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Title: Criticality Calculations with MCNP6 - Practical Lectures

Abstract

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.

Authors:
 [1];  [1];  [1]
  1. Los Alamos National Lab. (LANL), Los Alamos, NM (United States). Monte Carlo Methods, Codes, and Applications (XCP-3)
Publication Date:
Research Org.:
Los Alamos National Lab. (LANL), Los Alamos, NM (United States)
Sponsoring Org.:
USDOE National Nuclear Security Administration (NNSA). Nuclear Criticality Safety Program (NCSP)
OSTI Identifier:
1334108
Report Number(s):
LA-UR-16-29071
TRN: US1700798
DOE Contract Number:  
AC52-06NA25396
Resource Type:
Technical Report
Country of Publication:
United States
Language:
English
Subject:
73 NUCLEAR PHYSICS AND RADIATION PHYSICS; 22 GENERAL STUDIES OF NUCLEAR REACTORS; 42 ENGINEERING; M CODES; MONTE CARLO METHOD; CRITICALITY; SOLUTIONS; VALIDATION; LECTURES; TRIGA TYPE REACTORS; NEUTRONS; MATHEMATICAL SOLUTIONS; RADIATION ACCIDENTS; ABSORPTION; DIFFUSION; TANKS; ALARM SYSTEMS; CROSS SECTIONS; NUCLEAR FUELS; COMPUTER CALCULATIONS; FISSILE MATERIALS; SAFETY; SENSITIVITY; VOIDS; STORAGE FACILITIES; BW STANDARD REACTOR; DISCRETE ORDINATE METHOD; ADJOINT DIFFERENCE METHOD; TRAINING; Monte Carlo; neutron transport

Citation Formats

Brown, Forrest B., Rising, Michael Evan, and Alwin, Jennifer Louise. Criticality Calculations with MCNP6 - Practical Lectures. United States: N. p., 2016. Web. doi:10.2172/1334108.
Brown, Forrest B., Rising, Michael Evan, & Alwin, Jennifer Louise. Criticality Calculations with MCNP6 - Practical Lectures. United States. doi:10.2172/1334108.
Brown, Forrest B., Rising, Michael Evan, and Alwin, Jennifer Louise. Tue . "Criticality Calculations with MCNP6 - Practical Lectures". United States. doi:10.2172/1334108. https://www.osti.gov/servlets/purl/1334108.
@article{osti_1334108,
title = {Criticality Calculations with MCNP6 - Practical Lectures},
author = {Brown, Forrest B. and Rising, Michael Evan and Alwin, Jennifer Louise},
abstractNote = {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.},
doi = {10.2172/1334108},
journal = {},
number = ,
volume = ,
place = {United States},
year = {2016},
month = {11}
}

Technical Report:

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