核电厂燃料包壳破损情况下反应堆停堆过程中水化学监测与控制

燃料包壳破损情况下反应堆停堆过程水化学监测与控制,对核安全、降低源项、减少人员照射剂量、提高换料大修经济效益有重要意义。本文简述了反应堆停堆过程水化学监测与控制方法,通过宁德核电厂燃料包壳破损情况下,首次大修停堆过程水化学监控的实践效果分析,并对此次反应堆停堆过程中遇到异常现象进行分析,提出了解决的建议。     

水化学革命正在减少职业照射

【英国《国际核工程》1990年2月号第27页报道】英国核能学会最近组织了一次水化学会议,会议着重讨论了反应堆水化学革命的最新动向。用水化学方法可以在反应堆水循环系统中防止辐射场的形成,并减少设备的腐蚀损坏;因此,在修理和检查时     

根据水化学和职业辐射照射剂量进行先进压水堆材料选择

一回路材料选择准则是根据降低反应堆运行期间及完成使用寿命以后的职业照射剂量和形成的放射性废物量来评定的，原因分析涉及到最佳一回路水化学环境选择、材料的耐腐蚀性及一回路内表面预处理几个方面。捷克斯洛伐克就一回路材料成份对核电站辐射情况的影响进行了讨论，并将实验结果与对腐蚀产物在一回路的迁移建立的数学模型得出的计算结果进行了对比。     

反应堆结构的磨蚀研究现状

蒸汽发生器传热管与支承结构、燃料棒包壳与格架等反应堆结构在高温高压水环境下,由于流致振动导致的微动磨蚀破坏是影响反应堆安全运行的重要因素之一。温度、水化学条件是影响结构磨蚀特性的关键参数,文中对各国的磨蚀研究现状进行了介绍。     

西安脉冲堆实验装置参数测试

反应堆各实验孔道的参数是反应堆的重要指标，也是反应堆应用的重要参数。在简单介绍西安脉冲反应堆参数测量方法、测量设计的基础上，详细给出了该堆上一些实验孔道的参数测试结果，为在该堆的应用工作提供参考。     

日本超临界压水堆的研究现状

在2002年第四代堆（Gen-Ⅳ）国际论坛上，超临界压水堆被选定为6种未来发展堆型之一，日本在超临界压水堆方面发挥了主导性作用，其研究开发工作也走在了世界的前例。本文介绍了日本超临界压水堆实用化的研究开发，内容包括反应堆概念设计、传热与流动、材料与水化学的研究开发计划与成果。     

实验反应堆功率调节系统PID控制器的解析设计方法

通过对实验反应堆系统模型进行线性化和模型降阶、再对降阶模型进行PID控制器的理论设计的方法，可使反应堆功率调节系统的设计最优化、客观化和解析化。将此解析化方法应用于—实验反应堆功率调节系统的设计，通过数值仿真证明：设计的控制器控制效果良好，该解析化设计方法可行?     

秦山第二核电厂混合堆芯水化学控制技术探讨

根据秦山第二核电厂混合堆芯的特点,对其一回路水化学优化控制方法进行了研究,提出采用4段的硼-锂(B-Li)协调方案控制最高锂浓度,以降低反应堆结构材料的腐蚀风险.对不同燃料循环腐蚀活化产物进行跟踪分析,结果表明:在混合堆芯运行期问,采用优化的B-Li协调水化学控制对抑制一回路结构材料腐蚀和降低一回路辐射场是有效的.     

