Sampling-Based Nuclear Data Uncertainty Quantification for Continuous Energy Monte Carlo Codes
Author:
Publisher:
Published: 2015
Total Pages: 150
ISBN-13:
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A Monte Carlo Framework for Nuclear Data Uncertainty Propagation Via the Windowed Multipole Formalism
Author: Abdulla Abdulaziz Alhajri
Publisher:
Published: 2020
Total Pages: 228
ISBN-13:
DOWNLOAD EBOOKA new framework has been developed that calculates the uncertainty in calculated quantities, such as K[subscript eff], reactivity coefficients, multigroup cross sections, and reaction rate ratios, that arise due to uncertainties in the underlying nuclear data. This framework relies on first order uncertainty analysis using sensitivity methods. The major innovation in the proposed framework is the use of the windowed multipole formalism for calculating the sensitivities. The use of the windowed multipole formalism provides a natural, physics-inspired binning strategy for the sensitivity coefficients, while also aiding in the statistical convergence of the calculated sensitivity tallies. Additionally, our framework improves on existing methods by fully accounting for temperature effects. The proposed method allows for identifying exactly the resonances and parameters that are driving the uncertainty, and thus provides guidance to nuclear data evaluators and experimenters on how to reduce the uncertainty in the most efficient manner. Calculating the uncertainty requires two key pieces of information; the windowed multipole sensitivity coefficients, and the windowed multipole covariance matrix. A sensitivity coefficient calculation algorithm based on the CLUTCH-FM methodology was implemented in OpenMC. Several methods for obtaining the windowed multipole covariance matrix from the resonance parameter covariance matrix were explored, and ultimately an approach based on random-sampling was selected. Along the way, an analytical benchmark was developed for the purposes of validating the framework, as well as the implementation. This analytical benchmark consists of a solution to the forward and adjoint neutron transport equations. The windowed multipole covariance matrix was calculated for three isotopes; 238U , 157Gd , and 23Na . The uncertainty in K[subscript eff] due to the uncertainty in the 238U and 157Gd cross sections was calculated for two criticality safety benchmarks, and a beginning-of-life PWR model. The uncertainty of several reaction rate ratios due to the uncertainty in the 157Gd cross section was also calculated for the PWR model. The resonances of 238U and 157Gd that have the largest contribution to the uncertainty were identified for the criticality safety benchmarks.
Enhancements in Continuous-Energy Monte Carlo Capabilities for SCALE 6.2
Author:
Publisher:
Published: 2014
Total Pages:
ISBN-13:
DOWNLOAD EBOOKSCALE is a widely used suite of tools for nuclear systems modeling and simulation that provides comprehensive, verified and validated, user-friendly capabilities for criticality safety, reactor physics, radiation shielding, and sensitivity and uncertainty analysis. For more than 30 years, regulators, industry, and research institutions around the world have used SCALE for nuclear safety analysis and design. SCALE provides a plug-and-play framework that includes three deterministic and three Monte Carlo radiation transport solvers that are selected based on the desired solution. SCALE includes the latest nuclear data libraries for continuous-energy and multigroup radiation transport as well as activation, depletion, and decay calculations. SCALE s graphical user interfaces assist with accurate system modeling, visualization, and convenient access to desired results. SCALE 6.2 provides several new capabilities and significant improvements in many existing features, especially with expanded continuous-energy Monte Carlo capabilities for criticality safety, shielding, depletion, sensitivity and uncertainty analysis, and improved fidelity in nuclear data libraries. A brief overview of SCALE capabilities is provided with emphasis on new features for SCALE 6.2.
Nuclear Computations Under Uncertainty
Author: Pablo Philippe Ducru
Publisher:
Published: 2021
Total Pages: 343
ISBN-13:
DOWNLOAD EBOOKThese contributions are documented in nine peer-reviewed journal articles (eight published and one under review) and seven conference articles (six published and one under review), constituting the core of this thesis.
