Thesis (Nucl.E.)--Massachusetts Institute of Technology, Dept. of Nuclear Engineering, 1981. / MICROFICHE COPY AVAILABLE IN ARCHIVES AND SCIENCE. / Includes bibliographical references. / by John Edward Rivera. / Nucl.E.

Thesis. 1980. Ph.D.--Massachusetts Institute of Technology. Dept. of Nuclear Engineering. / MICROFICHE COPY AVAILABLE IN ARCHIVES AND SCIENCE. / Bibliography : leaves 138-146. / Ph.D.

Thesis. 1978. Nucl.E.--Massachusetts Institute of Technology. Dept. of Nuclear Engineering. / MICROFICHE COPY AVAILABLE IN ARCHIVES AND SCIENCE. / Includes bibliographical references. / Nucl.E.

The fluoride-salt-cooled high-temperature reactor (FHR) is a novel reactor design benefitting from passive safety features, high operating temperatures with corresponding high conversion efficiency, to name a few key features. The fuel is a layered graphite plank configuration containing enriched uranium oxycarbide (UCO) tri-structural isotropic (TRISO) fuel particles. Fuel cycle cost (FCC) models have been used to analyze and optimize fuel plate thicknesses, enrichment, and packing fraction as well as to gauge the economic competitiveness of this reactor design. Since the development of the initial FCC model, many corrections and modifications have been identified that will make the model more accurate. These modifications relate to corrections made to the neutronic simulations and the need for a more accurate fabrication costs estimate. The former pertains to a MC Dancoff factor that corrects for fuel particle neutron shadowing that occurs for double-heterogeneous fuels in multi-group calculations. The latter involves a detailed look at the fuel fabrication process to properly account for material, manufacturing, and quality assurance cost components and how they relate to the heavy metal loading in a FHR fuel plank. It was found that the fabrication cost may be a more significant portion of the total FCC than was initially attributed. TRISO manufacturing cost and heavy metal loading via packing fraction were key factors in total fabrication cost. This study evaluated how much neutronic and fabrication cost corrections can change the FCC model, optimum fuel element parameters, and the economic feasibility of the reactor design.

