核电厂水压试验设备设计方案优化研究

核岛容器水压试验是核电厂在役检查的重要方法之一,能够验证容器在持续承压状态下的完整性和密封性。根据不同类型容器对试验临时特殊装置的安装要求,在设计阶段优化容器本体及管道布置方案,从而降低在役水压试验人员受照剂量、缩减试验工期、减少对设备和相关附件的破坏,降低核电厂检修成本。      

10 MW高温气冷堆反应堆压力容器的出厂水压试验

根据ASME规范第Ⅲ卷NB－6200节的规定，对10MW高温气冷堆压力容器的水压试验要求、试验过程、试验结果及评价进行了叙述。用清华大学核能技术设计研究院研制的液压张拉机对主螺栓实施了合理及有效的张拉，对压力容器进行了应变和不变形测量，取得了反应堆压力容器水压试验的圆满成功。     

反应堆压力容器密封环国产化替代研究

C型密封环是保证反应堆压力容器顶盖与筒体密封的核心部件，其密封性能直接关系到核电厂安全稳定运行。长期以来，C型密封环制造技术被外国公司垄断，单套售价高，供货周期长。通过密封特性试验、氦气检漏试验、水压试验和冷热循环试验，验证了国产C型密封环具备商用条件。通过功能影响分析论证了国产C型密封环的等效性。核电厂严控国产C型密封环安装质量，并通过在役水压试验、运行试验验证了国产C型密封环的性能，实现了国产C型密封环商用。      

方家山核电工程一回路主设备役前检查实施优化

核电厂一回路主设备在首次装料前必须完成役前检查,役前检查一直以来被安排在一回路水压试验后实施。方家山核电工程根据工程进度情况,通过分析对比法规标准,探讨主设备役前检查与主系统水压试验之间的关系,打破以往的观念,创造性地将主设备役前检查的时机进行优化,并通过国家核安全局(NNSA)审评。按此优化方案,方家山核电工程在国内首次在一回路水压试验前实施了主设备役前检查,不但保证了水压试验顺利进行,还缩短检查工期,有效达到了优化目标。     

压力容器水压试验压力及其利弊分析

对现行压力容器规范规定的压力容器水压试验的压力、利弊等进行讨论,指出水压试验压力约为塑性失稳压力的40%~45%,其应力准则允许的压力约为塑性失稳力的50%~75%;水压试验是压力容器检漏、强度验证的一种实用有效的试验方法和检查技术;水压试验可以改善和提高压力容器承载能力,减小破坏可能性;在役水压试验可以确定可能存在的最大缺陷或最大承受的压力,为后续安全运行和分析提供依据和数据,但压力不宜超过首次强度水压试验压力;对超载和温态预应力水压试验以及水压试验有害作用应进一步开展科研。     

M310堆型核电站役前检查与系统水压试验实施时机探讨

通过对核安全导则HAD103/07、RSE-M规范、ASME规范第XI卷和法国1999法令等相关文件的解读,结合目前国内压水堆核电站役前检查的工程实践,分析役前检查与水压试验的相互影响,形成对机组一回路役前检查与系统水压试验实施时机的技术方案。     

AP1000与二代压水堆核电厂的严重事故预防与缓解策略比较

根据压水堆核电厂严重事故发生机理,基于高压堆熔、压力容器失效以及安全壳失效三个关键阶段,针对AP1000和二代核电厂进行比较,在系统结构设计上分析两者在严重事故预防与缓解策略方面的异同,最后对我国在役核电厂的严重事故预防与缓解提出建议。     

AP1000与二代压水堆核电厂的严重事故预防与缓解策略比较

根据压水堆核电厂严重事故发生机理,基于高压堆熔、压力容器失效以及安全壳失效三个关键阶段,针对AP1000和二代核电厂进行比较,在系统结构设计上分析两者在严重事故预防与缓解策略方面的异同,最后对我国在役核电厂的严重事故预防与缓解提出建议。     

