ML20101N014

Forwards Info Associated W/Analysis Performed for Canopy Seal Weld Repair on Facility During Cycle 7 Refueling Outage,As Requested During Telcon W/Nrc

ML20101N014
Person / Time
Site:	Sequoyah
Issue date:	04/03/1996
From:	Shell R TENNESSEE VALLEY AUTHORITY
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 9604080195
Download: ML20101N014 (9)

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IUA Tennessee VaUey Ae'nority. Post Othce Box 2000. Soddytesy. Tennessee 37379 l

April 3,1996 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555 Gentlemen:

In the Matter of ) Docket No. 50-327 Tennessee Valley Authority )

SEQUOYAH NUCLEAR PLANT (SON)- ADDITIONAL INFORMATION FOR AMERICAN SOCIETY OF MECHANICAL ENGINEERS (ASME) ALTERNATIVE CANOPY SEAL WELDS - UNIT 1

References:

1. TVA letter to NRC dated October 11,1995, "Sequoyah Nuclear Plant (SON)- Request for Approval of Alternative to American Society of

Mechanical Engineers (ASME) Code Requirements - Canopy Seal Welds - Unit 1" l 2. TVA letter to NRC dated December 19,1995, "Sequoyah Nuclear i

Plant (SON) - Additional Information for American Society of Mechanical Engineers (ASME) Alternative Canopy Seal Welds -

Unit 1" l

As requested durmg a telephone conversation with the NRC staff, we are enclosing information associated with the analysis that was performed for the canopy seal weld repair on SON Unit *> during the Unit 1 Cycle 7 refueling outage. The analysis demonstrates that under a variety of conservative assumptions, the critical flaw size predicted for the repair geometry is, in all cases, of significant length and would have been detected by the enhanced visual examination that was performed. This information is provided in conjunction with the above referenced letters.

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i U.S. Nuclear Regulatory Commission Page 2 April 3,1996 Please direct questions concerning this issue to D. V. Goodin at (423) 843-7734.

Sincerely, krY. A R. H. Shell  :

i Manager I

SON Site Licensing Enclosure cc (Enclosure): '

Mr. D. E. LaBarge, Project Manager ,

Nucle 9r Regulatory Commission l

One White Flint, North l 11555 Rockville Pike Rockville, Maryland 20852-2739 ~ ,

i NRC Resident inspector Sequoyah Nuclear Plant 2600 Igou Ferry Road Soddy-Daisy, Tennessee 37379-3624 Regional Administrator U.S. Nuclear Regulatory Commission Region 11 i

101 Marietta Street, NW, Suite 2900 Atlanta, Georgia 30323-2711 3

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.g 4 ENCLOSURE ANALYSES ASSOCIATED WITH THE WELD OVERLAY REPAIR e

FOR THE SEQUOYAH NUCLEAR PLANT UNIT 1 CONTROL ROD DRIVE MECHANISM LOWER CANOPY SEAL WELDS 1 i l l l

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NRC Reauest NRC requested additionalinformation concerning the repair of SON's canopy seal welds. NRC's questions were (1) how big a flaw the repair weld can tolerate in the heat affected zone going into the pressure boundary and (2) how big a flaw could the remote video camera detect?

TVA Resoonse in response to NRC questions, StructuralIntegrity Associates performed a calculation l of the critical flaw lengths for four cases (see attachment). The lengths of flaws ranged from 4.4 inches to 7.3 inches. The results of these calculations demonstrate, under a variety of conservative assumptions, the.the critical flaw size predicted for j the repair geometry of the canopy seal weld is, h all cases, of significant length. I Welding Services Incorporated (WSI) performed a test of the video system for Prairie Island where the same type of canopy seal weld repairs were made. The test demonstrated that two wires of 0.005 and 0.001 inch in diameter and 0.4 inch in

'ength could be adequately viewed. The test was performed on a mockup of a canopy seal housing similar in configuration to the Prairie Island design, which is also similar ;

to SON's design. WSI's quality control personnel were able to adequately resolve the i two wires on the mockup using the remote video equipment. WS1 used equipment at )

SON identical to that used at Prairie Island. TVA performed a demonstration examination for the Authorized Nuclear inspector using the remote video equipment at SON prior to using it for examination purposes. The demonstration was performed j using a machinist scale to determine if a 1/32 of an inch graduation could be '

distinguished. This is the normal requirement for a visual examination. The demonstration examination was found acceptable.

Therefore, based upon the above, TVA finds that the remote video equipment could adequately detect the postulated cracks. This method provides adequate assurance of ,

safety and is considered by TVA to be an acceptable alternative for the dye penetrant I examination. l l

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StructuralIntegrity Associates, Inc.  ;

3315 Almaden Expressway Suite 24 February 15,1996 sanJose CA95118-1557 HLC-96-022 Pnone: 408 4 78-8200 Fax: 408-978-8964 Mr. Ken Wilson Tennessee Valley Authority Sequoyah Nuclear Power Plant Sequoyah Access Road Soddy Daisy, TN 37379

Subject:

Evaluation ofLimiting Flaws for Structural Adequacy in Canopy Seal Repairs at Sequoyah Nuclear Power Plant

Dear Ken,

At the request of the NRC and in support of the use of visual examination rather than dye penetrant examination of the completed weld overlay repairs to canopy seal welds at Sequoyah, Structural Integrity Associates (SI) performed several analyses to dete,nine the critical flaw size in the repaired  ;

location. The purpose of these analyses was to demonstrate tuat a through wall flaw which has a flaw size which is sufficiently smaller than the critical flaw size , could i s detected by visual examination 1 thus assuring sufficient safety margins. Tennesse Valley Authority (FA) will review the critical flaw sizes determined in this calculation to confirm than the resulting sizes are detectable with margin by -

the visual technique.

