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Improvements And Applications Of The Uniform Fission Site Method In Monte Carlo
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Book Synopsis Improvements and Applications of the Uniform Fission Site Method in Monte Carlo by : Jessica Lynn Hunter
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.
Book Synopsis Monte Carlo Simulation Research on the Spontaneous Fission Yield of 240Pu by : Tianyou Xie
Download or read book Monte Carlo Simulation Research on the Spontaneous Fission Yield of 240Pu written by Tianyou Xie and published by . This book was released on 2015 with total page 103 pages. Available in PDF, EPUB and Kindle. Book excerpt: This research describes results of mathematical simulations to predict the spontaneous fission yield of 240Pu thereby guiding selection of a suitable fission product for use as a radiochronometer for weapons grade plutonium. The PUREX process is effective in separating plutonium and uranium in irradiated nuclear fuel from fission products, which may make it possible to use one or more of the spontaneous fission products from 240Pu as a radiochronometer. However, the fission product inventory from the spontaneous fission of 240Pu reported in the literature is quite variable. The non-uniform induced fission probability at multiple neutron energies makes the induced fission yield too complex to be listed as a single value. The neutron flux distribution from a source containing 240Pu is a combination of both spontaneous and induced fissions. Fission products generated in this source are due to a combination of spontaneous and induced fission processes, each with a unique neutron energy distribution and fission product yield. The physical conditions of the source (i.e., size, shape, etc.) will have a significant influence on the neutron flux distribution and the induced fission probability associated with 240Pu. Thus, it is likely that there is a relationship between the combined (spontaneous and induced) fission yield and the physical conditions of the source. Exploring this relationship, a table of values will be generated to predict the combined fission product yield for any source geometry containing 240Pu. Minimizing variations in the source geometry should stabilize the combined fission product yield and provide a means to determine the spontaneous fission probability for 240Pu. This research has identified that the combined fission product yields from 97Zr and 138Xe exhibit less sensitivity to the physical source parameters than other fission products making them good candidates as radiochronometers for age dating weapons-grade plutonium.
Book Synopsis Monte Carlo Source Convergence and the Whitesides Problem by :
Download or read book Monte Carlo Source Convergence and the Whitesides Problem written by and published by . This book was released on 2000 with total page 13 pages. Available in PDF, EPUB and Kindle. Book excerpt: The issue of fission source convergence in Monte Carlo eigenvalue calculations is of interest because of the potential consequences of erroneous criticality safety calculations. In this work, the authors compare two different techniques to improve the source convergence behavior of standard Monte Carlo calculations applied to challenging source convergence problems. The first method, super-history powering, attempts to avoid discarding important fission sites between generations by delaying stochastic sampling of the fission site bank until after several generations of multiplication. The second method, stratified sampling of the fission site bank, explicitly keeps the important sites even if conventional sampling would have eliminated them. The test problems are variants of Whitesides' Criticality of the World problem in which the fission site phase space was intentionally undersampled in order to induce marginally intolerable variability in local fission site populations. Three variants of the problem were studied, each with a different degree of coupling between fissionable pieces. Both the superhistory powering method and the stratified sampling method were shown to improve convergence behavior, although stratified sampling is more robust for the extreme case of no coupling. Neither algorithm completely eliminates the loss of the most important fissionable piece, and if coupling is absent, the lost piece cannot be recovered unless its sites from earlier generations have been retained. Finally, criteria for measuring source convergence reliability are proposed and applied to the test problems.
Book Synopsis Acceleration Methods for Monte Carlo Particle Transport Simulations by : Lulu Li (Ph. D.)
Download or read book Acceleration Methods for Monte Carlo Particle Transport Simulations written by Lulu Li (Ph. D.) and published by . This book was released on 2017 with total page 175 pages. Available in PDF, EPUB and Kindle. Book excerpt: Performing nuclear reactor core physics analysis is a crucial step in the process of both designing and understanding nuclear power reactors. Advancements in the nuclear industry demand more accurate and detailed results from reactor analysis. Monte Carlo (MC) eigenvalue neutron transport methods are uniquely qualified to provide these results, due to their accurate treatment of space, angle, and energy dependencies of neutron distributions. Monte Carlo eigenvalue simulations are, however, challenging, because they must resolve the fission source distribution and accumulate sufficient tally statistics, resulting in prohibitive run times. This thesis proposes the Low Order Operator (LOO) acceleration method to reduce the run time challenge, and provides analyses to support its use for full-scale reactor simulations. LOO is implemented in the continuous energy Monte Carlo code, OpenMC, and tested in 2D PWR benchmarks. The Low Order Operator (LOO) acceleration method is a deterministic transport method based on the Method of Characteristics. Similar to Coarse Mesh Finite Difference (CMFD), the other acceleration method evaluated in this thesis, LOO parameters are constructed from Monte Carlo tallies. The solutions to the LOO equations are then used to update Monte Carlo fission sources. This thesis deploys independent simulations to rigorously assess LOO, CMFD, and unaccelerated Monte Carlo, simulating up to a quarter of a trillion neutron histories for each simulation. Analysis and performance models are developed to address two aspects of the Monte Carlo run time challenge. First, this thesis demonstrates that acceleration methods can reduce the vast number of neutron histories required to converge the fission source distribution before tallies can be accumulated. Second, the slow convergence of tally statistics is improved with the acceleration methods for the earlier active cycles. A theoretical model is developed to explain the observed behaviors and predict convergence rates. Finally, numerical results and theoretical models shed light on the selection of optimal simulation parameters such that a desired statistical uncertainty can be achieved with minimum neutron histories. This thesis demonstrates that the conventional wisdom (e.g., maximizing the number of cycles rather than the number of neutrons per cycle) in performing unaccelerated MC simulations can be improved simply by using more optimal parameters. LOO acceleration provides reduction of a factor of at least 2.2 in neutron histories, compared to the unaccelerated Monte Carlo scheme, and the CPU time and memory overhead associated with LOO are small.
