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Fridman E, Nikitin E, Ponomarev A, Di Nora A, Kliem S, Mikityuk K. Extension of the DYN3D/ATHLET code system to SFR applications: models description and initial validation. ANN NUCL ENERGY 2023. [DOI: 10.1016/j.anucene.2022.109619] [Citation(s) in RCA: 0] [Impact Index Per Article: 0] [Reference Citation Analysis] [Track Full Text] [Journal Information] [Subscribe] [Scholar Register] [Indexed: 12/13/2022]
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2
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Jeon S, Hong H, Choi N, Joo HG. Methods and performance of a GPU-based pinwise two-step nodal code VANGARD. PROGRESS IN NUCLEAR ENERGY 2023. [DOI: 10.1016/j.pnucene.2022.104528] [Citation(s) in RCA: 0] [Impact Index Per Article: 0] [Reference Citation Analysis] [Track Full Text] [Subscribe] [Scholar Register] [Indexed: 12/15/2022]
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3
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Gong H, Zhu T, Chen Z, Wan Y, Li Q. Parameter identification and state estimation for nuclear reactor operation digital twin. ANN NUCL ENERGY 2023. [DOI: 10.1016/j.anucene.2022.109497] [Citation(s) in RCA: 0] [Impact Index Per Article: 0] [Reference Citation Analysis] [Track Full Text] [Journal Information] [Subscribe] [Scholar Register] [Indexed: 11/06/2022]
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4
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Bläsius C, Herb J, Sievers J, Knospe A, Viebach M, Lange C. Mechanical model for the motion of RPV internals affecting neutron flux noise. ANN NUCL ENERGY 2022. [DOI: 10.1016/j.anucene.2022.109243] [Citation(s) in RCA: 1] [Impact Index Per Article: 0.5] [Reference Citation Analysis] [Track Full Text] [Journal Information] [Subscribe] [Scholar Register] [Indexed: 11/01/2022]
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5
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Arkadov GV, Pavelko VI, Povarov VP, Slepov MT. Phenomenology of acoustic standing waves as applied to the VVER-1200 reactor plant. NUCLEAR ENERGY AND TECHNOLOGY 2022. [DOI: 10.3897/nucet.8.82755] [Citation(s) in RCA: 0] [Impact Index Per Article: 0] [Reference Citation Analysis] [Abstract] [Track Full Text] [Journal Information] [Subscribe] [Scholar Register] [Indexed: 11/12/2022] Open
Abstract
The insufficiently studied issues of acoustic standing waves (ASW) in the main circulation circuits of the VVER reactor plants are considered. For a long time no proper attention has been given to this phenomenon both by the researchers and NPP experts. In general, generation of ASWs requires the acoustic inhomogeneities of the medium in the planes perpendicular to the direction of propagation of the longitudinal wave, in which a jump in acoustic resistance occurs, this is shown by the authors based on an example of the wave equation solution (D’Alembert equation) for a certain function of two variables. The ASW classification has been developed based on the obtained experimental material, 6 ASW types have been described, and their key parameters have been specified. The amplitude distributions have been plotted for all major ASW types proceeding from the phase relations of signals from the pressure pulsation detectors and accelerometers installed on the MCC pipelines. The nature of these distributions is general and they are valid for all VVER types. For the first time the globality of all lowest ASW types is identified. Four attribute properties of the ASWs have been formulated. The first attribute is the regular ASW temperature dependences, which is the source of the diagnostic information in the process of heating/cooling of the VVER unit. The linear experimental dependences of the ASW frequencies on coolant temperature have been obtained. The frequencies, at which the MCC resonant excitation due to coincidence of the ASW frequencies with the RCP rotational frequency harmonics, have been found experimentally. The ASW energy, which origin has resulted from the RCP operation, is estimated. The RCP operation can be presented as continuous generation of pressure pulsations, which fall onto the acoustic path inhomogeneities in the form of a traveling wave and generate a standing wave after reflection from them.
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Current Status and On-Going Development of VTT’s Kraken Core Physics Computational Framework. ENERGIES 2022. [DOI: 10.3390/en15030876] [Citation(s) in RCA: 2] [Impact Index Per Article: 1.0] [Reference Citation Analysis] [Abstract] [Track Full Text] [Subscribe] [Scholar Register] [Indexed: 11/16/2022]
Abstract
The Kraken computational framework is a new modular calculation system designed for coupled core physics calculations. The development started at VTT Technical Research Centre of Finland in 2017, with the aim to replace VTT’s outdated legacy codes used for the deterministic safety analyses of Finnish power reactors. In addition to conventional large PWRs and BWRs, Kraken is intended to be used for the modeling of SMRs and emerging non-LWR technologies. The main computational modules include the Serpent Monte Carlo neutron and photon transport code, the Ants nodal neutronics solver, the FINIX fuel behavior module and the Kharon thermal hydraulics code, all developed at VTT. The core physics solution can be further coupled to system-scale simulations. In addition to development, significant effort has been devoted to verification and validation of the implemented methodologies. The reduced-order Ants code has been successfully used for steady-state, transient and burnup simulations of PWRs with rectangular and hexagonal core geometry. The Ants–Kharon–FINIX code sequence is actively used for the core design tasks in VTT’s district heating reactor project. This paper is a general overview on the background, functional description, current status and future plans for the Kraken framework. Due to the short history of development, Kraken has not yet been comprehensively validated or applied to full-scale core physics calculations. A review of previous studies is instead provided to exemplify the practical use.
