Dynamics modeling and stability analysis of a fluidized bed nuclear reactor

A theoretical model describing the coupling of neutronics, thermohydraulics and fluidization in a fluidized bed nuclear reactor is presented. The stability of the system is investigated by linearizing and perturbing the system around its equilibrium points and identifying the root loci of the sytem. It is found that within the operational range, the eigenvalues are located in the negative part of the phase plane, implying linear stability. Simulations of transient conditions are performed, viz. a hypothetical startup transient and a quasistatic transient related to noise resulting from stochastic movements of the fuel particles. These simulations show that although the total power of the reactor may reach high values, the fuel temperature is well below safety limits at all times.     

Application of On-Line Digital Noise Analysis to Reactor Diagnosis in JMTR

The reactor noise analysis technique is particularly useful in reactor diagnosis for on-line monitoring if the raw noise signals can be processed in almost real time.

An on-line reactor noise analysis system has been developed with use made of the mini-computer HITAC-10. This system utilizes functions for calculating the power spectral density in almost real time, plots the output by digital incremental plotter, and displays the results by means of color graphic display equipment, in order to detect anomalous reactor conditions with the statistical technique.

Using this system, reactor noise signals have been measured and analyzed under various operational conditions in the JMTR. The variance of the power spectral density is found to fit a logarithmic probability density function. This function is independent of the frequency, but is dependent on the number of sampling functions.

A logical procedure for anomaly detection based on statistical characteristics has been developed. It is applied to a case where it is supposed that the PWR operating mode in the OWL-1 is the normal process and that the BWR mode is the anomalous. It is demonstrated as a result, that this procedure can successfully detect anomalous processes.     

Application of a nonlinear dynamical descriptor to BWR stability analyses

Reactor noise analyis method based on nonlinear dynamical theory is applied to the Forsmarks 1&2 BWR stability benchmark organized by OECD/NEA. The method utilizes the determination process of the fractal dimension of oscillatory neutron-flux signals. For practical application, the fractal dimension is expected to classify different asymptotic regimes of nonlinear dynamical systems. In this case, each signal is classified into stable, quasi-stable, and unstable states. It was confirmed that the result was consistent to that of decay ratios. In addition, because the data processing does not include fitting calculation often used by the decay ratio, it is surmised that the result hardly varies with analyzers. This can be the prominent advantage of this methodology compared to decay ratio.     

Real time reactor noise diagnostics for the Borssele (PWR) nuclear power plant

After the upgrade of Borssele NPP in 1997, core cycle 24, the power plant operated three years more with 91% availability. The authority of the power plant decided to enhance and upgrade the reactor trend monitoring and plant information recording system with higher frequencies than the plant data processing system (PPS) as well as installing a flexible and multiple-purpose reactor noise analysis system which may support the reactor maintenance group with on-line and off-line capabilities for several different signal processing applications. Two measuring and monitoring systems were built in 2001 and fully taken in implementation during the start-up of the new core 28. In this sense, the new system was used in power operation during the 29^th of September 2001. This paper will introduce the measuring system, the operational tasks, and the results obtained so far on the real-time core-barrel motions (CBM) and the two-primary coolant pump vibrations measured through the reactor noise analysis.     

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.     

数字化核测量系统反应堆倍周期计算的敏感性分析

反应堆倍周期是核反应堆工程中的一个重要参数。在反应堆启动和功率提升过程中,操纵员可通过反应堆倍周期来了解反应堆的运行状态,并据此控制反应性。数字化核测量系统通过对与反应堆功率成正比的电压信号进行采样和处理,计算得到反应堆倍周期。在实际的应用中,电压信号往往包含测量噪声,对计算结果带来较大的不确定性。针对数字化核测量系统的倍周期计算问题,对其敏感性进行了分析,并给出相应的算例。     

Research activities on nuclear reactor physics and thermal-hydraulics in Japan after Fukushima-Daiichi accident

Research and development in nuclear reactor physics and thermal-hydraulics continue to be vital parts of nuclear science and technology in Japan. The Fukushima accident not only brought tremendous change in public attitudes towards nuclear engineering and technology, but also had huge influence towards the research and development culture of scientific communities in Japan. After the Fukushima accident, thorough accident reviews were completed by independent committees, namely, Tokyo Electric Power Company (TEPCO), the Japanese government, the Diet of Japan, the Rebuild Japan Initiative Foundation, and the Nuclear and Industrial Safety Agency. Reactor physics and thermal-hydraulics divisions of Atomic Energy Society of Japan (AESJ) also issued the roadmaps after the accident. As a result, lessons learned from the accident were made clear, and a number of new research activities were initiated. The present paper reviews ongoing nuclear engineering research activities in Japanese institutes, universities, and corporations, focusing on the areas in reactor physics and thermal-hydraulics since the Fukushima accident to the present date.     

