This paper describes the analysis of the failure of a Zr-2.5Nb pressure tube in a CANDU reactor. The failure sequence was established as: (1) the existence of an undetected manufacturing flaw in the form of a lamination, (2) in-service development of the flaw by oxidation of the lamination, (3) delayed hydride cracking, which extended the flaw through the wall of the tube, resulting in leakage, and (4) rupture of the tube by cold pressurization while the reactor was shut down. The comprehensive failure analysis led to a remedial action plan that permitted the reactor to be returned to full-power operation and ensured a low probability of a similar occurrence for all CANDU reactors.

Jet pumps, which have no moving parts, provide a continuous circulation path for a major portion of the core coolant flow in a boiling water reactor. Part of the pump is held in place by a beam-and-bolt assembly, wherein the beam is preloaded by the bolt. The Alloy X-750 beams had been heat treated by heating at 885 deg C (1625 deg F) for 24 h and aging at 705 deg C (1300 deg F) for 20 h. Jet pump beams were found to have failed in two nuclear reactors, and other beams were found to be cracked. Investigation (visual inspection, metallurgical examination, tension testing, and simulated service testing in oxygenated water) supported the conclusion that intergranular SCC under sustained bending loading was responsible for the failure. The location of the cracking was consistent with the results of stress analysis of the part. Recommendations included either replacing the beams, reheat treatment, or preload reduction.

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.

Two AISI type 316 stainless steel components intended for use in a reducer section for sodium piping in a fast breeder test reactor were found to be severely corroded—the first soon after pickling, and the second after passivation treatments. Metallographic examination revealed that one of the components was in a highly sensitized condition and that the pickling and passivation had resulted in severe intergranular corrosion. The other component was fabricated from thick plate and, after machining, the outer surface represented the transverse section of the original plate. Pickling and passivation resulted in severe pitting because of end-grain effect. Strict control of heat treatment parameters to prevent sensitization and modification of pickling and passivating conditions for machined components were recommended.

Cracking occurred in an ASME SB166 Inconel 600 safe-end forging on a nuclear reactor coolant water recirculation nozzle while it was in service. The safe-end was welded to a stainless-steel-clad carbon steel nozzle and a type 316 stainless steel transition metal pipe segment. An Inconel 600 thermal sleeve was welded to the safe-end, and a repair weld had obviously been made on the outside surface of the safe-end to correct a machining error. Initial visual examination of the safe-end disclosed that the cracking extended over approximately 85 deg of the circular circumference of the piece. Investigation (visual inspection, on-site radiographic inspection, limited ultrasonic inspection, chemical analysis, 53x metallographic cross sections and SEM images etched in 8:1 phosphoric acid) supported the conclusion that the cracking mechanism was intergranular SCC. No recommendations were made.

Argonne National Laboratory has conducted analyses of failed components from nuclear power-generating stations since 1974. The considerations involved in working with and analyzing radioactive components are reviewed here, and the decontamination of these components is discussed. Analyses of four failed components from nuclear plants are then described to illustrate the kinds of failures seen in service. The failures discussed are (1) intergranular stress-corrosion cracking of core spray injection piping in a boiling water reactor, (2) failure of canopy seal welds in adapter tube assemblies in the control rod drive head of a pressurized water reactor, (3) thermal fatigue of a recirculation pump shaft in a boiling water reactor, and (4) failure of pump seal wear rings by nickel leaching in a boiling water reactor.

AM350 stainless steel bellows used in the control rod drive mechanism of a fast breeder reactor failed after 1000 h of service in sodium at 550 deg C (1020 deg F). Helium leak testing indicated that leaks had occurred at various regions of the welded joints between the convolutes in the bellows. The weld failure was attributed to poor quality assurance during fabrication, which resulted in cracklike openings at the fusion zone. The openings extended during tensile loading. Use of proper welding procedures and quality control measures were recommended to prevent future failures.

A gear belonging to a pressurized heavy-water reactor refueling machine failed after 10 years in service. The material specified for the gear was a type C90700 bronze. Macroscopic examination focused on three gear teeth that had fractured completely at the roots, and fracture zones were examined using SEM microscopy. Failure of the gears was attributed to heavy wear resulting from misalignment. A lack of adequate lubrication was also noted. Periodic alignment adjustment and lubrication were recommended.

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.

