| 1. |
- Calivá, Francesco, et al.
(författare)
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A deep learning approach to anomaly detection in nuclear reactors
- 2018
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Ingår i: Proceedings of the International Joint Conference on Neural Networks. ; 2018-July
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Konferensbidrag (refereegranskat)abstract
- In this work, a novel deep learning approach to unfold nuclear power reactor signals is proposed. It includes a combination of convolutional neural networks (CNN), denoising autoencoders (DAE) and k-means clustering of representations. Monitoring nuclear reactors while running at nominal conditions is critical. Based on analysis of the core reactor neutron flux, it is possible to derive useful information for building fault/anomaly detection systems. By leveraging signal and image pre-processing techniques, the high and low energy spectra of the signals were appropriated into a compatible format for CNN training. Firstly, a CNN was employed to unfold the signal into either twelve or forty-eight perturbation location sources, followed by a k-means clustering and k-Nearest Neighbour coarse-to-fine procedure, which significantly increases the unfolding resolution. Secondly, a DAE was utilised to denoise and reconstruct power reactor signals at varying levels of noise and/or corruption. The reconstructed signals were evaluated w.r.t. their original counter parts, by way of normalised cross correlation and unfolding metrics. The results illustrate that the origin of perturbations can be localised with high accuracy, despite limited training data and obscured/noisy signals, across various levels of granularity.
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| 2. |
- Chambon, Amalia, 1986, et al.
(författare)
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A deterministic against Monte-Carlo depletion calculation benchmark for JHR core configurations
- 2017
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Ingår i: Int. Conf. Mathematics & Computational Methods Applied to Nuclear Science & Engineering (M&C 2017), Jeju, Korea, April 16-20, 2017.
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Konferensbidrag (refereegranskat)abstract
- The Jules Horowitz Reactor (JHR) is the next international Material-Testing Reactor (MTR) under construction in the south of France at CEA Cadarache research center. Its first criticality is foreseen by the end of the decade. The innovative character of the JHR led to the development of a specific neutronic calculation scheme called HORUS3D/N for performing design and safety studies. HORUS3D/N is based onthe deterministic codes APOLLO2 and CRONOS2 and on the European nuclear data library JEFF-3.1.1. Up to now, the biases and uncertainties due to the HORUS3D/N calculation scheme in depletion have been assessed by comparing HORUS3D/N deterministic calculations with 2D APOLLO2-MOC reference route calculations. The recent development of the Monte-Carlo code TRIPOLI-4® in its depletion mode(TRIPOLI-4®D) offers the opportunity to study the JHR 3D core configurations under fuel depletion conditions. This paper presents the first CRONOS2/TRIPOLI-4®D benchmark results obtained for 3 core configurations of interest including control rods and experimental devices up to a burnup value of 60 GWd/tHM. The main parameters of interest are the reactivity and the isotopic concentrations as functions of burnup. This first study of actual JHR configurations in depletion demonstrates that CRONOS2underestimates the reactivity for burnups lower than 8 GWd/tHM and overestimates it for higher burnups, with respect to the TRIPOLI-4®D predictions. A good agreement between the two codes is observed concerning the 235U consumption with discrepancies values less than -0.5% at 60 GWd/tHM. Nevertheless, a global CRONOS2 overestimation of the plutonium inventory can be noticed. Compared with 3D assembly calculation in an infinite lattice, this overestimation was tracked down to the condensation of the nuclear constants provided by APOLLO2, showing the limits of a two steps calculation.
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| 3. |
- Chambon, Amalia, 1986, et al.
