Fully Automated 3-D Parallel Simulation and Optimization of a Full Scale Pressurized Water Reactor Fuel Assembly with Burnup Corrected Cross Sections PDF Download
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Author: Thomas Joseph Plower Publisher: ISBN: Category : Languages : en Pages :
Book Description
ABSTRACT: Computational nuclear fuel burnup analysis is an essential field within the Nuclear Engineering discipline, since it plays important functions in core reactivity management, criticality safety, Special Nuclear Materials management, and fuel assembly reload design of commercial power and research reactors. Three dimensional (3-D) deterministic transport methods provides unique advantages in the fuel burnup analysis field and the intention of this thesis is to demonstrate the author's contributions to the development of a novel 3-D deterministic fuel burnup package called the PENTRAN /PENBURN (Parallel Environment Neutral particle Transport/Parallel Environment Burnup) suite. Specifically, cross section generation procedures will be presented including discussions on development of a coupled cross section interpolator code called INTERP-XS. Additionally, detailed fuel burnup analysis of a 17x17 PWR assembly will be presented. Finally, the development of an automated sequence driver called BURNDRIVER will be shown. Major conclusions include: excellent agreement between INTERP-XS generated cross sections and those generated by SCALE, demonstration of 3-D burnup effects captured by PENTRAN/PENBURN through PWR assembly analysis, and successful creation of a user-friendly burnup sequence driver.
Author: Thomas Joseph Plower Publisher: ISBN: Category : Languages : en Pages :
Book Description
ABSTRACT: Computational nuclear fuel burnup analysis is an essential field within the Nuclear Engineering discipline, since it plays important functions in core reactivity management, criticality safety, Special Nuclear Materials management, and fuel assembly reload design of commercial power and research reactors. Three dimensional (3-D) deterministic transport methods provides unique advantages in the fuel burnup analysis field and the intention of this thesis is to demonstrate the author's contributions to the development of a novel 3-D deterministic fuel burnup package called the PENTRAN /PENBURN (Parallel Environment Neutral particle Transport/Parallel Environment Burnup) suite. Specifically, cross section generation procedures will be presented including discussions on development of a coupled cross section interpolator code called INTERP-XS. Additionally, detailed fuel burnup analysis of a 17x17 PWR assembly will be presented. Finally, the development of an automated sequence driver called BURNDRIVER will be shown. Major conclusions include: excellent agreement between INTERP-XS generated cross sections and those generated by SCALE, demonstration of 3-D burnup effects captured by PENTRAN/PENBURN through PWR assembly analysis, and successful creation of a user-friendly burnup sequence driver.
Author: Publisher: ISBN: Category : Languages : en Pages :
Book Description
Computational capability has been developed to adjust multi-group neutron cross-sections to improve the fidelity of boiling water reactor (BWR) modeling and simulation. The method involves propagating multi-group neutron cross-section uncertainties through BWR computational models to evaluate uncertainties in key core attributes such as core k-effective, nodal power distributions, thermal margins, and in-core detector readings. Uncertainty-based inverse theory methods are then employed to adjust multi-group cross-sections to minimize the disagreement between BWR modeling predictions and measured plant data. For this work, measured plant data were virtually simulated in the form of perturbed 3-D nodal power distributions with discrepancies with predictions of the same order of magnitude as expected from plant data. Using the simulated plant data, multi-group cross-section adjustment reduces the error in core k-effective to less than 0.2% and the RMS error in nodal power to 4% (i.e. -- the noise level of the in-core instrumentation). To ensure that the adapted BWR model predictions are robust, Tikhonov regularization is utilized to control the magnitude of the cross-section adjustment. In contrast to few-group cross-section adjustment, which was the focus of previous research on BWR adaptive simulation, multi-group cross-section adjustment allows for future fuel cycle design optimization to include the determination of optimal fresh fuel assembly designs using the adjusted multi-group cross-sections. The major focus of this work is to efficiently propagate multi-group neutron cross-section uncertainty through BWR lattice physics calculations. Basic neutron cross-section uncertainties are provided in the form of multi-group cross-section covariance matrices. For energy groups in the resolved resonance energy range, the cross-section uncertainties are computed using an infinitely-dilute approximation of the neutron flux. In order to accurately account for spatial a.
Author: Publisher: ISBN: Category : Languages : en Pages :
Book Description
Adaptive simulation (AS) is an algorithm utilizing a regularized least squares methodology to correct for the discrepancy between core simulators predictions and actual plant measurements. This is an inverse problem that will adjust the cross sections input to a core simulator within their range of uncertainty to obtain better agreement with the plant measurements. The cross section adjustments are constrained to their range of uncertainty using the covariance matrix of the few-group cross sections and in imposing the regularization on the least squares solution. This few-group covariance matrix is obtained using the covariance matrix of the multi-group cross sections and the corresponding lattice physics sensitivity matrix. To perform the adaption, one must also have the sensitivity matrix of the core simulator. Constructing the sensitivity matrix of both the lattice physics code and core simulator would be a daunting task using the traditional brute-force method of computing a forward solve for a perturbation of every input. To avoid this, a singular value decomposition (SVD) is used to construct a low rank approximation of the covariance matrices, thus drastically reducing the number of required forward solves. Until now, AS has been used on a single depletion cycle to correct for discrepancies resulting from errors introduced by incorrect cross sections only. Adapting to a single depletion cycle means that the cross sections of cycle m were adjusted so that the core simulator better predicts the actual measurements of cycle m (and future cycles if the algorithm is robust). This, however, does not account for the reloaded burnt fuel number density errors at the beginning-of-cycle (BOC) m. By definition a burnt assembly has been used and depleted in a previous cycle. If adaption changes the cross sections of that burnt assembly in cycle m, those cross sections should have also been changed in any cycle preceding m which would have resulted in different BOC m numbe.
Author: Thomas Joseph Plower
Author: Thomas Joseph Plower
Author: U.S. Atomic Energy Commission. Division of Reactor Development
Author: Barry N. Naft
Author: John Gittus
Author: L. E. Strawbridge
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Author: Stone and Webster Engineering Corporation