Asymmetric Transients Research Articles

Clarke Transformation Solution of Asymmetrical Transients in Three-Phase Circuits

This work deals with the use of the Clarke transformation for the theoretical derivation of circuit models for the analysis of asymmetrical transients in three-phase circuits. Asymmetrical transients occur when only one or two phases of a three-phase power system are involved in a switch operation. Such a condition is critical from a theoretical viewpoint since the Clarke transformation is based on the assumption of circuit symmetry between the three phases. If the symmetry assumption is not met, the equivalent circuits in the transformed variables α, β, and 0 are not uncoupled. The literature concerning numerical approaches for asymmetrical transient analysis is very rich, but a comprehensive and rigorous analytical investigation of circuit models within the framework of the Clarke transformation is still lacking. Contrary to numerical approaches, analytical solutions provide deeper insight into the phenomenon and allow for theoretical interpretation and better understanding of the transient behavior. The proposed circuit models show that the β variables are always uncoupled with α and 0 variables, whereas coupling between α and 0 variables can be properly represented by an ideal transformer. Moreover, in the case of single-line switching, the β variables have no transient, i.e., they keep the steady-state behavior. Transient properties can be readily and effectively observed by representing the trajectory of the space vector on the complex plane. All the analytical results have been checked numerically through the Simulink (Matlab R2020a, The MathWorks, Inc., Natick, MA, USA) implementation of a specific three-phase circuit introduced to illustrate the main theoretical issues.

Spatially asymmetric transients of propagating exciton-polariton modes in a planar CdZnTe/CdMgTe guiding structure

We report on ultrafast time-resolved pump-probe studies in a CdZnTe/CdMgTe planar guiding structure covered with a metallic grating. The one-dimensional periodic gold structure allows for efficient coupling into the guiding layer for $p$-polarized 30 fs optical pulses with a large spectral bandwidth of about 60 nm. The resulting spectral width of optical pulses propagating inside the guiding layer corresponds to 20--30 nm. We demonstrate that the excitation of exciton-polariton modes in the guiding layer leads to a modulation of the optical response in the vicinity of the excitonic resonance. Spatially resolved pump-probe measurements show an asymmetric behavior in the optical response when the relative position of the pump and probe spots is varied on the scale of ten micrometers perpendicular to the metal ridges. This is attributed to the excitation of resonant and off-resonant exciton-polariton modes which propagate in opposite directions inside the guiding layer in accordance with their dispersion relations. Two main mechanisms are considered and evaluated, namely, Pauli blocking and excitation-induced dephasing, which are shown to be responsible for the pump-induced changes in the exciton absorption spectrum. While both of these processes lead to the generation of photoexcited carriers in the guiding layer, their impact on the optical properties (transmission and reflection) are different which leads to the asymmetric behavior of the spatially resolved transients.

Multi-physics models for design basis accident analysis of sodium fast reactors. Part I: Validation of three-dimensional TRACE thermal-hydraulics model using Phenix end-of-life experiments

The demonstrated technological feasibility of Sodium-cooled Fast Reactors (SFRs) makes them stand out among the other reactor concepts proposed by Generation IV International Forum (GIF) for short-term deployment. The availability of reliable computational tools in support of safety analyses and plant simulations under complex transient scenarios is essential to assure SFR’s compliance with the highest safety goals.Answering this need, a multi-physics three-dimensional core and system model is being developed to enable a more detailed representation of the physics of the plant and to anticipate more accurately plant behaviour, even under wider three-dimensional scenarios, such as asymmetric transients. The coupling will be performed using the U.S.NRC system codes TRACE-PARCS, modified to simulate more accurately when using sodium as coolant.The publicly available end-of-life tests conducted in the French SFR Phenix were chosen as baseline to perform a first validation of the computational model. The development of the tool started with a three-dimensional thermal-hydraulic nodal system of Phenix using the TRACE system code.The system simulates the Phenix end-of-life natural circulation test and the result have been compared with published experimental and benchmark results. The main physical phenomena of the 3 phases of the transient (rise in temperature in the low part of the reactor vessel, establishment of natural convection and subsequent cooling of the lower and upper part of the vessel) are predicted by the developed nodal system. More specifically, the analysis of parameters such as Intermediate Heat Exchangers (IHX), primary pumps and core temperatures, shows that the developed system is able to predict and study natural convection phenomena in Phenix-type reactors.The three-dimensional nodal system is able to clearly illustrate the existing thermal stratification in the hot pool, which is neglected by one-dimensional systems and enables the modelling of thermal hydraulic asymmetric behaviour, as it is shown by the uneven flow distribution in Phenix’s primary IHXs as they are asymmetrically located in the reactor vessel.

