ML20205P989

Requests Approval for Proposed Alternatives to Liquid Penetrant Requirements of N-518.4 of 1968 ASME Boiler & Pressure Vessel Code.Results of Analysis & Summary of Tests Performed & Tests Results Are Encl

ML20205P989
Person / Time
Site:	Prairie Island
Issue date:	04/12/1999
From:	Sorensen J NORTHERN STATES POWER CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20205P991	List: ML20205P989 ML20205Q004
References
NUDOCS 9904210054
Download: ML20205P989 (7)

Text

u Northern States Power Company Prairie Island Nuclear Generating Plant 1717 Wakonade Dr. East Welch. Minnesota 55089 April 12,1999 10 CFR 50.55a(a)(3)

U S Nuclear Regulatory Commission Attn: Document Control Desk {

Washington, DC 20555 PRAIRIE ISLAND NUCLEAR GENERATING PLANT Docket Nos. 50-282 License Nos. DPR-42 50-306 DPR-60 Reauest for Approval of Alternative to ASME Code Recuirements Prairie Island Unit 2 shut down on January 24,1998 to repair a small RCS leak. The source of this leakage was identified as a flaw in the wall of the part length control rod drive mechanism (CRDM) at location G9 on the reactor vessel head.

Because the part length CRDMs are not used and are abandoned in place, it was decided that all of the Unit 2 part length CRDMs would be permanently removed.

Removal of the part length CRDMs required capping of the associated reactor vessel 7 head penetrations. The method for capping of the penetrations was the installation of a / l cap that was screwed onto the threaded end of the penetration and then seal welded.

During the upcoming Unit 1 refueling outage, the four Unit 1 part length CRDMs will also be removed and capped. /[j2)t/7 l Based on N-518.4 of the 1968 ASME Boiler and Pressure Vessel Code, a liquid '

penetrant examination of the seal weld is required. However, liquid penetrant examination of the seal weld would be difficult. The CRDM penetrations being repaired are located in a high radiation area, with radiation fields of approximately 1000 mr/hr.

Additionally, access to the seal welds is difficult due to the limited clearance between the adjacent control rod drive housings. The separation between the outer rod travel housings is approximately 7.2 inches. This is not adequate clearance to gain complete access to the inner rod travel housings to perform the liquid penetrant examination of the seal welds. Final weld surface preparation, the liquid penetrant examination and the subsequent cleanup would be difficult and time consuming due to the limited access and personnel performing these operations would incur substantial radiation exposure.

While the liquid penetrant examination specified by N-518.4 would provide indication of surface cracks, the processes used to perform the seal welds and the visual 9904210054 990412 PDR ADOCK 05000282 P PDR ,

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USNRC NORTHERN STATES POWER COMPANY April 12,1999 Page 2 examination of the welds provide the best measure of the seal weld acceptability due to the limited accessibility and high radiation fields. A similar alternative to liquid penetrant examination was requested by NSP letter dated February 13,1998 and approved for Unit 2 by NRC letter dated February 20,1998.

The surface to be seal welded is examined with an 8x camera during weld surface i preparation. The weld is deposited using a fully automatic TIG process. All welding parameters are controlled within the qualified range from a remote panel. The weld puddle / deposit is observed via an 8x camera during every phase of the welding. A final visual examination of the weld surface is completed using the same 8x camera. In j addition, the post outage in-service pressure test of the Unit 1 reactor coolant system l will include VT-2 inspection of the seal weld for leakage.

10 CFR Part 50, Section 50.55a(a)(3) allows the use of alternatives to the ASME Code requirements, when authorized by the Director of the Office of Nuclear Reactor Regulation, if it can be demonstrated that:

1. The proposed alternatives would provide an acceptable level of quality and safety, or

2. Compliance with the specified requirements of this section would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety. l In accordance with the provisions of 10 CFR Part 50, Section 50.55a(a)(3), we are proposing the following alternatives to the liquid penetrant testing requirements of N-518.4 of the 1968 ASME Boiler and Pressure Vessel Code for the weld repairs described above:

1. Use of a controlled automatic welding process.

2. Observation of the weld puddle / deposit via an 8x camera during the welding process.

3. A final visual examination of the weld surface using the same 8x camera.

4. Performance of a VT-2 inspection of the seal weld for leakage during the post outage in-service pressure test.

5. Authorized Nuclear Inservice Inspector approval of alternative testing and NIS-2 4

acceptance.

A liquid penetrant examination would provide a more stringent verification of the final weld surface condition and therefore afford an added measure of the quality and safety of the completed seal weld. However, the liquid penetrant examination does not UIPt.CRDM relief req. DOC j

USNRC NORTHERN STATES POWER COMPANY April 12,1999 Page 3 provide a substantial increase in quality and safety above what is provided by the measures (controlled process, observation of weld process using 8x camera, final 8x visual inspection and in-service pressure test inspection) that have been and will be taken in lieu of the liquid penetrant examination, in addition, due to the time consuming nature of the examination process, personnel would incur substantial radiation J

exposure during the performance of liquid penetrant examinations. I An analysis was performed by Struct' ral Integrity Associates to demonstrate that (for Unit 2) a through-wall flaw could be detected by visual examination which has' a flaw size which is sufficiently smaller than the critical flaw size, thus assuring sufficient safety ,

margins. The analysis demonstrated that, under a variety of conservative assumptions, I the critical flaw size predicted for the repair geometry is in all cases of significant length.

