Download or read book Generation of Nuclear Data for Lead cooled Fast Reactors Using the Monte Carlo Method written by Carlos Garcia Domínguez and published by . This book was released on 2012 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: The LEADER project goal is to improve and develop a scaled demonstrator of the LFR technology, ALFRED. The work in this thesis is focused in the ALFRED project framework and its mission is to obtain few-group cross section data for LFRs. Cross sections The neutron transport problem is crucial in nuclear engineering and nuclear reactor physics. Neutron transport theory and the diffusion theory applied to neutron reactions are briefly described, including their principles and hypothesis. The two different computational approaches to solve the neutron transport problem are summarized. The software used to obtain the data is based on a modification of the Monte Carlo method. Thus, some basic probability theory concepts are introduced. This section follows with the discussion of the Monte Carlo method and its principles, and how it can be applied to solve the neutron transport problem. Afterwards, the Serpent code is explained, as well as its features and characteristics. The process of creating a 2-dimension model of ALFRED fuel assembly and the elaboration of Serpent input files are detailed. Cross section data for five neutron energy groups and at different material temperatures is obtained by running several simulations using Serpent. The last section includes a brief description of LFR technology and some specific ALFRED features. Some advantages and disadvantages of LFRs are included, along with some proposals to solve the disadvantages. The last part of this section illustrates the proposed ALFRED core scheme, using the data publicly available to date.

Download or read book On the fly Nuclear Data Processing Methods for Monte Carlo Simulations of Intermediate and Fast Spectrum Systems written by Jonathan Alan Walsh and published by . This book was released on 2016 with total page 212 pages. Available in PDF, EPUB and Kindle. Book excerpt: Computational methods for on-the-fly representation and processing of nuclear data within Monte Carlo neutron transport simulations of intermediate and fast spectrum systems are developed and implemented in a continuous-energy Monte Carlo code. First, a capability to compute temperature-dependent unresolved resonance region (URR) cross sections directly from zero-temperature average resonance parameters is presented. The use of this capability in benchmarking both evaluated and processed URR data is demonstrated. Results of this benchmarking lead to a partial resolution of a longstanding discrepancy between experiment and calculation results for a well-known fast critical assembly. Next, an on-the-fly probability table interpolation scheme for computing temperature-dependent URR cross sections is developed and used in analyses which show that interpolation on a relatively coarse temperature mesh (>100 K) can be used to reproduce results obtained with cross sections generated at an exact temperature. This enables the simulation of systems having detailed temperature distributions using probability table data which require significantly less memory than data generated on a fine temperature mesh. Additional methods for use in the investigation of two common approximations that are made in representing URR cross section data are developed. Namely, a multi-level URR cross section calculation capability is used to show that level-level interference effects in elastic scattering cross sections are negligible in many cases of interest. A capability to generate resonance structure in competitive reaction cross sections is used to show that neglecting cross section structure for reactions other than elastic scattering, capture, and fission can lead to non-negligible, unconservative biases (>100 pcm) in criticality safety calculations. The principal underlying assumption of the probability table method is also tested by comparing the results it yields with results that are averaged over many independent simulations, each using a single, independent realization of URR resonance parameters. Unknown URR resonance structure is observed to induce an uncertainty on the multiplication factor for intermediate and fast spectrum systems that is nearly an order of magnitude greater than that which is purely stochastic. This significantly increases the uncertainty to which results of simulations of those systems should be stated. Finally, a procedure for consistent, on-the-fly sampling of temperature-dependent neutron reaction kernels which requires no additional secondary distribution data is presented. It is used to show that Doppler effects may have only a small impact on elastic scattering secondary angular distributions at typical power reactor operating temperatures but can be appreciable at astrophysical temperatures.

Download or read book Handbook of Generation IV Nuclear Reactors written by Igor Pioro and published by Woodhead Publishing. This book was released on 2022-12-07 with total page 1112 pages. Available in PDF, EPUB and Kindle. Book excerpt: Handbook of Generation IV Nuclear Reactors, Second Edition is a fully revised and updated comprehensive resource on the latest research and advances in generation IV nuclear reactor concepts. Editor Igor Pioro and his team of expert contributors have updated every chapter to reflect advances in the field since the first edition published in 2016. The book teaches the reader about available technologies, future prospects and the feasibility of each concept presented, equipping them users with a strong skillset which they can apply to their own work and research. - Provides a fully updated, revised and comprehensive handbook dedicated entirely to generation IV nuclear reactors - Includes new trends and developments since the first publication, as well as brand new case studies and appendices - Covers the latest research, developments and design information surrounding generation IV nuclear reactors

