Study of the MTC estimation by noise analysis in 2-D heterogeneous systems

The effect of a heterogeneous distribution of the temperature noise on the MTC estimation by noise analysis is investigated. This investigation relies on 2-group diffusion theory, and all the calculations are performed in a 2-D realistic heterogeneous core. It is shown, similarly to the 1-D case, that the main reason of the MTC underestimation by noise analysis compared to its design-predicted value lies with the fact that the temperature noise might not be homogeneous in the core, and therefore using the local temperature noise in the MTC noise estimation gives erroneous results. A new MTC estimator, which was previously proposed for 1-D 1-group homogeneous cases and which is able to take this heterogeneity into account, was extended to 2-D 2-group heterogeneous cases. It was proven that this new estimator is always able to give a correct MTC estimation with an accuracy of 3%. This small discrepancy comes from the fact that the reactor does not behave in a point-kinetic way, contrary to the assumptions used in the noise estimators. This discrepancy is however quite small.     

Study of the bias of the MTC estimate by noise analysis due to the presence of feedback

This paper discusses the bias of the non-parametric Moderator Temperature Coefficient (MTC) estimate due to the presence of feedback. Up to now the non-parametric estimation of the Frequency Response Function (FRF) is the most commonly used method to estimate the MTC by noise analysis. This estimation method is proportional to the Cross Power Spectral Density between the total neutron flux variation and total temperature variation divided by the auto power spectral density of the total temperature variation. The estimation method is very popular since feedback is considered to be negligible in the frequency band of interest. Unfortunately this is not the case in practice. Measurements at a Nuclear Power Plant in Belgium will be used to confirm that this feedback cannot be neglected. In case of feedback the chosen estimator always results in a biased estimate when there are external neutron flux variations present. It will be seen that the ratio between the external neutron flux and external temperature variation in combination with the amplitude of the feedback determines the bias. The theoretical analysis of the bias is based on a simplified scheme of the MTC measurement setup. A simulation in MATLAB is used to confirm the theoretical results. In order to avoid a biased estimate due to the feedback we will advise to measure the external temperature variation and to use another non-parametric estimator.     

Development and application of core diagnostics and monitoring for the Ringhals PWRS

Noise analysis and reactor diagnostics have been applied at the Ringhals PWRs for a long time. Through a collaboration with the Department of Reactor Physics, Chalmers University of Technology, methods for treating new problems were elaborated, and known methods were developed further to make them more effective or to suit specific applications. All these methods were tested in real measurements, and many of them have been used routinely afterwards. In this paper two particular new methods are described in detail: 1) the determination of the axial position of control rods from the axial shape of the neutron flux with neural network methods, and 2) the use of gamma thermometers for the determination of the MTC and for core flow estimation.     

Diagnostics of core barrel vibrations by in-core and ex-core neutron noise

Diagnostics of core-barrel vibrations has traditionally been made by use of ex-vessel neutron detector signals. We suggest that in addition to the ex-core noise, also the in-core noise, induced by core barrel vibrations, be also used. This would enhance the possibilities of diagnostics where the number of the ex-core detectors is not sufficient or their positions are disadvantageous for effective diagnostics, especially for shell-mode vibrations.

To this order, the theory of in-core noise induced by a fluctuating core boundary has been elaborated and applied to the diagnostics of beam and shell mode vibrations. The formulas were tested on some measurements taken in the Ringhals PWRs. The results confirm the validity of the model itself, and the possibilities for enhanced diagnostics were demonstrated. A more effective use of these novel possibilities requires more in-core detectors and/or better detector positioning.     

Monitoring MTC using cold leg thermocouple and background neutron detector noise signals in a VVER-440 reactor during a whole fuel cycle

Monitoring of the Moderator Temperature Coefficient (MTC) was performed from the noise signals of cold leg thermocouples and background neutron detectors in a VVER-440 type reactor during a whole fuel cycle. A modified traditional noise based estimator was applied: the estimator was extended in order to take into account the effects of measurement geometry, coolant velocity and the relatively long time constant of the thermocouples. A systematic evaluation of measurement settings and evaluation parameters was carried out in order to determine optimal parameters. Optimal evaluation parameters were determined by considering the frequency dependence of the estimator, and by minimizing the statistical and systematic errors of the results. It can be concluded that the modified estimator provides adequate results which are close to the MTC given by the core design code calculations. It was found that relatively long FFT window sizes are needed to obtain correct results. The method needs long but industrially acceptable measurements for robust operation.     

Indicator to estimate temperature sensitivity of resonance in temperature measurement by neutron resonance spectroscopy

Through series of simplification and simulation, we find that it is the ratio of z (=Δ/Γ) and σ₀ (cross-section at the exact resonance energy in the absence of temperature broadening) that contains the most information of temperature sensitivity for a resonance. We consequently define a factor h called the effective fitness indicator to represent the lower limit of temperature error for each resonance. h bears succinct forms and is tested against numerical simulations as well as gathered experimental data. Our further analysis and simulation justify the use of h at high temperatures (above several hundreds degrees centigrade). When the transmission of a resonance is free from suffering a flat-bottom (without ntσ₀ ? 1), h can be used to estimate the temperature sensitivity of individual resonance with an analytic formula constructed from fitting, telling a relation between the temperature error and h. Moreover, spread of emission time caused by the moderator, phosphorescent decay of the scintillator, and background fraction are all included in numerical simulations to reveal their influences.     

