A neutronic and thermal hydraulic analysis of the 1-MW TRIGA research reactor at the Texas A&M University Nuclear Science Center using a new low enriched uranium fuel (named 30/20 fuel) was completed. This analysis provides safety assessment for the change out of the existing high enriched uranium fuel to this high-burnup, low enriched uranium fuel design. The codes MCNP and Monteburns were utilized for the neutronic analysis while the code PARET was used to determine fuel and cladding temperatures. All of these simulations used improved zirconium hydride cross sections that were provided by Dr. Ayman Hawari at North Carolina State University. The neutronic and thermal analysis showed that the reactor will operate with approximately the same fuel lifetime as the current high enriched uranium fuel and stay within the thermal and safety limits for the facility. It was also determined that the control rod worths and the temperature coefficient of reactivity would provide sufficient negative reactivity to control the reactor during the fuel’s complete lifetime. An assessment of the fuel’s viability for use with the Advanced Fuel Cycle Initiative’s Reactor Accelerator Coupling Experiments program was also performed. The objective of this study was to confirm the continued viability of these experiments with the reactor operating using this new fuel. For these experiments, the accelerator driven system must produce fission heating in excess of 1 kW when driven by a 20 kW accelerator system. This criterion was met using the new fuel. Therefore the change out of the fuel will not affect the viability of these experiments.

Thesis (M.S.)--Oregon State University, 2008. / Printout. Includes bibliographical references (leaves 77-83). Also available on the World Wide Web.

Throughout the history of the Annular Core Research Reactor (ACRR), Transient Rod (TR) A has experienced an increased rate of failure versus the other two TRs (B and C). Either by pneumatic force or electric motor, the transient rods remove the poison rods from the ACRR core allowing for the irradiation of experiments. In order to develop causes for why TR A is failing (rod break) more often, a better understanding of the whole TR system and its components is needed. This study aims to provide a foundational understanding of how the TR pneumatic system affects the motion of the TRs and the resulting effects that the TR motion has on the neutronics of the ACRR. Transient rod motion profiles have been generated using both experimentally-obtained pressure data and by thermodynamic theory, and input into Razorback, a SNL-developed point kinetics and thermal hydraulics code, to determine the effects that TR timing and pneumatic pressure have on reactivity addition and reactivity feedback. From this study, accurate and precise TR motion profiles have been developed, along with an increased understanding of the pulse timing sequence. With this information, a safety limit within the ACRR was verified for different TR travel lengths and pneumatic system pressures. In addition, longer reactivity addition times have been correlated to cause larger amounts of reactivity feedback. The added clarity on TR motion and timing from this study will pave the way for further study to determine the cause for the increased failure rate of TR A.

Master of Science / Department of Mechanical and Nuclear Engineering / Jeremy Roberts / The reactivity method is an indirect nondestructive technique to estimate integral burnup in fuel elements. In this method, the assumption is made that reactivity worth of a fuel element is a known function of burnup, often a linear relationship. When a fuel element burns, reactivity is reduced due to depletion of fissile actinides and generation of neutron-absorbing fission products. Currently, there is a lack of experimental data to verify the current composition of the KSU TRIGA (Training Research Isotopes General Atomics) fuel. Moreover, the KSU TRIGA Mark II staff method of estimating burnup is admittedly inaccurate due to its simple approximations. This work presents the positive period technique as convenient method use only the excess reactivity of the KSU core to compute reactivity via the inhour equation. Period measurements are determined via extraction and manipulation of the time dependent power data in the measurements. MCNP and Serpent modeling codes are both used extract the neutron kinetics parameters necessary in the inhour equation. Seven axial discretization of the KSU fuel was modeled, which minimizes the reactivity biases as function of burnup. Moreover, two unit cell models of the KSU TRIGA fuel were investigated. Modeled reactivity worths were computed using the KCODE in MCNP for comparative analysis. The burnup steps using two power peaking factor methods were developed to account for the biases introduced initial burnup of fuel prior to installation at KSU. By using the error distribution given by the two method to generate 200 test cases of the burnup steps can yield to reactivity worths as a function of burnup with quantifiable uncertainties. Finally, the results suggest that validation from another nondestructive technique such as gamma spectroscopy is necessary to asses the reactivity biases observed for higher burnup fuel elements due to unknown radial orientations. This work ultimately supports the production of a high-fidelity model of the KSU reactor.

