Validation of a methodology for fuel management analysis of Laguna Verde nuclear power plant

This paper shows the validation of the fuel management methodology based on the state of the art lattice physics code HELIOS and the CM-PRESTO code, for the fuel management analysis of the Laguna Verde nuclear power plant (LVNPP). The validation of these codes is performed with data from the first five operating cycles of LVNPP Unit 1. HELIOS calculations were performed for three different BWR standard type assemblies and compared with Monte Carlo, RECORD and fuel vendors’ results. The CM-PRESTO results are compared against plant information, such as K-effective at hot and cold conditions, thermal limits calculated by the process computer and instruments readings from the traveling in-core probes (TIP) and the local power range monitor (LPRM). Results show an improvement compared with those obtained with the previous methodology based on RECORD/PRESTO-B.     

Bistable flow spectral analysis. Repercussions on jet pumps

There have been many attempts at characterizing and predicting bistable flow in boiling water reactors (BWRs). Nevertheless, in most cases the results have only managed to develop models that analytically reproduce the phenomenon (Gavilán Moreno, 2009). Modeling has been forensic in all cases, while the capacity of the model focus on determining the exclusion areas on the recirculation flow map. The bistability process is known by its effects given there is no clear definition of its causal process.In the 1980s, Hitachi technicians (Miura et al., 1986) managed to reproduce bistable flow in the laboratory by means of pipe geometry, similar to that which is found in recirculation loops. The result was that the low flow pattern is formed by the appearance of a quasi stationary, helicoidal vortex in the recirculation collector's branches. This vortex creates greater frictional losses than regions without vortices, at the same discharge pressure. Neither the behavior nor the dynamics of these vortices were characterized in this paper. The aim of this paper is to characterize these vortices in such a way as to enable them to provide their own frequencies and their later effect on the jet pumps.The methodology used in this study is similar to the one used previously when analyzing the bistable flow in tube arrays with cross flow (Olinto, 2006). The method employed makes use of the power spectral density function. What differs is the field of application. We will analyze a Loop B with a bistable flow and compare the high and low flow situations. The same analysis will also be carried out on the loop that has not developed the bistable flow (Loop A) at the same moments in time.The next analysis will be carried out by comparison between Loops A and B for the same time periods, one with a stable flow and the other with a bistable one.Another analysis, arising from the previous ones, will study the autocorrelation coefficient of the flows’ time series, both of the bistable flow (Loop B), as well as of the stable flow (Loop A). In this way certain information will be forthcoming on the series structure, and therefore, about the process.Finally, an analysis will be carried out on the jet pump flows, selecting the end (1 and 10 for Loop A and 11 and 20 for Loop B) and the middle ones (5 for Loop A and 15 for Loop B). The aim is to determine what is the effect that the bistable flow, has on the jet pumps and their response in terms of frequency.     

Azcatl-CRP: An ant colony-based system for searching full power control rod patterns in BWRs

We show a new system named AZCATL-CRP to design full power control rod patterns in BWRs. Azcatl-CRP uses an ant colony system and a reactor core simulator for this purpose. Transition and equilibrium cycles of Laguna Verde Nuclear Power Plant (LVNPP) reactor core in Mexico were used to test Azcatl-CRP. LVNPP has 109 control rods grouped in four sequences and currently uses control cell core (CCC) strategy in its fuel reload design. With CCC method only one sequence is employed for reactivity control at full power operation. Several operation scenarios are considered, including core water flow variation throughout the cycle, target different axial power distributions and Haling conditions. Azcatl-CRP designs control rod patterns (CRP) taking into account safety aspects such as k_eff core value and thermal limits. Axial power distributions are also adjusted to a predetermined power shape.     

Novel fluctuation reduction procedure for nuclear reactivity calculations based on the discrete fourier transform method

A new method for calculating nuclear reactivity based on the Discrete Fourier Transform (DFT) – with two filters: a first-order delay low-pass filter and a Savitzky-Golay filter – is presented. The reactivity is calculated from an integrodifferential equation known as the inverse point kinetic equation, which contains the history of neutron population density. The new method can be understood as a convolution between the neutron population density signal and the response to the characteristic impulse of a linear system. The proposed method is based on the discrete Fourier transform (DFT) that performs a circular convolution. The fast Fourier transform algorithm (FFT) with the zero-padding technique is implemented to reduce the computational cost.     

