@article{chen_hou_ivanov_2023, title={A hybrid neutronics method with novel fission diffusion synthetic acceleration for criticality calculations}, volume={55}, ISSN={["1738-5733"]}, url={https://doi.org/10.1016/j.net.2022.12.022}, DOI={10.1016/j.net.2022.12.022}, abstractNote={A novel Fission Diffusion Synthetic Acceleration (FDSA) method is developed and implemented as a part of a hybrid neutronics method for source convergence acceleration and variance reduction in Monte Carlo (MC) criticality calculations. The acceleration of the MC calculation stems from constructing a synthetic operator and solving a low-order problem using information obtained from previous MC calculations. By applying the P1 approximation, two correction terms, one for the scalar flux and the other for the current, can be solved in the low-order problem and applied to the transport solution. A variety of one-dimensional (1-D) and two-dimensional (2-D) numerical tests are constructed to demonstrate the performance of FDSA in comparison with the standalone MC method and the coupled MC and Coarse Mesh Finite Difference (MC-CMFD) method on both intended purposes. The comparison results show that the acceleration by a factor of 3–10 can be expected for source convergence and the reduction in MC variance is comparable to CMFD in both slab and full core geometries, although the effectiveness of such hybrid methods is limited to systems with small dominance ratios.}, number={4}, journal={NUCLEAR ENGINEERING AND TECHNOLOGY}, author={Chen, Jiahao and Hou, Jason and Ivanov, Kostadin}, year={2023}, month={Apr}, pages={1428–1438} }
@article{vaglio-gaudard_destouches_hawari_avramova_ivanov_valentine_blaise_hudelot_2023, title={Challenge for the validation of high-fidelity multi-physics LWR modeling and simulation: Development of new experiments in research reactors}, volume={11}, ISSN={["2296-598X"]}, DOI={10.3389/fenrg.2023.1110979}, abstractNote={Current approaches to validate multi-physics coupling mainly rely upon experimental data from the operation of the current reactor fleet. These data allow global experimental validation based on Light Water Reactor (LWR) macroscopic physical parameters of interest. However, they are insufficient for validating detailed coupling at the assembly and pin level. The use of well-controlled experimental data provided by research reactors is essential to implement a rigorous and consistent step-wise validation process of high-fidelity multi-physics coupling. That is why experimental data, such as the core power evolution in a transient-state coming from the SPERT-III experimental program and the CABRI research reactor, are analyzed as a first step towards this objective for the simulation of LWR transients initiated by reactivity insertion. The analysis of the state-of-the-art shows no existing experimental benchmark available worldwide for LWRs to consistently and rigorously validate advanced reactor physics/thermal-hydraulics/fuel performance coupling at the pin- or sub-channel scale. In this context, a discussion is therefore initiated in this paper on the perspective of developing new experiments dedicated to high-fidelity multi-physics tools, focusing on a first application: the validation of reactivity feedback effects. Very few existing light-water experimental reactors containing UO 2 fuel could today have the capacity to host these experiments. The development of a new validation experiment could only be achievable by considering a two-stage process for the experiment realization: a first stage involving a distributed network of sensors in the reactor core using instrumentation commonly used in research reactors, and a second stage implementing an instrumented fuel pin and innovative experimental techniques, in the longer term. Even if the OECD/NEA activities in the Expert Group on Multi-Physics Experimental Data, Benchmarks and Validation (EGMPEBV) (currently merged in the Expert Group on Multi-Physics of Reactor Systems – EGMUP) have started to pave the way for the development of such a high-fidelity multi-physics experiment, most of the work is still ahead of us.}, journal={FRONTIERS IN ENERGY RESEARCH}, author={Vaglio-Gaudard, Claire and Destouches, Christophe and Hawari, Ayman and Avramova, Maria and Ivanov, Kostadin and Valentine, Timothy and Blaise, Patrick and Hudelot, Jean-Pascal}, year={2023}, month={Jan} }
@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{faure_delipei_petruzzi_avramova_ivanov_2023, title={Fuel performance uncertainty quantification and sensitivity analysis in the presence of epistemic and aleatoric sources of uncertainties}, volume={11}, ISSN={["2296-598X"]}, DOI={10.3389/fenrg.2023.1112978}, abstractNote={Fuel performance modeling and simulation includes many uncertain parameters from models to boundary conditions, manufacturing parameters and material properties. These parameters exhibit large uncertainties and can have an epistemic or aleatoric nature, something that renders fuel performance code-to-code and code-to-measurements comparisons for complex phenomena such as the pellet cladding mechanical interaction (PCMI) very challenging. Additionally, PCMI and other complex phenomena found in fuel performance modeling and simulation induce strong discontinuities and non-linearities that can render difficult to extract meaningful conclusions form uncertainty quantification (UQ) and sensitivity analysis (SA) studies. In this work, we develop and apply a consistent treatment of epistemic and aleatoric uncertainties for both UQ and SA in fuel performance calculations and use historical benchmark-quality measurement data to demonstrate it. More specifically, the developed methodology is applied to the OECD/NEA Multi-physics Pellet Cladding Mechanical Interaction Validation benchmark. A cold ramp test leading to PCMI is modeled. Two measured quantities of interest are considered: the cladding axial elongation during the irradiations and the cladding outer diameter after the cold ramp. The fuel performance code used to perform the simulation is FAST. The developed methodology involves various steps including a Morris screening to decrease the number of uncertain inputs, a nested loop approach for propagating the epistemic and aleatoric sources of uncertainties, and a global SA using Sobol indices. The obtained results indicate that the fuel and cladding thermal conductivities as well as the cladding outer diameter uncertainties are the three inputs having the largest impact on the measured quantities. More importantly, it was found that the epistemic uncertainties can have a significant impact on the measured quantities and can affect the outcome of the global sensitivity analysis.}, journal={FRONTIERS IN ENERGY RESEARCH}, author={Faure, Quentin and Delipei, Gregory and Petruzzi, Alessandro and Avramova, Maria and Ivanov, Kostadin}, year={2023}, month={Mar} }
@article{moloko_bokov_wu_ivanov_2023, title={Prediction and uncertainty quantification of SAFARI-1 axial neutron flux profiles with neural networks}, volume={188}, ISSN={["1873-2100"]}, url={https://doi.org/10.1016/j.anucene.2023.109813}, DOI={10.1016/j.anucene.2023.109813}, abstractNote={Artificial Neural Networks (ANNs) have been successfully used in various nuclear engineering applications, such as predicting reactor physics parameters within reasonable time and with a high level of accuracy. Despite this success, they cannot provide information about the model prediction uncertainties, making it difficult to assess ANN prediction credibility, especially in extrapolated domains. In this study, Deep Neural Networks (DNNs) are used to predict the assembly axial neutron flux profiles in the SAFARI-1 research reactor, with quantified uncertainties in the ANN predictions and extrapolation to cycles not used in the training process. The training dataset consists of copper-wire activation measurements, the axial measurement locations and the measured control bank positions obtained from the reactor's historical cycles. Uncertainty Quantification of the regular DNN models' predictions is performed using Monte Carlo Dropout (MCD) and Bayesian Neural Networks solved by Variational Inference (BNN VI). The regular DNNs, DNNs solved with MCD and BNN VI results agree very well among each other as well as with the new measured dataset not used in the training process, thus indicating good prediction and generalization capability. The uncertainty bands produced by MCD and BNN VI agree very well, and in general, they can fully envelop the noisy measurement data points. The developed ANNs are useful in supporting the experimental measurements campaign and neutronics code Verification and Validation (V&V).}, journal={ANNALS OF NUCLEAR ENERGY}, author={Moloko, Lesego E. and Bokov, Pavel M. and Wu, Xu and Ivanov, Kostadin N.}, year={2023}, month={Aug} }
@article{takasugi_martin_laboure_ortensi_ivanov_avramova_2023, title={Preservation of kinetics parameters generated by Monte Carlo calculations in two-step deterministic calculations}, volume={9}, ISSN={["2491-9292"]}, DOI={10.1051/epjn/2022056}, abstractNote={The generation of accurate kinetic parameters such as mean generation time Λ and effective delayed neutron fraction β eff via Monte Carlo codes is established. Employing these in downstream deterministic codes warrants another step to ensure no additional error is introduced by the low-order transport operator when computing forward and adjoint fluxes for bilinear weighting of these parameters. Another complexity stems from applying superhomogenization (SPH) equivalence in non-fundamental mode approximations, where reference and low-order calculations rely on a 3D full core model. In these cases, SPH factors can optionally be computed for only part of the geometry while preserving reaction rates and K -effective, but the impact of such approximations on kinetics parameters has not been thoroughly studied. This paper aims at studying the preservation of bilinearly-weighted quantities in the Serpent–Griffin calculation procedure. Diffusion and transport evaluations of IPEN/MB-01, Godiva, and Flattop were carried out with the Griffin reactor physics code, testing available modeling options using Serpent-generated multigroup cross sections and equivalence data. Verifying Griffin against Serpent indicates sensitivities to multigroup energy grid selection and regional application of SPH equivalence, introducing significant errors; these were demonstrated to be reduced through the use of a transport method together with a finer energy grid.}, journal={EPJ NUCLEAR SCIENCES & TECHNOLOGIES}, author={Takasugi, Cole and Martin, Nicolas and Laboure, Vincent and Ortensi, Javier and Ivanov, Kostadin and Avramova, Maria}, year={2023}, month={Feb} }
@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{altahhan_geemert_avramova_ivanov_2022, title={Extending a low-order inhomogeneous adjoint equations model to a higher-order model with verification on integral applications}, volume={177}, ISSN={["1873-2100"]}, DOI={10.1016/j.anucene.2022.109277}, abstractNote={• Development and verification of a NEM-M2B2 mathematical inhomogeneous-adjoint nodal diffusion solver. • Application of the Lagrangian multipliers method to derive the nodal mathematical inhomogeneous-adjoint. • Derived the local linear prediction formula, specific for the forward NEM-M2B2 model, and utilized it to study the repercussions of perturbations in the IAEA-3D benchmark on the Axial Offset (AO). • Verified the generalized adjoint code developed while showing detailed steps of how inhomogeneous adjoint codes can be verified. • Compared between the low-order inhomogeneous NEM-M0 adjoint and the developed higher order inhomogeneous NEM-M2B2 model for the AO as a RoI. • Introduced the Mantissa theory to explain the behavior of the linear adjoint models and prediction formulas. A higher-order nodal mathematical inhomogeneous adjoint model conjugate to the NEM-M2B2 nodal diffusion forward model is developed and introduced in this research. Verification of the developed model is presented through applications in perturbation analysis and the IAEA-3D benchmark including adjusted forms of it. This paper’s objective is to explore ways of extending and optimizing a mathematical adjoint capability suitable for use in an industrial reactor code, such that it becomes not merely an approximate but rather the exact adjoint counterpart to the typically used higher-order nodal forward solvers used in mature industrial reactor codes. Specifically, it is investigated how to upgrade an already available lower-order nodal mathematical adjoint solver towards higher-order accuracy. An example of the latter is the lower-order nodal adjoint solver used in the ARTEMIS reactor code, in the technical context of stabilization and acceleration of embedded control rod search mechanisms. Though the latter adjoint solver proved suitable for the needed preconditioning purposes, while also enabling the benefit of computationally very lean adjoint iterations, several future developments could benefit from having a higher-order adjoint nodal solver available as well. By using a preconditioned form of the base NEM-M2B2 nodal diffusion forward model and by using variational analysis, we have obtained a higher-order nodal mathematical adjoint that can have a physical interpretation associated with it as a Lagrangian multiplier. The nodal mathematical adjoint is then developed for the Axial Offset (AO) as a Response of Interest (RoI) which leads to an inhomogeneous adjoint system of equations. A solution verification of the adjoint developed is done through analyzing the effects coming from perturbations in the absorption and the scattering cross-sections. The applications investigated include axially and radially traveling perturbations along the reactor’s core. Several locations for the traveling perturbations are chosen to represent important locations in the core. Comparison between the low-order and the higher-order adjoint models is conducted. The forward model is set to the NEM-M2B2 nodal diffusion equation for both adjoints during the comparison. The higher-order adjoint model developed show consistent results in comparison to its lower-order sibling, suggesting the preference of using the developed higher-order model for adjoint computations.}, journal={ANNALS OF NUCLEAR ENERGY}, author={Altahhan, Muhammad Ramzy and Geemert, Rene and Avramova, Maria and Ivanov, Kostadin}, year={2022}, month={Nov} }
@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 (α T ) 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 k eff and α T uncertainties, while modeling these uncertainties with random independent values leads to negligible keff and α T 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, α T ) 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{altahhan_geemert_avramova_ivanov_2021, title={Development and verification of a higher-order mathematical adjoint nodal diffusion solver}, volume={163}, ISSN={["1873-2100"]}, DOI={10.1016/j.anucene.2021.108548}, abstractNote={In this paper, we derive a mathematical formulation of the higher order adjoint NEM-M2B2 equations by preconditioning the nodal interface neutron currents equations of the forward equations system, and by using the Lagrangian Multipliers analysis method. In the NEM-M2B2 system of equations, the quadratic transverse leakage approximation is used to model the leakage of neutrons between each node in the system. The solution of the adjoint equation can be used to perform adjoint-based predictive sensitivity/perturbation analysis. As an example, we use the mathematical adjoint solution as sensitivity weighting for predicting the response of the IAEA-3D benchmark’s eigenvalue to a perturbation in the independent parameters of the system (i.e., cross-sections). We also derive perturbation equations associated with the particular NEM-M2B2 model we are using. These perturbation-equations are used in predicting the model eigenvalue change without resorting to recalculating the forward NEM-M2B2 system of equations again (labeled as exact calculations). They also enabled construction of a reactivity sensitivity map showing the importance of each calculation node of the benchmark depending on its spatial and spectral coordinates. Perturbations were imposed on both the absorption cross-sections (fast and thermal) and the scattering cross-section of the IAEA-3D benchmark problem. Several verification steps were taken to ensure that the developed mathematical adjoint solver is adequate for adjoint analysis (e.g., commutativity checks, and comparison against exact calculations).}, journal={ANNALS OF NUCLEAR ENERGY}, author={Altahhan, Muhammad Ramzy and Geemert, Rene and Avramova, Maria and Ivanov, Kostadin}, year={2021}, month={Dec} }
@article{ivanov_sargeni_ivanov_bruna_2021, title={Evidence-based background for constrained uncertainty quantification in a core transient analysis}, volume={164}, ISSN={["1873-2100"]}, DOI={10.1016/j.anucene.2021.108606}, abstractNote={The paper discusses some topics related to a validation of multi-physics modeling. Since validation belongs to a category of decision-making processes, it should be prepared by dedicated scientific researches. In particular, the validation means a kind of characterization of the predictive capability maturity of given tools, libraries and calculational models. The compliance criteria might be expressed in terms of achievable accuracy or, inverse, in terms of somehow identified and quantified uncertainties. Despite the uncertainties, as such, might not be measured or compared with something measurable all the judgments should rely on reality, i.e., be supported by an evidence-based background. In practice it does require to eliminate subjective statements, if any, replacing them with something inferred from objective observations, including representative integral experiments. Unfortunately, in many fields like, among others, multi-physics simulations, we have not statistically sufficient number of high-fidelity and confident experiment-based benchmarks. In addition, because of technological and safety constraints, what is needed lies, largely, beyond the experimental domain. This is why, assessors have to rely on numerous, but partially representative experiments. These data could be treated using one or other Data Assimilation techniques to provide correction factors and uncertainties to single- and few-physics modules all having an evidence-based background. Then, coupling these pre-validated modules and their uncertainties, we could estimate uncertainties (and accuracies) in an application domain using one of wide range error propagation techniques. Thus, combining experiments-grounded and calibrated uncertainties, we are providing consistent, evidence-based background for validation. The last phase – constrained uncertainty propagation – was illustrated with an example of one international standard problem on transient initiated in LWR core by inadvertent Control Rod ejection.}, journal={ANNALS OF NUCLEAR ENERGY}, author={Ivanov, Evgeny and Sargeni, Antonio and Ivanov, Kostadin and Bruna, Giovanni}, year={2021}, month={Dec} }
