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。     

基于动态差分法的压水堆一回路放射性核素浓度分析

本文通过分析一回路冷却剂在堆芯辐照区、非辐照区、稳压器及化容控制系统中的流动特性,建立核素浓度的动态差分数学模型,模型特征参数可根据实际操作情况进行调整,将每次取水测量值对数学模型计算初始值进行修正,以准确地反映核素浓度变化情况。应用所建立的动态差分数学模型针对某一典型压水堆的实际运行工况进行计算,并将计算结果与Profip5程序计算值进行对比,验证了所建立的数学模型的准确性。然后,对压水堆一回路放射性核素浓度进行计算分析,得到一回路冷却剂核素浓度和辅助系统中核素平衡浓度,以及各系统核素浓度随时间的变化规律和停堆时一回路核素的浓度变化规律。结果表明,所建立的动态差分数学模型冷却剂核素计算值与Profip5计算值相差不大,化容控制系统对一回路放射性核素的净化率与国家标准中提供的净化率相吻合,方程组可用于压水堆不同工况下冷却剂核素浓度计算,在燃料破损监测时,对分析破损发生的时间、预估破损后冷却剂核素浓度峰值、计算破口所在燃耗区域及大小均有重要意义。     

高温气冷堆核电站安全级湿度仪研制

华能石岛湾高温气冷堆核电站与现役压水堆核电站一回路冷却剂不同,其采用氦气做为一回路冷却剂。氦气中水分含量过高可能导致安全事故,在运行时需要实时监控一回路冷却剂中的水分含量,当水分含量超过限值时发出停堆信号传送到保护系统。高温气冷堆湿度仪工作在高压环境下,输出信号与温度、湿度、压力等有关。通过试验压力、温度、湿度等物理量对湿度传感器输出信号的影响的分析研究,建立水分含量补偿模型,研制出基于纯硬件补偿电路的湿度仪样机并通过1E级鉴定试验验证。1E级湿度仪实现了实时监控一回路冷却剂中的水分含量,在水分含量超过限值时触发保护系统实现保护逻辑的目标,为反应堆安全运行提供了技术保障。     

压水堆核电厂一回路冷却剂系统中锕系核素质量评估方法研究

为评估压水堆核电厂燃料包壳破损时的工作人员辐射风险和燃料包壳破损程度,基于特征物理量建立一回路冷却剂系统中锕系核素质量评估方法。本文基于锕系核素的生成和迁移机理,建立了一回路冷却剂系统中锕系核素的平衡方程组,并选取3种易监测的特征物理量用以评估锕系核素向一回路冷却剂系统的释放量及其分布,并建立了一回路冷却剂系统中锕系核素质量的评估方法。然后分别采用国内在役压水堆核电厂无燃料包壳破损和有燃料包壳破损的实测数据对建立的评估方法进行了验证,验证结果表明：建立的评估方法可在无燃料包壳破损和有燃料包壳破损的情况下对一回路冷却剂系统中锕系核素质量进行评估,评估结果和预期符合。本文研究成果可为压水堆核电厂运行期间一回路冷却剂系统中锕系核素质量及其分布评估提供指导,从而优化后端的工作人员防护措施,降低辐射风险。     

基于RELWWER程序的WWER型核电厂燃料棒破损分析

裂变产物是一回路冷却剂中放射性核素的重要组成部分,在压水堆核电厂的运行过程中,需对一回路冷却剂进行放射性测量,并根据其中的裂变产物活度监控燃料组件的运行状态。本文通过对比分析RELWWER程序的计算结果和WWER型核电厂一回路冷却剂裂变产物比活度的实测数据,给出了初步判断堆芯中燃料棒的破损情况的方法,可为停堆换料方案的制定提供参考。     

压水堆—回路用不锈钢钴含量的分析探讨

本文推导了一套反应堆冷却剂腐蚀产物比活性的计算公式,进行了计算分析,验证了国外有关参考数据;并在此基础上提出“一般压水型核动力装置中堆内部件不锈钢的钻含量最好小于0.02%;堆外一回路不锈钢部件的钴含量可以不加特殊限制要求”的观点。这种观点的成立,将可能在满足辐射安全要求的前提下降低堆外一回路不锈钢部件的建造成本(对每一中小型装置来说,可节省3—20万元),并可简化生产工艺。     

