ML20209D003

Supplemental Application for Amend to License NPF-1, Incorporating License Change Application 152 Re Unneccessary Plugging of Steam Generator Tubes Which Increases Personnel Exposure & Reduces RCS Flow

ML20209D003
Person / Time
Site:	Trojan File:Portland General Electric icon.png
Issue date:	04/20/1987
From:	Cockfield D PORTLAND GENERAL ELECTRIC CO.
To:	NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM)
Shared Package
ML20209D009	List: ML20209D003 ML20209D025
References
TAC-64756, NUDOCS 8704290145
Download: ML20209D003 (7)

Text

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m e Pbrilar1d Gerleral M Coiissaisy David W. Cockfield Vice President, Nuclear April 20, 1987 Trojan Nuclear Plant Docket 50-344 License NPF-1 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington DC 20555

Dear Sir:

License Change Application 152 Attached are revised pages to License Change Application 152. The revision is consistent with the changes agreed to by Portland Cenoral Electric Company in meetings and telephone conversations with the Nuclear Regulctory Commission. The changes are denoted by side bars.

Sincerely, Attachments c: Mr. John B. Martin Regional Administrator, Region V U.S. Nuclear Regulatory Commission Mr. R. C. Barr Resident Inspector Trojan Nuclear Plant 1

Mr. David Kish, Director State of Oregon Department of Energy 0 0\

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121 S.W Sa! mon Street, Portand. Oreoon 97204

LCA 152 Page 1 of 6 LICENSE CHANGE APPLICATION 152 DESCRIPTION OF CHANGE The proposed replacement pages to Appendix A of Facility Operating License NPF-1 are provided as Attachment 1. A description of the changes to the existing Technical Specification is as follows:

Page 3/4 4 Changes to Section 4.4.5.2 were made to add any tubes which have defects below the F* distance but which have not been plugged to the 3 percent sample to be inspected.

Page 3/4 4 Changes to Section 4.4.5.4.a.6 changes the plugging limit to exclude imperfections in the area of the tube sheet region below the F* distance provided the tube has no indications of cracking within the F* distance.

Page 3/4 4-9a - Changes to Section 4.4.5.4.a add definitions of tube roll expansion, and F* distance. Section 4.4.5.4.b was revised to delete the requirement to plug all tubes with through-wall cracks.

Changes to Section 4.4.5.5 add a requirement to report the results of inspections of tubes which have defects below the F* distance but which were not plugged to the NRC prior to restart of the unit fol-lowing the inspection. Conforming changes were made to the Bases for Trojan Technical Specification 3/4.4.5.

REASON FOR CHANGE Steam generator tube inspections at Trojan have shown ambiguous indica-tions 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, reduces RCS flow, increases moisture carryover in the steam lines and shortens component life.

As one remedy, Westinghouse Electric Corporation performed an analysis and testing (WCAP-11307) to determine if the steam generator tube plugging criteria could be modified for that portion of the tube which had been 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 explcsively 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 prevent plugging tubes with defects below the F* distance, there can be no indications of cracking in the F* distance. Accordingly, the Trojan Technical Specifi-cations and Bases have been revised to modify the tube plugging limit.

l V. l LCA 152  ;

Page 2 of 6

, SIGNIFICANT HAZARDS CONSIDERATION / DETERMINATION As required tnr 10 CFR 50.91(a)(1) this analysis is provided to demonstrata the proposed changes do not represent a significant hazards considera-tion. In accordance with 10 CFR 50.92(c), implementation of the proposed l

license amendment was analyzed using the following standards and found not to: 1) involve a significant increase in the probability or conse-

.quences of an accident previously evaluated; or 2) create the possibility of a new or different kind of accident from any. accident previously.

evaluated; or 3) involve a significant reduction in a margin of safety.