@article{xu_hou_ivanov_2021, title={Methodology for Discontinuity Factors Generation for Simplified P-3 Solver Based on Nodal Expansion Formulation}, volume={14}, ISSN={["1996-1073"]}, url={https://doi.org/10.3390/en14206478}, DOI={10.3390/en14206478}, abstractNote={The Simplified Spherical Harmonic (SPN) approximation was first introduced as a three-dimensional (3D) extension of the plane-geometry Spherical Harmonic (PN) equations. A third order SPN (SP3) solver, recently implemented in the Nodal Expansion Method (NEM), has shown promising performance in the reactor core neutronics simulations. This work is focused on the development and implementation of the transport-corrected interface and boundary conditions in an NEM SP3 solver, following recent published work on the rigorous SPN theory for piecewise homogeneous regions. A streamlined procedure has been developed to generate the flux zero and second order/moment discontinuity factors (DFs) of the generalized equivalence theory to minimize the error introduced by pin-wise homogenization. Moreover, several colorset models with varying sizes and configurations are later explored for their capability of generating DFs that can produce results equivalent to that using the whole-core homogenization model for more practical implementations. The new developments are tested and demonstrated on the C5G7 benchmark. The results show that the transport-corrected SP3 solver shows general improvements to power distribution prediction compared to the basic SP3 solver with no DFs or with only the zeroth moment DF. The complete equivalent calculations using the DFs can almost reproduce transport solutions with high accuracy. The use of equivalent parameters from larger size colorset models show a slightly reduced prediction error than that using smaller colorset models in the whole-core calculations.}, number={20}, journal={ENERGIES}, publisher={MDPI AG}, author={Xu, Yuchao and Hou, Jason and Ivanov, Kostadin}, year={2021}, month={Oct} }
@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} }
@article{nyalunga_naicker_ivanov_2019, title={Quantification and propagation of neutronics uncertainties of the Kozloduy-6 VVER-1000 fuel assembly using SCALE 6.2.1 within the NEA/OECD benchmark for uncertainty analysis in modelling of LWRs}, volume={133}, ISBN={0306-4549}, DOI={10.1016/j.anucene.2019.07.016}, abstractNote={Abstract This work is based on the benchmark for uncertainty analysis in modelling of light water reactors compiled by the Nuclear Energy Agency within the Organisation for Economic Cooperation and Development (OECD/NEA). The objective of the benchmark is to determine and verify uncertainty bounds for results of calculations of LWRs based on operating data using best-estimate codes. The main contribution of this paper is the quantification of uncertainties in the Kozloduy-6 VVER-1000 fuel assembly using SCALE-6.2.1 methodology. The benchmark consists of three phases, each with three exercises. Three reactor systems are also studied, viz. the PWR, VVER and BWR reactors. In this study, Phase I of the benchmark was considered for the uncertainty quantification. The sources of uncertainties are classified into three groups, namely uncertainties due to nuclear data, manufacturing tolerances and numerical uncertainties due to methods’ implementations. The calculations are carried out using KENO-VI to perform the neutronics calculations and TSUNAMI-2D/3D and SAMPLER to perform the sensitivity and uncertainty analysis. Nuclear data uncertainty has been identified to be the highest contributor of uncertainty of the VVER-1000 fuel assembly. Although this is true, the uncertainty due to other parameters must always be considered together with nuclear data, since some of them could be significant.}, journal={ANNALS OF NUCLEAR ENERGY}, author={Nyalunga, G. P. and Naicker, V. V. and Ivanov, K.}, year={2019}, month={Nov}, pages={732–749} }
@article{petruzzi_ivanov_ivanov_2019, title={Selected papers from the 2018 Best Estimate Plus Uncertainty International Conference (BEPU 2018) Foreword}, volume={205}, ISSN={["1943-7471"]}, DOI={10.1080/00295450.2019.1676080}, number={12}, journal={NUCLEAR TECHNOLOGY}, author={Petruzzi, Alessandro and Ivanov, Kostadin and Ivanov, Evgeny}, year={2019}, month={Dec}, pages={III-IV} }
@article{zeng_hou_ivanov_jessee_2019, title={Uncertainty Quantification and Propagation of Multiphysics Simulation of the Pressurized Water Reactor Core}, volume={205}, ISSN={0029-5450 1943-7471}, url={http://dx.doi.org/10.1080/00295450.2019.1580533}, DOI={10.1080/00295450.2019.1580533}, abstractNote={In recent years, the demand to provide best-estimate predictions with confident bounds is increasing for the nuclear reactor performance and safety analysis. The Organisation for Economic Co-operation and Development Nuclear Energy Agency has been developing an international benchmark of the light water reactor (LWR) uncertainty analysis in modeling (UAM) for the examination of uncertainty quantification and propagation methodologies with various modeling and simulation code systems. The objective of the present work is to develop an uncertainty propagation mechanism based on the stochastic sampling method by taking into account the uncertainties of both basic nuclear data and fuel modeling parameters in the simulation of pressurized water reactors (PWRs) that can be incorporated in the conventional LWR simulation approach. More specifically, the Three Mile Island Unit 1–related exercises from the LWR-UAM benchmark have been modeled using the coupled TRACE/PARCS code system in the three-dimensional core representation. The input uncertainties of the neutronics simulation include few-group cross sections and kinetics parameters generated using the Sampler/Polaris sequence of SCALE 6.2.1. Several heat transfer–related variables for the fuel modeling were considered as sources of input uncertainty of the thermal-hydraulics simulations, including the thermal conductivity of fuel and cladding, fuel heat capacity, and the gap conductance. Dakota was used to sample input parameters of the coupled code system and to perform the uncertainty analysis. Two types of simulations were conducted: steady-state calculation at hot full-power condition and transient scenario initiated by the spatially asymmetric rod ejection accident. Quantities of interest for the steady-state calculation, including core multiplication factor and power peaking factors, were calculated with associated uncertainties. For transient calculations, best-estimate plus uncertainty results of the time evolution of core reactivity, core power, and peak fuel temperature were generated and analyzed. The Wilks’ formula was used to determine the necessary sample size to achieve a 95% confidence of the 95% limit of output quantities of interest. Although the uncertainty propagation and quantification method presented in this paper was developed for PWRs, it could be in general applicable to the multiphysics uncertainty quantification of other types of LWR cores.}, number={12}, journal={Nuclear Technology}, publisher={Informa UK Limited}, author={Zeng, Kaiyue and Hou, Jason and Ivanov, Kostadin and Jessee, Matthew Anderson}, year={2019}, month={Mar}, pages={1618–1637} }
@article{rouxelin_strydom_alfonsi_ivanov_2018, title={The IAEA CRP on HTGR uncertainties: Sensitivity study of PHISICS/RELAP5-3D MHTGR-350 core calculations using various SCALE/NEWT cross-section sets for Ex. II-1a}, volume={329}, ISSN={0029-5493}, url={http://dx.doi.org/10.1016/J.NUCENGDES.2017.12.008}, DOI={10.1016/J.NUCENGDES.2017.12.008}, abstractNote={Uncertainty and sensitivity analysis is an indispensable element of any substantial attempt in reactor simulation validation. A Coordinated Research Program (CRP) supervised by the International Atomic Energy Agency was started to investigate the various uncertainty quantification methodologies for High Temperature Gas-cooled Reactors. The first phase of the CRP is dedicated to the estimation of cell and lattice model uncertainties due to the neutron cross sections. Phase II is oriented towards the investigation of propagated uncertainties from the lattice to the coupled neutronics/thermal hydraulics core calculations. Best-estimate results for the prismatic single block (Ex. I-2a) and super cell models (Ex. I-2c) were obtained using the SCALE 6.2.0 two-dimensional lattice code NEWT. A reference spectrum was obtained with Serpent 2.1.27 for the single block, super cell and core models. The flux spectrum in the system of interest plays a primary role in the quantification of uncertainties caused by cross sections. In this work, the TRITON/NEWT-flux-weighted cross sections obtained for Ex. I-2a and various models of Ex.I-2c are utilized to perform a sensitivity analysis of the MHTGR-350 core power densities and eigenvalues. The core solutions are obtained with the Idaho National Laboratory (INL) coupled code PHISICS/RELAP5-3D. It is observed that axial shape of the core power density does not vary significantly with the various lattice cell libraries utilized. The use of cross section libraries originating from super cells induces changes of the core power density by 1–10% radially as compared to the Ex.I-2a cross sections. The magnitude of these variations increases as the moderator-to-fuel ratio increases in the super cell lattice models. A lattice flux spectrum resembling the core spectrum is hence necessary for correct predictions in nodal core calculations.}, journal={Nuclear Engineering and Design}, publisher={Elsevier BV}, author={Rouxelin, Pascal and Strydom, Gerhard and Alfonsi, Andrea and Ivanov, Kostadin}, year={2018}, month={Apr}, pages={156–166} }
@article{georgieva_dinkov_ivanov_2017, title={Benchmarking the Real-Time Core Model for VVER-1000 Simulator Application on Asymmetric Core Load}, volume={3}, ISSN={["2332-8975"]}, DOI={10.1115/1.4035550}, abstractNote={The aim of this paper is to summarize authors' experience in adaptation of an existing plant-specific VVER-1000/V320 model for simulation of a rare example of a Kalinin 3 nuclear power plant (NPP) transient of “switching-off of one of the four operating main circulation pumps at nominal reactor power” with an asymmetric core configuration. The fidelity and accuracy of simulation with emphasis on reactor core model is illustrated through comparison with plant-specific data. Simulation results concerning fuel assembly (FA) power and axial power distribution during the transient are compared with records from Kalinin 3 in-core monitoring system (ICMS). Main operating parameters of nuclear steam supply system of a VVER-1000/V320 series units vary to a considerable degree. While Kalinin 3 benchmark specification contains very good description of the transient, as well as record of many parameters of the unit, the document provides only superficial description of the reference unit. In such a case, an approach based on a “generic” V320 model by default introduces deviations which are difficult to quantify. There are several examples which warrant discussion. Some of the most important lessons learned are as follows. (1) individual characteristics of all the main circulation pumps and the reactor coolant loops are quite important for the quality of simulation and should be accounted for in the model; (2) variations in fuel assembly characteristics should be accounted for not only in terms of macroscopic cross section library but also in terms of local pressure loss coefficients and mixing factors in the case of mixed core loads; (3) comprehensive plant-specific model of dynamic response of instrumentation and control (I&C) systems is a necessity; dynamic characteristics of individual measurement channels (nuclear instrumentation, pressure, temperature) should be accounted for; and (4) comprehensive plant-specific model of balance-of-plant equipment, instrumentation, and control is a necessity. Above requirements impose a difficult task to comply with. Nevertheless, any individual nuclear power unit is supposed to maintain a detailed design database and data requirements for plant-specific model development should be considered.}, number={3}, journal={JOURNAL OF NUCLEAR ENGINEERING AND RADIATION SCIENCE}, author={Georgieva, Emiliya and Dinkov, Yavor and Ivanov, Kostadin}, year={2017}, month={Jul} }
@article{pericas_ivanov_reventós_batet_2017, title={Comparison of Best-Estimate Plus Uncertainty and Conservative Methodologies for a PWR MSLB Analysis Using a Coupled 3-D Neutron-Kinetics/Thermal-Hydraulic Code}, volume={198}, ISSN={0029-5450 1943-7471}, url={http://dx.doi.org/10.1080/00295450.2017.1299493}, DOI={10.1080/00295450.2017.1299493}, abstractNote={This paper compares the Best-Estimate Plus Uncertainty (BEPU) methodology with the Conservative Bounding methodology for design-basis-accident analysis. Calculations have been performed with TRACE [for thermal-hydraulic (TH) system calculations] and PARCS [for neutron-kinetics (NK) modeling] under the SNAP platform. DAKOTA is used under the SNAP interface for uncertainty and sensitivity analysis. A simplified three-dimensional (3-D) neutronics model of the Ascó II nuclear power plant is used as the core kinetic model. The TH model is a one-dimensional representation of the primary and secondary systems except for the vessel, which is represented by a 3-D VESSEL component. The design-basis transient selected for the comparison is a main steam line break (MSLB) in a pressurized water reactor. This transient is characterized by space-time effects and requires coupled 3-D kinetics and TH modeling, especially for the recriticality scenario. The comparison methodology is as follows. Once the models are created, a best-estimate base case calculation is performed. The model is modified by selecting the most important parameters and assigning conservative values to them in order to obtain a conservative calculation. Several parameters are modified in this conservative way. These parameters are then perturbed in BEPU calculations. At the end, a comparison is made between results obtained in the conservative calculation and the BEPU methodology, respectively. As a general conclusion the BEPU method has been successfully illustrated in a coupled 3-D kinetics and TH system. Also, the study is an effective test for the adequacy of nodalizations for the neutronic and TH utilized codes. The BEPU methodology gives more margins, which allows for higher operational flexibility of the plant. The results of the BEPU methodology help improve the plant economics while meeting the safety standards. As a conclusion, the BEPU methodology has been successfully tested in NK-TH calculations. Narrow margins between the upper and lower BEPU cases are a consequence of the few perturbed parameters chosen and the transient boundary conditions.}, number={2}, journal={Nuclear Technology}, publisher={Informa UK Limited}, author={Pericas, R. and Ivanov, K. and Reventós, F. and Batet, L.}, year={2017}, month={May}, pages={193–201} }
@article{shi_levine_ivanov_2017, title={New techniques for designing the initial and reload cores with constant long cycle lengths}, volume={99}, ISSN={0306-4549}, url={http://dx.doi.org/10.1016/J.ANUCENE.2016.08.018}, DOI={10.1016/J.ANUCENE.2016.08.018}, abstractNote={Abstract Several utilities have increased the output power of their nuclear power plant to increase their income and profit. Thus, the utility increases the power density of the reactor, which has other consequences. One consequence is to increase the depletion of the fuel assemblies (FAs) and reduce the end-of-cycle (EOC) sum of fissionable nuclides in each FA, ∑EOC. The power density and the ∑EOC remaining in the FAs at EOC must be sufficiently large in many FAs when designing the loading pattern, LP, for the first and reload cycles to maintain constant cycle lengths at minimum fuel cost. Also of importance is the cycle length as well as several other factors. In fact, the most important result of this study is to understand that the ∑EOCs in the FAs must be such that in the next cycle they can sustain the energy during depletion to prevent too much power shifting to the fresh FAs and, thus, sending the maximum peak pin power, PPPmax, above its constraint. This paper presents new methods for designing the LPs for the initial and follow on cycles to minimize the fuel costs. Studsvik’s CMS code system provides a 1000 MWe LP design in their sample inputs, which is applied in this study. The first 3 cycles of this core are analyzed to minimize fuel costs, and all three cycles have the same cycle length of ∼650 days. Cycle 1 is designed to allow many used FAs to be loaded into cycles 2 and 3 to reduce their fuel costs. This could not be achieved if cycle 1 was a low leakage LP (Shi et al., 2015). Significant fuel cost savings are achieved when the new designs are applied to the higher leakage LP designs. There are many factors, such as the core power density, cycle length, fuel cost, time between core shutdown and return to power, cost of replacement power during shutdown, loss of income during shutdown, cost of storing used FAs, and the income accrued over the same period of operation, which the utilities must consider when trying to increase their profit. The purpose of this paper is to provide the information to give guidance in making these decisions. Relative cost calculations are presented in establishing this guidance by comparing the utility’s profit accrued over a total cycle for the different core designs and cycle lengths.}, journal={Annals of Nuclear Energy}, publisher={Elsevier BV}, author={Shi, Jun and Levine, Samuel and Ivanov, Kostadin}, year={2017}, month={Jan}, pages={165–173} }
@article{hou_ivanov_boyarinov_fomichenko_2017, title={OECD/NEA benchmark for time-dependent neutron transport calculations without spatial homogenization}, volume={317}, ISSN={["1872-759X"]}, url={https://doi.org/10.1016/j.nucengdes.2017.02.008}, DOI={10.1016/j.nucengdes.2017.02.008}, abstractNote={A Nuclear Energy Agency (NEA), Organization for Economic Co-operation and Development (OECD) benchmark for the time-dependent neutron transport calculations without spatial homogenization has been established in order to facilitate the development and assessment of numerical methods for solving the space-time neutron kinetics equations. The benchmark has been named the OECD/NEA C5G7-TD benchmark, and later extended with three consecutive phases each corresponding to one modelling stage of the multi-physics transient analysis of the nuclear reactor core. This paper provides a detailed introduction of the benchmark specification of Phase I, known as the “kinetics phase”, including the geometry description, supporting neutron transport data, transient scenarios in both two-dimensional (2-D) and three-dimensional (3-D) configurations, as well as the expected output parameters from the participants. Also presented are the preliminary results for the initial state 2-D core and selected transient exercises that have been obtained using the Monte Carlo method and the Surface Harmonic Method (SHM), respectively.}, journal={NUCLEAR ENGINEERING AND DESIGN}, publisher={Elsevier BV}, author={Hou, Jason and Ivanov, Kostadin N. and Boyarinov, Victor F. and Fomichenko, Peter A.}, year={2017}, month={Jun}, pages={177–189} }
