Download or read book Multigroup Monte Carlo Reactor Calculation with Coarse Mesh Finite Difference Formulation for Real Variance Reduction written by and published by . This book was released on 2010 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: The coarse mesh finite difference (CMFD) formulation has been applied to Monte Carlo (MC) simulations in order to mitigate the issue of large real variances of pin power tallies in full-core problems. In this work, a parallelized multigroup (MG) two-dimensional (2-D) MC code named PRIDE (Probabilistic Reactor Investigation with Discretized Energy), which is capable of handling lattices of square pin cells within which circular substructures can be modeled, has been developed as a tool for the investigations of the new method. In this code, a scheme to construct a CMFD linear system is based on the MC tallies of coarse mesh average fluxes and the net currents at coarse mesh interfaces. These tallies are accumulated over the MC cycles to get more stable CMFD solutions which are used for feedback to MC fission source distribution (FSD). The feedback scheme in this code employs a weight adjustment of fission source neutrons for the next MC cycle that is to reflect the global CMFD FSD into the MC FSD. The performance of CMFD feedback has been investigated in terms of the number of inactive cycles required for the convergence of FSD and also the reduction of real variances of local property tallies in active cycles. The applications to 2-D multigroup full-core pressurized water reactor problems have demonstrated that the MC FSD converges considerably faster and the real variances of pin powers are smaller by a factor of 4 with CMFD FSD feedback. It is also noted that the large real variances of pin powers are caused mainly by the global assembly-wise fluctuations of power distributions in a large core rather than local fluctuations.
Download or read book Deterministic Numerical Methods for Unstructured Mesh Neutron Transport Calculation written by Liangzhi Cao and published by Woodhead Publishing. This book was released on 2020-08-30 with total page 294 pages. Available in PDF, EPUB and Kindle. Book excerpt: Deterministic Numerical Methods for Unstructured-Mesh Neutron Transport Calculation presents the latest deterministic numerical methods for neutron transport equations (NTEs) with complex geometry, which are of great demand in recent years due to the rapid development of advanced nuclear reactor concepts and high-performance computational technologies. This book covers the wellknown methods proposed and used in recent years, not only theoretical modeling but also numerical results. This book provides readers with a very thorough understanding of unstructured neutron transport calculations and enables them to develop their own computational codes. The fundamentals, numerical discretization methods, algorithms, and numerical results are discussed. Researchers and engineers from utilities and research institutes are provided with examples on how to model an advanced nuclear reactor, which they can then apply to their own research projects and lab settings. Combines the theoretical models with numerical methods and results in one complete resource Presents the latest progress on the topic in an easy-to-navigate format
Download or read book Improvements and Applications of the Uniform Fission Site Method in Monte Carlo written by Jessica Lynn Hunter and published by . This book was released on 2014 with total page 63 pages. Available in PDF, EPUB and Kindle. Book excerpt: Monte Carlo methods for reactor analysis have been in development with the eventual goal of full-core analysis. To attain results with reasonable uncertainties, large computational resources are needed. Variance reduction methods have been developed in order to reduce the computational resources required to obtain results in a practical amount of time. This work seeks to expand research in the Uniform Fission Site (UFS) method, a variance reduction technique recently developed that causes uniformity in uncertainty distributions by forcing uniformity in source distributions. This work aims to both improve the method as well as investigate its use with a source acceleration method, Coarse Mesh Finite Difference (CMFD) acceleration. Both techniques have been implemented into OpenMC, a continuous energy Monte Carlo code. The UFS method uses weights to alter the number of neutrons born at a fission site. It operates on a superimposed mesh, in which each mesh cell contains a different weight. These weights use an estimate of the source fraction and fuel volume fraction within the cell to produce uniformity. In current implementations, the fuel volumes are assumed to be dispersed equally over all mesh cells. This work aims to provide an estimate of the fuel volume fraction in each cell in order to improve the accuracy of the method for irregular geometries. The new fuel volume approximation method is tested on a toy problem and on a model of the Advanced Test Reactor, a core with highly irregular geometry. Figures of merit were calculated for a basic Monte Carlo simulation, a simulation with the standard UFS implementation, and the new UFS method with estimated volume fractions. With the toy problem, the new method showed significant improvement and had the highest figure of merit. In the case of the ATR, the long run time for the approximation lowered the figure of merit. Both problems demonstrated that the use of the standard UFS implementation on an irregular geometry produced higher uncertainties than not using the method at all. The UFS method, when used with the estimated volume fractions, behaved as expected and produced uniform uncertainty distributions. The investigation of the use of the UFS method with CMFD acceleration was conducted using the 3-D BEAVRS benchmark. Results showed that keeping CMFD acceleration on during active batches maintained a stationary source and reduced the variance for assembly results. The UFS method stacked on this, reducing the maximum relative uncertainties. The UFS method had variable results with different tallies, but no interference between the two methods was observed.
