Download or read book Advances in Fast Reactor Sensitivity and Uncertainty Analysis written by and published by . This book was released on 1978 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: A review of present methods and existing computer codes indicates an enormous capability not only to calculate sensitivity coefficients but also to apply them to a variety of purposes. However, there are still many limitations to our present capabilities. One of these limitations has been our inability to include in a complete and systematic way the effect of methods uncertainties on the determination of adjusted data, which depends, in general, not only on experimental measurements, but also on estimates of covariances associated with the measurements and the methods. Also, the uncertainty in adjusted data contains contributions from uncertainties in covariance estimates which contributions we have heretofore neglected. A new and comprehensive approach to include effects of methods uncertainties is presented here, and all sources which contribute to the uncertainty of the adjusted data are considered. This new approach is demonstrated using rough estimates for the methods uncertainties as applied to a simplified representation of the ZPR-6/7 fast benchmark. The results indicate that it may be essential to include methods uncertainties if integral experiments are to be used for the creation of adjusted nuclear data libraries. A careful evaluation of methods bias and uncertainties must still be performed.

Download or read book Advances in Nuclear Science and Technology written by Jeffery Lewins and published by Springer Science & Business Media. This book was released on 2012-12-06 with total page 384 pages. Available in PDF, EPUB and Kindle. Book excerpt: We have pleasure in presenting Volume Fourteen to our readers. Volume Fourteen signifies a new dimension for our series, a volume devoted to the development of a single timely topic, that of sensitivity to uncertainty. This is still a broad topic and has been treated as such by the se veral distinguished authors contributing to the volume from their extensive experience both in theory and practice. While the theme running through the volume emphasizes uncertainties in areas related to reactor physics, it is true to say that this field of application has much to offer other disciplines as well. Some of the authors are engaged in ex tensions to other areas. The volume may therefore appeal to a much wider audience who will appreciate a single and compre hensive overview of a methodology that is applicable to other fields. Notable developments in the field of nuclear engineering have included the formatting in recent versions of Evaluated Nuclear Data Files (e.g., ENDF/B and its variants) of cross section uncertainty, the general acceptance of good practice in the representation of error correlation matrices, and more recent developments in the application of Monte Carlo tech niques to sensitivity analysis in complex geometries.

Download or read book Systematic Sensitivity and Uncertainty Analysis of Sodium Cooled Fast Reactor Systems written by Friederike Bostelmann and published by . This book was released on 2020 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: Mots-clés de l'auteur: sodium-cooled fast reactor ; uncertainty analysis ; sensitivity analysis ; nuclear data ; random sampling ; XSUSA ; SCALE.

Download or read book Sensitivity and Uncertainty Analysis of Nuclear Data for the Metallic Fueled ABR 1000 Sodium Cooled Fast Reactor written by Jun Shi and published by . This book was released on 2016 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: The 1000 MWt Advanced Burner Reactor (ABR-1000) is a concept of Sodium-cooled Fast Reactor (SFR) developed at Argonne National Laboratory for the study of future reactor designs under the Global Nuclear Energy Partnership (GNEP) program. It was investigated within the OECD/NEA Working Party on Reactor Systems (WPRS) under the Sodium-cooled Fast Reactor core Feed-back and Transient response (SFR-FT) task force benchmark, which was completed in 2014. The results revealed that different nuclear data libraries contribute to the large discrepancies in some calculated neutronic parameters. This task force is followed up by another on-going OECD/NEA WPRS activity entitled as SFR Uncertainty Analysis in Modeling (SFR-UAM). In order to further investigate the properties of the ABR core, the impact of nuclear data uncertainties on the performance of a SFR is analyzed in detail in this master thesis using a "Best Estimate Plus Uncertainty" (BEPU) approach along with the nuclear data from the ENDF/B-VII.1 library. Several computer codes, including MC2-3, TWODANT, DIF3D, REBUS-3, PERSENT, DPT, and SAS4A/SASSYS-1, were used in this study. Significant uncertainties on neutronic parameters (e.g., sodium density coefficient, sodium void coefficient, structure density coefficient, Doppler coefficient) are found due to nuclear data, but thanks to the excellent reactor design, the margins to sodium boiling and fuel melting during the accidents are still large even if these non-negligible nuclear data uncertainties are considered.

