Instrument failure detection and reconstruction methodology inspace-time nuclear reactor dynamic systems with fixed in-core detectors

The detection and isolation of instrument failures in nuclear reactors equipped with fixed in-core detectors were studied in order to improve reliability and safety. This was done by representing the reactor as a linear stochastic distributed parameter system. A bank of detection observers based on the Kalman filter concept was constructed in order to isolate component failures via robust observation. Each observer was sensitive to only one specified component failure. This was done to minimize the covariance matrix error. However, because observed deviations may be attributed to changes in system behavior, failure decisions are confirmed by a multiple consecutive miscomparison (MCM) counter. Following failure detection, the failed sensor's output is replaced with an estimate from the failure-free filter. Various simulations were performed to verify this failure detection method as applied to reactor instrument failures. It was demonstrated that the method can detect both single- and common-mode failures. Also, without hardware redundancy, it can describe the system dynamics in the event of failures     

智能化反应堆物理启动装置的数据采集板的设计

介绍了智能化反应堆启动装置的关键组成部分数据采集板的研制。重点是以μPSD3234A为核心器件的硬件设计,以及以USB通讯接口为核心的固件设计。     

A cost-effective methodology to internalize nuclear safety in nuclear reactor conceptual design

A new methodology to perform nuclear reactor design, balancing safety and economics at the conceptual engineering stage, is presented in this work. The goal of this integral methodology is to take into account safety aspects in an optimization design process where the design variables are balanced in order to obtain a better figure of merit related with reactor economic performance. Design parameter effects on characteristic or critical safety variables, chosen from reactor behavior during accidents (safety performance indicators), are synthesized on Design Maps. These maps allow one to compare the safety indicator with limits, which are determined by design criteria or regulations, and to transfer these restrictions to the design parameters. In this way, reactor dynamic response and other safety aspects are integrated in a global optimization process, by means of additional rules to the neutronic, thermal-hydraulic, and mechanical calculations.An application of the methodology, implemented in Integrated Reactor Evaluation Program 3 (IREP3) code, to optimize safety systems of CAREM prototype is presented. It consists in balancing the designs of the Emergency Injection System (EIS), the Residual Heat Removal System (RHRS), the primary circuit water inventory and the containment height, to cope with loss of coolant and loss of heat sink (LOHS) accidental sequences, taking into account cost and reactor performance.This methodology turns out to be promising to internalize cost-efficiently safety issues. It also allows one to evaluate the incremental costs of implementing higher safety levels.     

Rationale for nuclear reactor control room data displays

One of the ‘lessons learned’ from the Three Mile Island accident focuses upon the need for a validated source of plant-status information in the control room. The utilization of computer-generated graphics to display the reduced readings of the plant instrumentation has introduced the need for a set of guidelines that focuses upon the mental image of plant conditions that the operators receive from viewing the computer monitors. The principles that govern the design of displays are similar to those employed in the education process because the objective is the same, namely, to transfer information.The philosophy for the development of displays to portray the status of major plant systems bases the level of detail upon the needs of the user. Graphic displays that relate the plant system parameter to the plant systems are recommended along with bar-chart type representations of the relationship between plant parameters and allowable limits.     

Computational fluid dynamic investigations of partial blockage detection by core-temperature monitoring system of a sodium cooled fast reactor

Thermal hydraulic studies have been carried out to understand temperature dilution suffered by core-temperature monitoring system of a sodium cooled fast reactor. The three-dimensional computational model is validated against experimental results of a water model. Jet mixing phenomenon as predicted by different turbulence models is compared and RNG k–? model is found to be better than other models. A comprehensive parametric study considering: (i) effects of construction/manufacturing tolerances on thermocouple positions with respect to subassembly positions, (ii) thermal/irradiation bowing of subassemblies, and (iii) changes in core power profile during reactor operation cycles has been carried out. The studies indicate the maximum possible dilution in fuel and blanket subassemblies to be 2.63 K and 46.84 K, respectively. Shifting of thermocouple positions radially outward by 20 mm with respect to subassembly centers leads to an overall improvement in accuracy of thermocouple readings. It is also seen that subassembly blockage that leads to 7% flow reduction in fuel subassembly and 12% flow reduction in blanket subassembly can be detected effectively by the core-temperature monitoring system.     

