ML19312C422
| ML19312C422 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 10/04/1977 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19312C411 | List: |
| References | |
| NUDOCS 7912130944 | |
| Download: ML19312C422 (6) | |
Text
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%,s UNITE 3 STATES NUCLEAR RECULATORY COMMIS$10N f
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CASHINGTON, C. C. 20666 O,,
f SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION DUKE POWER COMPANY OCONEE NUCLEAR STATION, UNITS 1, 2 & 3 DOCKET N,S. 50-269, 50-270 AND 50-287 Introduction By letter dated September 24, 1977, the licensee informed the NRC of the discovery of an apparently different steam generator tube degradation pheiomenon at Oconee Nuclear Station Unit 1.
This phenomenon was discovered during the steam generator tube inservice inspection completed in September 1977.
Results of this inspection were not included in Duke Power's Safety Assessment Report, submitted in August 1977, which is currently under review. The new degradation phenomenon is described as a localized erosion or cavitation mechanism resulting in a tube wall thinning.
This thinning has been detected by eddy current measurements and a total of eighty-nine tubes in Oconee Unit 1 steam generators "l A" and "18" have been affected by this phenomenon.
Discussion Inspection Results The eddy current inspection initially performed during this outage in each steam generator included all of the tubes adjacent to the open tube lane, a 3.0 percent sample randomly selected throughout the steam aenerator and a 2.5 percent sample randomly selected within the peripheral,egion of the steam generator. As a result of this inspection five defective tubes (tubes with eddy current indications greater than 40% wall thinning) were identified in the "lB" steam generator and three defective tubes were identified in the "l A" steam generator. These tubes were located in the peripheral region away from the open tube lane and were predominately at the 14th support plate elevation.
Based on these results a second tube sample consisting of 3 percent of the tubes in each steam generator was inspected.
l This second sample was concentrated in the areas around the defects and in l
the peripheral region.
The results of this sample inspection revealed one i
additional defective tube in the "lA" steam generator and 10 additional i
defective tubes in the "18" steam generator.
For the "lB" steam generator the majority of the additional defects were located in the half of the steam generator 6;sosite the open tube lane, l
consisting of quadrants WX and XY as i% hit ad in the licensee's submittal.
7912130
- The steam outlet lines leave the steam generator shell from these quadrants.
A third 6 percent sample was examined in that region of the "lB" ste,m generator which detected 10 more defective tubes.
In an effort to validate that the problem regions of the steam generator had been identified, a third 6 percent sample was inspected in the periphery of the WX-XY quadrants in steam generator "lA" and a fourth 6 percent sample was examined in the periphery on the open tube lane side of steam generator "lB".
These samples revealed one defective tube in steam generator "lA" and two additional defective tubes in steam generator "lB".
Due to the large number of defective eddy current indications, some of which were interpreted as 90% - 100% wall thinning, the licensee considered it essential to obtain tube samples for direct examination to help identify the degradation phenomenon.
Therefore, the licensee removed two peripheral tubes from steam generator "lB" for visual and laboratory examination.
The first tube removed had an eddy current indication just above the 14th support plate which interpreted as 45% - 50% wall thinning. Visual inspection of this tube revealed an eroded slot area about 1/8 inch long and 1/16 inch wide and approximately.020 inch deep.
The second tube removed had an eddy current indication interpreted to be 90% wall thinning just above the 14th support plate.
Visual inspection of this tube revealed a shallower erosion wear spot covering substantially more tube surface area.
This spot was about 1 inch long and 0.3 inches wide.
As a result of these observations, the licensee became aware of a form of tube degradation that is of a different nature than that previously observed in Oconee steam generators.
Ir. view of this, the licensee conducted eddy current testing on an additional 11% of the tubes in steam generator "lB".
This 11% sample included all of the tubes in the periphery of quadrants WX and XY and one-third of the periphery tubes in the open tube lane side of the steam generator.
This inspection included examination of four tubes with 14th tube support plate ir.dications which had been eddy current tested four months previmisly.
Three of these four tubes showed no change in degradation size while for the fourth tube the eddy current test indicated that tube thinning had gone from less than 20 percent to 35 percent of wall thickness.
In total, the licensee has eddy current inspected 33% of the tubes in steam generator "lB" and 16% of the tubes in steam generator "lA".
Of the approximately 7,350 tubes thus inspected in both steam generators, 32 were classified as defective tubes and 44 degraded (greater than 20% and less than 40% wall thinning) tubes were found in steam generator "18"; in steam generator "1A" the totals were 5 defective tubes and 8 degra ded.
All of the 37 tubes identified as defective were plugged.
i I
, Burst Test Data The licensee's submittal also included the results regarding B&W tube rupture test.
