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.     

Fatigue life estimation for internal threads in class 1 components

Heat exchangers, steam generators and other pressure vessels in nuclear power plants are equipped with bolted closures for the purpose of in service inspection and maintenance. The ASME Boiler and Pressure Vessel Code specifies that all Class 1 components meet the fatigue life requirements for Level A and B Service Conditions. In the case of bolted closures, it is often found that the bolt/stud is the most critical part. In many situations, the bolts fail to meet the fatigue requirements for the design life of the equipment. In such cases, the bolts can be replaced after certain duration based upon their fatigue life. However, the mating threads in the flange (which is an integral part of the vessel) are still a concern. While the replacement of the bolts is relatively easy and inexpensive, the corrective action (e.g. replacement or repair) for the flange is usually difficult and expensive, or impossible. Hence, it is important to have a reasonable estimate of the fatigue life of internal threads to alleviate or minimize the concern. In this paper, a simplified approach is presented for this purpose. Considering various bolt sizes, commonly used thread series and typical Class 1 component materials, it is shown that the fatigue life of the internal threads is about three times the fatigue life of the bolt threads. This conclusion greatly reduces or eliminates the concern for in service replacement or repair of the components with internal threads.     

核电厂一回路压力边界止回阀在线密封性能测试

320 MW压水堆一回路压力边界止回阀为核Ⅰ级关键设备,严密性要求非常高,直接关系到主系统的内泄漏率.焊接式止回阀维修后常采用密封面色印检查的方式,对其密封性能进行判断.如果管道内有存水或者湿热水汽,会影响到色印检查的准确度.针对在线止回阀密封性试验的特殊性,有的核电厂采用水压压降法试验设计过在线检测装置,但存在一些缺点和使用上的限制.文章采用低压气密封试验流量测定法,设计出可靠、便携的试验装置,对压力边界止回阀检修后密封性做出准确、定量的判断.     

核电厂电仪设备延寿再鉴定试验研究

在核电厂电气仪表设备（简称电仪设备）环境鉴定研究成果的基础上,开展核电厂电仪设备延寿再鉴定分析和试验研究。以秦山第一核电厂DDG-1型电气贯穿件（EPA）为研究对象,根据运行实际制定了再鉴定试验研究的遵循原则,在此原则下结合分析法确定了试验方案和试验项目序列以及EPA修复依据和方案,并在此基础上开展再鉴定试验研究。适当修复后的DDG-1型EPA按试验大纲依次通过了设备性能随时间变化的试验、抗震试验、设计基准事故（DBA）条件下热力学试验和DBA后极限电性能试验,试验后状态完好,表明该DDG-1型EPA经适当修复后能够完成继续延寿20 a的预期目标,可为核电厂其他电仪设备再鉴定试验研究提供指导和借鉴。      

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。      

Use of safety valves in the first loop of a nuclear power plant with a water-moderated-water-cooled power reactor

Conclusions It is not useful to transfer the requirements of the supervising committees on the safety of the usual vessels and high-pressure systems against their exceeding the allowable value of the pressure in nuclear installations with the conditions of radioactivity in the first loop of nuclear power plants, especially those onboard a vessel, with such a system as the active zone, which does not permit exposure and prolonged liberation of heat. Protection against an increase in pressure can be more effectively provided by reliable systems for shutting down the reactor, removing heat, and others, which is indicated by the comparative experience in using nuclear-powered icebreakers and some nuclear power plants abroad.Translated from Atomnaya Énergiya, Vol 50, No. 5, pp. 308–310, May, 1981.     

破损燃料组件修复后的物理特性分析

破损燃料组件修复后再次入堆使用是必须进行安全评估,以确保核安全。本文以采用AFA3G燃料组件的CPR1000机组为研究对象,对装入反应堆后的正常燃料组件和修复燃料组件的核物理和功率分布进行分析评估。结果表明:燃料组件内更换一根燃料棒对燃料组件反应性的影响很小,该影响可以忽略。更换不锈钢棒的数量越大,燃料组件反应性变化幅度越大。随着燃耗的加深,燃料组件反应性变化幅度也增大。修复的燃料组件虽然在换棒位置局部区域发生功率畸变,相对功率略微的升高,但离换棒位置较远的燃料棒的相对功率没有变化,换棒不会导致组件内功率峰发生象限的偏移。     

堆内捕集器对压力容器中子辐照影响分析

严重事故下为实现堆内熔融物滞留,可采用堆内捕集器(IVCC)的策略。捕集器属压力容器的一部分,属不可更换设备,需长期在堆内受中子辐照。本文通过对典型压水堆压力容器模型和带IVCC的压力容器模型的比较,发现IVCC不会改变压力容器内快中子通量,不会对压力容器的辐照造成影响。且IVCC使得压力容器的热中子通量明显减小,降低了压力容器的整体辐照水平。这说明IVCC对压力容器的辐照性能不会产生不利影响,反而有助于防止压力容器的老化。     

