Age-related degradation of boiling water reactor vessel internals

Researchers at the Idaho National Engineering Laboratory performed an assessment of the aging of the reactor internals in boiling water reactors (BWRs), and identified the unresolved technical issues related to the degradation of these components. The overall life-limiting mechanism is intergranular stress corrosion cracking (IGSCC). Irradiation-assisted stress corrosion cracking, fatigue, and thermal aging embrittlement are other potential degradation mechanisms. Several failures in BWR internals have been caused by a combination of factors such as environment, high residual or preload stresses, and flow-induced vibration. The ASME Code Section XI in-service inspection requirements are insufficient for detecting aging-related degradation at many locations in reactor internals. Many of the potential locations for IGSCC or fatigue are not accessible for inspection.     

Probability of crack-induced failure in BWR recirculation piping

The Lawrence Livermore National Laboratory (LLNL) has estimated the probability of double-ended guillotine break (DEGB) in the reactor coolant piping of Mark I boiling water reactor (BWR) plants. Two causes of pipe break are considered: crack growth at welded joints and the seismically-induced failure of component supports. For the former a probabilistic fracture mechanics model is used, for the latter a probabilistic support reliability model. This paper describes a probabilistic model developed to account for effects of intergranular stress corrosion cracking (IGSCC). The IGSCC model, based on experimental and field data compiled from several sources, correlates times to crack initiation and crack growth rates for Types 304 and 316NG stainless steel against material-specific ‘damage parameters’ which consilidate the separate effects of coolant environment (temperature, dissolved oxygen content, level of impurities), stress (including residual stress), and degree of sensitization. Application of this model to actual BWR recirculation piping shows that IGSCC clearly dominates the probability of failure in 304SS piping, mainly due to cracks that initiate within a few years after plant operation has begun. Replacing Type 304 piping with 316NG reduces failure probabilities by several orders of magnitude.     

Fracture mechanics investigation regarding the effects of cracks on the structural integrity of a BWR-core shroud

As a consequence of core shroud intergranular stress corrosion cracking (IGSCC) detected in the course of inservice inspections, a fracture mechanics analysis was carried out to evaluate the effects of postulated cracks on the structural integrity. In this study, critical crack sizes and crack growth were calculated. Due to the comparatively low stress acting on the core shroud during normal operation, the residual stresses in the welds make up the major proportion of the tensile stresses responsible for IGSCC. In order to consider residual stresses of the lower core support ring welds, a finite element analysis was performed at MPA Stuttgart using the FE-code ANSYS. The crack growth computed on the basis of USNRC crack growth rates da/dt demonstrated that crack growth in depth direction increases quickly at first, then retards and finally comes almost to a standstill. The cause of this ‘quasi-standstill’ is the residual stress pattern across the wall, being characterized by tensile stresses in the outer areas of the wall and compressive stresses in the middle of the wall. Crack growth in circumferential direction remains more or less constant after a slow initial phase. As the calculation of stress intensity factors K_I of surface flaws under normal conditions demonstrated, a ‘lower bound’ fracture toughness value is only exceeded in the case of very long and deep surface flaws. It can be inferred from crack growth calculations that under the assumption of intergranular stress corrosion cracking, the occurrence of such deep and at the same time long flaws is unlikely, regardless of the initial crack length. Irrespective of the above, the calculated critical throughwall crack lengths, which were determined using a ‘lower bound’ fracture toughness value, demonstrated that even long throughwall cracks will not affect the component’s integrity under full load. Moreover, it can be concluded from the studies of crack growth that—assuming intergranular stress corrosion cracking—a sufficiently long period will elapse before a crack which has just been initiated reaches a relevant size. Therefore, it can be stated that these cracks will likely be detected during periodic inservice inspections.     

