| 1. |
- Mikityuk, K., et al.
(author)
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Horizon-2020 ESFR-SMART project on Sodium Fast Reactor Safety: status after 18 months
- 2019
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Conference paper (other academic/artistic)abstract
- To improve the public acceptance of the future nuclear power in Europe we have to demonstrate that the new reactors have significantly higher safety level compared to traditional reactors. The ESFR-SMART project (European Sodium Fast Reactor Safety Measures Assessment and Research Tools) aims at enhancing further the safety of Generation-IV SFRs and in particular of the commercial-size European Sodium Fast Reactor (ESFR) in accordance with the European Sustainable Nuclear Industrial Initiative (ESNII) roadmap and in close cooperation with the Advanced Sodium Technological Reactor for Industrial Demonstration (ASTRID) program. The project aims at 5 specific objectives: 1. Produce new experimental data in order to support calibration and validation of the computational tools for each defence-in-depth level. 2. Test and qualify new instrumentations in order to support their utilization in the reactor protection system. 3. Perform further calibration and validation of the computational tools for each defence-in-depth level in order to support safety assessments of Generation-IV SFRs, using the data produced in the project as well as selected legacy data. 4. Select, implement and assess new safety measures for the commercial-size ESFR, using the GIF methodologies, the FP7 CP-ESFR project legacy, the calibrated and validated codes and being in accordance with the update of the European and international safety frameworks taking into account the Fukushima accident. 5. Strengthen and link together new networks, in particular, the network of the European sodium facilities and the network of the European students working on the SFR technology. By addressing the industry, policy makers and general public, the project is expected to make a meaningful impact on economics, environment, EU policy and society. Selected results and milestones achieved during the first eighteen months of the project will be briefly presented, including − proposal of new safety measures for ESFR; − evaluation of ESFR core performance; − benchmarking of codes; − experimental programs; and − education and training.
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| 2. |
- Demazière, C., et al.
(author)
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Development of three-dimensional capabilities for modelling stationary fluctuations in nuclear reactor cores
- 2015
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In: Annals of Nuclear Energy. - : Elsevier Ltd. - 0306-4549 .- 1873-2100. ; 84, s. 19-30
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Journal article (peer-reviewed)abstract
- This paper presents the development of a numerical tool meant at modelling the effect of stationary fluctuations in nuclear cores for systems cooled with either liquid water or boiling water. The originating fluctuations are defined for the variables describing the boundary conditions of the system, i.e. inlet velocity, inlet enthalpy, and outlet pressure. The tool then determines in the frequency domain the three-dimensional distributions within the core of the corresponding fluctuations in neutron flux, coolant density, coolant velocity, coolant enthalpy, and fuel temperature. The tool is thus based on the simultaneous modelling of neutron transport, fluid dynamics, and heat transfer in a truly integrated and fully coupled manner. The modelling of neutron transport relies on the two-group diffusion approximation and a spatial discretization based on finite differences. The modelling of fluid dynamics is performed using the Homogeneous Equilibrium Model, with a void fraction correction based on a pre-computed distribution of the static slip ratio (when two-phase flow conditions are encountered). Heat conduction in the fuel pins is also accounted for, and the heat transfer between the fuel pins and the coolant is modelled also using a pre-computed distribution of the heat transfer coefficient. The spatial discretization of the fluid dynamic and heat transfer problems is carried out using finite volumes. The tool, currently entirely Matlab based, requires minimal input data, mostly in form of the three-dimensional distributions of the macroscopic cross-sections and their relative dependence on coolant density and fuel temperature, the point-kinetic parameters of the core, as well as the three-dimensional distribution of the slip ratio (in case of two-phase flow conditions) and of the heat transfer coefficient. Such data can be provided by any static core simulator that thus needs to be run prior to using the present tool. In addition to briefly summarizing the different test cases used to verify the code, the paper also presents the results of simulations performed for a typical Pressurized Water Reactor and for a typical Boiling Water Reactor, as illustrations of the capabilities of the tool.
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| 3. |
- Durrant, A., et al.