Download or read book Nuclear Data Uncertainty Propagation for a Lead cooled Fast Reactor Combining TMC with Criticality Benchmarks for Improved Accuracy written by and published by . This book was released on 2014 with total page 100 pages. Available in PDF, EPUB and Kindle. Book excerpt:

Download or read book Monte Carlo Method written by and published by . This book was released on 1959 with total page 14 pages. Available in PDF, EPUB and Kindle. Book excerpt:

Download or read book Nuclear Reactors written by Amir Mesquita and published by BoD – Books on Demand. This book was released on 2012-02-10 with total page 354 pages. Available in PDF, EPUB and Kindle. Book excerpt: This book presents a comprehensive review of studies in nuclear reactors technology from authors across the globe. Topics discussed in this compilation include: thermal hydraulic investigation of TRIGA type research reactor, materials testing reactor and high temperature gas-cooled reactor; the use of radiogenic lead recovered from ores as a coolant for fast reactors; decay heat in reactors and spent-fuel pools; present status of two-phase flow studies in reactor components; thermal aspects of conventional and alternative fuels in supercritical water?cooled reactor; two-phase flow coolant behavior in boiling water reactors under earthquake condition; simulation of nuclear reactors core; fuel life control in light-water reactors; methods for monitoring and controlling power in nuclear reactors; structural materials modeling for the next generation of nuclear reactors; application of the results of finite group theory in reactor physics; and the usability of vermiculite as a shield for nuclear reactor.

Download or read book Improving Fuel Cycle Design and Safety Characteristics of a Gas Cooled Fast Reactor written by Willem Frederik Geert van Rooijen and published by IOS Press. This book was released on 2006 with total page 160 pages. Available in PDF, EPUB and Kindle. Book excerpt: The Generation IV Forum is an international nuclear energy research initiative aimed at developing the fourth generation of nuclear reactors, envisaged to enter service halfway the 21st century. One of the Generation IV reactor systems is the Gas Cooled Fast Reactor (GCFR), the subject of study in this thesis. The Generation IV reactor concepts should improve all aspects of nuclear power generation. Within Generation IV, the GCFR concept specifically targets sustainability of nuclear power generation. The Gas Cooled Fast Reactor core power density is high in comparison to other gas cooled reactor concepts. Like all nuclear reactors, the GCFR produces decay heat after shut down, which has to be transported out of the reactor under all circumstances. The layout of the primary system therefore focuses on using natural convection Decay Heat Removal (DHR) where possible, with a large coolant fraction in the core to reduce friction losses.

Download or read book Computational Methods for Efficient Nuclear Data Management in Monte Carlo Neutron Simulations written by Jonathan Alan Walsh and published by . This book was released on 2014 with total page 133 pages. Available in PDF, EPUB and Kindle. Book excerpt: This thesis presents the development and analysis of computational methods for efficiently accessing and utilizing nuclear data in Monte Carlo neutron transport code simulations. Using the OpenMC code, profiling studies are conducted in order to determine the types of nuclear data that are used in realistic reactor physics simulations, as well as the frequencies with which those data are accessed. The results of the profiling studies are then used to motivate the conceptualization of a nuclear data server algorithm aimed at reducing on-node memory requirements through the use of dedicated server nodes for the storage of infrequently accessed data. A communication model for this algorithm is derived and used to make performance predictions given data access frequencies and assumed system hardware parameters. Additionally, a new, accelerated approach for rejection sampling the free gas resonance elastic scattering kernel that reduces the frequency of zero-temperature elastic scattering cross section data accesses is derived and implemented. Using this new approach, the runtime overhead incurred by an exact treatment of the free gas resonance elastic scattering kernel is reduced by more than 30% relative to a standard sampling procedure used by Monte Carlo codes. Finally, various optimizations of the commonly-used binary energy grid search algorithm are developed and demonstrated. Investigated techniques include placing kinematic constraints on the range of the searchable energy grid, index lookups on unionized material energy grids, and employing energy grid hash tables. The accelerations presented routinely result in overall code speedup by factors of 1.2-1.3 for simulations of practical systems.

