Sensitivity and Uncertainty Analysis of Multiphysics Nuclear Reactor Core Depletion PDF Download
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Author: Andrew Scott Bielen
Publisher:
ISBN:
Category : Burnup
Languages : en
Pages :
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Author: Andrew Scott Bielen
Publisher:
ISBN:
Category : Burnup
Languages : en
Pages :
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ISBN: 9789462950719
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Languages : en
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ISBN:
Category :
Languages : en
Pages : 68
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The continued development of High Temperature Gas Cooled Reactors (HTGRs) requires verification of HTGR design and safety features with reliable high fidelity physics models and robust, efficient, and accurate codes. The predictive capability of coupled neutronics/thermal-hydraulics and depletion simulations for reactor design and safety analysis can be assessed with sensitivity analysis (SA) and uncertainty analysis (UA) methods. Uncertainty originates from errors in physical data, manufacturing uncertainties, modelling and computational algorithms. (The interested reader is referred to the large body of published SA and UA literature for a more complete overview of the various types of uncertainties, methodologies and results obtained). SA is helpful for ranking the various sources of uncertainty and error in the results of core analyses. SA and UA are required to address cost, safety, and licensing needs and should be applied to all aspects of reactor multi-physics simulation. SA and UA can guide experimental, modelling, and algorithm research and development. Current SA and UA rely either on derivative-based methods such as stochastic sampling methods or on generalized perturbation theory to obtain sensitivity coefficients. Neither approach addresses all needs. In order to benefit from recent advances in modelling and simulation and the availability of new covariance data (nuclear data uncertainties) extensive sensitivity and uncertainty studies are needed for quantification of the impact of different sources of uncertainties on the design and safety parameters of HTGRs. Only a parallel effort in advanced simulation and in nuclear data improvement will be able to provide designers with more robust and well validated calculation tools to meet design target accuracies. In February 2009, the Technical Working Group on Gas-Cooled Reactors (TWG-GCR) of the International Atomic Energy Agency (IAEA) recommended that the proposed Coordinated Research Program (CRP) on the HTGR Uncertainty Analysis in Modelling (UAM) be implemented. This CRP is a continuation of the previous IAEA and Organization for Economic Co-operation and Development (OECD)/Nuclear Energy Agency (NEA) international activities on Verification and Validation (V & V) of available analytical capabilities for HTGR simulation for design and safety evaluations [1], [2], [3]. Within the framework of these activities different numerical and experimental benchmark problems were performed and insight was gained about specific physics phenomena and the adequacy of analysis methods.
Author: Hayes Franklin Stripling
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Category :
Languages : en
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Depletion calculations for nuclear reactors model the dynamic coupling between the material composition and neutron flux and help predict reactor performance and safety characteristics. In order to be trusted as reliable predictive tools and inputs to licensing and operational decisions, the simulations must include an accurate and holistic quantification of errors and uncertainties in its outputs. Uncertainty quantification is a formidable challenge in large, realistic reactor models because of the large number of unknowns and myriad sources of uncertainty and error. We present a framework for performing efficient uncertainty quantification in depletion problems using an adjoint approach, with emphasis on high-fidelity calculations using advanced massively parallel computing architectures. This approach calls for a solution to two systems of equations: (a) the forward, engineering system that models the reactor, and (b) the adjoint system, which is mathematically related to but different from the forward system. We use the solutions of these systems to produce sensitivity and error estimates at a cost that does not grow rapidly with the number of uncertain inputs. We present the framework in a general fashion and apply it to both the source-driven and k-eigenvalue forms of the depletion equations. We describe the implementation and verification of solvers for the forward and ad- joint equations in the PDT code, and we test the algorithms on realistic reactor analysis problems. We demonstrate a new approach for reducing the memory and I/O demands on the host machine, which can be overwhelming for typical adjoint algorithms. Our conclusion is that adjoint depletion calculations using full transport solutions are not only computationally tractable, they are the most attractive option for performing uncertainty quantification on high-fidelity reactor analysis problems. The electronic version of this dissertation is accessible from http://hdl.handle.net/1969.1/151312
Author: Nicholas Wozniak
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Category :
Languages : en
Pages : 0
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Author: A. Hogenbirk
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Category :
Languages : en
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Author: Martin Becker
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Category : Nuclear physics
Languages : en
Pages : 371
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Category :
Languages : en
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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: Thomas F. DeLorey
Publisher:
ISBN:
Category : Nuclear reactors
Languages : en
Pages : 264
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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.