Multiscale neutronics/thermal-hydraulics coupling with COBAYA4 code for pin-by-pin PWR transient analysis

At UPM, in-depth modifications of the core simulator COBAYA, able to perform neutronics diffusion calculations at both nodal and pin-by-pin levels, have been accomplished during the 7th Framework EURATOM NURESAFE project. The main goal was to upgrade its integration in the European Platform for Nuclear Reactor Safety Simulation in order to facilitate the coupling with any other code of the platform for multi-physics analysis, focusing also on the code legibility and maintainability.An external and flexible coupling with the thermal-hydraulics code COBRA-TF was designed. As a result, COBAYA4/COBRA-TF allows multiscale coupled calculations, enabling both nodal and pin-by-pin neutronics resolutions using both assembly-based channels and pin-based subchannels at the thermal-hydraulics domain. Flexible mapping schemes also in axial direction can be defined.The coupled system was applied to a Main Steam Line Break transient benchmark. Pin-by-pin 3D simulations using one thermal-hydraulic channel per assembly were carried out in a reasonable computing time, and results compared to nodal solutions demonstrating the multiscale coupling capability for full core transients. Pin-by-pin calculations using thermal-hydraulics subchannels will be performed in a near future to assess the role that a very detailed mapping can play to predict realistic local parameters. While in asymmetric transients the effect can be important, it is expected that in symmetric transients assembly-based thermal-hydraulics channels can provide accurate pin-by-pin solutions in execution times suitable for routine analysis.The performed work will bring the ability to explore in an easy way multiscale effects on safety transient evaluations and give recommendations for the neutronics/thermal-hydraulics mapping depending on the application.

Development, validation and application of multi-point kinetics model in RELAP5 for analysis of asymmetric nuclear transients

Point kinetics approach in system code RELAP5 limits its use for many of the reactivity induced transients, which involve asymmetric core behaviour. Development of fully coupled 3D core kinetics code with system thermal-hydraulics is the ultimate requirement in this regard; however coupling and validation of 3D kinetics module with system code is cumbersome and it also requires access to source code. An intermediate approach with multi-point kinetics is appropriate and relatively easy to implement for analysis of several asymmetric transients for large cores. Multi-point kinetics formulation is based on dividing the entire core into several regions and solving ODEs describing kinetics in each region. These regions are interconnected by spatial coupling coefficients which are estimated from diffusion theory approximation. This model offers an advantage that associated ordinary differential equations (ODEs) governing multi-point kinetics formulation can be solved using numerical methods to the desired level of accuracy and thus allows formulation based on user defined control variables, i.e., without disturbing the source code and hence also avoiding associated coupling issues. Euler's method has been used in the present formulation to solve several coupled ODEs internally at each time step. The results have been verified against inbuilt point-kinetics models of RELAP5 and validated against 3D kinetics code TRIKIN. The model was used to identify the critical break in RIH of a typical large PHWR core. The neutronic asymmetry produced in the core due to the system induced transient was effectively handled by the multi-point kinetics model overcoming the limitation of in-built point kinetics model of RELAP5 and standalone 3D core kinetics codes.

