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Series: ASM Failure Analysis Case Histories
Publisher: ASM International
Published: 01 June 2019
DOI: 10.31399/asm.fach.power.c9001536
EISBN: 978-1-62708-229-7
... Abstract 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...
Abstract
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.
Book Chapter
Series: ASM Failure Analysis Case Histories
Publisher: ASM International
Published: 01 June 2019
DOI: 10.31399/asm.fach.power.c0048814
EISBN: 978-1-62708-229-7
.... Copper ions in solution should be eliminated or minimized. Chlorides Nuclear power generation Pressure vessels Welded joints ASME SA302-8 Stress-corrosion cracking A small leak was found in a nuclear steam-generator vessel constructed of 100-mm (4-in.) thick SA302, grade B, steel [0.25% C...
Abstract
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.
Book Chapter
Series: ASM Failure Analysis Case Histories
Publisher: ASM International
Published: 01 June 2019
DOI: 10.31399/asm.fach.power.c0090277
EISBN: 978-1-62708-229-7
... Abstract 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...
Abstract
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.
Series: ASM Failure Analysis Case Histories
Publisher: ASM International
Published: 01 June 2019
DOI: 10.31399/asm.fach.power.c9001676
EISBN: 978-1-62708-229-7
... Abstract 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...
Abstract
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.
Series: ASM Failure Analysis Case Histories
Volume: 2
Publisher: ASM International
Published: 01 December 1993
DOI: 10.31399/asm.fach.v02.c9001326
EISBN: 978-1-62708-215-0
... components Pinhole Thermocouples 316L UNS S31603 (Other, general, or unspecified) corrosion Background A batch of AISI type 316L stainless steel thermowells ( Fig. 1 ) was procured for a nuclear power plant to house thermocouples at 612 locations. Pressure testing revealed that about 20...
Abstract
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.
Series: ASM Failure Analysis Case Histories
Publisher: ASM International
Published: 01 June 2019
DOI: 10.31399/asm.fach.power.c9001146
EISBN: 978-1-62708-229-7
... a thermal shock in the pipe which, when cooled below its NDTT, cracked in a brittle manner. Low temperature Nil ductility transition temperature Nuclear power generation Piping Thermal shock ASTM A516 grade 70 UNS K02700 Brittle fracture Introduction On February 3, 1984, Georgia Power...
Abstract
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.
Series: ASM Failure Analysis Case Histories
Volume: 2
Publisher: ASM International
Published: 01 December 1993
DOI: 10.31399/asm.fach.v02.c9001327
EISBN: 978-1-62708-215-0
... nuclear power plants. The location of the large, single crack at the counterbore was similar to that observed at other PWRs. This crack and the smaller cracks exhibited crack arrests and a tight oxide film. The nature of the nozzle/reducer area of the steam generator was such that the mean stress...
Series: ASM Failure Analysis Case Histories
Volume: 1
Publisher: ASM International
Published: 01 December 1992
DOI: 10.31399/asm.fach.v01.c9001044
EISBN: 978-1-62708-214-3
... Stress-corrosion cracking C44300 UNS C44300 Stress-corrosion cracking Mixed-mode fracture Background Inhibited admiralty brass (UNS C44300) condenser tubes failed during testing. Applications The condenser was part of a repowering project in which an uncompleted nuclear power plant...
Abstract
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.
Series: ASM Failure Analysis Case Histories
Volume: 2
Publisher: ASM International
Published: 01 December 1993
DOI: 10.31399/asm.fach.v02.c9001334
EISBN: 978-1-62708-215-0
... Abstract 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...
Abstract
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.
Series: ASM Handbook
Volume: 11
Publisher: ASM International
Published: 15 January 2021
DOI: 10.31399/asm.hb.v11.a0006796
EISBN: 978-1-62708-295-2
... Power Generation Conf. , 1987 , Pwr-35 59. Keck R.G. and Griffith P. , “ Prediction and Mitigation of Erosion-Corrosive Wear in Secondary Piping Systems of Nuclear Power Plants ,” NUREG/CR-5007, 1987 60. McClintock F.A. and Argon A.S. , Mechanical Behavior...
Abstract
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.
Series: ASM Failure Analysis Case Histories
Volume: 1
Publisher: ASM International
Published: 01 December 1992
DOI: 10.31399/asm.fach.v01.c9001051
EISBN: 978-1-62708-214-3
.... Oxygen is a strong influencing factor in the general corrosion of carbon steels in nuclear reactor enviromnents. Oxygen content exerts considerable control on the incubation time of crack growth in A302 grade B material subjected to corrosion fatigue testing. These observations stemmed from a testing...
Abstract
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.
Series: ASM Failure Analysis Case Histories
Volume: 2
Publisher: ASM International
Published: 01 December 1993
DOI: 10.31399/asm.fach.v02.c9001370
EISBN: 978-1-62708-215-0
... included use of a case-hardened En 8 steel with the correct composition and regular maintenance of the pump. Cooling systems Nuclear reactors Shafts (power) Water cooling EN8 Fatigue fracture Corrosion fatigue Background After 7 years of operation, a service water pump in a nuclear...
