@article{porter_mousseau_avramova_2019, title={CTF-R: A novel residual-based thermal hydraulic solver}, volume={348}, ISSN={["1872-759X"]}, DOI={10.1016/j.nucengdes.2019.04.006}, abstractNote={The traditional scientific process has been revolutionized by the advent of computational modeling, but the nuclear industry generally uses “legacy codes,” which were developed early in the evolution of computers. One example of a legacy code, the thermal hydraulic subchannel code CTF, is modernized in this work through the development of a novel residual-based version, CTF-R. Unlike its predecessor, CTF-R is not limited by the strict computational limitations of the early 1980’s, and can therefore be designed such that it is inherently flexible and easy to use. A case study is examined to demonstrate how the flexibility of the code can be used to improve simulation results. In this example, the coupling between the solid and liquid fields is examined. Traditionally, this coupling is modeled explicitly, which imposes numerical stability limits on the time step size. These limits are derived and it is shown that they are removed when the coupling is made implicit. Further, the development of CTF-R will enable future improvements in next generation reactor modeling, numerical methods, and coupling to other codes. Through the further development of CTF-R and other residual-based codes, state-of-the-art simulation is possible.}, journal={NUCLEAR ENGINEERING AND DESIGN}, author={Porter, N. W. and Mousseau, V. A. and Avramova, M. N.}, year={2019}, month={Jul}, pages={37–45} }
@article{porter_mousseau_avramova_2019, title={Quantified Validation with Uncertainty Analysis for Turbulent Single-Phase Friction Models}, volume={205}, ISSN={["1943-7471"]}, DOI={10.1080/00295450.2018.1548221}, abstractNote={Abstract This paper introduces a framework for model selection that includes parameter estimation, uncertainty propagation, and quantified validation. The framework is applied to single-phase turbulent friction modeling in CTF, which is a thermal-hydraulic code for nuclear engineering applications. The friction model is chosen because it is well understood and easy to separate from other physics, which allows focus to be on the model selection framework instead of on the particulars of the chosen model. Two different empirical models are compared: the McAdams Correlation and the Simplified McAdams Correlation. The parameter estimation is performed by calibrating each of the friction models to experimental data using the Delayed Rejection Adaptive Metropolis algorithm, which is a Markov Chain Monte Carlo method. State point uncertainties are also considered, which are determined based on measurement errors from the experiment. The input parameter distributions are propagated through CTF using a statistical method with samples. A variety of validation metrics is used to quantify which empirical model is more accurate. It is shown that model form uncertainty can be quantified using validation once all other sources of uncertainty—numerical, sampling, experimental, and parameter—have been quantitatively addressed. When multiple models are available, the one that has the smallest model form error can be selected. Though the framework is applied to a simple example here, the same process can quantify the model form uncertainty of more complicated physics, multiple models, and simulation tools in other fields. Therefore, this work is a demonstration of best practices for future assessments of model form uncertainty.}, number={12}, journal={NUCLEAR TECHNOLOGY}, author={Porter, Nathan W. and Mousseau, Vincent A. and Avramova, Maria N.}, year={2019}, month={Dec}, pages={1607–1617} }
@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} }
@article{porter_avramova_mousseau_2018, title={Uncertainty Quantification Study of CTF for the OECD/NEA LWR Uncertainty Analysis in Modeling Benchmark}, volume={190}, ISSN={["1943-748X"]}, DOI={10.1080/00295639.2018.1435135}, abstractNote={Abstract This work describes the results of a quantitative uncertainty analysis of the thermal-hydraulic subchannel code for nuclear engineering applications, Coolant Boiling in Rod Arrays-Three Field (COBRA-TF). CTF is used, which is a version of COBRA-TF developed in cooperation between the Consortium for Advanced Simulation of Light Water Reactors and North Carolina State University. Four steady-state cases from Phase II Exercise 3 of the Organisation for Economic Co-operation and Development/Nuclear Energy Agency Light Water Reactor Uncertainty Analysis in Modeling (UAM) Benchmark are analyzed using the statistical analysis tool, Design Analysis Kit for Optimization and Terascale Applications (Dakota). The input parameters include boundary condition, geometry, and modeling uncertainties, which are selected using a sensitivity study and then defined based on expert judgment. A forward uncertainty quantification method with Latin hypercube sampling (LHS) is used, where the sample size is based on available computational resources. The means and standard deviations of thermal-hydraulic quantities of interest are reported, as well as the Spearman rank correlation coefficients between the inputs and outputs. The means and standard deviations are accompanied by their respective standard errors, and the correlation coefficients are tested for statistical significance. The quantities of interest include void fractions, temperatures, and pressure drops. The predicted uncertainty in all parameters remains relatively low for all quantities of interest. The dominant sources of uncertainty are identified. For cases based on experiments, two different validation metrics are used to quantify the difference between measured and predicted void fractions. The results compare well with past studies, but with a number of improvements: the use of an updated CTF input deck using the current UAM specification and the most recent version of CTF, the use of an LHS method, an analysis of standard errors for the statistical results, and a quantitative comparison to experimental data. Though the statistical uncertainty analysis framework presented herein is applied to thermal-hydraulic analyses, it is generally applicable to any simulation tool. Given a specified amount of computational resources, it can be used to quantify statistical significance through the use of fundamental statistical analyses. This is in contrast with the prevailing methods in nuclear engineering, which provide a sample size necessary to achieve a specified level of statistical certainty.}, number={3}, journal={NUCLEAR SCIENCE AND ENGINEERING}, author={Porter, Nathan W. and Avramova, Maria N. and Mousseau, Vincent A.}, year={2018}, pages={271–286} }
@article{porter_avramova_2018, title={Validation of CTF pressure drop and void predictions for the NUPEC BWR database}, volume={337}, ISSN={["1872-759X"]}, DOI={10.1016/j.nucengdes.2018.07.018}, abstractNote={A quantified validation of CTF pressure drop, equilibrium quality, and void fraction predictions is performed using the Japanese Nuclear Power Engineering Corporation boiling water reactor database. Four quantities of interest are compared between the experiments and the code predictions: pressure drop, average exit equilibrium quality, average exit void fraction, and subchannel exit void fraction. These four quantities of interest have root-mean-squared errors of 0.124, 0.005, 0.065, and 0.089, respectively. Pressure drop predictions are generally underpredicted for single phase cases and overpredicted for two phase cases. The equilibrium quality predictions are mostly within 0.01 of the designed experimental values, which indicates proper energy conservation. The void fraction results tend to be overpredicted by about 0.06, which is attributed to the interfacial modeling in the code. By splitting the subchannels into different groups, it is shown that those near unheated surfaces are the least accurately modeled, especially for cases that have a high concentration of unheated surfaces. The results are consistent with past validation analyses using a variety of subchannel codes. Unlike past studies with CTF, this work incorporates all 69 pressure drop experiments and 392 steady state void experiments from the database. A new method for quantification of asymmetries is proposed and applied, but the interpretation of the results depends on the definition of the measurement error.}, journal={NUCLEAR ENGINEERING AND DESIGN}, author={Porter, Nathan W. and Avramova, Maria N.}, year={2018}, month={Oct}, pages={291–299} }