核电站氧化运行及效果分析

本文根据秦山核电厂第一次（没有实施氧化运行）和实施氧化运行的第二、三、四次换料大修期间主冷却剂放射性水平、现场辐射水平的测量结果的分析，就停堆过程中实施氧化运行对降低大修现场辐射水平的效果进行分析研究。分析表明，在停堆过程中实施氧化运行能有效地降低辐射源项，降低大修现场的辐射水平，是降低大修集体剂量、实现辐射防护最优化的有效手段之一。     

核电站某起源项异常事件调查研究

国内某电厂大修机组辐射水平偏高、大修累积集体剂量偏高,下行至热停平台时RCP系统辐射指数很高,以放射性核素Co-60为主,氧化净化后整体辐射指数偏高,对比另外3台机组及该机组大修数据,机组整体辐射情况处于最高水平。本文主要介绍该机组辐射源项偏高原因及调查过程。     

停堆氧化期间加大净化流量改进可行性分析

使用化学和容积控制(RCV)系统水力模型,研究M310机组停堆氧化期间提高净化流量以缩短冷却剂净化时间的可行性,并提出初步的工程改造方案。研究结果表明,停堆氧化净化流量从27.2 m3/h提高到40 m3/h,1000 MW级核电机组单次大修至少可节约2 h工期。     

压水堆核电机组停堆氧化过程中溶解氢控制策略优化

在压水堆核电机组功率运行状态下,反应堆冷却剂系统内始终保持氢覆盖,然而机组在进行停堆氧化过程中,因反应堆需开口,为避免氢氧混合爆炸,需要首先除去氢气,将一回路的溶解氢含量降低到规范值以下才能开展氧化运行工作.在压水堆核电机组停堆氧化过程中,一回路溶解氢的有效控制能够决定化学控制过程是否会成为大修下行的关键路径.福清核电...     

核电厂大修期间一回路冷却剂中放射性指标监测与控制

核电厂大修期间,从机组降功率至卸料结束,由于一回路冷却剂温度和压力不断降低、pH和氧化还原环境的改变,冷却剂中裂变产物和活化腐蚀产物比活度会发生系列变化。结合海南核电三次大修经验,阐述了降功率期间存在小缺陷燃料元件的氙和碘释放规律、一回路冷却剂中活化腐蚀产物的释放与净化过程、稳压器开人孔阶段一回路冷却剂放射性指标反弹现象及原因分析、卸料结束后乏池放射性指标反弹现象及原因分析,为后续机组大修期间一回路冷却剂放射性指标监督与控制提供借鉴。     

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。     

压水堆核电站氧化停堆过程一回路冷却剂中58Co活度浓度分析

压水堆核电站氧化停堆过程中,一回路冷却剂中⁵⁸Co的停堆释放峰值可达上百个GBq/t,对工作人员的职业照射剂量及停堆进程都有很大影响。本文介绍了压水堆核电站氧化停堆过程,分析了对⁵⁸Co活度浓度变化有显著影响的因素,如一回路水化学、蒸汽发生器传热管材料、循环中停堆、化学和容积控制系统的净化等,同时提出了相关建议。     

压水堆一回路冷却剂活化腐蚀产物钴银锑

介绍了压水堆核电厂一回路冷却剂中主要活化腐蚀产物钴、银、锑源项的产生和对于停堆机组剂量大幅增加的影响。研究这些核素在反应堆运行和停堆期间的行为并尽早探知这些污染物的出现,以便确定相应的解决办法。它包括：从源头做起,与一回路冷却剂系统接触的设备和部件尽量不采用含有钴、银、锑的材料;制定严格的水化学和停堆程序,使得对这些核素污染的净化能力最佳化和对过度污染最小化;根据具体情况改进净化工艺,限制污染带来的影响。实践证明,这些措施对减少或限制钴、银、锑的污染是行之有效的。     

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.     

VVER机组大修辐射源项控制体系研究和实践

基于VVER机组停机过程中辐射源项的释放和迁移原理，本文结合系统的设计功能建立了一套覆盖机组状态的大修全过程辐射源项控制方法，提出了一套覆盖机组状态的大修全过程辐射源项控制体系。该体系经某VVER核电机组验证，通过一回路pH和溶氢等水化学控制措施，可以降低设备的腐蚀速率和腐蚀产物被活化的几率。使用一回路冷却剂净化系统（KBE）、冷却剂贮存系统（KBB）树脂床对一回路介质可以实现对放射性核素的有效净化，其中一回路贮存水箱的净化效率可以达到90%以上；系统介质或者外接冲洗设备对高剂量率系统设备进行冲洗、净化，净化效率可以达到50%以上。结合VVER机组辐射源项控制经验和最新的源项控制技术，提出了后续VVER机组辐射源项控制的优化和研究方向。     

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.     

