压水堆核电站一回路主要活化腐蚀产物及水化学控制措施

压水堆核电站在未发生燃料包壳破损的情况下,85%的堆芯外辐射场是由活化腐蚀产物造成的。本论文分析讨论了一回路冷却剂中主要的活化腐蚀产物及其产生的原理,提出了通过水化学控制来降低活化腐蚀产物的几种主要措施,包括pH控制、氧化运行以及注锌技术等。核电站相关分析数据表明,这些水化学控制措施对于控制活化腐蚀产物均取得了良好的效果。     

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。     

秦山第二核电厂混合堆芯水化学控制技术探讨

根据秦山第二核电厂混合堆芯的特点,对其一回路水化学优化控制方法进行了研究,提出采用4段的硼-锂(B-Li)协调方案控制最高锂浓度,以降低反应堆结构材料的腐蚀风险.对不同燃料循环腐蚀活化产物进行跟踪分析,结果表明:在混合堆芯运行期问,采用优化的B-Li协调水化学控制对抑制一回路结构材料腐蚀和降低一回路辐射场是有效的.     

压水堆核电站一回路~(16)N源项计算模型的优化

~(16)N是压水堆一回路冷却剂中的主要活化产物,也是一回路中的主要辐射源。本文在传统~(16)N源项计算模型的基础上,根据堆芯内冷却剂的流向,考虑堆芯区域以及下降段区域的中子通量差异,将堆芯划分为活化区域以及反射区域,并建立了相应的计算模型,以典型三代压水堆核电站为例进行了计算与验证,计算结果与技术文件吻合良好,偏差在10%以内,验证了模型的正确性。最后分析了一回路典型部位的~(16)N平衡放射性活度浓度,发现在反应堆堆芯出口处最高,随着冷却剂流向逐步减少。研究结果表明,优化的计算模型可更准确计算压水堆核电站冷却剂的~(16)N源项,为分析反应堆一回路的辐射源项提供参考依据。     

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。     

华龙一号活化腐蚀产物沉积源项评估

根据国内外核电厂主管道上沉积源项的运行经验数据,分析了两种主要核素Co-58和Co-60的沉积活度随电厂运行时间的变化趋势。在此基础上,采用一回路活化腐蚀产物源项计算软件预估了华龙一号的活化腐蚀产物沉积源项。在参考国内广泛运行的M310机型设计源项确定方法的基础上,分析给出了华龙一号活化腐蚀产物沉积源项的设计源项和现实源项,并与国内二代核电机组和国际三代核电机组进行对比,结果显示三者均处于同一量级水平,华龙一号与国际三代核电机组相差不大,且优于国内二代核电机组。分析结果显示本文预估的沉积源项具有一定的可靠性,华龙一号核电机组在活化腐蚀产物源项控制方面具有一定的先进性。     

压水堆核电厂采用富集硼的优势与可行性分析

压水堆核电厂通常采用天然硼进行反应性的化学补偿控制,研究表明采用富集硼替代天然硼,可以优化一回路水化学,降低一回路结构材料腐蚀风险和堆芯沉积风险,降低职业照射剂量。本文分析压水堆核电厂采用富集硼的优势与可行性,介绍国内外核电厂富集硼的应用情况。最后对我国华龙一号堆型和在役压水堆核电厂富集硼的应用提出建议。     

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。     

混合堆芯中腐蚀产物沉积和一回路放射性水平的影响研究

为提升核电站的经济效益和满足高燃耗的需要，越来越多先进高效的核燃料组件被研究制造，并采用与原有组件混合布置的方式进行性能评估，从而形成混合堆芯。燃料污垢（CRUD,Chalk RiversUnidentifiedDeposit）的沉积及一回路冷却剂中源项水平对于大修人员辐射防护有着重要影响，由于两种组件结构和材料上的差异，造成混合堆芯热工水力条件的变化，可能会对一回路腐蚀产物沉积和一回路放射性水平造成一定影响。本文针对某压水堆引入改进型AFA-3G燃料组件的过渡循环，对比计算了不同混合堆芯方案下的冷却剂源项和主管道停堆沉积源项（dose rate）水平。计算结果表明，不同混合堆芯方案下一回路冷却剂源项变化较小，引入改进型AFA-3G燃料组件对于降低主管道dose rate具有积极作用。     

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.     

一种基于混合传导模型的一回路结构材料腐蚀-活化-迁移模型及其应用

为了定量分析反应堆冷却剂加锌工艺对一回路系统堆芯外放射性水平的影响,本文结合描述材料微观腐蚀过程的混合传导模型(MCM)和描述腐蚀产物活化、迁移及沉积的宏观输运模型,形成了能够系统性描述一回路结构材料腐蚀-活化-迁移的联合模型,并通过遗传算法分析及文献调研确定模型各主要参数。经校验表明该模型能够有效计算正常运行工况下一回路中结构材料的均匀腐蚀程度,同时也能给出结构材料表面沉积层的放射性活度分布。使用该模型对加锌前后系统内不同分区的活度分别进行了计算,结果表明加锌工艺能显著降低一回路堆芯外放射性水平。     

