Thesis (Ph. D.)--Technische Universiteit Delft, 2006. / Includes bibliographical references (p. 121-124).

A new method for constructing probability tables in the Unresolved Resonance Region (URR) has been developed. This new methodology is an extensive modification of the Single-Level Breit-Wigner (SLBW) resonance-pair sequence method commonly used to generate probability tables in the URR. Using a Monte Carlo process, many resonance-pair sequences are generated by sampling the average resonance parameter data for the unresolved resonance region from the ENDF data file. The resonance parameters are then converted to the Reich-Moore format to take advantage of the more robust R-Matrix Limited (RML) format. For each sampled set of resonance-pair sequences, the temperature-dependent cross sections are calculated on a small grid around the energy of reference using the RML formalism and the Leal-Hwang Doppler broadening methodology. The effective cross sections calculated at the energy of reference are then used to construct probability tables in the unresolved resonance region. The RML cross section reconstruction algorithm has been rigorously tested for a variety of isotopes, including O-16, F-19, Cl-35, Fe-56, Cu-63, and Cu-65. The new URR method also produced normalized cross-section factor probability tables for U-238 that were found to be in agreement with current standards. The modified U-238 probability tables were shown to produce k-eff results in excellent agreement with several standard benchmarks, including the IEU-MET-FAST-007, IEU-MET-FAST-003, and IEU-COMP-FAST-004 benchmarks.

The Xilinx Multiprocessor System-on-Chip (MPSoC) is a complex device that uses 16nm FinFET technology to combine multiple processors, a large amount of FPGA resources, and many I/O interfaces on a single chip die. These features make the MPSoC a high-performance and architecturally flexible device. The potential computing power makes the MPSoC ideal for many embedded applications including terrestrial and space applications. The MPSoC, however, does not have extensive radiation history as many other devices have. The extent of the effect that ionized particles may have on the MPSoC is not well established. To solve this problem, neutron radiation testing can be used to determine the device's susceptibility to single-event upsets (SEUs). Though this thesis is not intended to qualify the MPSoC for space, this work does provide useful neutron radiation test data that helps to characterize the susceptible nature of the device. This thesis summarizes the SEU results obtained from neutron testing on the UltraScale+ MPSoC ZU9EG device. A series of three neutron beam tests were performed on the MPSoC ZU9EG at Los Alamos National Laboratories (LANL). Testing was performed using a novel testing methodology to collect SEU counts on the programmable logic and the processing system simultaneously. These results show a 10.1× improvement of the programmable logic CRAM over the previous Xilinx UltraScale device series.

The Xilinx Multiprocessor System-on-Chip (MPSoC) is a complex device that uses 16nm FinFET technology to combine multiple processors, a large amount of FPGA resources, and many I/O interfaces on a single chip die. These features make the MPSoC a high-performance and architecturally flexible device. The potential computing power makes the MPSoC ideal for many embedded applications including terrestrial and space applications.The MPSoC, however, does not have extensive radiation history as many other devices have. The extent of the effect that ionized particles may have on the MPSoC is not well established. To solve this problem, neutron radiation testing can be used to determine the device's susceptibility to single-event upsets (SEUs). . Though this thesis is not intended to qualify the MPSoC for space, this work does provide useful neutron radiation test data that helps to characterize the susceptible nature of the device. This thesis summarizes the SEU results obtained from neutron testing on the UltraScale+ MPSoC ZU9EG device. A series of three neutron beam tests were performed on the MPSoC ZU9EG at Los Alamos National Laboratories (LANL). Testing was performed using a novel testing methodology to collect SEU counts on the programmable logic and the processing system simultaneously. These results show a $10.1 timess improvement of the programmable logic CRAM over the previous Xilinx UltraScale device series.

Calculations of reactor parameters of interest (such as neutron multiplication factors, decay heat, reaction rates, etc.), are often based on models which are dependent on groupwise neutron cross sections. The uncertainties associated with these neutron cross sections are propagated to the final result of the calculated reactor parameters. There is a need to characterize this uncertainty and to be able to apportion the uncertainty in a calculated reactor parameter to the different sources of uncertainty in the groupwise neutron cross sections, this procedure is known as sensitivity analysis. The focus of this study is the application of a modified global sensitivity analysis technique to calculations of reactor parameters that are dependent on groupwise neutron cross–sections. Sensitivity analysis can help in identifying the important neutron cross sections for a particular model, and also helps in establishing best–estimate optimized nuclear reactor physics models with reduced uncertainties. In this study, our approach to sensitivity analysis will be similar to the variance–based global sensitivity analysis technique, which is robust, has a wide range of applicability and provides accurate sensitivity information for most models. However, this technique requires input variables to be mutually independent. A modification to this technique, that allows one to deal with input variables that are block–wise correlated and normally distributed, is presented. The implementation of the modified technique involves the calculation of multi–dimensional integrals, which can be prohibitively expensive to compute. Numerical techniques specifically suited to the evaluation of multidimensional integrals namely Monte Carlo, quasi–Monte Carlo and sparse grids methods are used, and their efficiency is compared. The modified technique is illustrated and tested on a two–group cross–section dependent problem. In all the cases considered, the results obtained with sparse grids achieved much better accuracy, while using a significantly smaller number of samples. / Thesis (M.Sc. Engineering Sciences (Nuclear Engineering))--North-West University, Potchefstroom Campus, 2011.

