@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{lin_athe_rouxelin_avramova_gupta_youngblood_lane_dinh_2022, title={Digital-twin-based improvements to diagnosis, prognosis, strategy assessment, and discrepancy checking in a nearly autonomous management and control system}, volume={166}, ISSN={["1873-2100"]}, url={http://www.scopus.com/inward/record.url?eid=2-s2.0-85115958204&partnerID=MN8TOARS}, DOI={10.1016/j.anucene.2021.108715}, abstractNote={The Nearly Autonomous Management and Control System (NAMAC) is a comprehensive control system that assists plant operations by furnishing control recommendations to operators in a broad class of situations. This study refines a NAMAC system for making reasonable recommendations during complex loss-of-flow scenarios with a validated Experimental Breeder Reactor II simulator, digital twins improved by machine-learning algorithms, a multi-attribute decision-making scheme, and a discrepancy checker for identifying unexpected recommendation effects. We assess the performance of each NAMAC component, while we demonstrate and evaluated the capability of NAMAC in a class of loss-of-flow scenarios.}, journal={ANNALS OF NUCLEAR ENERGY}, author={Lin, Linyu and Athe, Paridhi and Rouxelin, Pascal and Avramova, Maria and Gupta, Abhinav and Youngblood, Robert and Lane, Jeffrey and Dinh, Nam}, year={2022}, month={Feb} }
 @article{rouxelin_alfonsi_strydom_avramova_ivanov_2022, title={Propagation of VHTRC manufacturing uncertainties with RAVEN/PHISICS}, volume={165}, ISSN={["1873-2100"]}, DOI={10.1016/j.anucene.2021.108667}, abstractNote={The International Atomic Energy Agency recently concluded a Coordinated Research Program (CRP) to evaluate the effect of propagation of uncertainties on design and safety parameters in High Temperature Gas-cooled Reactors (HTGRs). This CRP catalyzed the development of novel software and methods relevant to HTGR uncertainty analysis. In the framework of this CRP, the statistical analysis code RAVEN was coupled to the neutron transport code PHISICS, using 6-group cross section libraries generated with the modules TRITON/NEWT from SCALE 6.2.1. This article describes the mechanics of the RAVEN/PHISICS sequence, and reports the effects of manufacturing uncertainties on integral parameter uncertainties found in the Very High Temperature Reactor Critical (VHTRC) core. The VHTRC experimental results included propagation of manufacturing uncertainties to obtain eigenvalue (keff) and temperature coefficient (αT) uncertainties. RAVEN/PHISICS was used to reproduce this analysis and to compare the predicted output uncertainties to the experimental measurements on the three VHTRC cores (HC-I, HP, HC-II). Results from the sequence agree with the experimental values (σ[keff] ~ 0.00300). The analysis also focuses on the interpretation of input uncertainties. The simulations conducted with RAVEN/PHISICS demonstrated the input uncertainties can induce a threefold increase in the resulting output uncertainties, depending on the mathematical modeling of the raw input uncertainties. In particular, the use of a unique uncertainty value repeated over lattice elements constitutes the major contribution to the keff and αT uncertainties, while modeling these uncertainties with random independent values leads to negligible keff and αT uncertainties, due to cancellation of errors. The propagation of the manufacturing uncertainties was also repeated using 56 energy groups in the neutron transport calculations, and showed a moderate impact on the output (keff, αT) uncertainties (~10 % difference) compared to the base-case 6-group simulations.}, journal={ANNALS OF NUCLEAR ENERGY}, author={Rouxelin, Pascal and Alfonsi, Andrea and Strydom, Gerhard and Avramova, Maria and Ivanov, Kostadin}, year={2022}, month={Jan} }