Identification of reactor internals'' vibration modes of a Korean standard PWR using structural modeling and neutron noise analysis

The vibration characteristics of a Korean standard PWR reactor internals have been estimated through a three-dimensional finite element analyses and verified by using the mode separated power spectral density functions obtained from the ex-core neutron noise signals. Also the natural vibration modes of the fuel assembly have been identified measuring both the ex-core and the in-core neutron noise signals which are close to each other. As a result, the fundamental bending mode frequency of the reactor internal structure is found to be around 8 Hz and the fundamental shell mode frequency 14.5 Hz, respectively. It is also shown that the fundamental bending mode frequency of the fuel assembly is 2.3 Hz and the 2^nd bending mode frequency 5.8 Hz, respectively. These results can be used for the supplements of the Korean standard PWR's CVAP (Comprehensive Vibration Assessment Program) data.     

Predicting of slow noise and vibration spectra degradation in the IBR-2 pulsed neutron source using a neural network simulator

Employing a neural net model of the noise state of the IBR-2 reactor (JINR, Dubna, Russia) and a model of the vibration state of movable reflectors in the reactor we have predicted slow degradation in reactor noises. Operative diagnostics and prediction of the reactor noise behavior with time involves separating of the reflector degradation trend in power noises. We investigate two neural models. The first concerns the vibrations of the reflectors and the second is a simplified reactor noise model. The predicted results are close, in character, to the experimental data. They show that it is the movable reflectors that are mainly responsible for the degradation of power noises.     

Application of advanced core process monitoring procedures in German power reactors

The nuclear reactor core design and the nuclear fuel management have been changed remarkable during the last few years. This development was initiated by increasing costs for the fuel recycling and nuclear waste storage. The fuel material, the fuel pellet fabrication, the fuel assembly structure and the core composition have been varied to get an effective fuel exploitation. Based on advanced core process conditions the reactor power and the fuel burn-up have been increased at German plants in recent years. Improved dynamic process monitoring procedures are required to get more information about the varied core process behaviour during the reactor operation. Since several years ISTec has been performed investigations to the process monitoring based on process signal measurements in German nuclear power plants. Using the standard instrumentation of the plants process signals have been measured and analysed by means of the digital data acquisition system SIGMA. The measured time signals are influenced by core process transients, global and local process fluctuations and by signal line transfer functions. Advanced time series analysis methods have been applied to separate different process effects in the multiple signal matrix. The separation of different process influences can improve significantly the information about the process condition in the reactor core.     

汽水分离再热器的传热系数及出口温度随功率变化规律的分析

本文从传热学的基本原理出发,推导出核电站汽水分离再热器的传热系数随功率变化的关系式,并用该关系式导出了汽水分离再热器的出口参数随功率的变化,从而为计算和设计汽水分离再热器的温度控制系统提供了理论基础。     

噪声法测量临界装置温度系数实验研究

为了用噪声法测量核电站反应堆的负慢化剂反应性温度系数(MTC),本文在铀溶液临界装置上研究噪声法实验测量反应性温度系数αT,并与周期法测量的αT进行比较。结果表明,两种方法测量的αT趋势基本一致。由于铀溶液临界装置中溶液的反应性温度效应与核电站反应堆的慢化剂温度效应的机理相似,因此本文利用噪声法测量铀溶液临界装置的αT对于核电站反应堆利用噪声法测量MTC有一定的参考价值。     

New proposal of reactivity coefficient estimation method using a gray-box model in nuclear power plants

A new method for estimating reactivity parameters, such as moderator temperature coefficient (MTC) and void reactivity coefficient (VRC), is proposed using steady-state noise data. In order to solve the ill-posed problem of reactivity parameter estimation, a concept of a gray box model is newly introduced. The gray box model includes a first principle based model and a black-box fitting model. The former model acts as a priori knowledge based constraints in a parameter estimation problem. After establishing the gray box and noise source models, the maximum likelihood estimation method based on Kalman filter is applied. Furthermore, it is shown that the frequency domain approach of the gray box model is useful in the case of VRC estimation. The effectiveness of the proposed algorithms is shown through numerical simulation and actual plant data analysis.     