安全壳通风系统放射性净化定量分析

以先进压水堆核电厂为对象,开展了正常运行工况安全壳内气载放射性产生方式研究,并构建了分析模型,包括冷却剂泄漏及40Ar中子活化。在此基础上,定量的论证了安全壳空气过滤系统对放射性净化作用,结果表明:无排风净化情况下安全壳大气内放射性水平较高,可达DAC(导出空气浓度)限值15.5倍,应实行较严格的措施限制人员进入;通过敏感性分析,识别出85Kr及133Xe为主导核素,由于这些核素半衰期较长,仅依靠衰变较难去除,采用每周定期20 h净化方案可解决该问题。同时,进一步研究了降功率并发碘尖峰机理模型,论证了停堆工况通风策略的有效性,结果表明:实施大风量净化可在进入冷停堆状态时将安全壳内气载放射性降到DAC限值,为人员在安全壳内进行长期操作提供了条件。     

高浜核电站1、3、4号机组的自动停运与重新运行

【日本《KNIC新闻速报》 1 999年 1 0月2 9日报道】　高浜核电站 1、 3、 4号机组(压水堆 ,1号机组额定功率为 82 6 MW;3、4号机组额定功率为 870 MW)于 1 0月 2 7日1 1时 4 8分自动停堆。其原因是 ,西京都变电所的 50万伏变压器的保护装置发生了故障 ,输电系统的频率及电压发生了变化。高浜核电站的报警器发出了“一回路冷却剂泵频率低”及“一回路冷却剂泵电压低”的信号 ,于是反应堆自动停堆。高浜核电站受这次自动停堆的影响 ,主汽轮机及给水泵的轴承润滑油里混入了蒸汽(水 ) ,于是便全部更换了润滑油。此外 ,这次发生的问题没有对…     

停堆氧化运行中主回路活化腐蚀产物的迁移与控制

核电站主回路中腐蚀产物的氧化净化是核电站大修的源项控制中最重要的一环.本文综合分析氧化净化过程中pH值、H2O2和溶解氢浓度对主回路活化腐蚀产物迁移的影响,结合大亚湾核电站的实践经验,提出了低pH值、低H2和集中引入适量氧的控制手段.     

某三代压水堆机组停堆水化学控制工艺改进与应用

有效降低压水堆机组反应堆冷却剂系统（RCP）材料腐蚀速率的同时有效去除活化腐蚀产物,可降低堆芯外辐射场、减少工作人员受照剂量,从而确保核电机组大修工作的顺利展开。某三代PWR机组采用富集硼酸（EBA）进行反应性控制的同时,利用其在功率运行期间对RCP系统冷却剂实施水化学控制的显著优势,同时在机组首次大修期间对停堆水化学控制工艺采取的改进措施（包括碱性环境向酸性环境转换、还原环境向氧化环境转换、强制氧化期间多次向一回路添加双氧水维持氧化性、化学和容积系统混床最大流量净化等）,在机组停堆下行阶段实现了降低机组辐射剂量并减少工作人员受照剂量的目的。      

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。     

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.     

Developments in Practical Time Series Analysis Methods in the Physical Sciences

A calculation model was developed to predict shutdown dose rate around the recirculation pipes and components in boiling water reactors (BWRs) by simulating the generation, transport, activation and deposition of corrosion products in the primary cooling water. The model is characterized by separating cobalt species in the water into soluble and insoluble materials and then calculating the deposits on the pipe wall for each species using the following considerations:

1. Soluble cobalt (designated as ionic cobalt) is taken into a spinel structure on the surface after diffusing into the oxide layers.

2. Insoluble cobalt (designated as crud cobalt) deposits on the oxide layers.

3. A part of the ionic cobalt released from the crud on the oxide layers is taken into a spinel structure like (1).

The calculated results agree satisfactorily with measurements in two BWR plants. The shutdown dose rate around the recirculation pipes during the entire operating life was calculated to evaluate the effects of the radiation reduction procedures.     

Activation analysis of the water cooling system of the LIPAc beam dump

LIPAc stands for Linear IFMIF Prototype Accelerator. LIPAc generates a 9 MeV deuteron beam, which is stopped at a beam dump, depositing over 1 MW of thermal power. A water cooling system has been devised for extracting this energy while keeping operational temperatures within range. The existing high neutron fluxes in the beam dump during operation produce activation of both coolant and beam stopper, which also suffers from corrosion into the coolant. The presence of radioisotopes in the cooling water leads to a radiological hazard.Water purification systems are located outside the accelerator vault and accumulate activated products during filtration, requiring a specific radiological shield to comply with target dose rates. Also devices containing large volume of activated cooling water, like N-16 decay pipes, require specific radioprotection analysis and design. This work identifies the most relevant radiation sources due to the activated cooling fluid, which may result in radiation doses to workers, and propose radioprotection measures into the design to mitigate their effect.     

