Experience with identification of loose parts by acoustic monitoring of primary system

Substantial progress has been achieved in the identification of loose parts which had been detected by acoustic monitoring of reactor primary system. Several years of practical experience and the use of the offline digital analysis system MEDEA proved that acoustic monitoring is very successful for detecting component failures at an early stage. ISTec is involved in loose parts monitoring in several nuclear power plants in Germany. Advanced powerful tools for classification and evaluation of burst signals have been realised.

Loose parts monitoring systems, which are installed in all German nuclear power plants (NPPs), indicated specific impact conditions at lower plenum of two BWR's. Flow tests were carried out with various coolant flow rates of internal axial pumps and use of model nuts in one case. More than 2000 different bursts have been analysed to provide information in detail about impact occurrences, their spectral characteristics and impact sequences. Burst shape parameters could be determined and signal amplitudes have been trended. Determination of the sound origin — fixed origin in one case, flow-induced moving origin in the other case - and mass estimation of the loose parts could be performed by application of advanced burst analysis methods. Characteristics of the impact signals are presented in the paper.     

核电厂流出物中14C的管理控制

介绍了国内外压水堆(PWR)和重水堆(HWR)核电厂流出物中14C的产生和释放管理现状、减少14C产生和释放的方法以及14C的提取、净化和分析方法,为我国核电厂气态和液态流出物中14C的监测和控制提供基础资料.此外,针对我国核电厂14C的排放和监测情况,提出了几点建议.     

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.     

On-line neuro-expert monitoring system for Borssele Nuclear Power Plant

A new method for an on-line monitoring system for the nuclear power plants has been developed utilizing the neural networks and the expert system. The integration of them is expected to enhance a substantial potential of the functionality as operators support.

The recurrent neural network and the feed-forward neural network with adaptive learning are selected for the plant modeling and anomaly detection because of the high capability of modeling for dynamic behavior. The expert system is used as a decision agent, which works on the information space of both the neural networks and the human operators. The information of other sensory signals is also fed to the expert system, together with the outputs that the neural networks generate from the measured plant signals. The expert system can treat almost all known correlation between plant status patterns and operation modes as a priori set of rules.

From the off-line test at Borssele Nuclear Power Plant (PWR 480 MWe) in the Netherlands, it was shown that the neuro-expert system successfully monitored the plant status. The expert system worked satisfactorily in diagnosing the system status by using the outputs of the neural networks and a priori knowledge base from the PWR simulator. The electric power coefficient is simultaneously monitored from the measured reactive and active electric power signals.     

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。     

华龙一号LPD在线监测系统误差分析

核电厂采用线功率密度（LPD）在线监测系统对电站运行的实测参数在线计算并显示堆芯线功率密度,能够准确、及时地描述堆芯状态,提高核电厂运行的安全性和经济性。LPD在线监测系统报警限值的设定,需要考虑在线监测系统的总体误差,并留有一定裕量。通过研究华龙一号LPD在线监测系统的总体不确定度的分析方法,将系统的各部分误差通过统计方法综合起来,得到系统的总体误差限值。结果表明,华龙一号采用的LPD在线监测系统误差满足工程要求。      

Trends in computational intelligence in nuclear engineering

Several changes to the focus of Computational Intelligence in Nuclear Engineering have occurred in the past few years. With earlier activities focusing on the development of condition monitoring and diagnostic techniques for current nuclear power plants, recent activities have focused on the implementation of those methods and the development of methods for next generation plants and space reactors. These advanced techniques are expected to become increasingly important as current generation nuclear power plants have their licenses extended to 60 years and next generation reactors are being designed to operate for extended fuel cycles (up to 25 years), with less operator oversight, and especially for nuclear plants operating in severe environments such as space or ice-bound locations.     

论解决核电安全问题的途径

从现有水冷反应堆核电厂存在堆芯熔化危险这一安全问题的焦点出发,分析了改进型反应堆AP-600、SIR、非能动安全反应堆PIUS和具有固有安全的模块高温气冷堆MHTGR等的安全特性.按照下一代水冷反应堆的设计要求和用户要求,提出了解决水堆核电厂安全问题的新概念——自安全铀氢锆反应堆,该堆型可能成为世界水堆核电发展的一个方问。中国核动力研究设计院正在探讨这种堆型。     

Noise and DC balanced outlet temperature signals for monitoring coolant flow in LMFBR fuel elements

Local cooling disturbances in LMFBR fuel elements may have serious safety implications for the whole reactor core. They have to be detected reliably in an early stage of their formation therefore. This can be accomplished in principle by individual monitoring of the coolant flow rate or the coolant outlet temperature of the subassemblies with high precision.

