Download or read book Coupling Between Monte Carlo Neutron Transport and Thermal hydraulics for the Simulation of Transients Due to Reactivity Insertions written by Margaux Faucher and published by . This book was released on 2019 with total page 0 pages. Available in PDF, EPUB and Kindle. Book excerpt: One of the main issues for the study of a reactor behaviour is to model the propagation of the neutrons, described by the Boltzmann transport equation, in the presence of multi-physics phenomena, such as the coupling between neutron transport, thermal-hydraulics and thermomecanics. Thanks to the growing computer power, it is now feasible to apply Monte Carlo methods to the solution of non-stationary transport problems in reactor physics, which play an instrumental role in producing reference numerical solutions for the analysis of transients occurring during normal and accidental behaviour.The goal of this Ph. D. thesis is to develop, verify and test a coupling scheme between the Monte Carlo code TRIPOLI-4 and thermal-hydraulics, so as to provide a reference tool for the simulation of reactivity-induced transients in PWRs.We have first tested the kinetic capabilities of TRIPOLI-4 (i.e., time dependent without thermal-hydraulics feedback), evaluating the different existing methods and implementing new techniques. Then, we have developed a multi-physics interface for TRIPOLI-4, and more specifically a coupling scheme between TRIPOLI-4 and the thermal-hydraulics sub-channel code SUBCHANFLOW. Finally, we have performed a preliminary analysis of the stability of the coupling scheme. Indeed, due to the stochastic nature of the outputs produced by TRIPOLI-4, uncertainties are inherent to our coupling scheme and propagate along the coupling iterations. Moreover, thermal-hydraulics equations are non linear, so the prediction of the propagation of the uncertainties is not straightforward.

Download or read book Modeling Feedback Effects of Transient Nuclear Systems Using Monte Carlo written by Miriam A. Kreher and published by . This book was released on 2023 with total page 0 pages. Available in PDF, EPUB and Kindle. Book excerpt: Monte Carlo neutron transport is the gold standard for accurate neutronics simulation of nuclear reactors in steady-state because each term of the neutron transport equation can be directly tallied using continuous-energy cross sections rather than needing to make approximations in energy, angle, or geometry. However, the time dependent equation includes time derivatives of flux and delayed neutron precursors which are difficult to tally. While it is straightforward to explicitly model delayed neutron precursors, and thus solve the time dependent problem in Direct Monte Carlo, this is such a costly approach that the practical length of transient calculations is limited to about 1 second. In order to solve longer problems, a high-order/low-order approach was adopted that uses the omega method to approximate the time derivatives as frequencies. These frequencies are spatially distributed and provided by a low-order Time Dependent Coarse Mesh Finite Difference diffusion solver. While this scheme has been previously applied to prescribed transients, thermal feedback is now incorporated to provide a fully self-propagating Monte Carlo transient multiphysics solver which can be applied to transients of several seconds long. Several recently developed techniques are used in the implementation of the proposed coupling approaches. Firstly, underrelaxed Monte Carlo, which is a steady-state technique that stabilizes the search for temperature distributions, is applied to find initial conditions. Secondly, tally derivatives are a Monte Carlo perturbation technique that can identify how a tally will change with respect to a small change in the system. Test problems of varying complexity are carried out in flow-initiated transients to show the versatility of these methods. Overall, this multi-level, multiphysics, transient solver provides a bridge between high fidelity Monte Carlo neutronics and the fast multi-group diffusion methods that are currently used in safety analysis.

Download or read book Summary Review on the Application of Computational Fluid Dynamics in Nuclear Power Plant Design written by IAEA and published by International Atomic Energy Agency. This book was released on 2022-03-28 with total page 121 pages. Available in PDF, EPUB and Kindle. Book excerpt: This publication documents the results of an IAEA coordinated research project (CRP)on the application of computational fluid dynamics (CFD) codes for nuclear power plant design. The main objective was to benchmark CFD codes, model options and methods against CFD experimental data under single phase flow conditions. This publication summarizes the current capabilities and applications of CFD codes, and their present qualification level, with respect to nuclear power plant design requirements. It is not intended to be comprehensive, focusing instead on international experience in the practical application of these tools in designing nuclear power plant components and systems. The guidance in this publication is based on inputs provided by international nuclear industry experts directly involved in nuclear power plant design issues, CFD applications, and in related experimentation and validation highlighted during the CRP.