Time-dependent corrosion product activity in a typical PWR due to changes in coolant chemistry for long-term fuel cycles

Extended fuel cycles with 18–24-month duration have complicated effects on primary coolant chemistry and finding optimum pH concentration range is difficult. For extended burnup cycles, study of changes in coolant activity due to corrosion products have been carried out for a typical pressurized water reactor (PWR) under pH and boric acid variations for linearly as well as nonlinearly accelerating corrosion rates. The computer program CPAIR-P (Deeba et al., 1999) has been modified to accommodate for time-dependent corrosion coupled with the effects of coolant chemistry. Results for ²⁴Na, ⁵⁶Mn, ⁵⁹Fe, ⁶⁰Co and ⁹⁹Mo show that the specific activity in primary loop approaches equilibrium value under normal operating conditions fairly rapidly. During reactor operation, predominant corrosion product activity is due to ⁵⁶Mn and after shutdown cobalt activity dominates. These simulations suggest that the effect of increase in pH value for an extended 24-month cycle on specific activity in the form of a decrease in the activity smeared by a linearly rising corrosion. The new saturation values for activity at the end of cycle are lower (∼50%) than a reactor operated at constant low pH/natural boric acid in coolant. For linearly accelerated corrosion and an increase in pH value, the coolant activity shows an initial rise and then it falls to a lower saturation level at the end of cycle when pH becomes large and the activity follows the slope of corrosion rate. For nonlinear rise in corrosion rate coupled with the pH rise from 6.9 to 7.4 and use of enriched boric acid (20%–40%) results in coolant activity peak during the cycle and it also approaches much smaller saturation values at the end of cycle when compared with the activity for the system having constant low pH value (6.9) in the coolant. In this paper we show that the use of enriched boric acid as chemical shim actually lowers the primary coolant activity when higher pH values are employed in coolant rather than natural boric acid. For multiple long-term fuel cycles the corrosion product activity shows an initial rise to a maximum value and then it fall back to low saturation values due to high pH concentration for enriched boric acid (40%) as chemical shim. If the removal rates of ion-exchanger are low (ε_IQ_I ∼100–300), then the activity builds up. However, when ion-exchanger performs well (ε_IQ_I ∼600–900) then the overall cobalt activity remains low and saturation value does not grow much with cycles.     

A study on a magnetic separation of radioactive corrosion products from NPP using permanent magnets

It is important to emphasize that the current trend to longer fuel cycles (18–24 months) has complicated the dilemma of finding optimum chemical condition for the primary coolant because of some problems such as increase of radioactive corrosion products, possibility of axial offset anomaly and so on. Radioactive corrosion products which are generated by the neutron activation of general corrosion products at a nuclear power plant are the major source of occupational radiation exposure. Generally, radioactive corrosion products exit in soluble and insoluble forms, and are removed by ion exchangers and purification filters. Most of the insoluble radioactive corrosion products have the characteristic of showing strong ferrimagnetism. Along with the new development and production of permanent magnets (rare earth magnets) which generate much stronger magnetic fields than conventional permanent magnets, a new type of magnetic filter is suggested that can efficiently separate corrosion products using rotation of permanent magnets. This new magnetic filter reveals good performance results in filtering magnetite, cobalt ferrite and nickel ferrite from aqueous coolant simulation.     

Radionuclides in primary coolant of a fluoride salt-cooled high-temperature reactor during normal operation

The release of fission products from coated particle fuel to primary coolant,as well as the activation of coolant and impurities,were analysed for a fluoride saltcooled high-temperature reactor (FHR) system,and the activity of radionuclides accumulated in the coolant during normal operation was calculated.The release rate (release fraction per unit time) of fission products was calculated with STACY code,which is modelled mainly based on the Fick's law,while the activation of coolant and impurities was calculated with SCALE code.The accumulation of radionuclides in the coolant has been calculated with a simplified model,which is generally a time integration considering the generation and decay of radionuclides.The results show that activation products are the dominant gamma source in the primary coolant system during normal operation of the FHR while fission products become the dominant source after shutdown.In operation condition,health-impacts related nuclides such as 3H,and 14C originate from the activation of lithium and coolant impurities including carbon,nitrogen,and oxygen.According to the calculated effective cross sections of neutron activation,6Li and 14N are the dominant 3H production source and 14C production source,respectively.Considering the high production rate,3H and 14C should be treated before being released to the environment.