A metallurgical failure analysis was performed on pieces of the cracked vent header pipe from the Edwin I. Hatch Unit 2 Nuclear power plant. The analysis consisted of optical microscopy, chemical analysis, mechanical Charpy impact testing, and fractography. It was found that the material of the vent header met the mechanical and chemical properties of ASTM A516 Grade 70 carbon-manganese steel material and microstructures were consistent with this material. Fracture faces of the cracked pipe were predominantly brittle in appearance with no evidence of fatigue contribution. The NDTT (Nil ductility Transition Temperature) for this material was approximately -51 deg C (-60 deg F). The fact that the material's NDTT was significantly out of the normal operating range of the pipe suggested an impingement of low temperature nitrogen (caused by a faulty torus inerting system) induced a thermal shock in the pipe which, when cooled below its NDTT, cracked in a brittle manner.

Three ASME SA106 grade B carbon steel feed water piping reducers from a pressurized water reactor showed indications of flaws near welds during ultrasonic testing. Further examination and testing indicated that the cracks resulted from a low-cycle corrosion fatigue phenomenon.

This article provides information on life assessment strategies and conceptually illustrates the interplay of nondestructive evaluation (NDE) and fracture mechanics in the damage tolerant approach. It presents information on probability of detection (POD) and probability of false alarm (PFA). The article describes the damage tolerance approach to life management of cyclic-limited engine components and lists the commonly used nondestructive evaluation methods. It concludes with an illustration on the role of NDE, as quantified by POD, in fully probabilistic life management.

A double-walled, hemispherical metal beam exit window made of alloy 718 developed a crack during service, leading to coolant leakage. The window had been exposed to radiation damage from 800 MeV protons and a cyclic stress from 600 MPa tensile to near zero induced by numerous temperature cycles calculated to be from 400 to 30 deg C (752 to 86 deg F). The window was activated to >200 Sv/h. It was determined through analysis using remote handling techniques and hot cells that the crack initiated near a spot weld used to affix thermocouples to the window surface. In addition to analysis of the crack, some of the irradiated material from the window was used to measure mechanical properties. Hot cell techniques for preparation of samples and testing were developed to determine true operating conditions of radiation, strain, and temperature.

Liquid penetrant inspection of an ASTM A296 grade CA-15 residual heat removal pump impeller from a nuclear plant revealed a crack like indication that approximated the outer contour of the wear ring. Examination of a section containing the crack and three sections from near the main crack indication revealed that the failure was caused by hot cracking related to original weld repairs performed on the impeller casting.

Various aluminum bronze valves and fittings on the essential cooling water system at a nuclear plant were found to be leaking. The leakage was limited to small-bore socket-welded components. Four specimens were examined: three castings (an ASME SB-148 CA 952 elbow from a small-bore fitting and two ASME SB-148 CA 954 valve bodies) and an entire valve assembly. The leaks were found to be in the socket-weld crevice area and had resulted from dealloying. It was recommended that the weld joint geometry be modified.

A section of type 304 stainless steel pipe from a stand by system used for emergency injection of cooling water to a nuclear reactor failed during precommissioning. Leaking occurred in only one spot. Liquid penetrant testing revealed a narrow circumferential crack. Metallographic examination of the cracked area indicated stress-corrosion cracking, which had originated at rusted areas that had formed on longitudinal scratch marks on the outer surface of the pipe. The material was free from sensitization, and there was no significant amount of cold work. It was recommended that the stainless steel be kept rust free.

A service water pump in a nuclear reactor failed when its shaft gave way. The fracture originated in the threaded portion of the sleeve nut on the drive-end side of the shaft. Results of the failure analysis showed that the cracking initiated at the thread root as a result of corrosion fatigue. Crack propagation occurred either by corrosion or mechanical fatigue. Evidence was found indicating high rotary bending stresses on the shaft during operation. The nonstandard composition of the En 8 steel used in the shaft and irregular maintenance reduced the life of the shaft. Recommendations included use of a case-hardened En 8 steel with the correct composition and regular maintenance of the pump.