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Pradhan SK, Acharya D, Das DK. Internal model control based proportional-integral controller with class topper optimization for power control of molten salt breeder reactor core. ANN NUCL ENERGY 2022. [DOI: 10.1016/j.anucene.2021.108675] [Citation(s) in RCA: 4] [Impact Index Per Article: 2.0] [Reference Citation Analysis] [Track Full Text] [Journal Information] [Subscribe] [Scholar Register] [Indexed: 12/01/2022]
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10
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Ponomarev A, Mikityuk K, Zhang L, Nikitin E, Fridman E, Álvarez-Velarde F, Romojaro Otero P, Jiménez-Carrascosa A, García-Herranz N, Lindley B, Baker U, Seubert A, Henry R. Superphénix Benchmark Part I: Results of Static Neutronics. JOURNAL OF NUCLEAR ENGINEERING AND RADIATION SCIENCE 2022. [DOI: 10.1115/1.4051449] [Citation(s) in RCA: 0] [Impact Index Per Article: 0] [Reference Citation Analysis] [Abstract] [Track Full Text] [Journal Information] [Subscribe] [Scholar Register] [Indexed: 11/08/2022]
Abstract
Abstract
In the paper, the specification of a new neutronics benchmark for large sodium cooled fast reactor (SFR) core and results of modeling by different participants are presented. The neutronics benchmark describes the core of the French sodium cooled reactor Superphénix at its startup configuration, which in particular was used for experimental measurement of reactivity characteristics. The benchmark consists of the detailed heterogeneous core specification for neutronic analysis and the results of the reference solution. Different core geometries and thermal conditions from the cold “as fabricated” up to full power were considered. The reference Monte Carlo (MC) solution of serpent 2 includes data on multiplication factor, power distribution, axial and radial reaction rates distribution, reactivity coefficients and safety characteristics, control rods worth, kinetic data. The results of modeling with seven other solutions using deterministic and MC methods are also presented and compared to the reference solution. The comparisons results demonstrate appropriate agreement of evaluated characteristics. The neutronics results will be used in the second phase of the benchmark for the evaluation of transient behavior of the core.
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Affiliation(s)
- Alexander Ponomarev
- Laboratory for Scientific Computing and Modelling, Paul Scherrer Institute (PSI), Forschungsstrasse 111, Villigen PSI 5232, Switzerland
| | - Konstantin Mikityuk
- Laboratory for Scientific Computing and Modelling, Paul Scherrer Institute (PSI), Forschungsstrasse 111, Villigen PSI 5232, Switzerland
| | - Liang Zhang
- Laboratory for Scientific Computing and Modelling, Paul Scherrer Institute (PSI), Forschungsstrasse 111, Villigen PSI 5232, Switzerland
| | - Evgeny Nikitin
- Reactor Safety Division, Institute of Resource Ecology, Helmholtz-Zentrum Dresden-Rossendorf (HZDR), Bautzner Landstraße 400, Dresden DE-01328, Germany
| | - Emil Fridman
- Reactor Safety Division, Institute of Resource Ecology, Helmholtz-Zentrum Dresden-Rossendorf (HZDR), Bautzner Landstraße 400, Dresden DE-01328, Germany
| | - Francisco Álvarez-Velarde
- Centro de Investigaciones Energéticas, MedioAmbientales y Tecnológicas (CIEMAT) Avda., Complutense, 40, Madrid 28040, Spain
| | - Pablo Romojaro Otero
- Centro de Investigaciones Energéticas, MedioAmbientales y Tecnológicas (CIEMAT)—Currently at SCK·CEN Avda, Complutense, 40, Madrid 28040, Spain
| | - Antonio Jiménez-Carrascosa
- Instituto de Fusion Nuclear, Universidad Politécnica de Madrid (UPM) José Gutiérrez Abascal, 2, Madrid 28006, Spain
| | - Nuria García-Herranz
- Instituto de Fusion Nuclear, Universidad Politécnica de Madrid (UPM) José Gutiérrez Abascal, 2, Madrid 28006, Spain
| | - Ben Lindley
- Department of Engineering Physics, University of Wisconsin-Madison Engineering Research Building, 1500 Engineering Drive, Madison WI 53706
| | - Una Baker
- Department of Engineering, University of Cambridge, Trumpington Street, Cambridge CB2 1PZ, UK
| | - Armin Seubert
- Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) gGmbH, Boltzmannstraße 14, Garching bei München 85748, Germany
| | - Romain Henry
- Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) gGmbH, Boltzmannstraße 14, Garching bei München 85748, Germany
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11
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Thorough analyses and resolution of various errors in pin-homogenized multigroup core calculation. ANN NUCL ENERGY 2021. [DOI: 10.1016/j.anucene.2021.108502] [Citation(s) in RCA: 0] [Impact Index Per Article: 0] [Reference Citation Analysis] [Track Full Text] [Journal Information] [Subscribe] [Scholar Register] [Indexed: 11/30/2022]
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12
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Pradhan SK, Das DK. Explicit model predictive controller for power control of molten salt breeder reactor core. NUCLEAR ENGINEERING AND DESIGN 2021. [DOI: 10.1016/j.nucengdes.2021.111492] [Citation(s) in RCA: 0] [Impact Index Per Article: 0] [Reference Citation Analysis] [Track Full Text] [Subscribe] [Scholar Register] [Indexed: 10/20/2022]
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13
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DYN3D and CTF Coupling within a Multiscale and Multiphysics Software Development (Part I). ENERGIES 2021. [DOI: 10.3390/en14165060] [Citation(s) in RCA: 1] [Impact Index Per Article: 0.3] [Reference Citation Analysis] [Abstract] [Track Full Text] [Subscribe] [Scholar Register] [Indexed: 11/16/2022]
Abstract
Understanding and optimizing the relation between nuclear reactor components or physical phenomena allows us to improve the economics and safety of nuclear reactors, deliver new nuclear reactor designs, and educate nuclear staff. Such relation in the case of the reactor core is described by coupled reactor physics as heat transfer depends on energy production while energy production depends on heat transfer with almost none of the available codes providing full coupled reactor physics at the fuel pin level. A Multiscale and Multiphysics nuclear software development between NURESIM and CASL for LWRs has been proposed for the UK. Improved coupled reactor physics at the fuel pin level can be simulated through coupling nodal codes such as DYN3D as well as subchannel codes such as CTF. In this journal article, the first part of the DYN3D and CTF coupling within the Multiscale and Multiphysics software development is presented to evaluate all inner iterations within one outer iteration to provide partially verified improved coupled reactor physics at the fuel pin level. Such verification has proven that the DYN3D and CTF coupling provides improved feedback distributions over the DYN3D coupling as crossflow and turbulent mixing are present in the former.
