用RT_(T_0)取代RT_(NDT)分析国产某核压力容器压力-温度限值曲线的效益

ASME规范和美国联邦法规规定了核反应堆压力容器(Reactor pressure vessel,RPV)在正常启、停堆过程中及水压试验时的压力和温度限值,2013年版ASME规范直接纳入了Code Case N-629,即同时接受了RTT0和RTNDT两种参考温度表征的材料断裂韧性KIc下包络线。本文对比分析采用KIR-RTNDT、KIc-RTNDT和KIc-RTT0三种断裂韧性取值方法所确定的压力-温度限值曲线(P-T曲线),以国产某台RPV为研究对象,计算了在40年设计寿期末和延寿期的P-T曲线。结果表明三者差别很大,基于KIc-RTT0下包络线拓宽了P-T运行窗口,甚至无需担心该容器在启停堆过程中会发生脆断,KIc-RTNDT曲线的计算结果偏保守,而由KIR-RTNDT给出的结果过于保守。研究结果为该电站的运行和延寿的可能性提供了支持。     

主曲线方法在核电厂压力容器老化延寿中的应用

反应堆压力容器的压力-温度限值曲线（P-T限值曲线）方法是确保压力容器完整性的重要方法，在处理压力容器老化延寿问题中有着重要意义。传统的方法利用由t-R_TNDT曲线表征的材料准静态断裂韧性限值（K_Ic）绘制P-T曲线，这种方法不能直接测量材料辐照后的材料无延性转变温度的参考温度（R_TNDT），且过于保守。本文针对某核电厂压力容器，利用现有的辐照监督管数据估计50 a延寿期末主曲线参考温度R_TT0，并采用ASME Code Case N629中的主曲线应用方法，计算寿期末的P-T限值曲线。与传统方法得到的P-T限值曲线相比，利用主曲线方法可以得到更大的运行窗口，能够提高设备的经济性。      

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     

新颁核行业标准

序号D＆准县】株准名称I实施日期11EJ807—94铀矿冶工作人员辐射防护管理规定11994年8月1日ZIEJ810—94压水堆核电厂一次启动中子源＄131EJ/T808—94铀燃料元件厂设计准则14IEj/T809—94铀燃料元件厂抗震设计分级I50EJ/T811——941W衡算管理导则〔61EJ/T812——94氖衡算管理导则l7lEJ/T813—94核工业气体离心分离术语ISIEJ/T814—94【o矿石中钎的测定PMBP苹取分离们氖肿＊分光光度法1giEJ/T815一94]六氟化铀中钦的分光光K＆NM10EJ/T816—94压水堆核电厂应急堆芯冷却地坑设计准则l11EJ/T817—94压力堆核电厂燃料组件包袭…     

池式研究堆衰变热计算与实验研究

采用量热法测量反应堆额定功率运行75.0 h停堆后45 h内的衰变热功率,拟合出归一化衰变热功率的经验关系式.与反应堆衰变热几种半经验公式和标准对比结果表明,实验结果在经验公式计算值范围内,并与EJ/T 745标准预测值符合较好.     

反应堆压力容器辐照脆化状态评估

基于大量相似辐照脆化试验测试数据和实际辐照监督测试数据,采用统计分析的方法,选出适合于某核电厂反应堆压力容器（RPV）的辐照脆化评估公式。以该核电厂已经完成的辐照监督管测试数据为输入,对RPV当前的辐照脆化状态进行了评估,并推算、分析了RPV在寿期末的结构完整性;基于辐照脆化计算结果,绘制了各运行阶段RPV的压力-温度限值曲线（P-T曲线）,并给出运行建议。      

压水堆核电厂压力容器开盖时刻主冷却剂放射性浓度控制要求研究

压水堆核电厂停堆开盖时刻主冷却剂放射性浓度限值是核电厂的重要设计参数。本文基于停堆开盖后厂内辐射风险来源分析,建立了适用于压水堆核电厂停堆压力容器开盖时刻主冷却剂中的放射性浓度控制值评估方法,并采用欧洲第三代压水堆技术方案(EPR)堆型核电厂的设计参数对建立的方法进行了验证。验证结果表明:基于此方法得出的停堆开盖限值与EPR堆型核电厂原设计较接近。     

CAP1400 MSLB事故破口谱分析及极限工况瞬态分析

CAP1400核电厂与传统的"二代"核电厂区别较大。CAP1400反应堆在AP1000的基础上进行了一系列改进。采用RELAP5/MOD3.3程序建立CAP1400核电厂模型,对主蒸汽管道破裂事故的破口谱进行分析,结果表明,直到0.058m~2的蒸汽管道破口都不会触发反应堆停堆。对于0.059～0.105 m~2的蒸汽管道破口,反应堆由超功率△T信号触发停堆。对于0.106～0.15 m~2的蒸汽管道破口,反应堆由蒸汽管道低压力安注信号触发停堆。从DNB和燃料中心熔化保护角度考虑,极限工况是破口尺寸为超功率触发停堆的最大破口尺寸0.105 m~2。对极限工况的热工水力瞬态进行研究,分析堆芯流量、热流密度、温度、压力等关键参数随时间变化的趋势。采用VIPRE程序对DNBR进行计算,得到事故对应的最小DNBR为1.914,大于验收准则1. 45,表明CAP1400反应堆在主蒸汽管道破裂事故下安全可靠。     

