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  • 1.
    Dufek, Jan
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering.
    Mickus, Ignas
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering.
    Optimal time step length and statistics in Monte Carlo burnup simulations2020In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 139, article id 107244Article in journal (Refereed)
    Abstract [en]

    Monte Carlo burnup simulations continue to be seen as computationally very expensive numerical routines despite recent developments of associated methods. Here, we suggest a way of improving the computing efficiency via optimisation of the length of the time steps and the number of neutron histories that are simulated at each Monte Carlo criticality run. So far, users of Monte Carlo burnup codes have been required to set these parameters at will; however, an inadequate choice of these free parameters can severely worsen the computing efficiency. We have tested a large number of combinations of the free parameters on a simplified and fast solver, and we have observed that the computing efficiency was maximized when the computing cost of all Monte Carlo neutron transport calculations (summed over all time steps) was approximately comparable to costs of other procedures (all depletion simulations, the loading and processing of neutron cross sections, etc.). In this technical note, we demonstrate these results, and we also derive a simple theoretical model of the convergence of Monte Carlo burnup simulations that conforms to these numerical results. Here, we also suggest a straightforward way to automatise the selection of the optimal values of the free parameters for Monte Carlo burnup simulations.

  • 2.
    Grishchenko, Dmitry
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Mickus, Ignas
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Kudinov, Pavel
    Experimental activities report on TALL-3D2015Report (Other academic)
  • 3.
    Ignas, Mickus
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering.
    Response Matrix Reloaded: for Monte Carlo Simulations in Reactor Physics2019Licentiate thesis, comprehensive summary (Other academic)
    Abstract [en]

    This thesis investigates Monte Carlo methods applied to criticality and time-dependent problems in reactor physics. Due to their accuracy and flexibility, Monte Carlo methods are considered as a “gold standard” in reactor physics calculations. However, the benefits come at a significant computing cost. Despite the continuous rise in easily accessible computing power, a brute-force Monte Carlo calculation of some problems is still beyond the reach of routine reactor physics analyses. The two papers on which this thesis is based try to address the computing cost issue, by proposing methods for performing Monte Carlo reactor physics calculations more efficiently. The first method addresses the efficiency of the widely-used k-eigenvalue Monte Carlo criticality calculations. It suggests, that the calculation efficiency can be increased through a gradual increase of the neutron population size simulated during each criticality cycle, and proposes a way to determine the optimal neutron population size. The second method addresses the application of Monte Carlo calculations to reactor transient problems. While reactor transient calculations can, in principle, be performed using only Monte Carlo methods, such calculations take multiple thousands of CPU hours for calculating several seconds of a transient. The proposed method offers a middle-ground approach, using a hybrid stochastic-deterministic scheme based on the response matrix formalism. Previously, the response matrix formalism was mainly considered for steady-state problems, with limited application to time-dependent problems. This thesis proposes a novel way of using information from Monte Carlo criticality calculations for solving time-dependent problems via the response matrix.

     

  • 4.
    Mickus, Ignas
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering.
    Dufek, Jan
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering.
    Optimal neutron population growth in accelerated Monte Carlo criticality calculations2018In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 117, p. 297-304Article in journal (Refereed)
    Abstract [en]

    We present a source convergence acceleration method for Monte Carlo criticality calculations. The method gradually increases the neutron population size over the successive inactive as well as active criticality cycles. This helps to iterate the fission source faster at the beginning of the simulation where the source may contain large errors coming from the initial cycle; and, as the neutron population size grows over the cycles, the bias in the source gets reduced. Unlike previously suggested acceleration methods that aim at optimisation of the neutron population size, the new method does not have any significant computing overhead, and moreover it can be easily implemented into existing Monte Carlo criticality codes. The effectiveness of the method is demonstrated on a number of PWR full-core criticality calculations using a modified SERPENT 2 code.