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Di Nora V, Fridman E, Nikitin E, Bilodid Y, Mikityuk K. Optimization of multi-group energy structures for diffusion analyses of sodium-cooled fast reactors assisted by simulated annealing – Part I: Methodology demonstration. ANN NUCL ENERGY 2021. [DOI: 10.1016/j.anucene.2021.108183] [Citation(s) in RCA: 3] [Impact Index Per Article: 1.0] [Reference Citation Analysis] [Track Full Text] [Journal Information] [Subscribe] [Scholar Register] [Indexed: 12/01/2022]
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15
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CTF and FLOCAL Thermal Hydraulics Validations and Verifications within a Multiscale and Multiphysics Software Development. ENERGIES 2021. [DOI: 10.3390/en14051220] [Citation(s) in RCA: 3] [Impact Index Per Article: 1.0] [Reference Citation Analysis] [Abstract] [Track Full Text] [Subscribe] [Scholar Register] [Indexed: 11/16/2022]
Abstract
Simulation codes allow one to reduce the high conservativism in nuclear reactor design improving the reliability and sustainability associated with nuclear power. Full-core coupled reactor physics at the rod level are not provided by most simulation codes. This has led in the UK to the development of a multiscale and multiphysics software development focused on LWRS. In terms of the thermal hydraulics, simulation codes suitable for this multiscale and multiphysics software development include the subchannel code CTF and the thermal hydraulics module FLOCAL of the nodal code DYN3D. In this journal article, CTF and FLOCAL thermal hydraulics validations and verifications within the multiscale and multiphysics software development have been performed to evaluate the accuracy and methodology available to obtain thermal hydraulics at the rod level in both simulation codes. These validations and verifications have proved that CTF is a highly accurate subchannel code for thermal hydraulics. In addition, these verifications have proved that CTF provides a wide range of crossflow and turbulent mixing methods, while FLOCAL in general provides the simplified no-crossflow method as the rest of the methods were only tested during its implementation into DYN3D.
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16
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Nikitin E, Fridman E, Mikityuk K, Radman S, Fiorina C. NEUTRONIC MODELLING OF THE FFTF CONTROL ROD WORTH MEASUREMENTS WITH DIFFUSION CODES. EPJ WEB OF CONFERENCES 2021. [DOI: 10.1051/epjconf/202124710017] [Citation(s) in RCA: 1] [Impact Index Per Article: 0.3] [Reference Citation Analysis] [Abstract] [Track Full Text] [Journal Information] [Subscribe] [Scholar Register] [Indexed: 11/14/2022] Open
Abstract
This paper presents an assessment of three deterministic core simulators with the focus on the neutronic performance in steady-state calculations of small Sodium cooled Fast Reactor cores. The selected codes are DYN3D, PARCS and the novel multi-physics solver GeN-Foam. By using these codes, the multi-group diffusion solutions are obtained for the selected twenty control rod worth measurements performed during the isothermal physics tests of the Fast Flux Test Facility (FFTF). The identical set of homogenized few-group cross sections applied in the calculations is generated with the Serpent Monte Carlo code. The numerical results are compared with each other as well as with the measured values. The obtained numerical results, such as the multiplication factors and control rod worth values, are in good agreement as compared to the experimental data. Furthermore, a comparison of the radial power distributions is presented between DYN3D, PARCS and GeN-Foam. Ultimately, the power distributions are compared to the full core Serpent solution, demonstrating an adequate performance of the selected deterministic tools. In overall, this study presents a verification and validation of the neutronic solvers applied by DYN3D, PARCS and GeN-Foam to steady-state calculations of SFR cores.
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Criticality Analysis for BWR Spent Fuel Based on the Burnup Credit Evaluation from Full Core Simulations. APPLIED SCIENCES-BASEL 2021. [DOI: 10.3390/app11041498] [Citation(s) in RCA: 2] [Impact Index Per Article: 0.7] [Reference Citation Analysis] [Abstract] [Track Full Text] [Subscribe] [Scholar Register] [Indexed: 11/16/2022]
Abstract
This study performed criticality analysis for the GBC-68 storage cask loaded with boiling water reactor (BWR) spent fuel at the discharged burnups obtained from the full-core simulations. The analysis was conducted for: (1) different reloading scenarios; (2) target burnups; and (3) two fuel assembly types—GE14 and SVEA100—to estimate the impact each of the three factors has on the cask reactivity. The BWR spent fuel composition was estimated using the results of the nodal analysis for the advanced boiling water reactor (ABWR) core model developed in this study. The nodal calculations provided realistic operating data and axial burnup and coolant density profiles, for each fuel assembly in the reactor core. The estimated cask’s keff were compared with the fresh fuel and peak reactivity standards to identify the benefit of the burnup credit method applied to the BWR spent fuel at their potential discharge burnups. The analysis identified the significant cask criticality reduction from employing the burnup credit approach compared to the conventional fresh fuel approach. However, the criticality reduction was small compared to the peak reactivity approach, and could even disappear for low burnt fuel assemblies from non-optimal reloading patterns. In terms of cask manufacturing, the potential financial benefit from using the burnup credit approach was estimated to be USD 3.3 million per reactor cycle.
