View source for SVPLTR 04-0025, Relief Request CR-27, Inservice Inspection Program Relief Regarding Reactor Pressure Vessel Longitudinal Shell Weld Examination Coverage for Third 10- Year Inservice Inspection Interval

{{Adams
| number = ML041320453
| issue date = 05/04/2004
| title = Relief Request CR-27, Inservice Inspection Program Relief Regarding Reactor Pressure Vessel Longitudinal Shell Weld Examination Coverage for Third 10- Year Inservice Inspection Interval
| author name = Bost D
| author affiliation = Exelon Generation Co, LLC
| addressee name = 
| addressee affiliation = NRC/Document Control Desk, NRC/NRR
| docket = 05000237
| license number = DPR-019
| contact person = 
| case reference number = SVPLTR #04-0025
| document type = Letter type:SVPLTR
| page count = 6
| project = 
| stage = Request
}}

=Text=
{{#Wiki_filter:Exelkn Exelon Generation Company, LLC      www.exeloncorp.com                                        Nuclear Dresden Nuclear Power Station 6500 North Dresden Road Morris, IL 60450-9765 10 CFR 50.55a May 4, 2004 SVPLTR: #04-0025 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555-0001 Dresden Nuclear Power Station, Unit 2 Facility Operating License Nos. DPR-19 NRC Docket Nos. 50-237

==Subject:==
Relief Request CR-27, Inservice Inspection Program Relief Regarding Reactor Pressure Vessel Longitudinal Shell Weld Examination Coverage for Third 10-Year Inservice Inspection Interval

==Reference:==
: 1) Letter from A. J. Mendiola (U. S. NRC) to 0. D. Kingsley (Exelon Generation Company, LLC), "Dresden - Authorization for Proposed Alternative Reactor Pressure Vessel Circumferential Weld Examinations (TAC Nos. MA6228 and MA6229)," dated February 25, 2000
: 2) Letter from L. W. Rossbach (U. S. NRC) to 0. 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 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 Xl, "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 remaining terms of the DNPS Units 2 and 3 operating licenses. The approved alternative requires inspections of essentially 100 percent of all longitudinal welds, and inspections of approximately 2 to 3 percent of the circumferential welds at their points of intersection with the longitudinal welds.

May 4, 2004 U.S. Nuclear Regulatory Commission Page 2 DNPS is submitting the attached relief request for those Unit 2 ASME Section Xl RPV longitudinal shell weld examinations where the inspection coverage achieved was less than or equal to 90%. 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 September 30, 2003.
RPV longitudinal shell weld examinations were completed during the Unit 2 refueling outage, which began on October 14, 2003, and was completed on November 11, 2003. Approval to delay RPV longitudinal shell weld examinations was provided by the NRC in Reference 2.
Should you have any questions concerning his letter, please contact Mr. Jeff Hansen at (815) 416-2800.
Respectfully, Danny Bo Site Vice" resident Dresden Nuclear Power Station

==Attachment:==
10 CFR 50.55a Request Number CR-27 cc:     Regional Administrator - NRC Region III NRC Senior Resident Inspector - Dresden Nuclear Power Station

ATTACHMENT 10 CFR 50.55a Request Number CR-27 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 4 ASME Code Components Affected Components affected are American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Code), Section Xl, 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-27.1.

===Applicable Code Edition and Addenda===
The applicable ASME Code, Section Xl, for Dresden Nuclear Power Station (DNPS), Unit 2 Third 10-Year Inservice Inspection Interval is the 1989 Edition.

===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 2 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 2 obtained Construction Permit CPPR-18 on January 10, 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 Xl, 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-27 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 4 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 Xl, Appendix Vil.
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 2.
Precedents The NRC has previously approved similar relief for Quad Cities Nuclear Power Station, Units 1 and 2. The NRC granted relief for Quad Cities Nuclear Power Station in Reference 1.
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ATTACHMENT 10 CFR 50.55a Request Number CR-27 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 4 References
: 1. Letter from S. A. Richards (U. S. NRC) to 0. D. Kingsley (Exelon Generation Company, LLC), "Alternative to 10 CFR 50.55a(g)(6)(ii)(A) Augmented Reactor Pressure Vessel Examination for Quad Cities Nuclear Plant, Units 1 and 2 (TAC Nos. M97370 and M99659)," dated October 23, 1998.
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ATTACHMENT 10 CFR 50.55a Request Number CR-27 Relief Requested In Accordance with 10 CFR 50.55a(g)(6)(ii)(A)(5) and 10 CFR 50.55a(a)(3)(i)
Page 4 of 4 TABLE CR-27.1 UNIT 2 RPV LONGITUDINAL SHELL WELDS Weld               Weld              Relief                Condition Limiting Coverage                 Coverage Identification     Description         Requested                                                            Percent (Based on 90%
Coverage)     l SC1A          Shell Course 1          Yes                  Jet Pump Diffuser and Baffle Plate              88 Weld at 77 deg.
SC1B          Shell Course 1          Yes                     Core Shroud Repair Tie Rod                   35 Weld at 110 deg.
SCi C         Shell Course 1          Yes             Jet Pump Diffuser, Diffuser Support Pads,            38 Weld at 197 deg.                      Recirculation Nozzle, Core Shroud Repair Tie Rod SC1 D         Shell Course 1          Yes                  Jet Pump Diffuser and Baffle Plate              86 Weld at 317 deg.
SC2A          Shell Course 2          Yes           Jet Pump Riser Brace, Surveillance Specimen            72 Weld at 98 deg.                       Support Bracket, Core Shroud Repair Tie Rod SC2B          Shell Course 2          Yes           Jet Pump Riser Brace, Surveillance Specimen            76 Weld at 218 deg.                                      Support Bracket SC2C          Shell Course 2          Yes           Jet Pump Riser Brace, Surveillance Specimen            83 Weld at 250 deg.                                      Support Bracket SC2D          Shell Course 2           No                                None                             100 Weld at 338 deg.
SC3A          Shell Course 3          Yes                Core Spray and Feedwater Spargers                 76 Weld at 77 deg.
SC3B          Shell Course 3          Yes                Core Spray and Feedwater Spargers                 77 Weld at 197 deg.
SC3C          Shell Course 3          Yes                Core Spray and Feedwater Spargers                 78 Weld at 296 deg.                                                    .
SC3D          Shell Course 3          Yes                Core Spray and Feedwater Spargers                 83 Weld at 317 deg.
SC3E          Shell Course 3          Yes          Core Spray and Feedwater Spargers, Guide Rod            83 Weld at 353 deg.
SC4A          Shell Course 4           No                                None                             100 Weld at 99 deg.
SC4B          Shell Course 4           No                                None                             100 Weld at 150 deg.
SC4C          Shell Course 4          Yes                    Steam Dryer Support Bracket                   88 Weld at 210 deg.
SC4D          Shell Course 4           No                                None                             100 l Weld at 330 deg.                                                                       l SC4E          Shell Course 4           No                                None                             100 Weld at 339 deg.
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