Gamma spectroscopy and gamma emission tomography are two non-destructive measurement techniques for assessing the performance of nuclear fuel which have been investigated in this thesis for existing and novel applications through theoretical studies and experimental demonstrations. For assessment of individual fuel rods using gamma spectroscopy, fuel assemblies are dismantled so that the fuel rods may be measured separately, which is time-consuming and may cause damage to the fuel. Gamma tomography is more seldom used, but its application on complete fuel assemblies would enable the assessment of individual fuel rods without the need to disassemble the fuel. Both techniques are based on recording gamma rays, emitted at characteristic energies from decaying radioactive products in the fuel. The feasibility of measuring short-lived fission gasses in the gas plenum of fuel rods with short cooling time was experimentally investigated. Based on the feasibility demonstration, a method was proposed and experimentally demonstrated for determining the fission gas release fraction of 133Xe in fuel rods with short cooling time. Additionally, a method for investigating the origin of released fission gasses based on the measured ratio of 133Xe/85Kr in the fuel rod gas plenum was demonstrated. These methods may be employed at research reactors, where fuel with short cooling time is available for measurement. A gamma emission tomography instrument has been designed, constructed and experimentally demonstrated on a Halden Reactor fuel assembly. Simulation studies showed that the instrument and the tomographic reconstruction methods employed may be useful for: identifying a leaking fuel rod in an assembly by its lack of fission gas content; reconstruction of the rod-wise fission product distributions in the fuel stack and plenum regions of the assembly; and determining the rod-wise fission gas release fractions. In the experimental demonstration, the rod-wise distributions of the fission products 137Cs and 85Kr in the fuel stack and plenum regions of the assembly were reconstructed, as well as the distributions of the activation products 60Co and 178mHf in the plenum region, revealing the plenum springs and tie rods, respectively. The reconstructed data was in the form of images, useful for qualitative assessment of the fuel.

The Fukushima Daiichi nuclear accidents in 2011 sparked considerable interest in the U.S. to develop new nuclear fuel with enhanced accident tolerance. Throughout the development of these new fuel concepts they will be extensively modeled using specialized computer codes and experimentally tested for a variety of different postulated accident scenarios. One accident scenario of interest, reactivity-initiated accident, is a nuclear reactor event involving a sudden increase in fission rate that causes a rapid increase in reactor power and temperature of the fuel which can lead to the failure of the fuel rods and are lease of radioactive material. The focus of this work will be on the fuel performance modeling of reactivity-initiated accidents using the BISON code being developed at Idaho National Laboratory. The overall goal of this work is to provide the best possible modeling predictions for future experimental tests. Accurate predictive capability modeling using BISON is important for safe operation of these tests and provides a cheaper alternative to the expensive experiments.

<p dir="ltr">In the last few decades, the US Department of Energy established the Generation-IV Initiative to advance the design of nuclear energy systems with a focus on fast nuclear reactors. Special interest has been placed on Sodium-cooled Fast Reactors (SFRs) along with metallic fuels. SFRs are attractive because of their ability to utilize fast neutrons effectively, allowing for efficient transmutation of long-lived radioactive isotopes and a more complete use of fissile material. This capability significantly reduces nuclear waste and improves fuel sustainability compared to other Generation IV reactors. The metallic fuels, such as U-Zr and U-Pu-Zr, are attractive because of their higher thermal conductivity and higher density of fissile material, leading to improved breeding ratios and higher burnup rates. Through years of testing, SFRs, such as the EBR-II, could achieve a very high burnup (up to 750 GWD/tU) while traditional generation I to III+ reactors achieve around 30 GWD/tU. Since fast reactors, particularly SFRs, operate on a hard neutron spectrum, they utilize different geometries and cooling materials. This requires the use of different mechanistic models in nuclear codes to accurately capture the underlying physics. The NRC has recently shown interest in upgrading its diffusion codes to support the integration of SFRs. They have shown additional interest in improving the simulation capability of SFR metallic fuel, specifically U-10wt%Zr, even for high fuel burnups in excess of 10 at.%. </p><p dir="ltr">The purpose of this thesis is multifaceted. It serves to develop new mechanistic modelsto support the modeling and analysis of SFRs in existing nodal codes; it also serves to advance the current understanding of the principal effects of U-Zr fuel behavior during steady-state conditions. From the neutronics perspective, a new nodal method will be developed within the PARCS nodal code, along with generalized concepts such as reactivity feedback coefficients and thermal expansion, which will then be validated against the EBRII steady-state benchmark. From the fuel behavior perspective, mechanistic models that describe fuel redistribution, temperature distribution, fission gas, mechanical stresses, and point defect generation will be developed into the Purdue Fuel Performance (PFP) code for U-Zr fuel. The code will then be validated against the radial fuel concentration profile of two EBR-II U-Zr fuel pins.</p>

