@article{takasugi_aly_holler_abarca_beeler_avramova_ivanov_2023, title={Development of an efficient and improved core thermal-hydraulics predictive capability for fast reactors: Summary of research and development activities at the North Carolina state University}, volume={412}, ISSN={["1872-759X"]}, DOI={10.1016/j.nucengdes.2023.112474}, abstractNote={The improved understanding of the safety, technical gaps, and major uncertainties of advanced fast reactors will result in designing their safe and economical operation. This paper focuses on the development of efficient and improved core thermal-hydraulics predictive capabilities for fast reactor modeling and simulation at the North Carolina State University. The described research and development activities include applying results of high-fidelity thermal-hydraulic simulations to inform the improved use of lower-order models within fast-running design and safety analysis tools to predict improved estimates of local safety parameters for efficient evaluation of realistic safety margins for fast reactors. The above-described high-to-low model information improvements are being verified and validated using benchmarks such as the OECD/NRC Liquid Metal Fast Reactor Core Thermal-Hydraulic Benchmark and code-to-code comparisons.}, journal={NUCLEAR ENGINEERING AND DESIGN}, author={Takasugi, C. and Aly, A. and Holler, D. and Abarca, A. and Beeler, B. and Avramova, M. and Ivanov, K.}, year={2023}, month={Oct} }
 @article{delipei_rouxelin_abarca_hou_avramova_ivanov_2022, title={CTF-PARCS Core Multi-Physics Computational Framework for Efficient LWR Steady-State, Depletion and Transient Uncertainty Quantification}, volume={15}, ISSN={["1996-1073"]}, url={https://doi.org/10.3390/en15145226}, DOI={10.3390/en15145226}, abstractNote={Best Estimate Plus Uncertainty (BEPU) approaches for nuclear reactor applications have been extensively developed in recent years. The challenge for BEPU approaches is to achieve multi-physics modeling with an acceptable computational cost while preserving a reasonable fidelity of the physics modeled. In this work, we present the core multi-physics computational framework developed for the efficient computation of uncertainties in Light Water Reactor (LWR) simulations. The subchannel thermal-hydraulic code CTF and the nodal expansion neutronic code PARCS are coupled for the multi-physics modeling (CTF-PARCS). The computational framework is discussed in detail from the Polaris lattice calculations up to the CTF-PARCS coupling approaches. Sampler is used to perturb the multi-group microscopic cross-sections, fission yields and manufacturing parameters, while Dakota is used to sample the CTF input parameters and the boundary conditions. Python scripts were developed to automatize and modularize both pre- and post-processing. The current state of the framework allows the consistent perturbation of inputs across neutronics and thermal-hydraulics modeling. Improvements to the standard thermal-hydraulics modeling for such coupling approaches have been implemented in CTF to allow the usage of 3D burnup distribution, calculation of the radial power and the burnup profile, and the usage of Santamarina effective Doppler temperature. The uncertainty quantification approach allows the treatment of both scalar and functional quantities and can estimate correlation between the multi-physics outputs of interest and up to the originally perturbed microscopic cross-sections and yields. The computational framework is applied to three exercises of the LWR Uncertainty Analysis in Modeling Phase III benchmark. The exercises cover steady-state, depletion and transient calculations. The results show that the maximum fuel centerline temperature across all exercises is 2474K with 1.7% uncertainty and that the most correlated inputs are the 238U inelastic and elastic cross-sections above 1 MeV.}, number={14}, journal={ENERGIES}, author={Delipei, Gregory K. and Rouxelin, Pascal and Abarca, Agustin and Hou, Jason and Avramova, Maria and Ivanov, Kostadin}, year={2022}, month={Jul} }
 @article{shahid_liu_abarca_novog_2021, title={Incorporation and testing of refrigerant fluids (R-134a) in the subchannel analysis code CTF (COBRA-TF)}, volume={142}, ISSN={["1878-4224"]}, DOI={10.1016/j.pnucene.2021.104028}, abstractNote={The subchannel codes such as COBRA-TF (CTF) are often used in the simulation of thermal hydraulic parameters in a reactor core. Use of these codes for safety analysis requires extensive validation against experimental data. While typically validation is performed on experiments in water, a large number of experiments using refrigerant are available in literature and may be used to expand the validation range of these codes. This research is focused on the subchannel code CTF which has been modified to incorporate Refrigerant-134a fluid properties. Subsequently the modified code was tested against several experimental results available for R-134a. Evaluation of the Heat Balance Method (HBM) and Direct Substitution Method (DSM) for CHF was carried out. In the case of the CHF Look-Up Table (LUT), fluid-to-fluid scaling was performed to predict the local CHF phenomena. Overall the HBM predictions show better agreement (with the exception of Katto's correlation when applied to test section with a cold-wall) as compared to the CHF LUT results which tended to significantly overpredict dryout under high qualities.}, journal={PROGRESS IN NUCLEAR ENERGY}, author={Shahid, U. and Liu, Y. and Abarca, A. and Novog, D. R.}, year={2021}, month={Dec} }