Download or read book Numerical Methods and Computational Sciences Applied to Nuclear Energy written by Yue Jin and published by Frontiers Media SA. This book was released on 2022-11-11 with total page 153 pages. Available in PDF, EPUB and Kindle. Book excerpt:

Download or read book Nuclear Engineering written by Zafar Ullah Koreshi and published by Academic Press. This book was released on 2022-03-23 with total page 549 pages. Available in PDF, EPUB and Kindle. Book excerpt: Nuclear Engineering Mathematical Modeling and Simulation presents the mathematical modeling of neutron diffusion and transport. Aimed at students and early career engineers, this highly practical and visual resource guides the reader through computer simulations using the Monte Carlo Method which can be applied to a variety of applications, including power generation, criticality assemblies, nuclear detection systems, and nuclear medicine to name a few. The book covers optimization in both the traditional deterministic framework of variational methods and the stochastic framework of Monte Carlo methods. Specific sections cover the fundamentals of nuclear physics, computer codes used for neutron and photon radiation transport simulations, applications of analyses and simulations, optimization techniques for both fixed-source and multiplying systems, and various simulations in the medical area where radioisotopes are used in cancer treatment. - Provides a highly visual and practical reference that includes mathematical modeling, formulations, models and methods throughout - Includes all current major computer codes, such as ANISN, MCNP and MATLAB for user coding and analysis - Guides the reader through simulations for the design optimization of both present-day and future nuclear systems

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 Coupling Interface for Physics to System Simulations written by Michael Lee Leimon and published by . This book was released on 2013 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: A new interfacial code was developed to couple the reactor physics code PARCS/AGREE to the systems level code MELCOR, with a goal of enabling state- of-art transient event analysis for high temperature gas reactor designs. Following the completion of this new code, it was then demonstrated by running two different coupled simulations, one of which was a transient event. The resultant code is capable of coupling spatial power profiles, point kinetics information and transient reactivity values from PARCS/AGREE to MELCOR by means of input/output file manipulation. The coupling demonstrations were between PBMR400 models that were designed to have an equivalent core region nodalization to that which was used in the OECD/NEA PBMR400 benchmark, thus allowing for comparisons. The accessible coupled simulation output results as extracted from MELCOR appeared to be overly generalized. Even so, the axial profiles from the coupled steady-state demonstration were in good agreement with the axial profiles of other OECD/NEA participants. Conversely, the coupled transient simulations showed a suspect, maximum average nodal component temperature rise of approximately 0.4K from a 3+$ reactivity insertion. The electronic version of this dissertation is accessible from http://hdl.handle.net/1969.1/148367

Download or read book Nuclear Power written by Pavel Tsvetkov and published by BoD – Books on Demand. This book was released on 2011-09-06 with total page 208 pages. Available in PDF, EPUB and Kindle. Book excerpt: At the onset of the 21st century, we are searching for reliable and sustainable energy sources that have a potential to support growing economies developing at accelerated growth rates, technology advances improving quality of life and becoming available to larger and larger populations. The quest for robust sustainable energy supplies meeting the above constraints leads us to the nuclear power technology. Today's nuclear reactors are safe and highly efficient energy systems that offer electricity and a multitude of co-generation energy products ranging from potable water to heat for industrial applications. Catastrophic earthquake and tsunami events in Japan resulted in the nuclear accident that forced us to rethink our approach to nuclear safety, requirements and facilitated growing interests in designs, which can withstand natural disasters and avoid catastrophic consequences. This book is one in a series of books on nuclear power published by InTech. It consists of ten chapters on system simulations and operational aspects. Our book does not aim at a complete coverage or a broad range. Instead, the included chapters shine light at existing challenges, solutions and approaches. Authors hope to share ideas and findings so that new ideas and directions can potentially be developed focusing on operational characteristics of nuclear power plants. The consistent thread throughout all chapters is the "system-thinking" approach synthesizing provided information and ideas. The book targets everyone with interests in system simulations and nuclear power operational aspects as its potential readership groups - students, researchers and practitioners.

Download or read book A Monte Carlo Code for the Transport of Neutrons written by William V. Baxter and published by . This book was released on 1959 with total page 40 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 1973 with total page 1132 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 A Fully Coupled Monte Carlo discrete Ordinates Solution to the Neutron Transport Equation Final Report written by and published by . This book was released on 1990 with total page 211 pages. Available in PDF, EPUB and Kindle. Book excerpt: The neutron transport equation is solved by a hybrid method that iteratively couples regions where deterministic (S{sub N}) and stochastic (Monte Carlo) methods are applied. Unlike previous hybrid methods, the Monte Carlo and S{sub N} regions are fully coupled in the sense that no assumption is made about geometrical separation or decoupling. The hybrid method provides a new means of solving problems involving both optically thick and optically thin regions that neither Monte Carlo nor S{sub N} is well suited for by themselves. The fully coupled Monte Carlo/S{sub N} technique consists of defining spatial and/or energy regions of a problem in which either a Monte Carlo calculation or an S{sub N} calculation is to be performed. The Monte Carlo region may comprise the entire spatial region for selected energy groups, or may consist of a rectangular area that is either completely or partially embedded in an arbitrary S{sub N} region. The Monte Carlo and S{sub N} regions are then connected through the common angular boundary fluxes, which are determined iteratively using the response matrix technique, and volumetric sources. The hybrid method has been implemented in the S{sub N} code TWODANT by adding special-purpose Monte Carlo subroutines to calculate the response matrices and volumetric sources, and linkage subrountines to carry out the interface flux iterations. The common angular boundary fluxes are included in the S{sub N} code as interior boundary sources, leaving the logic for the solution of the transport flux unchanged, while, with minor modifications, the diffusion synthetic accelerator remains effective in accelerating S{sub N} calculations. The special-purpose Monte Carlo routines used are essentially analog, with few variance reduction techniques employed. However, the routines have been successfully vectorized, with approximately a factor of five increase in speed over the non-vectorized version.