@article{bratton_jessee_wieselquist_ivanov_2017, title={Rod Internal Pressure Distribution and Uncertainty Analysis Using FRAPCON}, volume={197}, ISSN={["1943-7471"]}, DOI={10.13182/nt16-75}, abstractNote={The discharge rod internal pressure (RIP) and cladding hoop stress (CHS) distributions are quantified for Watts Bar Nuclear Unit 1 (WBN1) fuel rods by modeling core cycle design data, operation data, and as-built fuel enrichments and densities of each fuel rod in FRAPCON-3.5. A methodology is developed that tracks intercycle assembly movements and assembly batch fabrication information to build individual FRAPCON inputs for each evaluated WBN1 fuel rod. An alternate model for the amount of helium released from the zirconium diboride (ZrB2) integral fuel burnable absorber (IFBA) layer is derived and applied to FRAPCON output data to quantify the RIP and CHS for these types of fuel rods. SCALE/Polaris is used to quantify fuel rod–specific spectral quantities and the amount of gaseous fission products produced in the fuel for use in FRAPCON inputs. Fuel rods with ZrB2 IFBA layers (i.e., IFBA rods) are determined to have RIP predictions that are elevated when compared to fuel rods without IFBA layers (i.e., standard rods) despite the fact that IFBA rods often have reduced fill pressures and annular fuel pellets. The primary contributor to elevated RIP predictions at burnups less than and greater than 30 GWd/tonne U is determined to be the total fuel rod void volume and the amount of released fission gas in the fuel rod, respectively. Cumulative distribution functions (CDFs) are prepared from the distribution of RIP and CHS predictions for all standard and IFBA rods. The provided CDFs allow for the determination of the portion of WBN1 fuel rods that exceeds a specified RIP or CHS limit. Results are separated into IFBA and standard rods so that the two groups may be analyzed individually. FRAPCON results are provided in sufficient detail to enable the recalculation of the RIP while considering any desired plenum gas temperature, total void volume, or total amount of gas present in the void volume. A method to predict the CHS from a determined or assumed RIP is also proposed that is based on the approximately linear relationship between the CHS and the RIP. Finally, improvements to the computational methodology of FRAPCON are proposed.}, number={1}, journal={NUCLEAR TECHNOLOGY}, author={Bratton, Ryan N. and Jessee, Matt A. and Wieselquist, William A. and Ivanov, Kostadin N.}, year={2017}, month={Jan}, pages={47–63} }
@article{thompson_ivanov_2016, title={Advances in the Pennsylvania State University NEM code}, volume={94}, ISSN={["0306-4549"]}, DOI={10.1016/j.anucene.2016.03.014}, abstractNote={The Pennsylvania State University NEM code has been updated in an attempt to enable the code to model more neutronically complex reactor cores, such as those containing mixed-oxide fuel, low leakage cores, and cores that contain multiple burnable poison types. Current nodal methods, which are primarily focused on solving the diffusion equation using a nodal expansion method with the transverse leakage term solved using the quadratic leakage approximation, are known to be inaccurate in such environments. The NEM code is updated with a transport capability based upon the SP3 approximation, a semi-analytical solution, and an advanced transverse leakage method based upon the use of analytic basis functions. Each of these new features is described followed by the results of benchmarks to test their effectiveness.}, journal={ANNALS OF NUCLEAR ENERGY}, author={Thompson, S. A. and Ivanov, K. N.}, year={2016}, month={Aug}, pages={251–262} }
@inproceedings{georgieva_dinkov_ivanov_stieglitz_2016, title={Benchmarking the NEM real-time core model for VVER-1000 simulator application - asymmetric core}, DOI={10.1115/icone24-60135}, abstractNote={A real-time version of the Nodal Expansion Method (NEM) code is developed and implemented into Kozloduy 6 full-scope replica control room simulator. Combined with an enhanced thermal-hydraulics and I&C models the whole package is a high-fidelity simulation tool for operator training and various other applications. The fidelity and accuracy of simulation with emphasis on reactor core model is illustrated through comparison with plant-specific data. The transient of ‘Switching-off of One of the Four Operating Main Circulation Pumps at Nominal Reactor Power’ as described in OECD/NEA Kalinin 3 Coolant Transient Benchmark is an example of an asymmetric core scenario with a range of parameter changes. Simulation results concerning fuel assembly power and axial power distribution during the transient are compared with records from Kalinin 3 in-core monitoring system. Main operating parameters of nuclear steam supply system of a VVER-1000/V320 series units vary to a considerable degree. While Kalinin 3 benchmark specification contains very good description of the transient, as well as record of many parameters of the unit, the document provides only superfluous description of the reference unit. In such a case, an approach based on a ‘generic’ V320 model by default introduces deviations which are difficult to quantify. There are several examples which warrant discussion. One example is core coolant flow and pressure loss during the transient. Pump head and pressure loss across reactor vessel are measured and recorded and in-core monitoring system provides estimation of core coolant flow, which is quite high in comparison with some other V320 units (e.g. by about 5 % larger). Without more detailed pressure loss data across the main circulation loop and specific pump characteristics, however, one can only guess how much simulation is off the mark. Another detail of the same problem is coolant flow through a specific fuel assembly. The presence of a fuel assembly of different design (TVS-M type) surrounded by TVSA type fuel assemblies shall be thoroughly considered, because secondary sources indicate significant differences in fuel assembly pressure loss coefficients between the two types. Coolant flow affects coolant (and fuel) temperature profile and thus neutron cross-sections. Yet another example, even more strongly affecting our ability to interpret simulation results is core power reconstruction provided by the in-core monitoring system of the unit. The SPND (Self-Powered Neutron Detector) current readings are subject of conversion by an algorithm based upon simulated spatial neutron flux distribution across the reactor core. While error estimation of the parameters in stationary conditions is available from secondary sources, there is no reliable estimation of error magnitude during the transient.}, booktitle={Proceedings of the 24th International Conference on Nuclear Engineering, 2016, vol 5}, author={Georgieva, E. and Dinkov, Y. and Ivanov, Kostadin and Stieglitz, R.}, year={2016} }
@article{pericas_ivanov_reventós_batet_2016, title={Code improvement and model validation for Ascó-II Nuclear Power Plant model using a coupled 3D neutron kinetics/thermal–hydraulic code}, volume={87}, ISSN={0306-4549}, url={http://dx.doi.org/10.1016/J.ANUCENE.2015.09.024}, DOI={10.1016/J.ANUCENE.2015.09.024}, abstractNote={This paper provides a Best Estimate validation calculation with a coupled thermal–hydraulic and 3D neutron kinetic model for Asco-II Nuclear Power Plant. Common NRC codes have been used for its purpose. TRACE is the code used for the thermal–hydraulic system calculations; PARCS is the code used for the 3D neutron kinetics calculations. Cross section calculations were performed with the HELIOS lattice physics code, finally GenPMAXS was used to convert the cross section into the PARCS format. A simplified three dimensional 3D neutronics model of the Asco II NPP is used as a core kinetics model. A 3D cylindrical thermal–hydraulic vessel plus 1D representation of the remainder of the full plant model is used as the thermal–hydraulic model. The transient selected to ensure the model validation is an actual 50% Loss of Load. This transient is characterized by space–time effects and was used to validate different thermal–hydraulic system models for the GET university group in the past. The scenario is also good to ensure the validation of a coupled 3D neutron kinetics code since it provides a transient situation between two stable regions at 100% and 50%. From the current code versions used, some source code modifications have been carried out in order to ensure the correct feedback between thermal–hydraulic and neutron kinetics code. In that sense, a dynamic control rod movement between TRACE and PARCS has been implemented. This is a complete control rod position feedback during transient scenarios. After all the work was performed, the important TH and NK time trend parameters were compared to the plant data and the comparison was reasonable with some discrepancy, thus the developed system models and the code modifications are robust enough to be used for future safety analysis. New coupled code capability has been tested and found as a required capability, when validating 3D NK–TH coupled calculations.}, journal={Annals of Nuclear Energy}, publisher={Elsevier BV}, author={Pericas, R. and Ivanov, K. and Reventós, F. and Batet, L.}, year={2016}, month={Jan}, pages={366–374} }
@article{bennett_avramova_ivanov_2016, title={Coupled MCNP6/CTF code: Development, testing, and application}, volume={96}, ISSN={["0306-4549"]}, DOI={10.1016/j.anucene.2016.05.008}, abstractNote={This paper presents the development and testing of a high fidelity Monte Carlo based multi-physics code. The coupling was done between the Monte Carlo neutronics code MCNP6 and the thermal-hydraulic sub-channel code CTF. The coupling for the MCNP6/CTF code was done internally at the pin level. On-The-Fly cross sections were used to decrease the complexity of the coupled code as well as to decrease the memory requirement. The relaxation acceleration technique was applied to the coupled code and was shown that it could satisfy much stricter convergence criterions. The technique can also guarantee convergence and be used as a tool to decrease the computational time. The coupled code was tested against two other coupled Monte Carlo/thermal-hydraulic sub-channel codes and the results were similar. The coupled code was also tested on a full assembly problem.}, journal={ANNALS OF NUCLEAR ENERGY}, author={Bennett, A. and Avramova, M. and Ivanov, K.}, year={2016}, month={Oct}, pages={1–11} }
@article{bostelmann_strydom_reitsma_ivanov_2016, title={The IAEA coordinated research programme on HTGR uncertainty analysis: Phase I status and Ex. I-1 prismatic reference results}, volume={306}, ISSN={["1872-759X"]}, DOI={10.1016/j.nucengdes.2015.12.009}, abstractNote={The quantification of uncertainties in design and safety analysis of reactors is today not only broadly accepted, but in many cases became the preferred way to replace traditional conservative analysis for safety and licensing analysis. The use of a more fundamental methodology is also consistent with the reliable high fidelity physics models and robust, efficient, and accurate codes available today. To facilitate uncertainty analysis applications a comprehensive approach and methodology must be developed and applied, in contrast to the historical approach where sensitivity analysis were performed and uncertainties then determined by a simplified statistical combination of a few important input parameters. New methodologies are currently under development in the OECD/NEA Light Water Reactor (LWR) Uncertainty Analysis in Best-Estimate Modelling (UAM) benchmark activity. High Temperature Gas-cooled Reactor (HTGR) designs require specific treatment of the double heterogeneous fuel design and large graphite quantities at high temperatures. The IAEA has therefore launched a Coordinated Research Project (CRP) on HTGR Uncertainty Analysis in Modelling (UAM) in 2013 to study uncertainty propagation specifically in the HTGR analysis chain. Two benchmark problems are defined, with the prismatic design represented by the General Atomics (GA) MHTGR-350 and a 250 MW modular pebble bed design similar to the Chinese HTR-PM. Work has started on the first phase and the current CRP status is reported in the paper. A comparison of the Serpent and SCALE/KENO-VI reference Monte Carlo results for Ex. I-1 of the MHTGR-350 design is also included. It was observed that the SCALE/KENO-VI Continuous Energy (CE) k∞ values were 395 pcm (Ex. I-1a) to 803 pcm (Ex. I-1b) higher than the respective Serpent lattice calculations, and that within the set of the SCALE results, the KENO-VI 238 Multi-Group (MG) k∞ values were up to 800 pcm lower than the KENO-VI CE values. The use of the latest ENDF-B-VII.1 cross section library in Serpent lead to ∼180 pcm lower k∞ values compared to the older ENDF-B-VII.0 dataset, caused by the modified graphite neutron capture cross section. The fourth beta release of SCALE 6.2 likewise produced lower CE k∞ values when compared to SCALE 6.1, and the improved performance of the new 252-group library available in SCALE 6.2 is especially noteworthy. A SCALE/TSUNAMI uncertainty analysis of the Hot Full Power variant for Ex. I-1a furthermore concluded that the 238U(n,γ) (capture) and 235U(v¯) cross-section covariance matrices contributed the most to the total k∞ uncertainty of 0.58%.}, journal={NUCLEAR ENGINEERING AND DESIGN}, author={Bostelmann, Friederike and Strydom, Gerhard and Reitsma, Frederik and Ivanov, Kostadin}, year={2016}, month={Sep}, pages={77–88} }
@article{hou_choi_ivanov_2015, title={Development of an iterative diffusion-transport method based on MICROX-2 cross section libraries}, volume={77}, ISSN={["0306-4549"]}, DOI={10.1016/j.anucene.2014.11.014}, abstractNote={This paper introduces an innovative online cross section generation method, developed based on Iterative Diffusion-Transport (IDT) calculation to minimize the inconsistency and inaccuracy in determining physics parameters by feeding actual reactor core conditions into the cross section generation process. A two-dimensional (2-D) pin-by-pin lattice program, NEMA, was developed to generate assembly lattice parameters using the refined MICROX-2 cross section libraries and Nodal Expansion Method (NEM). The proposed method was verified against a 2-D miniature core (mini-core) benchmark problem. First, the few-group cross sections generated by NEMA were compared with those calculated by a Monte Carlo method code Serpent. Next, the analysis of a 2-D Light Water Reactor (LWR) mini-core benchmark problem was carried out by the nodal transport code DIF3D using few-group cross sections generated by NEMA, and the results were compared with those obtained from the Serpent full core calculation. Finally, the same benchmark problem was solved by the NEMA-DIF3D approach using the IDT coupling method. The computational benchmark calculations have shown that the homogenization technique implemented in NEMA is reliable when producing the few-group cross sections for the reactor core calculation. The IDT method also improves the eigenvalue and power distribution predictions.}, journal={ANNALS OF NUCLEAR ENERGY}, publisher={Elsevier BV}, author={Hou, J. and Choi, H. and Ivanov, K. N.}, year={2015}, month={Mar}, pages={335–342} }
@article{ngeleka_ivanov_levine_2015, title={Examination and refinement of fine energy group structure for high temperature reactor analysis}, volume={80}, ISSN={0306-4549}, url={http://dx.doi.org/10.1016/J.ANUCENE.2015.01.038}, DOI={10.1016/J.ANUCENE.2015.01.038}, abstractNote={Multi-group energy structure SHEM-281 and -361 were refined using a Contributon and Point-Wise Cross Section Driven method (CPXSD). The Contributon and Point-Wise Cross Section Driven method was derived based on the product of the forward and adjoint angular fluxes, and the point-wise cross section of important isotope/material. It is an iterative method that selects effective fine- and broad-group energy structures for a problem of interest. The two selected criteria for determining fine energy group structure were 10 pcm relative deviation of Δk/k for k-effective and 1% relative deviation for reaction rates. The energy group structure refinement was subdivided into fast, epithermal and thermal regions. Firstly, the refinement was done for fast region and a new library was created and applied in the fuel cell unit until the target criteria’s are met. Similar procedure was repeated for epithermal and thermal regions. The dominant parameters for each region are considered as required, fission reaction rate for fast region, absorption reaction rate for epithermal region and absorption and fission reaction rates for thermal region.}, journal={Annals of Nuclear Energy}, publisher={Elsevier BV}, author={Ngeleka, Tholakele P. and Ivanov, Kostadin N. and Levine, Samuel}, year={2015}, month={Jun}, pages={279–292} }
@article{ivanov_sanchez_stieglitz_ivanov_2015, title={Large-scale Monte Carlo neutron transport calculations with thermal hydraulic feedback}, volume={84}, ISSN={0306-4549}, url={http://dx.doi.org/10.1016/J.ANUCENE.2014.12.030}, DOI={10.1016/J.ANUCENE.2014.12.030}, abstractNote={Abstract The Monte Carlo method provides the most accurate description of the particle transport problem. The criticality problem is simulated by following the histories of individual particles without approximating the energy, angle or the coordinate dependence. These calculations are usually done using homogeneous thermal hydraulic conditions. This is a very crude approximation in the general case. In this paper, the method of internal coupling between neutron transport and thermal hydraulics is presented. The method is based on dynamic material distribution, where coordinate dependent temperature and density information is supplied on the fly during the transport calculation. This method does not suffer from the deficiencies characteristic of the external coupling via the input files. In latter case, the geometry is split into multiple cells having distinct temperatures and densities to supply the feedback. The possibility to efficiently simulate large scale geometries at pin-by-pin and subchannel level resolution was investigated. The Wielandt shift method for reducing the dominance ratio of the system and accelerating the fission source convergence was implemented. During the coupled iteration a detailed distribution of the fission heat deposition is required by the thermal hydraulics calculation. Providing reasonable statistical uncertainties for tallies having large numbers of bins, is a complicated task. This problem was resolved by applying the Uniform Fission Site method. Previous investigations showed that the convergence of the coupled neutron transport/thermal hydraulics calculation is limited by the statistical uncertainty and exhibits strong nonuniform behavior. The stochastic approximation scheme was used to stabilize the convergence. In combination with the Uniform Fission Site method, uniform convergence was achieved.}, journal={Annals of Nuclear Energy}, publisher={Elsevier BV}, author={Ivanov, A. and Sanchez, V. and Stieglitz, R. and Ivanov, K.}, year={2015}, month={Oct}, pages={204–219} }