In this paper a method is proposed to increase the sensitivity of outlet temperature signals to cooling disturbances. Using balanced temperature signals provides a means for eliminating the normal variations from the original signals which limit the sensitivity and speed of response to cooling disturbances. It is shown that a balanced signal can be derived easily from the original temperature signal by subtracting an inlet temperature and a neutron detector signal with appropriate time shift.

The method was tested with tape-recorded noise signals of the KNK I reactor at Karlsruhe. The experimental results confirm the theoretical predictions. A significant reduction of the uncertainty of measured outlet temperatures was achieved. This enables very sensitive and fast response monitoring of coolant flow.

Furthermore, it was found that minimizing the variance of the balanced signal offers the possibility for a rough determination of the heat transfer coefficient of the fuel rods during normal reactor operation at power.     

Performance of a transmutation advanced device for sustainable energy application

Preliminary studies have been performed to design a device for nuclear waste transmutation and hydrogen generation based on a gas-cooled pebble bed accelerator driven system, TADSEA (Transmutation Advanced Device for Sustainable Energy Application). In previous studies we have addressed the viability of an ADS Transmutation device that uses as fuel wastes from the existing LWR power plants, encapsulated in graphite in the form of pebble beds, cooled by helium which enables high temperatures (in the order of 1200 K), to generate hydrogen from water either by high temperature electrolysis or by thermochemical cycles. For designing this device several configurations were studied, including several reflectors thickness, to achieve the desired parameters, the transmutation of nuclear waste and the production of 100 MW of thermal power. In this paper new studies performed on deep burn in-core fuel management strategy for LWR waste are presented. The fuel cycle on TADSEA device has been analyzed based on both: driven and transmutation fuel that had been proposed by the General Atomic design of a gas turbine-modular helium reactor. The transmutation results of the three fuel management strategies, using driven, transmutation and standard LWR spent fuel were compared, and several parameters describing the neutron performance of TADSEA nuclear core as the fuel and moderator temperature reactivity coefficients and transmutation chain, are also presented.     

铀水栅零功率反应堆的功率刻度

反应堆功率的测量,在堆功率高时一般用热工方法,功率低时,可用各种堆物理方法,如中子源引进法、中子统计法和全堆总裂变率法。 中子源引进法误差较大,中子统计法需知探测器在堆内的效率和堆的β_(aff)值,此二者都较难测量。全堆总裂变率法是由测量堆的总裂变率来求得堆功率,它可避免前面两种方法的缺点,但需依赖裂变率相对分布的     

RAPID Operator-Free Fast Reactor Concept without Any Control Rods Reactor Concept and Plant Dynamics Analyses

The 1,000kWe metal fueled sodium-cooled fast reactor concept “RAPID” to achieve highly automated reactor operation has been demonstrated. RAPID (Refueling by All Pins Integrated Design) is designed for a terrestrial power system which enables quick and simplified refueling. It is one of the successors of the RAPID-L, the operator-free fast reactor concept designed for lunar base power system. The essential feature of the RAPID concept is that the reactor core consists of an integrated fuel assembly instead of conventional fuel subassemblies. In this small-size reactor core, 14,000 fuel pins are integrated and encased in a fuel cartridge. Refueling is conducted by replacing a fuel cartridge. The reactor can be operated without refueling for up to 10 years.

Unique challenges in reactivity control systems design have been addressed in the RAPID concept. The reactor has no control rod but involves the following innovative reactivity control systems: lithium expansion modules (LEM) for inherent reactivity feedback, lithium injection modules (LIM) for inherent ultimate shutdown, and lithium release modules (LRM) for automated reactor startup. All these systems adopt ⁶Li as a liquid poison instead of B₄C rods. In combination with LEMs, LIMs and LRMs, RAPID can be operated without an operator. In this paper, the RAPID reactor concept and its transient characteristics are presented.     

压水堆核电站大破口失水事故分析

压水堆核电站安全分析报告是核安全监管部门对其进行安全审查的重要文件,大破口失水事故是核电站运行的设计基准事故,是安全分析报告中的重要内容。本文使用RELAP5/MOD3.2进行压水堆冷管段大破口失水事故的计算,对比发现一回路冷管段发生双端断裂大破口时燃料元件包壳温度峰值（PCT）最高,且长时间维持在较高温度,此条件下反应堆最危险。计算结果表明,事故发生后,一回路压力迅速下降,堆芯冷却剂的流动性变差,导致堆芯裸露,燃料包壳温度又重新回升。通过安注系统和辅助给水系统等一系列动作,能保证燃料元件包壳温度不超过1204 ℃的限值。     