(författare)
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VALIDATION OF HORUS3D/N AGAINST TRIPOLI-4®D FOR CORE DEPLETION CALCULATION OF THE JULES HOROWITZ REACTOR
- 2016
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Ingår i: Proceedings of PHYSOR 2016: Unifying Theory and Experiments in the 21st Century, Sun Valley, Idaho, USA, May 1-5, 2016; Paper No. 15947. - 9781510825734 ; 1, s. 140-151
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Konferensbidrag (refereegranskat)abstract
- The international Jules Horowitz material testing Reactor (JHR) is under construction at CEA Cadarache research center, in southern France. Its first criticality is foreseen by the end of the decade. In order to perform JHR design and safety studies, a specific neutronics calculation tool, HORUS3D/N, based on the deterministic codes APOLLO2 and CRONOS2 and on the European nuclear data library JEFF3.1.1, was developed to calculate JHR neutronics parameters taking into account fuel depletion: reactivity, power distribution, control rod reactivity worth, etc. Up to now, the biases and uncertainties on the different neutronics parameters computed with HORUS3D/N were assessed, in particular, by comparing HORUS3D/N deterministic calculations with reference route calculations based on APOLLO2-MOC and TRIPOLI-4®. The use for JHR of the recent Monte-Carlo TRIPOLI-4® in its new Depletion mode (TRIPOLI-4®D) will also allow providing biases for the main neutronics parameters under fuel depletion conditions. These biases will give a quantitative estimation of the impact of the approximations of the flux calculation in the deterministic route. This paper presents a contribution to the validation of HORUS3D/N based on the first comparisons between the calculations performed with APOLLO2-MOC and CRONOS2, and the ones from TRIPOLI-4®D. The study is performed on 2-D calculations for two different clusters in an infinite lattice configuration. It focuses on the main parameters of interest: isotopic concentrations, plate power distributions, reactivity, as functions of burnup. The results obtained show reasonable discrepancies with APOLLO2 calculation and allow to be confident on the APOLLO2.8/REL2005/CEA2005 package recommendations developed by CEA for light water reactor studies used in HORUS-3D/N. In particular, the main fuel isotopes are well predicted with TRIPOLI-4®D with discrepancies values lower than -1.5%.
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| 4. |
- Demaziere, Christophe, 1973, et al.
(författare)
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Noise-based core monitoring and diagnostics – Overview of the CORTEX project
- 2017
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Ingår i: Proc. 3rd DAE-BRNS Symposium on Advances in Reactor Physics (ARP-2017), Mumbai, India, December 6-9, 2017.
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Konferensbidrag (refereegranskat)abstract
- This paper gives an overview of the CORTEX project, which is a Research and Innovation Action funded by the European Union in the Euratom 2016-2017 work program, under the Horizon 2020 framework. CORTEX, which stands for CORe monitoring Techniques and EXperimental validation and demonstration, aims at developing an innovative core monitoring technique that allows detecting anomalies in nuclear reactors, such as excessive vibrations of core internals, flow blockage, coolant inlet perturbations, etc. The technique is based on primarily using the inherent fluctuations in neutron flux recorded by in-core and ex-core instrumentation (often referred to as neutron noise), from which the anomalies will be differentiated depending on their type, location and characteristics. In addition to be non-intrusive and not requiring any external perturbation of the system, the method allows the detection of operational problems at a very early stage. Proper actions could thus be taken by utilities before such problems have any adverse effect on plant safety and reliability.
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| 5. |
- Demaziere, Christophe, 1973, et al.
(författare)
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Overview of the CORTEX project
- 2018
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Ingår i: International Conference on Physics of Reactors, PHYSOR 2018: Reactor Physics Paving the Way Towards More Efficient Systems. ; Part F168384-5, s. 2971-2980
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Konferensbidrag (refereegranskat)abstract
- This paper gives an overview of the CORTEX project, which is a Research and Innovation Action funded by the European Union in the Euratom 2016-2017 work program, under the Horizon 2020 framework. CORTEX, which stands for CORe monitoring Techniques and Experimental validation and demonstration, aims at developing an innovative core monitoring technique that allows detecting anomalies in nuclear reactors, such as excessive vibrations of core internals, flow blockage, coolant inlet perturbations, etc. The technique is based on primarily using the inherent fluctuations in neutron flux recorded by in-core and ex-core instrumentation (often referred to as neutron noise), from which the anomalies will be differentiated depending on their type, location and characteristics. In addition to be non-intrusive and not requiring any external perturbation of the system, the method allows the detection of operational problems at a very early stage. Proper actions could thus be taken by utilities before such problems have any adverse effect on plant safety and reliability. In order to develop a method that can reach a high Technology Readiness Level, the consortium, made of 20 partners, was strategically structured around the required core expertise from all the necessary actors of the nuclear industry, both within Europe and outside. The broad expertise of the consortium members ensures the successful development of new in-situ monitoring techniques.
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| 6. |
- Ghione, Alberto, 1989, et al.