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A Comparison of Advanced Boiling Water Reactor Simulations between Serpent/CTF and Polaris/DYN3D: Steady State Operational Characteristics and Burnup Evolution. ENERGIES 2021. [DOI: 10.3390/en14040838] [Citation(s) in RCA: 2] [Impact Index Per Article: 0.7] [Reference Citation Analysis] [Abstract] [Track Full Text] [Subscribe] [Scholar Register] [Indexed: 11/16/2022]
Abstract
High fidelity modelling for nuclear power plant analysis is becoming more common due to advances in modelling software and the availability of high-performance computers. However, to design, develop and regulate new light water nuclear reactors there are, up until now, limited requirements for high fidelity methods due to the already well-established computational methods already being widely accepted. This article explores the additional detail which can be obtained when using high fidelity methods through Monte Carlo/Sub-channel analysis compared to industrial methods of cross-section/nodal analysis using the Advanced Boiling Water Reactor as a case study. This case study was chosen due to the challenges in modelling two phase flow and the high levels of heterogeneity within the fuel assembly design. The article investigates how to implement such an approach, from a bottom-up procedure, by analysing each stage of the modelling process.
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Périn Y, Travleev A, Zilly M. COUPLED TRANSIENT ANALYSIS OF A CORE WITH FUEL ASSEMBLY BOWING WITH A HYBRID CTF/DYN3D MODEL. EPJ WEB OF CONFERENCES 2021. [DOI: 10.1051/epjconf/202124706036] [Citation(s) in RCA: 0] [Impact Index Per Article: 0] [Reference Citation Analysis] [Abstract] [Track Full Text] [Journal Information] [Subscribe] [Scholar Register] [Indexed: 11/14/2022] Open
Abstract
Fuel assembly bowing is a known phenomenon observed in many PWR reactors all over the world. The phenomenon is relevant to safety as it can lead to increased water gaps between assemblies which results in higher pin peaking factors. The goal of the present study is to assess the effect of assembly bowing not only for stead-state nominal conditions but also during a transient. The selected transient is the loss of one reactor coolant pump as it can be limiting especially regarding the Departure from Nucleate Boiling (DNB) safety criterion. This study focuses on an extreme case where the bowing is simulated in the core hot assembly by keeping the water gap constant over the whole core active length. The resulting cross-sections and form functions obtained from a 2d infinite lattice model are used in the nodal diffusion code DYN3D applying its pin-by-pin reconstruction method. For the transient simulation, DYN3D is coupled with the thermal-hydraulics subchannel code CTF on the SALOME platform. Several modelling options are compared: nominal geometry for neutronics and thermal-hydraulics (TH); mixed: neutronics with increased water gap, TH with nominal geometry; and increased water gap for both neutronics and TH. The results confirm that the increased water gap should be considered in both models in order to reduce the conservatism.
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Pautz A, Zwermann W. TRANSIENT CALCULATIONS OF SPERT III EXPERIMENTS. EPJ WEB OF CONFERENCES 2021. [DOI: 10.1051/epjconf/202124707017] [Citation(s) in RCA: 0] [Impact Index Per Article: 0] [Reference Citation Analysis] [Abstract] [Track Full Text] [Journal Information] [Subscribe] [Scholar Register] [Indexed: 11/14/2022] Open
Abstract
Cold-startup and hot-standby reactivity accident tests conducted at the SPERT III E-core research reactor are analysed with the coupled neutron-kinetic/thermal-hydraulic code system DYN3D-ATHLET. Homogenised 2-group cross sections for DYN3D are thereby generated with the Monte Carlo neutron transport code Serpent 2 in combination with the ENDF/B-VII.1 cross section library. Results in terms of maximum power, energy release, and reactivity compensation are in good agreement with the experimental values. The time-dependent contributions to the reactivity feedback are investigated for both a cold-startup test and a hot-standby test. These findings prove the suitability of the combined application of the simulation codes to predict the reactor dynamic behaviour in the event of prompt-critical and super-prompt critical transients even for small reactor cores. Furthermore, static core characteristics of the SPERT III E-core reactor at cold-startup condition are analysed with using a static DYN3D model, a detailed Serpent reference model, and a simplified Serpent model consistent with the DYN3D model. The critical control rod position and the excess reactivities of both the control rods and the transient rod obtained with the Serpent reference model are consistent with the experimental values. For the same parameters, the DYN3D model is in good agreement with the Serpent simplified model.
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21
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Viebach M, Lange C, Seidl M, Bilodid Y, Hurtado A. NEUTRON NOISE PATTERNS FROM COUPLED FUEL-ASSEMBLY VIBRATIONS. EPJ WEB OF CONFERENCES 2021. [DOI: 10.1051/epjconf/202124702015] [Citation(s) in RCA: 1] [Impact Index Per Article: 0.3] [Reference Citation Analysis] [Abstract] [Track Full Text] [Journal Information] [Subscribe] [Scholar Register] [Indexed: 11/14/2022] Open
Abstract
The neutron flux fluctuation magnitude of KWU-built PWRs shows a hitherto unexplained correlation with the types of loaded fuel assemblies. Also, certain measured long-range neutron flux fluctuation patterns in neighboring core quadrants still lack a closed understanding of their origin. The explanation of these phenomena has recently revived a new interest in neutron noise research.