The effects of neutron displacement damage on materials are sensitive to neutron energy spectra. In controlled neutron damage experiments, a well characterized neutron flux spectrum is critical in determining the equivalent dose for displacement damage. Two techniques were used to characterize the neutron flux spectra in the University of Texas at Austin TRIGA research nuclear reactor. The first technique uses a standard method of measuring the reaction rates of two identical metal foils, one of which was irradiated in a Cd cover, the other of which was irradiated bare. Assuming an analytic form of the neutron spectrum the reaction rates were used to determine an approximate spectrum. The second technique uses the reaction rates measured from a set of activated metal foils along with two spectral unfolding techniques to approximate and then refine the neutron spectrum. A Matlab code was developed which fits radiative capture reaction rates to an approximate spectrum using a least squares approach. The result was used as an initial guess in a second Matlab code which refines the epithermal and fast energy ranges of the spectrum using reaction rates from threshold reactions. Errors in the reaction rates calculated from the resulting spectrum to the measured reaction rates were used to assess the accuracy of the final neutron spectrum. / text

Oregon State University is home to a TRIGA® Mark II reactor. In October of 2008, the reactor began operating on low enriched uranium fuel. A model of the facility exists in MCNP, a Monte Carlo code that can be used for criticality calculations. Until now, a bias in the calculation of the neutron multiplication factor has been carried forward from outdated core models. This work involves updating various aspects of the model, including the geometry of the facility as well as materials and their properties, in order to arrive at a more accurate representation of the facility as it is today. The individual effect that each change has on the results of MCNP calculations of the core is documented. Following the updates to the model, the model can emulate records that describe the startup of the reactor in October of 2008. The results of these calculations can be compared to actual data in order to establish a foundation for benchmarking the model and characterizing the reactor core. The deviation between calculated and expected results can be used to determine a single reactivity bias in the model. The bias determined as a result of this work can be applied to future calculations using the model developed as a part of this work. / Graduation date: 2013

In 2008, the Oregon State TRIGA�� Reactor (OSTR) was converted from highly-enriched uranium (HEU) fuel lifetime improvement plan (FLIP) fuel to low-enriched uranium (LEU) fuel. This effort was driven and supported by the Department of Energy's (DoE's) Reduced Enrichment for Research and Test Reactors (RERTR) program. The basis behind the RERTR program's ongoing conversion effort is to reduce the nuclear proliferation risk of civilian research and test reactors. The original intent of the HEU FLIP fuel was to provide fuel to research reactors that could be utilized for many years before a necessary refueling cycle. As a research reactor, the OSTR provides irradiation facilities for a variety of applications, such as: activation analysis, fission-track dating, commercial isotope production, neutron radiography, prompt gamma characterization, and many others. In order to accurately perform these research functions, several studies have been conducted on the HEU FLIP fuel core to characterize the neutron spectra in various experimental facilities of the OSTR. As useful as these analyses were, they are no longer valid due to the change in fuel composition and the resulting alteration of core performance characteristics. The purpose of this study is to characterize the neutron spectra in various experimental facilities within the new LEU core so as to provide data that is representative of the OSTR's current state. / Graduation date: 2013

