压水堆活化腐蚀产物源项~(58)Co和~(60)Co敏感性研究

以某压水堆核电厂为例,采用CORA程序分析压水堆核电厂一回路材料组成、蒸汽发生器传热管材料钴含量、冷却剂氢氧化锂浓度、净化效率和反应堆运行功率等因素变化对一回路腐蚀产物58Co和60Co活度浓度的影响。计算结果表明:通过限制蒸汽发生器传热管材料中钴元素的含量、提高冷却剂中氢氧化锂浓度、提高冷却剂净化效率和降低功率等措施可以有效降低活化腐蚀产物的活度浓度,为压水堆核电厂辐射剂量控制提供参考。     

B-Li水化学对核电金属材料腐蚀的影响

本文阐述了压水堆核电站一回路B-Li水化学工况控制的发展趋势,及其对腐蚀产物、降低剂量率的作用;概括了B浓度、Li浓度及pH值对镍基合金、不锈钢、锆合金的PWSCC敏感性、裂纹扩展速率、腐蚀产物释放速率等性能的影响;分析了核电站应用富集硼酸的积极作用。一回路水化学控制在较高pH有利于减少核电站金属材料的腐蚀,提高核电站的安全与可靠性。     

B-Li水化学对核电站金属材料腐蚀的影响

本文阐述了压水堆核电站一回路B-Li水化学工况控制的发展趋势,及其对腐蚀产物、降低剂量率的作用;概括了B浓度、Li浓度及pH值对镍基合金、不锈钢、锆合金的PWSCC敏感性、裂纹扩展速率、腐蚀产物释放速率等性能的影响;分析了核电站应用富集硼酸的积极作用。一回路水化学控制在较高pH有利于减少核电站金属材料的腐蚀,提高核电站的安全与可靠性。     

一回路注锌对腐蚀产物的影响分析

目前的压水堆中多采用注锌技术来降低一回路腐蚀产物的源项,然而关于注锌对腐蚀产物影响的理论机理以及计算分析研究较为欠缺。基于此,本文从理论机理、程序开发、数值计算分析和实验验证的角度论证分析注锌对一回路腐蚀产物以及源项的影响。理论计算表明：注锌能明显降低基体金属中镍和钴的溶解;随着运行时间的增加,注锌对一回路冷却剂中的⁵⁸Co和⁶⁰Co呈现出抑制作用;注锌实验结果与理论计算分析的比值在0.5～2.0范围内,符合情况良好。本研究能为核电厂合理地采取注锌技术提供理论支撑。     

压水堆核电站一回路活化腐蚀产物源项控制措施探讨

材料替代和一回路水化学控制是降低活化腐蚀产物源项的主要措施。本文介绍M310、AP1000和EPR三种压水堆核电站一回路水化学优化情况,比较三种压水堆一回路活化腐蚀产物源项,分析探讨水化学优化对源项降低的影响,最后对国内压水堆核电站一回路水化学优化提出建议。     

CPR1000停堆开盖一回路冷却剂中58Co放射性浓度分析

沉积于一回路系统设备内壁的活化腐蚀产物是压水堆核电厂停堆工况下的主要放射性来源.文中选择CPR1000停堆换料期间放射性浓度较高的活化腐蚀产物58Co作为研究对象,分析该核素在停堆开盖过程中放射性浓度变化的影响因素,并建立相应的放射性浓度计算模型.计算结果表明,一回路净化流量和附着于设备内壁的58Co释放率是影响停堆期间一回路冷却剂58Co放射性浓度变化的主要因素,同时从理论上得出了CPR1000机组停堆净化工序能够使得一回路冷却剂内58Co放射性浓度降至相关停堆放化控制限值内的结论.     

压水堆核电厂源项控制实践与改进

本文分析了福清核电厂1号机组停堆沉积源项调查发现的一回路管道内壁⁵⁸Co和⁶⁰Co表面活度水平、剂量率贡献以及随机组运行时间发生的变化情况,并介绍了压水堆核电厂活化腐蚀产物的形成、沉积及存在形式。通过分析201大修主泵停运对氧化运行效果及蒸汽发生器(SG)下封头辐射水平的影响,结合酸性氧化环境下腐蚀产物溶解度变化的特点,提出改进主泵停运时机以提高氧化运行效果的建议。另外,还分析了阀门密封面维修导致向一回路系统引入含钴金属颗粒对机组源项的影响,建议严格控制阀门维修过程以减少⁵⁹Co进入一回路系统。     

PWR核电站蒸汽发生器停堆湿保养工况联氨的缓蚀作用

一、引言压水堆(PWR)核电站蒸汽发生器(SG)管材因二回路系统腐蚀产物积累发生应力腐蚀开裂,这是SG传热管破损的主要原因之一。维修和更换SG使PWR停运期间所需要的替用电力对发电站造成很大的财政负担,同时,二回路系统的腐蚀产物沉积在蒸汽发生器内,降低     