The contribution at hand investigates the idea that a synchronized coolant-driven vibration of major parts of the fuel-assembly ensemble leads to these phenomena. Starting with an assumed mode of such collective vibration, the resulting effects on the time-dependent neutron-flux distribution are analyzed via a DYN3D simulation. A three-dimensional representation of the time-dependent bow of all fuel assemblies is taken into account as a nodal DYN3D feedback parameter by time-dependent variations of the fuel-assembly pitch. The impact of its variation on the cross sections is quantified using a cross-section library that is generated from the output of corresponding CASMO5 calculations.
The DYN3D simulation qualitatively reproduces the measured neutron-flux fluctuation patterns. The magnitude of the fluctuations and its radial dependence are comparable to the measured details. The results imply that collective fuel-assembly vibrations are a promising candidate for being the key to understand long-known fluctuation patterns in KWU built PWRs. Further research should elaborate on possible excitation mechanisms of the assumed vibration modes.
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Yu H, Ju H, Wang M, Zhang J, Qiu S, Tian W, Su G. Study of boron diffusion models and dilution accidents in nuclear reactor: A comprehensive review. ANN NUCL ENERGY 2020. [DOI: 10.1016/j.anucene.2020.107659] [Citation(s) in RCA: 4] [Impact Index Per Article: 1.0] [Reference Citation Analysis] [Track Full Text] [Journal Information] [Subscribe] [Scholar Register] [Indexed: 11/26/2022]
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23
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Litskevich D, Atkinson S, Davies S. Verification of the current coupling collision probability method with orthogonal flux expansion for the assembly calculations. PROGRESS IN NUCLEAR ENERGY 2020. [DOI: 10.1016/j.pnucene.2020.103562] [Citation(s) in RCA: 4] [Impact Index Per Article: 1.0] [Reference Citation Analysis] [Track Full Text] [Subscribe] [Scholar Register] [Indexed: 11/30/2022]
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24
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Burnup Credit Evaluation for BWR Spent Fuel from Full Core Calculations. APPLIED SCIENCES-BASEL 2020. [DOI: 10.3390/app10217549] [Citation(s) in RCA: 2] [Impact Index Per Article: 0.5] [Reference Citation Analysis] [Abstract] [Track Full Text] [Subscribe] [Scholar Register] [Indexed: 11/17/2022]
Abstract
Due to the challenges of spent fuel accumulation, the nuclear industry is exploring more cost-effective solutions for spent fuel management. The burnup-credit method, in application for storage and transport of the spent fuel, gained traction over recent decades since it can remove the over-conservatism of the “fresh-fuel” approach. The presented research is focused on creating an innovative, best estimate approach of the burnup-credit method for boiling water reactor (BWR) spent fuel based on the results of neutronic/thermal-hydraulic coupled full core simulations. The analysis is performed using a Polaris/DYN3D sequence. Four different shuffling procedures were used to estimate the possible range of the BWR fuel discharged burnup variation. The results showed a strong influence of the shuffling procedure on the final burnup distribution. Moreover, a comparison of the 2D lattice and 3D coupled nodal approaches was conducted for the criticality estimation of single fuel assemblies. The analysis revealed substantial improvement in criticality curves obtained with the full-core model. Finally, it was shown that the benefit from the burnup-credit method is larger in the case of more optimal fuel utilisation-based shuffling procedures. The new approach developed here delivers a promising basis for future industrial optimisation procedures and thus cost optimisation.
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25
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On Singular Perturbation of Neutron Point Kinetics in the Dynamic Model of a PWR Nuclear Power Plant. SCI 2020. [DOI: 10.3390/sci2020036] [Citation(s) in RCA: 1] [Impact Index Per Article: 0.3] [Reference Citation Analysis] [Abstract] [Track Full Text] [Journal Information] [Subscribe] [Scholar Register] [Indexed: 11/17/2022] Open
Abstract
This short communication makes use of the principle of singular perturbation to approximate the ordinary differential equation (ODE) of prompt neutron (in the point kinetics model) as an algebraic equation. This approximation is shown to yield a large gain in computational efficiency without compromising any significant accuracy in the numerical simulation of primary coolant system dynamics in a PWR nuclear power plant. The approximate (i.e., singularly perturbed) model has been validated with a numerical solution of the original set of neutron point-kinetic and thermal–hydraulic equations. Both models use variable-step Runge–Kutta numerical integration.
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Papadionysiou M, Seongchan K, Hursin M, Vasiliev A, Ferroukhi H, Pautz A, Joo HG. Assessment of nTRACER and PARCS Performance for VVER Configurations. NUCL SCI ENG 2020. [DOI: 10.1080/00295639.2020.1753418] [Citation(s) in RCA: 3] [Impact Index Per Article: 0.8] [Reference Citation Analysis] [Track Full Text] [Journal Information] [Subscribe] [Scholar Register] [Indexed: 10/24/2022]
Affiliation(s)
| | - Kim Seongchan
- Seoul National University, Department of Nuclear Engineering, Seoul, Korea
| | - Mathieu Hursin
- Paul Scherrer Institut, Nuclear Energy and Safety Division, Villigen AG, Switzerland
| | - Alexander Vasiliev
- Paul Scherrer Institut, Nuclear Energy and Safety Division, Villigen AG, Switzerland
| | - Hakim Ferroukhi
- Paul Scherrer Institut, Nuclear Energy and Safety Division, Villigen AG, Switzerland
| | - Andreas Pautz
- Paul Scherrer Institut, Nuclear Energy and Safety Division, Villigen AG, Switzerland
| | - Han Gyu Joo
- Seoul National University, Department of Nuclear Engineering, Seoul, Korea
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On Singular Perturbation of Neutron Point Kinetics in the Dynamic Model of a PWR Nuclear Power Plant. SCI 2020. [DOI: 10.3390/sci2020030] [Citation(s) in RCA: 3] [Impact Index Per Article: 0.8] [Reference Citation Analysis] [Abstract] [Track Full Text] [Journal Information] [Subscribe] [Scholar Register] [Indexed: 11/16/2022] Open
Abstract
This short communication makes use of the principle of singular perturbation to approximate the ordinary differential equation (ODE) of prompt neutron (in the point kinetics model) as an algebraic equation. This approximation is shown to yield a large gain in computational efficiency without compromising any significant accuracy in the numerical simulation of primary coolant system dynamics in a PWR nuclear power plant. The approximate (i.e., singularly perturbed) model has been validated with a numerical solution of the original set of neutron point-kinetic and thermal–hydraulic equations. Both models use variable-step Runge–Kutta numerical integration.