This article provides an outline of the issues to consider in performing a probabilistic life assessment. It begins with an historical background and introduces the most common methods. The article then describes those methods covering subjects such as the required random variable definitions, how uncertainty is quantified, and input for the associated random variables, as well as the characterization of the response uncertainty. Next, it focuses on specific and generic uncertainty propagation techniques: first- and second-order reliability methods, the response surface method, and the most frequently used simulation methods, standard Monte Carlo sampling, Latin hypercube sampling, and discrete probability distribution sampling. Further, the article discusses methods developed to analyze the results of probabilistic methods and covers the use of epistemic and aleatory sampling as well as several statistical techniques. Finally, it illustrates some of the techniques with application problems for which probabilistic analysis is an essential element.

Large quantities of aluminum-clad spent nuclear materials have been in interim storage in the fuel storage basins at The Savannah River Site while awaiting processing since 1989. This extended storage as a result of a moratorium on processing resulted in corrosion of the aluminum clad. Examinations of this fuel and other data from a corrosion surveillance program in the water basins have provided basic insight into the corrosion process and have resulted in improvements in the storage facilities and basin operations. Since these improvements were implemented, there has been no new initiation of pitting observed since 1993. This paper describes the corrosion of spent fuel and the metallographic examination of Mark 31A target slugs removed from the K-basin storage pool after 5 years of storage. It discusses the SRS Corrosion Surveillance Program and the improvements made to the storage facilities which have mitigated new corrosion in the basins.

Leaks developed in 22 admiralty brass condenser tubes. The tubes were part of a condenser that was being used to condense steam from a nuclear power plant and had been in operation for less than 2 years. Analysis identified three types of failure modes: stress-corrosion cracking, corrosion under deposit (pitting and crevice), and dezincification. Fractures were transgranular and typical of stress-corrosion cracking. The primary cause of the corrosion deposit was low-flow conditions in those parts of the condenser where failure occurred. Maintenance of proper flow conditions was recommended.

A 460 mm (18 in.) diam suction line to the main feed water pump for a nuclear power plant failed in a violent, catastrophic manner. Samples of pipe, elbow, and weld materials (ASTM A106 grade B carbon steel, ASTM A234 grade WPB carbon steel, and E7018 carbon steel electrode, respectively) from the suction line were analyzed. Evidence of overall thinning of the elbow and pipe material and ductile tearing of fractures indicated that the feed water pipe failed as a result of an erosion corrosion mechanism, which thinned the wall sufficiently to cause rapid, ductile tearing of the material after its design stress had been exceeded. It was recommended that steel with a higher chromium content be used to mitigate the erosion corrosion potential in the lines and that more rigorous nondestructive (ultrasonic) examinations be performed.

A pair of steam generators operating at a pressurized water reactor site were found to be leaking near a closure weld. The generators were the vertical U-tube type, constructed from ASTM A302 grade B steel. The shell material exhibited high hardness values prior to confirmatory heat treatment, indicating high residual stresses in the area of the weld. All cracks were transgranular and were associated with pits on the inside surfaces of the vessels. It was concluded that the cracking was caused by a low-cycle corrosion fatigue phenomenon, with cracks initiating at areas of localized corrosion and propagating by fatigue. The cause of the pitting/cracking was related to the unit's copper species in solution.

Nuclear power plants typically experience two or three high-cycle fatigue failures of stainless steel socket-welded connections in small bore piping during each plant-year of operation. This paper discusses fatigue-induced failure in socket-welded joints and the strategy Texas Utilities Electric Company (TU Electric) has implemented in response to these failures. High-cycle fatigue is invisible to proven commercial nondestructive evaluation (NDE) methods during crack initiation and the initial phases of crack growth. Under a constant applied stress, cracks grow at accelerating rates, which means cracks extend from a detectable size to a through-wall crack in a relatively short time. When fatigue cracks grow large enough to be visible to NDE, it is likely that the component is near the end of its useful life. TU Electric has determined that an inspection program designed to detect a crack prior to the component leaking would involve frequent inspections at a given location and that the cost of the inspection program would far exceed the benefits of avoiding a leak. Instead, TU Electric locates these cracks by visually monitoring for leaks. Field experience with fatigue-induced cracks in socket-welded joints has confirmed that visual monitoring does detect cracks in a timely manner, that these cracks do not result in catastrophic failures, and that the plant can be safely shut down in spite of a leaking socket-welded joint in a small bore pipe. Historical data from TU Electric and Southwest Research Institute are presented regarding the frequency of failures, failure locations, and the potential causes. The topics addressed include 1) metallurgical and fractographic features of fatigue cracks at the weld toe and weld root; 2) factors that are associated with fatigue, such as mechanical vibration, internal pulsation, joint design, and welding workmanship; and 3) implications of a leaking crack on plant safety. TU Electric has implemented the use of modified welding techniques for the fabrication of socket-welded joints that are expected to improve their ability to tolerate fatigue.