Leaks developed at random locations in aluminum brass condenser tubes within the first year of operation of a steam condenser in a nuclear power plant. One failed tube underwent scanning electron microscopy surface examination and optical microscope metallography. It was determined that the tube failed from crevice corrosion under seawater deposits that had formed on the inner surface. Mechanical cleaning of the condenser tubes every 6 months and installation of intake screens of smaller mesh size were recommended.

Pressure testing of a batch of AISI type 316L stainless steel thermowells intended for use in a nuclear power-plant resulted in the identification of leakage at the tips in 20% of the parts. Radiography at the tip region of representative thermowells showed linear indications along the axes. SEM examination revealed the presence of longitudinally oriented nonmetallic inclusions that were partly retained and partly dislodged. Electron-dispersive x-ray analysis indicated that the inclusions were composed of CaO. Based on the overall chemistry of the inclusion sites, the source of the CaO was determined to be slag entrapment during the steel making process. It was recommended that the thermowell blanks be ultrasonically tested prior to machining and that the design be modified to make internal pressurization possible.

A cast housing, part of a multi-shaft yoking mechanism, failed during assembly and installation of the equipment in which it was to be used. The housing, or yoke body, was cast from AISI 420 grade ferritic stainless steel. Analysis revealed that the failure was caused by the presence of shrinkage cavities, which lowered the load-bearing capability. The failure occurred at the location where there was an abrupt change in the section thickness. A redesign to provide a smooth contour at the section junction was recommended along with optimization of casting parameters to avoid shrinkage cavities.

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.

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.

A number of AISI 347 stainless steel bellows intended for use in the control rod drive mechanism of a fast breeder reactor were found to be leaking before being placed in service. The bellows, which had been in storage for one year in a seacoast environment, exhibited a leak rate on the order of 1 x 10−7 cu cm/s (6 x 10−8 cu in./s). Optical metallography revealed numerous pits and cracks on the surfaces of the bellow convolutes, which had been welded to one another using an autogenous gas tungsten arc welding process. Microhardness measurements indicated that the bellows had not been adequately stress relieved. It was recommended that a complete stress-relieving treatment be applied to the formed bellows. Improvement of storage conditions to avoid direct and prolonged contact of the bellows with the humid, chloride-containing environment was also recommended.

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.

The self-powered flux detectors used in some nuclear reactors are Pt or V-cored co-axial cables with MgO as an insulator and Inconel 600 as the outer sheath material. The detectors are designed to operate in a He atmosphere; to maximize the conduction of heat (generated from the interaction with gamma radiation) and to prevent corrosion. A number of failures have occurred over the years because of a loss of the He cover gas in the assembly. This has resulted in either acid attack on the Inconel 600 sheath in a wet environment or gaseous corrosion in a dry environment. In the latter case, nitriding and embrittlement occurred at temperatures as low as 300 to 400 deg C (determined from an examination of the oxidation of the Zircaloy-2 carrier rod on which the detectors were mounted). Recent results are described and discussed in terms of the oxidation and nitriding kinetics of Zircaloy-2 and Inconel 600, respectively.

Cracking occurred in type 304L stainless steel sheaths on nichrome wire heaters that had been in storage for about 5 years in a coastal atmosphere. The cracks were discovered when the heater coils were removed from storage in their original polyethylene packing materials and straightened for use. Fractography established that fracture occurred by stress-corrosion cracking. The cracks originated at rusted areas on the cladding that occurred under iron particles left on the surface during manufacture. High hardness values indicated that solution annealing following cold working had not been carried out as specified. It was recommended that the sheathing material be fully annealed and that the outer surface be pickled and passivated.

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.

The dished ends of a heavy water/helium storage tank manufactured from 8 mm (0.3 in.) thick type 304 stainless plate leaked during hydrotesting. Repeated attempts at repair welding did not alleviate the problem. Examination of samples from one dished end revealed that the cracking was confined to the heat affected zone (HAZ) surrounding circumferential welds and, to a lesser extent, radial welds that were part of the original construction. Most of the cracks initiated and propagated from the inside surface of the dished ends. Microstructures of the base metal, HAZ, and weld metal indicated severe sensitization in the HAZ due to high heat input during welding. An intergranular corrosion test confirmed the observations. The severe sensitization was coupled with residual stresses and exposure of the assembly to a coastal atmosphere during storage prior to installation. This combination of factors resulted in failure by stress-corrosion cracking. Implementation of a new repair procedure was recommended. Repairs were successfully made using the new procedure, and all cracks in the weld repair zones were eliminated.