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。     

核电厂常规运行工况下放射性惰性气体和碘的释放源项计算

本文对核素在反应堆内部的产生、转移和释放过程进行了分析研究,计算了一回路冷却剂中放射性核素的浓度。并考虑了一回路系统中放射性核素向安全壳及其辅助厂房的泄漏,计算了由此带来的气载放射性核素活度的变化。同时,对秦山二期核电厂常规运行工况下气载流出物释放源项进行了计算,并与设计值和实际测量值进行了比较分析。从结果来看,本文计算值比实测值和设计值约大一个量级,本文计算方法可为核电厂气载流出物释放源项提供一个上限值。     

压水堆核电站常规运行工况下气载放射性物质释放源项的计算

本文介绍一个自行编制的用于计算压水堆核电站在常规运行工况下气载放射性物质向环境释放量的计算机程序ＭＧＡＬＥＳ。采用ＯＲＩＧＥＮ２程序，根据燃料元件的成份和燃耗情况计算堆芯的放射性核素谱；用放射性物质经堆芯向一回路迁移的逃脱率系数计算一回路冷却剂中的放射性核素浓度；再考虑核电站实际运行过程中一、二回路冷却剂的泄漏以及通风、除气等过程，计算其正常运行工况下气载放射性物质向环境的释放量。     

二次中子源破损诊断方法研究与验证

压水堆核电机组使用的二次中子源存在破损风险,反应堆功率运行工况下无法对二次中子源的状态进行物理检查。根据二次中子源的活化特性将¹²²Sb和¹²⁴Sb作为诊断二次中子源破损的特征核素,对使用一回路冷却剂的γ放射性在线监测数据、一回路冷却剂中¹²²Sb和¹²⁴Sb的比活度诊断二次中子源破损的方法可行性进行了分析,设计了二次中子源破损诊断流程,并使用上述诊断方法对二代改进型1000 MW级压水堆核电机组二次中子源破损问题进行了诊断。验证结果表明,二次中子源破损后一回路冷却剂取样分析得出的¹²²Sb和¹²⁴Sb比活度变化趋势与核辐射监测设备监测到的一回路冷却剂γ放射性变化趋势在总体上吻合。因此,本研究提出的二次中子源破损诊断方法是有效的。     

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.     

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。     

反应堆燃料元件破损当量诊断方法

破损当量是衡量反应堆燃料元件破损严重程度的重要指标,但破损当量无法直接测量,在决策应用中不具有可操作性,需要建立与破损当量对应的可监测指标。本文结合实践经验,分析确定了可用于燃料元件破损诊断的典型核素,建立了反应堆一回路冷却剂中裂变产物核素活度浓度与燃料元件破损当量之间的传递关系;给出了一回路冷却剂取样分析实验方法,并指出实验过程中应注意的问题;建立了采用监测一回路冷却剂中典型裂变产物核素活度浓度诊断破损当量的方法,并分析了诊断中不确定度的主要影响因素。本研究为反应堆燃料元件破损当量诊断提供了技术方法。     

压水堆核电厂裂变产物源项计算方法研究

裂变产物作为一回路冷却剂中放射性核素的重要组成部分,在核电厂设计中具有非常重要的意义。文中对堆芯积存量计算模型、燃料包壳内裂变产物向一回路冷却剂释放模型、裂变产物在一回路中的平衡模型进行了分析与研究,并以典型压水堆核电厂为例进行了计算与验证,证实了本文中给出计算模型的合理性以及适用性,可供压水堆核电站裂变产物源项计算分析参考。     

HFETR随堆辐照组件核素释放对一回路水质影响研究

随堆辐照组件考验增加了反应堆一回路水质异常的不确定性,难以有效判断辐照组件是否是导致反应堆一回路水质异常的原因。本文基于一回路水中核素的产生与消失的平衡关系,建立了随堆辐照组件破损导致一回路水质异常上升的计算模型。结合堆运行一回路水比活度监测数据,该计算模型可作为随堆辐照组件考验过程中发生破损的判据,同时可用于随堆辐照组件对反应堆运行水质的影响进行分析。以HFETR某炉段随堆考验组件的水质异常升高现象为研究对象,通过计算模型所得计算结论与监测数据比对分析,表明该随堆辐照组件未发生破损,且不是导致一回路水总比活度异常的主要原因。该随堆辐照组件辐照后经堆外解体检测显示随堆考验组件未发生破损,与计算模型研究结论相同。     

压水堆核电厂冷却剂碘同位素活度比值131I/133I与燃料完整性关系研究

压水堆核电厂正常运行期间燃料元件破损会造成一回路裂变产物活度升高,碘同位素活度比值131I/133I是行业内最常用的判断燃料破损情况的指标之一。本文介绍了压水堆正常运行期间冷却剂131I和133I的产生来源和迁移过程,建立模型估算了燃料完整、小破口和大破口情况下131I/133I范围,并通过在运CPR1000型压水堆核电厂的运行监测数据对计算模型进行了验证,两者符合得较好。     

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.     

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 salt-cooled 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.     

Critical parameter study of PWR primary coolant pipe leak failure using probabilistic fracture mechanics

This paper discusses the critical parameters which influence the failure probabilities of a PWR primary coolant loop. Probabilistic fracture mechanics (PFM) is applied for the parametric study, using the Monte Carlo program PRAISE to predict the failure probabilities of a PWR primary coolant loop from various distributions of input parameters. Parameters such as nondetection probability of preservice and inservice inspection, vibratory stress, residual stresses, and their correlations are extensively studied. Critical crack depth which causes immediate failure are calculated in the presence of various vibratory stresses with and without residual stresses. Crack growth schemes are determined with various initial defect depth and depth-length ratio as a function of plant operation time. The results show quantitatively how PWR primary coolant loop reliability can be greatly improved by increasing the sensitivity and decreasing the uncertainty of preservice nondestructive inspection.     