The analysis results are summarized below.

1.0 GEOMETRY The design geometry of the repair is illustrated in Figure 1. For the purpose of the present evaluation, the component was modeled as a pipe with outside radius equal to that of the CRDM (3.22 inches) and wall thickness equal to the overlay thickness in the vicinity of the CRDM outside surface (0.36 inches minimum), as shown on Figure 1. Through wall axial and circumferential flaws were evaluated. These geometries are considered to be reasonable representations of the actual design geometry. The model geometries are shown in Figure 2.

2.0 APPLIED STRESSES For conservatism, the applied stress was assumed to be membrane strcss at the Code allowable membrane level (P, = S.). No distinction was made between the hoop and axial directions in this regard, although realistically, the axial direction should be half of this value. Based upon discussions Alrroe. OM Edvor Senas, een Ft Lassersaes. Fi fa pet, Taiwee setemoirtes. Int.

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I Page 2 February 15,1996 l Ken Wilson HLG-96-022 with plant personnel, there are no bending loads present. Therefore, bending stresses were not considered.

3.0 MATERIAL PROPERTIES The allowable stress, S,was taken to be 16.2 ksi at 650 F, which is typical of 304 stainless steel.

The Alloy 625 material of the weld overlay repair has a significantly higher allowable stress at this temperature, so use of the stainless steel value is conservative. The flow stress for this component  ;

was taken as 3 S,.

For linear clastic fracture mechanics evaluations, the Ku was taken as 135 ksi/i ii(as recommended in NRC Generic Letter 90-05) which is very conservative for this material at this temperature.

4.0 ANALYTICAL APPROACH Two analysis methodologies were employed. The limit load (net section collapse) method is considered most appropriate for evaluation of through wall flaws in this very ductile material. This i method is described in Appendix C of ASME Section XI. For comparison, linear elastic fracture mechanics (LEFM) methods were also applied.

A total of four cases were studied. These were:

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1. Through wall axial flaw. Limit load.  ;

2. Through wall axial flaw. LEFM

( 3. Through wall circumferential flaw. Limit load. ,

4. Through wall circumferential flaw. LEFM i

No Code safety margins were included in this evaluation, since the objective is to get a reasonable view of the relationship between detectable and critical flaw sizes.

The results from each case are summarized below.

I 5.0 RESULTS 5.1 Through wall axial flaw. Limit load.

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l This case follows the methodology underlying ASME Section XI, Tables IWB-3641 and  ;

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Appendix C. The SI program pc-CRACK was used to perform the analysis. The conclusion

is that an axial flaw could be at least 4.4 inches long before leac%g to incipient collapse. This j is much longer than is physically achievable, since cracking would be expected to be confined

{ to the weld overlay material and vicinity, which, in the axial direction, extends j approximately 1 *mch.

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f StructuralIntegntyAssociates,Inc.

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l Page 3 February 15,1996 Ken Wilson HLG-96-022 l

5.2 Through wall axial flaw. LEFM.

I This analysis assumes that brittle failure is the operative mechanism. The pc-CRACK program is used with this analysis. A fracture mechanics model of a through wall crack in a cylinder under internal pressure was used, together with an assumed fracture toughness Ku

= 135 ksi/Tii. The conclusion is that for this set of assumptions, the critical flaw length is greater than 5.0 inches.

i 5.3 Through wall circumferential flaw. Limit Load.

l This analysis used hand calculations using the methods of Section XI Appendix C. The SI l program ANSC was also used to perform a separate analysis of the same configuration. The analysis assumed a through wall circumferential flaw, and determined the critical flaw length l

using limit load techniques. The conclusion is that such a flaw could be 125* around the j cylinder before reaching a critical size. This corresponds to a flaw approximately 7.08 inches l long.

l 5.4 Through wall circumferential flaw. LEFM.

This analysis assumed that the failure mode was brittle failure. The pc-CRACK program was used with a through wall circumferential flaw in a cylinder under remote tension fracture mechanics model. A Ku = 135 ksi(in)^0.5 was conservatively assumed. The conclusion of l this analysis was that the critical flaw length for this set of assumptions was 7.3 inches.

6.0 CONCLUSION

S The above results demonstrate that, undet a variety ofconservative assumptions, the critical flaw size predicted for the repair geometry is in all cases of significant length. It is likely that a much smaller riaw could be detected by an enhanced visual examination. TVA should confirm their detectable flaw 1 size, to complete the demonstration that the visual examination provides s.dequate assurance of safety, l l Please call if you have any questions regarding the above.  :

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Sincerely,

)/ H l M Mn

! H. L. Gustin, P. E.

. Associate' i

i /cd i attachment i cc: WSI-20Q-102 f StructuralIntegrityAssociates,Inc.

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Figure 2. Modeled Geometries l

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-1 Attachment to: HLG.96-022 f StructuralIntegrityAssociates,Inc.

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