  • 5.
    Mickus, Ignas
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Reactor Technology.
    Dufek, Jan
    KTH, School of Engineering Sciences (SCI), Physics, Reactor Technology.
    Tuttelberg, Kaur
    PERFORMANCE OF THE EXPLICIT EULER AND PREDICTOR-CORRECTOR-BASED COUPLING SCHEMES IN MONTE CARLO BURNUP CALCULATIONS OF FAST REACTORS2015In: Nuclear Technology, ISSN 0029-5450, E-ISSN 1943-7471, Vol. 191, no 2, p. 193-198Article in journal (Refereed)
    Abstract [en]

    We present a stability test of the explicit Euler and predictor-corrector based coupling schemes in Monte Carlo burnup calculations of the gas fast reactor fuel assembly. Previous studies have identified numerical instabilities of these coupling schemes in Monte Carlo burnup calculations of thermal-spectrum reactors due to spatial feedback induced neutron flux and nuclide density oscillations, where only sufficiently small time steps could guarantee acceptable precision. New results suggest that these instabilities are insignificant in fast-spectrum assembly burnup calculations, and the considered coupling schemes can therefore perform well in fast-spectrum reactor burnup calculations even with relatively large time steps. Note: Some figures in this technical note may be in color only in the electronic version.

  • 6.
    Mickus, Ignas
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Reactor Technology.
    Kööp, Kaspar
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Jeltsov, Marti
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Grishchenko, Dmitry
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Lappalainen, J.
    Development of tall-3d test matrix for APROS code validation2015In: International Topical Meeting on Nuclear Reactor Thermal Hydraulics 2015, NURETH 2015, 2015, p. 4562-4575Conference paper (Refereed)
    Abstract [en]

    APROS code is a multifunctional process simulator which combines System Thermal-Hydraulic (STH) capabilities with ID,'3D reactor core neutronics and full automation system modeling. It is applied for various tasks throughout the complete power plant life cycle including R&D, process and control engineering, and operator training. Currently APROS is being developed for evaluation of Generation IV conceptual designs using Lead-Bismuth Eutectic (LBt) alloy coolant. TALL-3D facility has been built at KTH in order to provide validation data for standalone and coupled STH and Computational Fluid Dynamics (CFD) codes. The facility consists of sections with measured inlet and outlet conditions for separate effect and integral effect tests (SETs and lETs). The design is aimed at reducing experimental uncertainties and allowing fall separation of code validation from model input calibration. In this paper we present the development of experimental TALL-3D lest matrix for comprehensive validation of APROS code. First, the representative separate effect and integral system response quantities (SRQs) arc defined. Second, sources of uncertainties are identified and code sensitivity analysis is carried out to quantify the effects of code input uncertainties on the code prediction. Based on these results the test matrixes for calibration and validation experiments arc determined in order to minimize the code input uncertainties. The applied methodology and the results arc discussed in detail.

  • 7.
    Mickus, Ignas
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Kööp, Kaspar
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Jeltsov, Marti
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Vorobyev, Yuri
    Moscow Power Engineering Institute.
    Villanueva, Walter
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    An Approach to Physics Based Surrogate Model Development for Application with IDPSA2014Conference paper (Refereed)
    Abstract [en]

    Integrated Deterministic Probabilistic Safety Assessment (IDPSA) methodology is a powerful tool for identification of failure domains when both stochastic events and physical time dependent processes are important. Computational efficiency of deterministic models is one of the limiting factors for detailed exploration of the event space. Pool type designs of Generation IV heavy liquid metal cooled reactors introduce importance of capturing intricate 3D flow phenomena in safety analysis. Specifically mixing and stratification in 3D elements can affect efficiency of passive safety systems based on natural circulation. Conventional 1D System Thermal Hydraulics (STH) codes are incapable of predicting such complex 3D phenomena. Computational Fluid Dynamics (CFD) codes are too computationally expensive to be used for simulation of the whole reactor primary coolant system. One proposed solution is code coupling where all 1D components are simulated with STH and 3D components with CFD codes. However, modeling with coupled codes is still too time consuming to be used directly in IDPSA methodologies, which require thousands of simulations. The goal of this work is to develop a computationally efficient surrogate model (SM) which captures key physics of complex thermal hydraulic phenomena in the 3D elements and can be coupled with 1D STH codes instead of CFD. TALL-3D is a lead-bismuth eutectic thermal hydraulic loop which incorporates both 1D and 3D elements. Coupled STH-CFD simulations of TALL-3D typical transients (such as transition from forced to natural circulation) are used to calibrate the surrogate model parameters. Details of current implementation and limitations of the surrogate modeling are discussed in the paper in detail.