反应堆物理启动数字化在线实验系统

系统是基于微机在线数据采集和处理的综合实验系统，包含中子计数信号和堆功率信号采集模块、外推临界实验模块、周期反应性测量模块和反应堆中子通量水平实时监测模块。其数据及结果显示功能强大，包含中子探测信号显示、反应堆装载图显示、详细数据列表显示、次临界外推曲线和周期反应性内插曲线显示。微机操作界面友好，功能多且操作方便，可方便地设置数据采集卡参数、反应堆物理参数及与实验过程有关的其它参数。     

反应堆系统数字化实时仿真的高效实现

根据实验反应堆的物理特性，建立堆芯动态模型，探讨多种实时仿真算法的实现途径。研究提出了进行数字化实时仿真的一种高效实现方法。为配合功率调节系统半实物仿真试验而实现了一座实验反应堆在Windows平台下的实时仿真系统。      

Passive cooldown performance of a 65 MW integral reactor

Advanced integral-type pressurized water reactor with a maximum thermal power of 65 MW is under development at the Korea Atomic Energy Research Institute (KAERI). This 65 MW integral reactor incorporates a number of innovative design features. In the case of a transient, the passive residual heat removal system (PRHRS) is designed to cool the reactor coolant system (RCS) from a normal operation condition to a hot shutdown condition by a natural circulation, and the shutdown cooling system (SCS) is designed to cool the primary system from a hot shutdown condition to a refueling condition by a forced circulation. A realistic calculation has been carried out by using the TASS/SMR code and a sensitivity analysis has been performed to evaluate a passive cooldown capability for various system conditions such as natural and forced circulation conditions for the reactor coolant system or the passive residual heat removal system, and number of active PRHRS trains. The reactor coolant system and the passive residual heat removal system adequately remove the core decay heat by a natural circulation and the 65 MW integral reactor can cool the coolant to the SCS entry condition in the primary system for all the possible operational conditions.     

压水堆核电站一回路工况变化对主泵主要机械性能的影响

论述了大亚湾和岭澳1000MW压水堆核电站反应堆冷却剂回路(一回路)主要瞬态工况对反应堆冷却剂泵的主要机械性能参数的影响,为避免主泵受瞬态干扰,以及通过改变系统参数调整来改善主泵机械参数提供了理论依据。     

HFETR—回路水的辐解研究

本文对 HFETR 冷却剂——水的辐照分解做了细致的研究。监测了10—80MW 运行功率下氢及过氧化氢的生成量,以及同一功率状况下氢及过氧化氢随时间的变化情况、辐解量与水质的关系等.实验表明:辐解产物——氢及过氧化氢的量随功率的升高而增加,但功率提升至30MW 以后,其量变化缓慢。在同一功率状况下,自达到功率的瞬间开始,辐解量随时间迅速增加,但很快趋于稳定。反应堆在10—80MW 功率下运行,正常状况的氢含量不超过15mL/L(H_2O),过氧化氢最大为26.6ppm.     

10MW高温气冷堆一回路氦气品质研究

10MW高温气冷堆以氦气作为冷却剂,氦气中含有H2O、CO2、H2、CO、CH4、N2、O2等7种影响氦气品质的杂质。分析反应堆在不同工况下的氦气品质数据的变化规律,可证明一回路氦气在反应堆功率运行过程中经氦气净化系统净化后,氦气品质能够满足技术规格书要求。但随一回路氦气平均温度的升高,氦气品质呈下降趋势,并可初步判断存在缓慢变化的杂质源项为水。     

The impact of radiolytic yield on the calculated ECP in PWR primary coolant circuits

A code, PWR-ECP, comprising chemistry, radiolysis, and mixed potential models has been developed to calculate radiolytic species concentrations and the corrosion potential of structural components at closely spaced points around the primary coolant circuits of pressurized water reactors (PWRs). The pH(T) of the coolant is calculated at each point of the primary-loop using a chemistry model for the B(OH)₃ + LiOH system. Although the chemistry/radiolysis/mixed potential code has the ability to calculate the transient reactor response, only the reactor steady state condition (normal operation) is discussed in this paper. The radiolysis model is a modified version of the code previously developed by Macdonald and coworkers to model the radiochemistry and corrosion properties of boiling water reactor primary coolant circuits. In the present work, the PWR-ECP code is used to explore the sensitivity of the calculated electrochemical corrosion potential (ECP) to the set of radiolytic yield data adopted; in this case, one set had been developed from ambient temperature experiments and another set reported elevated temperatures data. The calculations show that the calculated ECP is sensitive to the adopted values for the radiolytic yields.     

