ML20236P673
| ML20236P673 | |
| Person / Time | |
|---|---|
| Site: | Trojan File:Portland General Electric icon.png |
| Issue date: | 08/07/1987 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20236P660 | List: |
| References | |
| NUDOCS 8708120463 | |
| Download: ML20236P673 (4) | |
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT N0.133 TO FACILITY OPERATING LICENSE NO. NPF-1 PORTLAND GENERAL ELECTRIC COMPANY THE CITY OF EUGENE, OREGON PACIFIC POWER AND LIGHT COMPANY
'ROJAN NUCLEAR PLANT DOCKET NO. 50-344 INTRODUCTION By letters dated February 20, and April 20, 1987, Portland General Electric Company (PGE) proposed changes to the Trojan Technical Specifications (TS)
, regarding steam generator tube plugging criteria. The amendment to the TS was proposed due to ambiguous eddy current indications of tube. degradation in the mechanical roll expanded portion of the tubes within the tube sheet in the steam generators. These steam generators were originally fabricated with a partial depth mechanical roll at the bottom of the tube. Subsequently, the full depth of the tube was expanded using a controlled explosive process.
Existing Technical Specification tube plugging criteria require all defective tubes to be plugged.
It can be shown that tube )1ugging or repair is not required in cases where there are indications in tie mechanical roll portion of the tube, and are at least a prescribed distance (F*) below the transition from the roll expansion to the explosive expansion region. The proposed amendment would preclude occupational radiation exposure that would otherwise be incurred by plant workers involved in tube plugging operations. The proposed amendment would minimize the loss of margin in the reactor coolant flow through the steam 9enerator in LOCA analyses. The proposed amendment would also avoid loss of margin in reactor coolant system (RCS) flow and therefore assist in assuring that minimum flow rates are maintained in excess of that required for operation at full power. Reduction in the amount of tube plugging required can reduce the length of plant outages and reduce the time that the steam generator is open to the containment environment during an outage.
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2 The possibility of tube repair by sleeving should not be considered a reason to exclude use of the modified tube sheet plugging limit, but should be considered one of the options used to address degradation in the region of the tube above F*. The disadvantages of tube plugs noted above also apply to some extent to sleeves. Additionally, installation of sleeves involves some impact on eddy current testing because of changes in geometry at the ends and expansions of the. sleeve and size of probe that can pass through the reduced diameter of the sleeve. The Trojan Technical Specifications & not currently contain provisions for repair of tubes by sleeving.
Steam generator tube inspections at Trojan have shown ambiguous indications in that section of the tube which has been mechanically expanded within the tube sheet. The existing tube plugging criteria apply throughout the tube length but do not take into account the reinforcing and retaining effect of the tube sheet on the external surface of the tube. Unnecessary plugging of steam generator tubes results in increased personnel exposure and reduces RCS flow.
As a remedy, Westinghouse Electric Corporation performed analysis and testing (WCAPe11307. Pro)rietary) to determine if the steam generator tube plugging criteria could >e modified for that portion of the tube which had been i mechanically expanded inside the tube sheet. The results of this analysis and testing show that if the defect in the tube is more than 0.91 inches below the bottom of the transition between the mechanically expanded section and the explosively expanded section of the tube, the tube can be left in service. To this distance, an eddy current uncertainty of 0.5 inches has been added and the result rounded to 1.4 inches, which is the F* distance. In order to avoid plugging tubes with defects below the F* distance, there can be no indications of cracking in the F* distance. Accordingly, the Trojan Technical Specifications-and Bases have been revised to modify the tube plugging limit.
Existing tube repair or plugging criteria, i.e., current applications of Regulatory Guide 1.121, do not take into account the effect of the tube sheet on the external surface of the tube.
The proposed change designates a portion of the tube for which tube degradation does not necessitate remedial action except as indicated for compliance with tube leakage limits as set forth in the Trojan Technical Specifications. As noted above, the area subject to this char.ge is in the mechanically expanded portion of the tube within the tube sheet of the steam generators. For the purpose of the evaluation of F*, the explosively expanded portion of the tube was conservatively assumed not to provide resistance to pullout or leakage.
The F* criteria do not depend on any determination of the condition of tube degradation in the portion of the tube below the F* distance.
