Theses - Nuclear Engineering
http://hdl.handle.net/1721.1/7853
2017-12-20T17:31:05ZReactor agnostic multi-group cross section generation for fine-mesh deterministic neutron transport simulations
http://hdl.handle.net/1721.1/112525
Reactor agnostic multi-group cross section generation for fine-mesh deterministic neutron transport simulations
Boyd, William Robert Dawson, III
A key challenge for full-core transport methods is reactor agnostic multi-group cross section (MGXS) generation. Monte Carlo (MC) presents the most accurate method for MGXS generation since it does not require any approximations to the neutron flux. This thesis develops novel methods that use MC to generate the fine-spatial mesh MGXS that are needed by high-fidelity transport codes. These methods employ either engineering-based or statistical clustering algorithms to accelerate the convergence of MGXS tallied on fine, heterogeneous spatial meshes by Monte Carlo. The traditional multi-level approach to MGXS generation is replaced by full-core MC calculations that generate MGXS for multi-group deterministic transport codes. Two pinwise spatial homogenization schemes are introduced to model the clustering of pin-wise MGXS due to spatial self-shielding spectral effects. The Local Neighbor Symmetry (LNS) scheme uses a nearest neighbor-like analysis of a reactor geometry to determine which fuel pins should be assigned the same MGXS. The inferential MGXS (iMGXS) scheme applies unsupervised machine learning algorithms to "noisy" MC tally data to identify clustering of pin-wise MGXS without any knowledge of the reactor geometry. Both schemes simultaneously account for spatial self-shielding effects while also accelerating the convergence of the MC tallies used to generate MGXS. The LNS and iMGXS schemes were used to model MGXS clustering from radial geometric heterogeneities in a suite of 2D PWR benchmarks. Both schemes reduced U-238 capture rate errors by up to a factor of four with respect to schemes which neglect to model MGXS clustering. In addition, the schemes required an order of magnitude fewer MC particle histories to converge MGXS for multi-group deterministic calculations than a reference MC calculation. These results demonstrate the potential for single-step MC simulations of the complete heterogeneous geometry as a means to generate reactor agnostic MGXS for deterministic transport codes. The LNS and iMGXS schemes may be valuable for reactor physics analyses of advanced LWR core designs and next generation reactors with spatial heterogeneities that are poorly modeled by the engineering approximations in today's methods for MGXS generation.
Thesis: Ph. D., Massachusetts Institute of Technology, Department of Nuclear Science and Engineering, 2017.; Cataloged from PDF version of thesis.; Includes bibliographical references (pages 487-495).
2017-01-01T00:00:00ZAcceleration methods for Monte Carlo particle transport simulations
http://hdl.handle.net/1721.1/112521
Acceleration methods for Monte Carlo particle transport simulations
Li, Lulu, Ph. D. Massachusetts Institute of Technology
Performing nuclear reactor core physics analysis is a crucial step in the process of both designing and understanding nuclear power reactors. Advancements in the nuclear industry demand more accurate and detailed results from reactor analysis. Monte Carlo (MC) eigenvalue neutron transport methods are uniquely qualified to provide these results, due to their accurate treatment of space, angle, and energy dependencies of neutron distributions. Monte Carlo eigenvalue simulations are, however, challenging, because they must resolve the fission source distribution and accumulate sufficient tally statistics, resulting in prohibitive run times. This thesis proposes the Low Order Operator (LOO) acceleration method to reduce the run time challenge, and provides analyses to support its use for full-scale reactor simulations. LOO is implemented in the continuous energy Monte Carlo code, OpenMC, and tested in 2D PWR benchmarks. The Low Order Operator (LOO) acceleration method is a deterministic transport method based on the Method of Characteristics. Similar to Coarse Mesh Finite Difference (CMFD), the other acceleration method evaluated in this thesis, LOO parameters are constructed from Monte Carlo tallies. The solutions to the LOO equations are then used to update Monte Carlo fission sources. This thesis deploys independent simulations to rigorously assess LOO, CMFD, and unaccelerated Monte Carlo, simulating up to a quarter of a trillion neutron histories for each simulation. Analysis and performance models are developed to address two aspects of the Monte Carlo run time challenge. First, this thesis demonstrates that acceleration methods can reduce the vast number of neutron histories required to converge the fission source distribution before tallies can be accumulated. Second, the slow convergence of tally statistics is improved with the acceleration methods for the earlier active cycles. A theoretical model is developed to explain the observed behaviors and predict convergence rates. Finally, numerical results and theoretical models shed light on the selection of optimal simulation parameters such that a desired statistical uncertainty can be achieved with minimum neutron histories. This thesis demonstrates that the conventional wisdom (e.g., maximizing the number of cycles rather than the number of neutrons per cycle) in performing unaccelerated MC simulations can be improved simply by using more optimal parameters. LOO acceleration provides reduction of a factor of at least 2.2 in neutron histories, compared to the unaccelerated Monte Carlo scheme, and the CPU time and memory overhead associated with LOO are small.
