• Nuclear Engineering and Technology
• /
• v.42 no.3
• /
• pp.271-278
• /
• 2010

While the deterministic lattice physics/depletion codes use leakage-corrected critical spectrum (although approximate due to the B1 buckling search employed), Monte Carlo depletion codes currently in use do not have such a feature in spite of their heterogeneity and continuous-energy modeling capability. This paper describes an approach to Monte Carlo depletion with leakage-corrected critical spectrum derived from first principles. This is based on the concept of albedo eigenvalue treated as weight of the reflected neutron in Monte Carlo simulation.

• Nuclear Medicine and Molecular Imaging
• /
• v.42 no.2
• /
• pp.127-136
• /
• 2008

Monte Carlo simulation methods are especially useful in studying a variety of problems difficult to calculate by experimental or analytical approaches. Nowadays, they are extensively applied to simulate nuclear medicine instrumentations such as single photon emission computed tomography (SPECT) and positron emission tomography (PET) for assisting system design and optimizing imaging and processing protocols. The goal of this paper is to address the practical issues, a potential user of Monte Carlo simulations for nuclear medicine can encounter, to help them to choose a code. This review introduces the different types of Monte Carlo codes currently available for nuclear medicine, comments main features and properties for a code to be proper for a given purpose, and discusses current research trends in Monte Carlo codes.

• Nuclear Engineering and Technology
• /
• v.35 no.6
• /
• pp.581-595
• /
• 2003

Transport lattice codes require the resonance integral tables for the resonant nuclides where the resonance integral is a function of the background cross section and can be prepared through a special program solving the slowing down equation. In case the cross section libraries do not include the resonance integral table for the resonant nuclides, the computational prediction produces a large error. We devised a new method using a Monte Carlo calculation for the effective resonance cross sections to solve this problem provisionally. We extended this method to obtain the resonance integral table for general purpose. The MCNP code is used for the effective resonance integrals and the LIBERTE code for the effective background cross sections. We modified the HELIOS library with the effective cross sections and the resonance integral tables obtained by the newly developed Monte Carlo method, and performed sample calculations using HELIOS and LIBERTE. The results showed that this method is very effective for the resonance treatment.

• Nuclear Engineering and Technology
• /
• v.51 no.4
• /
• pp.954-962
• /
• 2019

It is well known that the variance of tally is biased in a Monte Carlo calculation based on the power iteration method. Several studies have been conducted to estimate the real variance. Among them, the batch method, which was proposed by Gelbard and Prael, has been utilized actively in many Monte Carlo codes because the method is straightforward, and it is easy to implement the method in the codes. However, there is a problem when utilizing the batch method because the estimated variance varies depending on batch size. Often, the appropriate batch size is not realized before the completion of several Monte Carlo calculations. This study recognizes this shortcoming and addresses it by permitting selection of an appropriate batch size.

• Nuclear Engineering and Technology
• /
• v.49 no.6
• /
• pp.1211-1218
• /
• 2017

Monte Carlo neutron transport codes generally use the method of successive generations to converge the fission source distribution to-and then maintain it at-the fundamental mode. Recently, a phenomenon called "clustering" has been noted, which produces fission distributions that are very far from the fundamental mode. In this study, a mathematical model of clustering in Monte Carlo has been developed. The model draws on previous work for continuous-time birth-death processes, as well as methods from the field of population genetics.

• Nuclear Engineering and Technology
• /
• v.38 no.3
• /
• pp.225-230
• /
• 2006

With the advent of inexpensive computing power over the past two decades, applications of Monte Carlo radiation transport techniques have proliferated dramatically. At Los Alamos, the Monte Carlo codes MCNP5 and MCNPX are used routinely on personal computer platforms for radiation shielding analysis and dosimetry calculations. These codes feature a rich palette of variance reduction (VR) techniques. The motivation of VR is to exchange user efficiency for computational efficiency. It has been said that a few hours of user time often reduces computational time by several orders of magnitude. Unfortunately, user time can stretch into the many hours as most VR techniques require significant user experience and intervention for proper optimization. It is the purpose of this paper to outline VR strategies, tested in practice, optimized for several common radiation shielding tasks, with the hope of reducing user setup time for similar problems. A strategy is defined in this context to mean a collection of MCNP radiation transport physics options and VR techniques that work synergistically to optimize a particular shielding task. Examples are offered in the areas of source definition, skyshine, streaming, and transmission.

• Nuclear Engineering and Technology
• /
• v.54 no.5
• /
• pp.1902-1908
• /
• 2022

In many of the complex systems modeled in the field of nuclear engineering, it is often useful to use linear regression-based analyses to analyze relationships between model parameters and responses of interests. In cases where the response of interest is calculated by a simulation which uses Monte Carlo methods, there will be some uncertainty in the responses. Further, the reduction of this uncertainty increases the time necessary to run each calculation. This paper presents some discussion on how the Monte Carlo error in the response of interest influences the error in computed linear regression coefficients. A mathematical justification is given that shows that when performing linear regression in these scenarios, the error in regression coefficients can be largely independent of the Monte Carlo error in each individual calculation. This condition is only true if the total number of calculations are scaled to have a constant total time, or amount of work, for all calculations. An application with a simple pin cell model is used to demonstrate these observations in a practical problem.

• Proceedings of the Korean Society of Medical Physics Conference
• /
• 2006.08a
• /
• pp.67-67
• /
• 2006

• Nuclear Engineering and Technology
• /
• v.53 no.10
• /
• pp.3158-3163
• /
• 2021

Chromium material is commonly used for fusion plasma facing applications because of the low neutron activation property. The Monte Carlo method is one of the useful ways to investigate the ion-target interactions. In this study, Chromium target irradiated by protons was investigated using Monte Carlo based simulation tools. In this context, the calculations of radiation damage on Chromium material irradiated with protons at 17.9 and 22.3 MeV energies were carried out using GEANT4 and SRIM codes. Besides, the cross sections for proton interaction with Chromium target were calculated by the TALYS 1.9 code using CTM + FGM, BSFGM, and GSFM level densities. As a result, GEANT4, SRIM and TALYS 1.9 codes provide a suitable tool for the predictions of radiation damage and cross cross section with proton irradiation.

• The Journal of Korean Institute of Communications and Information Sciences
• /
• v.35 no.5C
• /
• pp.415-422
• /
• 2010

In this paper, we derive exact closed-form formulas, in terms of Meijer's G-function, for the ergodic capacity of space-time block codes (STBCs) from generalized linear complex orthogonal designs (GLCODs) and generalized coordinate interleaved orthogonal designs (GCIODs) in quasi-static frequency-nonselective i.i.d. Nakagami-m fading channels. The derived analytical results show an excellent agreement with Monte-Carlo simulation results. Thus, a useful means for analyzing and predicting the ergodic capacity performance of STBCs from GLCODs or GCIODs can be provided in various antenna configurations and different channel conditions without extensive Monte-Carlo simulations. We present some numerical results to verify the accuracy of the derived formulas.