压水堆核电厂液态流出物中55Fe的排放与监测方法

本文调研分析压水堆核电厂液态流出物中排放⁵⁵Fe的来源、排放的统计参考值和⁵⁵Fe的分析方法,提出开展核电厂液态流出物中⁵⁵Fe监测的建议。统计分析了美国41座压水堆核电厂在2005～2017年液态流出物中⁵⁵Fe的排放量,其发电量归一化排放量的几何平均值范围为5.18×10^-6～8.14×10^-5 GBq/GWh,所有压水堆电厂液态流出物中⁵⁵Fe排放量的几何平均值为1.52×10^-5 GBq/GWh,各年度⁵⁵Fe排放量在液态流出物中占比在12%以上,排第1至第4位。根据我国典型压水堆核电厂液态流出物排放体积,估算了液态流出物中⁵⁵Fe的排放浓度,约10.7 Bq/L。建议推进核电厂液态流出物中⁵⁵Fe监测方法的建立和完善。通过对⁵⁵Fe监测方法的调研,推荐采用固相萃取树脂的快速分析方法。     

Investigation of Resonant Power Oscillation in Halden Boiling Water Reactor Based on Noise Analysis

In the Halden Boiling Water Reactor (HBWR), a resonant power oscillation with a period around 0.04 Hz is observed at power levels higher than about 9.5 MWt. Although this resonant oscillation is not so strong as to affect the normal reactor operation, it is significant, from the viewpoint of reactor diagnosis, to reveal the cause of this oscillation as well as to understand its characteristics.

Noise analysis based on the autoregressive (AR) modeling technique together with spectral and correlation analyses is performed to investigate the driving source, which indicates that it is attributed to the dynamic interference with the reactor of heat exchange process in two parallel-connected steam transformers.

The present study demonstrates the effectiveness of the technique applied here for determining the so-called noise source inducing variations of quantities in a system together with its applicability to various problems in the field of reactor noise analysis and diagnosis.     

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.     

On-line monitoring applications in nuclear power plants

The nuclear power industry is working to reduce generation costs by adopting condition-based maintenance strategies and automating testing activities. These developments have stimulated great interest in on-line monitoring (OLM) technologies and new diagnostic and prognostic methods to anticipate, identify, and resolve equipment and process problems and ensure plant safety, efficiency, and immunity to accidents. This paper provides examples of these technologies with particular emphasis on eight key OLM applications: detecting sensing-line blockages, testing the response time of pressure transmitters, monitoring the calibration of pressure transmitters on-line, cross-calibrating temperature sensors in situ, assessing equipment condition, performing predictive maintenance of reactor internals, monitoring fluid flow, and extending the life of neutron detectors. These applications are discussed in the following sections. Emphasis is placed on the principles of a core OLM method - noise analysis - and the technical requirements for an integrated OLM system are summarized.     

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.     

Measurement of large negative reactivity of an accelerator-driven system in the Kyoto University Critical Assembly

Large negative reactivity of a subcritical system driven by a pulsed 14 MeV neutron source has been measured in the Kyoto University Critical Assembly. The subcriticality of the accelerator-driven system (ADS) ranged in effective multiplication factor roughly from 0.98 to 0.92, which corresponded to an operational range of an actual ADS proposed by Japan Atomic Energy Agency. As the measurement technique, pulsed neutron method, power spectral analysis for pulsed neutron source, accelerator-beam trip method were employed. From neutron count decay data obtained by the pulsed neutron experiment, not only the prompt-neutron decay constant of fundamental mode but also a higher spatial mode could be derived. The subcriticality was also determined from the fundamental decay constant. The measured cross-power spectral density consisted of a familiar correlated reactor-noise component and many uncorrelated delta-function-like peaks at the integral multiple of pulse repetition frequency. The fundamental prompt-neutron decay constant, i.e., the subcriticality determined from the latter uncorrelated peaks was consistent with that obtained by the above pulsed neutron experiment. However, the magnitude of the former correlated component was reduced with an increase in the subcriticality and eventually this component became almost white at deeply subcritical state ranging in the multiplication factor under 0.95. Consequently, the determination of the decay constant from the correlated component was impossible under such a subcritical state. As data analysis method for the beam trip experiment, both the conventional integral count method and the least-squares inverse kinetics method (LSIKM) were employed. The LSIKM analysis led to the consistent subcriticality with that obtained by the pulsed neutron experiment, while the integral count method significantly underestimated the subcriticality. This underestimation originated from a residual background count, which was maintained after the beam trip. The LSIKM was mostly not influenced by such a slight count rate.     