(författare)
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Assessment of criteria for Onset of Flow Instability in vertical narrow rectangular channels with downward flow
- 2018
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Ingår i: 12th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, Operation and Safety (NUTHOS-12).
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Konferensbidrag (refereegranskat)abstract
- This paper presents an assessment study of criteria for the prediction of the Onset of Flow Instability (OFI) in heated vertical narrow rectangular channels with downward flow. The onset of flow instability is a limiting safety issue in nuclear research reactors, since it may lead to flow starvation and eventually boiling crisis in some of the core channels. The experimental database consists of OFI tests at low pressure, with gap sizes between 1.37 and 3.23 mm, and with uniform and non-uniform heat flux profiles. Two typologies of criteria were tested, following a global and local strategy, respectively. According to the global approach, the Whittle-Forgan and Stelling criteria only use global system parameters to predict OFI. Relatively good results can be obtained over the whole database. The local approach is based on the prediction of the Net Vapor Generation (NVG) along the channel. The standard Saha-Zuber correlation fails to capture OFI for Peclet numbers lower than 70000. On the other hand, the Saha-Zuber KIT correlation identifies OFI in all the experiments with uniform heat flux. In the case of a non-uniform heat flux along the width of the rectangular channel, the use of thermalhydraulic parameters averaged over the cross section is not sufficient to predict the local onset of NVG. Nevertheless, this issue can be solved if a local estimate of the flow conditions is employed, so that the effect of the non-uniform heat flux can be accounted for. These results are consistent with the ones found for upward flow in previous studies. No significant effect of the flow direction is therefore observed.
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| 7. |
- Ghione, Alberto, 1989, et al.
(författare)
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Assessment of Critical Heat Flux correlations in narrow rectangular channels
- 2016
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Ingår i: Proceedings of the 11th International Topical Meeting on Nuclear Thermal Hydraulics, Operation and Safety (NUTHOS-11), Gyeongju, Korea, Oct. 9-13 2016.
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Konferensbidrag (refereegranskat)abstract
- The aim of the work is to assess different CHF correlations when applied to vertical narrow rectangular channels with upward low-pressure water flow. This is a contribution to the improvement of the thermal-hydraulic modeling of the Jules Horowitz Reactor, which is a research reactor under construction at CEA-Cadarache (France). For this purpose, 46 CHF tests from the SULTAN-JHR experimental database were used. These experiments were performed at CEA-Grenoble in two vertical uniformly heated rectangular channels with gaps of 1.51 (SE3: 20 tests) and 2.16 mm (SE4: 26 tests). The experimental conditions ranged between 0.38 and 0.87 MPa for the outlet pressure, between 1200 and 6600 kg/m2/s for the mass flux, between 56.4 and 156.4 °C for the inlet liquid sub-cooling and between -0.01 and 0.12 for the outlet steam quality. Several models were tested. The Groeneveld look-up tables, which were developed mainly with experiments in pipes, significantly over-estimate the CHF. Furthermore, they fail to predict the decrease of the CHF with the reduction of the gap size. Doerffer’s modification of Groeneveld look-up table for internally heated annuli and the Sudo correlation for nuclear research reactors with plate-type fuel, give better results. In particular, Doerffer’s formula predicts the experimental data with a mean error of -10 % for SE4 and +17 % for SE3, while the Sudo relationship gives mean errors equal to -2.3 % and +32 %.
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| 8. |
- Ghione, Alberto, 1989, et al.
(författare)
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Assessment of thermal–hydraulic correlations for narrow rectangular channels with high heat flux and coolant velocity
- 2016
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Ingår i: International Journal of Heat and Mass Transfer. - : Elsevier BV. - 0017-9310. ; 99, s. 344-356
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Tidskriftsartikel (refereegranskat)abstract
- The focus of the paper is on the evaluation of the correlations for predicting single-phase friction, single- and two-phase forced convection heat transfer coefficients in rectangular narrow channels, where the wall heat flux and the coolant flow can reach relatively high values. For this purpose, several correlations are reviewed and assessed against the SULTAN-JHR experiments. These tests were performed at CEA-Grenoble with upward water flow in two vertical uniformly heated narrow rectangular channels with gap of 1.51 and 2.16 mm. The experimental conditions range between 0.2 and 0.9 MPa for the pressure; 0.5–18 m/s for the coolant velocity and between 0.5 and 7.5 MW/m2 for the heat flux. The use of an appropriate turbulent friction factor leads to good comparison with the experimental data. The analysis of the single-phase turbulent heat transfer coefficient shows that the standard correlations (e.g. Dittus–Boelter) significantly under-estimate the heat transfer coefficient, especially at high Reynolds number. Therefore, new best-fitting correlations are derived. It is also observed that a reduction in gap size may lead to the enhancement of the heat transfer. The heat transfer is also under-estimated in two-phase flow if standard correlations (e.g. Jens–Lottes) are employed; however, good comparison with the experimental data are obtained with more appropriate models for fully developed boiling, such as the Forster–Greif correlation. The global accuracy associated to these correlations is also quantified in a rigorous manner.
