A review of work related to reheat cracking in nuclear reactor pressure vessel steels

This review was completed in 1976 and describes the then-current knowledge on the problem of reheat cracking in weldments of nuclear pressure vessel steels. The incidence of underclad cracks and structural weld cracks in heat affected zones (HAZ) is described and current theories on the cracking mechanism and controlling factors discussed. Problems concerned with detecting the cracks by destructive and non-destructive methods are outlined. The paper then deals in detail with testing techniques for reheat crack susceptibility and with methods of controlling the problem of underclad and structural weld cracking. Finally, the engineering significance of the cracks in relation to hydrotesting and service is discussed. An appendix gives recommendations for additional work on the subject.     

Application of advanced master curve approaches on WWER-440 reactor pressure vessel steels

The master curve (MC) approach used to measure the transition temperature, T₀, was standarised in the ASTM Standard Test Method E 1921 in 1997. The basic MC approach for analysis of fracture test results is intended for macroscopically homogeneous steels with a body centred cubic (ferritic) structure only. In reality, due to the manufacturing process, the steels in question are seldom fully macroscopically homogeneous.     

Fatigue crack growth behavior of reactor pressure vessel steels in air and high-temperature water environments

Fatigue tests under constant amplitude load were conducted on CT specimens of A533B3 steels with four levels of sulfur content at different temperatures in air and high-temperature water environments. A modified capacitance-type COD gauge was shown to be suitable for fatigue crack length measurement at high temperatures in air. The observation that the Young's moduli measured at a strain rate of 4 × 10⁻³ s⁻¹ for the A533B3 steels at 150 °C and 300 °C did not decrease with an increase in temperature seemed to be related to the presence of dynamic strain aging. The fatigue crack growth rates at 150 °C and 300 °C in air were about two and half times slower than those tested at 400 °C, because dynamic strain aging prevailed at 150 °C and 300 °C. Fractographic examination results suggested that inclusions embedded in secondary cracks enhanced the fatigue crack initiation rather than the fatigue crack growth. The fatigue crack growth rates taken in the oxygen-saturated water environment were one order of magnitude faster than those obtained in air.     

Stress corrosion cracking of power boiler drums

This paper deals with the study, analysis and technical diagnosis fundamentals concerning damage induced by stress corrosion cracking. The main repair and safe operation methods for power boiler drums are described; this work being based on plant experience.     

Recent Japanese research activities on probabilistic fracture mechanics for pressure vessel and piping of nuclear power plant

This paper describes a review of recent Japanese activities on probabilistic fracture mechanics (PFM) analyses. Japan Atomic Energy Agency (JAEA: previously JAERI) had sponsored research committees on PFM organized by Japan Society of Mechanical Engineers (JSME) and Japan Welding Engineering Society (JWES) for more than a decade. This work still continues with the same members in JWES. The purpose of the continuous activity is to provide probabilistic approaches in several fields of integrity problems of nuclear power plant. This paper shows some of the newest results of the JWES research committee. First topic is evaluation of the new JSME code case with rules of Fitness-For-Service from the view of PFM, including reactor pressure vessel subject to pressurized thermal shock loading, piping with a crack of the allowable size and effect of sizing accuracy for piping integrity. The next one is development of new PFM techniques including reliability assessment of piping with domestic (Japanese) SCC data and maintenance optimization of LWRs based on risk and economic models. The last topic is the international round robin program just starting from 2008.     

Mechanical methods of improving resistance to stress corrosion cracking in BWR piping systems

Pipelocks and the mechanical stress improvement process (MSIP) have been applied in BWR plants. Pipelocks restore the integrity of the weldments with identified cracks. MSIP removes residual tensile stresses from weldments, thus preventing initiation of cracks or retarding growth of pre-existing flaws in piping systems. The first 12-in pipelock installed at Commonwealth Edison's Quad Cities plant was inspected after operating for 18 months. Pipelocks for 10-in, 12-in and 28-in reactor safe-ends were fabricated for Carolina Power & Light's Brunswick plant. MSIP was applied at Commonwealth Edison's Dresden and LaSalle BWR plants. Extensive qualification has been completed for MSIP under US Nuclear Regulatory Commission and Electric Power Research Institute sponsorship. Sectioning of the pipe wall by Argonne National Laboratories provided stress distribution before and after MSIP for the 12-in and 28-in pipes. Measured ‘as welded’ tensile stresses before MSIP were within the range 30–50 000 psi. Compressive stresses after MSIP at the inside surface of the weldment reached more than 30 000 psi in both hoop and axial directions. The axial compressive stresses extended to the middle plane of the wall. Hoop stresses remained compressive through the wall. The stresses were uniformly distributed around the circumference of the pipe. J.A. Jones Applied Research Company completed an evaluation of MSIP applied to a precracked weldment between 28-in pipe and elbow. The pipe was squeezed to about 1·7% in the presence of cracks 25%, 50% and 90% through-wall. High compressive stresses were measured after MSIP. The cracks did not extend and could be identified after completion of the process by the usual UT technique. The use of mechanical methods becomes especially adequate for reactor safe-ends including bi- or tri-metallic joints. The use of overlay technique or induction heat stress improvement is more difficult due to high thermally induced strains at the strong discontinuity interface between materials of different thermal expansion. Basic concepts and practical application of mechanical methods to inhibit stress corrosion attack are described.     

