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.     

Prediction of structural integrity of steam generator tubes under severe accident conditions

Available models for predicting failure of flawed and unflawed steam generator tubes under normal operating and design-basis accident conditions are reviewed. These rate-independent flow stress models are inadequate for predicting failure of steam generator tubes under severe accident conditions because the temperature of the tubes during such accidents can reach as high as 800°C where creep effects become important. Therefore, a creep rupture model for predicting failure was developed and validated by tests on unflawed and flawed specimens containing axial and circumferential flaws and loaded by constant as well as ramped temperature and pressure loadings. Finally, tests were conducted using pressure and temperature histories that are calculated to occur during postulated severe accidents. In all cases, the creep rupture model predicted the failure temperature and time more accurately than the flow stress models.     

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.     

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。     

Finite element analysis for creep failure of coolant pipe in light water reactor due to local heating under severe accident condition

During severe accident of a light water reactor (LWR), the piping of the reactor cooling system would be damaged when the piping is subjected to high internal pressure and very high temperature, resulted from high temperature gas generated in a reactor core and decay heat released from the deposit of fission products. It is considered that, under such a condition, short-term creep at high temperatures would cause the piping failure. For the evaluation of piping integrity under a severe accident, a method to predict such high temperature short-term creep deformation should be developed, using a creep constitutive equation considering tertiary creep. In this paper, the creep constitutive equation including tertiary creep was applied to nuclear-grade cold-drawn pipe of 316 stainless steel (SUS316), based on the isotropic damage mechanics proposed by Kachanov and Ravotnov. Tensile creep test data for the material of a SUS316 cold-drawn pipe were used to determine the coefficients of the creep constitutive equation. Using the constitutive equation taking account of creep damage, finite element analyses were performed for the local creep deformation of the coolant piping under two types of conditions; uniform temperature (isothermal condition) and temperature gradient of circumferential direction (non-isothermal condition). The analytical results show that the damage variable integrated into the creep constitutive equation can predict the pipe failure in the test performed by Japan Atomic Energy Research Institute, in which failure occurred from the outside of the pipe wall.     

Creep Properties of Base Metal and Welded Joint of Hastelloy XR Produced for High-Temperature Engineering Test Reactor in Simulated Primary-Coolant Helium

Creep tests of base metal, weld metal and welded joint of Hastelloy XR, which had the same chemical composition as Hastelloy XR produced for an intermediate heat exchanger of the High-Temperature Engineering Test Reactor, were conducted in simulated primary coolant helium. The weld metal and welded joint showed almost equal to or longer rupture time than the base metal of Hastelloy XR at 850 and 900°C, although they gave shorter rupture time at 950° C under low stress and at 1,000°C. The welded joint of Hastelloy XR ruptured at the base metal region at 850 and 900°C. On the other hand, it ruptured at the weld metal region at 950 and 1,000°C. The steady-state creep rate of weld metal of Hastelloy XR was lower than that of base metal at 850, 900 and 950°C. The creep rupture strengths of base metal, weld metal and welded joint of Hastelloy XR obtained in this study were confirmed to be much higher than the design allowable creep-rupture stress (S_R ) of the Design Allowable Limits below 950°C.     

A high-temperature creep model for Zr-2.5 wt% Nb pressure tubes

For a postulated loss-of-coolant accident in a CANDU reactor, in which the primary cooling circuit fails to remove the heat generated in the core, the temperature of the pressure tubes could rise very quickly. Since any deformation of the pressure tubes would control how the core heat is transferred to the surrounding moderator, which is a large heat sink, the accurate prediction of this transient deformation is essential. The majority of the pressure tubes in CANDU reactors are cold-worked Zr-2.5 wt% Nb and creep equations for this material have been developed from uniaxial creep tests. These creep equations were successful in predicting the creep strain in constant-stress uniaxial tests in which the temperature was ramped at rates ranging from 1° C/s to 50° C/s. They also successfully predicted the ballooning of internally pressurized sections of pressure tube that were heated at about 5° C/s.     