 @misc{hidalga_abarca_miro_sekrhi_verdu_2019, title={A multi-scale and multi-physics simulation methodology with the state-of-the-art tools for safety analysis in Light Water Reactors applied to a Turbine Trip scenario (Part II)}, volume={350}, ISSN={["1872-759X"]}, DOI={10.1016/j.nucengdes.2019.05.009}, abstractNote={The development of the computer technology, as well as the research in the different science fields governing the core behavior of a Light Water Reactor, allows implementing all the known physics and consider detailed scales of analysis. Conversely to conservative approaches, the Best Estimate approach applies the available science by means of models and correlations that are applied in different scales using simulation tools. With this approach, the critical elements of the core can be evaluated with realistic predictions that can adjust the operation conditions and core design to more cost-efficient values without compromising the safety of the Nuclear Power Plant. The authors of this paper present the second part of a multi-scale and multi-physics methodology for the evaluation of fast transients in Light Water Reactors. In this part, the results obtained from the coupled Neutron Kinetics and Thermal-Hydraulics channel-by-channel core model are used for a detailed thermal-hydraulic pin-by-pin analysis and thermomechanics pin model. The aim of this work is to evaluate the safety analysis of the critical fuel rod in Turbine Trip scenario. For that purpose, the critical fuel rod is located using the minimum Critical Power Ratio. This safety variable is predicted in a thermal-hydraulic pin-by-pin model using CTF-UPVIS code. Afterwards, the conditions of the critical rod are loaded in a pin model for a simulation with FRAPCON/FRAPTRAN. Moreover, this paper proves the Best Estimate capability of the presented methodology by means of comparing the results with equivalent simulations that are more conservative, or consist of more limited simulation scales. On the one hand, the Best Estimate prediction is compared against the envelope of the minimum Critical Power Ratio along the axial nodal distribution of the simulated fuel rod. In addition, another comparison is made against assuming constant fuel-cladding gas conductance, showing the enhancement added by considering the axial distribution of this parameter, provided by FRAPCON/FRAPTRAN. On the other hand, the results of this methodology are compared against the limitation of accounting only the bundle radial average value of the minimum Critical Power Ratio. Furthermore, the Best Estimate results are complemented with an Uncertainty and Sensitivity analysis that will define the statistical boundaries of the prediction according to the 95/95 criterion.}, journal={NUCLEAR ENGINEERING AND DESIGN}, author={Hidalga, Patricio and Abarca, Agustin and Miro, Rafael and Sekrhi, Abdelkrim and Verdu, Gumersindo}, year={2019}, month={Aug}, pages={205–213} }
 @misc{hidalga_abarca_miro_sekrhi_verdu_2019, title={A multi-scale and multi-physics simulation methodology with the state-of-the-art tools for safety analysis in light water reactors applied to a turbine trip scenario (PART I)}, volume={350}, ISSN={["1872-759X"]}, DOI={10.1016/j.nucengdes.2019.05.008}, abstractNote={The simulation of transient events is a requirement in the evaluation of the safety of Nuclear Power Plants. The Nuclear Authority request the operators to report the prediction of the evolution of the corresponding safety variables using simulation codes and methodologies that have proved to be validated against real data, whether experiments or plant measurements. Moreover, these simulation codes are used in the engineering work that a Nuclear Power Plant needs for planning a competitive and safe operation strategy. The available resources in simulation tools make possible complex analysis that can be used to predict realistic results. The consequence is the opportunity of making a safe and cost-efficient evaluation of the safety margins. Operators can use these tools for licensing to the Nuclear Authority and for calculation support of the operation of the reactor in whichever considered case. This paper presents a methodology that takes advantage of different simulation tools to join the capabilities in the Best Estimate (BE) simulation of transients for Light Water Reactors. This methodology works in different steps to account all the physics using the proper scale in a multi-physics and multi-scale approach. An automatic tool manages the data pre- and post-processing the corresponding input and output files. The purpose is to simulate the transient case in a coarse mesh and generate the boundary conditions for a simulation in more detailed scale with a finer mesh in the next step. Therefore, this methodology works generating the corresponding nodal cross section data to be used in coupled 3D thermal-hydraulics and neutron kinetics simulations run with system codes. A channel-by-channel core model is used in order to identify the critical fuel channel. Finally, the boundary conditions of the critical fuel channel are loaded in a pin-by-pin thermal-hydraulic model to perform the definitive Safety Analysis of the target variable, that is selected by the user. The methodology presented in this paper, is applied to a real fast transient case, a Turbine Trip event of fuel cycle 18 in Kernkraftwerk Leibstadt, KKL. The results of each step of this methodology have been validated against the available plant data and the selected target safety variable, the Critical Power Ratio at pin level, has been code-to-code verified. The results show good agreement proving the effectivity of this methodology.}, journal={NUCLEAR ENGINEERING AND DESIGN}, author={Hidalga, Patricio and Abarca, Agustin and Miro, Rafael and Sekrhi, Abdelkrim and Verdu, Gumersindo}, year={2019}, month={Aug}, pages={195–204} }