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| 9. |
- Ghione, Alberto, 1989, et al.
(författare)
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Criteria for onset of flow instability in heated vertical narrow rectangular channels at low pressure: an assessment study
- 2017
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Ingår i: International Journal of Heat and Mass Transfer. - : Elsevier BV. - 0017-9310. ; 105, s. 464-478
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Tidskriftsartikel (refereegranskat)abstract
- In this paper, an assessment of the criteria for the prediction of the flow excursion instabilities in vertical narrow rectangular channels is presented. The experimental database consists of 166 flow excursion points at low pressure with upward flow, and with uniform or non-uniform heat flux profiles. The test sections have gap sizes between 1.27 and 3.6 mm, hydraulic diameters between 2.37 and 6.58 mm, and length to heated diameter ratio between 70.9 and 196.8. A wide range of parameters is covered: the mass flux is between 740 and 20,325 kg/m2/s; the outlet pressure is between 0.12 and 1.73 MPa; the heat flux is between 0.4 and 14.9 MW/m2; the Peclet number is between 15,889 and 358,460; the outlet sub-cooling is between 4.8 and 35.1 °C. None of the tests reaches saturation at the exit of the test section. Several criteria for identifying the Onset of Flow Instability (OFI), were tested. Such criteria can rely on correlations for the Onset of Nucleate Boiling (ONB), the Net Vapor Generation (NVG), the onset of Fully Developed Boiling (FDB), or can relate global parameters of the system. All these models have good performances on average, with both uniform and non-uniform axial heat fluxes. The ONB-based relationships are largely conservative as expected since the ONB always precedes the OFI. The NVG criteria can provide relatively good results, but a crucial issue is related to the value of Peclet number at which the transition between the thermally and the hydro-dynamically driven bubble detachment takes place in narrow rectangular channels. In view of this, the standard Saha–Zuber correlation cannot predict OFI for Peclet numbers lower than 70,000; while the Saha–Zuber KIT correlation, whose transition Peclet number is smaller, identifies the OFI for all the experiments. The approach that makes use of a FDB correlation can capture the OFI in most of the cases, although its performance also depends on the type of correlation that is applied for the single-phase heat transfer. The Flow Instability Ratios (FIRs) like the ones developed by Whittle–Forgan or Stelling et al., are of particular interest because they only require global system parameters, and because they are shown to be a valid option for determining the flow excursion in the experiments included in this study. For instance, the Stelling FIR with the Saha–Zuber KIT correlation estimates OFI in all the tests. Finally, best-fitting procedures of the available data were also introduced in order to optimize such FIRs.
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| 10. |
- Ghione, Alberto, 1989, et al.
(författare)
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Uncertainty and sensitivity analysis for the simulation of a station blackout scenario in the Jules Horowitz Reactor
- 2017
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Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 104, s. 28-41
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Tidskriftsartikel (refereegranskat)abstract
- An uncertainty and sensitivity analysis for the simulation of a station blackout scenario in the Jules Horowitz Reactor (JHR) is presented. The JHR is a new material testing reactor under construction at CEA on the Cadarache site, France. The thermal-hydraulic system code CATHARE is applied to investigate the response of the reactor system to the scenario. The uncertainty and sensitivity study was based on a statistical methodology for code uncertainty propagation, and the 'Uncertainty and Sensitivity' platform URANIE was used. Accordingly, the input uncertainties relevant to the transient, were identified, quantified, and propagated to the code output. The results show that the safety criteria are not exceeded and sufficiently large safety margins exist. In addition, the most influential input uncertainties on the safety parameters were found by making use of a sensitivity analysis.
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