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Nelson, Andrew T.

Publications and source records attributed to Nelson, Andrew T..

57 records · Page 4

Impact of uranium oxide (UO 2 ) fuel with molybdenum (Mo) inserts on pressurized water reactor performance and safety

This work investigates nuclear reactor performance and safety characteristics of UO 2 with high thermal conductivity Mo insert structures by using multiphysics modeling techniques. Additionally, the purpose of this study is to use scoping analyses to quantify the impact of using Mo inserts from neutronic and heat transfer standpoints. Attention is given to reactor performance parameters, such as cycle length, maximum fuel temperature, temperature gradients in the fuel, and stored energy in the fuel. The finite-element code BISON and the Monte Carlo particle transport code Serpent were used to perform sensitivity analyses on the Mo insert geometry to optimize the insert design and inform larger scale modeling that required the homogenization of the UO 2 and Mo. Although BISON is often used as a fuel performance analysis tool, it is used in this context for heat transfer analysis only. Fuel performance optimization is outside the scope of the current study, but would be important for future work focused on this concept. The results showed that the insert had little impact on neutronic performance and that homogenizing the UO 2 and Mo was acceptable for reactor physics calculations. Reactivity temperature coefficients calculated using homogeneous UO 2 -Mo were shown to be relatively similar to UO 2 , but higher Mo content and 235 U enrichment can reduce the worth of soluble boron and control rods. The effect of insert geometry on heat transfer was much greater, and an approximately 15–20% difference in maximum fuel temperature was predicted between the best and worst performing heat transfer geometries. Furthermore, thermal conductivity calibration based on the finite element analysis results was performed to improve the accuracy of temperature predictions in reactor analysis models that homogenized the UO 2 -Mo fuel. Compared with UO 2 in a pressurized water reactor (PWR), the optimized UO 2 -Mo design increased the margin to fuel melt by 13–32% across the fuel cycle, but it requires the 235U enrichment to exceed 5% to match the cycle length of conventional UO 2 .

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Hydration of α-UO 3 following storage under controlled conditions of temperature and relative humidity

Changes in chemical speciation of uranium oxides following storage under varied conditions of temperature and relative humidity are valuable for characterizing material provenance. In this study, subsamples of high purity α-UO 3 were stored under four sets of controlled conditions of temperature and relative humidity over several years, and then measured periodically for chemical speciation. Powder X-ray diffraction (XRD) analysis and extended X-ray absorption fine structure spectroscopy confirm hydration of α-UO 3 to a schoepite-like end product following storage under each of the varied storage conditions, but the species formed during exposure to the lower relative humidity and lower temperature condition follows different trends from those formed under the other three storage conditions (high relative humidity with high or low temperatures, and low relative humidity with a high temperature). Thermogravimetry coupled with XRD analysis was carried out to distinguish desorption pathways of water from the hydrated end products. Density functional theory calculations discern changes in the structure of α-UO 3 following incorporation of 1, 2 or 3 H 2 O molecules or 1, 2 or 3 OH groups into the orthorhombic lattice, revealing differences in lattice constants, U–O bond lengths, and U–U distances. The collective results from this analysis are in contrast to analogous studies that report that U 3 O 8 is oxidized and hydrated in air during storage under high relative humidity conditions.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

Accelerating nuclear fuel development and qualification: Modeling and simulation integrated with separate-effects testing

In this work, an approach to transform and accelerate nuclear fuel development and qualification is outlined. The approach exploits advanced modeling and simulation at the outset to inform constituent and system selection and to enable integral fuel performance analyses. Analyses using these tools identify and prioritize the most important fuel performance parameters and phenomena for subsequent targeted characterization with separate-effects tests. Separate-effects testing spans out-of-pile and in-pile tests and is meant to iterate with and inform engineering-scale integral fuel performance analyses throughout the development process. Exercising this cycle in an agile fashion will increase confidence in the integral fuel performance predictions while reducing uncertainties. This process sets the stage for executing a much more limited set of well-defined integral irradiation tests designed to validate engineering-scale fuel performance codes and to confirm the performance and safety of the fuel system under prototypic conditions. This approach will reduce the time for development and qualification of a new fuel system, and it will also reduce associated costs.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