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pressurized water reactor

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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
... 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...
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
... Abstract 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...
Series: ASM Failure Analysis Case Histories
Volume: 1
Publisher: ASM International
Published: 01 December 1992
DOI: 10.31399/asm.fach.v01.c9001078
EISBN: 978-1-62708-214-3
... Abstract 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...
Series: ASM Failure Analysis Case Histories
Volume: 1
Publisher: ASM International
Published: 01 December 1992
DOI: 10.31399/asm.fach.v01.c9001056
EISBN: 978-1-62708-214-3
... Abstract Type 347 stainless steel moderator circuit branch piping in a pressurized hot water reactor was experiencing frequent leakage. Investigation of the problem involved failure analysis of leaking pipe specimens, analytical stress analysis, and determination of “leak-before-break...
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
... 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. Corrosion Decontamination Intergranular stress corrosion...
Series: ASM Failure Analysis Case Histories
Publisher: ASM International
Published: 01 June 2019
DOI: 10.31399/asm.fach.power.c0091528
EISBN: 978-1-62708-229-7
... Abstract 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...
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...
Series: ASM Failure Analysis Case Histories
Publisher: ASM International
Published: 01 June 2019
DOI: 10.31399/asm.fach.power.c9001682
EISBN: 978-1-62708-229-7
... Abstract 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...
Series: ASM Handbook
Volume: 11
Publisher: ASM International
Published: 15 January 2021
DOI: 10.31399/asm.hb.v11.a0006785
EISBN: 978-1-62708-295-2
... rivet holes and other areas of high stress when boiler water contained caustic soda (NaOH) or the boiler water chemistry was conducive to the formation of caustic soda ( Ref 4 ). In their 1926 paper, “The Cause and Prevention of Embrittlement of Boiler Plate,” researchers at the University of Illinois...
Series: ASM Failure Analysis Case Histories
Publisher: ASM International
Published: 01 June 2019
DOI: 10.31399/asm.fach.power.c9001515
EISBN: 978-1-62708-229-7
... Abstract 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...
Series: ASM Failure Analysis Case Histories
Volume: 3
Publisher: ASM International
Published: 01 December 2019
DOI: 10.31399/asm.fach.v03.c9001828
EISBN: 978-1-62708-241-9
... the pumps operate, is presented in the following section. Design Description Heat Transport Loop The heat generated in a CANDU reactor is removed by circulation of pressurized heavy water (D 2 O) through the reactor core. There are four HT pumps, each pump circulating 2,205 l/s (29,400 IGPM) of D...
Series: ASM Failure Analysis Case Histories
Volume: 2
Publisher: ASM International
Published: 01 December 1993
DOI: 10.31399/asm.fach.v02.c9001343
EISBN: 978-1-62708-215-0
.... Applications The pipe was part of the stand by system for emergency injection of cooling water to a nuclear reactor in case of a failure in the primary heat transport system Circumstances Leading to Failure Pressure testing revealed a leak during precommissioning of the system. The exact time...
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
... 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...
Series: ASM Failure Analysis Case Histories
Volume: 2
Publisher: ASM International
Published: 01 December 1993
DOI: 10.31399/asm.fach.v02.c9001349
EISBN: 978-1-62708-215-0
... during hydrotesting. Circumstances Leading to Failure The heavy water/helium storage tank was designed for a pressure of 0.1 MPa (14.5 psi) and a temperature of 67 °C (150 °F) to contain helium gas at a maximum pressure of 0.035 MPa (5 psi) and a temperature of 40 °C (105 °F), as well...
Series: ASM Failure Analysis Case Histories
Publisher: ASM International
Published: 01 June 2019
DOI: 10.31399/asm.fach.power.c9001571
EISBN: 978-1-62708-229-7
... J. M. et al. , Nuclear Technology , August 1989 , 87 , 34 – 53 . 10.13182/NT89-A27637 3. Three Mile Island Reactor Pressure Vessel Investigation Project. Achievements and Significant Results , Organization for Economic Co-Operation and Development , Paris , 1994 . 4...
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
... 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. Leakage Nonmetallic inclusions Nuclear reactor...
Series: ASM Handbook Archive
Volume: 11
Publisher: ASM International
Published: 01 January 2002
DOI: 10.31399/asm.hb.v11.a0001818
EISBN: 978-1-62708-180-1
... 1984 by the National Board of Boiler and Pressure Vessel Inspectors Initial part failure Causes Type of failures Numbers Low water cut-off Faulty design fabrication or installation Corrosion or erosion Operator error or poor maintenance Burner failure Pressure control failure Other...
Series: ASM Handbook Archive
Volume: 11
Publisher: ASM International
Published: 01 January 2002
DOI: 10.31399/asm.hb.v11.a0003553
EISBN: 978-1-62708-180-1
... is an additional example. Example 1: SCC of a Type 304 Stainless Steel Pipe Caused by Residual Welding Stresses 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...
Series: ASM Handbook
Volume: 11A
Publisher: ASM International
Published: 30 August 2021
DOI: 10.31399/asm.hb.v11A.a0006812
EISBN: 978-1-62708-329-4
... part failure Causes Type of failure Numbers Low water cut-off Faulty design fabrication or installation Corrosion or erosion Operator error or poor maintenance Burner failure Pressure control failure Other Burned or overheated Collapsed inward Combination...
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...