@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={Abstract 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={195–204} }