Download or read book Criticality Calculations with MCNP written by Charles D. Harmon II and published by . This book was released on 1994 with total page 340 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Download or read book Criticality calculations with MCNP a primer written by Charles D. Harmon and published by . This book was released on 1974 with total page 0 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Download or read book Criticality Calculations with MCNP sup TM written by and published by . This book was released on 1994 with total page 175 pages. Available in PDF, EPUB and Kindle. Book excerpt: The purpose of this Primer is to assist the nuclear criticality safety analyst to perform computer calculations using the Monte Carlo code MCNP. Because of the closure of many experimental facilities, reliance on computer simulation is increasing. Often the analyst has little experience with specific codes available at his/her facility. This Primer helps the analyst understand and use the MCNP Monte Carlo code for nuclear criticality analyses. It assumes no knowledge of or particular experience with Monte Carlo codes in general or with MCNP in particular. The document begins with a Quickstart chapter that introduces the basic concepts of using MCNP. The following chapters expand on those ideas, presenting a range of problems from simple cylinders to 3-dimensional lattices for calculating keff confidence intervals. Input files and results for all problems are included. The Primer can be used alone, but its best use is in conjunction with the MCNP4A manual. After completing the Primer, a criticality analyst should be capable of performing and understanding a majority of the calculations that will arise in the field of nuclear criticality safety.
Download or read book Criticality Calculations with MCNP trademark written by and published by . This book was released on 1994 with total page 174 pages. Available in PDF, EPUB and Kindle. Book excerpt: With the closure of many experimental facilities, the nuclear criticality safety analyst increasingly is required to rely on computer calculations to identify safe limits for the handling and storage of fissile materials. However, in many cases, the analyst has little experience with the specific codes available at his/her facility. This primer will help you, the analyst, understand and use the MCNP Monte Carlo code for nuclear criticality safety analyses. It assumes that you have a college education in a technical field. There is no assumption of familiarity with Monte Carlo codes in general or with MCNP in particular. Appendix A gives an introduction to Monte Carlo techniques. The primer is designed to teach by example, with each example illustrating two or three features of MCNP that are useful in criticality analyses. Beginning with a Quickstart chapter, the primer gives an overview of the basic requirements for MCNP input and allows you to run a simple criticality problem with MCNP. This chapter is not designed to explain either the input or the MCNP options in detail; but rather it introduces basic concepts that are further explained in following chapters. Each chapter begins with a list of basic objectives that identify the goal of the chapter, and a list of the individual MCNP features that are covered in detail in the unique chapter example problems. It is expected that on completion of the primer you will be comfortable using MCNP in criticality calculations and will be capable of handling 80 to 90 percent of the situations that normally arise in a facility. The primer provides a set of basic input files that you can selectively modify to fit the particular problem at hand.
Download or read book MCNP sup TM Criticality Primer and Training Experiences written by and published by . This book was released on 1995 with total page 8 pages. Available in PDF, EPUB and Kindle. Book excerpt: With the closure of many experimental facilities, the nuclear criticality safety analyst is increasingly required to rely on computer calculations to identify safe limits for the handling and storage of fissile materials. However, the analyst may have little experience with the specific codes available at his or her facility. Usually, the codes are quite complex, black boxes capable of analyzing numerous problems with a myriad of input options. Documentation for these codes is designed to cover all the possible configurations and types of analyses but does not give much detail on any particular type of analysis. For criticality calculations, the user of a code is primarily interested in the value of the effective multiplication factor for a system (k{sub eff}). Most codes will provide this, and truckloads of other information that may be less pertinent to criticality calculations. Based on discussions with code users in the nuclear criticality safety community, it was decided that a simple document discussing the ins and outs of criticality calculations with specific codes would be quite useful. The Transport Methods Group, XTM, at Los Alamos National Laboratory (LANL) decided to develop a primer for criticality calculations with their Monte Carlo code, MCNP. This was a joint task between LANL with a knowledge and understanding of the nuances and capabilities of MCNP and the University of New Mexico with a knowledge and understanding of nuclear criticality safety calculations and educating first time users of neutronics calculations. The initial problem was that the MCNP manual just contained too much information. Almost everything one needs to know about MCNP can be found in the manual; the problem is that there is more information than a user requires to do a simple k{sub eff} calculation. The basic concept of the primer was to distill the manual to create a document whose only focus was criticality calculations using MCNP.