On the use of the multivariate autoregressive and relative power contribution models for determination of transient root cause

The multivariate autoregressive (MAR) model and the relative power contribution (RPC) ratio are used in this work to determine the root causes of a power oscillation event and an apparent positive reactivity insertion transient occurred at the at the BWR/5 Units of the Laguna Verde Nuclear Power Plant (LVNPP) of México. The application of the MAR and PRC models leads to identify dominant frequencies and the contribution from other different signals to the dominant frequencies. The methodology firstly uses a linear model to estimate the response characteristics of the system and the spectra of the noise sources. The estimate of the process linear predictor is obtained by the ordinary least squares method. Then, the model performs a MAR analysis, and the RPC ratio is computed to determine the inter-relationship between the different reactor noise signals. The RPC ratio is an indication of how the fluctuation of one variable depends on other variables, at each frequency.Reactor signals acquired during the two transient operational events are used in the analysis. In the first event, a problem on the position controller of the flow control valve's stem induced the appearance of a power peak of 12% amplitude on the average power range monitors. Actual insertion of positive reactivity did not occur. The signals used for the analysis come from an average power range monitor, the position of the stem in the valve, controller of stem position, and controller of the recirculation flow. For the second transient, power oscillations of about 12% amplitude occurred. Signals from an average power range monitor, total flow through the core and flow through the 10 jet pump of each loop are analyzed. In both cases, some other signals were also used, but since they did not show appreciable influence on the RPC results, they were not considered for final analysis.The RPC results obtained in here confirmed previously known facts about the origin of the transients analyzed. Specifically, for the first transient, a dominant frequency of 1.7 Hz appeared on the power spectral densities of different signals from sensors on the recirculation loop B. At this 1.7 Hz frequency, the RPC ratio showed influence of such loop B spectra to the average reactor power spectrum, but no influence at all of the average reactor power to any of such loop B signals. The root cause, although not a real power transient event, therefore was not of neutronic nature, but related to recirculation flow. In the second transient, the prominent power oscillation frequency (0.54 Hz) was tracked within the spectral data of other signals. The RPC results for this case showed a strong influence of the average reactor power on the flow signals, but only a modest contribution from the flow signals to the average reactor power. The root cause therefore was of neutronic nature, due mainly to a combination of a particular core configuration and control rod pattern change at the moment of the event.     

Wavelet time–frequency analysis of accelerating and decelerating flows in a tube bank

In the present work, the steady approximation for accelerating and decelerating flows through tube banks is discussed. With this purpose, the experimental study of velocity and pressure fluctuations of transient turbulent cross-flow in a tube bank with square arrangement and a pitch-to-diameter ratio of 1.26 is performed. The Reynolds number at steady-state flow, computed with the tube diameter and the flow velocity in the narrow gap between the tubes, is 8 × 10⁴. Air is the working fluid. The accelerating and decelerating transients are obtained by means of start and stop of the centrifugal blower. Wavelet and wavelet packet multiresolution analysis were applied to decompose the signal in frequency intervals, using Daubechies 20 wavelet and scale functions, thus allowing the analysis of phenomena in a time–frequency domain. The continuous wavelet transform was also applied, using the Morlet function. The signals in the steady state, which presented a bistable behavior, were separated in two modes and analyzed with usual statistic tools. The results were compared with the steady-state assumption, demonstrating the ability of wavelets for analyzing time varying signals.     

Semi-implicit modeling of condensation in the presence of non-condensables in the computer code

A methodology is developed for modeling condensation in the presence of non-condensables based on the stagnant film (couette flow) model. The methodology, which is fully compatible with the mathematical representation and numerical solution scheme for the two-fluid conservation equations in the computer code, rigorously treats the coupling between the heat and mass transfer processes and allows for implicit treatment of the interphase mass and momentum exchange terms without iterations and without adversely affecting the stability of the code's solution scheme. The computer code is accordingly modified and is successfully validated against published experimental data.     