Download or read book Structural Materials for Generation IV Nuclear Reactors written by Pascal Yvon and published by Woodhead Publishing. This book was released on 2016-08-27 with total page 686 pages. Available in PDF, EPUB and Kindle. Book excerpt: Operating at a high level of fuel efficiency, safety, proliferation-resistance, sustainability and cost, generation IV nuclear reactors promise enhanced features to an energy resource which is already seen as an outstanding source of reliable base load power. The performance and reliability of materials when subjected to the higher neutron doses and extremely corrosive higher temperature environments that will be found in generation IV nuclear reactors are essential areas of study, as key considerations for the successful development of generation IV reactors are suitable structural materials for both in-core and out-of-core applications. Structural Materials for Generation IV Nuclear Reactors explores the current state-of-the art in these areas. Part One reviews the materials, requirements and challenges in generation IV systems. Part Two presents the core materials with chapters on irradiation resistant austenitic steels, ODS/FM steels and refractory metals amongst others. Part Three looks at out-of-core materials. Structural Materials for Generation IV Nuclear Reactors is an essential reference text for professional scientists, engineers and postgraduate researchers involved in the development of generation IV nuclear reactors. - Introduces the higher neutron doses and extremely corrosive higher temperature environments that will be found in generation IV nuclear reactors and implications for structural materials - Contains chapters on the key core and out-of-core materials, from steels to advanced micro-laminates - Written by an expert in that particular area

Download or read book Fast Spectrum Reactors written by Alan E. Waltar and published by Springer Science & Business Media. This book was released on 2011-09-28 with total page 717 pages. Available in PDF, EPUB and Kindle. Book excerpt: This book is a complete update of the classic 1981 FAST BREEDER REACTORS textbook authored by Alan E. Waltar and Albert B. Reynolds, which , along with the Russian translation, served as a major reference book for fast reactors systems. Major updates include transmutation physics (a key technology to substantially ameliorate issues associated with the storage of high-level nuclear waste ), advances in fuels and materials technology (including metal fuels and cladding materials capable of high-temperature and high burnup), and new approaches to reactor safety (including passive safety technology), New chapters on gas-cooled and lead-cooled fast spectrum reactors are also included. Key international experts contributing to the text include Chaim Braun, (Stanford University) Ronald Omberg, (Pacific Northwest National Laboratory, Massimo Salvatores (CEA, France), Baldev Raj, (Indira Gandhi Center for Atomic Research, India) , John Sackett (Argonne National Laboratory), Kevan Weaver, (TerraPower Corporation) ,James Seinicki(Argonne National Laboratory). Russell Stachowski (General Electric), Toshikazu Takeda (University of Fukui, Japan), and Yoshitaka Chikazawa (Japan Atomic Energy Agency).

Download or read book Development of a Monte Carlo Model of the University of Wisconsin Nuclear Reactor written by Paul W. Humrickhouse and published by . This book was released on 2006 with total page 164 pages. Available in PDF, EPUB and Kindle. Book excerpt:

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 Development of a New Monte Carlo Reactor Physics Code written by Jaakko Leppänen and published by . This book was released on 2007 with total page 236 pages. Available in PDF, EPUB and Kindle. Book excerpt: Monte Carlo neutron transport codes are widely used in various reactor physics applications, traditionally related to criticality safety analyses, radiation shielding problems, detector modelling and validation of deterministic transport codes. The main advantage of the method is the capability to model geometry and interaction physics without major approximations. The disadvantage is that the modelling of complicated systems is very computing-intensive, which restricts the applications to some extent. The importance of Monte Carlo calculation is likely to increase in the future, along with the development in computer capacities and parallel calculation. An interesting near-future application for the Monte Carlo method is the generation of input parameters for deterministic reactor simulator codes. These codes are used in coupled LWR full-core analyses and typically based on fewgroup nodal diffusion methods. The input data consists of homogenised fewgroup constants, presently generated using deterministic lattice transport codes. The task is becoming increasingly challenging, along with the development in nuclear technology. Calculations involving high-burnup fuels, advanced MOX technology and next-generation reactor systems are likely to cause problems in the future, if code development cannot keep up with the applications. A potential solution is the use of Monte Carlo based lattice transport codes, which brings all the advantages of the calculation method. So far there has been only a handful of studies on group constant generation using the Monte Carlo method, although the interest has clearly increased during the past few years. The homogenisation of reaction cross sections is simple and straightforward, and it can be carried out using any Monte Carlo code. Some of the parameters, however, require the use of special techniques that are usually not available in general-purpose codes. The main problem is the calculation of neutron diffusion coefficients, which have no continuous-energy counterparts in the Monte Carlo calculation. This study is focused on the development of an entirely new Monte Carlo neutron transport code, specifically intended for reactor physics calculations at the fuel assembly level. The PSG code is developed at VTT Technical Research Centre of Finland and one of the main applications is the generation of homogenised group constants for deterministic reactor simulator codes. The theoretical background on general transport theory, nodal diffusion calculation and the Monte Carlo method are discussed. The basic methodology used in the PSG code is introduced and previous studies related to the topic are briefly reviewed. PSG is validated by comparison to reference results produced by MCNP4C and CASMO-4E in infinite two-dimensional LWR lattice calculations. Group constants generated by PSG are used in ARES reactor simulator calculations and the results compared to reference calculations using CASMO-4E data.