小型压水堆严重事故下一回路承压管道蠕变预测分析模型开发

一回路承压管道蠕变是压水堆核电厂严重事故重要现象之一。针对小型压水堆,本文基于SCDAP/RELAP5程序开发了严重事故分析模型,利用实验拟合方法得到了一回路主管道(SA321)、自然循环式蒸汽发生器传热管(00Cr25Ni35Al Ti)两种材料蠕变预测分析模型,改进了SCDAP/RELAP5程序蠕变预测分析功能模块,并通过假想事故序列验证了SA321、00Cr25Ni35Al Ti蠕变预测分析模型的合理性。为后续开展小型压水堆严重事故下一回路承压管道蠕变规律研究提供基础参考。     

Creep Properties with Short Period Excessive Loadings on a Nickel-Base Heat-Resistant Alloy Hastelloy XR

A series of constant load & temperature creep tests and constant temperature creep tests with short period excessive loadings was carried out on a nickel-base heat-resistant alloy Hastelloy XR, which was developed for applications in the High-Temperature Engineering Test Reactor (HTTR), at temperatures ranging from 900 to 1,000°C. The excessive loading levels were set at the design stress intensity values Sm for Hastelloy XR indicated in the HTTR high-temperature structural design code.

Five to six time excessive loadings did not cause significant changes of the minimum creep rate or the time to onset of tertiary creep. Excessive loadings repeated around ten times did not cause significant changes of the time to rupture or the rupture elongation. The results suggest that the design stress intensity values Sm for Hastelloy XR have been determined reasonably.     

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.     

Creep and creep rupture of a Zr-6%Sn-1%Mo alloy

Creep and creep rupture of a Zr-6%Sn-1%Mo alloy was investigated in a temperature interval 350 to 550°C using isothermal test and transmission electron microscopy techniques. The heat treatment of the alloy consisted in quenching to martensite and ageing at 550°C. The apparent activation energy of creep increases with temperature and applied stress from values lower to those significantly higher than the expected value of the activation enthalpy of the lattice diffusion to which the self-diffusion of all the three components of the solid solution (forming the matrix of the alloy) contributes. The parameter characterizing applied stress sensitivity of steady state creep rate increases with the stress from values only slightly higher than one to values close to thirty. The apparent activation energy of rupture life is close to that of creep and the stress sensitivity parameter of rupture life is close to that of steady state creep rate. The mechanisms likely to control creep rate under various external conditions are discussed. It is suggested that the mechanisms controlling the creep rate also control the rupture life.     

Material Properties of a Natural UF6 Transport Cylinder Vessel at High Temperature

Abstract

The IAEA Regulations for the Safe Transport of Radioactive Material are to be revised in 1996 and the fire test (800°C for 30 min) could become a requirement for the natural UF₆ transport cylinder. ASME SA 516 carbon steel is used as the structural material for this type of cylinder. It is very important to obtain high temperature data for SA 516 steel to be able to evaluate the integrity of the UF₆ transport cylinder vessel in the fire test. CRIEPI has therefore conducted material tests on SA 516 at high temperatures. The AC₁ and AC₃ transformation points of actual SA 516 steels have been measured. Tensile tests up to 900°C were conducted using USA, French and Japanese manufactured materials and the influence of phase transformation assessed. Preliminary creep tests show that assessment by creep strength can give a more conservative estimation than using the tensile strength. Creep deformation equations have been obtained using uniaxial creep tests and internal pressure creep tests. In addition, by the use of internal pressure creep rupture tests, the relation between the circumferential stress, the test temperature and the rupture time has been obtained.     

