Download or read book Overview of Nuclear Data Uncertainty in Scale and Application to Light Water Reactor Uncertainty Analysis written by W. Wieselquist and published by . This book was released on 2018 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt:
Download or read book Application of Perturbation Theory Methods to Nuclear Data Uncertainty Propagation Using the Collision Probability Method written by Pouya Sabouri and published by . This book was released on 2013 with total page 0 pages. Available in PDF, EPUB and Kindle. Book excerpt: This dissertation presents a comprehensive study of sensitivity/uncertainty analysis for reactor performance parameters (e.g. the k-effective) to the base nuclear data from which they are computed. The analysis starts at the fundamental step, the Evaluated Nuclear Data File and the uncertainties inherently associated with the data they contain, available in the form of variance/covariance matrices. We show that when a methodical and consistent computation of sensitivity is performed, conventional deterministic formalisms can be sufficient to propagate nuclear data uncertainties with the level of accuracy obtained by the most advanced tools, such as state-of-the-art Monte Carlo codes. By applying our developed methodology to three exercises proposed by the OECD (UACSA Benchmarks), we provide insights of the underlying physical phenomena associated with the used formalisms.
Download or read book Applying Uncertainty and Sensitivity on Thermal Hydraulic Subchannel Analysis for the Multi application Small Light Water Reactor written by Adam Brigantic and published by . This book was released on 2014 with total page 115 pages. Available in PDF, EPUB and Kindle. Book excerpt: Small modular reactors (SMRs) are a recent advancement in commercial nuclear reactor design with growing interest worldwide. New SMR concepts, such as the Multi-Application Small Light Water Reactor (MASLWR), must undergo a licensing processes established by the U.S. Nuclear Regulatory Commission (NRC) prior to commercial operation. Given the lack of historical, full scale operating experience, a general uncertainty and sensitivity analysis methodology was developed to help aid SMR designs through this process. Uncertainty was quantified through the empirical cumulative distribution function (ECDF) created from a desired data set. Linear regression techniques were applied to measure sensitivity. This methodology was demonstrated through the thermal hydraulic subchannel analysis of the MASLWR concept using RELAP5-3D Version 4.0.3 and VIPRE-01 Mod 2.2.1. Twelve uncertain input parameters were selected. System response uncertainty in the minimum departure from nucleate boiling ratio (MDNBR), maximum fuel temperature, and maximum clad temperature was evaluated. These figures were shown to satisfy U.S. NRC regulatory requirements for steady state operation at the 95 percent probability and 95 percent confidence level under the evaluated conditions. Sensitivity studies showed input parameters affecting local power generation within the core had a large influence on MDNBR, maximum fuel temperature, and maximum clad temperature.
Download or read book NUREG CR written by U.S. Nuclear Regulatory Commission and published by . This book was released on 1977 with total page 144 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Download or read book Development of an Uncertainty Methodology for Nuclear Applications written by Fatih Aydogan and published by LAP Lambert Academic Publishing. This book was released on 2011-10 with total page 372 pages. Available in PDF, EPUB and Kindle. Book excerpt: A new uncertainty methodology for not only nuclear engineering but also other disciplines is developed. This uncertainty methodology is analyzed by OECD/US-NRC Full Size Fine Mesh Bundle Tests (BFBT). Void distribution and critical power data for Boiling Water Reactor (BWR)bundles are utilized for the sensitivity and uncertainty analyses. Cumulative Distribution Functions for commonly used in two phase flow thermal hydraulic problems are developed and used as the inputs for the methodology. In addition Spanning Algorithm selects the limiting experimental cases to bound the experimental database. Finally, uncertainty methodology's feedback mechanism and flexibility provides a powerful tool for analyzers.
Download or read book The New Nuclear Data Sensitivity Analysis and Uncertainty Propagation Tool in NESTLE written by William A. Wieselquist and published by . This book was released on 2004 with total page 117 pages. Available in PDF, EPUB and Kindle. Book excerpt: Keywords: sensitivity analysis, nuclear engineering, nuclear data, NESTLE, reactor physics, uncertainty propagation.
Download or read book Quantification and Uncertainty Analysis of Source Terms for Severe Accidents in Light Water Reactors Quasar written by and published by . This book was released on 1986 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt:
Download or read book The New Nuclear Data Sensitivity Analysis and Uncertainty Propagation Tool in NESTLE written by and published by . This book was released on 2004 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: In support of the need for better design and evaluation tools for reactor-based transmutation systems we have upgraded NESTLE, the 2/4 energy group thermal reactor physics code of the Nuclear Engineering Department at North Carolina State University with: i) the ability to perform nuclide transmutation calculations for a general, user-defined field of nuclei and transmutation paths and ii) the ability to analyze sensitivities and propagate uncertainties in the end-of-cycle (EOC) nuclide inventory with respect to nuclear data and beginning-of-cycle (BOC) nuclide inventory. We present two methods of sensitivity analysis: i) direct perturbation and recalculation (DPAR) and ii) sensitivity analysis utilizing an adjoint system (AS). With DPAR, we simply perturb data and recalculate solutions of our system and thus may analyze sensitivity of all responses to perturbations in one data parameter per solution of the perturbed forward problem. With the AS, we form a system of equations, the solution of which may be used to estimate the first variation of a response with respect to any data parameters. For the AS, we have developed the equations for both the predictor and predictor-corrector neutron/nuclide field coupling methods in NESTLE. To our knowledge, the AS for the predictor-corrector coupling has never been presented. Then we used the tools we have developed to evaluate the sensitivity of EOC nuclide concentrations and SNF hazard measures with respect to nuclear data for a cycle 1 pressurized water reactor (PWR) core. In our study, we found that the nuclear data crucial to modeling US reactors' once-through cycle (fission cross sections of 235U and 239Pu, the main fuel nuclei, and capture cross sections for 238U) also has the highest impact on EOC nuclide inventory of so-called "problem nuclei" (e.g. Am, Cm, etc.) Note that these results only apply to cycle 1, in which fresh fuel is irradiated for the first time. Because.
Download or read book Uncertainty Analysis Framework for the Multi Physics Light Water Reactor Simulation written by Kaiyue Zeng and published by . This book was released on 2020 with total page 84 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Download or read book Handbook of Uncertainty Quantification written by Roger Ghanem and published by . This book was released on 2017 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt:
Download or read book Thermal Safety Margins in Nuclear Reactors written by Henryk Anglart and published by CRC Press. This book was released on 2024-06-11 with total page 387 pages. Available in PDF, EPUB and Kindle. Book excerpt: This book presents an overview of state-of-the art approaches to determine thermal safety margins in nuclear reactors. It presents both the deterministic and probabilistic aspects of thermal safety margins of nuclear reactors to facilitate the understanding of these two difficult topics at various academic levels, from undergraduates to researchers in nuclear engineering. It first sets out the theoretical background before exploring how to determine thermal safety margins in nuclear reactors, through examples, problems and advanced state-of-the-art approaches. This will help undergraduate students better understand the most fundamental aspects of nuclear reactor safety. For researchers and practitioners, this book provides a comprehensive overview of most recent achievements in the field, offering an excellent starting point to develop new methods for the assessment of the thermal safety margins. This book is written to bridge the gap between deterministic and appropriate treatment of uncertainties to assess safety margins in nuclear reactors, presenting these approaches as complementary to each other. Even though these two approaches are frequently used in parallel in real-world applications, there has been a lack of a consistent teaching approach in this area. This book is suitable for readers with a background in calculus, thermodynamics, fluid mechanics, and heat transfer. It is assumed that readers have previous exposure to such concepts as laws of thermodynamics, enthalpy, entropy, and conservation equations used in fluid mechanics and heat transfer. Key Features: Covers the theory, principles, and assessment methods of thermal safety margins in nuclear reactors whilst presenting the state-of-the-art technology in the field Combines the deterministic thermal safety considerations with a comprehensive treatment of uncertainties, offering a framework that is applicable to all current and future commercial nuclear reactor types Provides numerous examples and problems to be solved
Download or read book Liquid fueled Molten Salt Reactor Modeling and Uncertainty Analysis for Safeguards Purposes written by Andre Vidal Soares and published by . This book was released on 2022 with total page 0 pages. Available in PDF, EPUB and Kindle. Book excerpt: With the recent interest in advanced nuclear reactors, the need for developing nuclear material accounting strategies arise. Especially in liquid-fueled molten salt reactors, which preclude the ability to count fixed amount of material in a discrete way, as opposed to the traditional counting method widely used in the current nuclear fleet. Usually, these reactor concepts involve online fuel processing where fuel and its products flow within many streams along the plant. In computational modeling, this feature requires high fidelity depletion schemes, including the addition of feed and removal capabilities that can simulate their regular operation. Estimating isotopic inventory under regular and diversion scenarios would provide an insight on which reactor parameters could be tracked to timely detect diversion of special nuclear material. To achieve this, a depletion model was developed for the Molten Salt Demonstration Reactor (with minor adaptations) in Serpent 2 and SCALE. Feed and removal capabilities were implemented in the Serpent latest release and is under testing in SCALE 6.3 beta versions. Plutonium diversion scenarios are modeled and their impact on regular operation of the reactor is analyzed. The SCALE module Sampler is used to estimate nuclear data uncertainties propagated along the depletion interval. Several isotopes presented changes in their concentrations given a 10SQ plutonium diversion protracted scenario. Some examples are fission products such as 89Sr (1.06% change and 0.58% uncertainty), 91Y (1.92% change and 0.43% uncertainty), 113Cd (-5.00% change and 3.31% uncertainty), 151Eu (-3.84% change and 2.70% uncertainty) and also actinides like 241Am (-6.88% change and 2.23% uncertainty), 242mAm (-6.22% change and 4.88% uncertainty) and 242Cm (-4.50% change and 3.84% uncertainty). Preliminary sensitivity analyses were performed using TSUNAMI and Sampler. Results revealed that 238U(n,[gamma]) plays an important role in contributing to the uncertainty in parameters (e.g., k-inf) and nuclide concentrations (e.g., 239Pu and 241Pu). Future work includes the full analyses of scenarios - including abrupt ones, which are already modeled, and a comprehensive sensitivity study on nuclide concentrations - including fission products and higher actinides. Moreover, methods to improve nuclear covariance data - guided by the sensitivity study - will also be pursued using Bayesian methods and machine learning techniques.
Download or read book Quasar Uncertainty Study written by and published by . This book was released on 1986 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: Over the last decade, substantial development and progress has been made in the understanding of the nature of severe accidents and associated fission product release and transport. As part of this continuing effort, the United States Nuclear Regulatory Commission (USNRC) sponsored the development of the Source Term Code Package (STCP), which models core degradation, fission product release from the damaged fuel, and the subsequent migration of the fission products from the primary system to the containment and finally to the environment. The objectives of the QUASAR (Quantification and Uncertainty Analysis of Source Terms for Severe Accidents in Light Water Reactors) program are: (1) to address the uncertainties associated with input parameters and phenomenological models used in the STCP; and (2) to define reasonable and technically defensible parameter ranges and modelling assumptions for the use in the STCP. The uncertainties in the radiological releases to the environment can be defined as the degree of current knowledge associated with the magnitude, the timing, duration, and other pertinent characteristics of the release following a severe nuclear reactor accident. These uncertainties can be quantified by probability density functions (PDF) using the Source Term Code Package as the physical model. An attempt will also be made to address the phenomenological issues not adequately modeled by the STCP, using more advanced, mechanistic models.
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 A Proposed Approach to Uncertainty Analysis written by and published by . This book was released on 1983 with total page 84 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Download or read book Quantifying Uncertainty in Nuclear Analytical Measurements written by and published by . This book was released on 2004 with total page 268 pages. Available in PDF, EPUB and Kindle. Book excerpt: Dedicated specifically to nuclear analytical techniques, this publication is intended to assist scientists using alpha, beta and gamma spectrometries, neutron activation and XRF analyses, and other nuclear analytical methods, in assessing and quantifying the sources of uncertainty in their daily measurements.
Download or read book Uncertainty Quantification for Safety Verification Applications in Nuclear Power Plants written by Emmanuel Boafo and published by . This book was released on 2016 with total page 0 pages. Available in PDF, EPUB and Kindle. Book excerpt: There is an increasing interest in computational reactor safety analysis to systematically replace the conservative calculations by best estimate calculations augmented by quantitative uncertainty analysis methods. This has been necessitated by recent regulatory requirements that have permitted the use of such methods in reactor safety analysis. Stochastic uncertainty quantification methods have shown great promise, as they are better suited to capture the complexities in real engineering problems. This study proposes a framework for performing uncertainty quantification based on the stochastic approach, which can be applied to enhance safety analysis. Additionally, risk level has increased with the degradation of Nuclear Power Plant (NPP) equipment and instrumentation. In order to achieve NPP safety, it is important to continuously evaluate risk for all potential hazards and fault propagation scenarios and map protection layers to fault / failure / hazard propagation scenarios to be able to evaluate and verify safety level during NPP operation. In this study, the Fault Semantic Network (FSN) methodology is proposed. This involved the development of static and dynamic fault semantic network (FSN) to model possible fault propagation scenarios and the interrelationships among associated process variables. The proposed method was demonstrated by its application to two selected case studies. The use of FSN is essential for fault detection, understanding fault propagation scenarios and to aid in the prevention of catastrophic events. Two transient scenarios were simulated with a best estimate thermal hydraulic code, CATHENA. Stochastic uncertainty quantification and sensitivity analyses were performed using the OPENCOSSAN software which is based on the Monte Carlo method. The effect of uncertainty in input parameters were investigated by analyzing the probability distribution of output parameters. The first four moments (mean, variance, skewness and kurtosis) of the output parameters were computed and analyzed. The uncertainty in output pressure was 0.61% and 0.57% was found for the mass flow rate in the Edward's blowdown transient. An uncertainty of 0.087% was obtained for output pressure and 0.048% for fuel pin temperature in the RD-14 test case. These results are expected to be useful for providing insight into safety margins related to safety analysis and verification.