用数据处理标定核数据采集系统

利用计算机程序拟合数据采集系统获取的数据,改变拟合参数,找出最佳拟合值,从而标定该数据采集系统,所得标定值的系统误差最小。     

基于核电四极矩共振技术的爆炸物检测系统的数据处理及信号识别算法

分析了核电四极矩共振(NQR)信号的特点,利用傅里叶频谱分析对NQR信号进行处理的信噪比及其随测量时间的变化曲线,在此基础上,针对NQR信号提出一种具有更高信噪比的频谱分析方法(加权傅里叶频谱分析),同时提出一种可以有效区分NQR信号和正弦干扰信号的频谱差值分析方法。     

Optimal power control system of a research nuclear reactor

The power control system is a key control system for a nuclear reactor, which directly concerns the safe operation of a nuclear reactor. Much attention is paid to the power control system performance of nuclear reactor in engineering. The designers put a high value upon design of an optimal power control system. In this paper, a design method is applied to the design of power control system. According to the optimal control theory, an objective function, quadratic performance index with weight factors is proposed. Then, the objective function is transformed into frequency domain form by use of Paserval's theorem. In frequency domain, an optimal transfer function can be obtained at the lowest value of objective function. The system with optimal transfer function has an optimal performance. The transfer function of the power control system is derived from a typical research nuclear reactor. Using the state feedback theory, the transfer function is synthesized to the optimal transfer function. The simulative results with the optimal controller and with a conventional controller show that the performance of the optimal power control system is largely improved on dynamic characters. The method applied here not only can be used for research nuclear reactor but also can be easily extended to pressurized water reactor power plant and other fields.     

ADS加速器失束次临界反应堆动态特性研究

加速器驱动系统(ADS)中次临界堆芯的功率水平依靠强流质子轰击散裂靶产生的中子源来维持.加速器较为频繁的失柬问题,必将对ADS次临界反应堆安全性产生影响.研究了ADS系统失束事故特性,设计开发出具有较强针对性的用于ADS失束事故分析软件,对加速器驱动快中子次临界反应堆的动态响应开展了初步研究.结论表明仅靠断束停堆,仍有可能危及次临界反应堆的安全性.建议增设辅助停堆保护系统以提高ADS安全性.     

反应堆控制保护系统监测参数校准

如何开展反应堆控制保护系统测量设备的现场校准工作,特别是核测系统的2个非标设备-功率测量保护装置和周期测量保护装置,输入信号跨8个数量级,输出周期信号又按非线性刻度。由于信号在测量、转换和传输过程中引入的误差,使得同一个信号在保护系统和计算机系统的显示值不一致。作者根据研究堆调试阶段核测设备校准的经验,提出了核测系统测量设备的校准方法。核测功保输出的预报警信号向保护系统测量结果修正,这样能保证预报警触发信号与保护触发信号之比为1.05∶1.1的关系。保护系统校准周期保护触发动作点,保证堆功率增长周期为10 s时保护系统送出事故信号。校准结果不仅满足了保护系统触发精度的要求,也使同一个监测参数在不同位置的示值较为一致。     

反应堆控制棒驱动机构模拟负载系统的研制

介绍了反应堆控制棒驱动机构(CRDM)模拟负载装置的设计原理和方法,研制出了一种新型的模拟负载系统,用来模拟反应堆棒控系统对控制棒的控制过程。对设计的模拟负载系统进行了功能性试验和性能参数测试,并与实际运行系统进行比较后,发现该系统达到了各项功能控制要求,且性能稳定可靠,模拟负载的电磁线圈散热性能与负载特性良好,各项性能指标达到了设计要求。     

核医学数据获取动态存储器的设计

介绍了核医学图像数据获取的基本方法和与计算机的接口,重点介绍接口中数据获取动态存储器的设计,给出了硬件电路和设计原理。     

多群核数据不确定性对堆芯物理计算的影响

核数据不确定性是造成反应堆物理计算结果不确定性的重要因素之一。基于所需抽样核数据的协方差矩阵开发了随机抽样模块(Stochastic Sampling,SAMP),在此基础上利用SCALE(Standardized Computer Analyses for Licensing Evaluation)软件包实现了混合法和随机抽样法两种不确定性分析方法,以研究多群核数据不确定性对堆芯物理计算的影响。以3×3假想堆芯为对象,对两种方法进行了验证,然后应用于国际原子能机构(International Atomic Energy Agency,IAEA)燃料管理基准题中的Almaraz核电厂首循环堆芯。分析结果表明,两种方法结果符合良好,Almaraz核电厂堆芯keff不确定性约为0.5%,堆芯径向和轴向功率的最大不确定性分别为1.9%和0.45%。     