Nenhuma / The IPR-R1 TRIGA Mark I nuclear research reactor, at the Nuclear Technology Development Center (CDTN), is a pool type reactor cooling by light water. TRIGA reactors (Training, Research, Isotope, General Atomics) were designed for research, training and radioisotope production. The International Atomic Energy Agency (IAEA) recommends the use of safety and friendly interfaces for monitoring and controlling the operational parameters of the nuclear reactors. In this context, a system to simulate the neutron evolution flux of the TRIGA IPR-R1 reactor (TRIGA Simulator System - Sistema Simulador TRIGA) was developed using the LabVIEW software, considering the modern concept of virtual instruments (VIs) using electronic processor and visual interface in video monitor, with the objective of assisting the reactor operator training, allowing to study, to observe, and to analyze the behavior, and the tendency of some processes occurring in the reactor. Consequently the reactor operation parameters can be simulated and their relations can be visualized, supporting on the understanding of the interrelation of these parameters and their behavior, promoting a better knowledge of TRIGA IPR-R1 reactor processes. Some scenarios are presented to demonstrate that it is possible to use predetermined values in any parameters to verify its effect in the other ones. Therefore the TRIGA Simulator System (Sistema Simulador TRIGA) will allow the study of parameters, which affect the reactor operation, without the necessity of using the facility, avoiding risk, and reducing costs and operation time. / O reator nuclear de pesquisa TRIGA IPR-R1 Mark I, do Centro de Desenvolvimento da Tecnologia Nuclear (CDTN), é um reator do tipo piscina refrigerado à água leve. Os reatores TRIGA (Training, Research, Isotope, General Atomics) foram projetados para pesquisa, treinamento e produção de radioisótopos. A Agência Internacional de Energia Atômica (AIEA) recomenda o uso de interfaces amigáveis e seguras para o monitoramento e controle dos parâmetros operacionais dos reatores nucleares. Inserido neste contexto, um sistema para simulação da evolução do fluxo de nêutrons do reator nuclear de pesquisa TRIGA IPR-R1 (Sistema Simulador TRIGA) foi desenvolvido, utilizando o software LabVIEW, considerando o moderno conceito de instrumentos virtuais (VIs) por meio de processador eletrônico e interface visual em monitor de vídeo, cujo objetivo é auxiliar no treinamento de operadores de reatores, permitindo estudar, observar e analisar o comportamento e a tendência de alguns dos processos que acontecem em um reator. Deste modo, os parâmetros de operação do reator podem ser simulados e seus relacionamentos visualizados, auxiliando no entendimento de como estas variáveis estão interligadas e se comportam, promovendo melhor conhecimento dos processos do reator TRIGA IPR-R1. São apresentados cenários de utilização do Sistema, demonstrando que se podem usar valores determinados em qualquer um dos parâmetros, verificando seu efeito nos demais. Portanto o Sistema Simulador TRIGA possibilitará o estudo de parâmetros que afetam a operação do reator, sem a necessidade de usar a instalação, evitando riscos e minimizando custos e tempo de operação

The National Bureau of Standards Reactor (NBSR) located on the National Institute of Standards and Technology (NIST) Gaithersburg campus, is currently underway of fuel conversion from high enriched uranium (HEU) fuel to low enriched uranium (LEU) fuel. One particular challenging part of the conversion of the NBSR is the high average flux level (2.5×1014 n/cm2-s) required to maintain experimental testing capabilities of the reactor, without significant changes to the external structures of the reactor. Recently the General Atomics (GA) Training Research Isotopes General Atomics (TRIGA) fuel has shown some promising features as a LEU candidate for the high performance research reactors such as the NBSR. The GA fuel has a long history of success in conversion of research reactors since it was developed in 1980s. The UZrH compound in the GA fuel has seen success in long term TRIGA reactors, and is a proven safe LEU alternative. This study performs a neutronics evaluation of the TRIGA fuel under the schema of the NBSR’s heavy conversion requirements in order to examine whether the TRIGA fuel is a viable option for conversion of the NBSR. To determine the most optimal path of conversion, we performed a feasibility study with particular regard to the fuel dimensions, fuel rod configurations, cladding, as well as fuel structure selection. Based on the outcome of the feasibility study, an equilibrium core is then generated following the NBSR’s current fuel management schema. Key neutronics performance characteristics including flux distribution, power distribution, control rod (i.e., shim arms) worth, as well as kinetics parameters of the equilibrium core are calculated and evaluated. MCNP6, a Monte Carlo based computational modeling software was intensively used to aid in these calculations. The results of this study will provide important insight on the effectiveness of conversion, as well as determine the viability of the conversion from HEU to LEU using the GA fuel.

Thesis (Ph. D.)--University of Texas at Austin, 2008. / Vita. Includes bibliographical references.