Download or read book Criticality Calculations with MCNPtrademark A Primer written by and published by . This book was released on 1994 with total page 0 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Download or read book Criticality Calculations with MCNP6 Practical Lectures written by and published by . This book was released on 2016 with total page 746 pages. Available in PDF, EPUB and Kindle. Book excerpt: These slides are used to teach MCNP (Monte Carlo N-Particle) usage to nuclear criticality safety analysts. The following are the lecture topics: course information, introduction, MCNP basics, criticality calculations, advanced geometry, tallies, adjoint-weighted tallies and sensitivities, physics and nuclear data, parameter studies, NCS validation I, NCS validation II, NCS validation III, case study 1 - solution tanks, case study 2 - fuel vault, case study 3 - B & W core, case study 4 - simple TRIGA, case study 5 - fissile mat. vault, criticality accident alarm systems. After completion of this course, you should be able to: Develop an input model for MCNP; Describe how cross section data impact Monte Carlo and deterministic codes; Describe the importance of validation of computer codes and how it is accomplished; Describe the methodology supporting Monte Carlo codes and deterministic codes; Describe pitfalls of Monte Carlo calculations; Discuss the strengths and weaknesses of Monte Carlo and Discrete Ordinants codes; The diffusion theory model is not strictly valid for treating fissile systems in which neutron absorption, voids, and/or material boundaries are present. In the context of these limitations, identify a fissile system for which a diffusion theory solution would be adequate.
Download or read book MCNP Analyses of Criticality Calculation Results written by and published by . This book was released on 1995 with total page 13 pages. Available in PDF, EPUB and Kindle. Book excerpt: Careful assessment of the results of a calculation by the code itself can reduce mistakes in the problem setup and execution. MCNP has over four hundred error messages that inform the user of FATAL or WARNING errors that have been discovered during the processing of just the input file. The latest version, MCNP4A, now performs a self assessment of the calculated results to aid the user in determining the quality of the Monte Carlo results. MCNP4A, which was released to RSIC in October 1993, contains new analyses of the MCNP Monte Carlo calculation that provide simple user WARNINGs for both criticality and fixed source calculations. The goal of the new analyses is to provide the MCNP criticality practitioner with enough information in the output to assess the validity of the k{sub eff} calculation and any associated tallies. The results of these checks are presented in the k{sub eff} results summary page, several k{sub eff} tables and graphs, and tally tables and graphs. Plots of k{sub eff} at the workstation are also available as the problem is running or in a postprocessing mode to assess problem performance and results.
Download or read book Visualization and Analyses of MCNP Criticality Calculation Results written by and published by . This book was released on 1995 with total page 10 pages. Available in PDF, EPUB and Kindle. Book excerpt: Careful assessment of the results of a calculation by the code itself can detect mistakes in the problem setup and execution. MCNP has over four hundred error messages that inform the user of FATAL or WARNING errors that have been discovered during processing of just the input file. MCNP4A performs a self assessment of the calculated results to aid the user in determining the quality of the Monte Carlo results. MCNP4A contains new built-in sensitivity analyses of the Monte Carlo calculation that provide the user with simple WARNING messages for both criticality and fixed source calculations. The goal of the new analyses described in this paper is to provide the MCNP criticality practitioner with enough information in the output to assess the validity of the k{sub eff} calculation and any associated tallies. The results of these checks are presented in the k{sub eff} results summary, several k{sub eff} tables and graphs, and tally tables and graphs. Plots of k{sub eff} at the workstation are also available as the problem is running or in a postprocessing mode to assess problem performance and results. Plots of the fission source by cycle supply valuable visual information, although they are not yet available in the production version of MCNP.
