ML20196F203
| ML20196F203 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 11/30/1998 |
| From: | Sorensen J NORTHERN STATES POWER CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20196F208 | List: |
| References | |
| NUDOCS 9812040219 | |
| Download: ML20196F203 (5) | |
Text
j, Northern States Power Company Prairie Island Nuclear Generating Plant 1717 Wakonade Dr. East Welch, Minnesota 55089 November 30,1998 10 CFR 60.55a(a)(3)
U S Nuclear Regulatory Commission Attn: Document Control Desk i
Washington, DC 20555 PRAIRIE ISLAND NUCLEAR GENERATING PLANT I
Docket Nos. 50-282 License Nos. DPR-42 50-306 DPR-60 i
Request for Approval of Alternative to ASME Code Requirements During reactor head inspection during the current Unit 2 refueling shutdown, boric acid residue was noted at the lower canopy seal weld of full length control rod drive mechanism (CRDM) at location E11 (Attachment 3). The residue was confined to the lower canopy seal weld with minor spray on surrounding housings; no boric acid residue was noted below the upper coil housing on the penetration or reactor vessel head insulation.
The repair options were evaluated and it was determined that the most appropriate repair was the use of a weld buildup rather than removing the defect and performing a 7i ff)Q/I weld repair. Weld buildup was an acceptable repair technique because the canopy seal weld does not provide the structural strength or the pressure boundary for the joint.
A fracture mechanics analysis was performed to justify not removing the existing defect.
Even though the canopy seal does not provide structural strength for the joint, the weld buildup over the canopy seal is considered a repair under the rules of ASME Section XI, IWA-4000 because the welding is performed on pressure retaining components.
The need for NRC review and approval of the fracture mechanics analysis was discussed with the NRC Staff on June 1,1995 and it was determined that no NRC review was required.
Based on N-518.4 of the 1968 ASME Boiler and Pressure Vessel Code, a liquid penetrant examination of the weld buildup is required. However, liquid penetrant examination of the canopy seal weld buildup would be difficult. The canopy seal being 9812040219 981130
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USNRC NORTHERN STATES POWER COMPANY November 30,1998 Page 2 r'epaired is located in a high radiation area, with radiation fields of approximately 400 mr/hr. Additionally, access to the canopy seal being repaired is difficult for much of the circumference. The separation between the outer rod travel housings is approximately 7.2 inches. There is not adequate clearance to gain complete access to the inside surface to perform the liquid penetrant examination of the weld repairs. Final weld surface preparation (grinding), 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 weld buildup and the visual examination of the welds provide the best measure of the lower canopy seal weld buildup acceptability due to the limited accessibility and high radiation fields. The surface to be repaired is examined with an 8x camera during weld surface preparation.
The weld buildup 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 a 8x camera during every phase of the welding. A final visual examination of the weld surface is completed using the same 8x camera. In addition, the post outage system leakage test of the reactor coolant system will include a VT-2 inspection of the Icwar canopy seal weld area 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.
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-i 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 a 8x camera during the welding process.
- 3. A final visual examination of the weld surface using the same 8x camera.
canopysealregrelief1998 doc
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November 30,1998 i
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- 4. Performance of a VT-2 inspection of the canopy seal weld area for leakage during the post outage system leakage test inspection.
- 5. Authorized Nuclear Inservice inspector approval of alternative testing and NIS-2 acceptance.
We also propose using the above alternatives for the remaining lower and intermediate canopy seal welds on both units. These welds would be pre-emptively overlaid during i
the upcoming refueling outages. Prairie Island has performed this repair on several l
occasions and no cracks or leaks have been seen during subsequent shutdown and startup inspections.
l 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 weld buildup. However, the liquid penetrant examination does not l
provide a substantial increase in quality and safety above what is provided by the l
measures (controlled process, observation of weld process using 8x camera, final 8x visual inspection and post outage system leakage test inspection) that have been and l
will be taken in lieu of the liquid penetrant examination.
t An analysis was performed by Structural Integrity Associates to demonstrate that 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, 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 l
8x magnification. The analysis results are summarized in Attachment 1.
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 the performance of the weld repair. This testing confirmed that the critical flaw sizes resulting from the Structural Integrity analysis are detectable with margin by the visual inspection technique. A summary of the tests performed and the test results are provided as l.
In conclusion, the proposed alternatives (automatic weld process, observation of the process using 8x camera, final 8x visual examination and post outage system leakage test inspection) to the liquid penetrant requirements of N-518.4 of the 1968 ASME Boiler Code and Pressure Vessel provide an acceptable level of quality and safety for weld repairs to the lower canopy seal welds. Furthermore, compliance with the liquid penetrant examination requirements of N-518.4 of the 1968 ASME Boiler and Pressure canopyseatrogreliefi998. doc
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USNRC NORTHERN STATES POWER COMPANY November 30,1998 Page 4 essel Code would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.
We have made no new Nuclear Regulatory Commission commitments in this letter.
Please contact Jack Leveille (612-388-1121, Ext. 4142)if you have any questions related to this request.
Joel P. Sorensen Plant Manager Prairie Island Nuclear Generating Plant c: Regional Administrator-Region 111, NRC Senior Resident inspector, NRC NRR Project Manager, NRC J E Silberg Attachments: 1. Evaluation of Lirniting Flaws for Structural Integrity in Canopy Seal Repairs at Prairie Island Nuclear Plants
- 2. Summary of Camera Testing
- 3. Control Rod Locations Figures t
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canopysealregreheft998. doc
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t EVALUATION OF LIMITING FLAWS FOR STRUCTURAL INTEGRITY IN CANOPY SEAL REPAIRS AT PRAIRIE ISLAND NUCLEAR PLANT i
STRUCTURAL INTEGRITY ASSOCIATES,INC i
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