A cascade control strategy for a space nuclear reactor system

This work presents a linear feedback control for the space nuclear reactor power system TOPAZ II. The point-kinetics approximation with six-delayed-neutron-group is used to represent the neutron field dynamics. A favorable choice of input control variables is demostrated, which leads to a cascade control configuration with two classical either PI or P controllers. The strategy is based on linearizing-like feedback control endowed with a modeling error estimator via a reduced order-observer. The effectiveness of the control law to the tracking of a given thermal power profile in the start-up regime and the tracking of a given electric power profile in the operation regime are illustrated via numerical simulations.     

反中微子监测反应堆运行的数值模拟

利用反应堆出射反中微子计数监测反应堆运行,是国际上新兴的防扩散监测技术,已经过了实验检验。为研究该方法监测反应堆的能力,我们通过在现有MCNP5和MCORGS数值模拟软件中增加了蒙特卡罗方法模拟出射反中微子数目、能量和方向等信息的功能,开发出了用于模拟探测反应堆运行时出射反中微子的数值模拟软件。利用该软件我们研究了反应堆燃耗与出射反中微子计数关系、不同燃耗下铀和钚材料同位素比与出射反中微子计数关系、不同反应堆运行和换料条件下出射中子随反应堆运行时间的变化规律等问题。数值模拟结果表明,反应堆出射反中微子计数可以提供与反应堆运行情况相关的信息。     

Improved performances of the fast reactor calculational system ERANOS-ERALIB1 due to improved a priori nuclear data and consideration of additional specific integral data

A single consistent scheme of calculational methods and nuclear data called ERANOS-ERALIB1 was produced in 1996 to calculate fast reactor neutronic parameters. It represents a significant improvement on previous schemes such as CARNAVAL-IV, PROPANE and VASCO, each of which were required in order to calculate one specific application. The nuclear data library ERALIB1 has been obtained by a consistent statistical adjustment based on 355 integral data from 71 different systems. The performance of ERALIB1 is excellent, as demonstrated during its validation for which all the k_eff SUPER-PHENIX data were reproduced to within 70 pcm.

The only restriction on this satisfactory performance is related to the rather poor prediction of the sodium void reactivity effect. This was due to very bad nuclear data for ²³Na, and the unsatisfactory methods used to calculate the sensitivity coefficients for the sodium void reactivity variation Δρ_Na. To improve the performance relative to this point and to enlarge the domain of validation several actions have been undertaken:

•a revision of the formalism and algorithms used to calculate the derivatives of Δρ_Na to the sodium cross section data,

•a significant enlargement of the integral data base related to this aspect of the sodium void effect. Compared to the initial data base established in support of ERALIB1, several additional (18) sodium void configurations corresponding to voids of different volumes at different core locations have been studied.

In order to broaden the range of application of the improved library, which will be called ERALIB1.A, significant effort has been devoted to additional configurations which have firstly been evaluated, and then if judged suitable, included in the adjustment process. They are related to two specifically targeted experimental programmes:

•a study of neutron deep penetration. Several configurations of the JANUS experimental programme (shielding constructed of separate steel, iron and sodium plates) have been analysed. With this complementary information ERALIB 1A becomes applicable for accurate predictions of shielding configurations,

•a study of steel reflectors for a fast reactor of the SUPER-PHENIX type. The measurements performed in the MASURCA (the CIRANO experimental programme) and FCA facilities include spectral indices (F25(r)/F25(0),...) at different positions in the reflector. As a consequence of these measurements, important information has been obtained for additional “secondary” structural material isotopes, such as ⁵⁷Fe, ⁶⁰Ni and ⁵³Cr.

Significant effort has also been devoted to the analysis of 29 β_eff experiments. The result of this is an improvement of the uncertainty on ν_d(E) which guarantees a prediction of β_eff with the required accuracy (3% for critical configurations, and 5% for power reactors).

The consistent statistical adjustment method by Gandini et al. (1973) has been completed. Rigorous criteria have been introduced to identify any data which are suspect and/or inconsistent in the integral data base. These data may introduce additional bias in the adjusted library, and for that reason they must be discarded before adjustment.