Thesis: Ph. D., Massachusetts Institute of Technology, Department of Nuclear Science and Engineering, 2017.; Cataloged from PDF version of thesis.; Includes bibliographical references (pages 166-175).
2017-01-01T00:00:00ZQuantifying the adhesion of noble metal foulants on structural materials in a Molten Salt Reactor
http://hdl.handle.net/1721.1/112520
Quantifying the adhesion of noble metal foulants on structural materials in a Molten Salt Reactor
Tanaka, Reid S
As discovered during the Molten Salt Reactor (MSR) Experiment (MSRE), selected fission products deposited on the wetted surfaces throughout the reactor. Fission products such as molybdenum and ruthenium are noble with respect to the electrochemical potential of the fluoride fuel salt and therefore remain insoluble in their elemental forms rather than becoming ionic salts. Coalescing in the primary fluid, these noble metals then migrate and eventually deposit on internal reactor surfaces. Since the bulk of these noble metal fission products are also energetically unstable they bring not only physical fouling, but heat and radiation from decay as well. The adherence forces of five of the seven principal radioactive foulants discovered during the MSRE were measured on six different potential structural materials by atomic force microscope force spectroscopy (AFM-FS). The noble metals studied were niobium, molybdenum, ruthenium, antimony and tellurium. Structural materials measured were Hastelloy-N, the primary structural metal of the MSRE; two steels, SS316L and F91; commercially pure nickel and molybdenum; and silicon carbide. MSRs operate with surfaces free of passivating corrosion layers, so the measurements were conducted on bare metal surfaces. An argon ion gun chamber was constructed for removal of the oxide layers from mechanically polished substrate metals by sputtering. A combination vacuum chamber/glove box was crafted to accept the sputter polished substrates into a dry, inert environment where the force adhesion measurements were made. Data acquired in the last phase of the study partially demonstrated the concept. The measured particle-to-substrate attractive forces found antimony and tellurium to be generally more adherent to the bare metals than niobium or molybdenum. The finding is consistent with the fouling examined in the MSRE. If held, this correlation of laboratory measurement to actual fouling may aid the reactor designer in anticipating fouling to plan for the effects. Such knowledge would inform selection of plant materials, both for operating components such as flow detectors and heat exchangers, where fission product deposition would be undesired, and for processing components such as filters and metal collection systems, where adhesion would be preferred.
Thesis: S.M., Massachusetts Institute of Technology, Department of Nuclear Science and Engineering, 2017.; Cataloged from PDF version of thesis.; Includes bibliographical references (pages 145-149).
2017-01-01T00:00:00ZRadiation damage quantification in elemental copper using Wigner energy storage
http://hdl.handle.net/1721.1/112482
Radiation damage quantification in elemental copper using Wigner energy storage
Carter, Ki-Jana
Radiation damage in materials can cause critical components in fission and fusion reactors to fail with potentially catastrophic consequences. Radiation damage quantification is essential for understanding, predicting, and preventing such failures. The current unit of radiation damage, displacements per atom (DPA), is not a measurable quantity, and it is known to be an inaccurate measure of radiation damage. This project aims to quantify radiation damage accurately and measurably by characterizing the storage of energy in radiation-induced material defects, known as Wigner energy storage. In order to gain an atomistic understanding of radiation damage, the irradiation and calorimetry of elemental copper were simulated using molecular dynamics code. A custom defect analysis script was used to determine the energy stored as a function of irradiation energy and defect type. Wigner energy peaks were clearly visible in the calorimetry data, indicating that Wigner energy measurement is a plausible technique for quantifying radiation damage. Future work should focus on achieving more realistic heating rates and measuring Wigner energy storage experimentally using fast scanning calorimetry.
Thesis: S.B., Massachusetts Institute of Technology, Department of Nuclear Science and Engineering, 2017.; Cataloged from PDF version of thesis.; Includes bibliographical references (pages 55-58).
2017-01-01T00:00:00Z