@article{avramova_ivanov_kozlowski_pasichnyk_zwermann_velkov_royer_yamaji_gulliford_2015, title={Multi-physics and multi-scale benchmarking and uncertainty quantification within OECD/NEA framework}, volume={84}, ISSN={0306-4549}, url={http://dx.doi.org/10.1016/J.ANUCENE.2014.12.014}, DOI={10.1016/J.ANUCENE.2014.12.014}, abstractNote={• Presentation of latest multi-physics multi-scale NEA/OECD benchmarks. • Utilization of high-quality experimental data for detailed comparative analysis. • Including uncertainty and sensitivity analysis of modeling predictions. • Uncertainty propagation in LWR multi-physics and multi-scale simulations. The development of multi-physics multi-scale coupled methodologies for Light Water Reactor (LWR) analysis requires comprehensive validation and verification procedures, which include well-established benchmarks developed in international cooperation. The Nuclear Energy Agency (NEA) of the Organization for Economic Co-operation and Development (OECD) has provided such framework, and over the years a number of LWR benchmarks have been developed and successfully conducted. The first set of NEA/OECD benchmarks that permits testing of the neutronics/thermal–hydraulics coupling, and verifying the capability of the coupled codes to analyze complex transients with coupled core/plant interactions have been completed and documented. These benchmarks provided a validation basis for the new generation of coupled “best-estimate” codes. The above mentioned OECD/NEA LWR benchmark activities have also stimulated follow up developments and benchmarks to test these developments. The models utilized have been improved when moving from one benchmark to the next and this created a need to validate them using high-quality experimental data. Second set of the NEA/OECD benchmarks have been initiated by the Expert Group on Uncertainty Analysis in Modelling (EGUAM) at the Nuclear Science Committee (NSC), NEA/OECD to address the current trends in the development of LWR multi-physics and multi-scale modeling and simulation. These benchmarks include the following common features, which address some of the issues identified in the first set of OECD/NEA benchmarks: (a) utilization of high-quality experimental data; (b) refined local scale modeling in addition to global predictions; (c) more detailed comparisons and analysis; (d) including uncertainty and sensitivity analysis of modeling predictions. The paper presents each of these new benchmarks by providing description and discussion of comparative analysis of obtained results. Special attention is devoted to uncertainty propagation in LWR multi-physics and multi-scale simulations for design and safety evaluations.}, journal={Annals of Nuclear Energy}, publisher={Elsevier BV}, author={Avramova, M. and Ivanov, K. and Kozlowski, T. and Pasichnyk, I. and Zwermann, W. and Velkov, K. and Royer, E. and Yamaji, A. and Gulliford, J.}, year={2015}, month={Oct}, pages={178–196} }
@article{levine_blyth_ivanov_2015, title={The effect of changing enrichments on core performance}, volume={75}, ISSN={0306-4549}, url={http://dx.doi.org/10.1016/J.ANUCENE.2014.07.035}, DOI={10.1016/J.ANUCENE.2014.07.035}, abstractNote={Abstract The information presented in this paper has been developed as a follow on to two previous papers published using the same low leakage core configuration with the addition in this paper of evaluating fuel costs. The two previous publications studied the characteristics of this low leakage core with two different enrichment sets, where each enrichment set represents the three batches in the core. The purpose of the two previous papers proved the effectiveness of using the Haling Power Depletion (HPD) method as a guide. The first purpose of this paper is to extend this study to higher enrichment sets to finally attain a core having close to the highest possible cycle length. Three additional similar enrichment sets are studied increasing the number of enrichment sets to five. The ratio between the enrichment sets was maintained constant except for the highest enrichment set. This was done to increase the cycle length to approximately the longest possible cycle length of 800 days for a 1000 MWe reactor limited to a maximum 5% enrichment. The core reactor physics characteristics of these five cores are presented in this paper together with the evaluating of the fuel costs. These core characteristics include radial power fractions (RPF), Haling Power Depletion, RPF distributions, maximum pin peak powers (PPP MAX ), and other important data. The HPD RPFs of all 5 cores were similar and used to help develop the burnable poison placement designs for each core. The longest two cycles required an improved technique using more information than the HPD results to develop successful BP placement designs. Also, it was very difficult to find the correct soluble boron ppmB in the HPD input to have the Studvik HPD calculation converge. There is an error in this algorithm. The fuel costs for the five cores were calculated and the results prove that the fuel costs are lower with the cores having the longest cycle lengths. The details observed in this study are presented in this paper.}, journal={Annals of Nuclear Energy}, publisher={Elsevier BV}, author={Levine, S. and Blyth, T. and Ivanov, K.}, year={2015}, month={Jan}, pages={139–145} }
@article{georgieva_dinkov_ivanov_2014, title={A cycle-specific cross-section update for real-time simulation of VVER-1000 core}, volume={74}, ISSN={0149-1970}, url={http://dx.doi.org/10.1016/J.PNUCENE.2014.03.009}, DOI={10.1016/J.PNUCENE.2014.03.009}, abstractNote={A two-group cross-section generation methodology, cycle-specific cross-section update procedure and VVER-1000 reactor core model are described. The HELIOS lattice physics code is used to calculate the cross-section data tables according to a customized version of the cycle-specific cross-section modelling methodology of the Pennsylvania State University (PSU). A real-time version of the NEM code from PSU is developed for VVER-1000 full-scope simulator applications. The cross-section update procedure is tailored to meet cycle-specific reactor core simulation fidelity requirements as well as particular customer needs and practices. Combined with an enhanced thermal-hydraulics and Instrumentation (I&C) models the scope of Kozloduy 6 full-scope replica control room simulator is expanded to the whole range of plant operation modes ranging from cold shutdown (depressurized) state to rated power, as well as deviations from normal operating modes and even beyond-design basis accidents. The fidelity and accuracy of simulation is illustrated through comparison with plant-specific data.}, journal={Progress in Nuclear Energy}, publisher={Elsevier BV}, author={Georgieva, Emiliya L. and Dinkov, Yavor D. and Ivanov, Kostadin N.}, year={2014}, month={Jul}, pages={222–231} }
@article{hou_choi_ivanov_2014, title={ASSESSMENT OF MICROX-2 CODE WITH NEW ENDF/B-VII RELEASE 0 MASTER LIBRARIES}, volume={186}, ISSN={["1943-7471"]}, DOI={10.13182/nt12-137}, abstractNote={AbstractA lattice code, MICROX-2, was assessed for its neutronics calculation performance with new cross-section libraries. First, the new cross-section libraries were generated based on ENDF/B-VII release 0. A total of 386 nuclides were processed, including 10 thermal scattering nuclides. The updated NJOY system and MICROR code were used to process nuclear data and convert them into the MICROX-2 library format. The energy group structure of the new library was optimized for both the thermal and fast neutron spectrum systems based on the Contributon and Pointwise Cross Section Driven (CPXSD) method, resulting in a total of 1173 energy groups. Second, a series of pin-cell–level benchmark calculations was performed against experimental measurements and numerical calculations performed by the deterministic and Monte Carlo codes for multiplication factors and reaction rate ratios. Both the homogeneous and heterogeneous pin-cell calculations were conducted for 15 cases. The results of MICROX-2 calculations sho...}, number={3}, journal={NUCLEAR TECHNOLOGY}, publisher={American Nuclear Society}, author={Hou, Jia and Choi, Hangbok and Ivanov, Kostadin}, year={2014}, month={Jun}, pages={305–316} }
@article{kozlowski_wysocki_gajev_xu_downar_ivanov_magedanz_hardgrove_march-leuba_hudson_et al._2014, title={Analysis of the OECD/NRC Oskarshamn-2 BWR stability benchmark}, volume={67}, ISSN={0306-4549}, url={http://dx.doi.org/10.1016/J.ANUCENE.2013.09.028}, DOI={10.1016/J.ANUCENE.2013.09.028}, abstractNote={On February 25, 1999, the Oskarshamn-2 NPP experienced a stability event which culminated in diverging power oscillations. The event was successfully modeled by the TRACE/PARCS coupled system code, and further uncertainty analysis of the event is described in this paper. The results show very good agreement with the plant data, capturing the entire behavior of the transient including the onset of instability, growth of the oscillations, and oscillation frequency. This provides confidence in the prediction of other parameters which are not available from the plant records. The event provides coupled code validation for a challenging BWR stability event, which involves the accurate simulation of neutron kinetics (NK), thermal–hydraulics (TH), and TH/NK coupling. The success of this work has demonstrated the ability of the 3-D coupled systems code TRACE/PARCS to capture the complex behavior of BWR stability events. The problem was released as an international OECD/NEA benchmark, and it is the first benchmark based on measured plant data for a stability event with a DR greater than one.}, journal={Annals of Nuclear Energy}, publisher={Elsevier BV}, author={Kozlowski, Tomasz and Wysocki, Aaron and Gajev, Ivan and Xu, Yunlin and Downar, Thomas and Ivanov, Kostadin and Magedanz, Jeffrey and Hardgrove, Matthew and March-Leuba, Jose and Hudson, Nathanael and et al.}, year={2014}, month={May}, pages={4–12} }
@article{rosenkrantz_avramova_ivanov_prinsloo_botes_elsakhawy_2014, title={Coupled 3D neutronics/thermal hydraulics modeling of the SAFARI-1 MTR}, volume={73}, ISSN={0306-4549}, url={http://dx.doi.org/10.1016/J.ANUCENE.2014.06.018}, DOI={10.1016/J.ANUCENE.2014.06.018}, abstractNote={Abstract The purpose of this study was to develop a coupled accurate multi-physics model of the SAFARI-1 Material Testing Reactor (MTR), a facility that is used for both research and the production of medical isotopes. The model was developed as part of the SAFARI-1 benchmarking project as a cooperative effort between the Pennsylvania State University (PSU) and the South African Nuclear Energy Corporation (Necsa). It was created using a multi-physics coupling of state of the art nuclear reactor simulation tools, consisting of a neutronics code and a thermal hydraulics code. The neutronics tool used was the PSU code NEM, and the results from this component were verified using the Necsa neutronics code OSCAR-4, which is utilized for SAFARI-1 core design and fuel management. On average, the multiplication factors of the neutronics models agreed to within 5 pcm and the radial assembly-averaged powers agreed to within 0.2%. The thermal hydraulics tool used was the PSU version of COBRA-TF (CTF) sub-channel code, and the results of this component were verified against another thermal hydraulics code, the RELAP5-3D system code, used at Necsa for thermal–hydraulics analysis of SAFARI-1. Although only assembly-averaged results from RELAP5-3D were available, they fell within the range of values for the corresponding assemblies in the comprehensive CTF solution. This comparison allows for the first time to perform a quantification of steady-state errors for a low-powered MTR with an advanced thermal–hydraulic code such as CTF on a per-channel basis as compared to simpler and coarser-mesh RELAP5-3D modeling. Additionally, a new cross section representation was used to ensure that the thermal hydraulic feedback effects on the core neutronics were captured as accurately as possible. This cross section representation was applied to SAFARI-1 core calculations for the first time in this work. Such implementation helps to quantify the effect of detailed modeling of thermal–hydraulics feedback effects on neutronics results in multi-physics simulations. The outcome of the study is the intended coupled neutronics/thermal–hydraulics model of the SAFARI-1 reactor.}, journal={Annals of Nuclear Energy}, publisher={Elsevier BV}, author={Rosenkrantz, Adam and Avramova, Maria and Ivanov, Kostadin and Prinsloo, Rian and Botes, Danniëll and Elsakhawy, Khalid}, year={2014}, month={Nov}, pages={122–130} }
@inproceedings{ivanov_sanchez_ivanov_2014, title={High-Fidelity Coupled Monte-Carlo/Thermal-Hydraulics Calculations}, ISBN={9782759812691}, url={http://dx.doi.org/10.1051/SNAMC/201402304}, DOI={10.1051/SNAMC/201402304}, abstractNote={Monte Carlo methods have been used as reference reactor physics calculation tools worldwide. The advance in computer technology allows the calculation of detailed flux distributions in both space and energy. In most of the cases however, those calculations are done under the assumption of homogeneous material density and temperature distributions. The aim of this work is to develop a consistent methodology for providing realistic three-dimensional thermal-hydraulic distributions by coupling the in-house developed sub-channel code SUBCHANFLOW with the standard Monte-Carlo transport code MCNP. In addition to the innovative technique of on-the fly material definition, a flux-based weight-window technique has been introduced to improve both the magnitude and the distribution of the relative errors. Finally, a coupled code system for the simulation of steady-state reactor physics problems has been developed. Besides the problem of effective feedback data interchange between the codes, the treatment of temperature dependence of the continuous energy nuclear data has been investigated.}, booktitle={SNA + MC 2013 - Joint International Conference on Supercomputing in Nuclear Applications + Monte Carlo}, publisher={EDP Sciences}, author={Ivanov, Aleksandar and Sanchez, Victor and Ivanov, Kostadin}, editor={Caruge, D. and Calvin, C. and Diop, C.M. and Malvagi, F. and Trama, J.-C.Editors}, year={2014} }
@article{ivanov_sanchez_stieglitz_ivanov_2014, title={Internal multi-scale multi-physics coupled system for high fidelity simulation of light water reactors}, volume={66}, ISSN={0306-4549}, url={http://dx.doi.org/10.1016/J.ANUCENE.2013.12.003}, DOI={10.1016/J.ANUCENE.2013.12.003}, abstractNote={In order to increase the accuracy and the degree of spatial and energy resolution of core design studies, coupled 3D neutronic (multi-group deterministic and continuous energy Monte-Carlo) and 3D thermal–hydraulic (CFD and subchannel) codes are being developed worldwide. At KIT, both deterministic and Monte-Carlo codes were coupled with subchannel codes and applied to predict the safety-related design parameters such as minimal critical power ratio (MCPR), maximal cladding and fuel temperature, departure from nuclide boiling ratio (DNBR). These coupling approaches were revised and considerably improved. Innovative method of internal on-the-fly thermal feedback interchange between the codes was implemented. It no longer relies on explicit material definitions and allows the modeling of temperature and density distributions based on the cell coordinates. In contrast to all existing coupled schemes, this method uses only standard MCNP geometry input and requires only proper definition of the geometrical dimensions. The initial material definition is arbitrary and is determined on-the-fly during the neutron transport by the thermal–hydraulic feedback. Another key issue addressed is the optimal application of parallel computing and the implementation of less time consuming tally estimators. Using multi-processor computer architectures and implementing collision density flux estimator, it is possible to reduce the Monte-Carlo running time and obtain converged results within reasonable time limit. The coupled calculation was accelerated further, by implementing stochastic approximation-based relaxation technique. Further, it is shown that large fuel assemblies can be analyzed on subchannel level.}, journal={Annals of Nuclear Energy}, publisher={Elsevier BV}, author={Ivanov, A. and Sanchez, V. and Stieglitz, R. and Ivanov, K.}, year={2014}, month={Apr}, pages={104–112} }
@inproceedings{colameco_ivanov_beacon_ivanov_2014, title={Iterative Transport-Diffusion Methodology For LWR Core Analysis}, ISBN={9782759812691}, url={http://dx.doi.org/10.1051/SNAMC/201401104}, DOI={10.1051/SNAMC/201401104}, abstractNote={This paper presents an update on the development of an advanced methodology for core calculations that uses local heterogeneous solutions for on-the-fly nodal cross-section generation. The Iterative Transport-Diffusion Method is an embedded transport approach that is expected to provide results with near 3D transport accuracy for a fraction of the time required by a full 3D transport method. In this methodology, the infinite environment used for homogenized nodal cross-section generation is replaced with a simulated 3D environment of the diffusion calculation. This update focuses on burnup methodology, axial leakage and 3D modeling.}, booktitle={SNA + MC 2013 - Joint International Conference on Supercomputing in Nuclear Applications + Monte Carlo}, publisher={EDP Sciences}, author={Colameco, David and Ivanov, Boyan D. and Beacon, Daniel and Ivanov, Kostadin N.}, editor={Caruge, D. and Calvin, C. and Diop, C.M. and Malvagi, F. and Trama, J.-C.Editors}, year={2014} }
@article{ghrayeb_ougouag_ouisloumen_ivanov_2014, title={Multi-group formulation of the temperature-dependent resonance scattering model and its impact on reactor core parameters}, volume={63}, ISSN={0306-4549}, url={http://dx.doi.org/10.1016/J.ANUCENE.2013.07.031}, DOI={10.1016/J.ANUCENE.2013.07.031}, abstractNote={A multi-group formulation for the exact neutron elastic scattering kernel is developed. It incorporates the neutron up-scattering effects stemming from lattice atoms thermal motion and it accounts for them within the resulting effective nuclear cross-section data. The effects pertain essentially to resonant scattering off of heavy nuclei. The formulation, implemented into a standalone code, produces effective nuclear scattering data that are then supplied directly into the DRAGON lattice physics code where the effects on Doppler reactivity and neutron flux are demonstrated. The correct accounting for the crystal lattice effects influences the estimated values for the probability of neutron absorption and scattering, which in turn affect the estimation of core reactivity and burnup characteristics. The results show an increase in values of Doppler temperature feedback coefficients up to −10% for UOX and MOX LWR fuels compared to the corresponding values derived using the traditional asymptotic elastic scattering kernel. This paper also summarizes research performed to date on this topic.}, journal={Annals of Nuclear Energy}, publisher={Elsevier BV}, author={Ghrayeb, Shadi Z. and Ougouag, Abderrafi M. and Ouisloumen, Mohamed and Ivanov, Kostadin N.}, year={2014}, month={Jan}, pages={751–762} }