Download or read book Accuracy and Efficiency of a Coupled Neutronics and Thermal Hydraulics Model written by and published by . This book was released on 2007 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: The accuracy requirements for modern nuclear reactor simulation are steadily increasing due to the cost and regulation of relevant experimental facilities. Because of the increase in the cost of experiments and the decrease in the cost of simulation, simulation will play a much larger role in the design and licensing of new nuclear reactors. Fortunately as the work load of simulation increases, there are better physics models, new numerical techniques, and more powerful computer hardware that will enable modern simulation codes to handle the larger workload. This manuscript will discuss a numerical method where the six equations of two-phase flow, the solid conduction equations, and the two equations that describe neutron diffusion and precursor concentration are solved together in a tightly coupled, nonlinear fashion for a simplified model of a nuclear reactor core. This approach has two important advantages. The first advantage is a higher level of accuracy. Because the equations are solved together in a single nonlinear system, the solution is more accurate than the traditional "operator split" approach where the two-phase flow equations are solved first, the heat conduction is solved second and the neutron diffusion is solved third, limiting the temporal accuracy to 1st order because the nonlinear coupling between the physics is handled explicitly. The second advantage of the method described in this manuscript is that the time step control in the fully implicit system can be based on the timescale of the solution rather than a stability-based time step restriction like the material Courant. Results are presented from a simulated control rod movement and a rod ejection that address temporal accuracy for the fully coupled solution and demonstrate how the fastest timescale of the problem can change between the state variables of neutronics, conduction and two-phase flow during the course of a transient.

Download or read book Energy Research Abstracts written by and published by . This book was released on 1993 with total page 694 pages. Available in PDF, EPUB and Kindle. Book excerpt:

Download or read book Thermal Hydraulics Aspects of Liquid Metal Cooled Nuclear Reactors written by Ferry Roelofs and published by Woodhead Publishing. This book was released on 2018-11-30 with total page 464 pages. Available in PDF, EPUB and Kindle. Book excerpt: Thermal Hydraulics Aspects of Liquid Metal cooled Nuclear Reactors is a comprehensive collection of liquid metal thermal hydraulics research and development for nuclear liquid metal reactor applications. A deliverable of the SESAME H2020 project, this book is written by top European experts who discuss topics of note that are supplemented by an international contribution from U.S. partners within the framework of the NEAMS program under the U.S. DOE. This book is a convenient source for students, professionals and academics interested in liquid metal thermal hydraulics in nuclear applications. In addition, it will also help newcomers become familiar with current techniques and knowledge. - Presents the latest information on one of the deliverables of the SESAME H2020 project - Provides an overview on the design and history of liquid metal cooled fast reactors worldwide - Describes the challenges in thermal hydraulics related to the design and safety analysis of liquid metal cooled fast reactors - Includes the codes, methods, correlations, guidelines and limitations for liquid metal fast reactor thermal hydraulic simulations clearly - Discusses state-of-the-art, multi-scale techniques for liquid metal fast reactor thermal hydraulics applications

Download or read book Monte Carlo Simulation of Neutron Transport in a Homogeneous Reactor with a Partially Inserted Control Rod written by J.K-H. Karlsson and published by . This book was released on 1997 with total page 30 pages. Available in PDF, EPUB and Kindle. Book excerpt:

Download or read book Computational Treatments for Neutron Resonance Elastic Scattering in Monte Carlo Nuclear Simulations written by Vivian Y. Tran and published by . This book was released on 2016 with total page 60 pages. Available in PDF, EPUB and Kindle. Book excerpt: Simulations are vital to the safe design and operation of nuclear reactors. It is therefore important that they accurately treat the physics of nuclear interactions. This work investigates the phenomenon of neutron resonance scattering with a moving target, which can affect the post-collision properties of the neutron and macroscopic values such as temperature reactivity coefficients. First, this research validates a faster computational treatment for resonance scattering--the accelerated resonance elastic scattering (ARES) kernel sampling method--against the already verified Doppler broadening rejection correction (DBRC) treatment in the open-source OpenMC Monte Carlo neutron transport code being developed at the Massachusetts Institute of Technology. In an effort to improve computational efficiency, the optimal energy limits where this phenomenon should be treated are determined and compared to the less costly but inaccurate approximation of assuming a constant constant cross section for determining the reaction kinematics. To reduce memory requirements and facilitate coupling with heat transfer, a new data representation was recently adopted in OpenMC based on the multipole formalism. However, this new approach invalidates the previous implementations of DBRC and ARES. This thesis thus developed a modified DBRC algorithm compatible with the new data representation. This new method is also validated against the previous DBRC method. While more computationally costly, the use of the multipole representation in treating resonance scattering reduces memory requirements by a hundred-fold and facilitates the representation of temperature dependent cross sections.