Download or read book Methodology for Uncertainty Analysis of Advanced Fuel Cycles and Preliminary Results written by and published by . This book was released on 2006 with total page 85 pages. Available in PDF, EPUB and Kindle. Book excerpt: This report assesses the sensitivity and uncertainty associated with certain advanced nuclear fuel cycles due to the variance of chosen parameters and how these results relate to the deep geological nuclear waste repository. High burn up uranium oxide, mixed oxide, and fast spectrum nuclear fuels are the advanced fuel cycles considered. The parameters that are varied in these cases are: the time of advanced fuel implementation, energy growth rate, fuel burn up, and reprocessing introduction and capacity. The results analyzed are the amount of spent fuel and the amount of Pu in spent fuel in the year 2099. The advanced fuel cycle scenarios are modeled using the DANESS code developed by Argonne National Laboratory. All the fuel cycles modeled in this report are highly sensitive to the above-mentioned varied parameters. In a 0% energy growth rate case the plutonium fast burner reactor significantly reduces the amount of waste destined to the repository. Compared to current once-through fuel cycle practices, the fast reactor reduces waste by 50-52 percent. As energy demand grows, the high burn up case of 100 (GWd per ton heavy metal) fuel, as modeled in this thesis, reduces the mass destined for the repository greatest. In the 1.5% energy growth rate, spent fuel mass is reduced 32-44 percent, and in the 3.0% energy growth rate those numbers are 43-49 percent.

Download or read book Efficient Uncertainty Quantification for a Fast Spectrum Generation IV Reactor System written by and published by . This book was released on 2004 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: This research is part of on-going research on the management of uncertainties in simulator predictions for Generation IV systems via optimum experimental design. The focus is on uncertainties originating due to input data. The objective is to devise an algorithmic framework for quantification of uncertainties, identification of their key sources, and ultimately guiding the design of validating experiments for their reduction. An integral part of this research is the development of uncertainty quantification algorithms for models involving many input data and output responses. This represents the focus of the research reported here. Uncertainty Quantification (UQ) in nuclear systems simulation is playing an increasing role in supporting decisions related to the research and development of advanced nuclear energy systems, especially those of interest to the Global Nuclear Energy Partnership (GNEP) and Next Generation Nuclear Plant (NGNP) programs. UQ will help assess the adequacy of existing simulation tools and associated databases, e.g. nuclear cross-section data, and provide guidance to areas of models and/or data where further development and/or measurements should be prioritized. A sensitivity and uncertainty analysis has been conducted to study the effects of neutron microscopic cross-section data uncertainty on macroscopic attributes that influence reactor core design, performance and safety for a Generation IV reactor concept. In the realm of reactor engineering, neutron cross-section data represents the basic physics of neutron interactions with matter and therefore have large impacts on evolution of flux, power, reactivity and other reactor performance attributes. Currently, we focus on uncertainties originating from cross-section data uncertainties, believed to be of primary significance for fast reactor calculations. This thesis presents a recent development of an UQ algorithm for increasing the efficiency of UQ to a level that enables its execution on a r.

Download or read book Sensitivity and Uncertainty Analysis of Reactor Performance Parameters written by Martin Becker and published by . This book was released on 1982 with total page 371 pages. Available in PDF, EPUB and Kindle. Book excerpt:

Download or read book Report on INL Activities for Uncertainty Reduction Analysis of FY12 written by and published by . This book was released on 2012 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: The work scope of this project related to the Work Packages of "Uncertainty Reduction Analyses" with the goal of reducing nuclear data uncertainties is to produce a set of improved nuclear data to be used both for a wide range of validated advanced fast reactor design calculations, and for providing guidelines for further improvements of the ENDF/B files (i.e. ENDF/B-VII, and future releases). Recent extensive sensitivity/uncertainty studies, performed within an international OECD-NEA initiative, have quantified for the first time the impact of current nuclear data uncertainties on design parameters of the major FCR & D and GEN-IV systems, and in particular on Na-cooled fast reactors with different fuels (oxide or metal), fuel composition (e.g. different Pu/TRU ratios) and different conversion ratios. These studies have pointed out that present uncertainties on the nuclear data should be significantly reduced, in order to get full benefit from the advanced modeling and simulation initiatives. Nuclear data plays a fundamental role in performance calculations of advanced reactor concepts. Uncertainties in the nuclear data propagate into uncertainties in calculated integral quantities, driving margins and costs in advanced system design, operation and safeguards. This package contributes to the resolution of technical, cost, safety, security and proliferation concerns in a multi-pronged, systematic, science-based R & D approach. The Nuclear Data effort identifies and develops small scale, phenomenon-specific experiments informed by theory and engineering to reduce the number of large, expensive integral experiments. The Nuclear Data activities are leveraged by effective collaborations between experiment and theory, between DOE programs and offices, at national laboratories and universities, both domestic and international. The primary objective is to develop reactor core sensitivity and uncertainty analyses that identify the improvement needs of key nuclear data which would facilitate fast spectrum system optimization and assure safety performance. The inclusion of fast spectrum integral experiment data is key to minimizing the impact of nuclear data uncertainties on reactor core performance calculations, thus providing the best nuclear data needs assessment. This report presents the status of activities performed at INL under the ARC Work Package previously mentioned. As major achievement this year a comprehensive adjustment, including 87 experiments, was carried out. The results of this adjustment provide useful insights and helpful feedback to both nuclear data evaluation and measurer communities. In the following, we will document first the theory that underlines the adjustment methodology, and then we will illustrate the sensitivity coefficient computation and the nuclear data and experiment selection. Subsequently, the adjustment results will be shown, and, finally, conclusions, including future work, will be provided.

Download or read book Methods and Computer Codes for Probabilistic Sensitivity and Uncertainty Analysis written by J. K. Vaurio and published by . This book was released on 1984 with total page 10 pages. Available in PDF, EPUB and Kindle. Book excerpt:

Download or read book Sensitivity Analyses of Fast Reactor Systems Including Thorium and Uranium written by and published by . This book was released on 1978 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: The Cross Section Evaluation Working Group (CSEWG) has, in conjunction with the development of the fifth version of ENDF/B, assembled new evaluations for 232Th and 233U. It is the purpose of this paper to describe briefly some of the more important features of these evaluations relative to ENDF/B-4 to project the change in reactor performance based upon the newer evaluated files and sensitivity coefficients for interesting design problems, and to indicate preliminary results from ongoing uncertainty analyses.

Download or read book Methods for Quantifying Uncertainty in Fast Reactor Analyses written by and published by . This book was released on 2008 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: Liquid-metal-cooled fast reactors in the form of sodium-cooled fast reactors have been successfully built and tested in the U.S. and throughout the world. However, no fast reactor has operated in the U.S. for nearly fourteen years. More importantly, the U.S. has not constructed a fast reactor in nearly 30 years. In addition to reestablishing the necessary industrial infrastructure, the development, testing, and licensing of a new, advanced fast reactor concept will likely require a significant base technology program that will rely more heavily on modeling and simulation than has been done in the past. The ability to quantify uncertainty in modeling and simulations will be an important part of any experimental program and can provide added confidence that established design limits and safety margins are appropriate. In addition, there is an increasing demand from the nuclear industry for best-estimate analysis methods to provide confidence bounds along with their results. The ability to quantify uncertainty will be an important component of modeling that is used to support design, testing, and experimental programs. Three avenues of UQ investigation are proposed. Two relatively new approaches are described which can be directly coupled to simulation codes currently being developed under the Advanced Simulation and Modeling program within the Reactor Campaign. A third approach, based on robust Monte Carlo methods, can be used in conjunction with existing reactor analysis codes as a means of verification and validation of the more detailed approaches.