  • 8.
    Mickus, Ignas
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Lappalainen, Jari
    VTT Technical Research Centre of Finland.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Validation of APROS Code Against Experimental Data from a Lead-Bismuth Eutectic Thermal-Hydraulic Loop2015In: Proceedings of ICAPP 2015, May 03-06, 2015 - Nice (France), 2015Conference paper (Refereed)
    Abstract [en]

    APROS is a multifunctional simulator, suitable for various tasks throughout the complete power plant life cycle from design to operator training. The code combines the System-Thermal Hydraulic (STH) capabilities with 1D/3D reactor core neutronics and full automation system modelling. APROS is widely used for dynamic studies in light water nuclear reactor technology. Currently the simulation capabilities of the code are being developed for evaluation of Generation IV nuclear reactor designs that use liquid metals (e.g. Lead-Bismuth Eutectic (LBE) alloy) as cooling media. However, validation of APROS code has been limited to light water cooled systems (test facilities and reactors). Therefore validation is required for the LBE thermal-hydraulic models. In this work, we use data from the Thermal-hydraulic Accelerator-Driven System (ADS) Lead bismuth Loop (TALL) facility constructed at KTH. In total, 10 transient data sets from TALL experiments were used for APROS LBE fluid model integral validation. TALL facility model consisting of the primary LBE loop, the secondary oil loop and the ultimate water heat sink was developed using APROS Graphical User Interface (GUI). Five System Response Quantities (SRQs) were considered in the validation process: LBE temperatures at four locations of the facility and the LBE mass flow. Uncertainty analysis was employed in the validation process. Results of quantitative code validation are presented in the paper. Suggestions for improvements of the validation methodology, experimental conditions and approaches to reduction of experimental uncertainty are discussed in detail.

  • 9.
    Mickus, Ignas
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering.
    Roberts, Jeremy A.
    Department of Mechanical and Nuclear Engineering, Kansas State University.
    Dufek, Jan
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering.
    Stochastic-deterministic response matrix method for reactor transients2019In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, article id 107103Article in journal (Refereed)
    Abstract [en]

    Presented is a stochastic-deterministic, response matrix method for transient analyses of nuclear systems. The method is based on the response matrix formalism, which describes a system by a set of response functions. We propose an approach in which these response functions are computed during a set of Monte Carlo criticality calculations and are later used to formulate a deterministic set of equations for solving a space-time dependent problem. Application of the response matrix formalism results in a set of loosely connected equations, which leads to a favourable linear scaling of the problem. The method offers a simplified approach compared to previously proposed response matrix methods by avoiding phase-space expansions in sets of basis functions. We describe the method starting with the fundamental neutron transport considerations, provide a demonstration on two absorber movement transients in a 3 × 3 assembly PWR mini-core geometry, and compare the solutions against time-dependent Monte Carlo simulations.

  • 10.
    Wallenius, Janne
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering. LeadCold Reactors, Valhallavagen 79, S-11428 Stockholm, Sweden.
    Qvist, S.
    LeadCold Reactors, Valhallavagen 79, S-11428 Stockholm, Sweden..
    Mickus, Ignas
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering. LeadCold Reactors, Valhallavagen 79, S-11428 Stockholm, Sweden.
    Bortot, Sara
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering. LeadCold Reactors, Valhallavagen 79, S-11428 Stockholm, Sweden.
    Szakalos, Peter
    KTH, School of Engineering Sciences in Chemistry, Biotechnology and Health (CBH), Chemistry, Surface and Corrosion Science. LeadCold Reactors, Valhallavagen 79, S-11428 Stockholm, Sweden.
    Ejenstam, Lina
    KTH, School of Engineering Sciences in Chemistry, Biotechnology and Health (CBH), Chemistry, Surface and Corrosion Science.
    Design of SEALER, a very small lead-cooled reactor for commercial power production in off-grid applications2018In: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 338, p. 23-33Article in journal (Refereed)
    Abstract [en]

    In this paper, the conceptual design of a small lead-cooled nuclear reactor intended to replace diesel-power in off-grid applications is presented. In a vessel of dimensions making it transportable by air, the targeted design performance is to produce 3 MW of electrical power for up to 30 years without reloading of fuel. Consequently, the inner vessel can be sealed, delaying malevolent access to the nuclear fuel and improving security. Alumina forming alloys are applied to ensure long term corrosion protection of fuel cladding tubes, steam generator tubes and primary vessel over the operational temperature regime. Moreover, decay heat can be removed in a completely passive manner by natural convection from the core to the primary coolant and by thermal radiation from the primary vessel to the environment. Finally, the source term is such that relocation of population residing beyond 1 km from the reactor will not be required even in the case of a complete core melt.

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