蒸汽发生器管束区泥渣沉积特性研究

泥渣颗粒在蒸汽发生器(SG)管束区的沉积是传热管发生腐蚀的主要原因。本文通过数值模拟来研究泥渣颗粒在SG管束区的沉积特性以及不同粒径的泥渣颗粒在管束区不同位置的沉积特性。研究表明:低流速和回流是泥渣颗粒在管束区滞留的主要原因;当颗粒粒径较大时,泥渣颗粒在底板、流量分配板和支撑板上的沉积较为分散;随着粒径的减小,泥渣颗粒在底板上向中心处沉积,在流量分配板上向筒体处沉积,在支撑板上向中心处沉积。     

法国压水堆核电站的水化学

【瑞士《欧洲原子》1984年第4卷第9期第12页报道】压水堆核电站一回路和二回路的腐蚀,是使这类核电站中断运行的主要原因,而腐蚀则与其中用水的质量有直接关系。为了限制锈层的形成,抑制局部腐蚀现象发生,一回路与二回路用水的质量,在所有工况下都应符合一定的标准和规格。二回路用水的主要参数有 pH 值,氧含量、阳离子电导率和钠含量。法马通压水堆     

自然循环条件下蒸汽发生器倒U型管流量分配特性研究

以欧洲压水堆热工实验装置（PWR PACTEL）一回路系统蒸汽发生器为研究对象,首先,基于流体一维流动模型的质量、动量和能量守恒方程建立管道进出口压降以及传热与流体流量之间的关系;其次,以遗传算法为基础开发倒U型管蒸汽发生器流量分配计算程序,采用基准实验对程序正确性和可靠性开展验证;最后,利用流量分配程序计算蒸汽发生器倒U型管管组的流量分布情况,研究管高、管长以及一/二次侧换热系数对蒸汽发生器内流量分配的影响。结果表明,所开发流量分配程序计算结果与实验吻合良好;在选定的自然循环工况下,该蒸汽发生器中长管更易发生倒流,且倒流现象呈现分布范围广、单管流量低的特点;倒U型管内正流流速与管长成反比,与管高成正比,倒流流速随着管长的增加保持不变,与管高呈反比关系;传热系数较低时,总流量与传热系数成反比关系,当传热系数高于特定值后部分管内发生倒流,总流量骤降。      

压水堆核电厂乏燃料水池γ剂量率变化分析

⁵⁸Co是压水堆核电厂活化腐蚀产物的核心γ源项核素,受pH值和温度变化影响,含⁵⁸Co的活化腐蚀产物溶解度将持续发生变化。福清核电厂在执行某次机组调停小修过程中,一回路冷却剂中的⁵⁸Co活度浓度,随冷却剂温度下降而持续上升;在完成某次换料大修卸料工作后,乏燃料水池水温上升,池内⁵⁸Co活度浓度也随之升高,导致乏池表面最高γ剂量率达到了设计值的10倍左右。通过分析两个案例中,⁵⁸Co活度浓度、γ剂量率水平和温度变化趋势,对比工艺系统的运行记录,可以确认:两次⁵⁸Co活度浓度的升高,均与溶液温度密切相关。分析结果表明,在酸性环境下,含⁵⁸Co的活化腐蚀产物,其溶解度在一定温度范围内具有正温度系数,溶解度将随温度上升而增大;达到最大值后,溶解度表现出负温度系数,溶解度随温度上升而减小。根据该结论,通过启动乏燃料水池备用冷却回路,降低乏池温度,成功减小了池内的⁵⁸Co活度浓度,乏池表面γ剂量率迅速恢复至正常水平,避免了后续燃料操作人员的额外剂量照射。该实践的成功,对抑制和去除压水堆核电厂活化腐蚀产物中的⁵⁸Co,提供了新的思路。     

Chemical cleaning of PWR steam generators with a low temperature process

Corrosion products generated in the steam, feedwater and condensate systems of a PWR will be transferred by the feedwater into the secondary side of the steam generators (SG). Up to several hundred kilograms of deposits may collect on the surfaces of the SG. These deposits not only reduce the efficiency of the SG by deterioration of the heat transfer, but also cause an acceleration of the corrosion of the SG tubes and blockage of support plate passages.The chemical removal of the corrosion products opens the possibility to eliminate causes of defects on heat transfer tubes by removing the corrosion products and trapped impurities.A low temperature cleaning process, based on the EPRI developed SGOG iron and copper solvents, was applied in November 1990 to remove the hard deposits from the tube sheets of a two-loop plant. The sludge containing over 60% copper was removed with the application of one iron removal step and several copper removal steps. Over 95% of the available sludge was removed. The corrosion of the unalloyed and low alloy materials was extremely low. The Incoloy 600 tubes showed no corrosion.In addition aspects of crevice cleaning at elevated temperatures are mentioned.     