Accelerator-Driven System at Kyoto University Critical Assembly
Author: Cheol Ho Pyeon
Publisher: Springer
Published: 2021-03-20
Total Pages: 352
ISBN-13: 9789811603433
DOWNLOAD EBOOKThis open access book is a unique compilation of experimental benchmark analyses of the accelerator-driven system (ADS) at the Kyoto University Critical Assembly (KUCA) on the most recent advances in the development of computational methods. It is devoted especially to nuclear engineers and scientists. Readers will find a detailed description of advanced measurement techniques and calculation methodologies for the ADS with 14 MeV neutrons and high-energy neutrons (with combined use of 100 MeV protons and Pb-Bi target) at KUCA. Additionally, experimental results of nuclear transmutation of minor actinides by ADS and at a critical state are included. Readers also have access to benchmarks of specific ADS experiments with raw data in the Appendix. The book is a valuable resource for the ADS experiments at KUCA which are globally recognized as both static and kinetic studies from the point of view of fundamental research.
Computational Methods for Efficient Nuclear Data Management in Monte Carlo Neutron Simulations
Author: Jonathan Alan Walsh
Publisher:
Published: 2014
Total Pages: 133
ISBN-13:
DOWNLOAD EBOOKThis thesis presents the development and analysis of computational methods for efficiently accessing and utilizing nuclear data in Monte Carlo neutron transport code simulations. Using the OpenMC code, profiling studies are conducted in order to determine the types of nuclear data that are used in realistic reactor physics simulations, as well as the frequencies with which those data are accessed. The results of the profiling studies are then used to motivate the conceptualization of a nuclear data server algorithm aimed at reducing on-node memory requirements through the use of dedicated server nodes for the storage of infrequently accessed data. A communication model for this algorithm is derived and used to make performance predictions given data access frequencies and assumed system hardware parameters. Additionally, a new, accelerated approach for rejection sampling the free gas resonance elastic scattering kernel that reduces the frequency of zero-temperature elastic scattering cross section data accesses is derived and implemented. Using this new approach, the runtime overhead incurred by an exact treatment of the free gas resonance elastic scattering kernel is reduced by more than 30% relative to a standard sampling procedure used by Monte Carlo codes. Finally, various optimizations of the commonly-used binary energy grid search algorithm are developed and demonstrated. Investigated techniques include placing kinematic constraints on the range of the searchable energy grid, index lookups on unionized material energy grids, and employing energy grid hash tables. The accelerations presented routinely result in overall code speedup by factors of 1.2-1.3 for simulations of practical systems.
Monte Carlo Capabilities of the SCALE Code System
Author:
Publisher:
Published: 2014
Total Pages: 12
ISBN-13:
DOWNLOAD EBOOKSCALE is a broadly used suite of tools for nuclear systems modeling and simulation that provides comprehensive, verified and validated, user-friendly capabilities for criticality safety, reactor physics, radiation shielding, and sensitivity and uncertainty analysis. For more than 30 years, regulators, licensees, and research institutions around the world have used SCALE for nuclear safety analysis and design. SCALE provides a "plug-and-play" framework that includes three deterministic and three Monte Carlo radiation transport solvers that can be selected based on the desired solution, including hybrid deterministic/Monte Carlo simulations. SCALE includes the latest nuclear data libraries for continuous-energy and multigroup radiation transport as well as activation, depletion, and decay calculations. SCALE's graphical user interfaces assist with accurate system modeling, visualization, and convenient access to desired results. SCALE 6.2 will provide several new capabilities and significant improvements in many existing features, especially with expanded continuous-energy Monte Carlo capabilities for criticality safety, shielding, depletion, and sensitivity and uncertainty analysis. Finally, an overview of the Monte Carlo capabilities of SCALE is provided here, with emphasis on new features for SCALE 6.2.
