Search Results for Nuclear power generation

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 nuclear steam-generator vessel constructed of 100-mm thick SA302, grade B, steel was found to have a small leak. The leak originated in the circumferential closure weld joining the transition cone to the upper shell. The welds had been fabricated from the outside by the submerged arc process with a backing strip. The backing was back gouged off, and the weld was completed from the inside with E8018-C3 electrodes by the shielded metal arc process. Striations of the type normally associated with progressive or fatigue-type failures including beach marks that allowed tracing the origin of the fracture to the pits on the inner surface of the vessel were revealed. Copper deposits with zinc were revealed by EDS examination of discolorations. Pitting was revealed to have been caused by poor oxygen control in the steam generators and release of chloride into the steam generators. It was concluded by series of controlled crack-propagation-rate stress-corrosion tests that A302, grade B, steel was susceptible to transgranular stress-corrosion attack in constant extension rate testing with as low as 1 ppm chloride present. It was recommended to maintain the coolant environment low in oxygen and chloride. Copper ions in solution should be eliminated or minimized.

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.

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.

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.

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.

Inhibited admiralty brass (UNS C44300) condenser tubes used in a natural-gas-fired cogeneration plant failed during testing. Two samples, one from a leaking tube and the other from an on leaking tube, were examined. Chemical analyses were conducted on the tubes and corrosion deposits. Stress-corrosion cracking was shown to have caused the failure. The most probable corrosive was ammonia or an ammonium compound in the presence of oxygen and water. All of the tubes were replaced.

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.

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.

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.

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 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.

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.

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.

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.

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.

Vibration analysis can be used in solving both rotating and nonrotating equipment problems. This paper presents case histories that, over a span of approximately 25 years, used vibration analysis to troubleshoot a wide range of problems.

Corrosion in a closed-loop cooling water system constructed of austenitic stainless steel occurred during an extended lay up of the system with biologically contaminated water. The characteristics of the failure were those of microbiologically influenced corrosion (MIC). The corrosion occurred at welds and consisted of large subsurface void formations with pinhole penetrations of the surfaces. Corrosive attack initiated in the heat affected zones of the welds, usually immediately adjacent to fusion lines. Stepwise grinding, polishing, and etching through the affected areas revealed that voids generally grew in the wrought material by uniform general corrosion. Tunneling or worm-holing was also observed, whereby void extension occurred by initiating daughter voids probably at flaws or other inhomogeneities. Selective attack occurred within the fusion zone, i.e., within the cast two-phase structure of the weld filler itself. The result was a void wall which consisted of a rough and porous ferritic material, a consequence of preferential attack of the austenitic phase and slightly lower rate of corrosive attack of the ferrite phase. The three-dimensional spongy surface was studied optically and with the scanning electron microscope.

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.