Outer Diameter Stress Corrosion Cracking Research Articles

Role of aluminum and silicon species in the formation of hard sludge on the secondary side of PWR steam generators: Potential geopolymers contribution

Despite the exhaustive chemical control, modern pressurized water reactors (PWRs) steam generators (SGs) have experienced damage associated with the formation of hard sludge on top of tubesheet. This phenomenon is detrimental because it is linked to tube degradation problems on the secondary side of SGs by corrosion processes, such as outer diameter stress corrosion cracking (ODSCC) enhanced by tube deformation or denting at the top of tubesheet. There is evidence of the contribution of aluminum and silicon species to the binding and adhesion of iron oxides particles in the formation and consolidation of hard sludge during normal operation. In fact, characterization of SG deposit samples mainly constituted by hard sludge determines high concentrations of these species (up to 6.0 ± 0.1 wt% of Al and 2.7 ± 0.1 wt% of Si), together with high microhardness values (above 800 HV) measured at particular locations within collars or hard sludge particles.Taking into account the physicochemical conditions (low temperature and high pH) in the gap between tube and tubesheet of SGs due to the thermal contraction of both materials during transient conditions (cooldown), this work proposes and discusses the possible formation of geopolymers (inorganic aluminosilicate polymers with cement-like mechanical properties formed at low temperatures and high alkaline conditions) during this period. In order to support this proposal, the role played by Si and Al compounds in the formation and consolidation of hard sludge is reviewed, considering information obtained from hard sludge characterization and from experimental studies reported in the literature.

Effect of Heat Treatment and Surface Condition on Inter-Granular Corrosion of Alloy 800

Abstract The alloy 800 is being used for steam generator tubings for pressurized heavy-water reactors (PHWRs). This material is susceptible to degradation because of severe operating conditions, like high temperature, stress, and corrosive environment. These degradation mechanisms include primary water stress corrosion cracking (PWSCC), secondary side or outer diameter stress corrosion cracking (ODSCC), inter-granular corrosion (IGC), fretting, wear, denting, high cycle fatigue, corrosion fatigue, etc. The present study was conducted to analyze the effect of solution annealing temperatures, sensitization treatments, and surface conditions on the corrosion rate. The alloy 800 tubular samples were investigated by means of a conventional corrosion test (according to ASTM G28-02) and an electrochemical potentiokinetic reactivation (EPR) test. Susceptibility to inter-granular corrosion under various experimental parameters was examined by using both the test methods and the results are compared. The observed trends in corrosion rate obtained by using conventional method and EPR test were found to be similar. These results were used to obtain most optimum heat-treatment parameters and surface condition, which will yield best corrosion resistance.

A Study on the Chemical Cleaning Process and Its Qualification Test by Eddy Current Testing

Steam Generator (SG) tube, as a barrier isolating the primary coolant system from the secondary side of nuclear power plants (NPP), must maintain the structural integrity for the public safety and their efficient power generation. So, SG tubes are subject to the periodic examination and the repairs if needed so that any defective tubes are not in service. Recently, corrosion related degradations were detected in the tubes of the domestic OPR-1000 NPP, as a form of axially oriented outer diameter stress corrosion cracking (ODSCC). According to the studies on the factors causing the heat fouling as well as developing corrosion cracking, densely scaled deposits on the secondary side of the SG tubes are mainly known to be problematic causing the adverse impacts against the soundness of the SG tubes [1]. Therefore, the processes of various cleaning methods efficiently to dissolve and remove the deposits have been applied as well as it is imperative to maintain the structural integrity of the tubes after exposing to the cleaning agent. So qualification test (QT) should be carried out to assess the perfection of the chemical cleaning and QT is to apply the processes and to do ECT. In this paper, the chemical cleaning processes to dissolve and remove the scaled deposits are introduced and results of ECT on the artificial crack specimens to determine the effectiveness of those processes are represented.

Manufacturing of representative axial stress corrosion cracks in tube specimens for eddy current testing

A crack manufacturing method is proposed for producing axial outer-diameter stress corrosion cracking (ODSCC) in Alloy 600 steam generator tube specimens. The process can be conducted at ambient conditions and, unlike conventional methods, it does not require complicated facilities. Eddy current testing and destructive examination data for the axial ODSCC specimens were compared with the available field crack data to determine whether those SCC specimens are representative in accordance with an existing industry guideline. Based on the results of our investigations, it was determined that the proposed method could be used to manufacture representative axial ODSCC in Alloy 600 tube specimens.

고온, 고압 알칼리 수용액에서의 Alloy 600 산화막 특성에 미치는 납 농도 영향

0.1 M NaOH 용액에 PbO첨가양이 증가함에 따라 Alloy 600에 형성되는 산화막의 부동태 피막 특성이 열화되었다. 또한 뚜렷한 2중층 구조의 산화막이 점차 사라지고, 산화막내 존재하는 납의 양이 증가하였다. 산화막 내부 납의 양이 증가함에 따라 산화막 내부 니켈의 결핍이 점차 커졌다. 납에 의해 산화막의 부동태 특성이 약화됨에 따라, 응력부식균열 저항성 또한 급감하였을 것으로 판단된다. Outer diameter stress corrosion cracking (ODSCC) has occurred for Alloy 600 (Ni 75 wt%, Cr 15 wt%, Fe 10 wt%) as a heat exchanger tube of the steam generator (SG) in nuclear power plants (NPP) during long term operation. Among many causes for SCC, lead (Pb) is known to be one of the most deleterious species in the secondary system. In the present work, the oxide formed on Alloy 600 was characterized as a function of the PbO content in 0.1 M NaOH at <TEX>$315^{\circ}C$</TEX> by using an electrochemical impedance spectroscopy (EIS), a transmission electron microscopy (TEM), equipped with an energy dispersive x-ray spectroscopy (EDS). The oxide property was analyzed in view of SCC susceptibility.