Download or read book Nuclear Science Abstracts written by and published by . This book was released on 1975 with total page 1146 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Download or read book Nuclear Science Abstracts written by and published by . This book was released on 1975-11 with total page 960 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Download or read book Monte Carlo and Thermal Hydraulic Coupling Using Low order Nonlinear Diffusion Acceleration written by Bryan Robert Herman and published by . This book was released on 2014 with total page 147 pages. Available in PDF, EPUB and Kindle. Book excerpt: Monte Carlo (MC) methods for reactor analysis are most often employed as a benchmark tool for other transport and diffusion methods. In this work, we identify and resolve a few of the issues associated with using MC as a reactor design tool. It is widely thought that MC tallies converge at an ideal rate proportional to the inverse of the square root of the number of tally batches. This is true only if tally batches are independent from one another. For a high dominance ratio light water reactor such as the BEAVRS model, significant correlation is present and the convergence rate was much slower. This work developed a means for analytically predicting tally convergence rates when batches are correlated. Analyses supported these findings and confirmed less than ideal convergence rates. For highly correlated problems, it is recommended to reduce error by running additional independent simulations, rather than increasing the number of neutrons in each individual simulation through additional batches. Before tallies can be accumulated, the fission source must be stationary. For the BEAVRS model, this took approximately 200 fission source generations. This process can be accelerated by using coarse mesh finite difference (CMFD), a nonlinear diffusion acceleration method. CMFD was implemented in the continuous-energy MC code OpenMC. When employing this technique, the number of inactive generations was reduced by a factor of 10. Realistic reactor calculations also require thermal hydraulic (TH) feedback which was integrated into the source convergence process. The use of CMFD in addition to TH reduced the number of fission source generations by a factor of 3. Further reduction was achieved by performing nonlinear iterations between the low-order CMFD operator and TH model. Support vector regression, a machine learning algorithm, was used to construct coolant density and fuel temperature dependencies of diffusion parameters between each TH update using MC tallies. A framework was introduced to obtain relative pin power distributions with 95% confidence intervals to 1% with continuous-energy Monte Carlo coupled to thermal hydraulics using low-order CMFD iterations.
Download or read book ERDA Energy Research Abstracts written by United States. Energy Research and Development Administration and published by . This book was released on 1976 with total page 1320 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Download or read book Discrete ordinates Cost Optimization of Weight dependent Variance Reduction Techniques for Monte Carlo Neutral Particle Transport written by Clell J. Jr Solomon and published by . This book was released on 2010 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: A method for deterministically calculating the population variances of Monte Carlo particle transport calculations involving weight-dependent variance reduction has been developed. This method solves a set of equations developed by Booth and Cashwell [1979], but extends them to consider the weight-window variance reduction technique. Furthermore, equations that calculate the duration of a single history in an MCNP5 (RSICC version 1.51) calculation have been developed as well. The calculation cost, defined as the inverse figure of merit, of a Monte Carlo calculation can be deterministically minimized from calculations of the expected variance and expected calculation time per history. The method has been applied to one- and two-dimensional multi-group and mixed material problems for optimization of weight-window lower bounds. With the adjoint (importance) function as a basis for optimization, an optimization mesh is superimposed on the geometry. Regions of weight-window lower bounds contained within the same optimization mesh element are optimized together with a scaling parameter. Using this additional optimization mesh restricts the size of the optimization problem, thereby eliminating the need to optimize each individual weight-window lower bound. Application of the optimization method to a one-dimensional problem, designed to replicate the variance reduction iron-window effect, obtains a gain in efficiency by a factor of 2 over standard deterministically generated weight windows. The gain in two dimensional problems varies. For a 2-D block problem and a 2-D two-legged duct problem, the efficiency gain is a factor of about 1.2. The top-hat problem sees an efficiency gain of 1.3, while a 2-D 3-legged duct problem sees an efficiency gain of only 1.05. This work represents the first attempt at deterministic optimization of Monte Carlo calculations with weight-dependent variance reduction. However, the current work is limited in the size of problems that can be run by the amount of computer memory available in computational systems. This limitation results primarily from the added discretization of the Monte Carlo particle weight required to perform the weight-dependent analyses. Alternate discretization methods for the Monte Carlo weight should be a topic of future investigation. Furthermore, the accuracy with which the MCNP5 calculation times can be calculated deterministically merits further study.