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28
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Wang J, Wang Q, Ding M. Review on neutronic/thermal-hydraulic coupling simulation methods for nuclear reactor analysis. ANN NUCL ENERGY 2020. [DOI: 10.1016/j.anucene.2019.107165] [Citation(s) in RCA: 18] [Impact Index Per Article: 4.5] [Reference Citation Analysis] [Track Full Text] [Journal Information] [Subscribe] [Scholar Register] [Indexed: 10/25/2022]
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29
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Uncertainty and sensitivity analysis of PWR mini-core transients in the presence of nuclear data uncertainty using non-parametric tolerance limits. ANN NUCL ENERGY 2020. [DOI: 10.1016/j.anucene.2019.107146] [Citation(s) in RCA: 5] [Impact Index Per Article: 1.3] [Reference Citation Analysis] [Track Full Text] [Journal Information] [Subscribe] [Scholar Register] [Indexed: 11/19/2022]
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30
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Reactor core analysis through the SP3-ACMFD approach Part II: Transient solution. NUCLEAR ENGINEERING AND TECHNOLOGY 2020. [DOI: 10.1016/j.net.2019.07.024] [Citation(s) in RCA: 0] [Impact Index Per Article: 0] [Reference Citation Analysis] [Track Full Text] [Subscribe] [Scholar Register] [Indexed: 11/18/2022]
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31
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On the numerical modelling of frozen walls in a molten salt fast reactor. NUCLEAR ENGINEERING AND DESIGN 2019. [DOI: 10.1016/j.nucengdes.2019.110290] [Citation(s) in RCA: 0] [Impact Index Per Article: 0] [Reference Citation Analysis] [Track Full Text] [Subscribe] [Scholar Register] [Indexed: 11/22/2022]
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32
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Viebach M, Lange C, Bernt N, Seidl M, Hennig D, Hurtado A. Simulation of low-frequency PWR neutron flux fluctuations. PROGRESS IN NUCLEAR ENERGY 2019. [DOI: 10.1016/j.pnucene.2019.103039] [Citation(s) in RCA: 5] [Impact Index Per Article: 1.0] [Reference Citation Analysis] [Track Full Text] [Subscribe] [Scholar Register] [Indexed: 10/26/2022]
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33
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Molnar B, Tolnai G, Legrady D. A GPU-based direct Monte Carlo simulation of time dependence in nuclear reactors. ANN NUCL ENERGY 2019. [DOI: 10.1016/j.anucene.2019.03.024] [Citation(s) in RCA: 6] [Impact Index Per Article: 1.2] [Reference Citation Analysis] [Track Full Text] [Journal Information] [Subscribe] [Scholar Register] [Indexed: 10/27/2022]
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34
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Nikitin E, Fridman E. Modeling of the FFTF isothermal physics tests with the Serpent and DYN3D codes. ANN NUCL ENERGY 2019. [DOI: 10.1016/j.anucene.2019.06.058] [Citation(s) in RCA: 4] [Impact Index Per Article: 0.8] [Reference Citation Analysis] [Track Full Text] [Journal Information] [Subscribe] [Scholar Register] [Indexed: 11/16/2022]
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35
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Hádek J, Meca R. Analysis of uncontrolled dilution of boric acid concentration in the reactor VVER-1000/320. KERNTECHNIK 2019. [DOI: 10.3139/124.190015] [Citation(s) in RCA: 0] [Impact Index Per Article: 0] [Reference Citation Analysis] [Abstract] [Track Full Text] [Journal Information] [Subscribe] [Scholar Register] [Indexed: 11/20/2022]
Abstract
AbstractThe paper gives a description of conservative analysis of initiating event associated with uncontrolled dilution of boric acid concentration in the reactor VVER-1000/320 of Temelín NPP. This event is included in the group of beyond design basis accidents. The aim of analysis is to determine also the time interval which is necessary for interventions leading to the deterrence of fuel damage. The failure of operator intervention to isolate dilution routes at intervals shorter than 30 min is assumed. Since the response of the whole NPP system influences the course of safety important parameters of the reactor core, the calculations were made by an externally coupled version of the 3D reactor dynamic code DYN3D and the thermohydraulic system code ATHLET. It is shown that, in addition to exceeding the DNBR limit of more than 99 min from the start of the transient, the remaining safety acceptance criteria will not be violated until the end of the calculation.