Investigation of nitride fuels and their properties has, for decades, been propelled on the basis of their desirable high metal densities and high thermal conductivities, both of which oer intrinsic advantages to performance, economy, and safety in fast and light water reactor systems. In this time several key obstacles have been identied as impeding the implementation of these fuels for commercial applications; namely chemical interactions with air and steam, the noted diculty in sintering of the material, and the high costs associated with the enrichment of 15N. The combination of these limitations, historically, led to the well founded conclusion that the most appropriate use of nitride fuels was in the fast reactor fuel cycle, where the cost burdens associated with them is substantially less. Indeed, it is within this context that the vast majority of work on nitrides has been and continues to be done. Nevertheless, following the 2011 Fukushima-Daiichi nuclear accident, a concerted governmental-industrial eort was embarked upon to explore the alternatives of so-called \accident tolerant" and \high performance" fuels. These fuels would, at the same time, improve the response of the fuel-clad system to severe accidents and improve the economy of operation for light water reactor systems. Among the various candidates proposed are uranium nitride, uranium silicide, and a third \uranium nitride-silicide" composite featuring a mixture of the former. In this thesis a method has been established for the synthesis, fabrication, and characterization of high purity uranium nitride, and uranium nitride-silicide composites, prepared by the spark plasma sintering (SPS) technique. A specic result has been to isolate the impact of the processing parameters on the microstructure of representative fuel pellets, essentially permitting any conceivable microstructure of interest to be fabricated. This has enabled the development of a highly reproducible technique for the production of pellets with microstructures tailored towards any desired porosity between 88-99.9%TD, any grain size between 6-24 μm, and, in the case of  the uranium nitride-silicide composite, a silicide-coated UN matrix. This has permitted the evaluation of these microstructural characteristics on the performance of these materials, specically with respect to their role as accident tolerant fuels. This has generated results which have tightly coupled nitride performance with pellet microstructure, with important implications for the use of nitrides in water-cooled reactors. / Under artionden har forskning om nitridbranseln och dess egenskaper bedrivits pa grundval av nitridbransletsatravarda egenskaper avseende dess hoga metall tathet och hog varmeledningsformaga. Dessa egenskaper besitter vasentliga fordelar avseende prestanda, ekonomi och sakerhet for metallkylda som lattvatten reaktorer. Genom forskning har aven centrala begr ansningar identierats for implementering av nitridbranslen for kommersiellt bruk. Begransningar avser den kemiska interaktionen med luft och vattenanga, en uppmarksammad svarighet att sintring av materialet samt hoga kostnader forknippade med den nodvandiga anrikningen av 15-N. Kombinationen av dessa begransningar resulterade, tidigare, i en valgrundad slutsats att nitridbranslet mest andamalsenliga anvandningsomrade var i karnbranslecykeln for snabba reaktorer. Detta da kostnaderna forenade med implementeringen av branslet ar avsevart lagre. Inom detta sammanhang har majoriteten av forskning avseende nitrider bedrivits och fortskrider an idag. Dock, efter karnkraftsolyckan i Fukushima-Daiichi 2011, inleddes en samlad industriell och statlig anstrangning for att undersoka alternativ till sa kallade \olyckstoleranta" och \hogpresterande" branslen. Dessa branslen skulle samtidigt forbattra reaktionstiden for bransleinkapsling systemet mot allvarliga olyckor samt forbattra driftsekonomin av lattvattenreaktorer. Foreslagna kandidater ar urannitrid, uransilicid och en tredje \uran nitrid-silicid", komposit bestaende av en blandning av de foregaende. Genom denna avhandling har en metod faststallts for syntes, tillverkning och karaktarisering av uran nitrid av hog renhet samt uran nitrid-silicid kompositer, forberedda med tekniken SPS (Spark Plasma Sintering). Ett specikt resultat har varit att isolera eekten av processparametrar pa mikrostrukturen pa representativa branslekutsar. Detta mojliggor, i princip, framstallningen av alla tankbara mikrostrukturer utav intresse for tillverkning. Vidare har detta mojliggjort utvecklingen av en hogeligen reproducerbar  teknik for framstallningen av branslekutsar med mikrostrukturer skraddarsydda for onskad porositet mellan 88 och 99.9 % TD, och kornstorlek mellan 6 och 24 μm. Dartill har en metod for att belagga en matris av uran nitrid-silicid framarbetats. Detta har mojliggjort utvarderingen av dessa mikrostrukturella parametrars paverkan pa materialens prestanda, sarskilt avseende dess roll som olyckstoleranta branslen. Detta har genererat resultat som ar tatt sammanlankat nitridbranslets prestanda till kutsens mikrostruktur, med viktiga konsekvenser for den potentiella anvandningen av nitrider i lattvatten reaktorer. / <p>QC 20170210</p>

The Digital Cherenkov Viewing Device (DCVD) is an instrument used by authority inspectors to assess irradiated nuclear fuel assemblies in wet storage for the purpose of nuclear safeguards. Originally developed to verify the presence of fuel assemblies with long cooling times and low burnup, the DCVD accuracy is sufficient for partial defect verification, where one verifies that part of an assembly has not been diverted. Much of the recent research regarding the DCVD has been focused on improving its partial defect detection capabilities. The partial-defect analysis procedure currently used relies on comparisons between a predicted Cherenkov light intensity and the intensity measured with the DCVD. Enhanced prediction capabilities may thus lead to enhanced verification capabilities. Since the currently used prediction model is based on rudimentary correlations between the Cherenkov light intensity and the burnup and cooling time of the fuel assembly, there are reasons to develop alternative models taking more details into account to more accurately predict the Cherenkov light intensity. This work aims at increasing our understanding of the physical processes leading to the Cherenkov light production in irradiated nuclear fuel assemblies in water. This has been investigated through simulations, which in the future are planned to be complemented with measurements. The simulations performed reveal that the Cherenkov light production depends on fuel rod dimensions, source distribution in the rod and initial decay energy in a complex way, and that all these factors should be modelled to accurately predict the light intensity. The simulations also reveal that for long-cooled fuel, Y-90 beta-decays may contribute noticeably to the Cherenkov light intensity, a contribution which has not been considered before. A prediction model has been developed in this work taking fuel irradiation history, fuel geometry and Y-90 beta-decay into account. These predictions are more detailed than the predictions based on the currently used prediction model. The predictions with the new model can be done quickly enough that the method can be used in the field. The new model has been used during one verification campaign, and showed superior performance to the currently used prediction model. Using the currently used model for this verification, the difference between measured and predicted intensity had a standard deviation of 15.4% of the measured value, and using the new model this was reduced to 8.4%.