十年大修项目的运行风险分析和风险控制

岭澳核电站L101大修(1号机第一次大修)实施了十年项目安全壳打压试验,L201大修(2号机第一次大修)实施了一回路水压试验、压力容器在役检查和安全壳打压试验三大十年项目。L201大修是大亚湾和岭澳两电站第一次完整实施代表性十年大修项目的大修,其中一回路水压试验为国内商运核电站首次实施。在研究各试验要求和进行风险分析的基础上,对核电站十年大修项目的运行活动和试验活动进行了有效的风险控制,编写了一整套十年大修专项总体运行程序,建立了以运行程序为主线,通过主隔离、TSD、TCA和工作票管理,有效地控制各项活动风险的运行控制体系,解决了不同试验状态之间以及试验和机组启动的接口问题,为十年大修的顺利完成打下了坚实的基础;在大修过程中,证明了这套程序和体系的有效性和科学性,这些成功经验为以后两电站及国内核电站十年大修提供了参考,具有广泛的社会价值。     

关于颁布《核电厂核岛机械设备无损检验规范》等两项核行业标准的通知

各有关单位：现将《核电厂核岛机械设备无损检验规范》等两项推荐性行业标准予以颁布，自1997年5月1日起实施，标准文本由核工业标准化研究所负责出版发行。附件：标准目录中国核工业总公司一九九六年十二月三十日附件标准目录序号标准号项目名称负责起草单位1EJ／T1039－1996核电厂核岛机械设备无损检验规范核工业无损检测中心2EJ／T1040-1996核电厂核岛设备材料理化检验方法上海核工程研究设计院关于颁布《核电厂核岛机械设备无损检验规范》等两项核行业标准的通知     

Proof pressure tests of the first reactor vessels at Hartlepool and Heysham nuclear power stations

The nuclear stations currently nearing completion at Hartlepool and Heysham are the world's first design of Advanced Gas Cooled Reactor to use podded boiler construction. The proof pressure tests on the first reactor pressure vessels at both stations were carried out during the early part of 1980. The object of the proof pressure test is to satisfy statutory and contractual requirements by demonstrating the integrity of the completed concrete pressure vessel, liners and pressurised penetrations when subjected to a pneumatic test pressure of 740 lbf/in² g, equivalent to 1.15 times the design pressure of 644 lbf/in² g. Testing was carried out in the latter period of the construction phase of each station with the majority of the plant and buildings in an advanced state of construction and the reactor internals essentially complete. This paper examines the results of both reactor 1 vessel tests and compares the behaviour of the vessels with one another and with analytical predictions. Possible sources of instrumentation error are identified and discussed. The conclusions are drawn from correlation of the test results with theoretical predictions.The vessel pressure was raised and lowered in six stages to and from the test pressure with a hold period between each stage to read instruments, examine the vessel surfaces and leak check the penetrations and closures. To obtain the relevant information at each pressure hold, the vessels' permanent instrumentation, consisting of vibrating wire strain gauges and their associated electrical resistance thermometers, and the liner and reactor internal thermocouples were monitored. In addition, strain gauges fixed to the concrete face of the liner during construction were monitored and liner strains measured during the test were compared with predictions. Vessel deflections were monitored by special equipment mounted on the reactor building to enable vessel dilation to be measured. The vessel top and bottom caps were instrumented by manometric systems in order that the cap profiles and deflections could also be determined.The measurements and observations made during the proof pressure tests adequately demonstrated that the vessels behaved in accordance with predictions. The concrete strains and vessel external deflections confirmed that the vessels behaved in a linear and elastic manner throughout the tests, with no cracking being observed in the concrete during the tests. Liner strains compared favourably with predicted values, exhibiting a linear behaviour under increasing pressure. The strain levels recorded gave complete confidence in the liner design.The tests confirmed the integrity of the vessels, thus enabling them to go forward to the engineering run stage in the commissioning programme leading ultimately to the raising of power.     

