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 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 MCNP Analyses of Criticality Calculation Results written by and published by . This book was released on 1995 with total page 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 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 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 Criticality Benchmark Results Using Various MCNP Data Libraries written by and published by . This book was released on 1999 with total page 5 pages. Available in PDF, EPUB and Kindle. Book excerpt: A suite of 86 criticality benchmarks has been recently implemented in MCNP{trademark} as part of the nuclear data validation effort. These benchmarks have been run using two sets of MCNP continuous-energy neutron data: ENDF/B-VI based data through Release 2 (ENDF60) and the ENDF/B-V based data. New evaluations were completed for ENDF/B-VI for a number of the important nuclides such as the isotopes of H, Be, C, N, O, Fe, Ni, {sup 235,238}U, 237Np, and {sup 239,240}Pu. When examining the results of these calculations for the five manor categories of 233U, intermediate-enriched 235U (IEU), highly enriched 235U (HEU), 239Pu, and mixed metal assembles, we find the following: (1) The new evaluations for 9Be, 12C, and 14N show no net effect on k{sub eff}; (2) There is a consistent decrease in k{sub eff} for all of the solution assemblies for ENDF/B-VI due to 1H and 16O, moving k{sub eff} further from the benchmark value for uranium solutions and closer to the benchmark value for plutonium solutions; (3) k{sub eff} decreased for the ENDF/B-VI Fe isotopic data, moving the calculated k{sub eff} further from the benchmark value; (4) k{sub eff} decreased for the ENDF/B-VI Ni isotopic data, moving the calculated k{sub eff} closer to the benchmark value; (5) The W data remained unchanged and tended to calculate slightly higher than the benchmark values; (6) For metal uranium systems, the ENDF/B-VI data for 235U tends to decrease k{sub eff} while the 238U data tends to increase k{sub eff}. The net result depends on the energy spectrum and material specifications for the particular assembly; (7) For more intermediate-energy systems, the changes in the {sup 235,238}U evaluations tend to increase k{sub eff}. For the mixed graphite and normal uranium-reflected assembly, a large increase in k{sub eff} due to changes in the 238U evaluation moved the calculated k{sub eff} much closer to the benchmark value. (8) There is little change in k{sub eff} for the uranium solutions due to the new {sup 235,238}U evaluations; and (9) There is little change in k{sub eff} for the 239Pu metal assemblies, but a decrease in k{sub eff} for the solution assemblies, moving them closer to the benchmark value.
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 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 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 MCNP Calculations for Criticality safety Benchmarks with ENDF written by and published by . This book was released on 1995 with total page 12 pages. Available in PDF, EPUB and Kindle. Book excerpt: The MCNP Monte Carlo code, in conjunction with its continuous-energy ENDF/B-V and ENDF/B-VI cross-section libraries, has been benchmarked against results from 27 different critical experiments. The predicted values of k{sub eff} are in excellent agreement with the benchmarks, except for the ENDF/B-V results for solutions of plutonium nitrate and, to a lesser degree, for the ENDF/B-V and ENDF/B-VI results for a bare sphere of 233U.
Download or read book Materials for Nuclear Waste Immobilization written by Michael I. Ojovan and published by MDPI. This book was released on 2020-01-09 with total page 220 pages. Available in PDF, EPUB and Kindle. Book excerpt: The book outlines recent advances in nuclear wasteform materials including glasses, ceramics and cements and spent nuclear fuel. It focuses on durability aspects and contains data on performance of nuclear wasteforms as well as expected behavior in a disposal environment.
Download or read book New Calculations for Critical Assemblies Using MCNP4B written by and published by . This book was released on 1997 with total page 9 pages. Available in PDF, EPUB and Kindle. Book excerpt: A suite of 41 criticality benchmarks has been modeled using MCNP{trademark} (version 4B). Most of the assembly specifications were obtained from the Cross Section Evaluation Working Group (CSEWG) and the International Criticality Safety Benchmark Evaluation Project (ICSBEP) compendiums of experimental benchmarks. A few assembly specifications were obtained from experimental papers. The suite contains thermal and fast assemblies, bare and reflected assemblies, and emphasizes 233U, 235U, 238U, and 239Pu. The values of k{sub eff} for each assembly in the suite were calculated using MCNP libraries derived primarily from release 2 of ENDF/B-V and release 2 of ENDF/B-VI. The results show that the new ENDF/B-VI. 2 evaluations for H, O, N, B, 235U, 238U, and 239Pu can have a significant impact on the values of k{sub eff}. In addition to the integral quantity k{sub eff}, several additional experimental measurements were performed and documented. These experimental measurements include central fission and reaction-rate ratios for various isotopes, and neutron leakage and flux spectra. They provide more detailed information about the accuracy of the nuclear data than can k{sub eff}. Comparison calculations were performed using both ENDF/B-V.2 and ENDF/B-VI. 2-based data libraries. The purpose of this paper is to compare the results of these additional calculations with experimental data, and to use these results to assess the quality of the nuclear data.
Download or read book A Suite of Criticality Benchmarks for Validating Nuclear Data written by and published by . This book was released on 2006 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: The continuous-energy neutron data library ENDF60 for use with MCNP[trademark] was released in the fall of 1994, and was based on ENDF/B-Vl evaluations through Release 2. As part of the data validation process for this library, a number of criticality benchmark calculations were performed. The original suite of nine criticality benchmarks used to test ENDF60 has now been expanded to 86 benchmarks. This report documents the specifications for the suite of 86 criticality benchmarks that have been developed for validating nuclear data.