Download or read book Handbook of Nuclear Engineering written by Dan Gabriel Cacuci and published by Springer Science & Business Media. This book was released on 2010-09-14 with total page 3701 pages. Available in PDF, EPUB and Kindle. Book excerpt: This is an authoritative compilation of information regarding methods and data used in all phases of nuclear engineering. Addressing nuclear engineers and scientists at all levels, this book provides a condensed reference on nuclear engineering since 1958.

Download or read book MONTE CARLO CALCULATION OF THE NUCLEAR TEMPERATURE COEFFICIENT IN FAST REACTORS written by and published by . This book was released on 1963 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt:

Download or read book The Preparation of Nuclear Data Into a Form Suitable for Monte Carlo Calculations Using the IBM7030 Computer written by W.M.M. Kerr and published by . This book was released on 1964 with total page 45 pages. Available in PDF, EPUB and Kindle. Book excerpt:

Download or read book A Computational Benchmark of S N and Monte Carlo Codes Using the Ohio State University Nuclear Reactor Laboratory written by Ryanne Ariel Kennedy and published by . This book was released on 2007 with total page 414 pages. Available in PDF, EPUB and Kindle. Book excerpt: Abstract: In support of the nation's nuclear energy industry, the Innovations in Nuclear Infrastructure and Education (INIE) program was established in 2002 by the Department of Energy. Its function is to strengthen university nuclear engineering education programs through improved and original use of university research and training reactors. The Ohio State University (OSU) is part of the INIE consortium consisting of Penn State University, OSU, Purdue University, University of Illinois (Urbana-Champaign), University of Michigan and University of Wisconsin -- Madison. For improving research reactor utilization and to meet objectives consistent with the goals of the INIE program, a full facility model of the OSU Research Reactor (OSURR) was assembled using the PENTRAN (Parallel Environment Neutral particle Transport) 3-D discrete ordinates code (version 9.36b). The focus of this thesis is the creation and benchmark of a full facility model of the OSU Research Reactor using the discrete ordinates transport code PENTRAN and the Monte Carlo code MCNP5. Storing the full phase-space information for an exact geometry model of the OSU Research Reactor using the discrete ordinates code PENTRAN would require a few thousand gigabytes of computer memory. This large memory requirement is a result of the fine spatial meshing essential for modeling the very thin layers of cladding and fuel over the whole core. Such a large model is unrealistically cumbersome even considering the parallel memory and phase-space decomposition capability of PENTRAN. Hence, it was essential to consider some level of homogenization of different material regions including fuel, clad, and/or moderator/coolant in the discrete ordinates model. Several parametric analyses were performed in an attempt to understand the impact of systematic uncertainties in the models that occur as a result of modeling approximations and the homogenization of core regions. These parametric studies were performed using PENTRAN and MCNP and included the analysis of several categories of uncertainty in the research reactor such as fuel impurities and uncertainty in the core's geometry. In addition to analyses of the PENTRAN model, several tests were performed during the construction of the MCNP model. Detailed studies of the control rods and irradiation facilities were performed and compared to experimental data. An irradiation experiment was performed to collect data in three irradiation facilities in the OSURR facility. The thermal flux was calculated in each of these locations using the experimental data and compared directly to the results of the full core models built with MCNP and PENTRAN. In addition to benchmarking the model flux results with this experiment, an eigenvalue comparison was made for three different rod configurations for both codes. Overall, agreement was seen between experimental data, MCNP results, and PENTRAN results. The eigenvalue results from different rod configurations were within the uncertainty that was calculated from parametric analyses of the OSURR core. The flux distributions over the core matched well between MCNP and PENTRAN, and all discrepancies were accounted for by analysis of the homogenization effects and other differences between the models.