秦山二期反应堆压力容器出厂水压试验

依靠国内技术与设备手段,成功完成了秦山核电站二期工程(QS-II)2号机组反应堆压力容器出厂水压试验的应变测试,给出了水压试验中各测点应变的测试结果。结果表明,在22.8MPa高压水下多数测点为弹性变形,容器进出口管嘴270o边缘变形局部进入屈服,但接近日本三菱重工对1号机组反应堆压力容器的测量结果。与有限元分析结果进行的比较表明,2号机组反应堆压力容器应变测试有良好的精度,强度符合出厂要求。     

反应堆压力容器密封环国产化替代研究

C-ring is the core part for seal of the reactor pressure vessel top cover and cylinder, and the sealing performance is directly related to the safety and stably operation of the nuclear power plant. For a long time, the manufacturing technology of C-ring was monopolized by foreign company, with high price and long supply cycle. It is verified that the C-ring made in China fulfill the commercial application, through sealing properties test, helium leak detection test, water pressure test and thermal-cold cycle test. The functional impact analysis proved the equivalence of the domesticated C-ring. The installation quality of domestic C-ring was strictly controlled in nuclear power plant, and the performance of the domestic C-ring was verified by the in-service hydraulic test and the running test.     

核电厂CRDM耐压壳焊缝超声检验工艺设计

控制棒驱动机构（CRDM）耐压壳属于核电厂主回路,其连接焊缝是整个放射性回路压力边界的薄弱环节,其安全性和可靠性直接影响反应堆的安全运行状态。针对CRDM耐压壳焊缝附近空间狭小、壁厚薄、可达性差等特点,本文采用仿真技术设计了一套专用的扁平型双晶聚焦超声探头和检验工艺,试验验证结果满足规程要求,解决了核电厂在役检查的监督难点,并获得了核电厂主回路Ⅰ级部件类似焊缝检验的工艺设计和验证方法。      

核电厂内部水淹确定论安全评价方法及应用

为确保核电厂在内部水淹情况下仍能保证安全功能,有必要在核电厂设计时考虑对内部水淹进行防护,并对核电厂进行内部水淹安全评价以验证内部水淹防护目标的实现。通过对内部水淹防护目的、要求以及内部水淹防护措施进行研究,探索和提出内部水淹确定论安全评价的设计假设、方法和步骤。以某百万kW级核电厂硼酸输送泵间、反应堆和乏燃料水池冷却和处理系统泵间为例对方法进行了应用和验证,分析表明无需对硼酸输送泵间进行特殊防护,但须对反应堆和乏燃料水池冷却和处理系统泵间采取适当的疏水措施以确保核安全功能的实现。     

核电厂BOSS焊缝在线堆焊修复相控阵超声检验技术

核电厂反应堆换料水池与乏燃料水池冷却和处理系统（PTR）及设备循环冷却系统（RRI）中含有大量管座接头（BOSS）焊缝,其安全性和可靠性直接影响所存储核燃料的安全状态,对其进行缺陷排查和在线修复是核电厂在役检查监督的重点和难点。本文针对BOSS焊缝在线堆焊修复的特殊要求和检验难点以及射线检验的局限性,设计了一套专用的相控阵超声探头和检验工艺,试验验证结果满足堆焊修复要求,并制订了核电厂BOSS焊缝堆焊修复无损检验的方法和在役检查监督的策略。     

Department of energy’s annealing prototype demonstration projects

One of the challenges utilities face in addressing technical issues associated with the aging of nuclear power plants is the long-term effect of plant operation on reactor pressure vessels. These vessels are exposed to neutrons during the operation of a reactor. For certain plants, this exposure can cause embrittlement of some of the vessel welds, which can shorten the useful life of the vessel. This reactor pressure vessel embrittlement issue has the potential to affect the continued operation of a number of US pressurized water reactor plants. However, the properties that are degraded by long-term irradiation can be recovered through a thermal annealing treatment of the vessel steel. Although a dozen Russian-designed and several US military vessels have been successfully annealed, US utilities concluded that an annealing demonstration using a US reactor pressure vessel was a prerequisite before annealing a licensed US nuclear power plant. In May 1995, the Department of Energy and Sandia National Laboratories initiated a program to evaluate the feasibility of annealing US licensed plants using two different heating technologies. One team completed its annealing prototype demonstration in July 1996, using an indirect gas-fired furnace at the uncompleted Public Service of Indiana’s Marble Hill nuclear power plant in southern Indiana. The second team’s annealing prototype demonstration using a direct heat electrical furnace at the uncompleted Consumers Power Company’s nuclear power plant at Midland, Michigan, was scheduled to be completed in early 1997, but has now been delayed indefinitely. This paper describes the Department of Energy’s annealing prototype demonstration program and the results to date for each project.     

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。