A modernized and versatile startup reactivity measuring system installed at NPP Paks and its application for subcritical systems

In 2004 the Hungarian Paks NPP completed a project for upgrading the reactivity measuring system applied during reactor startup experiments. Almost all components of the previous system were replaced, only ex-core ionisation chambers remained unaltered. New hardware and software components were introduced for neutron flux signal handling, for data acquisition, as well as for measurement evaluation and data presentation. High-precision picoamper meters were installed at each reactor unit, current signals are handled by a portable signal processing unit. The system applies an accurate on-line reactivity calculation algorithm based on the point-kinetic model with six delayed neutron groups. Detailed off-line evaluation and analysis of startup measurements can be performed on the portable unit, as well.The paper describes the architecture, data acquisition modules, services and man–machine interface of the new system. Functions and results are illustrated with measured data recorded during a startup of Unit 3. In 2003 and 2004 the RMR was installed and tested at all Paks NPP units successfully and now it is in regular use during unit startups.The second part of the paper illustrates an extension of the new system to perform reactivity measurements using the well-known Rossi-α and Feynman-α statistical methods. The modified system was needed to estimate the reactivity of a subcritical system formed by damaged fuel assemblies stored at the fuel service pit of Paks Unit 2. Theoretical background of the applied algorithms is outlined, then results of validation tests and on site measurements are treated. The measurements have shown that the subcriticality of the damaged fuel was sufficiently deep if the high boron concentration in the fuel service pit was maintained.     

CFD evaluation of hydrogen risk mitigation measures in a VVER-440/213 containment

In the PHARE project “Hydrogen Management for the VVER440/213” (HU2002/000-632-04-01), CFD (Computational Fluid Dynamics) calculations using GASFLOW, FLUENT and CFX were performed for the Paks NPP (Nuclear Power Plant), modelling a defined severe accident scenario which involves the release of hydrogen. The purpose of this work is to demonstrate that CFD codes can be used to model gas movement inside a containment during a severe accident. With growing experience in performing such analyses, the results encourage the use of CFD in assessing the risk of losing containment integrity as a result of hydrogen deflagrations. As an effective mitigation measure in such a situation, the implementation of catalytic recombiners is planned in the Paks NPP. In order to support these plans both unmitigated and recombiner-mitigated simulations were performed. These are described and selected results are compared. The codes CFX and FLUENT needed refinement to their models of wall and bulk steam condensation in order to be able to fully simulate the severe accident under consideration.Several CFD codes were used in parallel to model the same accident scenario in order to reduce uncertainties in the results.Previously it was considered impractical to use CFD codes to simulate a full containment subject to a severe accident extending over many hours. This was because of the expected prohibitive computing times and missing physical capabilities of the codes. This work demonstrates that, because of developments in the capabilities of CFD codes and improvements in computer power, these calculations have now become feasible.     

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.     

北京正负电子对撞机输运线磁铁电源控制系统

在北京正负电子对撞机重大改造工程中,新的控制系统将采用系统集成工具包EPICS进行开发.描述了在EPICS环境下开发的输运线磁铁电源控制系统,包括系统结构、控制软件和人机界面等.该系统已经正式投入运行,性能稳定可靠,为对撞机改造项目做出了贡献.     

核反应堆功率调节系统控制特性研究

某新型研究堆采用新型燃料元件和堆芯结构，反应堆的控制特性缺乏基本数据，功率调节系统的设计无可借鉴经验。通过半实物仿真试验，在同一功率定值下分别引入阶跃和斜坡反应性扰动，考虑调节棒在不同位置的价值影响，采用PD控制方案研究功率调节系统的调节特性和控制效果，并对控制方案和PD参数进行比较和优化，为反应堆功率调节系统的设计和投入运行奠定基础。半实物仿真试验结果表明：采用同一组控制器参数，无法满足预定的控制要求。充分发挥数字化控制器的优点，在同一功率定值、不同棒位下，采用不同的控制器参数能较好地满足预定的控制指标和性能要求。     

多功能网络化核数据获取和处理系统

应用网络技术将多个功能不同、相互独立的以PC微机为基础组成的核数据获取和处理子系统有机互联,组成了一种新型的多功能网络化核数据获取和处理系统,探讨了系统的最佳组成模式、工作原理及实现方法,给出了系统在2．5MeV静电加速器上进行团簇特性研究的初步结果。     

Analysis of low load transients for the Vandellòs-II NPP: Application to operation and control support