Evaluation of the burst characteristics for axial notches on SG tubings

Some events of steam generator tubes have been reported in some nuclear power plants around the world. Main causes of the leakage are from various types of corrosion in the steam generator (SG) tubing. Primary water stress corrosion cracking (PWSCC) of steam generator tubing have occurred in many tubes in Korean plants, and they were repaired using sleeves or plugs. In order to develop proper repair criteria, it is necessary to ascertain the leak behavior of the tubings. A high-pressure leak and burst testing system was manufactured. Various types of electro-discharged-machined (EDM) notches having different lengths were machined on the o.d. of test tubes to study SG tube behavior. Leak rate and ligament rupture pressure as well as the burst pressure were measured for the tubes at room temperature. Rupture pressure of the part through-wall defect tubes depends on the defect depth and length. Water flow rates after the rupture were independent of the flaw types; tubes having 20–60 mm long EDM notches showed similar flow rates regardless of the initial defect depth. A fast pressurization rate generated a lower burst pressure than the case of a slow pressurization.     

Operating performance and reliability of CANDU PHWR fuel channels in canada

CANDU nuclear plants use many, small-diameter high-pressure fuel channels unlike PWR nuclear plants which have a single, large pressure vessel. Good operating performance from the CANDU fuel channels has made a major contribution to the world-leading operating record of the CANDU nuclear power plants. As of 1982 December 31, there were 7,480 fuel channels installed in 18 CANDU reactors over 500 MW(e) in size. Eight of these reactors have been declared in-service and have accumulated 24,000 fuel channel-years of operation. The only significant operating problems with fuel channels have been the occurrence of leaking cracks in 70 fuel channels and a larger amount of axial creep on the early reactors than was originally provided for in the design. Both of these problems have been corrected on all CANDU reactors built since the Bruce GS ‘A’ station and the newer reactors should exhibit even better performance.     

新燃料组件运输容器加速度计跳离事件的处理方法研究

新燃料组件运输容器上的加速度计是用于监测燃料组件运输过程中的异常冲击。加速度计的跳离表示在运输过程中可能存在对燃料组件产生损坏的载荷。近年来,国内核电站发生多起新燃料组件运输容器的加速度计跳离事件。发生运输容器的加速度计跳离事件后,需对事发燃料组件的机械完整性以及可用性进行评估,并判断其是否可入堆使用。本文在对加速度计的作用原理及加速度计跳离过程进行深入分析基础上,提出了一种新燃料组件运输容器加速度计跳离事件的通用处理方法。利用该通用处理方法对某核电站近年来发生的新燃料组件运输容器加速度计跳离事件进行了处理,处理结果得到了业主的采纳。本文中提出的加速度计跳离事件通用处理方法,可为国内核电站后续加速度计跳离事件的处理提供重要的参考和依据。      

Corium retention for high power reactors by an in-vessel core catcher in combination with External Reactor Vessel Cooling

If there were inadequate cooling during a reactor accident, a significant amount of core material could become molten and relocate to the lower head of the reactor vessel, as happened in the Three Mile Island Unit 2 (TMI-2) accident. If it is possible to ensure that the vessel lower head remains intact so that relocated core materials are retained within the vessel, the enhanced safety associated with these plants can reduce concerns about containment failure and associated risk. For example, the enhanced safety of the Westinghouse Advanced 600 MWe Pressurized Water Reactor (AP600), which relied upon External Reactor Vessel Cooling (ERVC) for in-vessel retention (IVR), resulted in the United States Nuclear Regulatory Commission (US NRC) approving the design without requiring certain conventional features common to existing Light Water Reactors (LWRs). Accordingly, IVR of core melt is a key severe accident management strategy adopted by some operating nuclear power plants and proposed for some advanced light water reactors. However, it is not clear that currently-proposed methods to achieve ERVC will provide sufficient heat removal for higher power reactors. A US–Korean International Nuclear Energy Research Initiative (INERI) project has been initiated in which the Idaho National Engineering and Environmental Laboratory (INEEL), Seoul National University (SNU), Pennsylvania State University (PSU), and the Korea Atomic Energy Research Institute (KAERI) will determine if IVR is feasible for reactors up to 1500 MWe. This paper summarizes results from the first year of this 3-year project.     

General review of recent activities in acoustic emission technology in Italy

This paper describes the present situation in Italy in the field of Acoustic Emission researches and applications.Information on the level of instrumentation development is given. Both multichannel and multiparameter systems for large structure examination in real time and data logging systems for continuous surveillance purposes are considered.The expertise accumulated in the application of AE to pressure vessel examination during hydrotest is mentioned, this being oriented to pressure components of conventional power stations and chemical plants.Particular attention is recently paid to mechanical fatigue tests. These were conducted on intermediate PWR nuclear pressure vessel, reduced scale offshore nodes and full scale prototype aircraft.A considerable activity has been carried out on application of AE technique to the detection of fluid leakages in power plant components. Both intrusive and non-intrusive methods have been considered. Many boilers and pre-heaters of thermal power plants have been instrumented for an on-line AE monitoring during operation. The problem of the loose part monitoring has been also considered.Several basic researches for material characterization by AE have been also conducted. Different composite material, carbon and austenitic steels, metal alloys have been studied.     