It is likely that a much smaller flaw could be credibly detected by visual examination ,

under 8x magnification. Based on similarity of design, geometry, and materials, the '

. analysis also applies to Unit 1 cap welding. The analysis results are summarized in  !

Attachment 1. J l

In order to confirm the detectable flaw size, tests were performed by Welding Services Incorporated to evaluate the capabilities of the camera system used in ine performance of the weld repair. This testing confirmed that the critical flaw sizes iesulting from the Structural Integrity analysis are detectable with margin by the visual inspection i technique. Since the same camera will be used for the Unit 1 cap welding, these i camera test results are applicable to the Unit 1 cap welding. A summary of the tests performed and the test results are provided as Attachment 2.

In conclusion, the proposed alternatives (automatic weld process, observation of the .

process using 8x camera, final 8x visual examination and in-service pressure test  ;

inspection) to the liquid penetrant requirements of N-518.4 of the 1968 ASME Boiler '

and Pressure Vessel Code provide an acceptable level of quality and safety for the seal welds on the part length CRDM penetrations. Furthermore, compliance with the liquid penetrant examination requirements of N-518.4 of the 1968 ASME Boiler and Pressure Vessel Code would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.

UIPLCRDM relief req. DOC

USNRC NORTHERN STATES POWER COMPANY April 12,1999 Page 4 We have made no new Nuclear Regulatory Commission commitments in this letter.

Please contact Jeff Kivi(651-388-1121)if you have any questions related to this reque t.

Joel P. Sorensen Site General Manager Prairie Island Nuclear Generating Plant c

Regional Administrator- Region Ill, NRC l Senior Resident inspector, NRC NRR Project Manager, NRC l J E Silberg Attachments: 1. Calculation Package: Evaluation of Limiting Flaws for Structural Adequacy in CRDM Repair Adapter Plug Fillet Weld Evaluation at Prairie Island Unit 2 i

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2. Summary of Camera Testing j i

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UIPLCRDM relief req. DOC L

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' j ATTACHMENT 1 CALCUL.ATION PACKAGE:

EVALUATION OF LIMITING FLAWS FOR STRUCTURAL ADEQUACY IN CRDM REPAIR ADAPTER PLUG FILLET WELD EVALUATION AT PRAIRIE ISLAND UNIT 2 STRUCTURAL INTEGRITY ASSOCIATES, INC l

Structural Integrity Associates, Inc.

331s Almaden Erpresresy suite 24 San lame.CAtst18-1s57 Prene:

,,; 40,C-976

, , , . 8200 February 12,1998 MLH-98 006 Wtecrn Mr. Dick Cooper Northem States Power Company Prairie Island Nuclear Generation Plant 1717 Wakonade Drive East Welch,MN 55089 l

Subjec:: Evaluation of Critical Flaw Sizes for CRDM Canopy Fillet Weld

Dear Mr. Cooper:

Structural Integrity Associates has performed an evaluation to determine the critical flaw size for the subject location. The results of this evaluation can be us,ed to demonstrate that the cdtical flaw size is significantly larger than the flaw size observable'using visual inspection techniques.

This can serve to eliminate the need to perform dye penetrant testing of the fillet weld.

The evaluation was performed using limit load metheds since the materialis ductile and the fluence at this location is below that needed to impact the material fracture toughness. The evaluation was performed assuming a through-wall axial flaw and a through-wall circumferential flaw. The stress in the fillet weld was assumed equal to the design stress intensity of the material (S.). The S. for the stainless steel at 650*F was used in this calculation (16.2 ksi). A safety factor of 1.0 was used since the critical flaw size is being calculated.

The calculatiens were performed by assuming a pipe with rbius and thickness equa fillet weld location (0.265" equivalent thickness, mean radiu's =3.3 inches). The results for the critical flaw lengths (m terms of fraction of circumference add characteristic length parameter, fraction of 4Rt) are independent of the pipe geometry since'the stress is ret ta S The cdtica!

axial flaw was determined using the SI program, pc CRACK, and the critical circumferential flaw l size was determined using the EPRI DLL program.

I The results of the evaluation are shown below.

Critical Through-Wall Axial Flaw Length: 4.17 inch Critical Through-Wall Circumferential Flaw Length: 6.95 inch, 1

It should be noted that this calculation was performed using' an applied stress equivalent'to the S.

for the stainless steel material, which is conservative. If the' actual hi l i were used, the critical flaw lengths are expected to be even larger. pl l Osstens.fC sil,sr Sprise.NO Poneans leasa. H hiset istmas Bas Jena. CA Atron. CH Phone: 704 67J.t356 Phou:30148623Z3 Nee Et tT 27tt Ptaas#238455M Phons:4804?>8200 Phons 330-464 8086

Mr. Dick Cooper February 12,1998 Page 2 MLH-98-006 l I have included the output listing for the two cases considered in this calculation. I will forward the entire cakulation package to you shortly as soon as it is prepared.

S' el ,

L. errera, P. E.

, Senior Consultant Attachments H. L. Gustin i cc:

R. A.Mattson G. A.Miessi NSP-27Q 1

i f Structuralintegrity Associates, Inc.

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