증기발생기 전열관 재료의 2차측 응력부식균열 민감성

Alloy 600 (Ni 75 wt%, Cr 15 wt%, Fe 10 wt%) as a heat exchanger tube of the steam generator (SG) in nuclear power plants (NPP) has been degraded by various corrosion mechanism during the long-term operation. Especially lead (Pb) is known to be one of the most deleterious species in the secondary system causing outer diameter stress corrosion cracking (ODSCC). Oxide formation and breakdown is requisite for SCC initiation and propagation. Therefore it is expected that a property change of the oxide formed on SG tubing materials by lead addition into a solution is closely related to PbSCC. In the present work, the SCC susceptibility was assessed by using a slow strain rate test (SSRT) in caustic solutions with and without lead for Alloy 600 and Alloy 690 (Ni 60 wt%, Cr 30 wt%, Fe 10 wt%) used as an alternative of Alloy 600 because of outstanding superiority to SCC. The results were discussed in view of the oxide property formed on Alloy 600 and Alloy 690. The oxides formed on Alloy 600 and Alloy 690 in aqueous solutions with and without lead were examined by using a transmission electron microscopy (TEM), equipped with an energy dispersive x-ray spectroscopy (EDXS).

Degradation of Alloy 600 Steam Generator Tubes in Operating Pressurized Water Reactor Nuclear Power Plants

Abstract The Korea Atomic Energy Research Institute (KAERI) and some foreign laboratories have carried out six examinations of pulled tubes from Korean nuclear power plants, since the first commercial operation of Kori Unit 1 in 1978. A total of 15 tubes were extracted that exhibited different types of failures such as pitting, outer diameter stress corrosion cracking (ODSCC), primary water stress corrosion cracking (PWSCC), intergranular attack (IGA), etc. The pitting of the Plant A tube was related to high copper dissolved from the condenser material, and the presence of Cl− and high dissolved oxygen. Transgranular SCC observed at Plant B seemed to be related to lead compounds. ODSCC and IGA in Plant A were deduced to be connected to the formation of an aggressive environment in the crevices. PWSCC in Plant A and Plant C originated from the inherent characteristics of the materials, which were not properly thermally treated. After the failure analysis, the performance of nondestructive testing was evalu...

Overview of steam generator tube degradation and integrity issues

The degradation of steam generator tubes in pressurized water nuclear reactors, and, in particular, the stress corrosion cracking (SCC) of Alloy 600 tubes, continues to be a serious problem. Primary water SCC is commonly observed at the roll transition zone (RTZ), at U-bends and tube denting locations, and occasionally in plugs and sleeves. Outer-diameter SCC (ODSCC) and intergranular attack (IGA) commonly occur near tube support plate (TSP) crevices, near the tube sheet in crevices, or under sludge piles, and occasionally in the free span. A particularly troubling recent trend has been the increasing occurrence of axial and circumferential cracking at the RTZ on both the primary and secondary sides. Outer-diameter stress corrosion cracking in TSP crevices, commonly consisting of segmented axial cracks with interspersed uncracked ligaments, is also becoming more common. Despite recent advances in inservice inspection (ISI) technology, a clear need still exists for quantifying and improving the reliability of ISI methods with respect to the probability of detection of the various types of flaws and their accurate sizing. These improvements are necessary to permit an accurate assessment of the consequences of leaving degraded tubes in service over the next reactor operating cycle. Degradation modes such as circumferential cracking, intergranular attack, and ODSCC at the TSP have affected a large number of tubes. New regulatory guidance is being developed that requires the development and implementation of a steam generator management program that monitors tube condition against accepted performance criteria to ensure that the tubes will perform the required safety function over the next operating cycle. In formulating new guidance for the implementation of alternate repair criteria, the U.S. Nuclear Regulatory Commission is also evaluating the contribution to overall plant risk from severe accidents. Preliminary analyses are being performed for a postulated severe-accident scenario that involves station blackout and loss of primary feedwater, in order to determine the probability of failure for degraded tubes.

Control of Alkaline Stress Corrosion Cracking in Pressurized-Water Reactor Steam Generator Tubing

Abstract Outer-diameter stress corrosion cracking (ODSCC) of alloy 600 (UNS N06600) tubings in steam generators of the Kori-1 pressurized-water reactor (PWR) caused an unscheduled outage in 1994. Failure analysis and remedy development studies were undertaken to avoid a recurrence. Destructive examination of a removed tube indicated axial intergranular cracks developed at the top of sludge caused by a boiling crevice geometry. A high ODSCC propagation rate was attributed to high local pH and increased corrosion potential resulting from oxidized copper presumably formed during the maintenance outage and plant heatup. Remedial measures included: (1) crevice neutralization by crevice flushing with boric acid (H3BO3) and molar ratio control using ammonium chloride (NH4Cl), (2) corrosion potential reduction by hydrazine (H2NNH2) soaking and suppression of oxygen below 20 ppb to avoid copper oxide formation, (3) titanium dioxide (TiO2) inhibitor soaking, and (4) temperature reduction of 5°C. Since application o...