Modelling and simulations of reactor neutron noise induced by mechanical vibrations

Mechanical vibrations of core internals are among the main perturbations that induce oscillations in the neutron flux field, also known as neutron noise. In this work, different simulation models for the study of the influence of the mechanical vibrations of fuel assemblies on the neutron flux in the reactor core have been discussed. These methodologies employ the diffusion approximation, with or without a previous homogenization model, to simulate the neutron noise in the time or the frequency domain. The diffusion-based approach is expected to be less accurate in the vicinity of the vibrating fuel assemblies, but correct when considering distances larger than a few diffusion lengths away from the perturbation. All methodologies provide consistent results and can reproduce typical features of the neutron noise induced by mechanical vibrations of core components. First, FEMFFUSION can perform simulations in both the time and frequency domains. Second, CORE SIM + can be used to study various neutron noise scenarios in realistic three-dimensional reactor configurations. The third methodology is centred on using commercial codes as CASMO-5, SIMULATE-3 and SIMULATE-3K. This methodology allows time domain simulations of the neutron noise induced by different neutron noise sources in a nuclear reactor. Finally, a model for time-dependent geometry is implemented for the code system ATHLET/QUABOX-CUBBOX employing a cross-section-based approach for encoding water gap width variations at the reflector

Application of causality analysis on nuclear reactor systems

Causality analysis is a substantial tool for identifying cause-and-effect links between different components of a system and has been extensively used in various areas of science such as neuroscience, climatology, and econometrics. This analysis is carried out in terms of the renormalized partial directed coherence and the directed transfer function connectivity measures. Applying such analysis in the nuclear reactor field is of paramount importance since it can help in inferring cause-and-effect relationships between highly coupled processes, and consequently, it can assist on the safe and reliable operation of a nuclear power plant during the occurrence of possible disturbances or malfunctions. The effectiveness of the connectivity analysis is demonstrated through several simulated and measured test cases. Results show that the connectivity analysis is able to identify accurately the importance and central role of the activation signal when it is applied on a simple analytical model and a simulated nuclear reactor system. In addition, the application on more realistic and complex measured data sets of a Swiss boiling water reactor illustrates the capability of this analysis to indicate possible causes behind the observed anomalies or trends observed at certain conditions and, more importantly, allows a better understanding of the underlying interactions among different neutronic and thermal-hydraulic processes. Published under license by AIP Publishing

Nuclear data uncertainties for Swiss BWR spent nuclear fuel characteristics

The effect of nuclear data (fission yields, cross sections and emitted spectra) is quantified for spent nuclear fuel assemblies from a realistic boiling water reactor operated over 25 cycles. Nominal calculations are performed with the CASMO5, SIMULATE-3 and SNF codes and the ENDF/B-VII.0 nuclear data library. The uncertainties are calculated with the same codes, using a Monte Carlo propagation method, and the ENDF/B-VII.1 covariance matrices. The conclusions are that (1) the nuclear data have a non-negligible impact for spent fuel quantities (e.g., decay heat, neutron emission or isotopic contents); (2) the importance of varying all data together is demonstrated, showing an under- or overestimation of uncertainties if fission yields are sampled separately from the other nuclear data; and finally (3) the importance of considering the full irradiation history (multi-cycle assembly life) is also demonstrated, showing also an under- or overestimation of uncertainties when performing the nuclear data sampling for a single reactor cycle

MODELLING AND ANALYSIS OF FUEL ASSEMBLY VIBRATIONAL MODES IN PWRS USING SIMULATE-3K

Some of the KWU pre-KONVOI PWRs operating across Europe saw a systematic increase in the neutron noise levels over several cycles in the last decade, and subsequently, core internals’ movements, especially vibrations of fuel assemblies with specific designs were identified as one of the plausible causes. Therefore, it is important to develop computational methods that can allow to investigate and predict the reactor noise response to fuel assemblies vibrations. To this aim, the 3D nodal reactor dynamics code SIMULATE-3K is used at PSI with a special module called the ‘assembly vibration model’ that imitates time-dependent motions of fuel assemblies by dynamically modifying the water-gaps surrounding the laterally moving fuel assemblies. The varying water-gaps are represented by the variation in the corresponding two-group macroscopic cross sections generated using the lattice code CASMO-5 in 2D. The studies conducted so far to assess the methodology for full core noise simulations were based on assuming vibrations of a clamped-free cluster of fuel assemblies that are unsupported from both ends. However, as this represents a non-physical movement, further developments were made at PSI to allow simulating more realistic movements of fuel assemblies such as the cantilevered mode vibration. The updated methodology, along with evaluations of the simulated noise response to realistic vibration modes, is presented in this paper. Results show that, as expected, the radial and axial neutron noise behaviour follow the vibration pattern of the imposed time-dependent axial functions corresponding to the natural oscillation modes of the fuel assemblies, thereby providing confidence in the application of the developed methodology for numerical neutron noise analyses of the PWR cores