Experience and assessment of stress corrosion cracking in L-grade stainless steel BWR internals

The stress corrosion cracking (SCC) rate of reactor internals of boiling water reactors (BWR) continues to increase with on-line operating years. The recent occurrences of cracking in the weld heat affected zones of high carbon stainless steel core shrouds correlate with the years of operation and the water chemistry history. Recently, cracking has also been found in shrouds that were constructed of low carbon or stabilized stainless steels. While these steels are more resistant to intergranular stress corrosion cracking (IGSCC) in the as-fabricated condition, this field experience establishes that the conditions under which the materials will crack in core structures are attributable to the combined effects of high residual stresses, associated with the shroud construction, the presence of a more aggressive, oxidizing environment in the core and to microstructural changes in the material. These changes result from the manufacturing process as well as thermal exposure during operation. Studies of materials that have cracked in the field, as well as high temperature material studies in the laboratory, are being performed to understand the mechanisms. The use of highly oxidizing, high purity water environments is integral to reproducing the conditions for cracking. The status of the laboratory efforts to gain understanding and to verify the mechanisms are presented. Modeling of IGSCC is also a key tool used to understand the cracking behavior of the low carbon stainless steels. The PLEDGE (Plant Life Extension Diagnosis by GE) model is used to support the hypotheses that tie together the role of the different contributing elements: residual stress, core water chemistry and microstructural features. The crack growth modeling is also used to evaluate the benefits of different strategies to manage and mitigate cracking of reactor internals such as hydrogen water chemistry.     

Hydrogen water chemistry for BWRs: a status report on the EPRI development program

Many boiling water reactors (BWRs) have experienced extensive intergranular stress corrosion cracking (IGSCC) in their austenitic stainless steel reactor coolant system piping, resulting in serious adverse impacts on plant capacity factors, O&M costs, and personnel radiation exposures. A major research program to provide remedies for BWR pipe cracking was co-funded by EPRI, GE, and the BWR Owners Group for IGSCC Research between 1979 and 1988. Results from this program show that the likelihood of IGSCC depends on reactor water chemistry (particularly on the concentrations of ionic impurities and oxidizing radiolysis products) as well as on material condition and the level of tensile stress. Tests have demonstrated that the concentration of oxidizing radiolysis products in the recirculating reactor water of a BWR can be reduced substantially by injecting hydrogen into the feedwater. Recent plant data show that the use of hydrogen injection can reduce the rate of IGSCC to insignificant levels if the concentration of ionic impurities in the reactor water is kept sufficiently low. This approach to the control of BWR pipe cracking is called hydrogen water chemistry (HWC). This paper presents a review of the results of EPRI's HWC development program from 1980 to the present. In addition, plans for additional work to investigate the feasibility of adapting HWC to protect the BWR vessel and major internal components from potential stress corrosion cracking problems are summarized.     

Evaluation of stainless steel pipe cracking: Causes and fixes

This paper discusses (1) studies of impurity effects on susceptibility to intergranular stress corrosion cracking (IGSCC), (2) intergranular crack growth rate measurements, (3) finite-element studies of the residual stresses produced by induction heating stress improvement (IHSI) and the addition of weld overlays to flawed piping, (4) leak-before-break analyses of piping with 360° part-through cracks, and (5) parametric studies on the effect of through-wall residual stresses on intergranular crack growth behavior in large diameter piping weldments. The studies on the effect of impurities on IGSCC of Type 304 stainless steel show a strong synergistic interaction between dissolved oxygen and impurity concentration of the water. Low carbon stainless steel (Type 316NG) appear resistant to IGSCC even in impurity environments. However, they can become susceptible to transgranular SCC with low levels of sulfate or chloride present in the environment. The finite-element calculations show that IHSI and the weld overlay produce compressive residual stresses on the inner surface, and that the stresses at the crack tip remain compressive under design loads at least for shallow cracks.     

核级316L不锈钢管道射线插塞孔开裂原因分析

为查明某核电厂核级316L奥氏体不锈钢管道射线插塞孔裂纹显示的成因,对含插塞孔不锈钢管段的宏/微观形貌、化学成分、力学性能、维氏硬度、断口形貌、腐蚀产物、应力分布等进行了分析。结果表明：裂纹以沿晶方式扩展,断口呈冰糖块脆性断裂花样并伴有大量氧化腐蚀产物,属于典型的压水堆一回路水介质条件下由插塞孔局部应变-硬化导致的晶间应力腐蚀开裂。引起应变-硬化的主要原因是插塞孔和插塞的过盈配合以及射线插塞孔密封焊缝焊接残余应力过高。建议加强在役机组同类结构的检查,减少新建机组类似结构的使用。     

The evaluation of intergranular stress corrosion cracking problems of stainless steel piping in Taiwan BWR-6 nuclear power plant

Taiwan BWR-6 Kuosheng Nuclear Power Plant Unit 1 implemented the inspection of the intergranular stress corrosion cracking (IGSCC)-susceptible weldments of stainless steel piping in the reactor recirculation, reactor water clean-up, residual heat removal, core spray and feedwater systems. The purpose of this paper is to present the status of the fracture problems in the weldments. The crack growth analysis due to IGSCC and the standard weld overlay design based on the ASME Code Section XI and NUREG-0313 Rev. 2 for the fracture weldments are discussed in detail. Then, the contingent programs including the inspection program, fracture evaluation, and the standard weld overlay, are completely established to prevent pipe break during the reactor operation.     