双面慢化环形燃料反应堆物理实验

环形燃料零功率反应堆是首个双面慢化环形燃料作为核燃料的反应堆。本文采用周期法、落棒法获取环形燃料零功率反应堆的临界参数、控制棒价值、元件价值、含Gd元件的反应性效应等关键参数,对环形燃料零功率反应堆的物理性能进行实验研究,验证环形燃料反应堆堆芯物理设计计算程序。结果表明：根据外推过程确定堆芯临界装载环形燃料元件96根,实心燃料元件172根,此时k_eff为1.000 40,堆芯调节棒价值为-247.5 pcm,安全棒价值为-1 358.4 pcm;元件价值与理论值平均偏差为1.3 pcm,含Gd元件反应性效应与理论值平均相对偏差为8.8%。本文结果为环形燃料的工程化设计程序提供关键数据支撑。     

Control rod drop hydrodynamic analysis for a pressurized water reactor

During operation of nuclear power reactors, reactivity initiated accidents can take place such as a control rod drop. If this occurs, the reactivity increases significantly and leads to an enhancement in power, fuel temperature and damage of reactor eventually. Exact assessment of these accidents depends on the hydrodynamic information. In this research, it is tried to simulate the unsteady flow field around the control rod for a pressurized water reactor power plant. In order to simulate the flow field around the control rod inside the guide tube, averaged Navier–Stokes equations accompanied by the layering dynamic mesh strategy have been used. The information exchange between the two computational stationary and moving grids, the computational grid around the control rod and the grid next to the guide tube, has been taken place through the interface. It was concluded that the time duration of control rod to reach the bottom of the core depends on the leakage. It was also observed that the velocity and acceleration of the control rod would be reduced by decreasing leakage flow rate and in certain leakages, the acceleration of the control rod approaches zero due to equilibrium conditions. During this research, a correlation based on the achieved data was proposed which would provide useful information on the relation between the leakage and the time for control rod to reach the bottom of the core.     

MOX燃料在轻水堆核电站中的应用

目前MOX燃料已成为一种可用于轻水堆核电站成熟的核燃料。简要介绍了国外该领域的发展状况以及MOX燃料对反应堆性能的主要影响和应对措施。探讨了MOX燃料在国内压水堆核电站中的应用问题。     

Consequence analysis for nuclear reactors,Yongbyon

Since the Fukushima nuclear power plant accidents in 2011, there have been an increased public anxiety about the safety of nuclear power plants in Korea. The lack of safeguards and facility aging issues at the Yongbyon nuclear facilities have increased doubts. In this study, the consequence analysis for the 5-MWe graphite-moderated reactor in North Korea was performed. Various accident scenarios including accidents at the interim spent fuel pool in the 5-MWe reactor have been developed and evaluated quantitatively. Since data on the design and safety system of nuclear facilities are currently insufficient, the release fractions were set by applying the alternative source terms made for utilization in the analysis of a severe accident by integrating the results of studies of severe accidents occurred before. The calculation results show the early fatality zero deaths and latent cancer fatality about only 13 deaths in Seoul. Thus, actual impacts of a radiological release will be psychological in terms of downwind perceptions and anxiety on the part of potentially exposed populations. Even considering the simultaneous accident occurrence in both 5-MWe graphite-moderated reactor and 100-MWt light water reactor, the consequence analysis using the MACCS2 code shows no significant damage to people in South Korea.     

微型反应堆裂变率分布实验研究

利用固体径迹探测器测量处于微型反应堆不同益的燃料元件内单位体积的裂变率，得到了堆的裂变经分布和总裂变率，并与其它参数相结合求得了反应堆功率。同时，测量了对应功率下反应堆内辐照座的热中子通量密度得到单位功率的热中子能量密度，即额定中子能量密度下的运行功率。文章给出的测量方法，避免了金箔法测量反应堆功率所引入的近似假设。     

Transport of MOX fuel: continuous challenge

Abstract

For 45 years TN International has been involved in the radioactive materials transportation field. Since the beginning the spent nuclear fuel transportation has been its core business. During all these years TN International, now part of AREVA, has been able to anticipate and fulfil the needs for new transport or storage casks design to fit the nuclear industry evolutions. A whole fleet of casks able to transport all the materials of the nuclear fuel cycle has been developed. This paper focuses on the casks used to transport the fresh and used mix oxide (MOX) fuel. To transport the fresh MOX boiling water reactor and pressurised water reactors fuel, TN International has developed two designs of casks: the MX 6 and the MX 8. These casks are and have been used to transport MOX fuel for French, German, Swiss and in a near future Japanese nuclear power plants. A complete set of baskets have been developed to optimise the loading in terms of integrated dose and also of course capacity. Mixed oxide used fuel has now its dedicated cask: the TN 112 which certificate of approval has been obtained in July 2008. This cask is able to transport 12 MOX spent fuel elements with a short cooling time. The first loading of the cask has been performed in September 2008 in the Electricité de France nuclear power plant of Saint-Laurent-des-Eaux. By its continuous involvement in the nuclear transportation field, TN International has been able to face the many challenges linked to the radioactive materials transportation especially talking of MOX fuel. TN International will also have to face the increasing demand linked to the nuclear renaissance.