5MW核供热堆的水化学实验研究

实验研究了５ＭＷ核供热实验反应堆的水化学特性。该堆是一体化自稳压自然循环壳式轻水反应堆，测量了一回路冷却水水质的化学成分，如溶解氧、ｐＨ值、电导率、硝酸根、氯离子、氟离子及腐蚀产物、溶解氢等。讨论了反应堆结构对水化学的影响。     

Fission product behaviour and graphite corrosion under accident conditions in the HTR

This paper primarily gives an overview of methods and data in source term estimations for the HTR with pebble bed core. For medium size HTRs the risk dominating accidents are tied to core heat-up events, where a significant portion of the fission product inventory may be released from the coated fuel particles. Here the research mainly is focused on temperature-induced coated particle failure and the interaction of metallic fission products with the core graphite. For small HTRs, with their limitation of maximum temperatures below coated particle failure limits, core heat-up accidents virtually play no role with respect to source terms. Here the risk is dominated by accidents like water ingress or rapid depressurization which may lead to a partial release of fission products accumulated on primary circuit surfaces like the steam reformer. Deposition of fission products and remobilization under the conditions mentioned above are predominant research areas. It can be expected that the ongoing and planned improvements of models and data base, in particular for the medium size HTR, will result in a further reduction of the already low source terms.A principal possibility for core degradation and hence destruction of fission product barriers is graphite corrosion caused by massive air ingress. The research effort in this field as well as for graphite corrosion during water ingress accidents is described in Part B of this paper. From the viewpoint of risk for this type of accident no significant contribution to that of present reactor concepts was found.     

Doses from accidental releases of tritium and activation products into the atmosphere

In view of public acceptance and the licensing procedure of projected fusion reactors, the release of tritium and activation products during normal operation as well as after accidents is a significant safety aspect. Calculations have been performed under accidental conditions for unit releases of corrosion products from water coolant loops, of first wall erosion products including different coating materials, and of tritium in its chemical form of tritiated water (HTO). Dose assessments during normal operation have been performed for corrosion products from first wall primary coolant loop and for tritium in both chemical forms (HT/HTO). The two accident consequence assessment (ACA) codes UFOTRI and COSYMA have been applied for the deterministic dose calculations with nearly the same input variables and for several radiological source terms. Furthermore, COSYMA and NORMTRI have been applied for routine release scenarios. The paper analyzes the radioation doses to individuals and the population resulting from the different materials assumed to be released in the environment.D.T.I. Dr. Trippe Ing. GmbH, Karlsruhe.     

Modeling the effects of nuclear environments on crevice corrosion of pure nickel by coupling of phase and polarization behavior

As from long-term operating experience the high purity primary water cycle of light water nuclear reactors may exhibit excursions from the recommended water chemistry leading to potentially favorite conditions for stress corrosion cracking (SCC) which may be initiated and its propagation controlled by local pitting and crevice corrosion. Deterministic modeling of local corrosion including incubation times for crevice corrosion should therefore provide a basis for lifetime predictions of components, which have been subjected to sporadic intermediate water chemistry fluctuations. Based on previous work for room temperature (RT), the chloride-induced crevice corrosion at 288 °C of pure nickel as an important base element in respective high alloyed nuclear materials is modeled by coupling anodic polarization with the precipitation of nickel oxide and nickel chloride calculated from the water–hydrogen–nickel chloride heterogeneous phase equilibrium diagram. The surface corrosion potentials are fixed by bulk levels of hydrogen and oxygen contents as well as pH simulating hydrogen treatment of irradiation subjected cooling water for the reduction of corrosion potentials and mitigation of SCC at operating temperature 288 °C in Boiling Water Reactors (BWRs). Assuming chemical equilibrium conditions during the selected time steps in a relevant component crevice the calculated change of the crevice solution composition is quantitatively shown to initiate crevice corrosion by the breakdown of the passive nickel oxide layer followed by the formation of non-passive nickel chloride and the subsequent acidification of the crevice solution. The effects of corrosion potentials, bulk levels of pH and chlorides, are investigated. As a result, the reduction of corrosion potentials and increase in bulk pH provide significant increases in the passive layer breakdown times and acidification times inside the crevice. Depending on bulk pH and corrosion potentials the reduction of bulk chlorides down to recommended levels in BWRs retards crevice corrosion significantly. For a standard 100,000 h time for crevice acidification to locally less than pH = 0 the respective chloride–pH domain is evaluated. Such diagrams may be related to respective effects on stress corrosion cracking and its mitigation by hydrogen water chemistry (HWC).     

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。     

Decrease of the Corrosion Products in Water Coolant – a Basic Method for Increasing the Efficiency and Improving the Safety of Nuclear Power Plants

The corrosive behavior of oxidized steel in water to which inhibitors or oxygen have been added is studied. It is shown that preoxidation of steel makes it possible to decrease the amount of inhibitors or oxygen. The experiments have shown that the functional possibilities of an oxygen-neutral water regime can be expanded to achieve effective protection. The corrosion resistance of steel can be increased manyfold by managing the water chemistry correctly. Likewise, the concentration of corrosion products in the coolant and, correspondingly, their activation in the neutron flux of a nuclear reactor and the amount of radionuclides can be decreased.     

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。