BWR stability and bifurcation analysis using reduced order models and system codes: Identification of a subcritical Hopf bifurcation using RAMONA

The system code RAMONA, as well as a recently developed BWR reduced order model (ROM), are employed for the stability analysis of a specific operational point of the Leibstadt nuclear power plant. This has been done in order to assess the ROM's applicability and limitations in a quantitative manner. In the context of a detailed local bifurcation analysis carried out using RAMONA in the neighbourhood of the chosen Leibstadt operA ational point, a bridge is built between the ROM and the system code, This has been achieved through interpreting RAMONA solutions on the basis of the physical mechanisms identified in the course of applying the ROM. This leads, for the first time, to the identification of a subcritical Poincare-Andronov-Hopf (PAH) bifurcation using a system code. As a consequence, the possibility of the so-called correspondence hypothesis is suggested to underline the relationship between a stable (unstable) limit cycle solution and the occurrence of a supercritical (subcritical) PAH bifurcation in the modeling of boiling water reactor stability behaviour. (C) 2007 Elsevier Ltd. All rights reserved

Development and verification of a methodology for neutron noise response to fuel assembly vibrations

Nuclear reactors are stochastic systems, in which several parameters fluctuate even during steady-state conditions. In addition, structural components like fuel assemblies, vibrate due to strong hydraulic forces. This stochasticity is the cause of the neutron population fluctuating behavior; a phenomenon referred as neutron noise. This paper systematically analyzes the capabilities of the Studsvik's codes to model the fuel assembly vibration and to assess the neutron noise response. To this aim, the CASMO-5 code is utilized for generating cross sections by parametrizing the water-gap thickness of fuel segments. The fuel assembly vibration modelling is approximated by the dynamic modification of the water-gap widths between adjacent assemblies using the transient code SIMULATE-3K. The cross sections dependency on the water-gap width is analyzed and successfully compared against Serpent-2 reference results. Last, the utilized codes capability is verified qualitatively or/and quantitatively through a series of cases, at both lattice and nodal level. (C) 2020 The Author(s). Published by Elsevier Ltd

Electrochemical synthesis of ZnO thin film and nanowire for piezoelectric applications

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Uncertainties for Swiss LWR spent nuclear fuels due to nuclear data

This paper presents a study of the impact of the nuclear data (cross sections, neutron emission and spectra) on different quantities for spent nuclear fuels (SNF) from Swiss power plants: activities, decay heat, neutron and gamma sources and isotopic vectors. Realistic irradiation histories are considered using validated core follow-up models based on CASMO and SIMULATE. Two Pressurized and one Boiling Water Reactors (PWR and BWR) are considered over a large number of operated cycles. All the assemblies at the end of the cycles are studied, being reloaded or finally discharged, allowing spanning over a large range of exposure (from 4 to 60 MWd/kgU for ≃9200 assembly-cycles). Both UO2 and MOX fuels were used during the reactor cycles, with enrichments from 1.9 to 4.7% for the UO2 and 2.2 to 5.8% Pu for the MOX. The SNF characteristics presented in this paper are calculated with the SNF code. The calculated uncertainties, based on the ENDF/B-VII.1 library are obtained using a simple Monte Carlo sampling method. It is demonstrated that the impact of nuclear data is relatively important (e.g. up to 17% for the decay heat), showing the necessity to consider them for safety analysis of the SNF handling and disposal

NEUTRON NOISE SPECTRAL FEATURES OF SIMULATED MECHANICAL AND THERMAL-HYDRAULIC PERTURBATIONS IN A PWR CORE

KWU-PWR reactors (SIEMENS design) are commonly exhibiting high neutron noise levels that can lead to costly operational issues. Recent analysis seems to indicate that, coolant flow, temperature oscillations, and mechanical vibrations have a key impact on neutron noise phenomena. In order to advance in understanding this phenomenon, the transient nodal code SIMULATE-3K (S3K) has been already utilized to simulate scenarios with individual or combined types of perturbation sources: mechanical vibrations of fuel assemblies and thermal-hydraulic fluctuations at the core inlet. In this work, new simulations are performed with all the perturbations applied simultaneously. The simulated neutron detectors responses are then analyzed with noise analysis techniques. All the simulated spectral features of neutron noise are compared to those obtained from real plant data. Results show that the simulated neutron noise phenomenology behaves similarly to that obtained from real plant data by increasing the fluctuation amplitude in the inlet coolant flow in the S3K calculations