Intergranular stress corrosion cracking initiation model of austenite stainless steels used in BWRs for optimized water chemistry control

A calculation model on intergranular stress corrosion cracking (IGSCC) initiation time of materials used in boiling water reactors (BWRs) has been developed to evaluate effectiveness of water chemistry control for mitigation of the IGSCC. The model was composed of four terms which determine passive film break time: (1) a chemical term based on electrochemical corrosion potential (ECP) and impurity concentration; (2) a mechanical term based on strain rate; (3) a material term based on sensitization; and (4) an irradiation term based on acceleration of corrosion by γ-rays and neutron irradiation. The contribution of the chemical term in the passive film break was calculated based on a deterministic local corrosion model. Then, the local corrosion model was modified by adding mechanical acceleration of the film rupture to treat the IGSCC phenomenon. The model could reproduce the behavioral tendency seen in the slow strain rate tensile test on high carbon contents with sensitization heat treatment (for example, 620°C × 24 h). Under BWR operating conditions, IGSCC initiation time could be extended by a factor of 5 by lowering the electric conductivity from 1.0 to 0.06 μS/cm. If the ECP was reduced below the critical potential by a mitigation method, the IGSCC initiation time was predicted to become sufficiently long for pipings and components.     

Sensitization development in austenitic stainless steel piping

Pacific Northwest Laboratory and the Division of Engineering Technology of the US Nuclear Regulatory Commission are conducting a program to determine a method for evaluating welded and repair-welded stainless steel piping for light-water reactor service. Validated models, based on experimental data, are being developed to predict the degree of sensitization (DOS) and the intergranular stress corrosion cracking (IGSCC) susceptibility in the heat-affected zone (HAZ) of the SS weldments. The cumulative effects of material composition, past fabrication procedures, past service exposure, weldment thermomechanical (TM) history, and projected post-repair component life are being considered.     

Corrosion of ZrO2 treated type 304 stainless steels in high temperature pure water with various amounts of hydrogen peroxide

As boiling water reactors (BWRs) age, intergranular stress corrosion cracking (IGSCC) of the structural materials in the reactor piping systems and vessel internals has become a major degradation problem. Several approaches to mitigating IGSCC in the structural components have been developed and investigated. Among them, the technique of inhibitive protective coatings is deemed the most promising one since it is expected to work even in the absence of the well-known hydrogen water chemistry technology.Following our earlier work on exploring the electrochemical characteristics of important oxidizing species on zirconium oxide (ZrO₂) treated Type 304 stainless steels (SSs), we targeted on the characteristics of hydrogen peroxide, which is another strongly oxidizing species in the reactor coolant other than oxygen, in this study. Tests were conducted to determine electrochemical parameters such as electrochemical corrosion potential (ECP), corrosion current density, exchange current density and Tafel constant of the reduction reaction of hydrogen peroxide on 304 SS specimens before and after the ZrO₂ treatment. The surface morphologies of the treated and untreated specimens were examined by scanning electron microscopy, energy dispersive X-ray spectroscopy, and laser Raman spectra. Furthermore, the corrosion mitigation efficiency of ZrO₂ treatment was evaluated by electrochemical polarization tests in simulated BWR environments. Test results showed that there were no significant differences in ECP between the untreated and ZrO₂ treated specimens in the test environments of various hydrogen peroxide concentrations. However, it was found via polarization analysis that the exchange current density of the reduction reaction on and the corrosion current density of the treated specimens were markedly lower than those on and of the untreated ones in the same environments. The ZrO₂ treatment was able to deter the reduction rate of hydrogen peroxide on the Type 304 SS surface.     

