@article{toptan_salko_avramova_clarno_kropaczek_2019, title={A new fuel modeling capability, CTFFuel, with a case study on the fuel thermal conductivity degradation}, volume={341}, ISSN={["1872-759X"]}, DOI={10.1016/j.nucengdes.2018.11.010}, abstractNote={A new fuel modeling capability, CTFFuel, is developed from the subchannel code, CTF. This code is a standalone interface to the CTF fuel rod models, allowing for fuel rod simulations to be run independently from the fluid. This paper provides an overview of the code with a case study on the thermal conductivity degradation of LWR fuels to demonstrate its capabilities. The modeling of fuel thermal conductivity degradation in the code is improved through the addition of new modeling options to account for the irradiation effects via globally defined parameters. After the initial implementation, a variety of order-of-accuracy tests and code comparisons are performed to test software quality. A controlled analysis is allowed by CTFFuel to verify the numerical scheme of CTF’s conduction solution and to benchmark its fuel temperature predictions against FRAPCON-4.0’s. Overall, the software quality and verification procedure ensures that the new model is coded correctly, that it properly interacts with the rest of the code.}, journal={NUCLEAR ENGINEERING AND DESIGN}, author={Toptan, Aysenur and Salko, Robert K. and Avramova, Maria N. and Clarno, Kevin and Kropaczek, David J.}, year={2019}, month={Jan}, pages={248–258} }
 @article{toptan_kropaczek_avramova_2019, title={Gap conductance modeling I: Theoretical considerations for single- and multi-component gases in curvilinear coordinates}, volume={353}, ISSN={["1872-759X"]}, DOI={10.1016/j.nucengdes.2019.110283}, abstractNote={Abstract Accurate estimation of heat transfer across the gap is important in nuclear fuel performance because heat transfer across the fuel-to-cladding gap heavily impacts fuel temperatures and the thermo-mechanical performance of nuclear fuel rods. Better understood physics will allow a better prediction of the gap behavior. This paper focuses on providing an overview of the gap conductance model including theoretical considerations and underlying assumptions. The gap conductance is calculated considering three summed heat paths: fill gas conductance, direct thermal radiation, and solid contact conductance. Each heat transfer mechanism is described in detail. First, the models are generalized to curvilinear coordinates for diatomic/polyatomic molecules. Traditional models use one-dimensional Cartesian equations for a monatomic gas. Second, expressions for temperature jump distance and thermal accommodation coefficients are made consistent with the kinetic theory for both single- and multi-component gases. Lastly, fill gas thermal conductivity is updated to include its dependence on rod internal pressure.}, journal={NUCLEAR ENGINEERING AND DESIGN}, author={Toptan, Aysenur and Kropaczek, David J. and Avramova, Maria N.}, year={2019}, month={Nov} }
 @article{toptan_kropaczek_avramova_2019, title={Gap conductance modeling II: Optimized model for UO2-Zircaloy interfaces}, volume={355}, ISSN={["1872-759X"]}, DOI={10.1016/j.nucengdes.2019.110289}, abstractNote={The model conventionally used to calculate heat transfer across the fuel-cladding gap in light water nuclear reactors is a modified version of the Ross-Stoute model. The model was modified to include gap distance in the formulation, which introduced additional uncertainties because the model parameters were not adjusted after the modification. In this study, this conventional model is optimized for uranium dioxide-Zircaloy interfaces using experimental data at high pressure for single- and multi-component gases. First, a calibration is performed for single-component gases. Second, the calibration is extended to multi-component gases, which allows for a demonstration of sources of uncertainty in the model. Third, a general form of the gap conductance model is optimized by combining both data sets. Difficulties arise due to: (i) inaccurate estimation of contact characteristics (e.g., number of solid contacts, deformation mechanism of surface irregularities, contact shapes) that are different for each experimental setup; (ii) the non-physical ratio of temperature jump distance to the gap distance for postulated model function form; (iii) an insufficient description of the appropriate heat transfer regime; and (iv) the pressure dependence of thermal conductivity for inert gases aside from helium. Lastly, a general model is optimized by setting the temperature jump distance at the wall to zero, which reduces possible uncertainties. This final analysis results in a more accurate prediction of the available experimental data. The Associated parameter uncertainty of the model is estimated by performing uncertainty propagation. Overall, the optimized model results in a larger gap conductance with significantly reduced error.}, journal={NUCLEAR ENGINEERING AND DESIGN}, author={Toptan, Aysenur and Kropaczek, David J. and Avramova, Maria N.}, year={2019}, month={Dec} }
 @article{toptan_kropaczek_avramova_2019, title={On the validity of the dilute gas assumption for gap conductance calculations in nuclear fuel performance codes}, volume={350}, ISSN={["1872-759X"]}, DOI={10.1016/j.nucengdes.2019.04.042}, abstractNote={Fill gas thermal conductivity’s dependence on pressure is neglected in today’s nuclear fuel performance codes. Current codes assume that gas behaves as a dilute gas, but the pressure effect is more pronounced at temperatures lower than ten times the critical temperature of each pure gas. The validity of this assumption for nuclear fuel performance is examined herein. Theories related to dilute and dense gas properties are presented, along with their validation against literature data at up to 30 MPa for selected inert gases. Underlying assumptions are clearly described for each model, and their possible impacts on gap conductance calculations are discussed. The dilute gas assumption is valid for helium because it behaves as a dilute gas. However, the assumption is not valid in most gap conductance calculations when the gap is mostly occupied with either lower conductivity gaseous fission products or an initial fill gas other than helium.}, journal={NUCLEAR ENGINEERING AND DESIGN}, author={Toptan, Aysenur and Kropaczek, David J. and Avramova, Maria N.}, year={2019}, month={Aug}, pages={1–8} }
 @article{toptan_porter_salko_avramova_2018, title={Implementation and assessment of wall friction models for LWR core analysis}, volume={115}, ISSN={["0306-4549"]}, DOI={10.1016/j.anucene.2018.02.022}, abstractNote={The modeling of frictional pressure drop in the nuclear thermal hydraulics subchannel code, CTF, is improved through the addition of three new modeling options. Two of the new models allow the code to account for the effects of surface roughness and the third enables a user-supplied option. After the initial implementation, a variety of analyses are performed to test the software quality. First, a series of defect tests are designed for both single- and multi-channel configurations which compare simulated results to approximate solutions. The single-channel tests assess the friction model implementation; a suite of three-by-three bundle tests are used to ensure proper implementation of the roughness averaging scheme. The maximum relative error in the pressure drop over all defect tests is less than 0.15%. A solution verification test is performed to ensure that the first order numerical scheme in CTF is not significantly disrupted by the friction model. Finally, the wall friction model is validated using both separate and integral effects experimental data. Overall, the software quality, verification, and validation procedure ensures that the new model is coded correctly, that it properly interacts with the rest of CTF, and that it can be used to model real-world data for turbulent single-phase flow. The work completed herein provides a complete demonstration of modern coding practices. Future work could include a formal equation analysis of the numerical error in the friction model, as well as an analysis of validation data for one dimensional two-phase flow.}, journal={ANNALS OF NUCLEAR ENERGY}, author={Toptan, Aysenur and Porter, Nathan W. and Salko, Robert K. and Avramova, Maria N.}, year={2018}, month={May}, pages={565–572} }