Download or read book MCNP and OMEGA Criticality Calculations written by Eberhard Seifert and published by . This book was released on 1998 with total page 22 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Download or read book A Criticality Study of Fast Critical Experimental Benchmarks Using MCNP Code to Qualifying Different Evaluations written by Sanae Elouahdani and published by . This book was released on 2019 with total page 0 pages. Available in PDF, EPUB and Kindle. Book excerpt: In this chapter we present our MCNP modeling, concerning fast critical experimental benchmarks, about qualifying our libraries of cross-sections deduced from the evaluations ENDF/B-VII, JEFF-3.1, JENDL-3.3, JENDL-4 processed by the code NJOY. The benchmarks analyzed are characterized by simple geometries which help to have a precise calculation. In our neutron calculation, we used the MCNP code (version 5), the reference code for the neutron transport calculation with the Monte Carlo method. It is also very efficient for criticality calculation. The cross-section data for all the isotopes that make up the material of the studied benchmarks are processed in ACE format at 300 K temperature using the NJOY 99.9 modular system. A detailed comparison of the criticality results of our simulation was carried out to highlight the influence of these evaluations on the keff calculations.
Download or read book Site Characterization Progress Report written by and published by . This book was released on 1997 with total page 856 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Download or read book MCNP Perturbation Capability for Monte Carlo Criticality Calculations written by and published by . This book was released on 1999 with total page 5 pages. Available in PDF, EPUB and Kindle. Book excerpt: The differential operator perturbation capability in MCNP4B has been extended to automatically calculate perturbation estimates for the track length estimate of k{sub eff} in MCNP4B. The additional corrections required in certain cases for MCNP4B are no longer needed. Calculating the effect of small design changes on the criticality of nuclear systems with MCNP is now straightforward.
Download or read book Medical Physics Calculations with MCNP written by Alexis D Lazarine and published by . This book was released on 2006 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: The rising desire for individualized medical physics models has sparked a transition from the use of tangible phantoms toward the employment of computational software for medical physics applications. One such computational software for radiation transport modeling is the Monte Carlo N-Particle (MCNP) radiation transport code. However, no comprehensive document has been written to introduce the use of the MCNP code for simulating medical physics applications. This document, a primer, addresses this need by leading the medical physics user through the basic use of MCNP and its particular application to the medical physics field. This primer is designed to teach by example, with the aim that each example will illustrate a practical use of particular features in MCNP that are useful in medical physics applications. These examples along with the instructions for reproducing them are the results of this thesis research. These results include simulations of: dose fromTc-99m diagnostic therapy, calculation of Medical Internal Radiation Dose (MIRD) specific absorbed fraction (SAF) values using the ORNL MIRD phantom, x-ray phototherapy effectiveness, prostate brachytherapy lifetime dose calculations, and a radiograph of the head using the Zubal head phantom. Also included are a set of appendices that include useful reference data, code syntax, and a database of input decks including the examples in the primer. The sections in conjunction with the appendices should provide a foundation of knowledge regarding the MCNP commands and their uses as well as enable users to utilize the MCNP manual effectively for situations not specifically addressed by the primer.
