| 1. |
- al-Dbissi, Moad, 1994, et al.
(author)
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Conceptual design and initial evaluation of a neutron flux gradient detector
- 2022
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In: Nuclear Instruments and Methods in Physics Research, Section A: Accelerators, Spectrometers, Detectors and Associated Equipment. - : Elsevier BV. - 0168-9002. ; 1026
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Journal article (peer-reviewed)abstract
- Identification of the position of a localized neutron source, or that of local inhomogeneities in a multiplying or scattering medium (such as the presence of small, strong absorbers) is possible by measurement of the neutron flux in several spatial points, and applying an unfolding procedure. It was suggested earlier, and it was confirmed by both simulations and pilot measurements, that if, in addition to the usually measured scalar (angularly integrated) flux, the neutron current vector or its diffusion approximation (the flux gradient vector) is also considered, the efficiency and accuracy of the unfolding procedure is significantly enhanced. Therefore, in support of a recently started project, whose goal is to detect missing (replaced) fuel pins in a spent fuel assembly by non-intrusive methods, this idea is followed up. The development and use of a dedicated neutron detector for within-assembly measurements of the neutron scalar flux and its gradient are planned. The detector design is based on four small, fiber-mounted scintillation detector tips, arranged in a rectangular pattern. Such a detector is capable of measuring the two Cartesian components of the flux gradient vector in the horizontal plane. This paper presents an initial evaluation of the detector design, through Monte Carlo simulations in a hypothetical scenario.
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| 2. |
- al-Dbissi, Moad, 1994, et al.
(author)
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Identification of diversions in spent PWR fuel assemblies by PDET signatures using Artificial Neural Networks (ANNs)
- 2023
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In: Annals of Nuclear Energy. - 0306-4549 .- 1873-2100. ; 193
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Journal article (peer-reviewed)abstract
- Spent nuclear fuel represents the majority of materials placed under nuclear safeguards today and it requires to be inspected and verified regularly to promptly detect any illegal diversion. Research is ongoing both on the development of non-destructive assay instruments and methods for data analysis in order to enhance the verification accuracy and reduce the inspection time. In this paper, two models based on Artificial Neural Networks (ANNs) are studied to process measurements from the Partial Defect Tester (PDET) in spent fuel assemblies of Pressurized Water Reactors (PWRs), and thus to identify at different levels of detail whether nuclear fuel has been replaced with dummy pins or not. The first model provides an estimation of the percentage of replaced fuel pins within the inspected fuel assembly, while the second model determines the exact configuration of the replaced fuel pins. The two models are trained and tested using a dataset of Monte-Carlo simulated PDET responses for intact spent PWR fuel assemblies and a variety of hypothetical diversion scenarios. The first model classifies fuel assemblies according to the percentage of diverted fuel with a high accuracy (96.5%). The second model reconstructs the correct configuration for 57.5% of the fuel assemblies available in the dataset and still retrieves meaningful information of the diversion pattern in many of the misclassified cases.
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| 3. |
- al-Dbissi, Moad, 1994, et al.
(author)
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On the use of neutron flux gradient with ANNs for the detection of diverted spent nuclear fuel
- 2024
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In: Annals of Nuclear Energy. - 0306-4549 .- 1873-2100. ; 204
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Journal article (peer-reviewed)abstract
- One of the main tasks in nuclear safeguards is regular inspections of Spent Nuclear Fuel (SNF) assemblies to detect possible diversions of special nuclear material such as 235U and 239Pu. In these inspections, characteristic signatures of SNF such as emissions of neutrons and gamma rays from the radioactive decay, are measured and their consistency with the declared assemblies is verified to ensure that no fuel pins have been removed. Research in this field is focused on both the development of detection equipment and methods for the analysis of the acquired measurement data. In this paper, the use of the neutron flux gradient, which is not considered in regular SNF verification, is investigated in combination with the scalar neutron flux as input to artificial neural network models for the quantification of fuel pins in SNF assemblies. The training and testing of these ANN models rely on a synthetic dataset that is generated from Monte Carlo simulations of a typical intact pressurized water reactor assembly and with different patterns of fuel pins replaced by dummy pins. The dataset consists of unique scenarios so that the ANN can be assessed over “unknown” cases that are not part of the learning phase. Results show that the neutron flux gradient is advantageous for a more accurate reconstruction of diversions within SNF assemblies.
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| 4. |
- Calivá, Francesco, et al.
(author)
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A deep learning approach to anomaly detection in nuclear reactors
- 2018
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In: Proceedings of the International Joint Conference on Neural Networks. ; 2018-July
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Conference paper (peer-reviewed)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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| 5. |
- Chambon, Amalia, 1986, et al.
(author)
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A deterministic against Monte-Carlo depletion calculation benchmark for JHR core configurations
- 2017
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In: Int. Conf. Mathematics & Computational Methods Applied to Nuclear Science & Engineering (M&C 2017), Jeju, Korea, April 16-20, 2017.
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Conference paper (peer-reviewed)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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| 6. |
- Chambon, Amalia, 1986, et al.
(author)
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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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In: 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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Conference paper (peer-reviewed)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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| 7. |
- Demaziere, Christophe, 1973, et al.