Calculations of reactor parameters of interest (such as neutron multiplication factors, decay heat, reaction rates, etc.), are often based on models which are dependent on groupwise neutron cross sections. The uncertainties associated with these neutron cross sections are propagated to the final result of the calculated reactor parameters. There is a need to characterize this uncertainty and to be able to apportion the uncertainty in a calculated reactor parameter to the different sources of uncertainty in the groupwise neutron cross sections, this procedure is known as sensitivity analysis. The focus of this study is the application of a modified global sensitivity analysis technique to calculations of reactor parameters that are dependent on groupwise neutron cross–sections. Sensitivity analysis can help in identifying the important neutron cross sections for a particular model, and also helps in establishing best–estimate optimized nuclear reactor physics models with reduced uncertainties. In this study, our approach to sensitivity analysis will be similar to the variance–based global sensitivity analysis technique, which is robust, has a wide range of applicability and provides accurate sensitivity information for most models. However, this technique requires input variables to be mutually independent. A modification to this technique, that allows one to deal with input variables that are block–wise correlated and normally distributed, is presented. The implementation of the modified technique involves the calculation of multi–dimensional integrals, which can be prohibitively expensive to compute. Numerical techniques specifically suited to the evaluation of multidimensional integrals namely Monte Carlo, quasi–Monte Carlo and sparse grids methods are used, and their efficiency is compared. The modified technique is illustrated and tested on a two–group cross–section dependent problem. In all the cases considered, the results obtained with sparse grids achieved much better accuracy, while using a significantly smaller number of samples. / Thesis (M.Sc. Engineering Sciences (Nuclear Engineering))--North-West University, Potchefstroom Campus, 2011.

In zukünftigen kerntechnischen Anlagen können die Materialien Blei und Bismut eine größere Rolle spielen als heute. Für die Planung dieser Anlagen werden verlässliche Wirkungsquerschnittsdaten benötigt. Insbesondere der Neutronentransport in einem Blei-Spallationstarget eines beschleunigergetriebenen unterkritischen Reaktors hängt stark von den inelastischen Streuquerschnitten im Energiebereich von 0,5 MeV bis 6 MeV ab. In den vergangenen 20 Jahren wurden elastische und inelastische Neutronenstreuquerschnitte mit hoher Präzision für eine Vielzahl von Elementen am PTB-Flugzeitspektrometer gemessen. Zur Erzeugung der Neutronen wurde hauptsächlich die D(d,n)-Reaktion genutzt. Aufgrund des Q-Wertes der Reaktion und der verfügbaren Deuteronenenergien können Neutronen im Energiebereich von 6 MeV bis 16 MeV erzeugt werden. Die Messung von Wirkungsquerschnitten bei niedrigeren Energien erfordert somit die Verwendung einer anderen neutronenerzeugenden Reaktion. Hierfür wurde die 15N(p,n)15O-Reaktion ausgewählt, da sie die Erzeugung monoenergetischer Neutronen bis zu einer Energie von 5,7 MeV erlaubt. In dieser Arbeit wird die 15N(p,n)-Reaktion auf ihre Eignung als Quelle monoenergetischer Neutronen in Streuexperimenten untersucht. Die Untersuchung der Reaktion beinhaltet die Messung von differentiellen Wirkungsquerschnitten für ausgewählte Energien und die Auswahl von optimalen Targetbedingungen. Differentielle elastische und inelastische Neutronenstreuquerschnitte wurden unter Anwendung der Flugzeitmethode für Blei bei vier Energien zwischen 2 MeV und 4 MeV gemessen. Eine Bleiprobe mit natürlicher Isotopenzusammensetzung wurde verwendet. Für den Nachweis der gestreuten Neutronen wurden NE213 Flüssigszintillatoren verwendet, deren Nachweiswahrscheinlichkeit gut bekannt ist. Winkelintegrierte Wirkungsquerschnitte wurden mit einem Legendre-Polynomfit unter Verwendung der Methode der kleinsten Quadrate bestimmt. Zusätzlich erfolgten Messungen für die isotopenreinen Streuproben 209Bi und 181Ta bei 4 MeV Neutronenenergie. Die Ergebnisse werden mit denen früherer Experimente und aktuellen Evaluationen verglichen. / In future nuclear facilities, the materials lead and bismuth can play a more important role than in today’s nuclear reactors. Reliable cross section data are required for the design of those facilities. In particular the neutron transport in the lead spallation target of an Accelerator-Driven Subcritical Reactor strongly depends on the inelastic neutron scattering cross sections in the energy region from 0.5MeV to 6 MeV. In the recent 20 years, elastic and inelastic neutron scattering cross sections were measured with high precision for a variety of elements at the PTB time-of-flight spectrometer. The D(d,n) reaction was primarily used for the production of neutrons. Because of the Q value of the reaction and the available deuteron energies, neutrons in the energy range from 6MeV to 16MeV can be produced. For the cross section measurement at lower energies, however, another neutron producing reaction is required. The 15N(p,n)15O reaction was chosen, as it allows the production of monoenergetic neutrons with up to 5.7MeV energy. In this work, the 15N(p,n) reaction was studied with focus on the suitability as a source for monoenergetic neutrons in scattering experiments. This includes the measurement of differential cross sections for the neutron producing reaction and the choice of optimum target conditions. Differential elastic and inelastic neutron scattering cross sections were measured for lead at four energies in the region from 2MeV to 4MeV incident neutron energy using the time-offlight technique. A lead sample with natural isotopic composition was used. NE213 liquid scintillation detectors with well-known detection efficiencies were used for the detection of the scattered neutrons. Angle-integrated cross sections were determined by a Legendre polynomial expansion using least-squares methods. Additionally, measurements were carried out for isotopically pure 209Bi and 181Ta samples at 4MeV incident neutron energy. Results are compared with other measurements and recent evaluations.

