管壁减薄对严重事故工况下传热管断裂的影响及缓解措施的有效性

严重事故的恶劣条件(反复的冷热交替及一、二回路之间的压差)可能导致蒸汽发生器(SG)传热管发生蠕变断裂。本文基于一级概率安全分析(PSA)的分析结果确定的典型事故序列,计算分析SG传热管壁减薄对严重事故工况下诱发蒸汽发生器传热管断裂(SGTR)的影响,给出严重事故缓解措施,例如一回路降压和给SG补水的有效性计算。     

Structural analysis of electrosleeved tubes under severe accident transients

A flow stress model was developed for predicting failure of electrosleeved PWR steam generator tubing under severe accident transients. The electrosleeve, which is nanocrystalline pure nickel, loses its strength at temperatures greater than 400°C during severe accidents because of grain growth. A grain growth model and the Hall–Petch relationship were used to calculate the loss of flow stress as a function of time and temperature during the accident. Available tensile test data, as well as high-temperature failure tests, on notched electrosleeved tube specimens were used to derive the basic parameters of the failure model. The model was used to predict the failure temperatures of electrosleeved tubes with throughwall and part-throughwall axial cracks in the parent tube during a postulated severe accident transient.     

Estimation of temperature-induced reactor coolant system and steam generator tube creep rupture probability under high-pressure severe accident conditions

A severe accident has inherently significant uncertainties due to the complex phenomena and wide range of conditions. Because of its high temperature and pressure, performing experimental validation and practical application are extremely difficult. With these difficulties, there has been few experimental researches performed and there is no plant-specific experimental data. Instead, computer codes have been developed to simulate the accident and have been used conservative assumptions and margins. This study is an effort to reduce the uncertainty in the probabilistic safety assessment and produce a realistic and physical-based failure probability. The methodology was developed and applied to the OPR1000. The creep rupture failure probabilities of reactor coolant system (RCS) components were evaluated under a station blackout severe accident with all powers lost and no recovery of steam generator auxiliary feed-water. The MELCOR 1.8.6 code was used to obtain the plant-specific pressure and temperature history of each part of the RCS and the creep rupture failure times were calculated by the rate-dependent creep rupture model with the plant-specific data.     

Steam generator tube rupture (SGTR) scenarios

The steam generator tube rupture (SGTR) scenarios project was carried out in the EU 5th framework programme in the field of nuclear safety during years 2000–2002. The first objective of the project was to generate a comprehensive database on fission product retention in a steam generator. The second objective was to verify and develop predictive models to support accident management interventions in steam generator tube rupture sequences, which either directly lead to severe accident conditions or are induced by other sequences leading to severe accidents. The models developed for fission product retention were to be included in severe accident codes. In addition, it was shown that existing models for turbulent deposition, which is the dominating deposition mechanism in dry conditions and at high flow rates, contain large uncertainties. The results of the project are applicable to various pressurised water reactors, including vertical steam generators (western PWR) and horizontal steam generators (VVER).     

全厂断电引发的严重事故下蒸汽发生器传热管蠕变失效风险研究

全厂断电引发的严重事故若处置不当,可能发展为长期、高压的严重事故进程,此时堆芯冷却系统中的自然循环在导出部分堆芯余热的同时,也增加了蒸汽发生器（SG）传热管、稳压器波动管以及热管段出现蠕变失效的风险。本文基于两环路设计的秦山二期核电厂设计特点,结合蠕变失效风险模型,对全厂断电引发的严重事故后未能执行“严重事故管理导则中向蒸汽发生器注水（SAG-1）”时SG传热管的蠕变失效风险进行了研究,从而为全厂断电引发的严重事故的负面影响提供量化结果,为技术支持中心（TSC）最终决策提供参考依据。分析结果表明,全厂断电引发的严重事故后16 361 s可能出现蠕变失效;自事故后16 610 s,SG传热管出现蠕变失效的可能性均远低于稳压器波动管与热管段,秦山二期核电厂全厂断电引发的严重事故下因SG传热管蠕变失效而导致安全壳旁通的风险很小。     

Analysis of Overheating Rupture in Heat-Transfer Tubes Causing Corrosive High-Temperature Reaction

