dynamic range, and the excellent neutronics capabilities allow new opportunities to simulate devices of interest to experimental particle physics, particularly calorimetry This paper describes the capabilities of the current MCNPX version 26C, and also discusses ongoing code development

/pdf/the-mcnpx-monte-carlo-radiation-transport-code-qogkjw3dr7.pdf

The MCNPX Monte Carlo Radiation Transport Code

/pdf/mcnpx-2-7-0-extensions-57kx55q9jx.pdf

Laurie S. Waters

Papers

Initial MCNP6 Release Overview

Cross-Section Evaluations to 150 MeV for Accelerator-Driven Systems and Implementation in MCNPX

Initial MCNP6 Release Overview - MCNP6 version 1.0

The MCNPX Monte Carlo Radiation Transport Code

MCNPX 2.7.0 extensions