Pastilhas de nitrato de potássio foram irradiadas no núcleo do reator de pesquisas IEA-R1m do Instituto de Pesquisas Energéticas e Nucleares, IPEN/CNEN-SP, operando a 2 MW de potência, para determinar a secção de choque térmica e integral de ressonância da reação 41K(n,g)42K. O fluxo de nêutrons foi monitorado com folhas de liga ouro-alumínio. As atividades induzidas nos alvos foram determinadas por espectroscopia gama com detectores de germânio hiper puro. Os cálculos realizados se basearam no formalismo de Westcott. Foram realizadas simulações com o código MCNP (Monte Carlo N-Particle) para determinar a auto-blindagem e a depressão do fluxo de nêutrons nas pastilhas durante as irradiações e os fatores de correção da eficiência de detecção para fontes volumétricas, que leva em conta a absorção de raios gama nas mesmas. Foi efetuado um tratamento estatístico das incertezas envolvidas e determinadas as covariâncias entre os resultados, incluindo aquelas decorrentes das incertezas do padrão de referência (ouro). Os resultados obtidos foram comparados com os de outros autores. Foi testada a possibilidade de se observar o produto da reação 41K(n,g)42K(n,g)43K. / Pellets of potassium nitrate were irradiated in the IPEN/CNEN-SP (Instituto de Pesquisas Energeticas e Nucleares, Comissao de Energia Nuclear, Sao Paulo, SP) IEAR1m reactor core operating at 2 MW power in order to determine the 41K(n,g)42K reaction thermal cross-section and resonance integral. The neutron flux was monitored by Au-Al alloy foils, and the Westcott formalism was applied. Neutron self-shielding, flux depression and gamma-ray self-absorption in the relatively large samples, as well as the gamma-ray detection efficiency correction factor, were determined by simulation with MCNP code. The data reduction statistical methods included the determination of the covariances between the obtained results and the standard cross-sections used (Au). The results were compared to those already published. The observation of the consecutive neutron capture reaction leading to 43K was tried.