@article{hou_choi_ivanov_2014, title={Self-shielding models of MICROX-2 code: Review and updates}, volume={64}, ISSN={["0306-4549"]}, DOI={10.1016/j.anucene.2013.10.005}, abstractNote={Abstract The MICROX-2 is a transport theory code that solves for the neutron slowing-down and thermalization equations of a two-region lattice cell. The MICROX-2 code has been updated to expand its application to advanced reactor concepts and fuel cycle simulations, including generation of new fine-group cross section libraries based on ENDF/B-VII. In continuation of previous work, the MICROX-2 methods are reviewed and updated in this study, focusing on its resonance self-shielding and spatial self-shielding models for neutron spectrum calculations. The improvement of self-shielding method was assessed by a series of benchmark calculations against the Monte Carlo code, using homogeneous and heterogeneous pin cell models. The results have shown that the implementation of the updated self-shielding models is correct and the accuracy of physics calculation is improved. Compared to the existing models, the updates reduced the prediction error of the infinite multiplication factor by ∼0.1% and ∼0.2% for the homogeneous and heterogeneous pin cell models, respectively, considered in this study.}, journal={ANNALS OF NUCLEAR ENERGY}, publisher={Elsevier BV}, author={Hou, J. and Choi, H. and Ivanov, K. N.}, year={2014}, month={Feb}, pages={256–263} }
@article{clifford_ivanov_avramova_2013, title={A multi-scale homogenization and reconstruction approach for solid material temperature calculations in prismatic high temperature reactor cores}, volume={256}, ISSN={0029-5493}, url={http://dx.doi.org/10.1016/j.nucengdes.2012.11.016}, DOI={10.1016/j.nucengdes.2012.11.016}, abstractNote={Traditional full-core heat transfer analysis of high temperature reactors uses effective coarse mesh parameters that are typically derived from a priori analysis and/or simplified analytical models that approximate the subscale temperature response. Consequently, different assumptions are made on each spatial scale, potentially yielding inconsistent solution methods with large associated uncertainties. In contrast, homogenized cross-sections used in full-core neutronics analysis are obtained using consistent homogenization techniques applied in conjunction with unit cell calculations. This approach has been proven both efficient and accurate. It is therefore surprising that formal homogenization techniques are rarely used in heat transfer analysis of nuclear reactors. In this work we take advantage of distinct unit cells that can be identified on each spatial scale in the MHTGR reactor core with the view to develop a consistent and accurate methodology for constructing hierarchical coarse mesh models for solid heat conduction in this reactor type. Three techniques have been used: formal multi-scale expansion homogenization is applied to obtain effective unit cell thermodynamic parameters; coarse mesh temperature discontinuities are defined to ensure continuity of the fine-scale temperatures at interfaces; and reduced order models for the time-dependent temperature response of the unit cells are obtained using proper orthogonal decomposition applied to detailed unit cell simulation results. The result is an efficient method, aimed toward unstructured CFD frameworks, that accurately captures coarse mesh temperatures with the capability of fully reconstructing the fine scale solution at any hierarchical level. The advantages of this method are illustrated for a small prismatic HTGR core using a cascaded solution approach. Starting at the finest scale, the TRISO coated particles, high resolution unit cell calculations are performed in a hierarchical fashion to build up a library of homogenized coarse mesh parameters and reduced order models, which are then used for the full-core heat conduction analysis. We demonstrate the accuracy and efficiency of the method by comparing results for a typical HTR power excursion transient against detailed reference solutions.}, journal={Nuclear Engineering and Design}, publisher={Elsevier BV}, author={Clifford, Ivor and Ivanov, Kostadin N. and Avramova, Maria N.}, year={2013}, month={Mar}, pages={1–13} }
@article{ivanov_sanchez_stieglitz_ivanov_2013, title={High fidelity simulation of conventional and innovative LWR with the coupled Monte-Carlo thermal-hydraulic system MCNP-SUBCHANFLOW}, volume={262}, ISSN={0029-5493}, url={http://dx.doi.org/10.1016/J.NUCENGDES.2013.05.008}, DOI={10.1016/J.NUCENGDES.2013.05.008}, abstractNote={Abstract In order to increase the accuracy and the degree of spatial and energy resolution of core design studies, coupled 3D neutronic (multi-group deterministic and continuous energy Monte-Carlo) and 3D thermal-hydraulic (CFD and sub-channel) codes are being developed worldwide. At KIT, both deterministic and Monte-Carlo codes were coupled with sub-channel codes and applied to predict the safety-related design parameters such as critical power ratio, maximal cladding, fuel temperature and DNB. These coupling approaches were revised and improved based on the experience gained. One particular example is replacing COBRA-TF with SUBCHANFLOW, an in-house developed sub-channel code, in the COBRA-TF/MCNP coupling, accompanied with new way of radial mapping between the neutronic and thermal-hydraulic domains. The new coupled system MCNP5/SUBCHANFLOW makes it possible to investigate a variety of fuel assembly types. Key issues in such a coupled system are the implementation of the thermal-hydraulic/neutronic feedback mechanisms, the precision of the Monte-Carlo solutions, and the supervision of convergence during the iterative solution process. Another key issue considered is the optimal application of parallel computing. Using multi-processor computer architectures, it is possible to reduce the Monte-Carlo running time and obtain converged results within reasonable time limit. In particular, it is shown that by exploiting the capabilities of multi-processor calculation, large fuel assemblies on a pin-by-pin basis with a resolution at sub-channel level can be analyzed. One of the most important issues addressed in the current work is the temperature effects on nuclear data. For the particular studies pseudo material mixing approach was used to account for the temperature dependence of the nuclear data.}, journal={Nuclear Engineering and Design}, publisher={Elsevier BV}, author={Ivanov, A. and Sanchez, V. and Stieglitz, R. and Ivanov, K.}, year={2013}, month={Sep}, pages={264–275} }
@article{espel_avramova_ivanov_misu_2013, title={New developments of the MCNP/CTF/NEM/NJOY code system – Monte Carlo based coupled code for high accuracy modeling}, volume={51}, ISSN={0306-4549}, url={http://dx.doi.org/10.1016/j.anucene.2012.06.031}, DOI={10.1016/j.anucene.2012.06.031}, abstractNote={High accuracy code systems are necessary to model core environments with considerable geometry complexity and great material heterogeneity. These features are typical of current and innovative nuclear reactor core designs. Advanced methodologies and state-of-the art coupled code systems must be put into practice in order to model with high accuracy these challenging core designs. The presented research comprises the development and implementation of the thermal–hydraulic feedback to the Monte Carlo method and of speed-up mechanisms to accelerate the Monte Carlo criticality calculation. Coupled Monte-Carlo calculations can serve as reference solutions for verifying high-fidelity coupled deterministic neutron transport methods with detailed and accurate thermal–hydraulic models. The development and verification of such reference high-fidelity coupled multi-physics scheme is performed at the Pennsylvania State University (PSU) in cooperation with AREVA, AREVA NP GmbH in Erlangen, Germany, on the basis of MCNP5, NEM, NJOY and COBRA-TF (CTF) computer codes. This paper presents the latest studies and ameliorations developed to this coupled hybrid system, which includes a new methodology for generation and interpolation of Temperature-Dependent Thermal Scattering Cross Section Libraries for MCNP5, a comparison between sub-channel approaches, and acceleration schemes.}, journal={Annals of Nuclear Energy}, publisher={Elsevier BV}, author={Espel, Federico Puente and Avramova, Maria N. and Ivanov, Kostadin N. and Misu, Stefan}, year={2013}, month={Jan}, pages={18–26} }
@article{ellis_watson_ivanov_2013, title={Progress in the development of an implicit steady state solution in the coupled code TRACE/PARCS}, volume={66}, ISSN={0149-1970}, url={http://dx.doi.org/10.1016/J.PNUCENE.2013.02.009}, DOI={10.1016/J.PNUCENE.2013.02.009}, abstractNote={This paper describes the implementation of an implicit steady state solution method in the TRAC/RELAP Advanced Computational Engine (TRACE) thermal-hydraulics system code and Purdue Advanced Reactor Core Simulator (PARCS) code with the goal of improving solution stability and efficiency. The implicit steady state solution method has been implemented within the framework of the existing psuedo-transient solution method in TRACE and includes time-dependent thermal-hydraulic and heat transfer equations and time-independent neutron diffusion equations. The implicit solution method uses Newton's method to solve the thermal-hydraulic, heat transfer, and neutron diffusion equations during each psuedo-time step with an analytic construction of the Jacobian matrix. The linear system associated with each iteration of Newton's method is solved using a preexisting LU decomposition algorithm in TRACE. The implicit steady state solution was evaluated using two different meshes overlaid on a two-phase pipe model closely matching a boiling water reactor hydraulic channel. The implicit solution method reproduces the correct steady state solution for varying time step sizes for each mesh. An evaluation of the CPU runtime required to complete a steady state calculation using the implicit method shows that the well-developed and optimized explicit solution method currently requires less CPU runtime than the implicit solution method which has yet to be optimized. These results direct future development of the implicit solution method towards optimization strategies to reduce CPU runtime.}, journal={Progress in Nuclear Energy}, publisher={Elsevier BV}, author={Ellis, M.S. and Watson, J. and Ivanov, K.}, year={2013}, month={Jul}, pages={1–12} }
@article{levine_blyth_ivanov_2013, title={Understanding using the Haling Power Depletion (HPD) as a guide for designing PWR cores}, volume={53}, ISSN={0306-4549}, url={http://dx.doi.org/10.1016/j.anucene.2012.09.025}, DOI={10.1016/j.anucene.2012.09.025}, abstractNote={Abstract The Pennsylvania State University (PSU) is using the university version of the Studsvik Scandpower Code System (CMS) for research and education purposes. Preparations have been made to incorporate the CMS codes into the PSU Nuclear Engineering graduate class “Nuclear Fuel Management” course. The information presented in this paper has been developed during two phases of preparation of the material for the course. In the first phase, the Haling Power Depletion (HPD) was presented in the course for the first time. The HPD method has been criticized as not valid by many in the field even though it has been successfully applied at PSU for the past 20 years. It was noticed during the first phase that the radial power distribution (RPD) for low leakage cores during depletion remained similar to that of the HPD during most of the cycle and have close maximum normalized power values (NPmax’s) and cycle lengths. Thus, the Haling Power Depletion (HPD) may be used as a guide for conveniently designing mainly low leakage PWR cores because the HPD and actual core are similar. Studies were then made to better understand the HPD. Many different core configurations can be computed quickly with the HPD without using Burnable Poisons (BPs) to produce several excellent low leakage core configurations that are viable for power production. The first phase covered the solution to design a low leakage core for a cycle length of 16 GWD/MTU. Once the HPD core configuration is chosen as a potential core, it is followed by establishing a BP design to prevent violating any of the safety constraints during depletion. The problem to design a core for a cycle length of 18 GWD/MTU was not covered in the first phase. However, this cycle length has now been found to be beyond the scope of the course. Consequently, the cycle length is now assigned to be 17 GWD/MTU and the solution to this problem is covered in the second phase. Adding more BPs in this core for a solution than in phase one was more difficult to design. Consequently, techniques using the HPD results have been developed to help guide efficiently the placing of BPs in the core for a good solution. In this paper, the results of both phases of preparation for the course are presented.}, journal={Annals of Nuclear Energy}, publisher={Elsevier BV}, author={Levine, S. and Blyth, T. and Ivanov, K.}, year={2013}, month={Mar}, pages={120–128} }
@article{gomez-torres_sanchez-espinoza_ivanov_macian-juan_2012, title={DYNSUB: A high fidelity coupled code system for the evaluation of local safety parameters – Part I: Development, implementation and verification}, volume={48}, ISSN={0306-4549}, url={http://dx.doi.org/10.1016/j.anucene.2012.05.011}, DOI={10.1016/j.anucene.2012.05.011}, abstractNote={DYNSUB is a novel two-way pin-based coupling of the simplified transport (SP3) version of DYN3D with the sub-channel code SUBCHANFLOW. The new coupled code system allows for a more realistic description of the core behaviour under steady state and transient conditions. The details of the developed internal coupling approach of both codes together with its implementation are presented and discussed. Comparisons of the results predicted by DYNSUB with the ones of coarser coupled solutions have shown very good agreement in the global parameters (keff and power distribution at steady state and position and magnitude of the power peak in the transient cases) validating the correctness of the coupling strategy. At local level however, important (and expected) deviations in the local safety parameters (maximal clad, fuel and moderator temperatures) have arisen. Differences up to 150 K in the centreline fuel rod temperature were found. It demonstrates the novel capabilities of the developed coupled system DYNSUB. The more detailed coupling solution had also an important impact on the convergence process, mainly in the neutronics internal convergence due to a smoother gradient on the thermal-hydraulics feedback parameters between neighbour sub-channels, compared to the gradient between assembly level channels. DYNSUB has successfully been applied to analyze the behaviour of one eight of a PWR core during a REA transient by a pin-by-pin simulation consisting of a huge amount of nodes.}, journal={Annals of Nuclear Energy}, publisher={Elsevier BV}, author={Gomez-Torres, Armando Miguel and Sanchez-Espinoza, Victor Hugo and Ivanov, Kostadin and Macian-Juan, Rafael}, year={2012}, month={Oct}, pages={108–122} }
@article{gomez-torres_sanchez-espinoza_ivanov_macian-juan_2012, title={DYNSUB: A high fidelity coupled code system for the evaluation of local safety parameters – Part II: Comparison of different temporal schemes}, volume={48}, ISSN={0306-4549}, url={http://dx.doi.org/10.1016/j.anucene.2012.05.033}, DOI={10.1016/j.anucene.2012.05.033}, abstractNote={DYNSUB is a novel two-way pin-based coupling of the simplified transport (SP3) version of DYN3D with the subchannel code SUBCHANFLOW. The new coupled code system allows for a more realistic description of the core behaviour under steady state and transients conditions, and has been widely described in Part I of this paper. Additionally to the explicit coupling developed and described in Part I, a nested loop iteration or fixed point iteration (FPI) is implemented in DYNSUB. A FPI is not an implicit scheme but approximates it by adding an iteration loop to the current explicit scheme. The advantage of the method is that it allows the use of larger time steps; however the nested loop iteration could take much more time in getting a converged solution that could be less efficient than the explicit scheme with small time steps. A comparison of the two temporal schemes is performed. The results using FPI are very promising and represent a very good option in order to optimize computational times without losing accuracy. However it is also shown that a FPI scheme can produce inaccurate results if the time step is not chosen in agreement with the analyzed transient.}, journal={Annals of Nuclear Energy}, publisher={Elsevier BV}, author={Gomez-Torres, Armando Miguel and Sanchez-Espinoza, Victor Hugo and Ivanov, Kostadin and Macian-Juan, Rafael}, year={2012}, month={Oct}, pages={123–129} }
@article{ghrayeb_ouisloumen_ougouag_ivanov_2011, title={Deterministic modeling of higher angular moments of resonant neutron scattering}, volume={38}, ISSN={0306-4549}, url={http://dx.doi.org/10.1016/j.anucene.2011.04.025}, DOI={10.1016/j.anucene.2011.04.025}, abstractNote={An exact scattering kernel formulation for anisotropic scattering up to angular order 10 has been developed and implemented into a deterministic code. The effects of accounting for lattice dynamics on the modeling of neutron scattering in 235U, 238U, 238Pu, and other nuclides have been demonstrated. The new formulation essentially reproduces other investigators previous results for isotropic scattering and quantifies the departures from the isotropic values when higher angular orders are accounted for. The correct accounting for the lattice effects influences the estimated values for the probability of neutron absorption and scattering, which in turn affect the estimation of core reactivity and burnup characteristics. It is shown that, when using the exact scattering kernel formulation, the probability for upscattering significantly increases with increasing temperatures. For example, upscattering for 238U from below the 20.67 eV resonance increases from 5.57% at 300 K to 30.41% at 1000 K, respectively. Thus, it is shown that the exact scattering kernel is strongly sensitive to temperature, a fact of major importance for High Temperature Reactor fuels. The slowing down process is important in thermal reactors because it results in the neutrons entering the thermal energy range in which the majority of fission events occur. Correctly modeling the slowing down and hence slowing down source into the thermal energy range and consequently allowing the correct modeling of the thermal energy neutron fluxes (or the correct thermal range portion of the spectrum) is paramount to the correct prediction of criticality and safety features such as the Doppler effect. These artifacts are important for all thermal spectrum reactors. In High Temperature Reactors such as the NGNP or the Deep Burn HTR these effects are even more important.}, number={10}, journal={Annals of Nuclear Energy}, publisher={Elsevier BV}, author={Ghrayeb, Shadi Z. and Ouisloumen, Mohamed and Ougouag, Abderrafi M. and Ivanov, Kostadin N.}, year={2011}, month={Oct}, pages={2291–2297} }
