Failure Of Steam Generator Tubes Research Articles

Researches on thermal hydraulics and fuel performance of ATFs during extreme steam generator tube failure without ECCS and DHRS in NuScale

The small modular reactor NuScale has completed Standard Design Approval of Nuclear Regulatory Commission review by good safety performance and it makes history with the first SMR to receive NRC design approval. In this paper, to ensure the safety of the fuel rod in steam generator tube failure, accident tolerant fuels are selected as candidates for the traditional fuel rod. The primary and second circuit of NuScale were modeled, and compared with standard data to ensure accuracy. Thermal hydraulics and fuel performance of ATFs at NuScale during SGTF are analyzed. The results show that the fuel centerline temperature lower by loading with Th0.923U0.077O2-SiC ATF system in the steady-state. Moreover, it can prolong the time for radiotoxicity to leak during SGTF. In short, the safety of NuScale is improved with the replacement of ATF fuel system.

A Bayesian updating of crack distributions in steam generator tubes

The structural failure of steam generator tubes is a common problem that can a ect the availability and safety of nuclear power plants. To minimize the probability of occurrence of failure, it is needed to implement maintenance strategies such as periodic nondestructive inspections of tubes. Thus, a tube is repaired or plugged whenever it has detected a crack which a threshold size is overtaken. In general, uncertainties and errors in crack sizes are associated with the nondestructive inspections. These uncertainties and errors should be appropriately characterized to estimate the actual crack distribution. This work proposes a Bayesian approach for updating crack distributions, which in turn allows computing the failure probability of steam generator tubes at current and future times. The failure criterion is based on plastic collapse phenomenon, and the failure probability is computed by using the Monte-Carlo simulation. The failure probability at current and future times is in good agreement with the ones presented in the literature.

Improvement of steam generator tube failure propagation analysis code LEAP for evaluation of overheating rupture

ABSTRACT As a part of safety assessment or design of steam generators of sodium-cooled fast reactors, it is necessary to evaluate the water leak rate under sodium–water reaction accident. The computer code called LEAP-II calculating a design basis water leak rate during long-time event progress including self-wastage, target-wastage, wastage-type failure propagation, water leak detection, and water/steam blowdown was developed for the prototype fast reactor in the past studies. In this study, a numerical analysis method to predict occurrence of overheating tube rupture was constructed and integrated into this code to expand its application range. The newly constructed method consists of the elemental analysis models for temperature distribution formed by a reacting jet, water-side thermal hydraulics, heat transfer at the tube wall, temperature and stress of the tube, and failure of the tube. Applicability of the method was investigated through the numerical analysis of the experiment on water vapor discharging into liquid sodium pool under the actual condition of the steam generator. The numerical analysis demonstrated that the method could provide the appropriately conservative result on the overheating-rupture-type failure propagation.

Research on Fluid-Structure Interaction Dynamic Characteristics of Steam Generator Heat Exchanger Tubes

For heat exchanger tube of steam generator, the relation between heat exchanger tube and fluid is typical fluid-structure interaction problem. Flow induced vibration has been found so far to be responsible for fatigue damage and failure of steam generator tubes, which will result in large economic loss and radioactive pollution. So the steam generator tubes are the weakest link in the primary coolant loop. Based on the synthesis of all sorts of factors influencing the dynamic characteristics of steam generator heat transfer tubes, establishing the heat transfer tube model, research on the weakening effect of fluid hole on fluid, the natural frequencies of the heat transfer tubes are analyzed under different fluid holes and fluid hole distance by numerical simulation.

Investigation on droplet impingement erosion during steam generator tube failure accident

Abstract Droplet impingement erosion is one of the reasons causing the secondary heat transfer tube damage in case of the heat transfer tube failure in the steam generator of sodium-cooled fast reactor. This paper has been dedicated to investigate the impingement phenomena, pressure load and the damage rate by impingement. Single droplet impingement on the rigid wall was first simulated with MPS-AS (moving particle semi-implicit method for all speeds) method for water and sodium. The shockwave propagation during the impingement has been well captured. The pressure load by water impingement can be characterized with Heymann's correlation (1968) . A correlation for pressure load by sodium impingement is proposed based on the water hammer theory. The new correlation shows to be consistent with the simulation result. For prediction of erosion rate by sodium droplet impingement, Heymann's empirical correlations (1979) were extended by including the effects of liquid properties, i.e. density, sound speed and viscosity. The extended correlations show similar prediction accuracy to the original ones.

Role of oxide notching and degraded alloy microstructure in remarkably premature failure of steam generator tubes

Failure analysis investigation was carried out on a ruptured primary superheater tube of a boiler unit in a power plant. The rupture of 1Cr–0.5Mo steel tube had caused steam leakage and led to boiler shutdown after only 17,520 h of operation. The physical and mechanical characteristics of the alloy and the oxide scales in the area of failure were characterized by visual inspection, optical microscopy and microhardness testing, whereas the chemical variations were examined by optical emission spectrometry and energy dispersive spectrometry (EDS). A distinct surface/sub-surface zone of decarburization that was observed both on the inner and outer sides of the tube was developed before the tubes were put to the service. The fracture was attributed to a prolonged overheating as a result of the oxide scale build-up, causing loss of creep strength. The decarburized surface/sub-surface facilitated crack initiation/propagation.