停堆期间主系统冷态超压及未预期温度变化保护操作研究

冷态超压是指核电厂冷停堆期间一回路系统水密实状态下如果发生能量或质量注入,会造成一回路系统压力升高,为保证压力容器的完整性,RHR系统设置了3组安全阀,在压力超过其整定值时开启阀门。同时事故规程要求操纵员采取措施降低一回路系统的压力,稳定正常冷停堆状态。未预期温度变化是指RCS和RHR连接后,RHR系统及其支持系统(WCC系统、WES系统)的故障可能会造成一回路温度的异常变化。当温度变化率超过规定限值时,要求操纵员采取措施稳定一回路的温度。使用RELAP软件建立了停堆模型,对冷态超压事故及未预期温度变化事故后操纵员的主要操作进行了研究计算,提出了这两类事故后的操作指导,并提出了一个停堆工况下推荐运行范围。     

核电厂提前停堆燃料经济损失定量分析

由于重大设备故障或者电网应峰度夏等原因,核电厂可能会遇到需要提前停堆换料的运行情况。针对上述情况,电厂需定量评估提前停堆造成的燃料经济损失,以便于为后续核电厂的经济运行方式提供决策依据。本文以国内某核电厂为例提出了一种可以用于核电厂提前停堆换料燃料经济损失的定量评估方法,该方法适用于停堆换料的快堆、压水堆和沸水堆。     

核电厂核级承压设备对铁素体材料的断裂韧性基本要求

本文概要介绍了铁素体材料构件的低温脆断的理论基础和抗脆断设计．总结并评价了ASME规范中对核电厂核安全级别承压设备铁素体材料抗脆断的断裂韧性要求．即核安全级别与材料的缺口冲击韧性要求之间存在相应的关系，从而保证了核电厂承压边界不会发生脆性破裂。     

A parametric study on pressure–temperature limit curve using 3-D finite element analyses

In order to operate a reactor pressure vessel (RPV) safely, it is necessary to keep the pressure–temperature (P–T) limit during the heatup and cooldown process. While the ASME Code provides the P–T limit curve for safe operation, this limit curve has been prepared under conservative assumptions. In this paper, the effects of conservative assumptions involved in the P–T limit curve specified in the ASME Code Sec. XI were investigated. Three different parameters, the crack depth, the cladding thickness and the cooling rate, were reviewed based on 3-D finite element analyses. Also, the constraint effect on P–T limit curve generation was investigated based on J–T approach. It was shown that the crack depth and constraint effect change the safety region in P–T limit curve dramatically, and thus it is recommended to prepare a more precise P–T limit curve based on finite element analysis to obtain P–T limit for safe operation of a RPV.     

Mechanism based evaluation of materials behavior and reference curves

The safety assessment of nuclear pressure vessels and piping requires a quantitative estimation of defect growth by stable and unstable manner during service. This estimation is essential for determining whether the defect detected during inspection should be repaired or whether the size of the defect even after its expected growth is small enough to leave the integrity of the vessel unaffected.The most important stable defect growth mechanism is that of environmentally assisted cyclic crack growth. Recent results indicate that it is markedly affected by sulfur content and/or manganese sulfide morphology and distribution. This implies that an essential improvement in component safety has been gained by currently applied steelmaking practices, which result in extra low sulfur content, generally below 0.01 wt%, and in round shape and small size of inclusions, through, e.g., calcium treatment, hence considerably reducing the effect of environment on crack growth rate. This further implies that the ASME Section XI reference curves for environmentally accelerated cyclic crack growth are conservative for steels produced by current steelmaking practices.The ASME Section XI applies predominantly linear elastic fracture mechanics to assess the effects of cracks on the integrity of nuclear power plant components. Unstable linear elastic fracture often propagates by cleavage mechanism. The cleavage fracture process has recently been shown to be of statistical nature in both ferritic and bainitic steels. The carbide size distribution plays a dominant role in controlling the fracture toughness of these steels. A cleavage fracture model has been developed, based on carbide induced cleavage fracture in ferritic and bainitic steels, which can be used to estimate the expected value and probability limits of fracture toughness. This method has been utilized to evaluate the conservatism of the ASME reference fracture toughness curve. For this purpose a microstructural analysis was carried out for the HSST-02 plate material, with which a large amount of K_Ic data has previously been generated for reference curve purpose. The result of the statistical evaluation indicates that based on the 95% survival probability limit some parts of the ASME reference fracture toughness curve are unconservative.     

