ML20216G942
| ML20216G942 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 03/16/1998 |
| From: | Sorensen J NORTHERN STATES POWER CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NUDOCS 9803200140 | |
| Download: ML20216G942 (13) | |
Text
Northern States Power Company Prairie Island Nuclear Generating Plant i
1717 Wakonade Dr. East Welch. Minnesota 55089 March 16,1998 10 CFR Part 50 Section 50.55a j
U S Nuclear Regulatory Commission j
Attn: Document Control Desk I
Washington, DC 20555 l
i PRAIRIE ISLAND NUCLEAR GENERATING PLANT Docket Nos. 50-282 License Nos. DPR-42 50-306 DPR-60 Request for Relief No. 7 for the Unit 13rd 10-year Interval inservice Inspection Program I
On August 5,1994 we submitted for review our third 10-year Inservice Inspection i
Examination Plan for Unit 1 and, on March 28,1995, a response to a request for additional information to that plan. The NRC issued its evaluation of the 3rd 10-year interval Program Plan on February 22,1996.
The purpose of this letter is to submit a relief request for " limited examinations" associated with that plan. Attached is Unit 1 Relief Request No. 7, Revision 0 which addresses those limited examinations. We are requesting relief pursuant to 10 CFR Part 50, Section 50.55a(g)(5)(iii) due to the impracticality of obtaining "100%"
examination coverage for the affected items.
There are two items (B1.100: W-7 IR and W-101R, reactor vessel inner radii) where only 91.5% coverage was obtained for which we are not requesting relief at this time.
We believe that we will be able to achieve 100% coverage during the 10-year inspection when the core barrelis removed and we will have better access to examine the volume.
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.i,f 9803200140 980316 o"
^oockosoma aggpiMMI
USNRC NORTHERN STATES POWER COMPANY March 16,1998
' Page 2 in this letter we have made no new Nuclear Regulatory Ccmmission commitments.
Please contact Jack Leveille (612-388-1121, Ext. 4662) if you have any questions related to this lette.
oel P Sorensen Plant Manager Prairie Island Nuclear Generating Plant I
c: Regional Administrator - Region Ill, NRC Senior Resident inspector, NRC NRR Project Manager, NRC J E Silberg
Attachment:
ISI Relief Request No. 7 (Rev. 0) l U1RR7RD. DOC
' NORTHERN STATES POWER INSERVICE INSPECTION j
PRAIRIE ISLAND UNIT 1,3RD INTERVAL EXAMINATION PLAN 1
ISI Relief Request No. 7 (Rev. 0)
Limited Examination SYSTEM: Various Class: 1 and 2 Category: Various Item: Various Imoractical Examination Reauirements:
ASME Section XI (1989 no addenda) Code requires full examination of inservice l
inspection (ISI) components per Table IWB-2500-1, and IWC-2500-1. Reg. Guide 1.147 endorses Code Case N-460, " Alternative Examination Coverage for Class 1 and Class 2 Welds." This code case allows greater than 90% coverage of a weld to meet j
the " essentially 100%" requirement.
The Prairie Island construction permit was issued in 1967. This facility was designed and constructed with limited accessibility due to component configurations and/or physical barriers for which 100% coverage is not achievable on some ISI components examined for the Third Ten Year Interval.
Basis for Relief:
The following 10 CFR 50.55a paragraphs apply to the inservice inspection of components in accordance with the ASME Section XI code:
50.55a(g)(1): For a boiling or pressurized water-cooled nuclear power facility whose construction permit was issued prior to January 1,1971, components (including supports) must meet the requirements of paragraphs (g) (4) and (5) of this section to the extent practical.
50.55a(g)(4): Throughout the service life of a boiling or pressurized water-cooled nuclear power facility, components (including supports) which are classified as ASME Code Class 1, Class 2, and Class 3 must meet the requirements, except design and access provisbns and preservice examination requirements, set forth in Section XI of editions of the ASME Boiler and Pressure Vessel Code... to the extent practical within the limitations of design, geometry and materials of construction of the components.
50.55a(g)(5)(lv): Where an examination requirement by the code or addenda is determined to be impractical by the li;ensee and is not included in the revised inservice inspection program as permitted by paragraph (g)(4) of this section, the basis for this determination must be demonstrated to the satisfaction of the Commission...
Prairie Island was designed and constructed prior to development of ASME XI and therefore design for inspectability and inspection coverage is not in many cases, sufficient to permit satisfying the current Code requirements. Limitations to inspections are primarily due to obstructions and interferences.
