水冷反应堆主回路腐蚀产物活化及迁移模型的研究

发布时间：2018-03-10 21:48

本文选题：活化腐蚀产物　切入点：CATE程序　出处：《华北电力大学(北京)》2017年博士论文　论文类型：学位论文

【摘要】：放射性源项关系反应堆系统运行、维修维护及退役等环节,对辐射防护、个人和集体剂量以及安全分析有重大影响。水冷反应堆中,结构材料与冷却剂接触发生腐蚀,生成了较稳定的氧化层,金属离子穿过氧化层释放进入冷却剂。辐照区的氧化层以及由冷却剂携带进入辐照区的金属离子受中子辐照发生活化反应成为放射性物质,冷却剂中的放射性物质在冷却剂的携带下沉积到非辐照区形成了γ辐射场,对电厂检修维护及运行人员构成辐照危害。正常运行工况下,压水堆堆芯外90%的集体剂量是由与一回路冷却剂接触的管壁上沉积的活化腐蚀产物ACPs(Activated Corrosion Products)引起的。对于水冷聚变堆,不存在裂变产物,ACPs成为放射性的主要来源。无论压水堆还是水冷聚变堆,ACPs对正常运行工况下的ORE以及事故工况下的潜在放射性释放都存在着重大影响,直接影响工作人员的照射剂量水平。对ACPs的研究是反应堆事故分析、剂量与辐射防护优化、放射性废物管理等的重要技术基础,是反应堆审查取证的重要环节。目前国内外计算ACPs多数使用的是经验模型和半经验模型,其应用范围非常有限,依赖于电厂运行数据或试验数据,模拟温度、pH值等参数限制在一定范围内的变化,只适用于特定的堆型和工况;对放射性核素的种类和核反应的种类有极大的限制,只能计算Co-58、Co-60、Fe-59、Cr-51、Mn-54等几种放射性核素的核反应,不能满足聚变堆高能中子辐照下多种材料的源项分析需求,也不能满足事故瞬态下短寿命核素的计算需求;聚变堆独有的脉冲运行特点也对计算提出了新的要求。本论文开发了基于经典的经验模型的水冷反应堆主回路ACPs计算程序。对水冷反应堆主回路ACPs的产生与迁移机理开展研究,建立基于浓度差驱动原理的机理模型,开发了基于机理模型的水冷反应堆主回路ACPs计算程序。脱离了对核电厂及试验回路的经验系数的依赖,结合溶解度的计算成功实现了物质迁移方向的自动匹配功能,突破了以往程序对堆型及运行工况的限制。借助课题组中的沉积试验及测量结果,根据对模型计算值和试验测量结果的分析,对沉积模块进行修正,成功实现了pH值对沉积行为的影响的模拟;对多种结构材料进行了不同运行环境下的腐蚀行为模拟试验,解决了聚变堆工况下腐蚀模型计算不准确的问题;引入EAF-2007数据库,为活化及衰变反应提供核数据,实现了计算任意放射性核素的功能;加入多种脉冲等效模块,满足不同计算需求及聚变堆型的要求,保证计算精度的同时可以大幅提高计算效率;添加点核积分模块计算相应的剂量率及职业照射ORE(Occupational Radiation Exposure),实现了活度浓度与剂量率的转换。通过上述工作,克服对pH值变化范围的限制,突破了以往程序对材料及工况、放射性核素种类的限制,直接给出γ剂量场使得计算结果更加直观。基于以上工作,开发了适用于压水堆和水冷聚变堆的ACPs计算分析程序CATE。为充分验证模型的正确性及程序的适用性,分别从试验验证和程序验证两个角度选取了试验回路MIT-PCCL回路、水冷聚变堆ITER LIM-OBB回路和压水堆秦山二期核电厂一回路进行了模拟分析,并与公开发表的文献结果进行了比对。计算结果均能与试验测量值和程序计算值保持在同一数量级,在源项计算领域内可以认为计算结果是吻合的,从试验和程序的角度验证了模型的准确性和结果的可靠性。水冷聚变堆的高温高压环境、产生的高能量中子会对结构材料产生较强的腐蚀、活化作用,水冷聚变堆对结构材料提出了更高的要求,结合我国已生产的多种低活化材料,应用CATE程序首次实现了国际热核聚变实验堆ITER(International Thermonuclear Experimental Reactor)环境下国产低活化材料及传统奥氏体不锈钢对水冷聚变堆ACPs影响的对比分析;当前中国聚变工程试验堆CFETR(China Fusion Engineering Test Reactor)处于设计阶段,ACPs源项的水平是其颁证的关键影响因素,可能对聚变堆设计和运行有很大的影响,目前国内尚无对CFETR的ACPs水平计算分析的研究工作,本文应用CATE程序实现了对CFETR包层回路的ACPs进行计算分析。
[Abstract]:The radioactive source term relationship reactor system operation, maintenance and decommissioning and other sectors, have a significant impact on the radiation protection, analysis of individual and collective dose and safety. In water reactors, structural materials in contact with the coolant corrosion, formation of oxide layer is stable, metal ions through the oxide layer is released into the coolant. The oxide layer irradiated area and from the coolant carrying metal ions into the irradiated area irradiated by neutron activation reaction as radioactive substances, radioactive substances in the coolant in the coolant carrying deposition to the non irradiated area formed a gamma radiation field of power plant maintenance operation and maintenance personnel constitute radiation hazard. Under normal operating conditions, the collective dose 90% pressure the core is made of activated corrosion products ACPs deposition wall contact with a coolant on (Activated Corrosion Products) for the cause. Water cooled fusion reactor, there is no fission products, ACPs has become the main source of radioactivity. Both PWR or water-cooled fusion reactor under normal operating conditions, ACPs of ORE and the release of potential radioactive accident conditions have significant influence, directly affect the dose level of the staff. The study of ACPs reactor accident analysis dose, radiation protection and optimization, an important technical basis for management of radioactive waste, is an important part of the reactor. The review of evidence at home and abroad ACPs calculation is used mostly empirical and semi empirical models, the application range is very limited, depending on the power plant operation data and the test data, the simulation of temperature, pH value and other parameters change limit in a certain range, only suitable for specific reactor types and conditions; types of radionuclides and nuclear reactions are extremely limited, can only calculate the Co-58 Co-60, Fe-59, Cr-51, Mn-54 and other types of nuclear reactions of radionuclides, can not meet the high energy neutron irradiation of various materials in fusion reactor source analysis