Анализ динамических режимов работы трехфазных трехстержневых трансформаторов в пакете MATLAB

Background: Dynamic modes of various nonlinear electromagnetic devices regarding instantaneous currents and voltages are now calculated by numerical methods. That is why the software package MATLAB, a high-level programming environment, has become widely spread. However, its built-in features are not always sufficient to describe circuits containing non-linear electromagnetic devices, in particular three-core transformers. It becomes clear while studying asymmetrical transients which are caused by changing parameters of the transformer itself, such as breakage or short circuit of windings, or by using a connecting load. Asymmetric transients of transformers are normally analysed by the method of superposition of forced and free modes. And the forced mode is calculated by the method of symmetrical components, while the available free components are determined by introducing different coefficients identified by comparing the accumulated experimental and calculated data. This approach is based on rather rough assumptions. The goal of the present research is to develop methods, algorithms and programs which allow us to use Simulink for calculating nonlinear electromagnetic devices. Materials and methods: The study of dynamical modes in the transformer was performed by the alternating direction method in MATLAB environment that has a program for calculation of nonlinear magnetic circuit of the three-core transformer.  «Вестник ИГЭУ» Вып. 4 2016 г.  ФГБОУВО «Ивановский государственный энергетический университет имени В.И. Ленина» 12 Results: A model and algorithms have been developed extending the capabilities of Simulink for the calculation of circuits with non-linear electromagnetic devices. The asymmetrical modes for the three-core transformer with a Y/Yn winding connection scheme have been calculated. Conclusions: The proposed model is universal and allows analysing transient, steady-state and emergency modes under asymmetric loads in transformers with different connection diagrams. The evaluation was made by verifying the results of numerical experiments against the passport data of the transformer.

Control rod drop transient analysis with the coupled parallel code pCTF-PARCSv2.7

In order to reduce the response time when simulating large reactors in detail, a parallel version of the thermal–hydraulic subchannel code COBRA-TF (CTF) has been developed using the standard Message Passing Interface (MPI). The parallelization is oriented to reactor cells, so it is best suited for models consisting of many cells. The generation of the Jacobian matrix is parallelized, in such a way that each processor is in charge of generating the data associated with a subset of cells. Also, the solution of the linear system of equations is done in parallel, using the PETSc toolkit. With the goal of creating a powerful tool to simulate the reactor core behavior during asymmetrical transients, the 3D neutron diffusion code PARCSv2.7 (PARCS) has been coupled with the parallel version of CTF (pCTF) using the Parallel Virtual Machine (PVM) technology. In order to validate the correctness of the parallel coupled code, a control rod drop transient has been simulated comparing the results with the real experimental measures acquired during an NPP real test.

Contrasting H-mode behaviour with deuterium fuelling and nitrogen seeding in the all-carbon and metallic versions of JET

The former all-carbon wall on JET has been replaced with beryllium in the main torus and tungsten in the divertor to mimic the surface materials envisaged for ITER. Comparisons are presented between type I H-mode characteristics in each design by examining respective scans over deuterium fuelling and impurity seeding, required to ameliorate exhaust loads both in JET at full capability and in ITER. Attention is focused upon a common high-triangularity, single-null divertor configuration at 2.5 MA, q95 ≈ 3.5 yielding the most robust all-C performance. Contrasting results between the alternative linings are found firstly in unseeded plasmas, for which purity is improved and intrinsic radiation reduced in the ITER-like wall (ILW) but normalized energy confinement is ≈30% lower than in all-C counterparts, owing to a commensurately lower (electron) pedestal temperature. Divertor recycling is also radically altered, with slower, inboard–outboard asymmetric transients at ELMs and spontaneous oscillations in between them. Secondly, nitrogen seeding elicits opposite responses in the ILW to all-C experience, tending to raise plasma density, reduce ELM frequency, and above all to recover (electron) pedestal pressure, hence global confinement, almost back to previous levels. A hitherto unrecognized role of light impurities in pedestal stability and dynamics is consequently suggested. Thirdly, while heat loads on the divertor outboard target between ELMs are successfully reduced in proportion to the radiative cooling and ELM frequency effects of N in both wall environments, more surprisingly, average power ejected by ELMs also declines in the same proportion for the ILW. Detachment between transients is simultaneously promoted. Finally, inter-ELM W sources in the ILW divertor tend to fall with N input, although core accumulation possibly due to increased particle confinement still leads to significantly less steady conditions than in all-C plasmas. This limitation of ILW H-modes so far will be readdressed in future campaigns to continue progress towards a fully integrated scenario suitable for D–T experiments on JET and for ‘baseline’ operation on ITER. The diverse changes in behaviour between all-C and ILW contexts demonstrate essentially the strong impact which boundary conditions and intrinsic impurities can have on tokamak-plasma states.