A review of corrosion product transport and radiation field buildup in boiling water reactors

Cobalt-60 is the major radiation source in the boiling water reactor (BWR) for personnel exposure during shutdown maintenance. The Co-60 activity is produced by neutron activation of cobalt with other corrosion products deposit on fuel surfaces, and is released into the coolant and deposited on primary system piping walls in the system. The transport phenomena of corrosion products in the primary system and radiation field buildup are reviewed separately in three different areas: the behavior of corrosion products in the BWR coolant, including the chemistry of corrosion products and formation of mixed metal oxides; the transport of corrosion products on fuel cladding surfaces, and the mechanisms of deposition and release are discussed; and the transport of Co-60 and radiation field buildup on out-of-core surfaces under various chemistry conditions, including normal water chemistry, hydrogen water chemistry and with chemical additives. It is concluded that with understanding the mechanisms of transport, the radiation field buildup in most operating BWRs has been considerably reduced in recent years. The major factors are reduction of cobalt source reduction, control of Co-60 release from fuel surfaces with zinc addition and improvement in water quality to minimize the corrosion product input and the material corrosion.     

Predicted impact of power uprate on the water chemistry of Kuosheng boiling water reactor

A theoretical model was adapted to evaluate the impact of power uprate on the water chemistry of a commercial boiling water reactor (BWR) in this work. In principle, the power density of a nuclear reactor upon a power uprate would change immediately, followed by water chemistry variations due to enhanced radiolysis of water in the core and near-core regions. It is currently a common practice for commercial BWRs to adopt hydrogen water chemistry (HWC) for corrosion mitigation. The optimal feedwater hydrogen concentration may be different after a power uprate is implemented in a BWR. A computer code DEMACE was used in the current study to investigate the impact of various power uprate levels on major radiolytic species concentrations and electrochemical corrosion potential (ECP) behavior of components in the primary coolant circuit of a domestic BWR-6 type reactor operating under either normal water chemistry or HWC. Our analyses indicated that under a constant core flow rate the chemical species concentrations and the ECP did not vary monotonously with increases in reactor power level at a fixed feedwater hydrogen concentration. In particular, the upper plenum and the upper downcomer regions exhibited uniquely higher ECPs at 108% and 115% power levels than at the other evaluated power levels.     

破口大小和二回路运行状态对核供热堆自然循环断流瞬态的影响

为研究一体化布置的核供热堆在发生破口失水事故中破口大小和从中间回路排出热量减少对断流过程的影响，选用不同的破口尺寸和不同的二回路工作状态，在５ＭＷ核供热堆热工水力模拟回路ＨＲＴＬ－５上进行了实验研究。稳态运行工况的系统压力为１．５ＭＰａ，在发生小破口失水事故后，加热功率维持为额定功率的５％以模拟剩余发热情况。实验研究并比较了不同条件下压力、温度、循环流量、液位和失水量等重要参数的变化。这些实验数据为核供热堆的安全分析提供了实验依据。     

Dynamic behavior of upper hydraulic drive control rod

A 200 MW nuclear heating reactor (NHR) was developed based on the 5 MW testing heating reactor (THR-5). In order to improve its features, the INET was used to optimize its structure design. One task was to design the upper hydraulic drive control rod, which is located at the upper part of the reactor core, so that the fuel rod in each corner of the bundle will not be removed as they are in the THR-5. A program based on the method of characteristics was run to investigate the dynamic behavior of the upper hydraulic control rod drive system. The steady-state and dynamic characteristics of the control rod were studied. The influence of the reactor temperature and structure parameters on the operation reliability was investigated. Finally the optimal design parameters of the upper hydraulic drive control rod were obtained.     

核动力系统模拟技术的研究

简要回顾了清华大学核研院在系统模拟技术方面所开展的主要工作，重点介绍了基于ＲＥＴＲＡＮ－０２程序研究开发的２００ＭＷ核供热堆紧凑型模拟器和基于网络计算技术的开发的１０ＭＷ高温气冷堆网络并行模拟原型系统。