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Affiliation(s)
- J. Hádek
- 1ÚJV Řež, a. s., Hlavní 130, 250 68 Husinec-Řež Czech Republic
| | - R. Meca
- 1ÚJV Řež, a. s., Hlavní 130, 250 68 Husinec-Řež Czech Republic
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36
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Ieremenko M, Ovdiienko I. Adaptation of the gas gap simplified model in DYN3D code to new types of fuel. KERNTECHNIK 2019. [DOI: 10.3139/124.190029] [Citation(s) in RCA: 0] [Impact Index Per Article: 0] [Reference Citation Analysis] [Abstract] [Track Full Text] [Journal Information] [Subscribe] [Scholar Register] [Indexed: 11/20/2022]
Abstract
AbstractCurrently, new types of fuel are being considered to be introduced or already in the introduction process at Ukrainian NPPs with WWER. By means of a new version of the TRANSURANUS code, new functions of the gas gap thickness in dependence on the burnup have been created and implemented into the gas gap model of the reactor dynamics code DYN3D. These new functions cover all actual and perspective fuel types for the Ukrainian NPPs with WWER.
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Affiliation(s)
- M. Ieremenko
- 1State Scientific and Technical Centre on Nuclear and Radiation Safety (SSTC NRS), Stusa st. 35–37, 03142 Kyiv Ukraine
| | - Iu. Ovdiienko
- 1State Scientific and Technical Centre on Nuclear and Radiation Safety (SSTC NRS), Stusa st. 35–37, 03142 Kyiv Ukraine
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37
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Kliem S, Grahn A, Bilodid Y, Höhne T. A realistic approach for the assessment of the consequences of heterogeneous boron dilution events in pressurized water reactors. NUCLEAR ENGINEERING AND DESIGN 2019. [DOI: 10.1016/j.nucengdes.2019.04.038] [Citation(s) in RCA: 3] [Impact Index Per Article: 0.6] [Reference Citation Analysis] [Track Full Text] [Subscribe] [Scholar Register] [Indexed: 11/29/2022]
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38
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Sun L, Peng M, Xia G, Wang J, Li R. Coupling simulation of neutron kinetics core model with CFD of IPWR steam line break accident. KERNTECHNIK 2019. [DOI: 10.3139/124.110979] [Citation(s) in RCA: 0] [Impact Index Per Article: 0] [Reference Citation Analysis] [Abstract] [Track Full Text] [Journal Information] [Subscribe] [Scholar Register] [Indexed: 11/20/2022]
Abstract
AbstractIn this paper, development of coupled codes using two-group neutron diffusion kinetics code and computational fluid dynamics (CFD) solver Fluent has been introduced. Way of coupling, time step control algorithm and spatial mesh overlays have been summarized in detail which are basic components and challenges of the coupling methodologies. The implement and verification of coupled code have been modeled on integral pressurized water reactor (IPWR) IP200 with hexagonal fuel assembly in the core and once-through steam generators. The steam line break core transient was analyzed in coupled code simulation of a core boundary conditions derived from system code simulation results. The results presented transient three-dimensional distribution of the key operation parameters such as reactor power and coolant temperature, also demonstrated the inherent safety features of IP200. The current work will bring about the ability to explore multi-scale and multi-dimensional safety transient evaluations and give more precise neutronics/thermal-hydraulics mapping.
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Affiliation(s)
- L. Sun
- 1Fundamental Science on Nuclear Safety and Simulation Technology Laboratory Harbin Engineering University, Harbin, Heilongjiang, 150001 China
| | - M. Peng
- 1Fundamental Science on Nuclear Safety and Simulation Technology Laboratory Harbin Engineering University, Harbin, Heilongjiang, 150001 China
| | - G. Xia
- 1Fundamental Science on Nuclear Safety and Simulation Technology Laboratory Harbin Engineering University, Harbin, Heilongjiang, 150001 China
| | - J. Wang
- 2Department of Engineering Physics, University of Wisconsin-Madison, Madison, Wisconsin, 53706 United States
| | - R. Li
- 1Fundamental Science on Nuclear Safety and Simulation Technology Laboratory Harbin Engineering University, Harbin, Heilongjiang, 150001 China
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39
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Terlizzi S, Kotlyar D. Fission Matrix Decomposition Method for Criticality Calculations: Theory and Proof of Concept. NUCL SCI ENG 2019. [DOI: 10.1080/00295639.2019.1583948] [Citation(s) in RCA: 5] [Impact Index Per Article: 1.0] [Reference Citation Analysis] [Track Full Text] [Journal Information] [Subscribe] [Scholar Register] [Indexed: 10/27/2022]
Affiliation(s)
- Stefano Terlizzi
- Georgia Institute of Technology, Department of Nuclear and Radiological Engineering, 770 State Street, Atlanta, Georgia 30332-0745
| | - Dan Kotlyar
- Georgia Institute of Technology, Department of Nuclear and Radiological Engineering, 770 State Street, Atlanta, Georgia 30332-0745
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40
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Applying the Serpent-DYN3D code sequence for the decay heat analysis of metallic fuel sodium fast reactor. ANN NUCL ENERGY 2019. [DOI: 10.1016/j.anucene.2018.11.020] [Citation(s) in RCA: 4] [Impact Index Per Article: 0.8] [Reference Citation Analysis] [Track Full Text] [Journal Information] [Subscribe] [Scholar Register] [Indexed: 11/20/2022]
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41
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On a Roadmap for Future Industrial Nuclear Reactor Core Simulation in the U.K. to Support the Nuclear Renaissance. ENERGIES 2018. [DOI: 10.3390/en11123509] [Citation(s) in RCA: 7] [Impact Index Per Article: 1.2] [Reference Citation Analysis] [Abstract] [Track Full Text] [Subscribe] [Scholar Register] [Indexed: 11/16/2022]
Abstract
The U.K. has initiated the nuclear renaissance by contracting for the first two new plants and announcing further new build projects. The U.K. government has recently started to support this development with the announcement of a national programme of nuclear innovation. The aim of this programme with respect to modelling and simulation is foreseen to fulfil the demand in education and the build-up of a reasonably qualified workforce, as well as the development and application of a new state-of-the-art software environment for improved economics and safety. This document supports the ambition to define a new approach to the structured development of nuclear reactor core simulation that is based on oversight instead of looking at detail problems and the development of single tools for these specific detail problems. It is based on studying the industrial demand to bridge the gap in technical innovation that can be derived from basic research in order to create a tailored industry solution to set the new standard for reactor core modelling and simulation for the U.K. However, finally, a technical requirements specification has to be developed alongside the strategic approach to give code developers a functional specification that they can use to develop the tools for the future. Key points for a culture change to the application of modern technologies are identified in the use of DevOps in a double-strata approach to academic and industrial code development. The document provides a novel, strategic approach to achieve the most promising final product for industry, and to identify the most important points for improvement.