Detailed Evaluation of RCS Boundary Rupture during High-Pressure Severe Accident Sequences

A depressurization possibility of the reactor coolant system (RCS) before a reactor vessel rupture during a high-pressure severe accident sequence has been evaluated for the consideration of direct containment heating (DCH) and containment bypass. A total loss of feed water (TLOFW) and a station blackout (SBO) of the advanced power reactor 1400 (APR1400) has been evaluated from an initiating event to a creep rupture of the RCS boundary by using the SCDAP/RELAP5 computer code. In addition, intentional depressurization of the RCS using power-operated safety relief valves (POSRVs) has been evaluated. The SCDAPRELAP5 results have shown that the pressurizer surge line broke before the reactor vessel rupture failure, but a containment bypass did not occur because steam generator U tubes did not break. The intentional depressurization of the RCS using POSRV was effective for the DCH prevention at a reactor vessel rupture.     

An improvement in creep resistance of Zr-2.5%Nb tubes

Specimens from a Zr-2.5%Nb pressure tube were heat-treated in the range 650 to 1050°C and then creep-tested at 450°C. Water quenching produced anisotropic behaviour for soaking temperatures from 650 to 850°C, and isotropic behaviour above this range. A lower ‘intermediate cooling rate’ produced anisotropic behaviour for the whole soaking range. Creep resistance improved with increasing soaking temperature, particularly for transverse intermediate-cooled specimens. At soaking temperatures of 880°C and higher, a 100-fold reduction in creep rate was achieved with these specimens compared with as-cold-worked. An attempt is made to relate the creep data to crystallographic texture. In the second part of the program, the 880°C intermediate-cool heat treatment was chosen for further testing. It was confirmed that the material exhibits superior creep resistance compared with cold-worked Zr-2.5%Nb, at all stresses from 34 MN/m² to the ultimate tensile strength, in the temperature range from 300 to 450°C. As expected, the ultimate tensile strength was reduced by this heat treatment. At the upper end of the stress range, at 300°C, a discontinuity occurs in creep data. The instability causing the discontinuity and leading to failure at relatively low stress is attributed to a twinning mechanism. Large twins encompassing hundreds of grains are observed.     

堆芯熔化严重事故下反应堆压力容器下封头高温蠕变分析

核电厂在发生堆芯熔化严重事故时,采用堆内熔融物滞留（IVR）策略将熔融物包容在反应堆压力容器（RPV）内是一项重要缓解措施。在IVR策略期间,RPV下封头在熔融物的极高温度载荷和力学载荷的共同作用下很有可能因过度蠕变变形而失效。因此,有必要对熔融物滞留条件下RPV下封头进行蠕变变形分析,以保证RPV结构完整性。该文在假定IVR条件下,采用有限元方法对RPV下封头进行热-结构耦合分析,通过计算得到容器壁的温度场和应力场,以及下封头的塑性和蠕变变形,并结合塑性和蠕变断裂判据对下封头进行失效分析。结果表明,考虑蠕变影响后,结构的变形将大大增加;严重事故下采取熔融物滞留策略期间,RPV下封头的主要失效模式为蠕变失效而非塑性失效;内压对蠕变变形量和蠕变失效时间有较大影响。该文为严重事故下RPV下封头的蠕变和失效研究提供了分析方法。      

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。     

Analysis of the deformation and damage development of a dissimilar weld operating in the creep regime

The analysis of the deformation and damage behaviour of stress rupture tests with specimens out of the dissimilar metal weld seam 12% Cr-steel welded with a nickel base electrode for alloy 800 exhibits two competing processes:

• - Crack initiation occurs along the melting line due to high thermal stresses;

• - Creep deformation and damage concentrates in a heat affected zone of the ferritic 12% Cr steel due to long term stresses. The velocity of stress relaxation determines the resulting damage mechanism. At high temperatures with predominant creep deformation the cracks initiated in the melting line arrest and the creep deformation is concentrated in the heat affected zone (HAZ). At lower temperature the fracture area along the melting line increases. Long term tests at 535°C lead to lower stress rupture values compared to the scatterband of X 20 CrMoV 12 1 due to the reduced cross section after crack initiation in the melting line.

The analysis of stress rupture tests leads to the conclusion that grinding of melting line cracks is a reasonable measure because of sufficient stress relaxation.     