Page 1 of 4
j
' NORTHERN STATES POWER INSERVICE INSPECTION PRAIRIE ISLAND UNIT 1,3RD INTERVAL EXAMINATION PLAN l
Summary of the limited examinations are described below and also included in Table 1 attached:
Part A: Category B-A, " Pressure Retaining Welds in Reactor Vessel," Reactor Vessel; the volumetric coverage for the Head-to-Flange weld is 54.20%.
Examinations are performed with ultrasonic automated techniques from the outside diameter (OD). Additional coverage can not be obtained due to weld geometry and interference from three lifting lugs as identified in Figure 1.
Examination from the inside diameter (ID) is not practical due to ALARA constraints.
Part B: Category B-D, " Full Penetration Welds of Nonles in Vessels," Steam Generators (S/G) Primary Side; the coverage for Nonle Inside Radius Section is 91.50%. The limitations are 2.5 in. x 2.5 in, welded pads (5 locations) in scan area and generator support leg at 90,10 in. long. Examination from the inside diameter (ID) is not practical due to ALARA constraints.
Part C: Category B-J," Pressure Retaining Welds in Piping," RHR Loop A, Reactor Coolant A & B, and RTD Tak'e Off A& B Systems:
For RHR Loop A, the coverage for this Elbow to Valve (see Figure 2) is 71.55%.
Additional coverage can not be obtained due to valve geometry and welded lugs.
For Reactor Coolant A, the coverage for this Nonle to Elbow (see Figure 3) is 46.20%. Additional coverage can not be obtained due to OD counter bore and nonle contour.
For Reactor Coolant B, the coverage for this Elbow to Reactor Coolant Pump (see Figure 4) is 21.60% for circ. weld and the coverage is 73.20% for
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longitudinal seam. Additional coverage can not be obtained due to pump configuration and OD counter bore on pipe side.
For RTD Take Off A, the coverage for the Pipe to Elbow is 66.60%. Additional coverage can not be obtained due to base metal obstructed by a wek.> pipe hanger.
For RTD Take Off B, the coverage for the Elbow to Pipe welds is 86.60%.
Additional coverage can not be obtained due to a pipe support welded attachment.
Page 2 of 4
- NORTHERN STATES POWER INSERVICE INSPECTION PRAIRIE ISLAND UNIT 1,3RD INTERVAL EXAMINATION PLAN i
Part D: Category C-A, " Pressure Retaining Welds in Pressure Vessels," S/G and RHR Heat Exchanger 12:
For the S/G, the coverage for Shell to Transition (see Figure 5) is 54.30%.
Additional coverage can not be obtained due to restraint at lower toe of weld and 3 in. x 3 in. pads in four locations at upper toe of weld.
For RHR Heat Exchanger 12, the coverage for Head to Shell (see Figure 6) is 74.00%. Additional coverage can not be obtained due to an inlet and outlet nozzle reinforcing ring and vessel support interference.
Part E: Category C-B, " Pressure Retaining Nozzle Welds in Vessels," S/G; the coverage for Nozzle to Shell is 82.55%. Additional coverage can not be obtained due to Feedwater Ring and Ring support welded to the ID of shell and l
OD nozzle configuration.
Part F: Category C-C, " Integral Attachments for Vessels, Piping, Pumps, and Valves," Main Steam A & B System; the coverage for the limited integral attachments are 47.00%,22.50%, and 84.60%. Additional coverage can not be obtained for the 47.00% and 22.50% due to proximity of the adjacent support / lugs which limits the inspection to one direction, and due to the tack welded cover plate for 84.60%, respectively.
Part G: Category C-F-2, " Pressure Retaining Welds in Carbon Steel or Low Alloy Steel Piping," Main Steam B System; the coverage for Elbow to Pipe is 71.70% for the surface exam and 75.00% for the volumetric exam. Additional coverage can not be obtained due to a welded pipe support 1 in. from cire. weld covering the longseam.
Additional Means of Establishing Integritv:
In addition, hydrostatic tests are performed during regular inspection intervals to ensure the piping system is capable of maintaining pressure integrity. System integrity is monitored continuously during normel operation by many direct and indirect methods, e.g., containment radiation monitoring, containment air monitoring, containment leakage detection and monitoring, containment temperature monitoring, etc.
Alterriate Examination:
None, the nature of the limitations have been noted on the ISI examination reports and included in the ISI Outage Summary Report, NSP will continue to document the limitations.
Page 3 of 4
' NORTHERN STATES POWER INSERVICE INSPECTION 1
PRAIRIE ISLAND UNIT 1,3RD INTERVAL EXAMINATION PLAN All inservice inspection at Prairie Island has been done to the greatest extent practical.
Limitations are due to design, geometry, and materials of construction of the components or ALARA concerns. NSP will continue to utilize the most current techniques available for future examinations.
. Anoroval Status:
Not yet approved, submitted March 16,1998 i
4 Page 4 of 4
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