CATHARE Assessment of Natural Circulation in the PKL Test Facility during Asymmetric Cooldown Transients

Results of the CATHARE code calculations related to asymmetric cooldown tests in the PKL facility are presented. The test under consideration is the G2.1 experiment performed within the OECD/NEA PKL-2 project. It consists of carrying out a cooldown under natural circulation conditions in presence of two (out of four) emptied Steam Generators (SGs) and isolated on their secondary sides. The main goal of the current study is to assess the impact of a chosen cooldown strategy upon the occurrence of a Natural Circulation Interruption (NCI) in the inactive (i.e., noncooling) loops. For this purpose, three G2.1 test runs were investigated. The calculation results emphasize, mainly, the effect of the cooldown strategy, and the conditions that could lead to the occurrence of the NCI phenomenon.

Experimental investigation of in-vessel mixing phenomena in a VVER-1000 scaled test facility during unsteady asymmetric transients

In-vessel turbulent mixing phenomena affect the time and space distribution of coolant properties (e.g., boron concentration and temperature) at the core inlet which impacts consequently the neutron kinetics response. For reactor safety evaluation purposes and to characterize these phenomena it is necessary to set and validate appropriate numerical modelling tools to improve the current conservative predictions. With such purpose, an experimental campaign was carried out by OKB Gidropress, in the framework of the European Commission Project “TACIS R2.02/02 – Development of safety analysis capabilities for VVER-1000 transients involving spatial variations of coolant properties (temperature or boron concentration) at core inlet”. The experiments were conducted on a scaled facility representing the primary system of a VVER-1000 including a detailed model of the Reactor Pressure Vessel with its internals. The simulated transients involved perturbations of coolant properties distribution providing a wide validation matrix. The main achievements of the set of experiments featuring transient asymmetric pump behaviour are presented in this paper. The potential of the obtained experimental database for the validation of thermal fluid dynamics numerical simulation tools is also discussed and the role of computational fluid dynamics in supporting the experimental data analysis is highlighted.

Oxidation of carbon monoxide on poly-oriented and single-crystalline platinum electrodes over a wide range of pH

The oxidative stripping of pre-adsorbed carbon monoxide has been studied on poly-oriented platinum, and on Pt(111), Pt(100) and Pt(110) single-crystal electrodes in phosphate buffer solutions as a function of pH, both stripping voltammetry and chronoamperometry. It was found that the stripping peak potential has a tendency to decrease as a function of pH until a pH of ca. 10–11, which is ascribed to a weaker adsorption of phosphate on platinum with increasing pH. Above a pH of ca. 11, the stripping peak appears to stay constant or increase, depending on the surface structure. We hypothesize that this may be due to the fact that above a pH of ca. 10–11, the main product of carbon monoxide oxidation is carbonate, which may be block active sites on the surface. By comparison with the stripping on the single-crystalline Pt, the stripping voltammetry on the poly-oriented Pt electrode appears as a convolution of the oxidation on the different facets. A similar conclusion is drawn for the chronoamperometry experiments, and it is suggested that this may be the main reason for the asymmetrical transients observed on the poly-oriented Pt electrodes, as well as on other strongly heterogeneous Pt electrodes that have been studied in the literature.