The proposed amendment would modify TS Section 4.4.5, " Steam Generator Surveillance Requirements" and the bases for Technical Specification 3/4.4.5,
" Steam Generators", which provide tube inspection requirements and acceptance ;
criteria to determine the level of degradation for which the tube may remain i in service. The proposed amendment revises the definition for tube plugging limit to prescribe the portion of the tube subject to the F' criteria.
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DISCUSSION AND EVALUATION I l
The supporting technical and safety evaluation of the subject criteria (Westinghouse WCAP-11307 "Tubesheet Region Plugging Criterion for the Enrtland General Electric Company Trojan Nuclear Station" (Proprietary), and WCAP-11315 (Non-Proprietary) demonstrate that the presence of the tube sheet will enhance i the tube integrity in the region of the hardroll by precluding tube I deformation beyond its initial expanded outside diameter. The resistance to presence of the both tube rupture and tube collapse is strengthened by the tube sheet in that region. The result of the hardroll of the tube into the tube sheet is an interference fit between the tube and the tube sheet. Tube rupture can not occur because the contact between the tube and tube sheet does not permit suf ficient movement of the tube material. In a similar manner, the tube sheet does not permit sufficient movement of tube material to permit buckling collapse of the tube during postulated LOCA loadings.
Additionally through analysis and testing, Westinghouse has demonstrated that the roll expension over the F* distance is sufficient as long as it contains no indications of cracking to preclude pullout of the tube from the tube sheet. Even with the conservative assumption that a tube could completely sever circumferentially below the F* distance, test results demonstrate that pullout of the tube is precluded under normal and postulated accident condition loadings.
Relative to expected leakage, the length of the roll expansion in the F*
distance is sufficient as long as it contains no indications of cracking to preclude significant leakage from tube degradation located below the F*
distance. The existing Technical Specification leakage rate requirements and accident analysis assumptions remain unchanged in the unlikely event that significant leakage from this region does occur. As noted above, tube rupture and pullout is not expected for tubes using the F* criteria. Any leakage out of the tube from within the tube sheet at any elevation in the tube sheet is fully bounded by the existing steam generator tube rupture analysis included in the Updated Final Safety Analysis Report. The proposed F*
criteria do not adversely impact any other previously evaluated design basis accident.
The use of the F* criteria has been demonstrated to maintain the integrity of the tube bundle commensurate with the requirements of Regulatory Guide 1.121 for indications in the free span of tubes and the primary to secondary pressure boundary under normal and postulated accident conditions. Acceptable tube degradation is any degradation in the tube below the F* distance. The safety factors used in the detennination of the F* distance are consistent with the safety factors in the ASME Boiler and hessure Vessel Code used in steam generator design. The F* distance has been verified by testing to be greater than the length of roll expansion required to preclude significant leakage during nonnal and postulated accident conditions. Additionally, for axial or nearly axial indications in the tube sheet region, the tube end remains structurally intact further decreasing any potential for tube pullout.
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For tubes with axial or nearly axial cracks, the strength of the tube relative to an axial load would not be reduced below the strength required to resist potential axial loads. In this case leakage is the dominant consideration to determine the necessity of tube plugging or repairing. Again, based on testing, using the F* criteria would not be expected to result in significant leakage from through-wall cracks located below the F* distance.
It should be noted that the staff has previously issued similar amendments for the Virgil C. Summer Nuclear Station Unit I and McGuire Nuclear Station Units 1 and 2. These amendments were for full-depth hard-rolled tubes whereas this amendment request only involves approximately the bottom 3 inches of the tubes. Thus, there is much more of each tube in the tube sheet to resist tube pullout and to inhibit primary-to-secondary leakage.
Based on the above, we find that the proposed change meets the margins and the intent of Regulatory Guide 1.121, and is therefore, acceptable.
CONTACT WITH STATE OFFICIAL The NRC staff has notified the Oregon Department of Energy of the proposed issuance of this amendment along with the proposed determination of no significant hazards consideration. No comments were received.
ENVIRONMENTAL CONSIDERATION This amendment invcives a change in the installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20.
The staff has 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 offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously published a proposed finding that the 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 651.22(c)(9). Pursuant to 10 CFR 551.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.
CONCLUSION We have concluded, based on the considerations discussed above, that (1) there is reasonable 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 Ccmmission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and secu-rity or to the health and safety of the public.
PRINCIPAL CONTRIBUTOR:
C. Sellers Dated: August 7, 1987
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