The swelling mechanisms of U3Si2 under neutron irradiation in reactor conditions are not unequivocally known. The limited experimental evidence that is available suggests that the main driver of the swelling in this material would be fission gases accumulation at crystalline grain boundaries. The steps that lead to the accumulation of fission gases at these locations are multiple and complex. However, gradually, the gaseous fission products migrate by diffusion. Upon reaching a grain boundary, which acts as a trap, the gaseous fission products start to accumulate, thus leading to formation of bubbles and hence to swelling. Therefore, a quantitative model of swelling requires the incorporation of phenomena that increase the presence of grain boundaries and decrease grain sizes, thus creating sites for bubble formation and growth. It is assumed that grain boundary formation results from the conversion of stored energy from accumulated dislocations into energy for the formation of new grain boundaries.This thesis attempts to develop a quantitative model for grain subdivision in U3Si2 based on the above mentioned phenomena to verify the presence of this mechanism and to use in conjunction with swelling codes to evaluate the total swelling of the pellet in the reactor during its lifetime.

With the initiative taken by the SUNRISE project (Sustainable Nuclear Energy Research in Sweden) to construct a Lead-cooled research reactor, this thesis intends to extend the knowledge within nuclear fuel development. By using integral iterative modelling and simulating techniques that mimic real-world phenomena, novel fuel materials like uranium nitride are assessed for future validation.  The work deals with the fuel performance analysis of the SUNRISE LFR, employing the TRANSURANUS fuel performance code. This code contains a collection of model parameters that simulate the thermo-mechanical behaviour of the fuel cladding system on an engineering scale of the reactor core. A comparative study is performed for UO$_2$ and UN fuels using the same input data such as fuel geometry. In addition, predefined information relating to the neutronics analysis for the reactor was used as input to the TRANSURANUS code along with literature reviews to select the accurate models on the reactor, fuel, and its behaviour. Furthermore, a sensitivity study is carried out to assess the models and parameters affected by more significant uncertainty.  The uncertainty analysis of the UN fuel's swelling models is performed using the Dakota toolkit. The sampling of input data using the Dakota software coupled with the nuclear simulation program TRANSURANUS produced partial rank correlation coefficients significant to the modelling. However, since the assessed models displayed the same correlation coefficients, the results conclude that a deeper understanding of the theoretical swelling model might be required. / I samverkan med initiativet av SUNRISEprojektet (Sustainable Nuclear Energy Research inSweden) som syftar att bygga en blykyld forskningsreaktor, avser denna avhandling att utökakunskapen inom kärnbränsleutveckling. Med användning av integral iterativ modellering ochsimuleringstekniker som efterliknar verkliga fenomen bedöms nya bränslematerial somuranmononitrid för framtida validering. Arbetet behandlar analysen av bränsleprestanda för SUNRISE LFR, med användning avTRANSURANUS bränsleprestandakod. Denna kod innehåller en samling modellparametrarsom simulerar det termomekaniska beteendet hos bränslebetäckningssystemet i en tekniskskala för reaktorkärnan. En jämförande studie utförs för UO2 och UN-bränslen med sammaingångsdata som t.ex bränslegeometrin. Dessutom användes fördefinierad information om denneutroniska analysen för reaktorn som ingångsdata till TRANSURANUSkoden tillsammans medgranskning av litteratur för att välja lämpliga modeller för reaktorn, bränslet och dess beteende.Därtill genomfördes en känslighetsstudie för att bedöma de modeller och parametrar sompåverkas av mer betydande osäkerhet. Osäkerhetsanalysen av UN-bränslets svällningsmodeller utförs med hjälp av Dakota-verktyget.Samlingen av indata med Dakota-programmet i kombination medkärnkraftssimuleringsprogrammet TRANSURANUS gav korrelationskoefficienter för partiell rangviktiga för modelleringen. Eftersom de utvärderade modellerna visade sammakorrelationskoefficienter, tyder slutsatsen på att en djupare förståelse av den teoretiskasvällningsmodellen krävs