Este trabalho tem como objetivo estudar as modificações introduzidas, ao longo de sucessivas versões, nos modelos empíricos do programa computacional FRAPCON utilizado para a simulação do comportamento sob irradiação de varetas combustíveis de Reatores a Água Leve Pressurizada (Pressurized Water Reactor - PWR) em regime de estado estacionário e sob condições de alta queima. No estudo, foram analisados os modelos empíricos utilizados pelo FRAPCON e que são apresentados em sua documentação oficial. Um estudo bibliográfico foi conduzido sobre os efeitos da alta queima em combustíveis nucleares visando melhorar o entendimento dos modelos utilizados pelo FRAPCON nestas condições. Foram feitas simulações do comportamento sob irradiação de uma vareta combustível típica de um reator PWR utilizando as versões 3.3, 3.4 e 3.5 do FRAPCON. Os resultados apresentados pelas diferentes versões do programa foram comparados entre si de forma a verificar as consequências das mudanças de modelo nos parâmetros de saída do programa. Foi possível observar que as modificações introduzidas trouxeram diferenças significativas nos resultados de parâmetros térmicos e mecânicos da vareta combustível, principalmente quando se evoluiu da versão FRAPCON-3.3 para a versão FRAPCON-3.5. Nessa ultima versão, obteve-se menores temperaturas na vareta combustível, menores tensões e deformações no revestimento, menor espessura da camada de oxido formada no revestimento a altas queimas na vareta combustível. / This work studies the modifications implemented over successive versions in the empirical models of the computer program FRAPCON used to simulate the steady state irradiation performance of Pressurized Water Reactor (PWR) fuel rods under high burnup condition. In the study, the empirical models present in FRAPCON official documentation were analyzed. A literature study was conducted on the effects of high burnup in nuclear fuels and to improve the understanding of the models used by FRAPCON program in these conditions. A steady state fuel performance analysis was conducted for a typical PWR fuel rod using FRAPCON program versions 3.3, 3.4, and 3.5. The results presented by the different versions of the program were compared in order to verify the impact of model changes in the output parameters of the program. It was observed that the changes brought significant differences in the results of the fuel rod thermal and mechanical parameters, especially when they evolved from FRAPCON-3.3 version to FRAPCON-3.5 version. Lower temperatures, lower cladding stress and strain, lower cladding oxide layer thickness were obtained in the fuel rod analyzed with the FRAPCON-3.5 version.

Graduation date: 1994

At head of title: Energy Laboratory and Dept. of Nuclear Engineering. / Sponsored by Boston Edison Company and others under MIT Energy Laboratory Electric Utility Program.

A process for recycling spent nuclear fuel cladding, a zirconium alloy (Zircaloy), into a metal powder that may be used for advanced nuclear fuel applications, was investigated to determine if it is a viable strategy. The process begins with hydriding the Zircaloy cladding hulls after the spent nuclear fuel has been dissolved from the cladding. The addition of hydrogen atoms to the zirconium matrix stresses the lattice and forms brittle zirconium hydride, which is easily pulverized into a powder. The dehydriding process removes hydrogen by heating the powder in a vacuum, resulting in a zirconium metal powder. The two main objectives of this research are to investigate the dehydriding process and to design, build and demonstrate a specialized piece of equipment to process the zirconium from cladding hulls to metal powder without intermediate handling. The hydriding process (known from literature) took place in a 95 percent argon - 5 percent hydrogen atmosphere at 500 degrees C while the dehydriding process conditions were researched with a Thermogavimetric Analyzer (TGA). Data from the TGA showed the dehydriding process requires vacuum conditions (~0.001 bar) and 800 degrees C environment to decompose the zirconium hydride. Zirconium metal powder was created in two separate experiments with different milling times, 45 minutes (coarse powder) and 12 hours (fine powder). Both powders were analyzed by three separate analytical methods, X-Ray Diffraction (XRD), size characterization and digital micrographs. XRD analysis proved that the process produced a zirconium metal. Additionally, visual observations of the samples silvery color confirmed the presence of zirconium metal. The presence on zirconium metal in the two samples confirmed the operation of the hydriding / milling / hydriding machine. Further refining of the hydride / milling / dehydride machine could make this process commercially favorable when compared to the high cost of storing nuclear waste and its components. An additional important point is that this process can easily be used on other metals that are subject to hydrogen embrittlement, knowing the relevant temperatures and pressures associated with the hydriding / dehydriding of that particular metal.