Tensile and stress corrosion cracking properties of type 304 stainless steel irradiated to a very high dose

Certain safety-related core internal structural components of light water reactors, usually fabricated from Type 304 or 316 austenitic stainless steels (SSs), accumulate very high levels of irradiation damage (20–100 displacement per atom or dpa) by the end of life. Our databases and mechanistic understanding of the degradation of such highly irradiated components, however, are not well established. A key question is the nature of irradiation-assisted intergranular cracking at very high doses, i.e. is it purely mechanical failure or is it stress-corrosion cracking? In this work, hot-cell tests and microstructural characterization were performed on Type 304 SS from the hexagonal fuel can of the decommissioned EBR-II reactor after irradiation to ≈50 dpa at ≈370 °C. Slow-strain-rate tensile tests were conducted at 289 °C in air and in water at several levels of electrochemical potential (ECP), and microstructural characteristics were analyzed by scanning and transmission electron microscopies. The material deformed significantly by twinning and exhibited surprisingly high ductility in air, but was susceptible to severe intergranular stress corrosion cracking (IGSCC) at high ECP. Low levels of dissolved O and ECP were effective in suppressing the susceptibility of the heavily irradiated material to IGSCC, indicating that the stress corrosion process associated with irradiation-induced grain-boundary Cr depletion, rather than purely mechanical separation of grain boundaries, plays the dominant role. However, although IGSCC was suppressed, the material was susceptible to dislocation channeling at a low ECP, and this susceptibility led to a poor work-hardening capability and low ductility.     

Hydrogen and Hydrazine Co-injection to Mitigate Stress Corrosion Cracking of Structural Materials in Boiling Water Reactors, (VI) The Effect of Ammonia on Intergranular Stress Corrosion Cracking

Hydrogen and hydrazine co-injection into a boiling water reactor was considered as a new mitigation method of stress corrosion cracking (SCC). In this method, some amount of ammonia will be formed by the decomposition of hydrazine. The effect of ammonia on SCC susceptibility was studied over a wide range of electrochemical corrosion potentials (ECPs) in 288_C water by conducting slow strain rate technique SCC experiments (SSRTs). ECP was changed from _0:6V versus the standard hydrogen electrode (V(SHE)) to 0.1 V(SHE) by controlling dissolved oxygen concentration. Ammonia concentration was controlled to have values of 100 and 530 ppb. Similarly, sulfuric acid was injected to confirm the difference in the effect of injected chemical compounds on SCC susceptibility. The intergranular stress corrosion cracking (IGSCC) fraction, which was used as the index of SCC susceptibility, decreased with decreasing ECP for the case of no chemical injection. Sulfuric acid enhanced the IGSCC fraction. These data were in good agreement with literature data. On the other hand, ammonia at less than 530 ppb did not affect IGSCC fraction. It is expected that 51–280 ppb hydrazine and 0–53 ppb hydrogen will be injected into reactor water to mitigate SCC in BWRs. In the bottom region of the reactor pressure vessel, ECP and ammonia concentration will be _0:1 V(SHE) and 15–60 ppb, respectively. Under these conditions, ammonia did not affect SCC susceptibility. So SCC susceptibility will be mitigated by decreasing the ECP using hydrazine and hydrogen co-injection.     

Electrochemical noise measurements of stainless steel in high temperature water

Corrosion in a high purity aqueous environment simulating a boiling water reactor (BWR) is addressed in this work. This condition necessitates autoclave experiments under high pressure and temperature.Long-term electrochemical noise measurements were explored as a mean to detect and monitor stress corrosion cracking phenomenon. An experimental set up, designed to insulate the working electrode from external interference, made possible to detect and monitor stress corrosion cracking in slow strain rate tests for sensitized and solution annealed 304 stainless steel at 288 °C. Time-series analysis showed variations in the signature of the current density series due to transgranular stress corrosion cracking (TGSCC) and intergranular stress corrosion cracking (IGSCC).     