Acoustic Emission in Subcooled Nucleate Pool Boiling

Intensity and power spectral density of acoustic emission in subcooled nucleate pool boiling of water from boiling inception to the heater burnout was investigated experimentally. Platinum and Ni-Cr wires with dia. 1.0 mm or smaller were used as the heater. The measurements were performed in containers at atmospheric pressure and in a dam at pressures up to 3 atm. In the former case photographic observations of boiling bubbles were also made.

The acoustic emission in boiling had frequency components up to and sometimes beyond 50 kHz; the shape of power spectral densities behaved differently for heat fluxes above and below a certain value at which the overall acoustic intensity assumed a maximum value. With the increase in heat flux the acoustic emission increased in high frequencies below this heat flux, while it became eminent in low frequencies above the same heat flux. This phenomenon is related to the transition from the region with single bubbles to the region with coalesced bubbles. Resonance-like peaks found in the measurements with containers that may depend on the container geometry were not observed in the measurements in the dam and smoother power spectral densities were obtained.     

Application of reactor noise analysis in the candu reactors of Ontario hydro

Reactor noise analysis techniques are being applied in Ontario Hydro's CANDU nuclear generating stations to monitor the dynamic characteristics of critical plant components and processes. A comprehensive analysis of stationary signal fluctuations (noise) of the standard instrumentation of Pickering-B, Bruce-B and Darlington units has been carried out in the past two years. In these measurements the feasibility of applying noise analysis techniques to actual operating data has been demonstrated. The results indicated that the detection and characterization of instrument and process failures, and validation of process signals and instrument functionality can be based on the existence of certain statistical signatures derived from the measured reactor noise signals.     

Reactor noise measurements in the safety and regulating systems of CANDU stations

Reactor noise measurements of safety and regulating system intrumentation are performed in the CANDU nuclear power stations of Ontario Power Generation (OPG) and Bruce Power. Station signals included in the noise measurements are in-core flux detectors (ICFD), ion chambers (I/C), flow transmitters, pressure transmitters, and resistance temperature detectors (RTD). Their frequency dependent noise signatures are regularly measured during steady-state operation, and are used for parameter estimation and anomaly detection.

The specific applications include the following areas:

Flux noise measurements to detect and characterize (a) anomalies of in-core flux detectors, ion chambers and their electronics, (b) mechanical vibration of fuel channels and in-core detector tubes induced by coolant/moderator flow.

Pressure and flow noise measurements to estimate the in-situ response times of flow/pressure transmitters and their sensing lines installed in the reactor's coolant loops.

Temperature noise measurements to estimate the in-situ response times of thermal-well or strap-on type RTDs installed in the reactor's coolant and moderator loops.

Development and assessment of fire-related risk unavailability matrices to support the application of the maintenance rule in a PWR nuclear power plant

Two methods are presented which serve to incorporate the fire-related risk into the current practices in nuclear power plants with respect to the assessment of configurations. The development of these methods is restricted to the compulsory use of fire probabilistic safety assessment (PSA) models. The first method is a fire protection systems and key safety functions unavailability matrix which is developed to identify structures, systems, and components significant for fire-related risk. The second method is a fire zones and key safety functions (KSFs) fire risk matrix which is useful to identify fire zones which are candidates for risk management actions. Specific selection and quantification methodologies have been developed to obtain the matrices. The Monte Carlo method has been used to assess the uncertainty of the unavailability matrix. The analysis shows that the uncertainty is sufficiently bounded. The significant fire-related risk is localized in six KSF representative components and one fire protection system which should be included in the maintenance rule. The unavailability of fire protection systems does not significantly affect the risk. The fire risk matrix identifies the fire zones that contribute the most to the fire-related risk. These zones belong to the control building and electric penetrations building.     

Some Remarks on the Endogeneous Pulsed Source Technique in Reactor Noise Analysis

It was established that the dynamics of an evaporator show markedly asymmetric responses. These phenomena are closely related to the response of the liquid level which directly affects the heat capacity.

What is more, the time constants of an evaporator are so large that the transfer functions expressing uranium product evaporator are better approximated by a form representing no self-regulation. Examination of the controllability aspects of three different control algorithms, i.e. control by boiling point raising, cascade control and multivariable control, resulted in the conclusion that the last-mentioned algorithm is superior to the two others for controlling the uranium concentration which is the most important element in uranium product evaporation.