A portion of the wall of a reactor vessel used in burning impurities from carbon particles failed by localized melting. The vessel was made of Hastelloy X (Ni-22Cr-9Mo-18Fe). Considering the service environment, melting could have been caused either by excessive carburization (which would have lowered the melting point of the alloy markedly) or by overheating. A small specimen containing melted and unmelted metal was removed from the vessel wall and examined metallographically. It was observed that the interface between the melted zone and the unaffected base metal was composed of large grains and enlarged grain boundaries. An area a short distance away from the melted zone was fine grained and relatively free of massive carbides. This evidence supported the conclusion that the vessel failed by melting that resulted from heating to about 1230 to 1260 deg C (2250 to 2300 deg F), which exceeded normal operating temperatures, and carburization was not the principal cause of failure. No recommendations were made.

A low-carbon steel (St35.8) tube in a phthalic anhydride reactor system failed. Visual and stereomicroscopic examination of fracture surfaces revealed heavy oxide/deposits on the outer surface of the tube, tube wall thinning in the area of the fracture, and discolorations and oxides/deposits on the inner surface. Cross sections from the fracture surface were metallographically examined, and the deposits were analyzed. It was determined that the tube had thinned from the inner surface because of a localized overheating condition (probably resulting from a runaway chemical reaction within the tube) and then fractured, which allowed molten salt to flow into the tube.

The stub-shaft assembly which was part of the agitator shaft in a polyvinyl chloride reactor, fractured in service after a nut that retained a loose sleeve around the smaller-diam section of the shaft had been tightened several times to reduce leakage. The shaft was made of ASTM A105, grade 2 steel, and the larger-diam section was covered with a type 316 stainless steel end cap. The cap was welded to each end using type ER316 stainless steel filler metal. The forged steel shaft was revealed to have fractured at approximately 90 deg to the shaft axis in the weld metal and not in the heat-affected zone of the forged steel shaft. Microscopic investigation and chemical analysis of the steel shaft revealed presence of martensite (offered a path of easy crack propagation) around the fusion line and dilution of the weld metal by the carbon steel shaft. The microstructure was found to be martensitic as the fusion line was approached. The forged steel shaft was concluded to have failed by ductile fracture and possible reasons were discussed. Corrective measures adopted in the replacement shaft were specified.

A power plant using two steam generators (vertical U-tube and shell heat exchangers, approximately 21 m (68 ft) high with a steam drum diameter of 6 m (20 ft)) experienced a steam generator tube rupture. Each steam generator contained 11,012 Inconel alloy 600 (nickel-base alloy) tubes measuring 19 mm OD, nominal wall thickness of 1.0 mm (0.042 in.), and average length of 18 m (57.75 ft). The original operating temperature of the reactor coolant was 328 deg C (621 deg F). A tube removal effort was conducted following the tube rupture event. Investigation (visual inspection, SEM fractographs, and micrographs) showed evidence of IGSCC initiating at the OD and IGA under ridgelike deposits that were analyzed and found to be slightly alkaline to very alkaline (caustic) in nature. Crack oxide analysis indicated sulfate levels in excess of expected values. The analysis supported the conclusion that that the deposits formed at locations that experienced steam blanketing or dryout at the higher levels of the steam generators. Recommendations included steam generator water-chemistry controls, chemical cleaning, and reduction of the primary reactor coolant system temperature.

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.

A 150 mm (6 in.) schedule 80S type 304 stainless steel pipe (11 mm, or 0.432 in., wall thickness), which had served as an equalizer line in the primary loop of a pressurized-water reactor, was found to contain several circumferential cracks 50 to 100 mm (2 to 4 in.) long. Two of these cracks, which had penetrated the pipe wall, were responsible for leaks detected in a hydrostatic test performed during a general inspection after seven years of service. Investigation (visual inspection, visual and ultrasonic weld examination, water analysis, and chemical analysis) supported the conclusion that the failure was caused by SCC due to stress, sensitization, and environment. Recommendations included replacing all pipe sections and installing them using low-heat-input, multiple-pass welding procedures.