B&W has demonstrated that a tube with a flat defect 70 percent through the wall will not fail under 5,000 PSI internal pressure.
This is greater than twice the pressure which would occur during a postulated main steam line break accident.
Evaluation The earlier problem of fatigue related circumferential cracking in tubes along the missing tube lane at the upper support plate locations has been carefully followed by us for some time. The leaks experienced so far have been quite small and have entailed orderly planned shutdown to investigate the leak and to remove the leaking tube from service by plugging.
The licensee has recently, at our request, provided an extensive safety assessment of the effect of such leaks on reactor functions as part of a proposed revision of technical specifications relating to steam generator inspection aid integrity protection.
This is currently under review by us.
We have reviewed the information submitted by the licensee on September 24 regarding the newly identified steam generator tube degradation phenomenon.
The licensee has completed a comprehensive eddy current examination of both Oconee Unit 1 steam generators.
This eddy current inspection program began with a broad sampling plan which was expanded when the additional degradation phenomenon was identified and continued until the problem areas in the steam generators were identified and thoroughly inspected.
Based on our evaluation we conclude that the eddy current testing program conducted has been sufficiently extensive to identify areas of the steam generator where there is a high. probability of tubes being affected by this cavitation or erosion mechanism.
Furthermore, the 100% eddy current examination perfonned in the high probability areas (periphery of quadrants WX-XY) in steam generator "lB" and additional eddy current sempling in lower probability areas in steam generator "lB" and steam generator "l A" provide sufficient confidence that the defective tubes hcVe been identified.
The NRC has also reviewed photographic and measurements results of the visual examinations of the two tubes removed from steam generator "lB".
The photographs of the tube defect locations give the appearance of a cavitation and erosion phenomenon consistent with the mechanism suggested by Babcock and Wilcox.
Due to the nature of this phenomenon and experience in non-nuclear cases of this type of tube erosion a high degradation rata is not expected.
This view is reinforced by examining the data collected on the four tubes that previously had eddy current indications when inspected
e 4-in May and were reinspected during this latest inspection.
Three of these four tubes indicated no further degradation rate while the fourth showed an approximate 15% degradation increase in four months.
The effect of this type of degradation on steam generator tubes will be very similar to the wastage type of degradation previously observed in recirculating type of steam generators.
Experience has shown that this type of degradation has not lead to catastrophic tube failure but rather has resulted in a leak before break situation. This type of gradual degradation is predictable and rc~.lts, at worst, in an orderly plant shutdown with no danger to the publi. nealth and safety.
Moreover, from experience with wastage corrosion we would expect that a detectable leak will penetrate the tube wall before general tube wall thinning reaches a level at which the tube would be incapable of with-standing loads imposed by the full range of normal operating and accident conditions. As a result we believe that it is quite unlikely that a significant number of tubes (5 to 10) could reach a level of thinning at which they would fail in the event of an MSLB or LOCA, without prior detection by leakage in at least one tube.
Because of the importance of leakage detection in assuring steam generator integrity, particularly since we appear to be dealing with a new phenomenon for which rate data is not well developed, it is very important for the licensee to continue its program of rapid repair of any detectable leakage and consultation with the NRC staff in the event such leakage is detected.
This enables the staff, on the basis of the information available, to determine whether any additional investigation is required.
Based on our previous experience with similar forms of tube degradation, B&W's burst test data, and on the continuation of licensee's program of leakage detection and staff consultation, we believe that there is reasonable assurance that there will be no significant reduction in overall steam generator integrity resulting from the new erosion phenomenon in the period until the next inspection. Nevertheless, careful investigation of this dedradation phenomenon should continue in order to expand the preliminary data reported above.
In this respect the NRC has requested and the licensee has comrnitted to the following:
m
' i 1.
Information will be provided in a subsequent status report on the metallurgical examination conducted on removed tubes 43/108 and 83/117. This information is expected to be available by Decenber 15, 1977.
2.
Evaluations will be perfonned to evaluate a plugging limit criteria for defective tuber 3.
An attempt will be made to develop an inservice inspection calibration standard which will permit a more realistic, less conservative evaluation of large-area, shallow defects.
4.
An attempt will be made to detennine the rate of growth, if any, of indications at the 14th support plate at future Oconee 1 outages.
5.
At the next Oconee 1 outag:, additional peripheral tubes will be examined consistent with critical path scheduling.
6.
Technical Specifications concerning inservice inspection of steam generator tubing will be reevaluated, and resubmitted if necessary, to incorporate the most recent experience.
7.
Information will be provided in the near future concerning the visual examination of previously leaking, stabilized tube 114/109.
We conclude that the efforts represented by these consnitments constitute a responsible and necessary approach to further understanding this steam generator tube degradation phenomenon and assessing its long term significance for the safe operation of these steam generators.
Until the licensee completes the above investigation, we are adding the following Technical Specification to Unit 1 If at any time, the leakage through the Oconee Unit 1 steam generator tubes equals or exceeds 0.3 gpm, a reactor shutdown shall be initiated within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> and the reactor shall be in a cold condition within the next 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.
If the leakage is less than 0.3 gpm, an assessmenFshall bdiade wheth'er operations may be continued safely '
or the plant should be shutdown.
In either case, the NRC shall be notified in accordance with Section 6.6.2.1.
The 0.3 gpm is consistent with the limits imposed on other facilities with tube degradation problems.
A 6-Based on the above evaluation and comitments we conclude that Oconee Nuclear Station, Unit 1 is safe for continued operations.
Progress in the licensee's investigation of this tube degradation phenomenon as well as performance of the Oconee Nuclear Station steam generators will continue to be under close observation by the NRC staff, and appropriate actions will be taken in the event of any unexpected developments not bounded by the above evaluation.
For this reason we have concluded that,with respect to newly identified tube erosion phenomenon, continued operation of the facility can be authorized under the conditions discussed above without significant reduction in overall steam generator safety margin.
Environmental Consideration We have determined that these amendments do not authorize a change in effluent types or total amounts nor an increase in power level and will not result in any significant environmental impact.
Having made this determination, we have further concluded that these amendments involve an action which is insignificant from the standpoint of environmental impact and pursuant to 10 CFR 551.5(d)(4) that an environmental impact statement, negative declaratinn, or environmental impact appraisal need not be prepared in connection with the issuance of these amendments.
Conclusion We have concluded, based on the considerations discussed above, that:
(1) because the amendments do not involve a significant increase in the probability or consequences of accidents previously considered and do not involve a significant decrease in a safety margin, the amendments do not involve a significant hazards consideration, (2) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (3) such activities will be conducted in compliance with the Comission's regulations and the issuance of these amendments will not be inimical to the comon defense and secu:'ity or to the health and safety of the public.
Da te:
October 4,1977
UNITED STNTES NUCLEAR REGULATORY COPMISS.JN DOCKET NOS. 50-269, 50-270 AND 50-287 DUKE POWER COMPANY NOTICE OF ISSUANCE OF AMENDMENTS TO FACILITY OPERATING LICENSES The U.S. Nuclear Regulatory Commission (the Commission) has issued Amendment Nos. 47
, 47 and 44 to Facility Operating License Nos.
DPR-38, OPR-47 and DPR-55, respectively, issued to Duke Power Company which revised the Technical Specifications for operation of the Oconee Nuclear Station, Unit Nos.1, 2 and 3, located in Oconee County, South Carolina.
The amenanents are effective as of the date of issuance.
These amendments revise the Technical Specifications to establish operating limits for Unit 1 cycle 4 operation and tighten leakage limits through the Steam Generator tubes.
The application for the amendments complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act),
and the Commission's rules and regulations.
The Commission has made appropriate findings as required by the Act and the Commission's rules and regulations in 10 CFR Chapter I, which are set forth in the license anendnents.
Prior public notice of these amendments was not required since the amendments do not involve a significant hazards consideration.
The Coanission has determined that the issuance of these amendments will not result in any significant environmen'tal impact and that pursuant to 10 CFR 551.5(d)(4) an environmental impact statement or negative declaration and environmental impact appraisal need not be prepared in ccnnection with tho issuance of these amendments.
l 7 9 n29 om i
For further details with respect to this action, see (1) the application for amendments dated March 30, 1977, as supplemented June 21, August 23, September 8,14 and 24,1977, (2) Amendment Nos.47,
47 and 44 to License Nos. DPR-38, DPR-47, and DPR-55, respectively, and (3) the Comission's related Safety Evaluation.
All of these items are available for public inspection at the Comission's Public Document Room,1717 H Street, NW., Washington, D.C. and at the Oconee County Library, 201 South Spring Street, Walhalla, South Carolina 29691.
A copy of items (2) and (3) may be obtained upon request addressed to the U.S. Nuclear Regulatory Commission, Washington, D.C.
20555, Attention:
Director, Division of Operating Reactors.
Dated at Bethesda, Maryland, this 4th day of October 1977.
FOR THE NUCLEAR REGULATORY CCMMISSION l
ha%W62 's -
A. $chwencer, Chief, Operating Reactors Branch fl Division of Operating Reactors t
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