Pastilhas de nitrato de potássio foram irradiadas no núcleo do reator de pesquisas IEA-R1m do Instituto de Pesquisas Energéticas e Nucleares, IPEN/CNEN-SP, operando a 2 MW de potência, para determinar a secção de choque térmica e integral de ressonância da reação 41K(n,g)42K. O fluxo de nêutrons foi monitorado com folhas de liga ouro-alumínio. As atividades induzidas nos alvos foram determinadas por espectroscopia gama com detectores de germânio hiper puro. Os cálculos realizados se basearam no formalismo de Westcott. Foram realizadas simulações com o código MCNP (Monte Carlo N-Particle) para determinar a auto-blindagem e a depressão do fluxo de nêutrons nas pastilhas durante as irradiações e os fatores de correção da eficiência de detecção para fontes volumétricas, que leva em conta a absorção de raios gama nas mesmas. Foi efetuado um tratamento estatístico das incertezas envolvidas e determinadas as covariâncias entre os resultados, incluindo aquelas decorrentes das incertezas do padrão de referência (ouro). Os resultados obtidos foram comparados com os de outros autores. Foi testada a possibilidade de se observar o produto da reação 41K(n,g)42K(n,g)43K. / Pellets of potassium nitrate were irradiated in the IPEN/CNEN-SP (Instituto de Pesquisas Energeticas e Nucleares, Comissao de Energia Nuclear, Sao Paulo, SP) IEAR1m reactor core operating at 2 MW power in order to determine the 41K(n,g)42K reaction thermal cross-section and resonance integral. The neutron flux was monitored by Au-Al alloy foils, and the Westcott formalism was applied. Neutron self-shielding, flux depression and gamma-ray self-absorption in the relatively large samples, as well as the gamma-ray detection efficiency correction factor, were determined by simulation with MCNP code. The data reduction statistical methods included the determination of the covariances between the obtained results and the standard cross-sections used (Au). The results were compared to those already published. The observation of the consecutive neutron capture reaction leading to 43K was tried.

Amostras de nitrato de potássio e enxofre natural foram irradiadas no núcleo do reator de pesquisas IEA-R1 do Instituto de Pesquisas Energéticas e Nucleares, IPEN/CNEN-SP, operando entre 3,5 e 5 MW de potência, para determinar as secções de choque térmicas e integrais de ressonância das reações 34S(n,)35S e 42K(n,)43K. O fluxo de nêutrons foi monitorado com ligas ouro-alumínio. As atividades induzidas nos alvos de ouro-alumínio e nitrato de potássio foram medidas com um detector semicondutor de germânio hiper puro; as atividades dos alvos de enxofre foram determinadas com um sistema de coincidências 4\\pi\\beta - \\gamma. Os efeitos de depressão de fluxo, autoblindagem e autoabsorção nos alvos foram corrigidos com base em simulações com o método de Monte Carlo por meio do programa MCNP. O programa PENELOPE, também baseado no método de Monte Carlo, foi modificado para simular o comportamento do sistema de coincidências 4\\pi\\beta - \\gamma. O formalismo de Westcott e o método da razão de cádmio foram usados para determinar os fluxos de nêutrons térmicos e epitérmicos, assim como as secções de choque térmicas e integrais de ressonância de ambos nuclídeos. Foi efetuado um tratamento estatístico das incertezas envolvidas e determinadas as covariâncias entre os resultados, incluindo aquelas decorrentes das incertezas do padrão de referência (ouro). As reações 34S(n,)35S e 42K(n,)43K apresentaram, respectivamente, seções de choque térmicas de 228(14) mb e 44,8(9) b, e integrais de ressonância de 144(6) mb e 1635(75) b. Estes resultados são incompatíveis com aqueles obtidos com cálculos teóricos. A seção de choque térmica da reação 34S(n,)35S concorda com valores obtidos por outros autores, dentro das incertezas experimentais. / Samples of potassium nitrate and natural sulphur were irradiated in the IPEN/CNEN-SP IEA-R1 research reactor core, operating between 3.5 and 5 MW, to determine the thermal neutron cross sections and resonance integrals of 34S(n,)35S and 42K(n,)43K reactions. The neutron flux was monitored with gold-aluminium alloy. The activities induced in targets of gold-aluminium and potassium nitrate were measured with a high purity germanium detector. Sulphur targets activities were determined with a 4\\pi\\beta-\\gamma coincidences system by the tracer method. Flux depression, self-shielding and self-absorption in the targets was evaluated by simulations using the MCNP software. The PENELOPE software, also based on Monte Carlo method, was modified to simulate the behavior of the 4\\pi\\beta-\\gamma coincidence system. The Westcott formalism and the cadmium ratio method were used to determine epithermal and thermal neutrons flux as well as the thermal cross sections and resonance integrals of both nuclides. A statistical analysis of the uncertainties was performed and the covariance between the results was determined, including those arising from the uncertainties of the gold reference standard. The results were compared with experimental values and theoretical predictions obtained by other authors. The 34S(n,)35S and 42K(n,)43K reactions had, respectively, thermal cross sections of 228(14) mb and 44.8(9) b, and integral resonances of 144(6) mb and 1635(75) b. These results are incompatible with the obtained with theoretical calculations. The 34S(n,)35S reaction thermal cross section agrees with values obtained by other authors, within the experimental uncertainties.