Book Synopsis Monte Carlo Evaluation of the Improvements in Nuclear Materials Identification System (NMIS) Resulting From a DT Neutron Generator by :
Download or read book Monte Carlo Evaluation of the Improvements in Nuclear Materials Identification System (NMIS) Resulting From a DT Neutron Generator written by and published by . This book was released on 2002 with total page 5 pages. Available in PDF, EPUB and Kindle. Book excerpt: Nuclear safeguards active measurements that rely on the time correlation between fast neutrons and gamma rays from the same fission are a promising technique. Previous studies have shown the feasibility of this method, in conjunction with the use of artificial neural networks, to estimate the mass and enrichment of fissile samples enclosed in special, sealed containers. This paper evaluates the use of the associated particle sealed tube neutron generator (APSTNG) as the interrogation source in correlation measurements. The results show that its use is of particular importance when floor reflections are present. The Nuclear Materials Identification System (NMIS) presently uses 252Cf ionization chambers as interrogation sources for the time-dependent coincidence measurements. Because triggers from this source are associated with neutrons emitted in any direction, adjacent materials such as the floor and nearby containers could affect the measurements and should be accounted for. Conversely, the APSTNG, together with an alpha particle detector, defines a cone of neutrons that can be aimed at the item under verification, thus removing the effects of nearby materials from the time-dependent coincidence distributions. Monte Carlo calculations were performed using MCNP-POLIMI, a modified version of the standard MCNP code. The code attempts to calculate more correctly quantities that depend on the second moment of the neutron and gamma distributions. The simulations quantified the sensitivity enhancements and removal of the effects of nearby materials by substituting the traditional 252Cf source with the APSTNG.
Book Synopsis Nuclear Reactor Analysis by : James J. Duderstadt
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.
Book Synopsis Theory, Application, and Implementation of Monte Carlo Method in Science and Technology by :
Download or read book Theory, Application, and Implementation of Monte Carlo Method in Science and Technology written by and published by . This book was released on 2019 with total page 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 1976 with total page 964 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 1994 with total page 1032 pages. Available in PDF, EPUB and Kindle. Book excerpt: Semiannual, with semiannual and annual indexes. References to all scientific and technical literature coming from DOE, its laboratories, energy centers, and contractors. Includes all works deriving from DOE, other related government-sponsored information, and foreign nonnuclear information. Arranged under 39 categories, e.g., Biomedical sciences, basic studies; Biomedical sciences, applied studies; Health and safety; and Fusion energy. Entry gives bibliographical information and abstract. Corporate, author, subject, report number indexes.
Book Synopsis A Guide to Monte Carlo Simulations in Statistical Physics by : David P. Landau
Download or read book A Guide to Monte Carlo Simulations in Statistical Physics written by David P. Landau and published by Cambridge University Press. This book was released on 2000-08-17 with total page 402 pages. Available in PDF, EPUB and Kindle. Book excerpt: This book describes all aspects of Monte Carlo simulation of complex physical systems encountered in condensed-matter physics and statistical mechanics, as well as in related fields, such as polymer science and lattice gauge theory. The authors give a succinct overview of simple sampling methods and develop the importance sampling method. In addition they introduce quantum Monte Carlo methods, aspects of simulations of growth phenomena and other systems far from equilibrium, and the Monte Carlo Renormalization Group approach to critical phenomena. The book includes many applications, examples, and current references, and exercises to help the reader.
Download or read book INIS Atomindex written by and published by . This book was released on 1987 with total page 880 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Book Synopsis Scientific and Technical Aerospace Reports by :
Download or read book Scientific and Technical Aerospace Reports written by and published by . This book was released on 1992 with total page 292 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Download or read book TID. written by and published by . This book was released on 1964 with total page 478 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Book Synopsis Nuclear Performance of Power-reactor Cores by : American Nuclear Society. Topical Meeting
Download or read book Nuclear Performance of Power-reactor Cores written by American Nuclear Society. Topical Meeting and published by . This book was released on 1964 with total page 480 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Book Synopsis ANS Topical Meeting by : American Nuclear Society. Topical Meeting
Download or read book ANS Topical Meeting written by American Nuclear Society. Topical Meeting and published by . This book was released on 1964 with total page 476 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Author :George Vourvopoulos Publisher :SPIE-International Society for Optical Engineering ISBN 13 : Total Pages :666 pages Book Rating :4.3/5 (91 download)
Book Synopsis Neutrons in Research and Industry by : George Vourvopoulos
Download or read book Neutrons in Research and Industry written by George Vourvopoulos and published by SPIE-International Society for Optical Engineering. This book was released on 1997 with total page 666 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Book Synopsis Applications of Monte Carlo Methods to Finance and Insurance by : Thomas N. Herzog
Download or read book Applications of Monte Carlo Methods to Finance and Insurance written by Thomas N. Herzog and published by ACTEX Publications. This book was released on 2002 with total page 276 pages. Available in PDF, EPUB and Kindle. Book excerpt:
