Criticality Calculations with MCNP6 - Practical Lectures

Technical Report · Tue Nov 29 04:00:00 EST 2016

DOI:https://doi.org/10.2172/1334108· OSTI ID:1334108

Brown, Forrest B. ^[1]; Rising, Michael Evan ^[1]; Alwin, Jennifer Louise ^[1]

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1. Los Alamos National Lab. (LANL), Los Alamos, NM (United States). Monte Carlo Methods, Codes, and Applications (XCP-3)

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.

• Research Organization: Los Alamos National Lab. (LANL), Los Alamos, NM (United States)
• Sponsoring Organization: USDOE National Nuclear Security Administration (NNSA). Nuclear Criticality Safety Program (NCSP)
• DOE Contract Number: AC52-06NA25396
• OSTI ID: 1334108
• Report Number(s): LA-UR--16-29071
• Country of Publication: United States
• Language: English

Related Subjects

22 GENERAL STUDIES OF NUCLEAR REACTORS
42 ENGINEERING
73 NUCLEAR PHYSICS AND RADIATION PHYSICS
ABSORPTION
ADJOINT DIFFERENCE METHOD
ALARM SYSTEMS
BW STANDARD REACTOR
COMPUTER CALCULATIONS
CRITICALITY
CROSS SECTIONS
DIFFUSION
DISCRETE ORDINATE METHOD
FISSILE MATERIALS
LECTURES
M CODES
MATHEMATICAL SOLUTIONS
MONTE CARLO METHOD
Monte Carlo
NEUTRONS
NUCLEAR FUELS
Nuclear Criticality Safety Program (NCSP)
RADIATION ACCIDENTS
SAFETY
SENSITIVITY
SOLUTIONS
STORAGE FACILITIES
TANKS
TRAINING
TRIGA TYPE REACTORS
VALIDATION
VOIDS
neutron transport