Download or read book ERDA Energy Research Abstracts written by United States. Energy Research and Development Administration. Technical Information Center and published by . This book was released on 1976 with total page 588 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Download or read book Development of High Fidelity Methods for 3D Monte Carlo Transient Analysis of Nuclear Reactors written by Samuel Christopher Shaner and published by . This book was released on 2018 with total page 140 pages. Available in PDF, EPUB and Kindle. Book excerpt: Monte Carlo is increasingly being used to perform high-fidelity, steady-state neutronics analysis of power reactor geometries on today’s leadership class supercomputers. Extending Monte Carlo to time dependent problems has proven to be a formidable challenge due to the significant computational resource and data processing requirements. In this thesis, a transient methodology is proposed and implemented to enable accurate and computationally tractable time dependent Monte Carlo analysis. The frequency transform method has been described and implemented in Monte Carlo for the first time. The attractiveness of this method lies in its ability to accurately capture the space and time dependent distribution of the delayed neutron source throughout a transient. Nuances to the algorithmic implementation are described and validated through a series of simple analytical test problems. Comparison with the adiabatic method currently employed for Monte Carlo transient analysis shows significant improvement in the spatial distribution and magnitude of the power for a negative reactivity insertion transient in the 2D and 3D C5G7 geometry. To aid in understanding the effect of statistical uncertainty in the tallied quantities on the time dependent flux solution, a simplified point kinetics model was developed and used for insightful analysis on simple transient test problems. This revealed how the time dependent flux profiles for a series of independent trials can be approximated by a normal distribution at low uncertainties in the tallied reactivity, but deviates from a normal distribution at relatively modest uncertainties in reactivity. Given the compuational constraints of solving large problems, having a simple model that can provide insight on the expected behavior and flux distribution can be very valuable. The frequency transform methodology belongs to a class of indirect space-time factorization methods that perform high-order calculations (e.g. Monte Carlo) over long time steps and low-order, computationally-efficient calculations (e.g. Point Kinetics) over short time steps as an approach to balance performance and accuracy. The coarse mesh finite difference (CMFD) diffusion operator is employed as the low-order solver in Monte Carlo transient analysis for the first time. The CMFD diffusion operator is attractive due to its potential to increase the time step size between the computationally expensive high-order solves. Implementing this methodology is important as continuous energy Monte Carlo is reactor-agnostic and able to treat complex geometries without difficulty, opening up the possibility of solving transients on new experimental geometries for which there is little data.
Download or read book Advanced Modeling and Simulation of Nuclear Reactors written by Jingang Liang and published by Frontiers Media SA. This book was released on 2023-04-10 with total page 161 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Download or read book INIS Atomindex written by and published by . This book was released on 1986 with total page 1352 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Download or read book ERDA Energy Research Abstracts written by United States. Energy Research and Development Administration and published by . This book was released on 1976 with total page 584 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Download or read book Energy Research Abstracts written by and published by . This book was released on 1988 with total page 1684 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Download or read book Handbook of Nuclear Reactors Calculants Volume II written by Yigal Ronen and published by CRC Press. This book was released on 1987-06-30 with total page 544 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Download or read book Nuclear Reactor Analysis written by James J. Duderstadt and published by Wiley. This book was released on 1991-01-16 with total page 0 pages. Available in PDF, EPUB and Kindle. Book excerpt: Classic textbook for an introductory course in nuclear reactor analysis that introduces the nuclear engineering student to the basic scientific principles of nuclear fission chain reactions and lays a foundation for the subsequent application of these principles to the nuclear design and analysis of reactor cores. This text introduces the student to the fundamental principles governing nuclear fission chain reactions in a manner that renders the transition to practical nuclear reactor design methods most natural. The authors stress throughout the very close interplay between the nuclear analysis of a reactor core and those nonnuclear aspects of core analysis, such as thermal-hydraulics or materials studies, which play a major role in determining a reactor design.
Download or read book Fluka written by Alfredo Ferrari and published by . This book was released on 2005 with total page 410 pages. Available in PDF, EPUB and Kindle. Book excerpt:
