ML20043D160
| ML20043D160 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 05/29/1990 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20043D156 | List: |
| References | |
| NUDOCS 9006070189 | |
| Download: ML20043D160 (5) | |
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'~g UNITED STATES 8'
NUCLEAR REGULATORY COMMISSION J
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t WA6HINGTON, D. C. 30666 i
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION I
RELATED TO AMENDMENT N0._153 TO FACILITY OPERATING LICENSE NO. DPR-50
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METROPOLITAN ED150N COMPANY i
OERSEY CENTRA. POWER 5
.IGHT COMPANY MfiSYLVAld A
- . ECTR C COMFANY GPU, NUCLEAL CORPURATION THREE MILE ISLAND _ NUCLEAR STATION. tfMIT NO.1 DOCKET NO. 50-289 t
5 INTRODUCTION:
By letter of March 12, 1990, GPU Nuclear Corporation (GPL'N) requested a change in the Three Mile Island Nuclear Station relating to steam generator surveillance., Unit 1 Technical Specifications The chenge was requested to modify the TM1-1 Technical Specifications for unscheduled steam generator tube inspection requiremehts after a primary-to-secondary leak in excess cf the limits of Specification.
The proposed change specifies that:
(1) when a leaking tube is located in Group A-1 (" lane wedge" area) all tubes in this group in only the affected steam gerierator need be inspected (current Technical Specifications are not i
explicit in this regard) to include these portions of the tubes where the leak was found, and if the results of the inspection fall into the C-3 Category, additional inspections will be performed in the same group in the other steam generator; anc (2) when the leaking tube is not in Group A-1, an inspection will be performed on the affected steam generator in accordance with the Technical Specification.
BACKGROUND On March 6 1990 at 0912, TMI-1 began a plant shutdown because of a primary to secondary leak in the once through steam generator (OTSG) which occurred shortly after a refueling outage. Following cooldown the A OTSG was opened and a bubble test oerformed on March 8.
Thetestidentifiedtube1inrow77 (designated A77-1) as the leaking tube. This tube is in the " lane wedge" region of the OTSG and had been Eddy Current examined in January 1990 as part of the BR refueling inservice inspection program. The 8R inspection identified no recordable indication of degradation on tube A77-1. Post-leak Eddy Current inspection performed on fiarch 9 identified that A77-1 had a through wall defect at the point where the tube exits the bottom of the upper tube sheet.
9006070189 900529 PDR ADOCK 05000289 7
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The failure of the tube was identified as a circumferentially oriented 360" i
crack.
Thisisbelievedtobetheresultofenvironmentally(assistedhigh cycle fatigue (HCF). This is based on eddy current testing ECT) data, visual examination of the tube experience with HCF crac, king of OTSG tubes.and on a comparison of the failure wit The determination that " lane wedge" area tubes are susceptible to this failure mechanism is based on four 4
tube samples from the Oconee Nuclear Station removed and analyzed during the period from 1976 to 1982.
The ECT performed for the inservice inspection program on tube A77-1 during t
January 1990 yielded no recordable indications of degradation. B&W industry experience confirms that this type of failure occurs rapidly and therefore evidence of the condition may exist only shortly before leakage would be experienced. Tube inspection techniques do not effectively identify HCF precursorconditionsunlesstheyareperformedjustprior(e.g. hours)before tube failure. Mitigating actions in response to tube leakage are provided by plant normal and emergency procedures.
GPUN completed an inspection of all tubes in the " lane wedge" area of the A OTSG. No new imperfections of these tubes were identified which differed from the prict outage inspection results.
It should be noted that one tube defect was found during examination of the " lane wedge" area following the tube leak.
The defect was a shallow inside diameter pit at 41% through wall based on a less than 1 volt one coil indication on a 8x1 absolute ECT probe. A review.of the ECT data confirmed that this defect existed during the last inspection.
According to the licensee, it was not previously judged as defective because of i
its very low signal level and shallow phase angle, L
GPUN considered the tube failure to be caused by HCF, an industry identified l
problem. Since additional eddy current inspection in the " lane wedge" area of the A OTSG had essentially duplicated the resuits of inspections performed during the previous outage inservice inspection, it was unnecessary to expand the present ECT beyond the " lane wedge.'