The Technical University of Catalonia (UPC) has been jointly working with the Asociación Nuclear Ascó-Vandellòs (ANAV) for a number of years in order to establish, qualify and use best estimate (BE) models for the reactors under the control of ANAV. ANAV is the consortium that is responsible for operation of the Ascó and Vandellòs-II reactors. The reactors are Westinghouse-design three-loop PWRs with an approximate electrical power of 1000 MW. The existing integral plant models for each plant are currently used for many different purposes among which are support of plant operation and control. Quite a number of studies have been done in order to improve both safety and plant competitiveness. Most of these dynamic analyses were carried out in relation to transients starting at nominal full power or at least, very close to full power. This paper develops a specific use of the Vandellòs-II plant model for operation and control support at low power involving new ranges of system actuation parameters. It also examines scenarios that are somewhat different from those typically analysed. The study starts showing the results of an assessment case, which is a start-up test and provides some additional qualification, and subsequently attempts to establish calculations to support both an improvement in feed water controls and to set up operating recommendations for low-load manual operation of feed water turbo-pump. Both results hopefully, will produce an outcome, which leads to an improvement in safety and reduces reactor trip probability.     

数值反应堆原型系统开发及示范应用研究进展

数值反应堆是当前核反应堆工程与技术领域国内外研发热点之一。本文介绍了国家重点研发计划“数值反应堆原型系统开发及示范应用”重点专项总体研发进展，包括数值反应堆原型系统CVR1.0、核心软件开发与验证、多物理耦合进展，以及数值反应堆原型系统初步示范应用。     

Recent Development of Data Acquisition System on HL-2A

This paper introduced recent development of data acquisition system(DAS) on the HL-2A tokamak. The existing DAS has to be remodeled because of the evident improvement on HL-2A itself and the increasing amount of corresponding experimental data. First, the experimental network on HL-2A Tokamak as a communication center is described in detail. Secondly, "The File's Format & Criterion of Experimental Data on HL-2A Tokamak" as the core of the new software is also presented. Thirdly, not only a multi-screen image display system on HL-2A is discussed, but also the build-up of a new display wall system is dealt with. Finally, on the graphical WINDOWS platform and on the basis of the new Data File's Format & Criterion, the new software on HL-2A is described, including a new multi-layer database management program, a real-time processing program, a common interactive analytic processing program and a data post-processing program, which is now under development.     

Thermal hydraulic behaviors during loss of RHR system at mid-loop operation of Chinese 300 MWe PWR NPP

Thermal hydraulic behavior of nuclear power plant (NPP) is analyzed by using mechanistic computer code for loss of residual heat removal (RHR) system during mid-loop operation of Chinese 300 MWe two-loop pressurized water reactor is presented. In the absence of recovery of RHR or other accident management measures, the reactor core will be uncovered for a long term resulting in core heat-up, degradation and relocation to the lower plenum. The effectiveness of available mitigate measures, such as safety injection system, gravity feed from refueling water storage tank (RWST) and steam generator (SG) reflux-condensation, are investigated. Coolant injection is highly effective in halting the accident progression and make the core recovered. The cooling capability of SG reflux-condensation has a relationship with different availabilities of steam generators and decay heat power. 6 days after shutdown, 2SG operation can keep the water level at mid-line of hot leg. 12 days after shutdown, both 2SG operation and 1SG operation can keep the water level at mid-line of hot leg. The analyses also indicate that the cooling mechanism of safety injection system is more effective than gravity feed from RWST and SG reflux-condensation. Through confirming the success criteria of SG reflux-condensation, time windows can be devided. Then, event trees for loss of RHR system under mid-loop operation are built with considering the analysis results and abnormal procedure.     

Variations of a passive safety containment for a BWR with active and passive safety systems

The paper presents variations of a certain passive safety containment for a near future BWR. It is tentatively named Mark S containment in the paper. It uses the operating dome as the upper secondary containment vessel (USCV) to where the pressure of the primary containment vessel (PCV) can be released through the upper vent pipes. One of the merits of the Mark S containment is very low peak pressure at severe accidents without venting the containment atmosphere to the environment. Another merit is the capability to submerge the PCV and the reactor pressure vessel (RPV) above the core level by flooding water from the gravity-driven cooling system (GDCS) pool and the upper pool. The third merit is robustness against external events such as a large commercial airplane crash owing to the reinforced concrete USCV. The Mark S containment is applicable to a large reactor that generates 1830 MW electric power. The paper presents several examples of BWRs that use the Mark S containment. In those examples active safety systems and passive safety systems function independently and constitute in-depth hybrid safety (IDHS). The concept of the IDHS is also presented in the paper.     