Chemical aspects of iodine-induced stress corrosion cracking failure of zircaloy-4 tubing above 50°C

The chemical environment associated with iodine-induced SCC failure of Zircaloy-4 tubing above 500°C has been characterized. At the critical iodine concentrations which result in SCC initiation and propagation, most of the iodine is present as condensed zirconium subiodides (I/Zr ? 0.4). Only a small part of the iodine remains in the gas phase as ZrI₄. The gaseous ZrI₄ is probably responsible for crack initiation and propagation. The critical ZrI₄ pressures for SCC failure have been estimated in zircaloy/iodine reaction experiments performed with unstressed zircaloy tube specimens. These pressures were confirmed in additional creep rupture tests conducted under controlled ZrI₄ partial pressure conditions. The estimated critical ZrI₄ pressure above which low-ductility SCC failure of the zircaloy tubing always occurs, independent of time-to-failure, varies between 0.005 bar at 550°C and 0.043 bar at 800°C. Below the critical values, however, a rather wide range of ZrI₄ pressures is associated with the onset of the SCC, especially at temperatures below 800°C. A comparison of the experimental results with available thermochemical data in the Zr-I system indicates that the main reaction involved during crack propagation is chemisorption of iodine-containing species on the fresh zircaloy surfaces created by metal straining at the crack tip.     

The creep rupture properties of beta-phase Zircaloy-2 cladding in the region 1000–1500°C

Accurate creep parameters are essential for the prediction and assessment of fuel cladding behaviour under reactor accident conditions. To resolve uncertainties in the limited data available for beta-phase Zircaloy cladding, an extensive programme of isothermal stress rupture tests has been undertaken for Zircaloy-2 tubes in a non-oxidizing environment at temperatures from 1000 to 1500°C. The secondary creep parameters are calculated from an analysis of the strain histories obtained from each test using a novel photo-electronic technique for the continuous monitoring of tube deformation. The resulting creep equation is compared with recent rupture life data for Zircaloy tubes. The analysis also identifies an anomalous primary stage which significantly retards the early stage of deformation for low stress tests at 1200°C and above and which is found to be extremely sensitive to prior annealing. The influence of this primary component on predictions of cladding behaviour is assessed.     

Characterization of PWR vessel steel tearing under severe accident condition temperatures

In the event of a severe core meltdown accident in a pressurised water reactor (PWR), core material can relocate into the lower head of the vessel resulting in significant thermal and pressure loads being imposed on the vessel. In the event of reactor pressure vessel (RPV) failure there is the possibility of core material being released towards the containment.On the basis of the loading conditions and the temperature distribution, the determination of the mode, timing, and size of lower head failure is of prime importance in the assessment of core melt accidents. This is because they define the initial conditions for ex-vessel events such as core/basemat interactions, fuel/coolant interactions, and direct containment heating. When lower head failure occurs (i) the understanding of the mechanism of lower head creep deformation; (ii) breach stability and its kinetic of propagation leading to the failure; (iii) and developing predictive modelling capabilities to better assess the consequences of ex-vessel processes, are of equal importance.The objective of this paper is to present an original characterization programme of vessel steel tearing properties by carrying out high temperature tearing tests on Compact Tension (CT) specimens.The influence of metallurgical composition on the kinetics of tearing is investigated as previous work on different RPV steels has shown a possible loss of ductility at high temperatures depending on the initial chemical composition of the vessel material. Small changes in the composition can lead to different types of rupture behaviour at high temperatures.The experimental programme has been conducted on various French RPV 16MND5 steels for temperatures ranging from 900 °C to 1100 °C. Comparisons between the tests performed on these various 16MND5 steels show that this approach is appropriate to characterize the difference in ductility observed at high temperatures.The aim of this experimental study is also to contribute to the definition of a tearing criterion by identifying, on the basis of CT results, the related material parameters at temperatures representative of the real severe accident conditions.This experimental campaign has been carried out in partnership with IRSN in the framework of a research programme whose purpose is to complete the mechanical properties database of 16MND5 steel and to model tearing failure in French RPV lower head vessels under severe conditions (Koundy et al., 2008).