needs, can not meet the computing needs of short life nuclide transient accident; fusion pulse operation characteristic and puts forward new requirements for the calculation. This paper developed a calculation program the main circuit of water-cooled reactor ACPs empirical model based on the classical research. And the migration mechanism of water cooled reactor main loop of the ACPs was established based on the principle of differential drive mechanism model, a computer program is developed the main circuit of ACPs water cooled reactor based on mechanism model. From the experience dependent coefficient on the nuclear power plant and the test circuit. Automatic matching with the calculated solubility of the successful implementation of the migration direction, broke through the limitations of previous procedures on the reactor type and operating condition. By the research group The deposition of test and measurement results, according to the analysis of the calculated value and the experimental results of the model, the deposition module is modified, the successful implementation of the simulation effect of pH value on the deposition behavior of various structural materials; the corrosion behavior of different operation environment simulation test, solves the corrosion condition of fusion reactor model the problem of inaccurate calculation; introduction of the EAF-2007 database to provide data for the activation and nuclear decay reaction, the calculation of arbitrary radionuclides; adding a variety of pulse equivalent module, to meet the different computational requirements and fusion type requirements, and ensure the calculation accuracy can greatly improve the computational efficiency; calculation of dose rate and irradiation ORE corresponding occupation add the point kernel integral module (Occupational Radiation Exposure), the activity concentration and dose rate conversion. Through the above work, to overcome the pH value range In the limit, breaking the previous program of materials and conditions, radioactive nuclide types, directly given dose field makes the result more intuitionistic. Based on the above work, developed for PWR and water-cooled fusion reactor ACPs analysis the applicability of CATE. program is correct and full program verification model. From the test and verification procedures two angle test circuit MIT-PCCL circuit, ITER circuit and LIM-OBB pressurized water cooled fusion reactor Qinshan two nuclear power plant in a loop are simulated, and the results and the published literature were compared. The calculated results are with the experimental values and the calculated values remain in the same magnitude, that calculation results were consistent with the source term calculation domain, the reliability model and the accuracy of the results is verified by the test program and point of view. Water cooled fusion reactor The environment of high temperature and high pressure, high energy neutrons produced by strong corrosion of structural materials, activation, water-cooled fusion reactor has put forward higher requirements for structural materials, combined with a variety of production in China has low activation materials, application of CATE ITER for the first time the International Thermonuclear Experimental Reactor (International Thermonuclear Experimental Reactor) under the environment of domestic low activation of contrast material and traditional austenitic stainless steel pile of ACPs impact on water cooled fusion analysis; current China Fusion Engineering Test Reactor (CFETR China Fusion Engineering Test Reactor) in the design stage, ACPs source level is the key factor affecting the certification, may have a great impact on the fusion reactor design and operation, at present there is no domestic level of CFETR ACPs calculation and analysis of the research work, the application of CATE program on CFETR clad ACPs gauge circuit Calculation analysis.

【学位授予单位】：华北电力大学(北京)
【学位级别】：博士
【学位授予年份】：2017
【分类号】：TM623.2

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