SVPLTR 05-0018, Relief Request CR-28, Inservice Inspection Program Relief Regarding Reactor Pressure Vessel Longitudinal Shell Weld Examination Coverage for Third 10-Year Inservice Inspection Interval
| ML051260293 | |
| Person / Time | |
|---|---|
| Site: | Dresden  |
| Issue date: | 05/06/2005 |
| From: | Bost D Exelon Generation Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| SVPLTR: #05-0018 | |
| Download: ML051260293 (5) | |
Text
10 CFR 50.55a May 6, 2005 SVPLTR: #05-0018 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555-0001 Dresden Nuclear Power Station, Unit 3 Renewed Facility Operating License No. DPR-25 NRC Docket No. 50-249
Subject:
Relief Request CR-28, Inservice Inspection Program Relief Regarding Reactor Pressure Vessel Longitudinal Shell Weld Examination Coverage for Third 10-Year Inservice Inspection Interval
References:
- 1. Letter from G. Y. Suh (U. S. NRC) to C. M. Crane (Exelon Generation Company, LLC), "Dresden Nuclear Power Station, Units 2 and 3 and Quad Cities Nuclear Power Station, Units 1 and 2 - Authorization for Proposed Alternative Reactor Pressure Vessel Circumferential Shell Weld Examinations (TAC Nos. MC2190, MC2191, MC2192 and MC2193)," dated March 23, 2005
- 2. Letter from L. W. Rossbach (U. S. NRC) to O. D. Kingsley (Exelon Generation Company, LLC), Exemption from the Requirements of 10 CFR 50.55a(g)(6)(ii)(A)(2), Inservice Examination of the Reactor Pressure Vessel, dated September 28, 2001
- 3. Letter from A. J. Mendiola (U. S. NRC) to C. M. Crane (Exelon Generation Company, LLC), "Dresden Nuclear Power Station, Unit 2 - Relief Request CR-27 for Third 10-Year Inservice Inspection Interval (TAC No. MC3268),"
dated September 16, 2004 In accordance with 10 CFR 50.55a, Codes and standards, paragraphs (a)(3)(i) and (g)(6)(ii)(A)(5), Dresden Nuclear Power Station (DNPS) is requesting relief from American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code,Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components, and the augmented examinations specified in 10 CFR 50.55a(g)(6)(ii)(A)(2) on the basis that the proposed alternative provides an acceptable level of quality and safety.
In Reference 1, the NRC approved an alternative reactor pressure vessel (RPV) weld examination pursuant to the provisions of 10 CFR 50.55a paragraphs (a)(3)(i) and (g)(6)(ii)(A)(5) for DNPS Units 2 and 3. The alternative allows permanent deferral of requirements to perform a volumetric examination of RPV circumferential shell welds for the extended terms of the DNPS Units 2 and 3 renewed operating licenses. The approved alternative requires inspections of essentially 100 percent of all longitudinal welds, and
May 6, 2005 U. S. Nuclear Regulatory Commission Page 2 inspections of approximately 2 to 3 percent of the circumferential welds at their points of intersection with the longitudinal welds.
DNPS is submitting the attached relief request for those Unit 3 ASME Section XI RPV longitudinal shell weld examinations where the inspection coverage achieved was less than or equal to 90 percent. Specifically, this includes volumetric examination of RPV longitudinal shell welds examinations completed during the Third 10-Year Inservice Inspection Interval. The Third 10-Year Inservice Inspection Interval began on March 1, 1992, and ended on October 31, 2003.
RPV longitudinal shell weld examinations were completed during the Unit 3 refueling outage, which began on October 26, 2004, and was completed on December 7, 2004. Approval to delay RPV longitudinal shell weld examinations was provided by the NRC in Reference 2. The NRC has previously approved a similar relief for Dresden Nuclear Power Station, Unit 2 in Reference 3.
Should you have any questions concerning this letter, please contact Mr. Pedro Salas at (815) 416-2800.
Respectfully, Original Signed By Danny Bost Site Vice President Dresden Nuclear Power Station
Attachment:
10 CFR 50.55a Request Number CR-28 cc:
Regional Administrator - NRC Region III NRC Senior Resident Inspector - Dresden Nuclear Power Station
ATTACHMENT 10 CFR 50.55a Request Number CR-28 Relief Requested In Accordance with 10 CFR 50.55a(g)(6)(ii)(A)(5) and 10 CFR 50.55a(a)(3)(i)
Page 1 of 3 ASME Code Components Affected Components affected are American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Code),Section XI, Class 1 pressure retaining reactor pressure vessel (RPV) longitudinal shell welds, Examination Category B-A, Item No. B1.12. Components are listed in Table CR-28.1.
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Applicable Code Edition and Addenda===
The applicable ASME Code,Section XI, for Dresden Nuclear Power Station (DNPS), Unit 3 Third 10-Year Inservice Inspection Interval is the 1989 Edition.
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Applicable Code Requirement===
In accordance with the provisions of 10 CFR 50.55a, "Codes and standards," paragraphs (a)(3)(i) and (g)(6)(ii)(A)(5), Exelon Generation Company, LLC (EGC) requests relief for DNPS, Unit 3 from the requirements of the augmented examinations specified in 10 CFR 50.55a(g)(6)(ii)(A)(2), which was used as a substitute for the reactor vessel shell weld examination scheduled for the third Inspection Interval as allowed by 10 CFR 50.55a(g)(6)(ii)(A)(2).