Gyrotron and power supply development for upgrading the electron cyclotron heating system on DIII-D

An upgrade of the electron cyclotron heating system on DIII-D to almost 15 MW is being planned which will expand it from a system with six 1 MW 110 GHz gyrotrons to one with ten gyrotrons. A depressed collector 1.2 MW 110 GHz gyrotron is being commissioned as the seventh gyrotron. A new 117.5 GHz 1.5 MW depressed collector gyrotron has been designed, and the first article will be the eighth gyrotron. Two more are planned, increasing the system to ten total gyrotrons, and the existing 1 MW gyrotrons will subsequently be replaced with 1.5 MW gyrotrons.Communications and Power Industries completed the design of the 117.5 GHz gyrotron, and are now fabricating the first article. The design was optimized for a nominal 1.5 MW at a beam voltage of 105 kV, collector potential depression of 30 kV, and beam current of 50 A, but can achieve 1.8 MW at 60 A. The design of the collector permits modulation above 100 Hz by either the body or the cathode power supply, or both, while modulation below 100 Hz must use only the cathode power supply.General Atomics is developing solid-state power supplies for this upgrade: a solid-state modulator for the cathode power supply and a linear high voltage amplifier for the body power supply. The solid-state modulator has series-connected insulated-gate bipolar transistors that are switched at a fixed frequency by a pulse-width modulation regulator to control the output voltage. The design of the linear high voltage amplifier has series-connected transistors to control the output voltage, which was successfully demonstrated in a proof-of-principle test at 2 kV. The designs of complete power supplies are progressing.The design features of the 117.5 GHz 1.5 MW gyrotron and the solid-state cathode and body power supplies will be described and the current status and plans are presented.     

新型热管反应堆堆芯热工安全分析

针对新型空间热管反应堆,采用商用CFD软件FLUENT对其堆芯进行了稳态热工安全分析。根据MCNP物理计算的堆芯功率分布,选取功率份额最高的相邻3个燃料元件作为分析对象,对控制转鼓7种不同转动角度下的正常工况以及单根热管失效的事故工况进行计算分析,得到最热通道各层材料的温度分布。采用二维热管分析程序计算得到蒸汽区的温度分布,并作为三维计算模型的温度边界。堆芯功率分布采用用户自定义程序UDF进行添加。计算结果表明,在额定功率4.0 MW水平下,在正常工况以及单根热管失效事故工况下,热管具有足够的传热能力将堆芯裂变热导出,同时,堆芯最热通道各层材料温度均低于安全限值,且具有较大的安全裕度,满足设计要求。     

工业数据采集和智能数据处理系统在岭澳核电站的应用

对岭澳核电站1、2号机组采用的KIT工业过程数据采集系统的功能和应用程序接口软件进行了分析研究。介绍了自主开发的基于Web的实时数据采集和在线数据发布监控系统,该系统实现了实时采集工业数据并上网发布,对生产过程进行远程监控和仪器、仪表的故障诊断等功能另外还讨论了基于神经网络技术的智能数据处理系统在核电站发电功率预测中的应用。     

固态燃料熔盐堆自适应功率控制器设计及分析

钍基熔盐堆核能系统(Thorium-based Molten Salt Reactor,TMSR)是中国科学院首批启动实施的战略性先导科技专项,旨在研发第四代反应堆核能系统。固态燃料钍基熔盐实验堆(The Solid Fuel Thorium-based Molten Salt Experimental Reactor,TMSR-SF1)是一个10 MW热功率的氟盐冷却球床堆,目前已经完成方案设计和初步工程设计。功率控制系统是反应堆一个关键控制系统,实现反应堆正常启动、功率运行和正常停堆功能,对保证反应堆安全和稳定运行起着极其重要的作用。根据TMSR-SF1运行控制要求,结合自适应控制理论,基于Lyapunov稳定性理论设计了一种TMSR-SF1模型参考自适应功率控制器。基于TMSR仿真平台,使用MATLAB/Simulink建立了自适应功率控制系统模型,并开展了控制器特性分析。结果表明,自适应功率控制器具备良好的负荷跟随能力,抗干扰能力强、稳定性好、可靠性高,能够满足TMSR-SF1功率控制的要求,确保堆芯的输出功率与功率设定值相匹配。     

Safety analysis of the IAEA reference research reactor during loss of flow accident using the code MERSAT

Using the thermal hydraulic code MERSAT detailed model including primary and secondary loop was developed for the IAEA's reference research reactor MTR 10 MW. The developed model enables the simulation of expected neutronic and thermal hydraulic phenomena during normal operation, reactivity and loss of flow accidents.Two different loss of flow accident (LOFA) have been simulated using slow and fast decrease time of core mass flow. In both cases the expected flow reversal from downward forced to upward natural circulation has been successfully simulated. The results indicate that in both accidents the limit of onset of subcooled boiling was not arrived and consequently no exceed of design limits in term of thermal hydraulic instability or DNB is observed. Finally, the simulation results show good agreement with previous international benchmark analyses accomplished with other qualified channel and thermal hydraulic system codes.     