一回路冷却剂净化系统的优化运行

介绍了一回路冷却剂净化系统（KBE）的结构及陛能特点,研究分析了氨对硼酸型态及阴阳树脂的影响,冷却剂贮存系统（KBB）的设计缺陷。整理绘制了机组运行过程中碱金属、溶解氢的趋势图,结合机组在实际运行中出现的阴棚旨排带造成冷却剂氯离子超标、总碱金属偏离、溶解氢浓度下降等实际案例,总结优化了阳树脂氨钾饱和的开始时间、加钾量和氨浓度的控制;以及在不改变KBE初始设计的基础上增加KBE除碱金属功能,优化碱金属偏离的纠正措施。并根据实际运行结果对PUROLITE和BAYER两家公司生产的核级树脂性能进行了对比。     

一回路动态排气剩余空气体积标准值提升后冷却剂流动特性研究

针对动态排气后提升一回路剩余空气体积标准值的改进方案,提出含高溶解度空气的冷却剂在主泵启动瞬态下的压力预测方法和是否释放为两相分离流动的判断方法,对一回路及其辅助系统进行热工水力建模,空气体积标准值提升为24标准立方米（1标准立方米=1.293 kg）后,对主泵启动的瞬态过程进行了仿真,得到了一回路主要节点压力变化规律;结合冷却剂中气体溶解-释放模型,得到饱和氮气溶解度、氧气溶解度变化规律。结果表明,主泵启动瞬态过程中,一回路主要节点压力均在机组运行正常范围内,一回路中溶解的氮气、氧气不会释放成为两相流动。因此,就流动特性而言,空气体积标准值提升到24标准立方米可行。      

Calculation of the activity of fission products in the primary coolant of the eastern-type pressurized water reactor (VVER1000-V446) of the Bushehr Nuclear Power Plant at normal full power operational condition

Calculation of the primary circuit's coolant activation due to fission products (FPs) has been investigated for the eastern-type pressurized water reactor (VVER1000-V446). The reactor has been considered under normal full power operational condition for the first fuel cycle. Determination of the reactor coolant activity is based on time-dependent fission product core inventories. ORIGEN2.1 code has been used to determine the time-dependent fission product core inventories. The fission products activity in the primary coolant is calculated using a set of ordinary differential equations (ODEs) which governs the FPs concentration in the primary coolant. Results for 87 FPs have been calculated. The results of these calculations have been found to agree well with the corresponding available values found in the Final Safety Analysis Report (FSAR) of the Bushehr Nuclear Power Plant (BNPP).     

Mathematical Simulation and Analysis of the Behavior of Radionuclides in the System Krasnoyarsk Integrated Mining and Chemical Plant–Enisei–Kara Sea

The equations describing the transport of radionuclides in a system consisting of river water with a suspension and bottom deposits, taking account of the influx of radionuclides with the discharges from the plant and runoff from the basin area, are formulated. The coefficients in the equations are determined by analyzing the data from measurements. On this basis, the flood-plain deposits are dated and the time dependence of the radionuclide discharges over the entire period of operation of plant is determined and the concentration in water and bottom deposits is calculated. The balances between the discharge, the runoff from the basin area, and influx into the bottom deposits, the floodplain, and the Kara Sea are constructed for the past. A procedure is developed for estimating the outflow of mobile forms of radionuclides from the bottom deposits after a sharp decrease of discharges in 1992. The rate of self-purification of the river basin in the future is estimated taking account of this effect and the redistribution of radionuclides between the bottom and flood-plain deposits.     

全厂断电事故下AP1000非能动余热排出系统瞬态特性数值分析

在一维质量、动量和能量守恒方程基础上建立了AP1000反应堆主冷却剂系统及非能动余热排出系统数学模型,并编制了用于该系统瞬态特性分析的动态仿真程序PRHRSDSC。模拟了非能动余热排出系统在全厂断电事故下的瞬态响应过程,并将计算结果与西屋公司的LOFTRAN程序结果进行对比。结果表明：系统可依靠自然循环有效导出堆芯余热,一回路冷却剂温度维持在过冷状态,峰值压力未超过运行压力限值,各参数的变化趋势符合良好,证明了建模的合理性。     

压水堆主冷却剂中氚源项计算分析

压水堆主回路冷却剂流经堆芯时,水中固有及特加核素受中子辐照后会产生氚,氚几乎全部以气体和液体的形式排入环境,造成氚污染。因此,氚是压水堆辐射环境影响评价的主要关注内容之一。本文以AP1000为例,根据压水堆主回路冷却剂中氚的产生途径及其随时间的变化情况建立详细的计算模型,计算压水堆主回路冷却剂中的氚活度并分析各产氚途径对氚产生量的贡献。计算结果表明：主回路冷却剂中的氚主要来源于可溶性硼的中子活化和铀裂变,对氚产生量的贡献达80%以上;在⁷Li纯度为99.9%时,AP1000主回路中的年产氚量为5.23×10¹³ Bq,锂产氚量占总量的14.01%,随⁷Li纯度的增加,锂产氚量的贡献呈线性减小,在⁷Li纯度为99.99%时,锂产氚量占总量的3.18%。其他途径对氚的产生量贡献很小,可忽略。根据以上结果,可通过控制主回路冷却剂中添加的初始硼浓度、提高燃料包壳质量、增加LiOH中⁷Li的纯度等多种途径来降低主冷却剂中氚的产生量,从而减少氚对环境的放射性污染。