Critical Heat Flux Prediction Method Based on Two-Phase Turbulence Model

In this paper, we present an analytical methodology to predict forced convective CHF (Critical Heat Flux) for DNB (Departure from Nucleate Boiling) type boiling transition that occurs inside of uniformly heated round tubes. Axial directional two-phase flow analysis was conducted based on one-dimensional two-fluid model and typical constitutive models. At the same time, the radial directional distribution of void fraction at any axial location was calculated based on the bubble diffusion model, which was coupled with two-phase turbulence model for boiling bubbly flow. The calculated void fraction showed the wall peak distribution, and was compared with experimental data, which was derived from subcool boiling experiments. IPNVG (Incipient Point of Net Vapor Generation), which means the starting point of two-phase flow analysis, was also investigated well, since it was revealed that IPNVG had a significant influence on CHF prediction. By using this methodology for calculating radial directional void fraction distribution, we carried out CHF prediction for water on the assumption that DNB would occur when the local void fraction near the heated wall exceeds a critical value. The predicted CHF agreed well with experimental data, and the accuracy was within about 20%.     

Semi-implicit modeling of condensation in the presence of non-condensables in the relap5/mod3 computer code

A methodology is developed for modeling condensation in the presence of non-condensables based on the stagnant film (couette flow) model. The methodology, which is fully compatible with the mathematical representation and numerical solution scheme for the two-fluid conservation equations in the relap5/mod3 computer code, rigorously treats the coupling between the heat and mass transfer processes and allows for implicit treatment of the interphase mass and momentum exchange terms without iterations and without adversely affecting the stability of the code's solution scheme. The relap5/mod3 computer code is accordingly modified and is successfully validated against published experimental data.     

Effect of cooling temperature of electrodes on Joule-heating flow in cubic cavity

The effect of cooling condition of electrode plates on chaotic Joule-heating flow behavior in a cubic cavity has been experimentally investigated. Two electrode plates are immersed in the fluid body and placed on the opposing side wall. They are connected to an AC power source for internal volumetric ‘Joule-heating’ in the cavity. This chaotic flow which induced by Joule heating is observed within whole the cavity. The chaotic flow is investigated quantitatively by Ultrasonic Velocity Profiling (UVP) method. The chaotic flow behavior is also observed by using Particle Image Velocimetry (PIV) method to understand the effect of temperature distribution in this condition. The chaotic flow behavior generates fluctuations of temperature and velocity. As a result, cooling condition of electrode plates has a strong effect on the Joule-heating flow behavior that the vortex area occurred in the upper part of cavity. On the other hand, in the adiabatic condition, unstable flow appeared in whole cavity. In additional, velocity profiles are analyzed by Fast Fourier Transform (FFT) method about the frequencies of fluctuations. Furthermore, a numerical analysis using Finite Element Method, GSMAC-FEM, is also examined the Joule-heating flow behavior for cubic cavity.     

A generic semi-implicit coupling methodology for use in RELAP5-3D

A generic semi-implicit coupling methodology has been developed and implemented in the RELAP5-3D^© computer program. This methodology allows RELAP5-3D^© to be used with other computer programs to perform integrated analyses of nuclear power reactor systems and related experimental facilities. The coupling methodology potentially allows different programs to be used to model different portions of the system. The programs are chosen based on their capability to model the phenomena that are important in the simulation in the various portions of the system being considered and may use different numbers of conservation equations to model fluid flow in their respective solution domains. The methodology was demonstrated using a test case in which the test geometry was divided into two parts, each of which was solved as a RELAP5-3D^© simulation. This test problem exercised all of the semi-implicit coupling features that were implemented in RELAP5-3D^© The results of this verification test case show that the semi-implicit coupling methodology produces the same answer as the simulation of the test system as a single process.     

