Download or read book Systems Guide to MCNP Monte Carlo Neutron and Photon Transport Code written by and published by . This book was released on 1984 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: The subject of this report is the implementation of the Los Alamos National Laboratory Monte Carlo Neutron and Photon Transport Code - Version 3 (MCNP) on the different types of computer systems, especially the IBM MVS system. The report supplements the documentation of the RSIC computer code package CCC-200/MCNP. Details of the procedure to follow in executing MCNP on the IBM computers, either in batch mode or interactive mode, are provided.
Download or read book MCNP written by and published by . This book was released on 1985 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: MCNP is a very general Monte Carlo neutron photon transport code system with approximately 250 person years of Group X-6 code development invested. It is extremely portable, user-oriented, and a true production code as it is used about 60 Cray hours per month by about 150 Los Alamos users. It has as its data base the best cross-section evaluations available. MCNP contains state-of-the-art traditional and adaptive Monte Carlo techniques to be applied to the solution of an ever-increasing number of problems. Excellent user-oriented documentation is available for all facets of the MCNP code system. Many useful and important variants of MCNP exist for special applications. The Radiation Shielding Information Center (RSIC) in Oak Ridge, Tennessee is the contact point for worldwide MCNP code and documentation distribution. A much improved MCNP Version 3A will be available in the fall of 1985, along with new and improved documentation. Future directions in MCNP development will change the meaning of MCNP to Monte Carlo N Particle where N particle varieties will be transported.
Download or read book MCNP a General Monte Carlo Code for Neutron and Photon Transport written by and published by . This book was released on 1979 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: The general-purpose Monte Carlo code MCNP can be used for neutron, photon, or coupled neutron-photon transport, including the capability to calculate eigenvalues for critical systems. The code treats an arbitrary three-dimensional configuration of materials in geometric cells bounded by first- and second-degree surfaces and some special fourth-degree surfaces (elliptical tori).
Download or read book MCNP A GENERAL MONTE CARLO CODE FOR NEUTRON AND PHOTON TRANSPORT written by W. L. Thompson and published by . This book was released on with total page 0 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Download or read book MCNP written by Los Alamos Monte Carlo Group and published by . This book was released on 1981 with total page 512 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Download or read book MCNP a General Monte Carlo Code for Neutron and Photon Transport written by and published by . This book was released on 1986 with total page 0 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Download or read book MCNP a General Monte Carlo Code for Neutron and Photon Transport Version 3A written by Judith F. Briesmeister and published by . This book was released on 1986 with total page 582 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Download or read book MCNP written by Los Alamos Scientific Laboratory. Group X-6 and published by . This book was released on 1979 with total page 411 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Download or read book MCNP written by W. L. Thompson and published by . This book was released on 1979 with total page 10 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Download or read book MCNP Monte Carlo Neutron Photon Capabilities for Nuclear Well Logging Calculations written by and published by . This book was released on 1989 with total page 7 pages. Available in PDF, EPUB and Kindle. Book excerpt: The Los Alamos Radiation Transport Code System (LARTCS) consists of state-of-the-art Monte Carlo and discrete ordinates transport codes and data libraries. The general-purpose continuous-energy Monte Carlo code MCNP (Monte Carlo Neutron Photon), part of the LARTCS, provides a computational predictive capability for many applications of interest to the nuclear well logging community. The generalized three-dimensional geometry of MCNP is well suited for borehole-tool models. SABRINA, another component of the LARTCS, is a graphics code that can be used to interactively create a complex MCNP geometry. Users can define many source and tally characteristics with standard MCNP features. The time-dependent capability of the code is essential when modeling pulsed sources. Problems with neutrons, photons, and electrons as either single particle or coupled particles can be calculated with MCNP. The physics of neutron and photon transport and interactions is modeled in detail using the latest available cross-section data. A rich collections of variance reduction features can greatly increase the efficiency of a calculation. MCNP is written in FORTRAN 77 and has been run on variety of computer systems from scientific workstations to supercomputers. The next production version of MCNP will include features such as continuous-energy electron transport and a multitasking option. Areas of ongoing research of interest to the well logging community include angle biasing, adaptive Monte Carlo, improved discrete ordinates capabilities, and discrete ordinates/Monte Carlo hybrid development. Los Alamos has requested approval by the Department of Energy to create a Radiation Transport Computational Facility under their User Facility Program to increase external interactions with industry, universities, and other government organizations. 21 refs.
Download or read book Monte Carlo N Particle Simulations for Nuclear Detection and Safeguards written by John S. Hendricks and published by Springer Nature. This book was released on 2022-09-27 with total page 316 pages. Available in PDF, EPUB and Kindle. Book excerpt: This open access book is a pedagogical, examples-based guide to using the Monte Carlo N-Particle (MCNP®) code for nuclear safeguards and non-proliferation applications. The MCNP code, general-purpose software for particle transport simulations, is widely used in the field of nuclear safeguards and non-proliferation for numerous applications including detector design and calibration, and the study of scenarios such as measurement of fresh and spent fuel. This book fills a gap in the existing MCNP software literature by teaching MCNP software usage through detailed examples that were selected based on both student feedback and the real-world experience of the nuclear safeguards group at Los Alamos National Laboratory. MCNP input and output files are explained, and the technical details used in MCNP input file preparation are linked to the MCNP code manual. Benefiting from the authors’ decades of experience in MCNP simulation, this book is essential reading for students, academic researchers, and practitioners whose work in nuclear physics or nuclear engineering is related to non-proliferation or nuclear safeguards. Each chapter comes with downloadable input files for the user to easily reproduce the examples in the text.
Download or read book Applications Guide to the RSIC distributed Version of the MCNP Code coupled Monte Carlo Neutron photon Code written by and published by . This book was released on 1985 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: An overview of the RSIC-distributed version of the MCNP code (a soupled Monte Carlo neutron-photon code) is presented. All general features of the code, from machine hardware requirements to theoretical details, are discussed. The current nuclide cross-section and other libraries available in the standard code package are specified, and a realistic example of the flexible geometry input is given. Standard and nonstandard source, estimator, and variance-reduction procedures are outlined. Examples of correct usage and possible misuse of certain code features are presented graphically and in standard output listings. Finally, itemized summaries of sample problems, various MCNP code documentation, and future work are given.
Download or read book On the Utilization of the Monte Carlo Neutron and Photon Transport Code for the Design and Optimization of Nuclear Counting Systems written by Carey Nation and published by . This book was released on 1997 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt:
Download or read book Monte Carlo Calculations with the MCNP Code for Investigations of Neutron and Photon Transport at the ASDEX Upgrade Tokamak written by Günter Fieg and published by . This book was released on 1991 with total page 43 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Download or read book Monte Carlo Calculations with the MCNP Code for Investigationsof Neutron and Photon Transport at the ASDEX Upgrade Tokamak written by G. Fieg and published by . This book was released on 1991 with total page 43 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Download or read book Development and Implementation of Photonuclear Cross section Data for Mutually Coupled Neutron photon Transport Calculations in the Monte Carlo N Particle MCNP Radiation Transport Code written by Morgan C. White and published by . This book was released on 2000 with total page 1078 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Download or read book SORS Monte Carlo Photon transport Code for the CDC 6600 written by John Kimlinger and published by . This book was released on 1967 with total page 50 pages. Available in PDF, EPUB and Kindle. Book excerpt:
