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Author: William A. Wieselquist
Publisher:
ISBN:
Category :
Languages : en
Pages : 117
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Keywords: sensitivity analysis, nuclear engineering, nuclear data, NESTLE, reactor physics, uncertainty propagation.
Author: William A. Wieselquist
Publisher:
ISBN:
Category :
Languages : en
Pages : 117
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Book Description
Keywords: sensitivity analysis, nuclear engineering, nuclear data, NESTLE, reactor physics, uncertainty propagation.
Author:
Publisher:
ISBN:
Category :
Languages : en
Pages :
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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.
Author: American Nuclear Society
Publisher:
ISBN:
Category : Nuclear engineering
Languages : en
Pages : 850
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Author: Pouya Sabouri
Publisher:
ISBN:
Category :
Languages : en
Pages : 0
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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.
Author:
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Category :
Languages : en
Pages :
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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.
Author:
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Category :
Languages : en
Pages :
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Verification and validation (V & V) are playing more important roles to quantify uncertainties and realize high fidelity simulations in engineering system analyses, such as transients happened in a complex nuclear reactor system. Traditional V & V in the reactor system analysis focused more on the validation part or did not differentiate verification and validation. The traditional approach to uncertainty quantification is based on a 'black box' approach. The simulation tool is treated as an unknown signal generator, a distribution of inputs according to assumed probability density functions is sent in and the distribution of the outputs is measured and correlated back to the original input distribution. The 'black box' method mixes numerical errors with all other uncertainties. It is also not efficient to perform sensitivity analysis. Contrary to the 'black box' method, a more efficient sensitivity approach can take advantage of intimate knowledge of the simulation code. In these types of approaches equations for the propagation of uncertainty are constructed and the sensitivities are directly solved for as variables in the simulation. This paper presents the forward sensitivity analysis as a method to help uncertainty qualification. By including time step and potentially spatial step as special sensitivity parameters, the forward sensitivity method is extended as one method to quantify numerical errors. Note that by integrating local truncation errors over the whole system through the forward sensitivity analysis process, the generated time step and spatial step sensitivity information reflect global numerical errors. The discretization errors can be systematically compared against uncertainties due to other physical parameters. This extension makes the forward sensitivity method a much more powerful tool to help uncertainty qualification. By knowing the relative sensitivity of time and space steps with other interested physical parameters, the simulation is allowed to run at optimized time and space steps without affecting the confidence of the physical parameter sensitivity results. The time and space steps forward sensitivity analysis method can also replace the traditional time step and grid convergence study with much less computational cost. Several well defined benchmark problems with manufactured solutions are utilized to demonstrate the extended forward sensitivity analysis method. All the physical solutions, parameter sensitivity solutions, even the time step sensitivity in one case, have analytical forms, which allows the verification to be performed in the strictest sense.
Author:
Publisher:
ISBN:
Category :
Languages : en
Pages : 8
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Sensitivity and uncertainty methods have been developed to aid in the establishment of areas of applicability and validation of computer codes and nuclear data for nuclear criticality safety studies. A key component in this work is the generation of sensitivity and uncertainty parameters for typically several hundred benchmarks experiments used in validation exercises. Previously, only one-dimensional sensitivity tools were available for this task, which necessitated the remodeling of multidimensional inputs in order for such an analysis to be performed. This paper describes the development of the SEN3 Monte Carlo based sensitivity analysis sequence for SCALE. Two options in the SEN3 package for the reconstruction of angular-dependent forward and adjoint fluxes are described and contrasted. These options are the direct calculation of flux moments versus the calculation of angular fluxes, with subsequent conversion to flux moments prior to sensitivity coefficient generation. The latter technique is found to be significantly more efficient.
Author:
Publisher:
ISBN:
Category : Business & Economics
Languages : en
Pages : 216
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Deterministic safety analysis is an important tool for confirming the adequacy and efficiency of provisions within the defence in depth concept for the safety of nuclear power plants (NPPs). IAEA Safety Standards Series No. NS-R-1.2 and Safety Reports Series No. 23 recommend, as one of the options for demonstrating the inclusion of adequate safety margins, the use of best estimate computer codes with realistic input data in combination with the evaluation of uncertainties in the calculation results. The evaluation of uncertainties is an issue of considerable complexity, and this Safety Report has been developed to complement the existing publications. It provides more detailed information on the methods available for the evaluation of uncertainties in deterministic safety analysis of NPPs and practical guidance in the use of these methods.
Author: Britta Folmer
Publisher: Academic Press
ISBN: 0128035587
Category : Technology & Engineering
Languages : en
Pages : 558
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Book Description
The Craft and Science of Coffee follows the coffee plant from its origins in East Africa to its current role as a global product that influences millions of lives though sustainable development, economics, and consumer desire.For most, coffee is a beloved beverage. However, for some it is also an object of scientifically study, and for others it is approached as a craft, both building on skills and experience. By combining the research and insights of the scientific community and expertise of the crafts people, this unique book brings readers into a sustained and inclusive conversation, one where academic and industrial thought leaders, coffee farmers, and baristas are quoted, each informing and enriching each other.This unusual approach guides the reader on a journey from coffee farmer to roaster, market analyst to barista, in a style that is both rigorous and experience based, universally relevant and personally engaging. From on-farming processes to consumer benefits, the reader is given a deeper appreciation and understanding of coffee's complexity and is invited to form their own educated opinions on the ever changing situation, including potential routes to further shape the coffee future in a responsible manner. - Presents a novel synthesis of coffee research and real-world experience that aids understanding, appreciation, and potential action - Includes contributions from a multitude of experts who address complex subjects with a conversational approach - Provides expert discourse on the coffee calue chain, from agricultural and production practices, sustainability, post-harvest processing, and quality aspects to the economic analysis of the consumer value proposition - Engages with the key challenges of future coffee production and potential solutions
Author: U.S. Nuclear Regulatory Commission
Publisher:
ISBN:
Category : Nuclear energy
Languages : en
Pages : 144
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