(author)
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Detection and localisation of multiple in-core perturbations with neutron noise-based self-supervised domain adaptation
- 2021
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In: Proc. Int. Conf. Mathematics and Computational Methods Applied to Nuclear Science and Engineering (M&C2021). - 9781713886310
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Conference paper (peer-reviewed)abstract
- The use of non-intrusive techniques for monitoring nuclear reactors is becoming more vital as western fleets age. As a consequence, the necessity to detect more frequently occurring operational anomalies is of upmost interest. Here, noise diagnostics — the analysis of small stationary deviations of local neutron flux around its time-averaged value — is employed aiming to unfold from detector readings the nature and location of driving perturbations. Given that in-core instrumentation of western-type light-water reactors are scarce in number of detectors, rendering formal inversion of the reactor transfer function impossible, we propose to utilise advancements in Machine Learning and Deep Learning for the task of unfolding. This work presents an approach to such a task doing so in the presence of multiple and simultaneously occurring perturbations or anomalies. A voxel-wise semantic segmentation network is proposed to determine the nature and source location of multiple and simultaneously occurring perturbations in the frequency domain. A diffusion-based core simulation tool has been employed to provide simulated training data for two reactors. Additionally, we work towards the application of the aforementioned approach to real measurements, introducing a self-supervised domain adaptation procedure to align the representation distributions of simulated and real plant measurements.
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| 4. |
- Tran, Hoai Nam, 1981, et al.
(author)
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A multi-group neutron noise simulator for fast reactors
- 2013
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In: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 62, s. 158-169
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Journal article (peer-reviewed)abstract
- A neutron noise simulator has been developed for fast reactors based on diffusion theory with multi-energy groups and several groups of delayed neutron precursors. The tool is expected to be applicable for core monitoring of fast reactors and also for other reactor types with hexagonal fuel assemblies. The noise sources are modeled through small stationary fluctuations of macroscopic cross sections, and the induced first order noise is solved fully in the frequency domain. Numerical algorithms are implemented for solving both the static and noise equations using finite differences for spatial discretization, where a hexagonal assembly is radially divided into finer triangular meshes. A coarse mesh finite difference (CMFD) acceleration has been used for accelerating the convergence of both the static and noise calculations. Numerical calculations have been performed for the ESFR core with 33 energy groups and 8 groups of delayed neutron precursors using the cross section data generated by the ERANOS code. The results of the static state have been compared with those obtained using ERANOS. The results show an adequate agreement between the two calculations. Noise calculations for the ESFR core have also been performed and demonstrated with an assumption of the perturbation of the absorption cross section located at the central fuel ring.
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| 5. |
- Viebach, M., et al.
(author)
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Verification of the code DYN3D for calculations of neutron flux fluctuations
- 2022
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In: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 166
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Journal article (peer-reviewed)abstract
- Insufficiently explained magnitudes and patterns of flux fluctuation observed mainly in KWU PWRs are recently investigated by various European institutions. Among the numerical tools used to investigate the neutron flux fluctuations is the time-domain reactor dynamics code DYN3D. As DYN3D and comparable codes have not been developed with the primary intention to simulate low-amplitude neutron flux fluctuations, their applicability in this field has to be verified. In order to contribute to the verification of DYN3D for the simulation of neutron flux fluctuations, two special cases of perturbations of the neutron flux (a localized absorber of variable/oscillatory strength and a travelling oscillatory perturbation) are considered with DYN3D on the one hand and with the frequency-domain neutron noise tool CORE SIM as well as analytical frequency-domain approaches, respectively, on the other hand. The obtained results are compared with respect to the distributions of the amplitude and the phase of the induced neutron flux fluctuations. The comparisons are repeated with varied amplitudes and frequencies of the perturbation. The results agree well both qualitatively and quantitatively for each of the conducted calculations. The remaining deviations between the DYN3D results and the reference results exhibit a dependence on the perturbation magnitude, which is attributed to the neglect of higher-order terms (linear theory) of the perturbed quantities in the calculation of the reference solutions.
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