Safety analysis of the reference research reactor MTR during reactivity insertion accident using the code MERSAT

The IAEA’s reference research reactor MTR-10 MW has been modeled using the code MERSAT. The developed MERSAT model consists of detailed representation of primary and secondary loops including reactor pool, bypass, main pump, heat exchanger and reactor core with the corresponding neutronics and thermalhydraulic characteristics. Following the successful accomplishment of the steady state operation at nominal power of 10 MW, reactivity insertion accident (RIA) for three different initial reactivity values of $1.5/0.5 s, $1.35/0.5 s and $0.1/1.0 s have been simulated. The predicted peaks of reactor power, hot channel fuel, clad and coolant temperatures demonstrate inherent safety features of the reference MTR reactor. Only in case of the fast RIA of $1.5/0.5 s, where the peak power of 133.66 MW arrived 0.625 s after the start of the transient, the maximum hot channel clad temperature arrives at the condition of subcooled boiling with the subsequent void formation. However, due to the strong negative reactivity feedback effects of coolant and fuel temperatures the void formation persists for a very short time so that thermalhydraulic conditions remained far from exceeding the safety design limits of thermalhydraulic instability and DNB. Finally, the simulation results show good agreement with previous international benchmark analyses accomplished with other qualified channel and thermalhydraulic system codes.     

Model extension and improvement for simulator-based software safety analysis

One of the major concerns when employing digital I&C system in nuclear power plant is digital system may introduce new failure mode, which differs with previous analog I&C system. Various techniques are under developing to analyze the hazard originated from software faults in digital systems. Preliminary hazard analysis, failure modes and effects analysis, and fault tree analysis are the most extensive used techniques. However, these techniques are static analysis methods, cannot perform dynamic analysis and the interactions among systems. This research utilizes “simulator/plant model testing” technique classified in (IEEE Std 7-4.3.2-2003, 2003. IEEE Standard for Digital Computers in Safety Systems of Nuclear Power Generating Stations) to identify hazards which might be induced by nuclear I&C software defects. The recirculation flow system, control rod system, feedwater system, steam line model, dynamic power-core flow map, and related control systems of PCTran–ABWR model were successfully extended and improved. The benchmark against ABWR SAR proves this modified model is capable to accomplish dynamic system level software safety analysis and better than the static methods. This improved plant simulation can then futher be applied to hazard analysis for operator/digital I&C interface interaction failure study, and the hardware-in-the-loop fault injection study.     

An evaluation of reactor cooling and coupled hydrogen production processes using the modular helium reactor

The high-temperature characteristics of the modular helium reactor (MHR) make it a strong candidate for producing hydrogen using either thermochemical or high-temperature electrolysis (HTE) processes. Using heat from the MHR to drive a sulfur-iodine (SI) thermochemical hydrogen production process has been the subject of a U.S. Department of Energy sponsored Nuclear Engineering Research Initiative (NERI) project led by General Atomics, with participation from the Idaho National Laboratory (INL) and Texas A&M University. While the focus of much of the initial work was on the SI thermochemical production of hydrogen, recent activities included development of a preconceptual design for an integral HTE hydrogen production plant driven by the process heat and electricity produced by a 600 MW MHR.This paper describes ATHENA analyses performed to evaluate alternative primary system cooling configurations for the MHR to minimize peak reactor vessel and core temperatures while achieving core helium outlet temperatures in the range of 900–1000 °C that are needed for the efficient production of hydrogen using either the SI or HTE process. The cooling schemes investigated are intended to ensure peak fuel temperatures do not exceed specified limits under normal or transient upset conditions, and that reactor vessel temperatures do not exceed American Society of Mechanical Engineers (ASME) code limits for steady-state or transient conditions using standard light water reactor vessel materials. Preconceptual designs for SI and HTE hydrogen production plants driven by one or more 600 MW MHRs at helium outlet temperatures in the range of 900–1000 °C are described and compared. An initial SAPHIRE model to evaluate the reliability, maintainability, and availability of the SI hydrogen production plant is also described. Finally, a preliminary flowsheet for a conceptual design of an HTE hydrogen production plant coupled to a 600 MW modular helium reactor is presented and discussed.