A Waiver of Compliance was issued by the NRC on March 14, 1990, that permitted resumption of plant operation without completing Technical Specification required random ECT tube examinations of the affected 0TSG. Rather, a focused ECT-program was performed which inspected all unplugged " lane wedge" area tubes failur,e mechanism.which OTSG industry experience has demonstrated are prone to the HC This ECT was performed down to the 14th support plate which includes all of the HCF failure-prone tube portions. This inspection resulted in the two indications noted above. A drip test of the entire OTSG was performed to provide additional confidence in the integrity of the tubes, and showed no pro)1 ems. A post-repair bubble test was also performed.
EVALUATION TMI-1 Technical Specification Section 4.19.3.c.1 currently specifies that additional unscheduled inservice inspections shall be performed on each steam generator during shutdown following a primary-to-secondary tube leak (not including leaks originating from tube-to-tube sheet welds) in excess of the limits of the Technical Specification.
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The proposed change to limit the unscheduled inservice inspection to the leaking steam generator following primary-to-secondary les cage through the i
steam gener; tor tubes which exceeded Technical Specification limits will reduce personnel radiation exposure associated with the inspections without compromising the objective of these inspections.
If the leaking tube is located in the " lane wedge" area and the results of the unscheduled inspection of the affected steam generator fall into the C-3 category, additional inspections will be perforined in the same tube group in the other steam generator.
If the leaking tube is not located in the ' lane wedge" area the unscheduled inspection will be performed on the affected steam generator only, in accordance with existing Technical Specification.
OTSG industry experience has shown that the ' lane wedge" area has been experiencing corrosion assisted fatigue and fretting wear. This area is more susceptible to damage due to the proximity to the open lane which allows higher moisture carryover and highest cross flow since the steam changes direction from vertical to horizontal to exit the steam generators. Performing Technical Specification limited tube inspection in the area where leaks are found will identify potential additional tubes which may be experiencing similar degradation and enabling appropriate corrective action to be taken to prevent further tube leakage.
The licensee has stated that this approach is consistent with EPRI recomendations in the PWR Inspection Guidelines, which means' that the inspections would normally be performed with a 8x1 coil.
SUMMARY
Industry experience indicates that the failure of tubes in the area of the I
failed tube are due to HCF. This mechanism is a rapid failure mechanism ead precursors may exist only briefly before failure occurs. There is no method to predict failures with such a rapid development. Leakage monitoring may be one effective means of detection for slowly developing leaks.
Repeating random ECT of the A OTSG outside the " lane wedge
- area would provide no ad/.itional l
technical information relevant to this failure mechanism. We therr. fore conclude, l
based on the above, that the proposed Technical Specification change should be granted.
STAFF RECOMMENDATIONS In view of the fact that other B&W plants that have experienced fatigue failures at the upper tube sheet have implemented preventive tube sleeving programs, we recomend that GPUN consider such a program unless it can be shown that the probability of additional fatigue cracks leading to a steam generator tube rupture is very low. However, we believe that probably the only way this can be shown is to demonstrate that augmented leak detection and leak rate monitoring methods can detect incipient leaks due to fatigue crack initiation before the rapidly propagating crack has grown around the circumference of the tube leading to a potential tube rupture.
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4 ENVIRONMENTAL CONSIDERATION Tha amendment changes a requirement with respect to installation or use of a facti'ty component located within the restricted areas as defined in 10 CFR Part 20 and changes surveillance requirements. We have determined that the amendment involves no significant increase in the amounts and no significant change in the types of any effluents that may be released off site, and that there s; no significant increase in individual or cumulative occupational radiation exposure. The staff has previously issued a proposed finding that this amendment involves no significant hazards consideration, and there has been no public comment on such finding. Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).
Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental i
assessment need be prepared in connection with the issuance of this amendment.
CONCLUSION Wehaveconcluded,basedontheconsiderationsdiscussedabove,that(1)there is reaso: able assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public.
Principal Contributor: Herb Conrad Dated: May 29,1990 t
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' DATED:; May 29, 1990 TMI-1 AMENDMENT' NO.153 TO FACILITY OPERATING LICENSE NO.DPR-50 DISTRIBUTION fig Plant File S.Varga(14E4)
B.Boger(14A2)
.J. Stolz S. Norris R. Hernan OGC D.Hagan.(MNBB3302))
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E. Jordan (MNBB 3302 G..H111(4 (P1-137)
W. Jones P-130A)
J. Calvo 11F23)
HerbConrad(9H15)
ACRS(10)
GPA/PA ARM /LFMB E. Wenzinger, R1 cc: Licensee / Applicant Service List i
is.
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