Hydrogen embrittlement of CrMo and CrMoV pressure vessel steels

The variation of $\text{K}{}_{\mathrm{i}}$ with time to fracture and the threshold values of $\text{K}{}_{\mathrm{ISH}}$ in an hydrogen environment were measured for four CrMo and CrMoV pressure vessel steels in various conditions of heat treatment. The CrMoV grades displayed higher $\text{K}{}_{\mathrm{ISH}}$ thresholds, suggesting that they are less susceptible to hydrogen embrittlement than CrMo grades of comparable strength. The findings were analysed with regard to the concurrent presence of various alloying elements and their effects on the microsegregation of detrimental impurity elements at the austenitic grain boundaries, where these harmful elements can, in conjunction with hydrogen, cause intercrystalline embrittlement. Studies were also made of the kinetics of stable crack growth in these materials.     

Failure probability and optimum microstructure of pressure vessel steels

This paper presents a method for assessing the probability of brittle fracture in steels used for nuclear power generating equipment. The method is based on a statistical analysis of microstructural parameters and on physical models of the initiation and propagation of cleavage microcracks. The least probability of brittle failure ascertained by this method has been utilized for determining the optimum ferrite grain size. This method represents a foundation for systematic control of the brittle fracture characteristics of engineering steels employed for the components of nuclear power stations.     

Stress intensity factor solutions for arbitrarily shaped surface flaws in reactor pressure vessel nozzle corners

Extensive experimental results—based on the frozen stress photoelasticity technique for extracting stress intensities—for nozzle corner cracks in ITV and BWR geometries were reported by Smith et al.¹ Based on the above experimental studies, it was conjectured that if the crack shape inserted into a finite element model is not a real one, or if the inner fillet (for shallow flaws) or the outer boundary shape (for moderate to deep flaws) is improperly approximated, the obtained numerical results for stress intensity factors may differ significantly from the physical behaviour at the nozzle-vessel junction. On the other hand, almost all the numerical analyses published to date, based on finite elements, boundary integral equations or alternating techniques, considered only quarter-circular nozzle corner cracks.This paper presents stress intensity factor solutions for naturally shaped nozzle corner cracks in pressurised ITV and BWR vessels. Several actual crack geometries observed in the experimental work of Smith et al., cited above, are studied using the three-dimensional hybrid crack-element approach of Atluri et al.⁸ and Atluri and Kathiresan⁹ wherein the stress intensity factors and their variation along an arbitrarily shaped 3-D crack front are directly computed. In order to be able to compare the present results with the photoelastic experimental results (wherein the Poisson's ratio of the material is 0·5), some of the present numerical results are obtained for $\text{ν ? 0·5}$.In addition, some new solutions for stress intensity factors for pressurised thin (outer to inner radii ratios of $\text{～1·1}$) cylindrical vessels with belt-line flaws of semi-elliptical shapes of various aspect ratios and depth ratios are presented. Cases of surface flaws in the meridional direction—as well as in the circumferential direction—of the vessel are treated.     

Fracture toughness and other mechanical metallurgy characteristics of pressure vessel steels

The objective of this work is to investigate the effect of specimen size and evaluation method on estimates of fracture toughness for three commercial pressure vessel steels in different heat treatment conditions. It is found that the fracture toughness values K_C recorded on large, 150 mm thick, specimens were consistently close to the lower bound envelope of the K_δC, K_JC and K_E values established on 20 mm thick specimens. Comparison of the various elasto-plastic fracture criteria ascertained on a single specimen shows that the variations in these values are relatively slight which implies that the evaluation techniques used in this study can be considered as giving results that are, for practical purposes, the same.     

Combination of radiation and hydrogen damage of reactor pressure vessel materials

The hydrogen embrittlement of low-alloyed base steel, austenitic cladding and heat affected zone (HAZ) of a reactor pressure vessel was measured for both unirradiated and irradiated materials. The fracture toughness decreased with both hydrogen charging and neutron irradiation; the shift of the fracture toughness-temperature transient curve is influenced by both damage processes. The plastic zone in hydrogen-charged material becomes smaller. The total elongation of both CrMoV and CrNiMoV HAZ decreases with increasing hydrogen content. This influence is pronounced in the HAZ after a weld process without subsequent annealing, a total loss of plasticity being observed in this case. The properties of the austenitic layer are not influenced at comparable hydrogen contents.     

A new approach to estimate irradiation embrittlement of pressure vessel steels

The idea of an engineering version of the local approach to brittle fracture as well as the possibility for it to be employmed to estimate irradiation embrittlement of pressure vessel (PV) steels is considered. Unlike the conventional temperature-shift-based methodology, the approach presented utilises the concept of stability of the ductile state of the metal. A new characteristic “parameter of mechanical stability” P_ms is proposed. This characteristic enables quantification of the level of stability of the ductile state of an irradiated PV steel in a specimen or in a reactor vessel with a crack at the specified level of loading. Within the framework of the proposed concept, a value for end-of-life fluence for a reactor PV is predicted by the condition of exhaustion of stability of a ductile state of a steel ahead of the crack (P_ms=1).     