FRAMATOME's continous efforts to improve steam generator corrosion resistance

The steam generators of PWR nuclear reactors are among the primary components most affected by corrosion problems. Corrosion of the steam generator tubes, which assure heat transfer between the primary and secondary circuits, have been observed on a large number of operating steam generators, especially in the United States. According to an NRC survey, as of November 1981, forty PWR units with steam generators of the recirculation type were in operation in the US. Of these, 32 have been found to have one or more forms of tube degradation.Construction of the French PWR nuclear program started in the early 70s, at the time a number of operating plants in the US were being affected by the first corrosion problems. Since, at that time, its construction program was in an early stage, FRAMATOME was able to make modifications on the first units to improve steam generator resistance to corrosion. For instance, full depth expansion of the tubes in the tube-sheet using an explosive process (Westex) was performed on Fessenheim 1 steam generators already installed on site. Later on, continuous operating experience was being obtained in the US, before startup of the French units. This allowed FRAMATOME to react rapidly and take immediate corrective actions at the design stage, during fabrication and sometimes even on site in order to mitigate the risk of corrosion in the steam generators.FRAMTOME is confident that the present design of its steam generator models, including a large number of major improvements is adequate to prevent major corrosion problems to occur during operation. However, the company has embarked on an important development program to further improve the corrosion resistance and thereby the reliability of its steam generators. This program includes studies on new tube expansion techniques, alternate materials for steam generator tubes (Inconel 690), improved tube inspection methods, local thermohydraulic flow, tube vibrations, etc.     

The design and thermohydraulics study of the HTR-10 High Temperature Helium Experimental Loop

A High Temperature Helium Experimental Loop (HTHEL) for the purpose of studying the transportation and deposition behavior of solid fission products in high-temperature helium coming from the steam generator (SG) in the 10 MW High Temperature Gas-cooled Test Reactor (HTR-10) is studied and designed. Through the optimal design based on thermohydraulics analysis, the three-sleeve structure of deposition sampling device (DSD) could realize full-length temperature control evenly and simulate the physical environment of the heat transfer tube of SG in the HTR-10 in the sense of thermohydraulics, which could be used to study the original source term in the primary circuit. The simulation of the graphite dust particle trajectories in the DSD are shown and it is elucidated that DSD could also be used to study the behavior of graphite dust in the high temperature helium in the SG of HTR, which will provide deeper understanding for the analysis of source terms of HTR-10.     

Thermal-Hydraulic Study of Integrated Steam Generator in PWR

One of the safety aspects of innovative reactor concepts is the integration of steam generators (SGs) into the reactor vessel in the case of the pressurized water reactor (PWR). All of the reactor system components including the pressurizer are within the reactor vessel in the SG integrated PWR.

The simple heat transfer code was developed for the parametric study of the integrated SG. The code was compared to the once-through 19-tube SG experiment and the good agreement between the experimental results and the code predictions was obtained. The assessed code was used for the parametric study of the integrated once-through 16 m-straight-tube SG installed in the annular downcomer. The proposed integrated SG as a first attempt has approximately the same tube size and pitch as the present PWR and the SG primary and secondary sides in the present PWR is inverted in the integrated PWR. Based on the study, the reactor vessel size of the SG integrated PWR was calculated.     

A comparison between code calculation and blowdown experiments simulating a loss-of-coolant accident in a pressurized water reactor

Within the reactor safety programme of the EURATOM Joint Research Centre at Ispra the transient heat transfer phenomena during depressurization are experimentally investigated under PWR conditions. The special closed loop DHT-1 essentially represents one subchannel and the upper and lower plenum of a pressurized water reactor. A test series simulating rupture in the hot leg of a primary cooling circuit was carried out. Pressure and test tube temperatures were measured at various rupture cross-sections. Independently from these experiments, a blowdown computer code was developed by the Groupement Atomique Alsacienne Atlantique (GAAA). The core part of this code allows calculation of the thermohydraulic history of the coolant within the core after a rupture in the primary cooling circuit. It has been checked with regard to the hypothesis and correlations applied; the experiments and calculations are compared.     

Corrosion and deposition on the secondary circuit of steam generators

This study conducts a critical review on the studies of material corrosion and deposition on the secondary circuit of a pressurized water reactor, especially on the steam generators (SGs). Available knowledge has shown that the structural materials in the environment of the secondary circuit are susceptible to flow-accelerated corrosion and deposition-induced degradation. The deposition of the non-volatile impurities, especially the corrosion products, on the SG surfaces can be a primary cause of material degradations, including stress corrosion cracking. The review will analyze the fouling mechanisms and behaviors, the source of impurities, corrosion mechanisms, and the factors that affect the deposition and corrosion behaviors.