The U-0.8wt%Ti alloy is often used in weapon applications where high strength and fairly good ductility are necessary. Components are immersion quenched in water from the gamma phase to produce a martensitic structure that is amenable to aging. Undesirable conditions occur when a component occasionally cracks during the quenching process, and when tensile specimens fail prematurely during mechanical testing. These two failures prompted an investigative analysis and a series of studies to determine the causes of the cracking and erratic behavior observed in this alloy. Quench-related failures whereby components that cracked either during or immediately after the heat treatment/quenching operation were sectioned for metallographic examination of the microstructure to examine the degree of phase transformation. Examination of premature tensile specimen failures by scanning electron microscopy and X-ray imaging of fracture surfaces revealed pockets of inclusions at the crack origins. In addition, tests were conducted to evaluate the detrimental effects of internal hydrogen on ductility and crack initiation in this alloy.

A heat transport pump in a heavy water reactor failed (exhibiting excessive vibration) during a restart following a brief interruption in coolant flow due to a faulty valve. The pump had developed a large crack across the entire length of a bearing journal. An investigation to establish the root cause of the failure included chemical and metallurgical analysis, scanning electron fractography, mechanical property testing, finite element analysis of the shrink fitted journal, and a design review of the assembly fits. The journal failure was attributed to corrosion fatigue. Corrective actions to make the journals less susceptible to future failures were implemented and the process by which they were developed is described.

Erosion of a solid surface can be brought about by liquid droplet impingement (LDI), which is defined as "progressive loss of original material from a solid surface due to continued exposure to erosion by liquid droplets." In this article, the emphasis is placed on the damage mechanism of LDI erosion under the influence of a liquid film and surface roughness and on the prediction of LDI erosion. The fundamentals of LDI and processes involved in initiation of erosion are also discussed.

The secondary cooling water system pressure boundary of Savannah River Site reactors includes expansion joints utilizing a thin-wall bellows. While successfully used for over thirty years, an occasional replacement has been required because of the development of small, circumferential fatigue cracks in a bellows convolute. One such crack was recently shown to have initiated from a weld heat-affected zone liquation microcrack. The crack, initially open to the outer surface of the rolled and seam welded cylindrical bellows section, was closed when cold forming of the convolutes placed the outer surface in residual compression. However, the bellows was placed in tension when installed, and the tensile stresses reopened the microcrack. This five to eight grain diameter microcrack was extended by ductile fatigue processes. Initial extension was by relatively rapid propagation through the large-grained weld metal, followed by slower extension through the fine-grained base metal. A significant through-wall crack was not developed until the crack extended into the base metal on both sides of the weld. Leakage of cooling water was subsequently detected and the bellows removed and a replacement installed.

Two sand-cast low-alloy steel equalizer beams (ASTM A 148, grade 105-85) designed to distribute the load to the axles of a highway truck broke after an unreported length of service. Normal service life would have been about 805,000 km (500,000 mi) of truck operation. Investigation (visual inspection, chemical analysis, tensile testing, unetched 65x and 1% nital etched 65x magnification) supported the conclusions that the steel was too soft for the application – probably due to improper heat treatment. Fracture of the equalizer beams resulted from growth of mechanical cracks that were formed before the castings were heat treated. Recommendations included the following changes in processing: better gating and risering in the foundry to achieve sounder castings; better shakeout practice to avoid mechanical damage; better inspection to detect imperfections; and normalizing and tempering to achieve better mechanical properties.

A crack was discovered in a cast steel (ASTM A 356, grade 6) steam turbine casing during normal overhaul of the turbine. The mechanical properties of the casting all exceeded the requirements of the specification. When the fracture surface was examined visually, an internal-porosity defect was observed adjoining a tapped hole. A second, much larger cavity was also detected. Investigation (visual inspection and 7500x SEM fractographs) supported the conclusions that failure occurred through a zone of structural weakness that was caused by internal casting defects and a tapped hole. The combination of cyclic loading (thermal fatigue), an aggressive service environment (steam), and internal defects resulted in gradual crack propagation, which was, at times, intergranular-with or without corrosive attack-and, at other times, was transgranular.