 @article{delipei_hou_avramova_rouxelin_ivanov_2021, title={Summary of comparative analysis and conclusions from OECD/NEA LWR-UAM benchmark Phase I}, volume={384}, ISSN={["1872-759X"]}, url={http://dx.doi.org/10.1016/j.nucengdes.2021.111474}, DOI={10.1016/j.nucengdes.2021.111474}, abstractNote={In recent years, large efforts have been devoted to Light Water Reactor (LWR) Uncertainty Quantification (UQ). In 2006, the LWR Uncertainty Analysis in Modeling (UAM) benchmark was launched with an aim to investigate the uncertainty propagation in all modeling stages of the LWRs and guide uncertainty and sensitivity analysis methodology development. This article summarizes the benchmark activities for the standalone neutronics phase (Phase I), which includes three main exercises: Exercise I-1: “Cell Physics,” Exercise I-2: “Lattice Physics,” and Exercise I-3: “Core Physics.” A comparative analysis of the Phase I results is performed in this article for all the considered LWRs types: Three Mile Island – 1 Pressurized Water Reactor (PWR), Peach Bottom – 2 Boiling Water Reactor (BWR), Kozloduy – 6 Water - Water Energetic Reactor (VVER) and a Generation-III reactor. It was found, for all major exercises, that the predicted uncertainty of the system eigenvalue is highly dependent on the choice of the covariance libraries used in the UQ process and is less sensitive to the solution method, nuclear data library and UQ method. For all four reactor types, the observed relative standard deviation across all exercises is approximately 0.5% for the UO2 fuel. In the pin cell and lattice calculations with MOX fuel this uncertainty increases to 1%. The main reason is the larger Pu-239 nu-bar uncertainty compared to the U-235 nu-bar. The largest contributors to the eigenvalue uncertainties are the U-235 nu-bar and the U-238 capture in the UO2 fuel and the Pu-239 nu-bar in the MOX fuel. In the assembly lattice exercises, higher uncertainties are predicted for the fast group than the thermal group constants with differences up to one order of magnitude. This is attributed to the larger uncertainties of most cross-sections at high energies. The obtained correlation matrices share some common major trends but also exhibit strong differences in case by case comparisons indicating an impact of the selected neutronics modeling and nuclear data library. In the core exercises, the predicted relative standard deviation of the radial and axial power, for most of the cores, is below 10%. An exception is the radial power profile of the Generation-III core, when a mixture of UOX/MOX assemblies is considered. Finally, it is important to note that the bias in most of the studies is significant and up to the same order of the estimated uncertainty. This indicates a need for better quantification of the bias/variance through more code to code and code to experiments comparisons.}, journal={NUCLEAR ENGINEERING AND DESIGN}, publisher={Elsevier BV}, author={Delipei, Gregory Kyriakos and Hou, Jason and Avramova, Maria and Rouxelin, Pascal and Ivanov, Kostadin}, year={2021}, month={Dec} }
 @article{rouxelin_alfonsi_ivanov_strydom_2020, title={Energy group search engine based on surrogate models constructed with the RAVEN/NEWT/PHISICS sequence}, volume={356}, ISSN={["1872-759X"]}, DOI={10.1016/j.nucengdes.2019.110356}, abstractNote={Transient calculations with nodal neutronics codes entail few-group energy structures. In systems other than LWRs, significant efforts are devoted to obtain satisfying group structures. The energy cut-offs available in the literature do not always match the energy boundaries available in lattice codes. This paper demonstrates an automated sequence that searches for suitable coarse-group configurations. The sequence couples the lattice code T-XSEC/NEWT for cross section generation and collapsing, the nodal code PHISICS for core calculations and the software RAVEN for variable sampling and analytical purposes. T-XSEC/NEWT receives an energy group configuration from RAVEN to generate microscopic self-shielded cross sections in a coarse format. PHISICS provides the core solution using the microscopic libraries. The performances of the group structures in the core model are stored to train a Reduced Order Model (ROM) built on-the-fly. The ROM spares the necessity to survey the large input space of all possible energy group structures, or expert judgements. The solution provided by RAVEN is a Limit Surface of group structures fitting success criteria. The approach is tested on a simplified two-dimensional HTTR core model. The Limit Surface obtained by RAVEN derives a few six-group structures fitting the HTTR model.}, journal={NUCLEAR ENGINEERING AND DESIGN}, author={Rouxelin, Pascal and Alfonsi, Andrea and Ivanov, Kostadin and Strydom, Gerhard}, year={2020}, month={Jan} }