Materials aspects of BWR plant life extension

Considerable experience with plant equipment performance in nuclear power stations has indicated that the principal factors limiting the life of BWRs and PWRs are materials related. Specifically, for LWRs it is known that these materials issues generally include parameters related to stress corrosion cracking, corrosion fatigue, wear and radiation embrittlement. Not only do these parameters affect and limit the actual useful design life of plant components but also affect the plant's operating availability. In all these cases, the elimination or control of one or more of these critical parameters should improve the plants availability and significantly extend the useful service life.In the present paper, research performed to address the intergranular stress corrosion cracking (IGSCC) area is described. Specific emphasis is placed on Type 304 stainless steel which has suffered IGSCC in piping in the heat-affected-zone (HAZs) adjacent to the welds in the BWR primary system. Research has developed and qualified a number of techniques which address the three necessary conditions for IGSCC in BWRs: (1) sensitized microstructure, i.e., chromium depletion at the grain boundaries during welding; (2) over yield tensile stress; and (3) oxygenated (200 ppb) high temperature (288Another potential life-limiting IGSCC phenomenon for certain components, irradiation assisted stress corrosion cracking (IASCC) of stainless steel exposed to a high neutron flux, is also discussed. Unlike the IGSCC, IASCC results in intergranular cracking of annealed material at low stress. Fortunately, preliminary research has indicated that some of the techniques utilized for IGSCC control in piping as well as new controlled impurity level stainless steel alloys may reduce the future potential IASCC concern to an insignificant level.     

Current understanding of radiation-induced degradation in light water reactor structural materials

Current phenomenological knowledge and understanding of mechanisms are reviewed for radiation embrittlement of reactor pressure vessel low alloy steels and irradiation assisted stress corrosion cracking of core internals of stainless steels. Accumulated test data of irradiated materials in light water reactors and microscopic analyses by using state-of-the-art techniques such as a three-dimensional atom probe and electron backscatter diffraction have significantly increased knowledge about microstructural features. Characteristics of solute clusters and deformation microstructures and their contributions to macroscopic material property changes have been clarified to a large extent, which provide keys to understand in the degradation mechanisms. However, there are still fundamental research issues that merit study for long-term operation of reactors that requires reliable quantitative prediction of radiation-induced degradation of component materials in low-dose rate high-dose conditions.     

Environmental controls for higher temperature direct-cycle light water reactors

Radiolysis modeling is used to estimate the minimum hydrogen concentration to activate platinum catalysts and reduce the electrochemical corrosion potential in light water reactors. Platinum catalysts are used in boiling water reactors to catalyze hydrogen and oxygen recombination, which reduces the corrosion potential and the susceptibility of austenitic structural materials to intergranular stress corrosion cracking. Two environmental challenges for material performance in higher temperature light water reactors are the increased susceptibility of austenitic materials to stress corrosion cracking and the higher production rate of oxidizing radiolytic species. For a reference supercritical water reactor, a hydrogen addition rate of 2 standard cubic feet per minute is needed to significantly reduce the susceptibility of austenitic materials to stress corrosion cracking. Also, for a reference higher temperature boiling water reactor, a hydrogen addition rate of 10 standard cubic feet per minute of hydrogen reduces the stress corrosion crack susceptibility of austenitic materials located in the lower portion of the reactor vessel.     

压水堆压力壳堆焊不锈钢的动电位再活化行为与晶间应力腐蚀破裂敏感性

用动电位再活化方法研究了压水堆压力壳堆焊不锈钢衬里材料的活化与再活化行为以及晶界形貌,用高温水恒变形和慢应变速率应力腐蚀破裂试验比较了晶间应力腐蚀破裂的敏感程应。结果表明再活化行为与晶间应力腐蚀破裂敏感性之间有一致关系。动电位再活化有可能作为核动力装置中不锈钢焊接件晶间应力腐蚀破裂敏感程度的无损检验方法。     

Characterization of the stress corrosion cracking behavior of thermally sensitized 20Cr-25Ni stainless steel in a simulated cooling pond environment

Niobium stabilized 20Cr-25Ni stainless steel is used for nuclear fuel cladding in the UK's fleet of advanced gas cooled reactors (AGRs). The cladding can have chromium-depleted grain boundaries as a consequence of irradiation in a reactor core, rendering a small proportion of cladding susceptible to intergranular stress corrosion cracking in cooling pond waters after removal from the reactor. In this work, thermal sensitization was used to simulate chromium depletion and the sensitized material was assessed for its susceptibility to pitting corrosion and stress corrosion cracking using slow strain rate testing (SSRT). Elevated chloride concentrations were used to accelerate corrosion initiation and propagation. In 10 ppm chloride and 80 °C, the pitting potential was at potentials between +375 mV and +400 mV (SCE). SSRT appeared to lower the pitting potential, with intergranular corrosion and intergranular stress corrosion cracks observed to nucleate at potentials of +200 mV (SCE).