On-the-fly Nuclear Data Processing Methods for Monte Carlo Simulations of Intermediate and Fast Spectrum Systems
Author: Jonathan Alan Walsh
Publisher:
Published: 2016
Total Pages: 212
ISBN-13:
DOWNLOAD EBOOKComputational methods for on-the-fly representation and processing of nuclear data within Monte Carlo neutron transport simulations of intermediate and fast spectrum systems are developed and implemented in a continuous-energy Monte Carlo code. First, a capability to compute temperature-dependent unresolved resonance region (URR) cross sections directly from zero-temperature average resonance parameters is presented. The use of this capability in benchmarking both evaluated and processed URR data is demonstrated. Results of this benchmarking lead to a partial resolution of a longstanding discrepancy between experiment and calculation results for a well-known fast critical assembly. Next, an on-the-fly probability table interpolation scheme for computing temperature-dependent URR cross sections is developed and used in analyses which show that interpolation on a relatively coarse temperature mesh (>100 K) can be used to reproduce results obtained with cross sections generated at an exact temperature. This enables the simulation of systems having detailed temperature distributions using probability table data which require significantly less memory than data generated on a fine temperature mesh. Additional methods for use in the investigation of two common approximations that are made in representing URR cross section data are developed. Namely, a multi-level URR cross section calculation capability is used to show that level-level interference effects in elastic scattering cross sections are negligible in many cases of interest. A capability to generate resonance structure in competitive reaction cross sections is used to show that neglecting cross section structure for reactions other than elastic scattering, capture, and fission can lead to non-negligible, unconservative biases (>100 pcm) in criticality safety calculations. The principal underlying assumption of the probability table method is also tested by comparing the results it yields with results that are averaged over many independent simulations, each using a single, independent realization of URR resonance parameters. Unknown URR resonance structure is observed to induce an uncertainty on the multiplication factor for intermediate and fast spectrum systems that is nearly an order of magnitude greater than that which is purely stochastic. This significantly increases the uncertainty to which results of simulations of those systems should be stated. Finally, a procedure for consistent, on-the-fly sampling of temperature-dependent neutron reaction kernels which requires no additional secondary distribution data is presented. It is used to show that Doppler effects may have only a small impact on elastic scattering secondary angular distributions at typical power reactor operating temperatures but can be appreciable at astrophysical temperatures.
Sensitivity & Uncertainty Analysis, Volume 1
Author: Dan G. Cacuci
Publisher: CRC Press
Published: 2003-05-28
Total Pages: 304
ISBN-13: 0203498798
DOWNLOAD EBOOKAs computer-assisted modeling and analysis of physical processes have continued to grow and diversify, sensitivity and uncertainty analyses have become indispensable investigative scientific tools in their own right. While most techniques used for these analyses are well documented, there has yet to appear a systematic treatment of the method based
Nuclear Data for Science and Technology
Author: Syed M. Qaim
Publisher: Springer Science & Business Media
Published: 2012-12-06
Total Pages: 1041
ISBN-13: 3642581137
DOWNLOAD EBOOKThis book describes the Proceedings of the International Conference on Nuclear Data for Science and Technology held at Jillich in May 1991. The conference was in a series of application oriented nuclear data conferences organized in the past under the auspices of the Nuclear Energy Agency-Nuclear Data Committee (NEANDC) and with the support of the Nuclear Energy Agency-Committee on Reactor Physics (NEACRP). It was the fIrst international conference on nuclear data held in Germany, with the scientific responsibility entrusted to the Institute of Nuclear Chemistry of the Research Centre Jillich. The scientific programme was established by the International Programme Committee in consultation with the International Advisers, and the NEA and IAEA cooperated in the organization. A total of 328 persons from 37 countries and fIve international organizations participated. The scope of these Proceedings extends to a wide range of interdisciplinary topics dealing with measu rement, calculation, evaluation and application of nuclear data, with a major emphasis on numerical data. Both energy and non-energy related applications are considered and due attention is given to some fundamental aspects relevant to the understanding of nuclear data.