Assessment of Natural Circulation Interruption during Asymmetric Cooldown Transients

This paper presents assessment results for the natural circulation interruption (NCI) phenomenon during the cooldown phase in a nuclear pressurized water reactor. This phenomenon could take place because of several circumstances, such as an asymmetric cooldown after the loss of the forced primary flow. Under NCI conditions, the homogeneous boration of the reactor coolant system (RCS) and the connection of the RCS to the residual heat removal system could be hindered. Moreover, at very low or no primary flow rates and an operating safety injection system, a pressurized thermal shock could occur in the reactor vessel due to cold fluid stratification in the loops. It is therefore important to understand the cause of loop flow stagnation and to derive accordingly the appropriate operator actions to avoid such a phenomenon.The main goal of the current study is to assess the effect of a cooldown strategy upon the single-phase NCI occurrence. For this purpose, two scenarios with asymmetric cooling between the reactor cooling loops were investigated: The first one concerns a feedwater line break combined with a loss of offsite power (LOOP), while the second one is limited to the LOOP (or any other transient leading to the loss of the forced primary flow). The analyses were carried out using the best-estimate thermal-hydraulic system code CATHARE 2/V2.5_1mod8.1, developed by Commissariat à l’Energie Atomique, Electricité de France, AREVA, and Institut de Radioprotection et de Sûreté Nucléaire. The calculation results mainly emphasize the effect of the cooldown rate and the opening strategy of the main steam atmospheric discharge valve upon the occurrence of the NCI phenomenon.

ICONE15-10327 VVER SAFETY ANALYSIS BY THE COUPLED CODE ATHLET/BIPR-VVER

The paper describes the modelling features for a VVER-1000 nuclear power plant by the coupled system code ATHLET/BIPR-VVER. Special attention is paid on the reactor pressure vessel model and its nodalization schema. They play an important role for the realistic prediction of the fluid mixing phenomena in case of asymmetric transients. For such a transient - simultaneous isolation of one steam generator from the steam line and the feed water supply at low power level - validation to measured parameters are presented and analysed. The results of the simulation are compared with experimental local and integral plant loop data from NPP Kozloduy Unit 6 collected during the plant-commissioning phase. Additional examples of coupled code application for VVER-1000 reactor applying detailed reactor pressure vessel nodalization are also presented, choosing Exercise 3 of the CEA-NEA/OECD VVER-1000 Coolant Transient Benchmark (V1000CT-2) - a main steam line break.

05/00872 Computer codes for spectrum average cross section calculations

At UPM, in-depth modifications of the core simulator COBAYA, able to perform neutronics diffusion calculations at both nodal and pin-by-pin levels, have been accomplished during the 7th Framework EURATOM NURESAFE project. The main goal was to upgrade its integration in the European Platform for Nuclear Reactor Safety Simulation in order to facilitate the coupling with any other code of the platform for multi-physics analysis, focusing also on the code legibility and maintainability.An external and flexible coupling with the thermal-hydraulics code COBRA-TF was designed. As a result, COBAYA4/COBRA-TF allows multiscale coupled calculations, enabling both nodal and pin-by-pin neutronics resolutions using both assembly-based channels and pin-based subchannels at the thermal-hydraulics domain. Flexible mapping schemes also in axial direction can be defined.The coupled system was applied to a Main Steam Line Break transient benchmark. Pin-by-pin 3D simulations using one thermal-hydraulic channel per assembly were carried out in a reasonable computing time, and results compared to nodal solutions demonstrating the multiscale coupling capability for full core transients. Pin-by-pin calculations using thermal-hydraulics subchannels will be performed in a near future to assess the role that a very detailed mapping can play to predict realistic local parameters. While in asymmetric transients the effect can be important, it is expected that in symmetric transients assembly-based thermal-hydraulics channels can provide accurate pin-by-pin solutions in execution times suitable for routine analysis.The performed work will bring the ability to explore in an easy way multiscale effects on safety transient evaluations and give recommendations for the neutronics/thermal-hydraulics mapping depending on the application.