In the steam generator of a liquid metal fast breeder reactor, a defect penetrating through heat-transfer tube will cause high-pressure water/steam to spout into the low-pressure sodium filling the space outside the tube, to initiate sodium-water reactions. If the leak exceeds an intermediate level (~2kg/s), the reaction jet may rupture adjoining tubes with overheating in the event of insufficient cooling available inside the tubes. Such phenomenon of overheating tube rupture presents a serious problem to the economy and safety of steam generator. With a view to clarifying the failure behavior of steam generator heat-transfer tubes under such condition a model of the phenomenon is derived through a series of tests on sodium-water reactions making use of a test loop representing the scale model of an actual fast breeder steam generator. Comparison of actual test data with analysis based on the model has yielded the following information: The failure behavior of gas-pressurized tubes fall into two categories: (a) by creep failure—occurring upon increase of cumulative damage with tube wall wastage caused by the reaction jet and (b) by ductile failure accompanied by creep—upon tube heating with the reaction jet to the extent of lowering tube wall strength below the hoop stress exerted by tube pressure. Analysis of the two categories of failure results in estimation of the percentage difference between analyzed and measured times to failure of 35–50% in the case of creep failure and of 20–50% in the case of ductile failure accompanied by creep. In practical application to steam generators in order to provide a safety margin a time factor—i.e., the safety factor indicating multiple of actual time to failure—of 3 is adopted against 1.5–2 indicated from test to be the actually applicable value.     

严重事故下蠕变诱发RCS破裂的概率评估

严重事故下一回路管道可能会发生蠕变失效,若出现蠕变诱发的蒸汽发生器传热管破裂（SGTR）,则会导致安全壳旁路失效;若出现蠕变诱发热段或波动管的失效,则产生的破口将会使一回路迅速卸压。因此,评估严重事故下蠕变诱发反应堆冷却剂系统（RCS）破裂的可能性是开展严重事故分析、特别是二级概率安全分析（PSA）的重要基础。本工作基于蠕变失效模型,考虑传热管的缺陷,建立了评价蠕变诱发RCS破裂的确定论模型。在此基础上,运用拉丁超立方体抽样方法,考虑重要参数的不确定性,开发了严重事故下蠕变诱发RCS破裂的概率评估程序。随后对典型的事故序列进行了蠕变诱发RCS破裂的概率评估。结果表明,对于高压事故序列,存在一定的蠕变诱发SGTR概率,也存在较高的蠕变诱发热段或波动管失效概率。     

Potential steam generator tube rupture in the presence of severe accident thermal challenge and tube flaws due to foreign object wear

This study develops a methodology to assess the probability for the degraded PWR steam generator to rupture first in the reactor coolant pressure boundary, under severe accident conditions with counter-current natural circulating high temperature gas in the hot leg and SG tubes. The considered SG tube flaws are caused by foreign object wear, which in recent years has emerged as a major inservice degradation mechanism for the new generation tubing materials. The first step develops the statistical distributions for the flaw frequency, size, and the flaw location with respect to the tube length and the tube's tubesheet position, based on data of hundreds of flaws reported in numerous SG inservice inspection reports. The next step performs thermal-hydraulic analysis using the MELCOR code and recent CFD findings to predict the thermal challenge to the degraded tubes and the tube-to-tube difference in thermal response at the SG entrance. The final step applies the creep rupture models in the Monte Carlo random walk to test the potential for the degraded SG to rupture before the surge line. The mean and range of the SG tube rupture probability can be applied to estimate large early release frequency in probabilistic safety assessment.     

Effect of Microstructure on Failure Behavior of Light Water Reactor Coolant Piping under Severe Accident Conditions

In a severe accident of light water reactors, the reactor coolant system (RCS) piping might be subjected to thermal loads caused by the decay heat of the deposited fission products and the heat transfer from the hot gases, with an internal pressure in some accident sequences. Tests on the RCS piping failure were performed along with high temperature tensile and creep rupture tests including metallography to investigate the failure behavior. The prediction of the 0.2% proof stress by Arrhenius equation is in good agreement with the measured stress above 800°C for served RCS piping materials. The modified Norton's Law for the short term creep rupture model agrees with the experimental values between 800 and 1,150°C for type 316 stainless steel. The microstructural change was discussed with the effect of the very rapid formation and resolution of the precipitation on the strength at high temperature. The result of the piping failure tests which simulated the severe accident conditions, i.e., in short-term at high-temperature, could support the plastic limit load prediction of the flow stress model using the 0.2% proof stress.     

