Environmental Engineering Reference
In-Depth Information
nozzles would be supplemented with an Alloy 52M weld cover. Minor pres-
surizers have been replaced with Alloy 690 material.
Many incidences of cracks in Alloys 182 and 82 have been found in
in-service PWR power plants. In July 2000, cracks at the outlet nozzle to pipe
safe end weld of Ringhals Unit 4 were found. Many small axial cracks were
found and removed with a boat sample through electro discharge machining
(EDM). The fi rst cracking had in fact been discovered in Ringhals Unit 3 in
June of that year, but the power plant was permitted to continue to operate
without any repair because the crack seemed superfi cial and shallow in depth.
In both cases, welding used Alloy 182 and cracks were axial. The cracks in
Ringhals Unit 3 and Unit 4 were removed in 2003 and in 2004, respectively,
and the cracks were repaired by welding inlay using Alloy 52 M.
The next largest incident happened in October 2000. Penetrating cracks
and leakage were found in the V.C. Summer power plant in the same part as
in Ringhals Unit 3. Initial UT was carried out on the internal surface of the
pipe and as a result, an axial defect near the upper part of the pipe was dis-
covered. The next test was conducted in spring 2001 and many defects were
found. All of the defects were axial, and the largest defect was penetrated.
The defects were removed, and a new spool piece was welded. The part was
restored to its original condition. Alloy 52 was used for the V.C. Summer
repair from the exit nozzle to the pipe weld zone; Alloy 82 was used in some
parts for thickness and for the rest of the weld zone. The other V.C. Summer
exit nozzle was repaired by using mechanical stress improvement process
(MSIP).
7. 5 References
ASTM 1988c , Standard Guide for the In-Service Annealing of Light - Water Cooled
Nuclear Reactor Vessels Annual Book of ASTM Standard, ASTM E 509-86,
Vol.12.02, American Society for Testing and Materials, Philadelphia.
Badanin V.I. ( 1989 ), 'Application of annealing for WWER vessels life extension',
Transactions of the 10 th International Conference on Structural Mechanics
in Reactor Technology, August 1989, Atomic Energy Society of Japan, Tokyo,
pp. 129-34.
Cole N.M. and T. Friderichs ( 1991 ), Report on Annealing of the Novovorenezh
Unit-3 Reactor Vessel in the USSR, NUREG/CR-5760.
Electric Power Research Institute ( 2008 ), Materials Reliability Program:
Pressurized Water Reactor Internals Inspection and Evaluation Guidelines
(MRP-227-Rev. 0).
Hwang S.S. ( 2003 ), 'Degradation of alloy 600 steam generator tubes in operating
pressurized water reactor nuclear power plants'. Corrosion , 59 , 9, 821 .
Hwang S.S. et al . ( 2007 ), KAERI/RR-2903/2007 Failure Analysis of Retired Steam
Generator Tubings. Daejeon, Korea , KAERI .
Hwang S.S. et al . ( 2010 ), Guideline on Management of the Internals of the Nuclear
Reactor of Korean Power Plant. Daejeon, Korea , KAERI .
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