@article{espel_tippayakul_ivanov_misu_2011, title={MCOR – Monte Carlo depletion code for reference LWR calculations}, volume={38}, ISSN={0306-4549}, url={http://dx.doi.org/10.1016/j.anucene.2010.12.010}, DOI={10.1016/j.anucene.2010.12.010}, abstractNote={Abstract The MCOR (MCnp-kORigen) code system is a Monte Carlo based depletion system for reference fuel assembly and core calculations. The MCOR code is designed as an interfacing code that provides depletion capability to the LANL Monte Carlo code by coupling two codes: MCNP5 with the AREVA NP depletion code, KORIGEN. The physical quality of both codes is unchanged. The MCOR code system has been maintained and continuously enhanced since it was initially developed and validated. The verification of the coupling was made by evaluating the MCOR code against similar sophisticated code systems like MONTEBURNS, OCTOPUS and TRIPOLI-PEPIN. After its validation, the MCOR code has been further improved with important features. The MCOR code presents several valuable capabilities such as: (a) a predictor–corrector depletion algorithm, (b) utilization of KORIGEN as the depletion module, (c) individual depletion calculation of each burnup zone (no burnup zone grouping is required, which is particularly important for the modeling of gadolinium rings), and (d) on-line burnup cross-section generation by the Monte Carlo calculation for 88 isotopes and usage of the KORIGEN libraries for PWR and BWR typical spectra for the remaining isotopes. Besides the just mentioned capabilities, the MCOR code newest enhancements focus on the possibility of executing the MCNP5 calculation in sequential or parallel mode, a user-friendly automatic re-start capability, a modification of the burnup step size evaluation, and a post-processor and test-matrix, just to name the most important. The article describes the capabilities of the MCOR code system; from its design and development to its latest improvements and further ameliorations. Additionally, several interesting studies performed with MCOR are explained; its validation against MONTEBURNS, OCTOPUS and TRIPOLI-PEPIN; sensitivity studies with different voids and cross-section libraries; detailed comparison against measurement. Finally, the MCOR is used as a reference tool for benchmarking deterministic codes on the example of qualification of the spectral codes (APOLLO2-A and CASMO-4).}, number={4}, journal={Annals of Nuclear Energy}, publisher={Elsevier BV}, author={Espel, Federico Puente and Tippayakul, Chanatip and Ivanov, Kostadin and Misu, Stefan}, year={2011}, month={Apr}, pages={731–741} }
@article{mphahlele_ougouag_ivanov_gougar_2011, title={Spectral zone selection methodology for pebble bed reactors}, volume={38}, ISSN={0306-4549}, url={http://dx.doi.org/10.1016/j.anucene.2010.08.014}, DOI={10.1016/j.anucene.2010.08.014}, abstractNote={A methodology is developed for determining boundaries of spectral zones for pebble bed reactors. A spectral zone is defined as a region made up of a number of nodes whose characteristics are collectively similar and that are assigned the same few-group diffusion constants. The spectral zones are selected in such a manner that the difference (error) between the reference transport solution and the diffusion code solution takes a minimum value. This is achieved by choosing spectral zones through optimally minimizing this error. The objective function for the optimization algorithm is the total reaction rate error, which is defined as the sum of the leakage, absorption and fission reaction rates errors in each zone. The selection of these spectral zones is such that the core calculation results based on diffusion theory are within an acceptable tolerance as compared to a proper transport reference solution. Through this work, a consistent approach for identifying spectral zones that yield more accurate diffusion results is introduced.}, number={1}, journal={Annals of Nuclear Energy}, publisher={Elsevier BV}, author={Mphahlele, Ramatsemela and Ougouag, Abderrafi M. and Ivanov, Kostadin N. and Gougar, Hans D.}, year={2011}, month={Jan}, pages={80–87} }
@article{espel_tarantola_ghrayeb_ivanov_2010, title={Application of Global Sensitivity Analysis to Nuclear Reactor Calculations}, volume={2}, ISSN={1877-0428}, url={http://dx.doi.org/10.1016/j.sbspro.2010.05.199}, DOI={10.1016/j.sbspro.2010.05.199}, abstractNote={The studies presented in this paper, describe the application of global sensitivity analysis to the modeling of nuclear reactor physics for better model understanding. Specifically, we investigate how much criticality conditions are affected by uncertainties in various inputs, including nuclear cross-sections, at different energies, from several isotopes in the fuel, the absorber and the moderator. The sensitivity analysis uses the Sobol’ and Jansen formulas, which allow us to estimate, for each uncertain input, its main effect, its total effect (i.e. the overall effect, which includes all the interactions, at any order, with all the other uncertain inputs), and all two-way interactions among all possible pairs of uncertain inputs. The sensitivity analysis consists of a number of model simulations, which are performed using Monte Carlo code MCNP5.}, number={6}, journal={Procedia - Social and Behavioral Sciences}, publisher={Elsevier BV}, author={Espel, F. Puente and Tarantola, S. and Ghrayeb, S. and Ivanov, K.}, year={2010}, pages={7726–7727} }
@article{kriangchaiporn_ivanov_haghighat_sears_2010, title={Transport model based on three-dimensional cross-section generation for TRIGA core analysis}, volume={37}, ISSN={0306-4549}, url={http://dx.doi.org/10.1016/j.anucene.2010.04.011}, DOI={10.1016/j.anucene.2010.04.011}, abstractNote={Three-dimensional (3-D) transport model for the Pennsylvania State University Breazeale Reactor (PSBR) core analysis has been developed based on the discrete ordinates (S n ) method. The effective fine- and broad-group structures for the TRIGA cross-section libraries were selected based on CPXSD (Contributon and Point-wise Cross-Section Driven) methodology. The study shows results of the following effective broad-group energy structures – a 12-group structure in 2-D geometry vs. a 26-group structure in 3-D geometry. Different 3-D pin/core configurations were used to verify and validate the selected effective group structures. The results agree with continuous energy cross-section Monte Carlo calculations for eigenvalues and normalized pin-power distributions, which are used as a reference in this research.}, number={9}, journal={Annals of Nuclear Energy}, publisher={Elsevier BV}, author={Kriangchaiporn, N. and Ivanov, K. and Haghighat, A. and Sears, C.F.}, year={2010}, month={Sep}, pages={1254–1260} }
@article{avramova_ivanov_2010, title={Verification, validation and uncertainty quantification in multi-physics modeling for nuclear reactor design and safety analysis}, volume={52}, ISSN={0149-1970}, url={http://dx.doi.org/10.1016/j.pnucene.2010.03.009}, DOI={10.1016/j.pnucene.2010.03.009}, abstractNote={Abstract The qualification procedure of coupled multi-physics code systems is based on the qualification framework (verification and validation) of separate physics models/codes, and includes in addition Verification and Validation (V&V) of the coupling methodologies of the different physics models. The extended verification procedure involves testing the functionality, the data exchange between different physics models, and their coupling, which is designed to model combined effects determined by the interaction of models. The extended validation procedure compares the predictions from coupled multi-physics code systems to available measured data and reference results. It is important to emphasize that such validation should be based on a multi-level approach similar to the one utilized in validating coupled neutronics/thermal–hydraulics codes in international standard problems. Appropriate benchmarks have been developed in international co-operation led by the Nuclear Energy Agency (NEA) of the Organization for Economic Cooperation and Development (OECD) that permits testing the neutronics/thermal–hydraulics coupling, and verifying the capability of the coupled codes to analyze complex transients with coupled core/plant interactions. This paper describes the above-mentioned multi-level V&V approach along with examples of the OECD benchmarks. In recent years there has been an increasing demand from nuclear research, industry, safety, and regulation for best estimate predictions to be provided with their confidence bounds. The ongoing OECD Light Water Reactor (LWR) Uncertainty Analysis in Modeling (UAM) benchmark activities contribute to establishing an unified framework to estimate safety margins supplemented by Uncertainty Quantification (UQ), which would provide more realistic, complete and logical measure of reactor safety. The paper describes the progress of the OECD LWR UAM benchmark. This activity is designed to address current regulation needs and issues related to practical implementation of risk informed regulation. Establishing such internationally accepted LWR UAM benchmark framework offers the possibility to accelerate the licensing process when using best estimate methods.}, number={7}, journal={Progress in Nuclear Energy}, publisher={Elsevier BV}, author={Avramova, Maria N. and Ivanov, Kostadin N.}, year={2010}, month={Sep}, pages={601–614} }
@article{xu_downar_walls_ivanov_staudenmeier_march-lueba_2009, title={Application of TRACE/PARCS to BWR stability analysis}, volume={36}, ISSN={0306-4549}, url={http://dx.doi.org/10.1016/j.anucene.2008.12.022}, DOI={10.1016/j.anucene.2008.12.022}, abstractNote={The work described here is the validation of TRACE/PARCS for Boiling Water Reactor stability analysis. A stability methodology was previously developed, verified, and validated using data from the OECD Ringhals stability benchmark. The work performed here describes the application of TRACE/PARCS to all the stability test points from cycle 14 of the Ringhals benchmark. The benchmark points from cycle 14 were performed using a half-core symmetric, 325 channel TRACE model. Several parametric studies are performed on test point 10 of cycle 14. Two temporal difference methods, Semi-Implicit method (SI) and Stability Enhanced Two Step (SETS) method are applied to three different mesh sizes in heated channels with series of time step sizes. The results show that the SI method has a smaller numerical damping than the SETS method. When applying the SI method with adjusted mesh and Courant time step sizes (the largest time step size under the Courant limit), the numerical damping is minimized, and the predicted Decay Ratio (DR) agrees well with the reference values which were obtained from the measured noise signal. The SI method with adjusted mesh and Courant time step size is then applied to all test points of cycle 14 with three types of initiating perturbations, control rod (CR), pressure perturbation, and noise simulation (NS). There is good agreement between the decay ratios and frequencies predicted by TRACE/PARCS and those from the plant measurements. Sensitivities were also performed to investigate the impact on the decay ratio and natural frequency of the heat conductivity of the gap between fuel and clad, as well as the impact of the pressure loss coefficient of spacers.}, number={3}, journal={Annals of Nuclear Energy}, publisher={Elsevier BV}, author={Xu, Yunlin and Downar, Thomas and Walls, R. and Ivanov, K. and Staudenmeier, J. and March-Lueba, J.}, year={2009}, month={Apr}, pages={317–323} }
@article{xulubana_tippayakul_ivanov_levine_mahgerefteh_2008, title={Accuracy evaluation of pin exposure calculations in current LWR core design codes}, volume={35}, ISSN={0306-4549}, url={http://dx.doi.org/10.1016/j.anucene.2007.07.005}, DOI={10.1016/j.anucene.2007.07.005}, abstractNote={Abstract The deregulated utility environment and better utilization of fuel assemblies in nuclear power plants has allowed designers to burn fuel assemblies to maximum allowable exposures. Any uncertainties, associated with the technical approach and numerical methods used to perform pin exposure calculations may cause either peak power exposure to exceed the Nuclear Regulatory Commission (NRC) exposure limit or lead to excessive conservatism and thus inefficient fuel utilization. In this work, a Monte Carlo based coupled depletion code (MCNP5/ORIGEN-S) is utilized to provide reference solutions in order to assess the accuracy of pin power and pin exposure reconstruction methods in the current commercial and licensed three-dimensional (3D) nodal Light Water Reactor (LWR) core design codes. The developed at the Pennsylvania State University (PSU) MCNP5/ORIGEN-S coupled depletion code system was validated using measured data from the PSU TRIGA research reactor critical experiments. A number of test cases (modeling benchmarks) representative of LWRs were developed starting from the least complex model towards more complicated and more realistic models. The accuracy evaluation of the pin reconstruction methods was performed by using the CASMO-4 and SIMULATE-3 codes as the representative of current commercial LWR core design systems. Two-dimensional (2D) transport calculations with the TRITON module from the SCALE5 package were employed to produce the spectrum averaged cross-section libraries as a function of burnup for ORIGEN-S calculations. The burnup dependent cross-section libraries are specifically generated for each lattice configuration type. For the MCNP5 calculations continuous cross-section libraries for different isotopes at hot operating temperatures are generated and subsequently utilized. Realistic lattice configurations of the GE13 BWR fuel assemblies (unrodded and rodded) depleted under operating conditions were studied in this research because of their heterogeneous nature. The 2D model test cases are constructed prior to 3D model test cases to investigate in a consistent manner the approximations involved in pin reconstruction methods of the current commercial LWR core design codes. Discrepancies more than 11% were observed between the pin-wise power and exposure data calculated by the two different methods. The statistical uncertainties in the Monte Carlo calculations were analyzed and addressed. The statistical uncertainties in the MCNP5 results remain essentially less than 2% throughout the depletion history and there was no noticeable propagation with burnup.}, number={3}, journal={Annals of Nuclear Energy}, publisher={Elsevier BV}, author={Xulubana, Vuyani and Tippayakul, Chanatip and Ivanov, Kostadin and Levine, Samuel H. and Mahgerefteh, Moussa}, year={2008}, month={Mar}, pages={414–424} }
@article{tippayakul_ivanov_frederick sears_2008, title={Development of a practical Monte Carlo based fuel management system for the Penn State University Breazeale Research Reactor (PSBR)}, volume={35}, ISSN={0306-4549}, url={http://dx.doi.org/10.1016/j.anucene.2007.07.013}, DOI={10.1016/j.anucene.2007.07.013}, abstractNote={Abstract A practical fuel management system for the he Pennsylvania State University Breazeale Research Reactor (PSBR) based on the advanced Monte Carlo methodology was developed from the existing fuel management tool in this research. Several modeling improvements were implemented to the old system. The improved fuel management system can now utilize the burnup dependent cross section libraries generated specifically for PSBR fuel and it is also able to update the cross sections of these libraries by the Monte Carlo calculation automatically. Considerations were given to balance the computation time and the accuracy of the cross section update. Thus, certain types of a limited number of isotopes, which are considered “important”, are calculated and updated by the scheme. Moreover, the depletion algorithm of the existing fuel management tool was replaced from the predictor only to the predictor-corrector depletion scheme to account for burnup spectrum changes during the burnup step more accurately. An intermediate verification of the fuel management system was performed to assess the correctness of the newly implemented schemes against HELIOS. It was found that the agreement of both codes is good when the same energy released per fission (Q values) is used. Furthermore, to be able to model the reactor at various temperatures, the fuel management tool is able to utilize automatically the continuous cross sections generated at different temperatures. Other additional useful capabilities were also added to the fuel management tool to make it easy to use and be practical. As part of the development, a hybrid nodal diffusion/Monte Carlo calculation was devised to speed up the Monte Carlo calculation by providing more converged initial source distribution for the Monte Carlo calculation from the nodal diffusion calculation. Finally, the fuel management system was validated against the measured data using several actual PSBR core loadings. The agreement of the predicted core excess reactivities and the measured values is found to be good considering the measurement uncertainties.}, number={3}, journal={Annals of Nuclear Energy}, publisher={Elsevier BV}, author={Tippayakul, Chanatip and Ivanov, Kostadin and Frederick Sears, C.}, year={2008}, month={Mar}, pages={539–551} }
@article{tyobeka_ivanov_pautz_2008, title={Evaluation of PBMR control rod worth using full three-dimensional deterministic transport methods}, volume={35}, ISSN={0306-4549}, url={http://dx.doi.org/10.1016/j.anucene.2007.11.008}, DOI={10.1016/j.anucene.2007.11.008}, abstractNote={It is a well known fact that the neutron diffusion theory fails in the vicinity of strongly absorbing media, such as the control rods. This failure is much more pronounced in the PBMR because the location of control rods in the side reflector adds a directional dependence to the flux, and this complicates the problem further. Reactor control and safety can only be ensured if control rod worths are accurately predicted. In this work a thorough evaluation of different control rod models is performed by constructing an approximate two-dimensional (2D) model adopting the so-called grey curtain and this is compared to full 3D deterministic transport model. Both these models are compared to an explicit MCNP model. It is shown in this study that it is possible to have an accurate 2D model of control rods, utilizing appropriate equivalent cross sections and applying them to a control rod grey curtain. Further, this paper also shows that it is possible to obtain reasonably accurate control rod worths using a 3D deterministic transport method with the same fidelity of a Monte Carlo reference calculation, provided that the correct cross sections are used. These results confirm that the deterministic transport models can be successfully used for PBMR transient safety analysis.}, number={6}, journal={Annals of Nuclear Energy}, publisher={Elsevier BV}, author={Tyobeka, Bismark and Ivanov, Kostadin and Pautz, Andreas}, year={2008}, month={Jun}, pages={1050–1055} }