Simple Method for Evaluating Mesh Size Effects in Control Rod Worth Analysis

It has been pointed out that the reactor coolant system piping could fail prior to the meltthrough of the reactor pressure vessel in a high pressure sequence of pressurized water reactor severe accidents. In order to apply to the evaluation of the piping failure which influences the subsequent accident progression, models for the strength of piping materials at high temperatures were examined. It was found that 0.2% proof stress and ultimate tensile strength above 1,073 K obtained from tensile tests was reproduced by a quadratic equation of the reciprocal absolute temperature. Short-term creep rupture time and minimum creep rate at high temperatures were well correlated by the modified Norton's Law as a function of stress and temperature, which implicitly expressed the effect of the precipitation and the resolution of precipitates on the creep strength. The modified Norton's Law gave better results than the conventional Larson-Miller method. Relating applied stress vs. minimum creep rate and tensile properties vs. applied strain rate obtained from the creep and tensile tests, a temperature range where the dynamic recrystallization significantly occurred was evaluated.     

Analysis of severe core damage accident progression for the heavy water reactor

In this study,the severe accident progression analysis of generic Canadian deuterium uranium reactor 6 was preliminarily provided using an integrated severe accident analysis code.The selected accident sequences were multiple steam generator tube rupture and large break loss-of-coolant accidents because these led to severe core damage with an assumed unavailability for several critical safety systems.The progressions of severe accident included a set of failed safety systems normally operated at full power,and initiative events led to primary heat transport system inventory blow-down or boil off.The core heat-up and melting,steam generator response,fuel channel and calandria vessel failure were analyzed.The results showed that the progression of a severe core damage accident induced by steam generator tube rupture or large break loss-of-coolant accidents in a CANDU reactor was slow due to heat sinks in the calandria vessel and vault.     

Creep rupture assessment for Level-2 PSA of a 2-loop PWR: accounting for phenomenological uncertainties

The Level-2 probabilistic safety assessment (PSA) of pressurized water reactors studies the possibility of creep rupture for major reactor coolant system components during the course of high pressure severe accident sequences.The present paper covers this technical issue and tries to quantify its associated phenomenological uncertainties for the development of Level-2 PSA.A framework is proposed for the formal quantification of uncertainties in the Level-2 PSA model of a PWR type nuclear power plant using an integrated deterministic and PSA approach.This is demonstrated for estimation of creep rupture failure probability in station blackout severe accident of a 2-loop PWR,which is the representative case for high pressure sequences.MELCOR 1.8.6 code is employed here as the deterministic tool for the assessment of physical phenomena in the course of accident.In addition,a MATLAB code is developed for quantification of the probabilistic part by treating the uncertainties through separation of aleatory and epistemic sources of uncertainty.The probability for steam generator tube creep rupture is estimated at 0.17.     

Analysis of hot leg natural circulation under station blackout severe accident

Under severe accidents, natural circulation flows are important to influence the accident progression and result in a pressurized water reactor （PWR）. In a station blackout accident with no recovery of steam generator （SG） auxiliary feedwater （TMLB＇ severe accident scenario）, the hot leg countercurrent natural circulation flow is analyzed by using a severe-accident code, to better understand its potential impacts on the creep-rupture timing among the surge line, the hot leg, and SG tubes. The results show that the natural circulation may delay the failure time of the hot leg. The recirculation ratio and the hot mixing factor are also calculated and discussed.     

AP1000核电厂大量放射性释放源项分析

在AP1000核电厂的某些严重事故情景中,安全壳可能发生失效或旁通,导致大量放射性物质释放到环境中,造成严重的放射性污染。针对大量放射性释放频率贡献最大的3种释放类别（安全壳旁通、安全壳早期失效和安全壳隔离失效）,分别选取典型的严重事故序列（蒸汽发生器传热管破裂、自动卸压系统阀门误开启和压力容器破裂）,使用MAAP程序计算分析了释放到环境中的裂变产物源项。该分析结果为量化AP1000核电厂的放射性释放后果和厂外剂量分析提供了必要的输入。     

Structural integrity evaluation method for overheating rupture of FBR steam generator tube