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.

Several tubes in a 35 m 2 (115 ft 2 ) type 316 stainless steel shell-and-tube condenser leaked unexpectedly in an organic chemical plant that produces vinyl acetate monomer. Leaks were discovered after 5 years of operation and relocation of the condenser to another unit in the same plant. Examination of tubes and tube sheets revealed pitting damage on the OD surface. Some of the pits had penetrated fully, resulting in holes. Inside diameter surfaces were free of corrosion. Macro- and microexaminations indicated that the tubes had been properly manufactured. Pitting was attributed to stagnant water on the shell side. It was recommended that the surfaces not be kept in contacts with closed stagnant water for appreciable lengths of time.

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.

An HK-40 alloy tubing weld in a reformer furnace of a petrochemical plant failed by leaking after a shorter time than that predicted by design specifications. Leaking occurred because of cracks that passed through the thickness of the weldment. Analysis of the cracked tubing indicated that the sulfur and phosphorus contents of the weld metal were higher than specified, the thickness was narrower at the weld, and the mechanical resistance of the weld metal was lower than specified. Cracking initiated at the weld root by coalescence of creep cavities. Propagation and expansion was aided by internal carburization. Quality control of welding procedures and filler metal was recommended.

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.

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.

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.

The curved parts of exit pigtails made of wrought Incoloy 800H tubing used in steam reforming furnaces failed by performance after a period of service shorter than that predicted by the designers. Examination of a set of tubes consisting of both curved (perforated) and straight parts revealed that the cracks initiated at the outer surface by a combined mechanism of creep and intergranular embrittlement. A smaller grain size resulting from cold bending fabrication procedures for the curved parts was responsible for accelerating the embrittlement. It was recommended that hot bending be used for fabrication of the curved parts. A change of alloy to a low-alloy chromium-molybdenum allay to protect against heat was also suggested.

A spherical carbon steel fixed-catalyst bed reactor, fabricated from French steel A42C-3S, approximately equivalent to ASTM A201 grade B, failed after 20 years of service while in a standby condition. The unit was found to contain primarily hydrogen at the time of failure. The vessel had a type 304 stainless steel shroud around the catalyst bed as protection against the overheating that was possible if the gas bypassed the bed through the refractory material. The failure was observed to have begun at the toe of the shroud-support ring weld. The ring was found to have a number of small cracks at the root of the weld. The cleavage mode of fracture was confirmed by SEM. The presence of extensive secondary cracking and twinning (Neumann bands) where the fracture followed the line of the shroud-support ring was revealed by metallography. It was revealed by refinery maintenance records that the ring had been removed for hydrotest and welded without any postweld heat treatment. The final cause of failure was concluded to be cracking that developed during the installation of the new shroud ring. Stress-relief heat treatments were recommended to be performed to reduce residual-stress levels after welding.

The accident at Three Mile Island Unit No. 2 on 28 March 1979 was the worst nuclear accident in US history. By Jan 1990, it was possible to electrochemically machine coupons from the lower head using a specially designed tool. The specimens contained the ER308L stainless steel cladding and the A533 Grade B plate material to a depth of about mid-wall. The microstructures of these specimens were compared to that of specimens cut from the Midland, Michigan reactor vessel, made from the same grade and thickness but never placed in service. These specimens were subjected to known thermal treatments between 800 and 1100 deg C for periods of 1 to 100 min. Microstructural parameters in the control specimens and in those from TMI-2 were quantified. Selective etchants were used to better discriminate desired microstructural features, particularly in the cladding. This report is a progress report on the quantification of changes in both the degree of carbide precipitation and delta ferrite content and shape in the cladding as a function of temperature and time to refine the estimates of the maximum temperatures experienced.

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.

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.