Methodology to optimize radiation protection in radioactive waste disposal after closure of a disposal facility based on probabilistic approach

We propose a methodology to optimize radiation protection in radioactive waste disposal after the closure of a disposal facility based on a probabilistic approach. In this methodology, a set of alternative options of the disposal system design with the associated uncertainties estimated through a probabilistic approach are developed. Then, the methodology evaluates the advantages and disadvantages of each option for the optimization of radiation protection, in which it is possible to determine the compliance with a dose constraint of 0.3 mSv/y and to collect information for making a decision on the optimization. In particular, the obtained information on the exposure, which is the mode and width of the dose distribution, can be converted into information on engineering measures and site conditions for the options. This process allows us to discuss which option should be selected as the optimal option by considering the balance between the exposure and the engineering, economic, and social feasibility of the option. This methodology is helpful for providing clear reasons why an optimal disposal system design is selected by combining quantitative information on the exposure and the feasibility of each option.     

Study of visualized simulation and analysis of nuclear fuel cycle system based on multilevel flow model

Complex energy and environment system, especially nuclear fuel cycle system recently raised social concerns about the issues of economic competitiveness, environmental effect and nuclear proliferation. Only under the condition that those conflicting issues are gotten a consensus between stakeholders with different knowledge background, can nuclear power industry be continuingly developed. In this paper, a new analysis platform has been developed to help stakeholders to recognize and analyze various socio-technical issues in the nuclear fuel cycle system based on the functional modeling method named Multilevel Flow Models （MFM） according to the cognition theory of human being. Its character is that MFM models define a set of mass, energy and information flow structures on multiple levels of abstraction to describe the functional structure of a process system and its graphical symbol repre- sentation and the means-end and part-whole hierarchical flow structure to make the represented process easy to be understood. Based upon this methodology, a micro-process and a macro-process of nuclear fuel cycle system were selected to be simulated and some analysis processes such as economics analysis, environmental analysis and energy balance analysis related to those flows were also integrated to help stakeholders to understand the process of decision-making with the introduction of some new functions for the improved Multilevel Flow Models Studio, and finally the simple simulation such as spent fuel management process simulation and money flow of nuclear fuel cycle and its levelised cost analysis will be represented as feasible examples.     

基于图像小波包信息熵和遗传神经网络的气-液两相流流型识别

根据小波包变换能够将图像信号按不同尺度进行分解的特性,提出了基于图像小波包信息熵特征和遗传神经网络相结合的气-液两相流流型识别的新方法.该方法采用高速摄影系统获取水平管道内气-液两相流的流动图像,经过处理,对图像进行多分辨率分析,提取小波包变换系数的信息熵特征,用主成分分析法降低特征维数构成特征矢量,作为流型样本对遗传神经网络进行训练,实现了对流动图像的流型智能化识别.结果表明:图像小波包信息熵特征可以很好地反映各流型之间的差异;遗传神经网络结合遗传算法和BP算法各自优点,具有收敛速度快、不易陷入局部极小的特性,网络识别率为100%.     

Analysis of Unsteady Turbulent Triple Jet Flow with Temperature Difference

Suitable analysis methodology is required to obtain detailed information about magnitude and frequency of temperature variation of flow field for the study of thermal stripping phenomena. The large eddy simulation (LES) is applied to analyze unsteady turbulent triple jet water flow which can be a direct cause of thermal stripping. Current analyses are performed with different sub-grid scale models, number of grids, time increments, and inlet temperature intensities to find the effects of these on prediction. Predicted results of the LES are compared with experimental results. The LES successfully produces a time history of turbulence variables, which can be used to evaluate magnitude and frequency of instantaneous temperature. The LES tends to predict higher levels of root mean square temperatures compared to those of an experiment, indicating very active mixing effect among triple jets. The LES is found to be able to provide reliable frequency information about temperature fluctuation. The different sub-grid scale models show no significant difference in prediction ability and other variations of the LES prediction show no significant difference in prediction either. However, cases using the fine grid and the small time increment are slightly better than others. Further study is desired with different levels of inlet temperature intensities and separate sub-grid scale models for temperature field.     