Download or read book Nuclear Engineering written by Zafar Ullah Koreshi and published by Academic Press. This book was released on 2022-03-23 with total page 549 pages. Available in PDF, EPUB and Kindle. Book excerpt: Nuclear Engineering Mathematical Modeling and Simulation presents the mathematical modeling of neutron diffusion and transport. Aimed at students and early career engineers, this highly practical and visual resource guides the reader through computer simulations using the Monte Carlo Method which can be applied to a variety of applications, including power generation, criticality assemblies, nuclear detection systems, and nuclear medicine to name a few. The book covers optimization in both the traditional deterministic framework of variational methods and the stochastic framework of Monte Carlo methods. Specific sections cover the fundamentals of nuclear physics, computer codes used for neutron and photon radiation transport simulations, applications of analyses and simulations, optimization techniques for both fixed-source and multiplying systems, and various simulations in the medical area where radioisotopes are used in cancer treatment. Provides a highly visual and practical reference that includes mathematical modeling, formulations, models and methods throughout Includes all current major computer codes, such as ANISN, MCNP and MATLAB for user coding and analysis Guides the reader through simulations for the design optimization of both present-day and future nuclear systems
Download or read book Monte Carlo Techniques for Nuclear Systems Theory Lectures written by and published by . This book was released on 2016 with total page 606 pages. Available in PDF, EPUB and Kindle. Book excerpt: These are lecture notes for a Monte Carlo class given at the University of New Mexico. The following topics are covered: course information; nuclear eng. review & MC; random numbers and sampling; computational geometry; collision physics; tallies and statistics; eigenvalue calculations I; eigenvalue calculations II; eigenvalue calculations III; variance reduction; parallel Monte Carlo; parameter studies; fission matrix and higher eigenmodes; doppler broadening; Monte Carlo depletion; HTGR modeling; coupled MC and T/H calculations; fission energy deposition. Solving particle transport problems with the Monte Carlo method is simple - just simulate the particle behavior. The devil is in the details, however. These lectures provide a balanced approach to the theory and practice of Monte Carlo simulation codes. The first lectures provide an overview of Monte Carlo simulation methods, covering the transport equation, random sampling, computational geometry, collision physics, and statistics. The next lectures focus on the state-of-the-art in Monte Carlo criticality simulations, covering the theory of eigenvalue calculations, convergence analysis, dominance ratio calculations, bias in Keff and tallies, bias in uncertainties, a case study of a realistic calculation, and Wielandt acceleration techniques. The remaining lectures cover advanced topics, including HTGR modeling and stochastic geometry, temperature dependence, fission energy deposition, depletion calculations, parallel calculations, and parameter studies. This portion of the class focuses on using MCNP to perform criticality calculations for reactor physics and criticality safety applications. It is an intermediate level class, intended for those with at least some familiarity with MCNP. Class examples provide hands-on experience at running the code, plotting both geometry and results, and understanding the code output. The class includes lectures & hands-on computer use for a variety of Monte Carlo calculations. Beginning MCNP users are encouraged to review LA-UR-09-00380, "Criticality Calculations with MCNP: A Primer (3nd Edition)" (available at http:// mcnp.lanl.gov under "Reference Collection") prior to the class. No Monte Carlo class can be complete without having students write their own simple Monte Carlo routines for basic random sampling, use of the random number generator, and simplified particle transport simulation.
Download or read book Masters Theses in the Pure and Applied Sciences written by Wade H. Shafer and published by Springer Science & Business Media. This book was released on 2012-12-06 with total page 427 pages. Available in PDF, EPUB and Kindle. Book excerpt: Masters Theses in the Pure and Applied Sciences was first conceived, published, and disseminated by the Center for Information and Numerical Data Analysis and Synthesis (CINDAS)* at Purdue University in 1957, starting its coverage of theses with the academic year 1955. Beginning with Volume 13, the printing and dis semination phases of the activity were transferred to University Microfilms/Xerox of Ann Arbor, Michigan, with the thought that such an arrangement would be more beneficial to the academic and general scientific and technical community. After five years of this joint undertaking we had concluded that it was in the interest of all concerned if the printing and distribution of the volumes were handled by an international publishing house to assure improved service and broader dissemination. Hence, starting with Volume 18, Masters Theses in the Pure and Applied Sciences has been disseminated on a worldwide basis by Plenum Publishing Corporation of New York, and in the same year the coverage was broadened to include Canadian universities. All back issues can also be ordered from Plenum. We have reported in Volume 39 (thesis year 1994) a total of 13,953 thesis titles from 21 Canadian and 159 United States universities. We are sure that this broader base for these titles reported will greatly enhance the value of this impor tant annual reference work. While Volume 39 reports theses submitted in 1994, on occasion, certain uni versities do report theses submitted in previous years but not reported at the time.