The amendment has been proposed due to ambiguous eddy current. indications l

of tube degradation in the mechanical roll expanded' portion of the tubes within the tube sheet in the steam generators at. Trojan. These steam j 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 plugging or repair.ls not required in cases where there are-indications in the 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. The proposed amend-ment 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 generator in LOCA analyses. The proposed amendment would 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.

The possibility of tube repair by sleeving should not be considered a reason to exclude use of the modified tube sheet plugging limit, but l 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 due to the

changes in geometry at the ends and expansions of the sleeve and the size of probe that can pass through the reduced diameter of the sleeve. The Trojan Technical Specifications do not currently contain provisions for l

repair of tubes by sleeving.

The proposed license amendment addresses the action required when degrada-  !

i tion has been detected in the partial depth mechanically expanded hardroll portion of steam generator tubes within the steam generator tube sheet.

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4 LCA 152 Page 3 of 6 Existing tube repair or plugging criteria, ie, current applications of USNRC Regulatory Guide 1.121, do not take into account the effect of the tube sheet on the external surface of the tube. The presence of the tube sheet will enhance the integrity of degraded tubes in that region by precluding tube deformation beyond the expanded outside diameter. Addi-tionally, a portion of the expansion at the top end of the roll expansion is sufficient to preclude pullout of the tube during normal operaticn and postulated accident condition loadings if a tube were postulated.to sever circumferential1y during Plant operations in the portion of the tube covered by the proposed amendment. Finally, the roll expansion of the tube into the tube sheet provides a barrier to significant leakage if through-wall cracking of the tube occurs in this region. ,

The proposed change designates a portion of the tube for which tube degradation does not necessitate remedial action except as dictated for compliance with tube leakage limits as set forth in the Trojan Nuclear Plant Technical Specifications. As noted above, the area subject to this change is in the mechanically expanded portion of the tube within the i tube sheet of the steam generators. The length of. mechanical expansion i required to resist pullout for all postulated conditions has been deter-j mined to be 0.91 inches. To this length, a distance of 0.5 inches has been added for eddy current measurement uncertainty and the total rounded

! to 1.4 inches. The length of 1.4 inches is designated as the F* dis- l j tance. The definition of F* in WCAP-11307 does not include the eddy current measurement uncertainty. For the purposes of the evaluation of F*, the explosively expanded portion of the tube was conservatively l assumed not to provide resistance to pullout'or' leakage. Since the roll expansion of the tube in the F* distance is sufficient to preclude pull- I out of the tube, use of the F* criteria does not depend on any determina-tion of the condition of tube degradation in the portion of the tube below the F* distance.

The proposed amendment would modify Trojan Technical Specification 4.4.5, 4 " Steam Generator Surveillance Requirements" and the bases for Technical '

Specification 3/4.4.5, " Steam Generators", which provide tube inspection i requirements and acceptance criteria to determine the level of degrada-tion for which the tube may remain in service. The proposed amendment revises the definition for tube plugging limit to prescribe the portion of the tube subject to the F* criteria. The proposed Technical Specification changes accompany this analysis.

Conformance of the proposed amendments to the standards for a determina-tion of no significant hazards as defined in 10 CFR 50.92 is shown in the following:

1. Operation of the Trojan Plant in accordance with the proposed license amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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LCA 152 Page.4 of 6 The supporting technical and safety evaluations of the subject criteria.[ Westinghouse WCAP-11307 "Tubesheet. Region Plugging Criterion for the Portland General Electric Company Trojan Nuclear.

Station" (Proprietary), and WCAP-11315 (Non-Proprietary)] demonstrate that the presence of the tube sheet will enhance the tube integrity in the region of the hardroll by precluding tube deformation beyond its initial expanded outside diameter. The resistance to both tube rupture and tube collapse is strengthened by the presence of 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 sufficient movement of tube material. In a similar manner, the tube sheet does not permit sufficient movement of tube material to permit buckling collapse of- l the tube during postulated LOCA loadings.