(author)
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Comparison of the U.S. NRC PARCS Core Neutronics Simulator Against In-Core Detector Measurements for LWR Applications
- 2012
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Reports (other academic/artistic)abstract
- The safety analysis of a Nuclear Power Plant (NPP) is based on the application of complex computer codes that are able to simulate the physical behaviour of the system under normal operations and abnormal conditions. Therefore, such codes must be extensively and continuously verified and validated in order to demonstrate their reliability.In this context, the current report presents an assessment study for the U.S. NRC 3-D neutronic core simulator PARCS, and it includes an evaluation of the performances of the code for LWRs applications. For this purpose, the cores of the Swedish Ringhals-3 Pressurized Water Reactor (PWR) unit and the Forsmark-2 Boiling Water Reactor (BWR) unit were modeled with PARCS. As regards the cross-sections needed for this kind of calculations, they were prepared by following a special procedure developed by the present authors since core material data were only available in the format of library and restart files created by the SIMULATE-3 neutronic core simulator. Correspondingly, a new cross-section interface was developed and verified by the Division of Nuclear Engineering, Chalmers University of Technology, in order to convert the SIMULATE-3 data into data suitable for PARCS. Thereafter, the PARCS models developed for Ringhals-3 and Forsmark-2 were used for neutronic core analyses, at different operating conditions, along several fuel cycles. The results achieved from these simulations were then compared against the axial power and the radial power distribution estimated from the measurements that were provided by the owners of the plants.In the PWR case, the PARCS simulations predict satisfactorily both the core axial power profile and the core radial power distribution, although, in some cases, the deviations between calculated and measured data exhibit trends that need further investigations. For instance, the PARCS simulation at the beginning of a fuel cycle seems to overestimate the power in the center of the core and to underestimate the power at the periphery, whereas, at the end of a fuel cycle, the situation is opposite.In the BWR case, the core axial profile was predicted in a reasonable manner, but quite significant discrepancies for the radial power distribution was found. The current work suggests that such a disagreement might be due to the inability of PARCS to properly model multiple composition control rods. In fact the largest deviations in the computed power from the measurements were observed for those fuel assemblies placed in the neighborhood of control rods.
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| 8. |
- Demaziere, Christophe, 1973, et al.
(author)
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Development and test of a novel neutronic verification scheme for Molten Salt Reactors
- 2021
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In: Transactions of the American Nuclear Society. - 0003-018X. ; 124:1, s. 504-507
-
Conference paper (peer-reviewed)abstract
- This paper presents the extension of a verification method of transient neutron transport solvers earlier developed to the case of Molten Salt Reactor (MSR). This method is based on the extraction of the point-kinetic response of a nuclear reactor excited by a mono-chromatic perturbation and on its subsequent comparison with a closed analytical form. Whereas a closed analytical form exists for systems with fixed fuel, no closed analytical form exists in the case of MSR, as highlighted by many authors. A workaround is nevertheless proposed in this work, thus giving the possibility to use a similar verification method to the case of MSR.
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| 9. |
- Demaziere, Christophe, 1973, et al.
(author)
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Development and test of a novel verification scheme applied to the neutronic modelling of Molten Salt Reactors
- 2022
-
In: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 167
-
Journal article (peer-reviewed)abstract
- This paper presents the extension of a method to verify transient neutron transport solvers earlier developed for reactors with non-moving fuel, to the case of Molten Salt Reactors (MSRs). This method is based on the extraction of the point-kinetic response of a nuclear reactor excited by a mono-chromatic perturbation and on its subsequent comparison with its expected functional dependence. Whereas a simple expression for this dependence exists for systems with fixed fuel, this is not the case for MSRs, as highlighted in many past studies. A workaround is nevertheless proposed in this work, thus giving the possibility to use a similar verification method to the case of MSRs. The method is applied to a simple dynamic MSR solver, demonstrating the capabilities of the technique. Contrary to other verification methods for which the system has to be simplified so that analytical solutions can be derived, the present method can be applied to any heterogeneous system.
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| 10. |
- Demaziere, Christophe, 1973, et al.
(author)
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Monte Carlo-based dynamic calculations of stationary perturbations
- 2020
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In: International Conference on Physics of Reactors: Transition to a Scalable Nuclear Future, PHYSOR 2020. - : EDP Sciences.
-
Conference paper (peer-reviewed)abstract
- Capitalizing on some earlier work, this paper presents a novel Monte Carlo-based approach that allows estimating the neutron noise induced by stationary perturbations of macroscopic cross-sections in the frequency domain. This method relies on the prior computation using Monte Carlo of modified Green’s functions associated to the real part of the dynamic macroscopic cross-sections, mimicking equivalent subcritical problems driven by external neutron sources. Once such modified Green’s functions are estimated, the neutron noise induced by any type of perturbations can be recovered, by solving a linear algebra problem accounting for the interdependence between the real and imaginary parts of the governing balance equations. The newly derived method was demonstrated on a large homogeneous test system and on a small heterogeneous test system to provide results comparable to a diffusion-based solver specifically developed for neutron noise applications. The new method requires the specification by the user of the real part of the Fourier transform of the macroscopic cross-sections. This is accomplished using ACE-formatted cross-section files defined by the user. Beyond this input data preparation, no change to the Monte Carlo source code is necessary. This represents the main advantage of the proposed method as compared to similar efforts requiring extensive modifications to the Monte Carlo source code.
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