@article{alim_ivanov_yilmaz_levine_2008, title={New genetic algorithms (GA) to optimize PWR reactors}, volume={35}, ISSN={0306-4549}, url={http://dx.doi.org/10.1016/j.anucene.2007.05.004}, DOI={10.1016/j.anucene.2007.05.004}, abstractNote={Abstract In this paper, the GARCO–PSU (Genetic Algorithm Reactor Code Optimization–Pennsylvania State University) code simultaneously optimizes the core loading pattern (LP) and the burnable poison (BP) placement in a pressurized water reactor (PWR). The LP optimization and BP placement optimization are interconnected, but it is difficult to solve the combined problem due to its large size. Separating the problem into two sequential steps provides a practical way to solve the problem. However, the result of this method alone may not develop the real optimal solution. GARCO–PSU achieves solving the LP optimization and BP placement optimization simultaneously by developing an innovative genetic algorithm (GA). The classical representation of the genotype has been modified to incorporate in-core fuel management basic knowledge. GARCO has three modes; the first mode optimizes the LP only, the second mode optimizes the LP and BP placement in sequence. The third mode, which optimizes the LP and BP placement simultaneously, is described in this paper. GARCO, as stated in Part I, can be applied to all types of PWR core structures having different geometries with an unlimited number of fuel assembly (FA) types in the inventory.}, number={1}, journal={Annals of Nuclear Energy}, publisher={Elsevier BV}, author={Alim, Fatih and Ivanov, Kostadin N. and Yilmaz, Serkan and Levine, Samuel H.}, year={2008}, month={Jan}, pages={113–120} }
@article{alim_ivanov_levine_2008, title={New genetic algorithms (GA) to optimize PWR reactors}, volume={35}, ISSN={0306-4549}, url={http://dx.doi.org/10.1016/j.anucene.2007.05.007}, DOI={10.1016/j.anucene.2007.05.007}, abstractNote={The objective of this study was to develop a unique scientific methodology as well as a practical tool for designing the loading pattern (LP) and burnable poison (BP) pattern for a given Pressurized Water Reactor (PWR) core. Because of the large number of possible combinations for the fuel assembly (FA) loading in the core, the design of the core configuration is a complex optimization problem. It requires finding an optimal FA arrangement and corresponding BP placement design that will achieve maximum cycle length while satisfying the safety constraints. To solve this optimization problem, a core reload optimization package, GARCO (Genetic Algorithm Reactor Code Optimization) code is developed. This code is applicable for all types of PWR cores having different geometries and designs with an unlimited number of FA types in the inventory. GARCO has three modes: the user can optimize the core configuration (LP pattern) with or without BPs in the first mode; the second mode is the optimization of BP placement in the core and the last mode is the user can optimize LP and BP placements simultaneously in mode 3. In this study, the first mode finds the optimal LPs using the Haling Power Depletion Method (HPD) for placing BPs in the core. The second mode, which depletes the core accurately, places BPs in the selected optimum LP pattern. This methodology is applied only to the TMI-1 PWR. However, the improved Mode 1 GA option was applied to both the VVER-1000 and the TMI-1 to demonstrate and verify the advantages of the new enhancements in optimizing the LP pattern only. The “Moby-Dick” code is used as reactor physics code for VVER-1000 analysis in this research. The SIMULATE-3 code, which is an advanced two-group nodal code, is used to analyze the TMI-1. The libraries of the BP designs used in SIMULATE-3 in this study were produced by Yilmaz (2005) [Yilmaz, S., 2005. Multilevel optimization of burnable poison utilization for advanced PWR fuel management. Ph.D. Thesis in Nuclear Engineering. the Pennsylvania State University].}, number={1}, journal={Annals of Nuclear Energy}, publisher={Elsevier BV}, author={Alim, Fatih and Ivanov, Kostadin and Levine, Samuel H.}, year={2008}, month={Jan}, pages={93–112} }
@article{alim_ivanov_levine_2008, title={New genetic algorithms (GA) to optimize PWR reactors}, volume={35}, ISSN={0306-4549}, url={http://dx.doi.org/10.1016/j.anucene.2007.05.008}, DOI={10.1016/j.anucene.2007.05.008}, abstractNote={The Haling Power Distribution (HPD) has been applied in a unique process to greatly accelerate the in-core fuel management optimization calculations. These calculations involve; the arrangement of fuel assemblies (FAs) and the placement of Burnable Poisons (BPs) in the fresh FAs. The HPD deals only with the arrangement of FAs. The purpose of this paper is to describe past uses of the HPD, provide an example selected from many similar calculations to explain why and how it can be used, and also to show its effectiveness as a filter in the GARCO GA code. The GARCO (Genetic Algorithm Reactor Core Optimization) is an innovative GA code that was developed by modifying the classical representation of the genotype and GA operators. A reactor physics code evaluates the LPs in the population using the HPD Method, which rapidly depletes the core in a single depletion step with a constant power distribution. The HPD is used basically in GARCO as a filter to eliminate invalid LPs created by the genetic operators, to choose a reference LP for BP optimization, and to create an initial population for simultaneous optimization of the LP and BP placement into the core. The accurate depletion calculation of the LP with BPs is done with the coupled lattice and reactor physics CASMO-4/SIMULATE3 package. However, the fact that these codes validate safety of the core with the added BP placement design also validates the use of the HPD method. The calculations are applied to the TMI-1 core as an example PWR providing concrete results.}, number={1}, journal={Annals of Nuclear Energy}, publisher={Elsevier BV}, author={Alim, Fatih and Ivanov, Kostadin and Levine, Samuel H.}, year={2008}, month={Jan}, pages={121–131} }
@article{d’auria_soloviev_malofeev_ivanov_parisi_2008, title={The three-dimensional neutron kinetics coupled with thermal-hydraulics in RBMK accident analysis}, volume={238}, ISSN={0029-5493}, url={http://dx.doi.org/10.1016/j.nucengdes.2007.03.003}, DOI={10.1016/j.nucengdes.2007.03.003}, abstractNote={Abstract The RBMK core is a complex ensemble of high-pressure high-temperature tubes, graphite bricks, low-pressure low-temperature control rod tubes, graphite interstitial gas passages. An about 7 MPa boiling light water crosses the around 19 m long vertical tubes (7 m active length). The lattice consisting of graphite columns and hydraulic channels is bounded by the reactor cavity whose resistant elements are the metal cylindrical tank and thick circular top and bottom plates with proper holes for the passage of tubes. Related to a typical water cooled reactor, the peculiarities of the RBMK core can be summarized as follows: (a) large dimensions – the overall core volume is by far the largest for a nuclear power plant (NPP) producing electricity; (b) use of separate moderator and coolant constituted by graphite and light boiling water, respectively – the boiling water mostly absorbs neutrons in this environment leading to the (small) positive void reactivity coefficient; (c) presence of water channels very close to each other containing coolant at different temperatures (543–557 K and 350 K for fuel channels (FC) and control and protection system (CPS) channels, respectively); (d) presence of core-wide radial, core-wide axial and local temperature gradients in the graphite bricks with temperature values in the range 330–650 K with the high-temperature values justified by the neutron moderation and gamma-heating processes. Owing to the above peculiarities, the development and the use of a three-dimensional neutron kinetics code (3D NK) coupled with a one-dimensional thermal-hydraulic (TH) code is essential in RBMK safety analyses. Two approaches have been used within the present context, i.e. use of coupled 3D NK-TH codes to support the accident analysis in the RBMK as discussed in the first of the companion papers in this journal volume: application of Korsar-Bars making use of the Unk code to derive λ-matrices needed for Bars and of Relap5/3D-Nestle making use of the Helios code to derive the macroscopic cross-sections. Bounding transient analyses of accident scenarios including control rod withdrawal, various Loss of Coolant Accident (LOCA) and discharge of the control rod circuit, have been completed. In all of the analysed cases, starting from nominal operating conditions, modest fission power time gradients have been found, i.e. characterized by time derivative values for local and global power changes substantially smaller than current values accepted in safety analyses of light water reactors.}, number={4}, journal={Nuclear Engineering and Design}, publisher={Elsevier BV}, author={D’Auria, F. and Soloviev, S. and Malofeev, V. and Ivanov, K. and Parisi, C.}, year={2008}, month={Apr}, pages={1002–1025} }
@article{ivanov_avramova_2007, title={Challenges in coupled thermal–hydraulics and neutronics simulations for LWR safety analysis}, volume={34}, ISSN={0306-4549}, url={http://dx.doi.org/10.1016/j.anucene.2007.02.016}, DOI={10.1016/j.anucene.2007.02.016}, abstractNote={The simulation of nuclear power plant accident conditions requires three-dimensional (3D) modeling of the reactor core to ensure a realistic description of physical phenomena. The operational flexibility of Light Water Reactor (LWR) plants can be improved by utilizing accurate 3D coupled neutronics/thermal–hydraulics calculations for safety margins evaluations. There are certain requirements to the coupling of thermal–hydraulic system codes and neutron-kinetics codes that ought to be considered. The objective of these requirements is to provide accurate solutions in a reasonable amount of CPU time in coupled simulations of detailed operational transient and accident scenarios. These requirements are met by the development and implementation of six basic components of the coupling methodologies: ways of coupling (internal or external coupling); coupling approach (integration algorithm or parallel processing); spatial mesh overlays; coupled time-step algorithms; coupling numerics (explicit, semi-implicit and implicit schemes); and coupled convergence schemes. These principles of the coupled simulations are discussed in details along with the scientific issues associated with the development of appropriate neutron cross-section libraries for coupled code transient modeling. The current trends in LWR nuclear power generation and regulation as well as the design of next generation LWR reactor concepts along with the continuing computer technology progress stimulate further development of these coupled code systems. These efforts have been focused towards extending the analysis capabilities as well as refining the scale and level of detail of the coupling. This article analyses the coupled phenomena and modeling challenges on both global (assembly-wise) and local (pin-wise) levels. The issues related to the consistent qualification of coupled code systems as well as their application to different types of LWR transients are presented. Finally, the advances in numerical and computation techniques for coupled code simulations are summarized with outlining remaining challenges.}, number={6}, journal={Annals of Nuclear Energy}, publisher={Elsevier BV}, author={Ivanov, Kostadin and Avramova, Maria}, year={2007}, month={Jun}, pages={501–513} }
@article{tyobeka_ivanov_pautz_2007, title={Utilization of two-dimensional deterministic transport methods for analysis of pebble-bed reactors}, volume={34}, ISSN={0306-4549}, url={http://dx.doi.org/10.1016/j.anucene.2007.01.014}, DOI={10.1016/j.anucene.2007.01.014}, abstractNote={Abstract This paper presents an overview of the investigations on the need for deterministic transport methods for the analysis of pebble-bed reactors. To account for the transport effects present in the PBMR design that cannot be modeled accurately with the diffusion theory, a two-dimensional neutronics solver based on transport theory is implemented in the Penn State NEM-THERMIX code system. The necessity of equipping neutronics analysis codes with neutron transport theory capability is investigated along with the challenge to accomplish this in an efficient and versatile manner. For this purpose a time-dependent two-dimensional neutron transport code DORT is utilized as a first step. The developed benchmark test cases, based on the PBMR 268 MW design, are used for this work and results from the comparative analyzes of these test cases are presented. The results show clearly that even in steady-state calculations, the differences between diffusion and transport-based methods in analyzing the PBMR are observed and need to be addressed.}, number={5}, journal={Annals of Nuclear Energy}, publisher={Elsevier BV}, author={Tyobeka, Bismark and Ivanov, Kostadin and Pautz, Andreas}, year={2007}, month={May}, pages={396–405} }
@article{bousbia salah_vedovi_d'auria_galassi_ivanov_2006, title={Analysis of the VVER1000 coolant trip benchmark using the coupled RELAP5/PARCS code}, volume={48}, ISSN={0149-1970}, url={http://dx.doi.org/10.1016/j.pnucene.2006.06.005}, DOI={10.1016/j.pnucene.2006.06.005}, abstractNote={Abstract Incorporation of full three-dimensional models of the reactor core into system thermal–hydraulic transient codes allows better estimation of interactions between the core behavior and plant dynamics. Considerable efforts have been made in various countries and organizations to verify and validate the capability of the so-called coupled codes technique. For these purposes appropriate Light Water Reactor (LWR) transient benchmarks based upon programmed transients performed in Nuclear Power Plants (NPP) were recently developed on a higher ‘best-estimate’ level. The reference problem considered in the current framework is a Main Coolant Pump (MCP) switching-on transient in a VVER1000 NPP. This event is characterized by a positive reactivity addition as consequence of the increase of the core flow. In the current study the coupled RELAP5/PARCS code is used to reproduce the considered test. Results of calculation were assessed against experimental data and also through the code-to-code comparison.}, number={8}, journal={Progress in Nuclear Energy}, publisher={Elsevier BV}, author={Bousbia Salah, Anis and Vedovi, Juswald and D'Auria, Francesco and Galassi, Giorgio and Ivanov, Kostadin}, year={2006}, month={Nov}, pages={806–819} }
@article{yilmaz_ivanov_levine_mahgerefteh_2006, title={Application of genetic algorithms to optimize burnable poison placement in pressurized water reactors}, volume={33}, ISSN={0306-4549}, url={http://dx.doi.org/10.1016/j.anucene.2005.11.012}, DOI={10.1016/j.anucene.2005.11.012}, abstractNote={An efficient and a practical genetic algorithm (GA) tool was developed and applied successfully to Burnable Poison (BP) placement optimization problem in the reference Three Mile Island-1 (TMI-1) core. Core BP optimization problem means developing a BP loading map for a given core loading pattern that minimizes the total Gadolinium (Gd) amount in the core without violating any design constraints. The number of UO2/Gd2O3 pins and Gd2O3 concentrations for each fresh fuel location in the core are the decision variables. The objective function was to minimize the total amount of Gd in the core together with the residual Gd reactivity binding at the End-of-Cycle (EOC). The constraints are to keep the maximum peak pin power during the core depletion and soluble boron (SOB) concentration at the Beginning of Cycle (BOC) both less than their limit values. The innovation of this study was to search all of the possible UO2/Gd2O3 fuel assembly designs with variable number of UO2/Gd2O3 fuel pins and concentration of Gd2O3 in the overall decision space. The use of different fitness functions guided the solution towards desired (good solutions) region in the solution space, which accelerated the GA solution. The main objective of this study was to develop a practical and efficient GA tool and to apply this tool to designing an optimum BP pattern for a given core loading.}, number={5}, journal={Annals of Nuclear Energy}, publisher={Elsevier BV}, author={Yilmaz, Serkan and Ivanov, Kostadin and Levine, Samuel and Mahgerefteh, Moussa}, year={2006}, month={Mar}, pages={446–456} }
@article{shkarupa_kadenko_malanich_kovtonyuk_ivanov_2006, title={Comparative RELAP5-3D analysis in support of the NPP DBA analysis in Ukraine}, volume={48}, ISSN={0149-1970}, url={http://dx.doi.org/10.1016/j.pnucene.2006.06.001}, DOI={10.1016/j.pnucene.2006.06.001}, abstractNote={In the Ukrainian in-depth safety assessment (ISA) projects the computer code RELAP5/Mod3.2 with point kinetics approximation is being used in the deterministic safety analysis of VVERs. It is generally accepted that the use of this approximation, with the proper modeling assumptions, results in conservative results. However, only coupled three-dimensional codes are capable to estimate the real localized feedback effects for such VVER specific transients as control rod ejection or main steam line break. Some results of the comparative RELAP5-3D analysis for the scenarios, that present strong local reactivity effects, are discussed in this paper. The goal of this RELAP5-3D analysis is to examine the differences in results obtained by the three-dimensional approach and the methodology that was used in Ukrainian ISA projects.}, number={8}, journal={Progress in Nuclear Energy}, publisher={Elsevier BV}, author={Shkarupa, A. and Kadenko, I. and Malanich, A. and Kovtonyuk, A. and Ivanov, K.}, year={2006}, month={Nov}, pages={891–911} }
@article{yilmaz_ivanov_levine_mahgerefteh_2006, title={Development of enriched Gd-155 and Gd-157 burnable poison designs for a PWR core}, volume={33}, ISSN={0306-4549}, url={http://dx.doi.org/10.1016/j.anucene.2005.11.011}, DOI={10.1016/j.anucene.2005.11.011}, abstractNote={In this study, a genetic algorithm developed by the authors was applied to design the optimal enriched Gd-155 and Gd-157 burnable poisons in a reference PWR TMI-1 core. The CASMO-4/TABLES/SIMULATE-3 package calculated the neutronic performance of the enriched UO2/Gd2O3 fuel pin configurations. These configurations included different fractions of neutron absorbing isotopes Gd-155 and Gd-157, and 100 w/o enriched Gd-155 designs. Fuel cost analysis was performed to evaluate the economical benefits of these optimized enriched gadolinium designs. The break-even point for unit Gd-155 enrichment cost was determined to be around ∼$30/gram-Gd-155 with current unit cost scenario. The projected savings were 3.13% in gross and 2.08% in net compared to total fuel cycle cost of a reference TMI-1 core loading, if all of the 68 feed assemblies would be replaced with the optimized designs.}, number={5}, journal={Annals of Nuclear Energy}, publisher={Elsevier BV}, author={Yilmaz, Serkan and Ivanov, Kostadin and Levine, Samuel and Mahgerefteh, Moussa}, year={2006}, month={Mar}, pages={439–445} }