Study of the deuterated titanium Ti2Hx sample by using nuclear reaction analysis and material analysis methods

Titanium is one of the best hydrogen loading material.The predicted maximum loading ratio of hydrogen in titanium may reach to 2.0.In this work.a titanium layer on molydbenum substrate was deuterated with the atomic ratio X=^2H/Ti≥1.6,The change of the surface topography and microstructure of the sample before and after lading was observed by using Scan Electron Microscopy (SEM),The surface segregation of the samples after deuteron bombardment was also observed.A fluctuatingly-incrcasing trend of the deuterium density in titanium target was detected in the deuteron implantation experiments.which indicated a suddenly explosion(segregation)or fast diffusion of deuterium in the titanium.Significant amount of nitrogen contamination was found in the Ti^2Hx sample by nuclear reaction analysis(NRA),which indicated that the Ti^2Hx structure might have the feature to trap nitrogen from air.The nitrogen contamination in Ti^2Hx significantly affects the increase of the atomic ration X=^2H/Ti.     

考虑SSI效应的核电站泵房结构楼层反应谱分析

采用Super SAP和CLASSI程序对某在建核电站泵房结构进行了极限安全地震震动和运行安全地震震动情况下的土壤-结构相互作用(SSI)的地震分析,揭示了结构在时域内的特性;通过傅立叶变换(FFT)分析了结构在频域内的特性,求得建筑结构中各楼层反应谱,结合核电厂设计规范分别给出了在较硬地基岩土条件下,结构考虑和不考虑SSI时各楼层反应谱,并对其进行了比较分析.结果表明,SSI效应对结构楼层反应谱的谱形、谱值以及零周期平台高度有一定的影响.     

核电厂丧失厂外电源事件下柴油发电机系统动态响应分析

将系统可靠性分析方法GO法与Markov法相结合,对核电厂概率安全分析(PSA)中厂外电源丧失(LOOP)后柴油发电机应急响应系统在24h内缓解全厂断电(SBO)事件中的动态过程进行分析,解决了维修相关存在下可修系统可靠性精确计算问题,并通过创建GO法“备用门”操作符真实地模拟应急响应系统工作的逻辑关系.通过将2种可靠性分析方法相结合使用的尝试,使之与柴油发电机应急响应系统存在维修相关的实际情况相适应,拓展了2种方法的分析领域,同时能够更为精确地得出SBO对系统安全运行的影响.     

嫦娥一号和嫦娥二号月球γ能谱解析

为了从嫦娥系列绕月γ能谱测量实验获取月表γ能谱特征,应用CE1-GRS和CE2-GRSγ能谱2C级科学数据,首先从散射本底扣除后的谱线中提取出弱特征峰信息,然后对嫦娥系列仪器谱进行定性分析,推定月表存在的可能元素种类。结果表明:(1)从CE1-GRSγ能谱能识别U、Th、K、Fe、Mg、Si、Al和O等8种月表可能元素,从CE2-GRSγ能谱能识别U、Th、K、Fe、Mg、Si、Al、O、Ti和Ca等10种月表可能元素;(2)卫星绕行轨道距月表距离越近、探测器的能量分辨率越高,得到的月球轨道γ能谱将更精细,从而能获取更准确的月表元素种类及其分布特征。     

CAP1400非能动余热排出系统自然循环试验风险分析与应对措施

大型先进压水堆(CAP1400)非能动余热排出系统(PRHR)自然循环试验是CAP1400首堆试验项目之一,也是调试期间的重大瞬态试验。试验过程中,由于反应堆一回路温度、压力和液位等参数剧烈变化,增大了试验风险,对机组运行控制提出了较高要求。本文在AP1000调试实践的基础上,从降低自然循环试验风险角度分析提出利用功率运行后的真实衰变热执行本试验。同时针对试验过程一回路压力、温度,稳压器(PZR)液位及堆外源量程等参数剧烈变化产生的安全风险分析,并制定相应的应对措施,为后续CAP1400 PRHR自然循环试验安全实施提供有力支撑。     

压水堆SBLOCA事故环路水封清除现象机理与模型影响研究

环路水封清除（LSC）是压水堆冷管段小破口失水事故（SBLOCA）的典型事故特征之一。为确定LSC现象的物理模型影响,探究准确复现LSC现象的物理模型设置,从LSC现象物理机理的角度,对影响LSC的主要物理模型进行梳理和分析。结合LOBI台架SBLCOA系列实验,对LSC现象物理机理、物理模型影响进行模拟和验证。结果表明,在对影响LSC现象的物理模型进行合理设置后,RELAP5程序模型能较好地复现LOBI台架实验工况中的LSC现象,验证了LSC现象物理模型影响及模型设置的合理性。      

Analysis of an ex-vessel break in the ITER divertor cooling loop

In the present work the integrated ECART code, developed for severe accident analysis in LWRs, is applied on the analysis of a large ex-vessel break in the divertor cooling loop of the international thermonuclear experimental reactor (ITER). A comparison of the ECART results with those obtained by Studsvik Nuclear AB (S), utilizing the MELCOR code, was also performed in the general framework of the quality assurance program for the ITER accident analyses. This comparison gives a good agreement in the results, both for thermal-hydraulics and the environmental radioactive releases. Mainly these analyses, from the point of view of the ITER safety, confirm that the accidental overpressure inside the vacuum vessel and the Tokamak cooling water system (TWCS) Vault is always well below the design limits and that the radioactive releases are adequately confined below the ITER guidelines.