Augmented RPV examinations specified in 10 CFR 50.55a(g)(6)(ii)(A)(2) are subject to the conditions specified in 10 CFR 50.55a(g)(6)(ii)(A)(4) where examination of the reactor vessel may be satisfied by an examination of essentially 100% of the reactor vessel shell welds.
Determination of Limits of Weld Volume Examination DNPS Unit 3 obtained Construction Permit CPPR-22 on October 14, 1966. The RPV was designed and fabricated before the examination requirements of ASME Section XI were formalized and published. Since this plant was not specifically designed to meet the requirements of ASME Section XI, full compliance is not feasible or practical within the limits of the current plant design.
The RPV is examined from the internal surface to the extent practical. Further examination from the inside surface is not practical without disassembly of vessel internal components. The exterior vessel surface is covered with permanent insulation located in close proximity to the RPV outside surface. The lower exterior vessel surface is also covered with a structural steel biological shield wall. Supplemental manual examinations from the outside surface are not practical due to the biological shield wall, insulation, and dose considerations.
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ATTACHMENT 10 CFR 50.55a Request Number CR-28 Relief Requested In Accordance with 10 CFR 50.55a(g)(6)(ii)(A)(5) and 10 CFR 50.55a(a)(3)(i)
Page 2 of 3 Proposed Alternative and Basis for Use Proposed Alternative In accordance with 10 CFR 50.55a(a)(3)(i), and (g)(6)(ii)(A)(5), EGC proposes the following alternate provisions for the subject weld examinations since the proposed alternative provides an acceptable level of quality and safety.
The examination requirements specified in 10 CFR 50.55a(g)(6)(ii)(A)(2) for the RPV longitudinal shell welds shall be performed, to the extent possible. When this examination is performed, welds are examined from inside surfaces of the RPV using an automated ultrasonic inspection system, which provides the best possible examination of the RPV longitudinal shell welds. Additionally, a VT-2 examination is performed on the RPV during the system leakage test per examination category B-P each refueling outage.
Basis For Use The RPV longitudinal shell welds are ultrasonically examined utilizing a Performance Demonstration Initiative (PDI) qualified automated ultrasonic inspection system meeting the requirements of ASME Section XI, Appendix VIII.
All components received examination(s) to the extent practical due to the limited or lack of access. The examinations conducted, confirmed satisfactory results evidencing no unacceptable flaws present, even though essentially 100% coverage was not attained.
Based on the above, with our earlier design, the underlying objectives of the code required volumetric examinations have been met. The examinations were completed to the extent practical and evidenced no unacceptable flaws present. Additionally, a VT-2 examination performed during the system leakage test per examination category B-P each refueling outage provides additional assurance that the structural integrity of the RPV is maintained.
Duration of Proposed Alternative Relief is requested for the Third 10-Year Inservice Inspection Interval of the Inservice Inspection Program for DNPS Unit 3.
Precedents The NRC has previously approved the following similar relief for Dresden Nuclear Power Station, Unit 2:
Letter from A. J. Mendiola (U. S. NRC) to C. M. Crane (Exelon Generation Company, LLC),
"Dresden Nuclear Power Station, Unit 2 - Relief Request CR-27 for Third 10-Year Inservice Inspection Interval (TAC No. MC3268)," dated September 16, 2004.
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ATTACHMENT 10 CFR 50.55a Request Number CR-28 Relief Requested In Accordance with 10 CFR 50.55a(g)(6)(ii)(A)(5) and 10 CFR 50.55a(a)(3)(i)
Page 3 of 3 TABLE CR-28.1 UNIT 3 RPV LONGITUDINAL SHELL WELDS Weld Identification Weld Description Relief Requested (Based on 90%
Coverage)
Condition Limiting Coverage Coverage Percent SC1A Shell Course 1 Weld at 77 deg.
Yes Jet Pump Diffuser 86 SC1B Shell Course 1 Weld at 197 deg.
Yes Jet Pump Diffuser, Core Shroud Repair Tie Rod 59 SC1C Shell Course 1 Weld at 317 deg.
Yes Jet Pump Diffuser 86 SC2A Shell Course 2 Weld at 22 deg.
Yes Jet Pump Riser Brace, Surveillance Specimen Support Bracket, Core Shroud Repair Tie Rod 57 SC2B Shell Course 2 Weld at 142 deg.
No Surface Condition at Circumferential Weld Intersection 98 SC2C Shell Course 2 Weld at 262 deg.
No Core Spray Piping 91 SC3A Shell Course 3 Weld at 77 deg.
Yes Core Spray and Feedwater Spargers, Core Spray Piping, Core Spray Repair Coupling 21 SC3B Shell Course 3 Weld at 197 deg.
Yes Core Spray and Feedwater Spargers, Core Shroud Repair Tie Rod 75 SC3C Shell Course 3 Weld at 250 deg.
Yes Core Spray Piping and Feedwater Sparger 80 SC3D Shell Course 3 Weld at 317 deg.
Yes Core Spray Piping and Feedwater Sparger 81 SC4A Shell Course 4 Weld at 99 deg.
No Nozzle N5A, Surface Condition at Reactor Flange Weld Intersection 97 SC4B Shell Course 4 Weld at 219 deg.
No Steam Dryer Support Bracket 91 SC4C Shell Course 4 Weld at 261 deg.
No None 100 SC4D Shell Course 4 Weld at 339 deg.
No None 100 Revision 0