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42
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Bilodid Y, Fridman E, Kotlyar D, Shwageraus E. Explicit decay heat calculation in the nodal diffusion code DYN3D. ANN NUCL ENERGY 2018. [DOI: 10.1016/j.anucene.2018.07.045] [Citation(s) in RCA: 5] [Impact Index Per Article: 0.8] [Reference Citation Analysis] [Track Full Text] [Journal Information] [Subscribe] [Scholar Register] [Indexed: 11/28/2022]
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43
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44
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Nikitin E, Fridman E. Extension of the reactor dynamics code DYN3D to SFR applications – Part II: Validation against the Phenix EOL control rod withdrawal tests. ANN NUCL ENERGY 2018. [DOI: 10.1016/j.anucene.2018.05.016] [Citation(s) in RCA: 6] [Impact Index Per Article: 1.0] [Reference Citation Analysis] [Track Full Text] [Journal Information] [Subscribe] [Scholar Register] [Indexed: 10/16/2022]
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45
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Nikitin E, Fridman E. Extension of the reactor dynamics code DYN3D to SFR applications – Part III: Validation against the initial phase of the Phenix EOL natural convection test. ANN NUCL ENERGY 2018. [DOI: 10.1016/j.anucene.2018.05.017] [Citation(s) in RCA: 10] [Impact Index Per Article: 1.7] [Reference Citation Analysis] [Track Full Text] [Journal Information] [Subscribe] [Scholar Register] [Indexed: 11/16/2022]
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46
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Extension of the reactor dynamics code DYN3D to SFR applications – Part I: Thermal expansion models. ANN NUCL ENERGY 2018. [DOI: 10.1016/j.anucene.2018.05.015] [Citation(s) in RCA: 11] [Impact Index Per Article: 1.8] [Reference Citation Analysis] [Track Full Text] [Journal Information] [Subscribe] [Scholar Register] [Indexed: 11/20/2022]
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47
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Hádek J, Meca R. Contribution to the validation of the VVER-1000 Temelin NPP computing model for the ATHLET/DYN3D coupled codes. KERNTECHNIK 2018. [DOI: 10.3139/124.110902] [Citation(s) in RCA: 1] [Impact Index Per Article: 0.2] [Reference Citation Analysis] [Abstract] [Track Full Text] [Journal Information] [Subscribe] [Scholar Register] [Indexed: 11/20/2022]
Abstract
Abstract
This paper contains a description and evaluation of the thermal-hydraulic calculation of VVER-1000 transient connected with steam dump to atmosphere (SDA) opening during decreased reactor power to 20% of nominal power (Nnom). The calculation was performed with the thermal-hydraulic system program ATHLET coupled with the 3-D reactor dynamic code DYN3D. A comparison with the experiment was performed on the base of measured values during SDA project function test on the VVER-1000 Temelín NPP Unit 2. Results obtained from calculated vs. experimental values contribute to the validation of the ATHLET/DYN3D coupling.
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Affiliation(s)
- J. Hádek
- ÚJV Řež , a. s., Hlavní 130 , 250 68 Husinec-Řež Czech Republic
| | - R. Meca
- ÚJV Řež , a. s., Hlavní 130 , 250 68 Husinec-Řež Czech Republic
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48
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Bilodid Y, Grundmann U, Kliem S. The HEXNEM3 nodal flux expansion method for the hexagonal geometry in the code DYN3D. ANN NUCL ENERGY 2018. [DOI: 10.1016/j.anucene.2018.02.037] [Citation(s) in RCA: 8] [Impact Index Per Article: 1.3] [Reference Citation Analysis] [Track Full Text] [Journal Information] [Subscribe] [Scholar Register] [Indexed: 11/29/2022]
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49
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Henry R, Tiselj I, Snoj L. Transient CFD/Monte-Carlo Neutron Transport Coupling Scheme for simulation of a control rod extraction in TRIGA reactor. NUCLEAR ENGINEERING AND DESIGN 2018. [DOI: 10.1016/j.nucengdes.2018.03.015] [Citation(s) in RCA: 8] [Impact Index Per Article: 1.3] [Reference Citation Analysis] [Track Full Text] [Subscribe] [Scholar Register] [Indexed: 11/29/2022]
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50
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Validation of the DYN3D-Serpent code system for SFR cores using selected BFS experiments. Part II: DYN3D calculations. ANN NUCL ENERGY 2018. [DOI: 10.1016/j.anucene.2017.12.036] [Citation(s) in RCA: 1] [Impact Index Per Article: 0.2] [Reference Citation Analysis] [Track Full Text] [Journal Information] [Subscribe] [Scholar Register] [Indexed: 11/22/2022]
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