Additionally through analysis and testing,' Westinghouse has demon-j strated that the roll expansion over the F* distance is sufficient as j long as it contains no indications of cracking to preclude pullout of 1 the. tube from the tube sheet. Even with the conservative assumption that a tube could completely sever circumferential1y below the F* distance, test results demonstrate that pullout of the tube is precluded under normal and postulated accident condition loadings.

l 5 Relative to expected leakage, the length of the roll expansion in the F* distance is sufficient as long as it contains no indications of j cracking to preclude significant leakage from tube degradation j located below the F* distance. The existing Technical Specification leakage rate requirements and accident analysis assumptions remain unchanged in the unlikely event 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 l 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 I design basis accident.

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2. The proposed license amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated.

Implementation of the proposed _F* criteria does not introduce any significant changes to the Plant design basis. Use of the criteria 7

does not create the possibility of a new or different kind of acci-dent not previously evaluated. Any hypothetical accident.as a result of any tube degradation in the expanded portion of the tube would be bounded by the existing tube rupture accident analysis.

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4 LCA 152 Page 5 of 6

3. The proposed license amendment does not involve a significant a reduction in a margin of safety.

The use of the F* criteria has been demonstrated to maintain the~ l 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 postu-lated accident conditions. Acceptable tube degradation is any degradation in the tube below the F* distance. The safety factors used in the determination of the F* distance are consistent with the safety factors in the ASME Boiler and Pressure 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 normal and postulated accident condi-tions. Additionally, for axial or nearly axial indications in the j tube sheet region, the tube end remains structurally intact further decreasing any potential for tube pullout.

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*

i criteria would not be expected to result in significant leakage from i

through wall cracks located below the F* distance.

Implementation of the F* criteria will decrease the number of tubes l q which must be taken out of service with tube plugs or repaired with j sleeves. Both plugs and sleeves reduce the RCS flow margin, thus implementation of the F* criteria will maintain the margin of flow l that would otherwise be reduced in the event of increased plugging or sleeving. Based on the above, it is concluded that the proposed change does not result in a significant reduction in a loss of margin with respect to Plant safety as defined in the Updated Final Safety Analysis Report or the bases of the Plant Technical Specifications.

CONCLUSION Based on the preceding analysis it is concluded that operation of Trojan Nuclear Plant in accordance with the proposed amendment does not result in the creation of an unreviewed safety question, an increase in the probability of an accident previously evaluated, create the possibility of a new or different kind of accident from any accident previously

{ evaluated, nor reduce any margins to Plant safety. Therefore, the license amendment does not involve a Significant Hazards Consideration as defined in 10 CFR 50.92.

LCA 152 Page 6 of 6 In the April 6, 1983 Federal Register, the NRC published a list of examples of amendments that are not likely to involve Significant Hazards Considerations. Example Number.6 of that list applies to these proposed changes and states that a change similar to the following would not likely involve a Significant Hazards Consideration:

"A change which either may result in some increase in.the probabil-ity or consequences of a previously-analyzed accident or may reduce in some way a safety margin, but where the results of the change are clearly within all acceptable criteria with respect to the system or component specified in the Standard Review Plan, eg, a change resulting from the application of a small refinement of a previously used calculational model or design method."

In addition, the NRC has previously issued similar amendments for the l Virgil C. Summer Nuclear Station Unit 1 and McCuire Nuclear Station Units 1 and 2 with the conclusion that.the changes did not involve a i

Significant Hazards Consideration. These amendments were for full-depth hard-rolled tubes whereas this amendment request only involves approxi-mately 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.

l SAFETY / ENVIRONMENTAL EVALUATION Safety and environmental evaluations were performed as required by 10 CFR 50 and the Trojan Technical Specifications. This review deter-mined that an unreviewed safety question does not exist since Plant operations remain consistent with the Updated FSAR, adequate surveillance is maintained, and there is c.o conceivable impact upon the environment.

RWR/kal 5918k.487

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