Structure analysis of a reactor pressure vessel by two- and three-dimensional models

This paper investigates the reactor pressure vessel of a 1300 MW pressurised water reactor. In order to determine the stresses and deformations of the vessel, two- and three-dimensional finite element models are used which represent the real structure with different degrees of accuracy. The results achieved by these different models are compared for the case of the transient called ‘Start up of the nuclear power plant’. It was found that axisymmetric models, which consider non-axisymmetric components by correction factors, together with special attention to holes and other stress concentrations, allow a sufficient computation of stresses and deformations in the vessel, with the exception of the coolant nozzle region. In this latter case a fully three-dimensional analysis may be necessary.     

Stress corrosion cracking behavior of ZK60 magnesium alloy under different conditions

The stress corrosion cracking (SCC) behavior of ZK60 magnesium alloy was investigated under different conditions, i.e. thin electrolyte layer (TEL) and solution, by slow strain rate tensile tests, electrochemical techniques, Auger electron spectroscopy, scanning electron microscopy coupled with electron backscattered diffraction, and time of flight secondary ion mass spectrometry. Results indicated that the ZK60 magnesium alloy in solution exhibits a higher SCC susceptibility with a combined SCC mechanism of weaker anodic dissolution (AD) and stronger hydrogen embrittlement (HE) compared to under TEL. Moreover, the HE mechanism under various conditions was discussed.     

压力容器补强圈与筒体间接触特性的应力分析研究

本文采用有限元分析方法对接管补强圈和筒体的不同接触状态进行了应力分析,揭示了不同接触间隙下的接触应力分布的情况。计算结果表明,随着压力的加载,接触位置由内外焊缝中间逐步扩散到整个接触面,接触间隙与接触应力值呈正相关。     

Deterministic assessment of reactor pressure vessel integrity under pressurised thermal shock

Numerical investigations were carried out to assess the integrity of reactor pressure vessels under pressurised thermal shock (PTS). The 4-loop reactor pressure vessel with cladding was subjected to thermo-mechanical loading owing to loss of coolant accident. The loss of coolant accident corresponding to small break as well as hot leg breaks were considered separately, which led to axisymmetric and asymmetric thermal loading conditions respectively. Three different crack configurations, 360° circumferential part through, circumferential semi-elliptical surface and circumferential semi-elliptical under-clad cracks, were postulated in the reactor pressure vessel. Finite element method was used as a tool for transient thermo-elastic analysis. The various fracture parameters such as crack mouth opening displacement (CMOD), stress intensity factor (SIF), nil ductility transition temperature (RT_NDT) etc. were computed for each crack configuration subjected to various type of loading conditions. Finally for each crack a fracture assessment was performed concerning crack initiation based on the fracture toughness curve. The required material RT_NDT was evaluated to avoid crack initiation.     

Simplified fitness-for-service assessment of pressure vessels and piping systems containing thermal hot spots and corrosion damage

Fitness-for-service (FFS) assessment is a quantitative engineering evaluation of operational components. In the context of pressure vessels and piping systems FFS assessment is performed periodically to ensure the operational safety and structural integrity. In this paper, a simplified method is developed for Level 2 FFS assessment (as described in API 579) of pressure vessels and piping systems containing thermal hot spots or corrosion damage. The method is based upon variational principles in plasticity, the m_α-tangent method (an extension of the m_α method), the concept of decay length and reference volume. The use of the m_α-tangent method extends the range of applicability to components and structures experiencing significant stress gradients in and around the damaged spot. The method is shown to provide a reasonably accurate estimate of the remaining strength of ageing pressure components. The method is demonstrated through an example, and the results are compared with Level 3 inelastic finite element analyses.     

Stress intensity factors for underclad and through clad defects in a reactor pressure vessel submitted to a pressurised thermal shock

CEA has launched important work on the development of a Stress Intensity Factors compendium for cracks in a Reactor Pressure Vessel (RPV) taking into account the cladding.The work is performed by Finite Element analysis with a parametric mesh for two types of defects (under clad defect and through clad defect) and a wide range of geometrical and material parameters.In addition, an analytical stress solution for Pressurised Thermal Shock (PTS) on the RPV is proposed to allow a complete analytical estimation of the stress intensity factor K_I for the PTS problem.The results are validated by comparison with a complete 3D finite element calculation performed on a complex and realistic case study.     

Adequacy of estimates and variability of fracture-related properties for reactor pressure vessel materials

It is sometimes necessary to estimate K_Ic, NDTT, or RT_NDT based on other types of test data because actual toughness results are not available. Correlations used to estimate these toughness parameters are often based on minimal Charpy V-notch data. This paper evaluates the adequacy of correlating Charpy V-notch data with fracture toughness parameters for two pressure vessel materials and weldments. Statistically based correlations or ‘engineering judgments’ were used to establish a conservative prediction criterion.