This paper describes a structural integrity evaluation method for a SG tube of FBR in case of sodium–water reaction and creep rupture tests to obtain the strength of the tube material. In the SG of FBR, if intermediate size of water/steam leak (1–2 kg s⁻¹) would occur from a tube, it could cause overheating rupture of the multiple tubes surrounding the initially failed tube due to generated sodium–water reaction heat. In the ultra-high temperature condition, the creep strength of the material is one of the dominant factors for failure behavior. Accordingly, we tried to apply the creep failure criterion for the overheating rupture of the SG tube. The creep rupture tests have been performed at ultra-high temperature conditions ranging from 1223.2 to 1323.2 K. The test material is ‘Mod .9Cr–1Mo steel’ which is one of the candidate materials for the tubes of the future SG of FBR. The test results have shown that tube rupture depends on the creep strength of the material; hence, instantaneous rupture does not occur even if the stress exceeds the design value of ultimate tensile strength. The test data have been suitably expressed using the Larson–Miller Parameter, and a structural integrity evaluation method based on the sum of the use-fraction associated with the creep damage has been proposed. Based on this method, the structural integrity of the tube in the sodium–water reaction flame has been evaluated. The results show that it is important to detect the initial leak of the tube within a short period and to reduce the steam pressure more rapidly by SG blowdown.     

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.     

IRIS反应堆严重事故下堆内自然循环及下封头失效分析

使用REALP5/SCDAP分析了IRIS堆汽轮机停机和部分失流事故导致的严重事故进程及缓解措施。分析结果表明IRIS堆内水装量大,使得堆芯较长时间处于淹没状态,事故发生后近7个小时堆芯开始裸露,10小时后堆芯开始损坏。对于不卸压不安注的情况,压力容器会完全干涸,堆芯和蒸汽发生器之间形成蒸汽自然循环流动,堆芯温度缓慢升高,低熔点的控制棒金属首先熔化落入下腔室并加热下封头,使得下封头底部区域发生蠕变断裂失效。在不卸压的情况下一个上充泵的安注流量就能够缓解事故。     

AP1000核电厂蒸汽发生器传热管破裂事故的分析研究

使用RELAP程序对AP1000核电厂蒸汽发生器传热管破裂（SGTR）事故进行了分析研究,证明了AP1000核电站在SGTR事故下,不需要操纵员的干预就能依靠非能动安全系统在破损蒸汽发生器满溢之前终止破口流量。重点研究了不同的事故分析假设条件,如厂外电是否可用以及破损蒸汽发生器的释放阀是否打开后卡在开启位置对事故后果的影响。结果表明,即使在对破损蒸汽发生器满溢最不利的假设条件下,AP1000核电站也能避免破损蒸汽发生器满溢,且存在一定的裕量。     

AP1000核电厂蒸汽发生器传热管破裂事故的分析研究

使用RELAP程序对AP1000核电厂蒸汽发生器传热管破裂（SGTR）事故进行了分析研究,证明了AP1000核电站在SGTR事故下,不需要操纵员的干预就能依靠非能动安全系统在破损蒸汽发生器满溢之前终止破口流量。重点研究了不同的事故分析假设条件,如厂外电是否可用以及破损蒸汽发生器的释放阀是否打开后卡在开启位置对事故后果的影响。结果表明,即使在对破损蒸汽发生器满溢最不利的假设条件下,AP1000核电站也能避免破损蒸汽发生器满溢,且存在一定的裕量。     

Bottom nozzle failure mechanism of water reactor pressure vessel under severe accident conditions

Most of past studies devoted to the creep rupture of a nuclear reactor pressure vessel (RPV) lower head under severe accident conditions, have focused on global deformation and rupture modes. Limited efforts were made on local failure modes associated with penetration nozzles as a part of TMI-2 vessel investigation project (TMI-2 VIP) in 1990s. However, it was based on an excessively simplified shear deformation model. In the present study, the mode of nozzle failure has been investigated using data and nozzle materials from Sandia National Laboratory's lower head failure experiment (SNL-LHF). Crack-like separations were revealed at the nozzle weld metal to RPV interfaces indicating the importance of normal stress component rather than the shear stress in the creep rupture. Creep rupture tests were conducted for nozzle and weld metal materials, respectively, at various temperature and stress levels. Stress distribution in the nozzle region is calculated using elastic–viscoplastic finite element analysis (FEA) using the measured properties. Calculation results are compared with earlier results based on the pure shear model of TMI-2 VIP. It is concluded from both LHF-4 nozzle examination and FEA that normal stress at the nozzle/lower head interface is the dominant driving force for the local failure. From the FEA for the nozzle weld attached in RPV, it is shown that nozzle welds failure occur by displacement controlled fracture of nozzle hole not by load controlled fracture of internal pressure. Considering these characteristics of nozzle weld failure, new concept of nozzle failure time prediction is proposed.