Validation of coupled codes using VVER plant measurements

A data set of five transients at different VVER type nuclear power plants was collected in order to validate neutron kinetics/thermal hydraulics codes. Two of these transients ‘drop of control rod at nominal power at Bohunice-3’ of VVER-440 type and ‘coast-down of 1 from 3 working MCPs at Kozloduy-6’ of VVER-1000 type, were then utilised for code validation. Eight institutes contributed to the validation with 10 calculations using 5 different combinations of coupled codes. The thermal hydraulic codes were ATHLET, SMABRE and RELAP5 and the neutron kinetic codes DYN3D, HEXTRAN, KIKO3D and BIPR8. The general behaviour of both the transients was quite well calculated with all the codes. Even an elementary modelling of coolant mixing in reactor pressure vessel under asymmetric transients improved correspondence to the measurements. Some differences between the calculations seem to indicate that fuel modelling and treatment of VVER-440 control rods need further consideration. The simultaneous validation interacted with the data collection effort and thus improved its quality. The complexity of data collection systems and sometimes conflicting data, however, called for compromises and interpretation guides that also taught the analysts balanced plant modelling.

Investigation of spatial coupling aspects for coupled code application in PWR safety analysis

The simulation of nuclear power plant accident conditions requires three-dimensional (3-D) modeling of the reactor core to ensure a realistic description of physical phenomena. This paper describes a part of the research activities carried out on the sensitivity of coupled neutronics/thermal-hydraulic system code's results to the spatial mesh overlays used for modeling pressurized water reactor (PWR) cores for analysis of different transients. The coupled TRAC-PF1/NEM was used to model PWR rod ejection accident (REA). Modeling schemes for pressurized water reactor are described in detail, followed by a comparative analysis of both steady state and transient calculations. By using different TRAC-PF1/NEM vessel modeling options it was demonstrated that the geometric refinement plays a great role in determining the local parameters and control rod worth in the case of spatially asymmetric transients. The capability of TRAC-PF1/NEM to introduce local refinement of heat structure models was explored while preserving the original coarse-mesh structure of the hydraulic model. The obtained results indicated that the thermal-hydraulic feedback phenomenon is non-linear and cannot be separated even in rod ejection accident analysis, where the Doppler feedback plays a dominant role. While the impact of neutronics mesh refinement is well known, this research found that the local predictions, as well as the global predictions are also very sensitive to the thermal-hydraulic refinement.

Recombination mechanisms in AlGaNand their effects on the response of ultraviolet detectors

We have fabricated and characterized UV photoconductors based onAl0.09Ga0.91N grownby MOCVD. We show the mechanism of carrier recombinationincluding deep levels and their effects on the photoresponse.Photoluminescence and transmission experiments show theexistence of a large acceptor band. The spectral photoresponsealso shows a subbandgap contribution corresponding to theacceptor band absorption. Additional non-ideal phenomena wereobserved: response nonlinearity with power density,non-exponential decays and asymmetric transients when thelight is switched on or off. In addition, we could stronglymodulate the above bandgap response by a simultaneousillumination with photon energies corresponding to thetransition from the A band to the conduction band. A model ofrecombination involving the acceptor band quantitativelyexplains all these effects.

RELAP5/MOD2 Split Reactor Vessel Model and Steamline Break Analysis

A split reactor vessel model for the RELAP5/ MOD2 computer code is developed in an attempt to realize more realistic predictions of asymmetrical transients in a two-loop nuclear power plant. Based on this split reactor model, coolant mixing processes within the reactor vessel are examined. This study evaluates the model improvements in terms of thermal-hydraulic simulations of the reactor core inlet fluid condition and the consequent core behavior. Furthermore, the split reactor vessel model is introduced into an integral RELAP5/MOD2 power plant model, and a steamline break analysis is performed to determine the influence of the boron concentration in the boron injection tank on accident consequences.

Development of an Efficient Three-Dimensional Reactor Core Model for Simulator Applications

This paper reports that an efficient personal computer-based three-dimensional reactor core model TRIDA is developed to simulate the dynamics of the reactor for simulator applications. A fast running matrix factorization technique is used to solve the coupled nodal kinetics equations. Calculations for several off-normal and highly asymmetric transients are performed, and some of the results are compared with those of the Arkansas Nuclear One Unit 2 Final Safety Analysis Report. In addition, the computation time of the matrix factorization technique is compared with those of the Gaussian elimination and Jacobi iterative techniques. Results of comparisons are also given.