@article{cuervo_avramova_ivanov_miró_2006, title={Evaluation and enhancement of COBRA-TF efficiency for LWR calculations}, volume={33}, ISSN={0306-4549}, url={http://dx.doi.org/10.1016/j.anucene.2006.03.011}, DOI={10.1016/j.anucene.2006.03.011}, abstractNote={Abstract Detailed representations of the reactor core generate computational meshes with a high number of cells where the fluid dynamics equations must be solved. An exhaustive analysis of the CPU times needed by the thermal-hydraulic subchannel code COBRA-TF for different stages in the solution process has revealed that the solution of the linear system of pressure equations is the most time consuming process. To improve code efficiency two optimized matrix solvers, Super LU library and Krylov non-stationary iterative methods have been implemented in the code and their performance has been tested using a suite of five test cases. The results of performed comparative analyses have demonstrated that for large cases, the implementation of the Bi-Conjugate Gradient Stabilized (Bi-CGSTAB) Krylov method combined with the incomplete LU factorization with dual truncation strategy (ILUT) pre-conditioner reduced the time used by the code for the solution of the pressure matrix by a factor of 20. Both new solvers converge smoothly regardless of the nature of simulated cases and the mesh structures and improve the stability and accuracy of results compared to the classic Gauss–Seidel iterative method. The obtained results indicate that the direct inversion method is the best option for small cases.}, number={9}, journal={Annals of Nuclear Energy}, publisher={Elsevier BV}, author={Cuervo, Diana and Avramova, Maria and Ivanov, Kostadin and Miró, Rafael}, year={2006}, month={Jun}, pages={837–847} }
@article{ivanov_aniel_siltanen_royer_ivanov_2006, title={Impact of cross-section generation procedures on the simulation of the VVER-1000 pump startup experiment in the OECD/DOE/CEA V1000CT benchmark by coupled 3D thermal-hydraulics/neutron kinetics models}, volume={48}, ISSN={0149-1970}, url={http://dx.doi.org/10.1016/j.pnucene.2006.06.003}, DOI={10.1016/j.pnucene.2006.06.003}, abstractNote={In the framework of joint effort between the Nuclear Energy Agency (NEA) of OECD, the United States Department of Energy (US DOE), and the Commissariat a l'Enerige Atomique (CEA), France a coupled three-dimensional (3D) thermal-hydraulics/neutron kinetics benchmark was defined. The overall objective of OECD/NEA V1000CT benchmark is to assess computer codes used in analysis of VVER-1000 reactivity transients where mixing phenomena (mass flow and temperature) in the reactor pressure vessel are complex. Original data from the Kozloduy-6 Nuclear Power Plant are available for the validation of computer codes: one experiment of pump start-up (V1000CT-1) and one experiment of steam generator isolation (V1000CT-2). Additional scenarios are defined for code-to-code comparison. As a 3D core model is necessary for a best-estimate computation of all the scenarios of the V1000CT benchmark, all participants were asked to develop their own core coupled 3D thermal-hydraulics/neutron kinetics models using the data available in the benchmark specifications and a common cross-section library. The first code-to-code comparisons based on the V1000CT-1 Exercise 2 specifications exhibited unacceptable discrepancies between two sets of results. The present paper focuses on the analysis of the observed discrepancies. The VVER-1000 3D neutron kinetics models are based on cross-section data homogenized on the assembly level. The cross-section library, provided as part of the benchmark specifications, thus consists in a set of parameterized two group cross sections representing the different assemblies and the reflectors. The origin of the observed large discrepancies was found mainly to lie in the methods used to solve the diffusion equation. The VVER reflector properties were also found to enhance discrepancies by increasing flux gradients at the core/reflector interface thus highlighting more the difficulties in some codes to handle high exponential flux gradients. This paper summarizes the different steps applied to analyze the neutronic codes and their predictions as well as the impact of cross-section generation procedures.}, number={8}, journal={Progress in Nuclear Energy}, publisher={Elsevier BV}, author={Ivanov, Boyan D. and Aniel, Sylvie and Siltanen, Pertti and Royer, Eric and Ivanov, Kostadin N.}, year={2006}, month={Nov}, pages={746–763} }
@article{ivanov_2006, title={Introduction to the special issue on the OECD/DOE/CEA V1000CT-1 benchmark}, volume={48}, ISSN={0149-1970}, url={http://dx.doi.org/10.1016/j.pnucene.2006.07.002}, DOI={10.1016/j.pnucene.2006.07.002}, number={8}, journal={Progress in Nuclear Energy}, publisher={Elsevier BV}, author={Ivanov, Kostadin N.}, year={2006}, month={Nov}, pages={727} }
@article{alim_bekar_ivanov_unlu_brenizer_azmy_2006, title={Modeling and optimization of existing beam port facility of PSBR}, volume={33}, ISSN={0306-4549}, url={http://dx.doi.org/10.1016/j.anucene.2006.10.007}, DOI={10.1016/j.anucene.2006.10.007}, abstractNote={Due to inherited design issues with the current arrangement of beam ports (BPs) and reactor core-moderator assembly in The Perm State Breazeale Reactor (PSBR), the development of innovative experimental facilities utilizing neutron beams is extremely limited. Therefore, a study has started to examine the existing BPs for neutron and gamma outputs and develop a new core-moderator location and BP geometry in PSBR. Although 7 BPs are placed in PSBR, 2 of them are using currently. In this study BP 4, one of the currently being used BP, is examined. With changing the location of the BP 4 and structure of the core assembly, some artificial models are developed and compared with the original model.}, number={17-18}, journal={Annals of Nuclear Energy}, publisher={Elsevier BV}, author={Alim, Fatih and Bekar, Kursat and Ivanov, Kostadin and Unlu, Kenan and Brenizer, Jack and Azmy, Yousry}, year={2006}, month={Nov}, pages={1391–1395} }
@article{ivanov_ivanov_royer_aniel_bieder_kolev_groudev_2006, title={OECD/DOE/CEA VVER-1000 coolant transient (V1000CT) benchmark – A consistent approach for assessing coupled codes for RIA analysis}, volume={48}, ISSN={0149-1970}, url={http://dx.doi.org/10.1016/j.pnucene.2006.06.002}, DOI={10.1016/j.pnucene.2006.06.002}, abstractNote={The rod ejection accident (REA) and the main steam line break (MSLB) are two of the most important design basis accidents (DBA) for VVER-1000 exhibiting significant localized space-time effects. A consistent approach for assessing coupled three-dimensional (3-D) neutron kinetics/thermal-hydraulics codes for reactivity insertion accidents (RIA) is to first validate the codes using the available plant test (measured) data and after that to perform cross code comparative analysis for REA and MSLB scenarios. The coupled 3-D neutron kinetics/thermal-hydraulics benchmark presented in this paper is based on data from the Unit #6 of the Bulgarian Kozloduy Nuclear Power Plant (KNPP) and it is entitled the VVER-1000 coolant transient (V1000CT) benchmark. Two real plant transients are selected for simulation in the benchmark: main coolant pump start-up (Phase 1) and coolant mixing tests (Phase 2). In addition to these transients extreme scenarios were defined for better testing 3-D neutronics/thermal-hydraulics coupling: rod ejection simulation with control rod being ejected in the core sector cooled by the switched on MCP (Phase 1) and MSLB transient (Phase 2). The paper presents an overview of the Phase 1 (V1000CT-1) benchmark activities and describes the approach used for assessing the coupled neutron kinetics/thermal-hydraulics codes. Selected comparative analysis of currently submitted participants' results is presented with emphasis on the observed modeling issues and deviations from the measured data.}, number={8}, journal={Progress in Nuclear Energy}, publisher={Elsevier BV}, author={Ivanov, Boyan D. and Ivanov, Kostadin N. and Royer, Eric and Aniel, Sylvie and Bieder, Ulrich and Kolev, Nikola and Groudev, Pavlin}, year={2006}, month={Nov}, pages={728–745} }
@article{gozalvez_yilmaz_alim_ivanov_levine_2006, title={Sensitivity study on determining an efficient set of fuel assembly parameters in training data for designing of neural networks in hybrid genetic algorithms}, volume={33}, ISSN={0306-4549}, url={http://dx.doi.org/10.1016/j.anucene.2005.12.006}, DOI={10.1016/j.anucene.2005.12.006}, abstractNote={Abstract Neural Networks (NNs) were applied as a tool for simulating several nuclear reactor physics parameters during core depletion calculations. The main objective was to develop NNs models capable of simulating useful reactor physics parameters to filter out the bad designs created in Genetic Algorithms (GAs) run without the need to perform reactor physics calculations for evaluation of individuals. Applying GAs to optimize both the nuclear reactor Loading Pattern (LP) and Burnable Poison (BP) designs for their respective performance characteristics creates many unwanted results along the way. New population individuals are normally analyzed with a reactor physics code to determine its fitness or applicability for future use. Significant time was required for each reactor physics code calculation and because most of the solution individuals created by GAs result in unusable designs, analyzing every solutions involves prohibitive computational times. Such long computational times can be greatly reduced by applying NNs to filter out most of the unwanted designs. A detailed description of the selection process of the NN architecture, training method, and adequate ranges of data are also presented. Finally, a hybrid GA algorithm is proposed in which two NNs are used to discard most of the worse LP and BP designs.}, number={5}, journal={Annals of Nuclear Energy}, publisher={Elsevier BV}, author={Gozalvez, Jose M. and Yilmaz, Serkan and Alim, Fatih and Ivanov, Kostadin and Levine, Samuel H.}, year={2006}, month={Mar}, pages={457–465} }
@article{kolev_petrov_ivanov_ivanov_2006, title={Simulation of the VVER-1000 pump start-up experiment of the OECD V1000CT benchmark with CATHARE and TRAC-PF1}, volume={48}, ISSN={0149-1970}, url={http://dx.doi.org/10.1016/j.pnucene.2006.07.003}, DOI={10.1016/j.pnucene.2006.07.003}, abstractNote={In the framework of joint effort between the Nuclear Energy Agency (NEA) of OECD, the United States Department of Energy (US DOE), and the Commissariat a l'Energie Atomique (CEA), France a coupled three-dimensional (3D) thermal-hydraulics/neutron kinetics benchmark for VVER-1000 was defined. The benchmark consists of calculation of a pump start-up experiment labelled V1000CT-1 (Phase 1), as well as a vessel mixing experiment and main steam line break (MSLB) transient labelled V1000CT-2 (Phase 2), respectively. The reference nuclear plant is Kozloduy-6 in Bulgaria. The overall objective is to assess computer codes used in the analysis of VVER-1000 reactivity transients. A specific objective is to assess the vessel mixing models used in system codes. Plant data are available for code validation consisting of one experiment of pump start-up (V1000CT-1) and one experiment of steam generator isolation (V1000CT-2). The validated codes can be used to calculate asymmetric MSLB transients involving similar mixing patterns. This paper summarizes a comparison of CATHARE and TRAC-PF1 system code results for V1000CT-1, Exercise 1, which is a full plant point kinetics simulation of a reactor coolant system (RCS) pump start-up experiment. The reference plant data include integral and sector average parameters. The comparison is made from the point of view of vessel mixing and full system simulation. CATHARE used a six-sector multiple 1D vessel thermal-hydraulic model with cross flows and TRAC used a six-sector, 18-channel coarse-mesh 3D vessel model. Good agreement in terms of integral parameters and inter-loop mixing is observed.}, number={8}, journal={Progress in Nuclear Energy}, publisher={Elsevier BV}, author={Kolev, N. and Petrov, N. and Ivanov, B. and Ivanov, K.}, year={2006}, month={Nov}, pages={922–936} }
@article{reitsma_strydom_de haas_ivanov_tyobeka_mphahlele_downar_seker_gougar_da cruz_et al._2006, title={The PBMR steady-state and coupled kinetics core thermal-hydraulics benchmark test problems}, volume={236}, ISSN={0029-5493}, url={http://dx.doi.org/10.1016/j.nucengdes.2005.12.007}, DOI={10.1016/j.nucengdes.2005.12.007}, abstractNote={In support of the pebble bed modular reactor (PBMR) Verification and Validation (V&V) effort, a set of benchmark test problems has been defined that focus on coupled core neutronics and thermal-hydraulic code-to-code comparisons. The motivation is not only to test the existing methods or codes available for high-temperature gas-cooled reactors (HTGRs), but also to serve as a basis for the development of more accurate and efficient tools to analyse the neutronics and thermal-hydraulic behaviour for design and safety evaluations in future. The reference design for the PBMR268 benchmark problem is derived from the 268 MW PBMR design with a dynamic central column containing only graphite spheres. Several simplifications were made to the design in order to limit the need for any further approximations when defining code models. During this process, care was taken to ensure that all the important characteristics of the reactor design were preserved. The definition and initial phases of the benchmark were performed under a cooperative research project between NRG, Penn State University (PSU) and PBMR (Pty) Ltd. However, participation has been extended to include Purdue University and INL. All contributions to the benchmark effort were made in-kind by the participating members including the participation in four benchmark meetings over a period of 3 years. Based on the work performed in this benchmark the PBMR 400 MW design with fixed central reflector has been accepted as an OECD benchmark problem and work has already started. In this paper, the benchmark definition and the different test cases are described in some detail. Phase 1 focuses on steady-state conditions with the purpose of quantifying differences between code systems, models and basic data. It also serves as the basis to establish a common starting condition for the transient cases. In Phase 2, the focus is on performing coupled kinetics/core thermal-hydraulics test problems with a common cross-section and material property sets. The six events selected are described, and examples of some results are included to illustrate the behaviour of the transients. The final results of this work will be published in an NRG report and the focus will move to the OECD 400 MW benchmark problem.}, number={5-6}, journal={Nuclear Engineering and Design}, publisher={Elsevier BV}, author={Reitsma, F. and Strydom, G. and de Haas, J.B.M. and Ivanov, K. and Tyobeka, B. and Mphahlele, R. and Downar, T.J. and Seker, V. and Gougar, H.D. and Da Cruz, D.F. and et al.}, year={2006}, month={Mar}, pages={657–668} }
@article{guler_levine_ivanov_svarny_krysl_mikolas_sustek_2004, title={Development of the VVER core loading optimization system}, volume={31}, ISSN={0306-4549}, url={http://dx.doi.org/10.1016/j.anucene.2003.09.005}, DOI={10.1016/j.anucene.2003.09.005}, abstractNote={The optimization procedures previously applied to Western type PWRs are being used for the first time to optimize the VVER 440 nuclear power plant (NPP)—Dukovany, and VVER 1000—NPP Temelin in the Czech Republic. The objective of the calculation is to minimize fuel cost while preserving all the safety constraints and margins. Optimization with burnable poisons (BP) is simplified by reloading the core in two steps; first, by optimum fuel placement using the Haling Power Distribution (HPD), and second, by optimum placement of burnable poisons to meet the safety constraints. This two-step single cycle optimization method can be extended to the multiple cycle level. Application to the VVER reactors involves applying a hexagonal geometry core analysis model. A program has been initiated to develop a similar multi-cycle optimization system for the VVER reactors between the Penn State University and SKODA in the Czech Republic. This research has been successful in developing loading patterns for cycles 2–4 of the Temelin NPP. The package, which was developed during this research, can be applied to any VVER reactors to optimize the core loading.}, number={7}, journal={Annals of Nuclear Energy}, publisher={Elsevier BV}, author={Guler, C. and Levine, S. and Ivanov, K. and Svarny, J. and Krysl, V. and Mikolas, P. and Sustek, J.}, year={2004}, month={May}, pages={747–772} }
@article{ivanov_ivanov_stamm'ler_2004, title={Helios, current coupling collision probability method, applied for solving the NEA C5G7 MOX benchmark}, volume={45}, ISSN={0149-1970}, url={http://dx.doi.org/10.1016/j.pnucene.2004.09.004}, DOI={10.1016/j.pnucene.2004.09.004}, abstractNote={As part of an effort to test the ability of current transport codes to treat reactor core problems without spatial homogenization, the lattice code HELIOS was employed to perform criticality calculations. The test consists in seven-group calculations of the C5 MOX fuel assembly problem specified by Cavarec et. al. [1]. This problem, known as C5G7 MOX Benchmark, is described in the Benchmark Specification [2] and comprises two cases — two and three-dimensional geometry. There are four fuel assemblies — two with MOX fuel, the other two with UO2 fuel. Each fuel assembly is made up of a 17×17 lattice of square fuel-pin cells. Fuel pin compositions are specified in the Benchmark Specification, which also provides seven-group transport-corrected isotropic scattering cross-sections for U02, the three MOX enrichments, the guide tubes, the fission chamber and the moderator. This paper preset is the methodology employed in solving the C5G7 MOX Fuel Assembly Problem using the transport code HELIOS.}, number={2-4}, journal={Progress in Nuclear Energy}, publisher={Elsevier BV}, author={Ivanov, Boyan D. and Ivanov, Kostadin N. and Stamm'ler, Rudi J.J.}, year={2004}, month={Jan}, pages={119–124} }
@article{tippayakul_levine_ivanov_mahgerefteh_2004, title={Minimizing the gamma radiation effects to spent fuel pool walls}, volume={31}, ISSN={0306-4549}, url={http://dx.doi.org/10.1016/j.anucene.2003.09.006}, DOI={10.1016/j.anucene.2003.09.006}, abstractNote={This paper presents the study of the gamma exposure to the spent fuel pool walls from spent fuel assemblies. The Monte Carlo code, MCNP4C, was used to model and analyze the spent fuel pool walls. In addition, the gamma source from the spent fuel assembly was modeled by the isotope generation and depletion code, ORIGEN2.2. Both PWR (TMI-I) and BWR (PB-2) spent fuel pool models were investigated. The primary objective of the study was to determine the arrangement of the spent fuel assemblies to minimize the gamma radiation to the spent fuel pool walls. Different cases of spent fuel assembly configurations in the spent fuel pool were studied. The study results show the amount of gamma radiation deposited on each section of the spent fuel pool walls. Finally, the spent fuel assembly with axial peaking factor was also studied.}, number={5}, journal={Annals of Nuclear Energy}, publisher={Elsevier BV}, author={Tippayakul, Chanatip and Levine, Samuel and Ivanov, Kostadin N. and Mahgerefteh, Moussa}, year={2004}, month={Mar}, pages={459–480} }