Allowable minimum upper shelf toughness for nuclear reactor pressure vessels

The paper develops methodology and procedure for determining the allowable minimum upper shelf toughness for continued safe operation of nuclear reactor pressure vessels. Elastic-plastic fracture mechanics analysis method based on the J-integral tearing modulus (J/T) approach is used. Closed from expressions for the applied J and tearing modulus are presented for finite length, part-throughwall axial flaw with aspect ratio of . Solutions are then presented for Section III, Appendix G flaw. A simple flaw evaluation procedure that can be applied quickly by utility engineers is presented. An attractive feature of the simple procedure is that tearing modulus calculations are not required by the user, and a solution for the slope of the applied J/T line is provided. Results for the allowable minimum upper shelf toughness are presented for a range of reactor pressure vessel thickness and heatup/cooldown rates.     

Thermohydraulics in a high temperature gas-cooled reactor prestressed concrete reactor vessel during unrestricted core heatup accidents

The long term core and primary loop heatup of an HTGR subsequent to loss of all forced circulation has been analyzed using a modified version of the CORCON code. The results indicate that if the liner cooling system is operating, or can be restarted within about 60 h, safe cooldown can be achieved, but significant core damage will occur. Without functioning liner cooling system the core heatup will lead to PCRV concrete degradation and the resulting concrete gas releases will ultimately cause containment building failure after 6 to 10 days.     

Dynamic fracture toughness test and master curve method analysis of IAEA JRQ material

IAEA conducted a round-robin fracture test program to test and verify the Master Curve method. One of the materials selected for the round-robin is a A-533B1 plate designated as reference material JRQ. Unnotched Charpy-size specimens were fabricated and distributed to a number of testing laboratories. The three US Owners Groups received specimens for both Charpy impact and three-point bending tests to establish fracture toughness master curves. The B&W Owners Group elected to perform a dynamic fracture toughness test under a high loading rate using the JRQ specimens. The master curve method was successfully applied to numerous fracture toughness data sets of pressure vessel steels. Joyce [Small Specimens Test Technique, ASTM STP 1329, 1997, ASTM] applied this method to high loading rate fracture toughness data for A-515 steel and showed the applicability of this approach to dynamic fracture toughness data. This paper presents the data and the resulting reference temperature shift in the Master Curves from static to dynamic fracture data. Based on earlier PTS analyses performed in 1985, an appropriate T₀ shift value is selected for nuclear power plant applications. This shift in T₀ is compared with the temperature shift between K_Ic and K_Ia curves in ASME Boiler and Pressure Vessel Code.     

基于主曲线法的核电主管道材料断裂韧性研究

核电工程的防脆断设计和在役缺陷评价主要应用线弹性断裂力学,并基于材料断裂韧性进行评价。材料断裂韧性需通过试验测定,首先采用落锤试验和V型缺口冲击试验共同确定参考温度,或采用主曲线法确定参考温度,然后将参考温度和材料温度作为变量建立关系式描述材料的断裂韧性。主曲线法能通过较少的试样试验得到材料的断裂韧性,并具有较高的置信度,因此在工程中已得到越来越多的应用。文中采用ASTM E1921标准,应用主曲线法测量了某核电厂主管道材料的参考温度,确定了材料的断裂韧性,并与ASME第Ⅺ卷附录G中的断裂韧性进行比较。结果表明,采用主曲线法得到的材料断裂韧性更高,工程应用中减少了保守裕度,提高了经济性。     

Statistical re-evaluation of the ASME KIC and KIR fracture toughness reference curves

Historically the ASME reference curves have been treated as representing absolute deterministic lower bound curves of fracture toughness. In reality, this is not the case. They represent only deterministic lower bound curves to a specific set of data, which represent a certain probability range. A recently developed statistical lower bound estimation method called the ‘Master curve’, has been proposed as a candidate for a new lower bound reference curve concept. From a regulatory point of view, the master curve is somewhat problematic in that it does not claim to be an absolute deterministic lower bound, but corresponds to a specific theoretical failure probability that can be chosen freely based on application. In order to be able to substitute the old ASME reference curves with lower bound curves based on the master curve concept, the inherent statistical nature (and confidence level) of the ASME reference curves must be revealed. In order to estimate the true inherent level of safety, represented by the reference curves, the original database was re-evaluated with statistical methods and compared to an analysis based on the master curve concept. The analysis reveals that the 5% lower bound master curve has the same inherent degree of safety as originally intended for the K_IC-reference curve. Similarly, the 1% lower bound master curve corresponds to the K_IR-reference curve.     

Study on Fracture Toughness of Nuclear Power Main Pipeline Material Based on Master Curve Method

The design against brittle fracture and in-service defect evaluation of nuclear power engineering mainly use linear elastic fracture mechanics, and the evaluation is based on the fracture toughness of the material. The fracture toughness of the material needs to be determined by testing. First, the reference temperature is determined by the drop weight test and the V-notch impact test, or by master curve method, then the reference temperature and the material temperature are used as variables to establish a relationship to describe the fracture toughness of the material. The master curve method can obtain fracture toughness through the fewer sample tests and the higher confidence, so it has been used more and more in engineering. In this paper, the ASTM E1921 standard was used to measure the reference temperature of the main pipeline material of a nuclear power plant using the master curve method. The fracture toughness of the material was determined, and compared with the fracture toughness in Appendix G of ASME Volume Ⅺ. The results show that the fracture toughness of material obtained by the master curve method has higher value, and the conservative margin is reduced in engineering, so economy is improved.