Application of radiation chemistry research to the operation of nuclear reactors

The theme of this review is the application of radiation chemistry research to improve the operating efficiency of nuclear reactors. The intense radiation fields in reactor cores produce a hostile environment for incore materials; this report describes how recent research helped overcome the chemistry problems caused by the radiation.Examples discussed are the inhibition of graphite moderator corrosion and prevention of carbon deposition in gas-cooled reactors, suppression of radiolysis of the cooling water in concrete pressure vessels, hydrogen formation following a loss of coolant accident in a PWR and improving the stability of decontamination reagents for water reactors.     

Water Chemistry Experience of Nuclear Power Plants in Japan

Japanese LWRs have experienced several troubles caused by corrosions of structural materials in the past ca. 20 years of their operational history, among which are increase in the occupational radiation exposures, intergranular stress corrosion cracking (IGSCC) of stainless steel piping in BWR, and steam generator corrosion problems in PWR. These problems arised partly from the improper operation of water chemistry control of reactor coolant systems. Consequently, it has been realized that water chemistry control is one of the most important factors to attain high availability and reliability of LWR, and extensive researches and developments have been conducted in Japan to achieve the optimum water chemistry control, which include the basic laboratory experiments, analyses of plant operational data, loop tests in operating plants and computer code developments. As a result of the continuing efforts, the Japanese LWR plants have currently attained a very high performance in their operation with high availability and low occupational radiation exposures. A brief review is given here on the R & D of water chemistry in Japan     

Analysis of Operating Experience Involving Loss of Decay Heat Removal during Reactor Shutdown in Pressurized Water Reactors

In PWRs, loss of decay heat removal (DHR) during reactor shutdown with the reactor coolant system (RCS) partially drained may result in core boiling in a short time. The subsequent RCS pressurization could prevent water flow into the RCS by gravity feed and consequently the core would be uncovered. This paper analyzes U.S. PWR operating experience involving the DHR loss in such reduced inventory conditions.

Between 1976 and 1990, reported were a total of 63 loss of DHR events which occurred during reactor shutdown with the RCS inventory reduced. Review of the event reports indicated that many loss of DHR events in reduced inventory conditions resulted from air entrainment into the DHR pumps due to lowering the reactor water level too far, loss of coolant inventory, increased pump flow and so on.

The coolant heatup rates were evaluated for 12 events with use of the data such as the time elapsed from reactor shutdown actually reported. The calculated results were in reasonably good agreement with the observed ones and showed that core boiling would take place within 1 h even if the DHR loss would occur in the late stage of shutdown (for example, 30 days after the shutdown).     

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。     

水冷聚变堆主回路活化产物源项计算分析

水冷聚变堆中结构材料活化腐蚀产物和冷却剂活化产物是正常运行工况下的最主要放射性来源,也是反应堆运行及维护过程中工作人员辐照剂量的直接来源。本文使用CATE V2.1程序对国际热核聚变实验堆(International Thermonuclear Experimental Reactor,ITER)LIM-OBB(Limiter-Out-Board Baffle)冷却回路的活化腐蚀产物和水活化产物进行模拟计算,并根据CATE模拟得到的放射性活度通过点核积分程序分别计算正常运行1.2 a及停堆15 d的剂量率。计算结果表明,反应堆运行期间冷却剂活化产物比活度和剂量率远大于结构材料活化腐蚀产物,而停堆后冷却剂活化产物迅速衰变完,结构材料的活化腐蚀产物成为比活度和剂量率的主要来源。     

压水堆核电厂压力容器开盖时刻主冷却剂放射性浓度控制要求研究

压水堆核电厂停堆开盖时刻主冷却剂放射性浓度限值是核电厂的重要设计参数。本文基于停堆开盖后厂内辐射风险来源分析,建立了适用于压水堆核电厂停堆压力容器开盖时刻主冷却剂中的放射性浓度控制值评估方法,并采用欧洲第三代压水堆技术方案(EPR)堆型核电厂的设计参数对建立的方法进行了验证。验证结果表明:基于此方法得出的停堆开盖限值与EPR堆型核电厂原设计较接近。     

核电厂核岛主设备支承间隙稳定性研究

核电厂核岛主设备与其支承之间设置有一定的间隙,目的是允许主设备因反应堆冷却剂系统温度和压力的变化而引起的自由热位移和热膨胀。间隙稳定性对核电机组的安全运行有重要意义,因此在役核电机组的每个换料周期中都需对此支承间隙进行测量评估,该测量评估工作花费时间长、辐射风险高。本文分析了支承间隙构成的影响因素,并结合间隙测量历史数据及工程经验提出了间隙稳定性概念及其验收准则,确定了间隙稳定性的评价流程。结合某在运核电厂稳压器支承间隙实测数据进行验证,为缩短间隙测量周期提供依据。该方法可缩短停堆换料的周期,减少大修测量人员的辐照剂量,保证机组安全运行及提高经济性。