Uranium-zirconium (U-Zr) alloys comprise a class of metallic nuclear fuel that is regularly considered for application in fast nuclear energy systems. The U-10wt%Zr alloy has been demonstrated to very high burnup without cladding breach in the Experimental Breeder Reactor-II (EBR-II). This was accomplished by successfully accommodating gaseous fission products with low smear density fuel and an enlarged cladding plenum. Fission gas swelling behavior of the fuel has been experimentally revealed to be significantly affected by the temperature gradient within a fuel pin and the multiple phase morphologies that exist across the fuel pin. However, the phase effects on swelling behavior have not been yet fully accounted for in existing fuel performance models which tend to assume the fuel exists as a homogeneous single phase medium across the entire fuel pin. Phase effects on gas bubble nucleation and growth in the alloy were investigated using transmission electron microscopy (TEM). To achieve this end, a comprehensive examination of the alloy system was carried out. This included the fabrication of uranium alloys containing 0.1, 2, 5, 10, 20, 30, 40, and 50 wt% zirconium by melt-casting. These alloys were characterized using electron probe micro-analysis (EPMA), differential scanning calorimetry (DSC), and thermogravimetric analysis (TGA). Once the alloys were satisfactorily characterized, selected U-Zr alloys were irradiated with 140 keV He^(+) ions at fluences ranging from 1 × 10^(14) to 5 × 10^(16) ions/cm^(2). Metallographic and micro-chemical analysis of the alloys indicated that annealing at 600 °C equilibrates the alloys within 168 h to have stable α-U and δ-UZr_(2) phase morphologies. This was in contrast to some reported data that showed kinetically sluggish δ-UZr_(2) phase formation. Phase transformation temperatures and enthalpies were measured using DSC-TGA for each of the alloys. Measured temperatures from different time annealed alloys have shown consistent matches with most of the features in the current U-Zr phase diagram which further augmented the EPMA observed microstructural equilibrium. Nevertheless, quantitative transformation enthalpy analysis also suggests potential errors in the existing U-Zr binary phase diagram. More specifically, the (β-U, γ2) phase region does not appear to be present in Zr-rich (> 15 wt%) U-Zr alloys and so further investigation may be required. To prepare TEM specimens, characterized U-Zr alloys were mechanically thinned to a thickness of ~150 μm, and then electropolished using a 5% perchloric acid/95% methanol electrolyte. Uranium-rich phase was preferentially thinned in two phase alloys, giving saw-tooth shaped perforated boundaries; the alloy images were very clear and alloy characterization was accomplished. During in-situ heating U-10Zr and U-20Zr alloys up to 810 °C, selected area diffraction (SAD) patterns were observed as the structure evolved up to ~690 °C and the expected α-U → β-U phase transformation at 662 °C was never observed. For the temperature range of the (α-U, γ2) phase region, phase transformation driven diffusion was observed as uranium moved into Zr-rich phase matrix in U-20Zr alloy; this was noted as nonuniform bridging of adjacent phase lamellae in the alloy. From the irradiation tests, nano-scale voids were discovered to be evenly distributed over several micrometers in U-40Zr alloys. For the alloys irradiated at the fluences of 1 × 10^(16) and 5 × 10^(16) ions/cm^(2), estimated void densities were proportional to the irradiation doses, (250 ± 40) and (1460 ± 30) /μm^(2), while void sizes were fairly constant, (6.0 ± 1.5) and (5.2 ± 1.2) nm, respectively. Measured data could be foundational inputs to the further development of a semi-empirical metal fuel performance model.

The Multi-Application Small Light Water Reactor (MASLWR) is a small natural circulation pressurized light water reactor design that was developed by Oregon State University (OSU) and Idaho National Engineering and Environmental Laboratory (INEEL) under the Nuclear Energy Research Initiative (NERI) program to address the growing demand for energy and electricity. The MASLWR design is geared toward providing electricity to small communities in remote locations in developing countries where constructions of large nuclear power plants are not economical. The MASLWR reactor is designed to operate for five years without refueling and with fuel enrichment up to 8 %. In 2003, an experimental thermal hydraulic research facility also known as the OSU MASLWR Test Facility was constructed at Oregon State University to examined the performance of new reactor design and natural circulation reactor design concepts. This thesis is focused on the thermal hydraulics analysis and fuel performance analysis of the MASLWR prototypical cores with fuel enrichment of 4.25 % and 8 %. The goals of the thermal hydraulic analyses were to calculate the departure nucleate boiling ratio (DNBR) values, coolant temperature, cladding temperature and fuel temperature profiles in the hot channel of the reactor cores. The thermal hydraulic analysis was performed for steady state operation of the MASLWR prototypical cores. VIPRE Version 01 is the code used for all the computational modeling of the prototypical cores during thermal hydraulic analysis. The hot channel and hot rod results are compared with thermal design limits to determine the feasibility of the prototypical cores. The second level of analysis was performed with a fuel performance code FRAPCON for the limiting MASLWR fuel rods identified by the neutronic and thermal hydraulic analyses. The goals of the fuel performance analyses were to calculate the oxide thickness on the cladding and fission gas release (FGR). The oxide thickness results are compared with the acceptable design limits for standard fuel rods. The results in this research can be helpful for future core designs of small light water reactors with natural circulation. / Graduation date: 2012