Search Results for Nuclear reactor components, corrosion

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.

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.

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.

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.

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 rupture of a thirty-year-old U-tube on a steam generator for a closed-cycle pressurized-water nuclear power plant occurred, resulting in limited release of reactor water. A typical tube bundle can be over 9 m (30 ft) tall and 3 m (10 ft) in diam with over 3,000 22-mm (7/8-in.) diam Inconel Alloy 600 tubes. Tube support plates (TSP) separate the tubes and allow flow of the heating water/steam. Inconel Alloy 600 is susceptible to intergranular stress-corrosion cracking over time, so investigation included review of operational records, maintenance history, and procedures. It also included FEA (thermal gradients, nonlinear material behavior, residual stress, changes in wall thickness during the formation of U-bends, and TSP distortions near the ruptured tube) of three-dimensional solid models of the U-tubes. The conclusion was that distortion of the TSPs and resulting “pinching” of the U-tubes, combined with the operational stresses, caused high stresses at the location where the tube cracked. The stresses were consistent with those required to initiate and propagate a longitudinal crack.

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 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.

Aluminum-clad spent nuclear fuel is stored in water filled basins at the Savannah River Site awaiting processing or other disposition. After more than 35 years of service underwater, the aluminum storage racks that position the fuel bundles in the basin were replaced. During the removal of the racks from the basin, a failure occurred in one of the racks and the Savannah River Technology Center was asked to investigate. This paper presents the results of the failure analysis and provides a discussion of the effects of corrosion on the structural integrity of the storage racks.

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.

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.

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.

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.

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.

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.

One inch diam Type 304 stainless steel piping was designed to carry containment atmosphere samples to an analyzer to monitor hydrogen and oxygen levels during operational and the design basis accident conditions that are postulated to occur in a boiling water reactor. Only one of six lines in the system had thru-wall cracks. Shallow incipient cracks were detected at the lowest elevations of one other line. The balance of the system had no signs of SCC attack. Chlorides and corrosion deposits in varying amounts were found throughout the system. The failure mechanism was transgranular, chloride, stress-corrosion cracking. Replacement decisions were based on the presence of SCC attack or heavy corrosion deposits indicative of extended exposure time to chloride-contaminated water. The existing uncracked pipe, about 75 percent of the piping in the system, was retained despite the presence of low level surface chlorides. Controls were implemented to insure that temperatures are kept below 150 deg F, or, walls of the pipe are moisture-free or the cumulative wetted period will never exceed 30 h.

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.

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.