Investigation on behavior of trapezoidal through-wall cracks under constant pressure loading

In general, a crack found in a steam generator tube has an elliptical shape, and as soon as it penetrates through the wall, it is assumed to be rectangular for conservatism. However, the leak and crack growth behavior have not been clearly understood after the elliptical crack penetrates the tube wall. One steam generator tube failure showed the tube rupture behavior without any significant prior indication. Several experiments using steam generator tubes with a trapezoidal crack were performed under constant pressure loading by Argonne National Laboratory and exhibited crack growth behavior much faster than expected. It has been found to be caused by time-dependent fatigue crack growth. No K I solution is available for a trapezoidal crack which can be used for fatigue crack growth analysis. The objective of this study is, therefore, to develop a new K I solution for a trapezoidal crack using FEM.

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.

Fluidelastic instability analysis of operating nuclear steam generator U-tubes

This paper presents a systematic assessment methodology of the potential for steam generator tube failures caused by fluidelastic instability in operating nuclear power reactors and provides the results of assessment for the U-tube steam generator (UTSG) model being employed at Kori units 2, 3 and 4, and Yonggwang units 1 and 2 in Korea. The assessment process involves evaluation of anti-vibration bar insertion conditions for the UTSG, three-dimensional thermal-hydraulic analysis of the steam generator, determination of flow distributions along the length of a specific U-tube, calculation of natural frequencies and mode shapes of the tube, and fluidelastic tube instability analysis. The thermal-hydraulic analysis for providing the detailed three-dimensional two-phase flow field in the secondary side of the steam generator model was accomplished using the ATHOS3 steam generator thermal-hydraulic analysis code. The UTVA code designed for calculating both the free vibration responses and fluidelastic stability ratio of a specific U-tube was used to assess the potential for fluidelastic instability of the steam generator U-tubes at various conditions of anti-vibration bar (AVB) inactive modes. In addition, the effects of tube plugging on the forced response of either plugged or intact tubes were discussed.

Assessment of primary water stress corrosion cracking of PWR steam generator tubes

This paper presents a model for assessing primary water stress corrosion cracking (PWSCC) damage to pressurized water reactor (PWR) steam generator tubes. PWSCC damage has been detected at three locations in the recirculating steam generator: expansion transition regions, U-bends, and tube dents. The model accounts for residual stresses, microstructure, and primary coolant temperatures. High residual tensile stresses and low mill-annealing temperatures contribute significantly to PWSCC. PWSCC is a thermally activated process; small increases in operating temperature accelerate damage. The model employs a Weibull distribution to represent tube degradation results obtained from inservice inspections and from destructive examinations. The model can be used to estimate the times to steam generator tube failures caused by PWSCC. This paper also evaluates remedial measures for PWSCC damage and presents a procedure for life assessment of steam generator tubes.

Current and emerging pressure boundary issues — a US regulatory perspective

Abstract Concerns about pressure boundary integrity deal primarily with older plants, and establishing a basis for their continued safe operation. Pressure vessel problems stem from exposure to fast neutrons which changes the Nil-Ductility-Temperature (NDT) and the elevated temperature fracture energy of some vessels. The predicted shift in NDT has increased over the last decade as more has been learned about the effect of impurities (copper) and the synergism between nickel and copper. In PWRs this has lead to concern about excursion in which the a vessel remains at high pressure as the coolant temperature drops rapidly, that is the so-called Pressurized Thermal Shock (PTS) accident. In BWRs one cannot have PTS events, but the more rapid than expected rise in NDT due to irradiation is impacting operations. In another set of PWRs the upper shelf energy of the welds was initially low due to the use of a slag which led to many small inclusions in the weld. Radiation has lowered the Charpy fracture energy of these welds to below the 50 ft lb level at which there is concern that the vessel may undergo low energy ductile failure even if cleavage does not occur. Problems in pressure boundary piping has stemmed primarily from corrosion, that is, IGSCC in BWR recirculation piping, and steam generator tube failures in PWRs. These have made a large contribution to downtime and occupational exposure, but are not seen as significant contributors to risk. There has been some concern about the aging (loss of toughness) of cast stainless components with significant ferrite content, especially because inspection by UT is difficult.

Technical Note: Relationship between Chromium Depletion in Sensitized Alloy 600 and U-Bend Cracking in Sodium Thiosulfate Solution

In order to further investigate the stress corrosion cracking (SCC) of Alloy 600 steam generator tube failures as in the Three Mile Island nuclear power plant unit 1 (TMI-1), an Alloy 600 tube was directly sensitized from the mill annealed condition (MAS) by a thermal treatment at 621C for 18 h prior to corrosion cracking studies. Reverse U-bends of the MAS material did not suffer SCC from exposure to sodium thiosulfate solutions up to 0.01M concentration and 80C. Results of past experiments have indicated that the U-bends cracked readily in a solution containing a few ppM sodium tetrathionate leading to the conclusion that some reduced species of sulfur could be the cause for SCC of the TMI-1 tubes. Tests were conducted on various specimens of MAS material, and new results indicate that SCC was very rapid and occurred in very dilute levels of thiosulfate. It is also indicated that SCC initiated by thiosulfate probably occurs only after a critical or threshold level of chromium depletion is achieved. The results also most pointedly indicate that processing variations from mill to mill in MAS material cause wide variations in structure and related properties.