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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: 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: Jack Douglas Galloway Publisher: ISBN: Category : Languages : en Pages : 174
Book Description
The computational ability to accurately simulate boiling water reactor operation under the full range of standard steady-state operation, along with the capability to fully track the isotopic distribution of any fueled region in any location in the core has been developed. This new three-dimensional node-by-node capability can help designers track, for example, a full suite of minor and major actinides, fission products, and even light elements that result from depletion, decay, or transmutations. This isotopic tracking capability is not restricted to BWRs and can be employed in the modeling of PWRs, CANDUs, and other reactor types that can be modeled with the NESTLE code, the base core simulator employed in this research. To accurately simulate boiling water reactor operation, a major thermal-hydraulics upgrade was performed which involved the implementation of a drift-flux solution scheme to model steady-state boiling water flow. Sub-cooled boiling and bulk boiling are accurately modeled and a scheme for computing the correct flow distribution has been implemented. In addition, the incorporation of a nodal ORIGEN-based microscopic depletion solution has been included which allows for exceptional detail in tracking a large number of elements in every node of a core design, thus accounting for spectral dependencies such as moderator density effects, moderator temperature effects, fuel temperature effects, as well as controlled or uncontrolled conditions. The results of this study show the excellent fidelity of the two-phase solution for accurately predicting the boiling of water when compared to experimental results. Likewise, the isotopic inventory results show near-identical agreement with the well-established and validated ORIGEN-based SCALE/TRITON isotopic depletion sequence. The aim of these developments is to eventually produce a publicly available three-dimensional core simulator capable of assessing detailed isotopic inventories, a capability particularly valuable for the evaluation of recycling scenarios and actinide management in a variety of reactor types and fuel designs.
Author: Robert Allen Walls Publisher: ISBN: Category : Languages : en Pages : 95
Book Description
Reactor plant design often incorporates data and insight ascertained from computer code simulations of plant dynamics and reactor core behavior. Increasing utilization of data gathered from simulations aids operators and designers in planning for overall plant operation and most importantly, safety. The United States (US) Nuclear Regulatory Commission (NRC) in researching reactor plant safety utilizes several computer codes, or models. The two codes used for work on this thesis are TRACE and PARCS. TRACE (TRAC RELAP5 Advanced Computational Engine) is a thermal-hydraulic code that models the coolant system under numerous variables in operating conditions. Coolant flow is especially important and the ability to model two-phase flow is essential in modeling boiling water reactors. Two-phase flow modeling is integral as it models the vast differences in flow from the bottom of the core to the top at the steam separators. TRACE has the ability to reproduce these essential parameters. PARCS (Purdue Advanced Reactor Core Simulator) is a multi-dimensional reactor kinetics code. TRACE coupled with PARCS has the computing power to provide accurate coupled power and flow distributions under various reactor transients or casualties. TRACE/PARCS was previously validated for use with Pressurized Water Reactor (PWR) transient analysis using the OECD/NEA Main Steam Line Break Benchmark. This thesis focuses on the evaluation of Boiling Water Reactor (BWR) transient analysis, mainly the NEA Ringhals 1 Stability Benchmark from 1996. This benchmark performed a series of tests on the Ringhals 1 reactor during the beginning of cycles 14, 15, 16, and 17. Three techniques for initiating instabilities (pressure perturbation, control rod perturbation, and simulated noise) were performed on each test point during each cycle. The steady state data as well as the transient results predicted by TRACE/PARCS reasonably agree with the measured data from the NEA Ringhals 1 Stability Benchmark.
Author: Christopher Gosdin Publisher: ISBN: Category : Languages : en Pages :
Book Description
CTF, the version of the thermal-hydraulic sub-channel code COBRA-TF being jointly developed and maintained by Pennsylvania State University (PSU) and Oak Ridge National Laboratory (ORNL) for applications in the U.S. Department of Energy (DOE) supported Consortium for Advanced Simulation of Light Water Reactors (CASL) project, uses a two-fluid, three-field representation of two-phase flow, which makes the code capable of modeling two-phase flow in Boiling Water Reactors (BWR) during nominal operating conditions. The sub-channel code CTF is used for Pressurized Water Reactors (PWR) for best-estimate evaluations of the nuclear reactor safety margins; however, due to its capabilities, CTF is powerful and valuable computational tool for modeling BWRs. CTF has been subjected to a strict verification procedure, by addressing the mathematical accuracy of the numerical solutions on multiple stages. The code was then validated using numerous of experimental databases, including the U.S. Nuclear Regulatory Commission (NRC) / Nuclear Energy Agency of the Organization for Economic Co-operation and Development (NEA-OECD) Boiling Water Reactor Full Bundle Tests (BFBT) Benchmark. The BFBT benchmark contains a large amount of test cases representative of BWRs steady-state and off-nominal operating conditions, which makes it one of the most widely used benchmark for validating BWR modeling tools. Two of the main experimental tests involve critical power tests and void distribution tests. Specific experimental cases were chosen and simulated using CTF. Statistical studies were carried out on the void distribution cases to evaluate the code modeling uncertainties. This thesis also focuses on application of CTF to mini- and whole-core BWR calculations on a pin-cell resolved level; as well as on demonstrating that CTF can properly model bypass flow in BWR cores. To increase the confidence in the CTF's BWR modeling capabilities, extensive simulations have been performed using the international NEA-OECD / US NNRC Oskarshamn-2 benchmark, including modeling of a single and 2x2 assemblies on a pin-by-pin level, and a full core model on an assembly level. Each model is varied, with an increasing amount of detail. The results demonstrate that CTF is capable of modeling basic and complex BWR core configurations and operating conditions. Using the three Oskarshamn-2 simulations, CTF's capabilities of modeling BWRs was further verified.
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Author: Christopher Michael Briggs
Author: Hany Samy Abdel-Khalik
Author: Jack Douglas Galloway
Author: Robert Allen Walls
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Author: Christopher Gosdin
Author: J. A. Woolley
Author: W. K. Lam
Author: Chi Kao