Download or read book Sensitivity and Uncertainty Analysis of Multiphysics Nuclear Reactor Core Depletion written by Andrew Scott Bielen and published by . This book was released on 2015 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt:

Download or read book Statistically Based Uncertainty Analysis for Ranking of Component Importance in the Thermal hydraulic Safety Analysis of the Advanced Neutron Source Reactor written by and published by . This book was released on 1992 with total page 61 pages. Available in PDF, EPUB and Kindle. Book excerpt: The Analytic Hierarchy Process (AHP) has been used to help determine the importance of components and phenomena in thermal-hydraulic safety analyses of nuclear reactors. The AHP results are based, in part on expert opinion. Therefore, it is prudent to evaluate the uncertainty of the AHP ranks of importance. Prior applications have addressed uncertainty with experimental data comparisons and bounding sensitivity calculations. These methods work well when a sufficient experimental data base exists to justify the comparisons. However, in the case of limited or no experimental data the size of the uncertainty is normally made conservatively large. Accordingly, the author has taken another approach, that of performing a statistically based uncertainty analysis. The new work is based on prior evaluations of the importance of components and phenomena in the thermal-hydraulic safety analysis of the Advanced Neutron Source Reactor (ANSR), a new facility now in the design phase. The uncertainty during large break loss of coolant, and decay heat removal scenarios is estimated by assigning a probability distribution function (pdf) to the potential error in the initial expert estimates of pair-wise importance between the components. Using a Monte Carlo sampling technique, the error pdfs are propagated through the AHP software solutions to determine a pdf of uncertainty in the system wide importance of each component. To enhance the generality of the results, study of one other problem having different number of elements is reported, as are the effects of a larger assumed pdf error in the expert ranks. Validation of the Monte Carlo sample size and repeatability are also documented.

Download or read book Application of FORSS Sensitivity and Uncertainty Methodology to Fast Reactor Benchmark Analysis written by and published by . This book was released on 1976 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: FORSS is a code system used to study relationships between nuclear reaction cross sections, integral experiments, reactor performance parameter predictions, and associated uncertainties. This paper presents the theory and code description as well as the first results of applying FORSS to fast reactor benchmarks. Specifically, for various assemblies and reactor performance parameters, the nuclear data sensitivities were computed by nuclide, reaction type, and energy. Comprehensive libraries of energy-dependent coefficients have been developed in a computer retrievable format and released for distribution by RSIC and NNCSC. Uncertainties induced by nuclear data were quantified using preliminary, energy-dependent relative covariance matrices evaluated with ENDF/B-IV expectation values and processed for 238U(n, f), 238U(n, .gamma.), 239Pu(n, f), and 239Pu(.nu.). Nuclear data accuracy requirements to meet specified performance criteria at minimum experimental cost were determined.

Download or read book Sensitivity and Uncertainty Analysis of the High Conversion Reactor Concept written by Thomas F. DeLorey and published by . This book was released on 1989 with total page 264 pages. Available in PDF, EPUB and Kindle. Book excerpt:

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 Application of the Sensitivity and Uncertainty Analysis System LASS to Fusion Reactor Nucleonics written by and published by . This book was released on 1976 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: Sensitivity analysis, as applied to both nuclear design and data uncertainty, has developed into a valuable tool for fusion reactor nuclear analysis. Several such studies have been undertaken with the LASL sensitivity system LASS, which includes as its principal modules SENSIT-1D, ONETRAN, and ALVIN. These modules function in a multigroup environment using standard flux and data interface files for communication. The input multigroup cross-section data and uncertainties are obtained primarily from ENDF/B using the NJOY processing system. In particular cases, the input library can be modified by the ALVIN module to improve consistency with available integral experiments. The primary output from LASS is the uncertainty (or change) in important reactor parameters, as calculated in the SENSIT-1D module. Applications of LASS and its component parts have been made to the Tokamak Fusion Test Reactor (TFTR), the Reference Theta-Pinch Reactor (RTPR), and to an Experimental Power Reactor (EPR). This paper emphasizes the initial assessment of cross-section sensitivity for an EPR design. Nucleonic responses examined include neutron and gamma-ray kerma in the toroidal field coils and Mylar superinsulation, displacement damage and transmutation in the copper of the toroidal field coils, and activation of the outboard dewar. These sensitivities are now being used to narrow the range of uncertainty analyses required to quantitatively assess cross-section adequacy for EPR design calculations. Acceptable target uncertainties in nucleonic design parameters are simultaneously being formulated. Experience at LASL with sensitivity and uncertainty analysis techniques incorporated in LASS has provided convincing evidence of their value for fusion reactor studies. Many of these studies are of a shielding nature; e.g., deep penetrations of high-energy neutrons through steel, lead, boron carbide, and graphite, with responses such as activation and kerma.