ML20031F157
| ML20031F157 | |
| Person / Time | |
|---|---|
| Issue date: | 09/07/1977 |
| From: | Case E Office of Nuclear Reactor Regulation |
| To: | Levine S Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML13319A640 | List:
|
| References | |
| FOIA-81-313 NUDOCS 8110190277 | |
| Download: ML20031F157 (4) | |
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f SEP 1 1977 MEMOR/.NDUM FOR:
S. l.evine, Director Office of Nuclear Regulatory Research FROM:
E. G. Case, Acting Director Office of Nuclear Reactor Regulation SUBECT:
REQUEST FOR CONFIRMATORY RESEARCH FOR THE EXPERIMENTAL DETERMINATION OF SUSCEPTIBILITY OF STRESS CORROSION CRACP"lG 0F INCONEL-600 TUBING (RR-NRR-77-17)
The purpose of this memo is to provide background information related to the steam generator tube denting phenomenon which has been observed in operating PWR facilities and to request the initiation of'a confirmatory research program to determine the factors affecting the susceptability of stress corrosion cracking of Incone'. 600 steam generator *.bing.
Description of Problem The phenomenon of PWR steam generator tube " dent:ng" (reduction in tube diameter at tube / tube support plate locations) was first discovered in Westinghouse designed plants in late 1975.
In 1976, several of these plants started to experience primary-to-secondary leakage at tube / tube suppor t plate intersections where tube denting was occurring.
Early indications were that the tube denting phenomenon was restricted to plants whose steam generators had been operating for a significant period of time with phosphate secondary chemistry control prior to conversion to all volatile treatment (AVT). The denting phenomenon occurs as a result of the growth of magnetite at the tube / tube support plate annulus; such growth has been attributed to residual phosphates that were not removed from the annulus between the tubes and the tube support plate when steam generator secondary coolant chemistry control was switched from phosphate treatment to AVT. Contamination of the secondary coolant by in-leakage of condenser cooling water is believeo to have caused a catalytic reaction with residual phosphates to corrode the carbon steel support plates. The continued corrosion of the carbon steel support plate exerts sufficient forces to dent tne tube and/or crack the tube support plate ligaments between the tube holes and the water circulation flow holes.
In Westinghouse designed steam generators, "hourglassing" of the rectangular flow slots from the corrosion product expansion forces has also been observed.
Subsequent to these early findings denting has also been detected in Westinghouse plants with limited previous exposure to phosphate chemistry; and in May of 1977 two Combusion Engineering plants operating exclusively on AVT were found to have dented tubes when subject to an eddy-current inservice intFaction.
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SEP i 1977 In its extreme, denting lead to tube failures (e.g., leaks). Leaks have been associated with dented tubes found in Turkey Point 4 and' Surry 1 and 2.
These leaks have either been due to small longitudinal cracks masked within the tube / tube support plate annulus or due to axial cracks in tight-bend U-tubes, at the apex of the bend due to flow-slot "hourglassing" and inward movement of the legs of the tubes.
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In both cases the leaks originated on the primary side (inside surface of the tubes) due to stress corrosion cracking at locations where high hoop strains had developed.
While the probability and consequence of steam generator tube failures are being investigated analytically and the structure integrity of dented tubes is being investigated experimentally, there are no programs presently underway addressing those conditions necessary for the occurrence of stress corrosion cracking of Inconel 600 steam generator tubes.
Confirmatory Research Program Elements Recent experience at operating PWR facilities has indicated a continuing incidence of cracks in Inconel 600 steam generator tubing as a result of nrimary water stress corrosion. Although these cracks are known to result from pr.imary water stress corrosion, a need exists to develop an understanding of those conditions necessary for such cracking to occur. This includes the identification of the parameters that can affect the time period required to develop such cracks and the functional relationship between these parameters.
Therefore, it is recommended that an experimental program be initiated to obtain a better understanding of basic mechanisms and factors operative in the stress corrcsion cracking of Inconel 600 steam generator tubing. This program should include, but not be limited to, a study of those parameters which affect the susceptibility of Inconel 600 tubing to stress corrosion cracking, such as:
environment (temperature; water chemistry) stresses or strains; residual and/or load stresses strain rate, period of time for crack initiation material chemistry (varying carbon levels) material condition (cold worked, annealed, stress-relievedetc.)
The proq: am should be designed such that its results could be extrapolated to predlet the future performance of dented steam generator tubes.
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. SEP 7 1977 Licensing Impact Due to the number of PWR facilities which may be susceptible to the steam generator denting phenomenon and the fact that the extent of tube denting in those facilities already affected is increasing, the results of this proposed program would have direct application in the resolution of NRR safety concerns.
It is anticipated that the results of this proposed experimental study would provide a basis for determining the expected life before failure.(i.e., cracking) of tubes subjected to continued denting.
The experimental data on the expected life of dented tubes, in conjunction with data on the rate of denting obtained from steam generator inservice inspections could then be used to establish inservice '
inspection frequency requirements and preventive plugging criteria for plants that have experienced tube denting. The inservice inspection requirements and tube plugging criteria would be incorporated into existing Regulatory Guides, principally 1.83 and 1.121.
Magnitude of_Research Effort NRR estimates that the confirmatory research program discussed above would cost approximately 600K and would encompass a time period of approximately 3 years.
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E. G. Case, Acting Director Office of Nuclear Reactor Regulation cc: See next page.
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