Analysis of Population Control Techniques for Time-Dependent and Eigenvalue Monte Carlo Neutron Transport Calculations

نویسندگان

چکیده

An extensive study of population control techniques (PCTs) for time-dependent and eigenvalue Monte Carlo (MC) neutron transport calculations is presented. We define PCT as a technique that takes censused returns controlled, unbiased population. A new perspective based on an abstraction particle census explored, paving the way to improved understanding application concepts. Five distinct PCTs identified from literature are reviewed: Simple Sampling (SS), Splitting-Roulette (SR), Combing (CO), modified (COX), Duplicate-Discard (DD). theoretical analysis how much uncertainty introduced by each Parallel algorithms applicable both MC simulations proposed. The relative performances runtime tally mean error or standard deviation assessed solving test problems. It found SR CO equally most performant techniques, closely followed DD.

برای دانلود رایگان متن کامل این مقاله و بیش از 32 میلیون مقاله دیگر ابتدا ثبت نام کنید

اگر عضو سایت هستید لطفا وارد حساب کاربری خود شوید

منابع مشابه

Neutron Transport Calculations Using Monte-Carlo Methods

Fusion reactions from direct-drive inertial confinement fusion experiments at the Laboratory for Laser Energetics (LLE) release large amounts of stored nuclear energy. Most of it is given off as energetic neutrons. Plastic scintillators are used as detectors for these neutrons. The neutrons that interact with the scintillator produce photons that are recorded using a photomultiplier tube on a f...

متن کامل

A Monte Carlo Neutron Transport Code for Eigenvalue Calculations on a Dual-gpu System and Cuda Environment

Monte Carlo (MC) method is able to accurately calculate eigenvalues in reactor analysis. Its lengthy computation time can be reduced by general-purpose computing on Graphics Processing Units (GPU), one of the latest parallel computing techniques under development. The method of porting a regular transport code to GPU is usually very straightforward due to the “embarrassingly parallel” nature of...

متن کامل

Convergence Testing for Mcnp5 Monte Carlo Eigenvalue Calculations

Determining convergence of Monte Carlo criticality problems is complicated by the statistical noise inherent in the random walks of the neutrons in each generation. The latest version of MCNP5 incorporates an important new tool for assessing convergence: the Shannon entropy of the fission source distribution, Hsrc. Shannon entropy is a well-known concept from information theory and provides a s...

متن کامل

Wielandt Acceleration for Mcnp5 Monte Carlo Eigenvalue Calculations

Monte Carlo criticality calculations use the power iteration method to determine the eigenvalue (keff) and eigenfunction (fission source distribution) of the fundamental mode. A recently proposed method for accelerating convergence of the Monte Carlo power iteration using Wielandt’s method has been implemented in a test version of MCNP5. The method is shown to provide dramatic improvements in c...

متن کامل

Perturbation Theory Eigenvalue Sensitivity Analysis with Monte Carlo Techniques

Methodologies to calculate adjoint-based first-order-linear perturbation theory sensitivity coefficients with multigroup Monte Carlo methods are developed, implemented, and tested in this paper. These techniques can quickly produce sensitivity coefficients for all nuclides and reaction types for each region of a system model. Monte Carlo techniques have been developed to calculate the neutron f...

متن کامل

ذخیره در منابع من


  با ذخیره ی این منبع در منابع من، دسترسی به آن را برای استفاده های بعدی آسان تر کنید

ژورنال

عنوان ژورنال: Nuclear Science and Engineering

سال: 2022

ISSN: ['0029-5639', '1943-748X']

DOI: https://doi.org/10.1080/00295639.2022.2091906