Abstract
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.
Series: ASM Handbook Archive
Volume: 11
Publisher: ASM International
Published: 01 January 2002
DOI: 10.31399/asm.hb.v11.a0003515
EISBN: 978-1-62708-180-1
... down and assembly operations. Similar situations exist in the nuclear power industry, when the inspection interval is linked to other operations such as refueling. There is an obvious motivation to extend the inspection interval. An important objective in the design and operation of structures...
Abstract
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.
Series: ASM Failure Analysis Case Histories
Volume: 2
Publisher: ASM International
Published: 01 December 1993
DOI: 10.31399/asm.fach.v02.c9001310
EISBN: 978-1-62708-215-0
... Abstract 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...
Abstract
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.
Series: ASM Failure Analysis Case Histories
Volume: 2
Publisher: ASM International
Published: 01 December 1993
DOI: 10.31399/asm.fach.v02.c9001329
EISBN: 978-1-62708-215-0
... Leakage Nuclear reactor components Welded joints ASME SB148 C95200 UNS C95200 C95400 UNS C95400 Dealloying/selective leaching Background Leaks were found in 90 out of 782 aluminum bronze valves and fittings in an essential cooling water (ECW) system at a nuclear power plant. The leakage...
Abstract
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.
Series: ASM Failure Analysis Case Histories
Publisher: ASM International
Published: 01 June 2019
DOI: 10.31399/asm.fach.modes.c9001632
EISBN: 978-1-62708-234-1
... for a variety of experiments in nuclear physics, solid state physics, and neutron science. At LANSCE, for example, a high-power beam of 800 MeV protons at a current level of 1 mA (800 kW) can be transported in vacuum through the experimental area to a beam stop. Because the beam stop must be heavily water...
Abstract
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.
Series: ASM Failure Analysis Case Histories
Publisher: ASM International
Published: 01 June 2019
DOI: 10.31399/asm.fach.power.c0091807
EISBN: 978-1-62708-229-7
... Corrosion Cracking Failure of Admiralty Brass Condenser Tubes in a Nuclear Power Plant Cooled by Freshwater , Corros. Sci. Vol 40 ( No. 11 ), 1998 , p 1821 – 1836 10.1016/S0010-938X(98)00079-1 Selected Reference Selected Reference • Warke W. R. , Stress-Corrosion Cracking...
Abstract
Failures occurred in admiralty brass condenser tubes in a nuclear plant cooled by freshwater. About 2500 tubes had to be replaced over a span of six years. Investigation (visual inspection, chemical analysis, water chemistry (for both intake and outfall), and corrosion products in the operating system and on test coupons exposed to the operating environment) supported the conclusion that the failure was caused by microbe-initiated SCC. No recommendations were made.
Series: ASM Failure Analysis Case Histories
Publisher: ASM International
Published: 01 June 2019
DOI: 10.31399/asm.fach.design.c9001578
EISBN: 978-1-62708-233-4
... height prevented the excitation of the window, and these combined actions solved both the noise and vibration issues. Part II—Turbines, Pumps, and Compressors Case #1: High Vibration on a High Pressure Core Injection Pump at a Nuclear Plant Operators of nuclear power plants are required...
Abstract
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.
Series: ASM Failure Analysis Case Histories
Volume: 1
Publisher: ASM International
Published: 01 December 1992
DOI: 10.31399/asm.fach.v01.c9001079
EISBN: 978-1-62708-214-3
... 1570-2, EPRI Nuclear Power Division, Sept 1989 . 4. Jhansale H.R. and McCann D.R. , Bore Inspection and Life Evaluation of Vintage Steam Turbine/Generator Rotors , ASTM J. Test. Eval. , Vol 18 ( No. 6 ), Nov 1990 , p 446 – 453 . 10.1520/JTE12514J 5. Jhansale H.R...
Abstract
Numerous flaws were detected in a steam turbine rotor during a scheduled inspection and maintenance outage. A fracture-mechanics-based analysis of the flaws showed that the rotor could not be safely returned to service. Material, samples from the bore were analyzed to evaluate the actual mechanical properties and to determine the metallurgical cause of the observed indications. Samples were examined in a scanning electron microscope and subjected to chemical analysis and several mechanical property tests, including tensile, Charpy V-notch impact, and fracture toughness. The material was found to be a typical Cr-Mo-V steel, and it met the property requirements. No evidence of temper embrittlement was found. The analyses showed that the observed flaws were present in the original forging and attributed them to lack of ingot consolidation. A series of actions, including overboring of the rotor to remove indications close to the surface and revision of starting procedures, were implemented to extend the remaining life of the rotor and ensure its fitness for continued service.
Series: ASM Failure Analysis Case Histories
Publisher: ASM International
Published: 01 June 2019
DOI: 10.31399/asm.fach.power.c9001559
EISBN: 978-1-62708-229-7
... cost of all the piping was estimated at over $1,000,000, not including potential cost penalties for delayed startup of the nuclear power plant. Pipe replacement costs were particularly high (approximately $2,000/ft.) due to the necessity to erect scaffolding, extensive welding in potentially...
Abstract
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.