CFD methodology and validation for single-phase flow in PWR fuel assemblies

This paper presents the CFD modeling methodology and validation for steady-state, normal operation in a PWR fuel assembly. This work is part of a program that is developing a CFD methodology for modeling and predicting single-phase and two-phase flow conditions downstream of structural grids that have mixing devices. The purpose of the mixing devices (mixing vanes in this case) is to increase turbulence and improve heat transfer characteristics of the fuel assembly. The detailed CFD modeling methodology for single-phase flow conditions in PWR fuel assemblies was developed using the STAR-CD CFD code. This methodology includes the details of the computational mesh, the turbulence model used, and the boundary conditions applied to the model. The methodology was developed by benchmarking CFD results versus small-scale experiments. The experiments use PIV to measure the lateral flow field downstream of the grid, and thermal testing to determine the heat transfer characteristics of the rods downstream of the grid. The CFD results and experimental data presented in the paper provide validation of the single-phase flow modeling methodology. Two-phase flow CFD models are being developed to investigate two-phase conditions in PWR fuel assemblies, and these can be presented at a future CFD Workshop.     

Contribution of the mathematical modelling of knowledge to the evaluation of uncertainty margins of a LBLOCA transient (LOFT-L2-5)

This work is devoted to some recent developments in uncertainty analysis of the computer code responses used for accident management procedures in nuclear industry. The classical probabilistic approach to evaluate uncertainties is recalled. In this case, the statistical treatment of the code responses is based on the use of order statistics. It provides direct estimations of relevant statistical measures for safety studies. However, the lack of knowledge about uncertainty sources can deteriorate the decision-making. To respect the real state of knowledge, a second model, based on the Dempster–Shafer theory is introduced. It allows to mix the probabilistic approach with the possibility theory that is more appropriate when few information is available. An application of both methodologies to the uncertainty analysis of a LBLOCA transient (LOFT-L2-5) is given.     

Hot point detection method

A method is described by which the information obtained on-line through a system of neutron measuring devices such as self-powered neutron detectors (SPNDs, also called collectrons) inserted in the core of a nuclear power reactor (in particular, a PWR) allows the on-line detection of a possible hot spot during plant operation. The method is based on the generalized perturbation method (GPT) techniques, for the calculation of the sensitivity coefficients of the integral quantities measured with the collectrons with respect to parameters representative of the hot spot, and on the use of statistical inference techniques, taking into account the errors associated with the measurements. The methodology allows to assess the effect on the quality of the hot point detection system following possible failures of the measuring devices during the core life cycle. Such an assessment may be useful for defining an adequate protection strategy in terms of quality, number and distribution of the collectrons.     

钠冷快堆中间回路建模仿真及流量控制研究

针对钠冷快堆中间回路泵、管道、换热器等,采用Matlab/Simulink软件建立了一种仿真模型,对回路的流量和管道换热进行了计算。根据相似理论、泵水力特性曲线及回路压力损失等计算流量。编制了SFAC V1.0程序,该程序的计算结果与实验值符合较好,最大相对误差为5%。将管道划分为不同节段,在各节段上建立能量守恒微分方程组,从而建立了管道换热计算的模型。同时,对钠流量的控制方式进行了设计和改进,对控制参数进行了整定,并对流量需求进行了计算。计算结果表明,该控制方式的控制品质较高。     

Development of natural circulation analysis methods for a sodium cooled fast reactor

A natural circulation evaluation methodology has been developed to insure safety of a sodium cooled fast reactor (SFR) of 1500 MWe adopting a natural circulation decay heat removal system (NC-DHRS). The methodology consists of a one-dimensional safety analysis which can be applied to safety evaluation for SFR licensing taking into account the temperature flattening effect due to buoyancy force in the core, and a three-dimensional fluid flow analysis which can evaluate thermal-hydraulics for local convection and thermal stratification in the primary system and DHRSs. The one-dimensional safety analysis method and the three-dimensional fluid flow analysis method have been validated using the test results of a water test apparatus and a sodium test loop for some typical transient events selected from the design basis events of the SFR. Finally, it has been confirmed that a good agreement between the test results and analysis results has been obtained, and reliability of each method has been demonstrated.