Intermediate size pressure vessel programme (ZB 1)

One of the focal points in the discussion about the safety of nuclear power plants is the integrity of the reactor pressure vessel.In order to prove its integrity tests are in progress in an underground test facility of the main power station in Mannheim with an intermediate size vessel from the research programme “Integrity of Components”. Patches of A 533 B and modified A 508 B material were welded into the vessel ZB 1, the test temperatures are approximately 70 and 290°C. The main goal of the tests is to measure the behaviour of artificial and natural flaws during static hydrotests and simulated operational (cyclic) conditions.In the first half of the research programme the objective is to produce a crack growth of some centimetres by cyclic loading between a variable minimum pressure and a maximum pressure of about 24 MPa. The total number of load cycles will be approximately 30 000.In the second half of the tests the vessel will be loaded by a number of pressure cycles which correspond to the loading a reactor pressure vessel experiences during 40 years of operation.During the static and cyclic loading acoustic emission monitoring is being made by German and American laboratories.This paper presents details of the vessel, the test loop, results of the nondestructive examinations conducted to quantify the crack depths and results of the acoustic emission monitoring.     

国内在役核电厂首例电气贯穿件更换改造

核电厂电气贯穿件作为安全壳上的关键设备,承担着核岛内外各种电力和信号传输以及保证安全壳压力边界完整性的重要功能。通过秦山核电厂一期工程30万千瓦机组第18次大修期间国产在役DDG-1型电气贯穿件更换改造项目的实施,分析了秦山核电厂一期工程在役电气贯穿件设备现状和改造的必要性;针对在役核电厂更换改造工期短和贯穿件密封性能验证难等问题,通过优化检验工序、制作专用检漏工装的方法,缩短了贯穿件改造的工期并验证了贯穿件密封性能。     

用小尺寸R－CT试样研究核压力容器钢断裂韧性及与Charpy试样的比较

提出了一种用双边带深侧槽的小尺寸圆形紧凑拉伸试样评定核压力容器（ＲＰＶ）钢断裂韧性的单试作试验方法，给出了用该方法测定的两个厂家生产的核压力容器用Ａ５０８ＣＬ３钢的断裂韧性参数，还与Ｃｈａｒｐｙ试样的试验结果及大尺寸标准试样的试验结果进行了比较。研究结果表明：用双边带深侧槽的小尺寸Ｒ－ＣＴ试样测得的断裂韧性值比相同恻槽深度预制疲劳裂纹Ｃｈａｒｐｙ试样的测试值更接近有效断裂韧性值，所以，用于核压力容器断裂韧性的监测是可行的。     

核电厂反应堆压力容器顶盖开关的过程控制

对于核电厂的建设和维护,反应堆压力容器顶盖开关工作是一个关键、繁琐且容易被人忽视的工序。很多的核电厂建设往往由于对此不够重视而吃尽苦头。恰希玛核电站二期项目（简称C2项目）,由于准备得相对充分,所以在这方面取得了一定的成果,这其中有很多做法和经验值得总结和借鉴。     

AP1000依托项目中压移动电源应用研究

福岛核事故后,世界各国对核电厂的安全十分关注,外部事件是否会引起核电厂大量放射性释放是关注的焦点.2012年,国家核安全局发布了《通用技术要求》,要求核电厂设置中压移动电源.针对AP1000依托项目非能动安全特性,选择中压移动电源为1E级蓄电池充电、需要时为正常余热排出系统(Normal Residual Heat R...     

Lead-bismuth cooled fast reactor for regional power generation

A condition for adopting low-capacity nuclear power plants in regional power generation is that they be competitive with thermal power plants. However, it is much more difficult to make low-capacity plants competitive than the power-generating units of high-capacity nuclear power plants, because as the capacity of a power source decreases the specific capital investments and power generation costs increase much more rapidly. It is shown that the innovative nuclear power technology of lead-bismuth cooled fast reactors, such as SVBR, based on the experience gained in operating nuclear submarines with chemically inert lead-bismuth coolant, which does not require high pressure in the first loop, satisfies all requirements for low-capacity nuclear power plants for regional power generation.     

反应堆压力容器内壁面异种金属焊接缺陷的修复及力学评价

某核电站反应堆压力容器(RPV)制造期间超声检测（UT）显示,顶盖法兰内壁面堆焊层熔合线附近出现大范围连续焊接缺陷,环向跨度大于8°,造成大范围低合金钢母材减薄。针对上述缺陷的产生开展了根本原因分析,结合技术现状给出补焊不锈钢的修复方案并展开详细的力学评价,从应力、疲劳和密封角度分析该缺陷对RPV性能的影响,论证了该修复方案的可行性。补焊不锈钢方案已得到工程应用,可为工程上类似问题的处理提供借鉴。