NG-07-0368, Relief Request for 3rd Period Limited Weld Examinations
| ML071510076 | |
| Person / Time | |
|---|---|
| Site: | Duane Arnold  |
| Issue date: | 05/18/2007 |
| From: | Vanmiddlesworth G Duane Arnold |
| To: | Document Control Desk, NRC/NRR/ADRO |
| References | |
| IR-04-062, IR-06-103, NG-07-0368, TAC MC2374, TAC MC2375 | |
| Download: ML071510076 (7) | |
Text
FPL Energy Duane Arnold, LLC 3277 DAEC Road Palo, Iowa 52324 FPL E May 18, 2007 NG-07-0368 10 CFR 50.55a U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D.C. 20555-0001 Duane Arnold Energy Center Docket 50-331 License No. DPR-49 Relief Request for 3rd Period Limited Weld Examinations Pursuant to 10 CFR 50.55a(g)(5)(iii), FPL Energy Duane Arnold, LLC, (hereafter, FPL Energy Duane Arnold) hereby requtests NRC approval of the enclosed relief request from IWB-2500 and IWC-2500 to allow performance of limited examinations of various welds.. ,This relief isýrequested for the Third Ten year Interval of the Inservice Inspection Program for the Duane Arnold Energy Center (DAEC), which ended on. October 31, 2006.
FPL Energy Duane Arnold requests approval of this request by the end of October 2007.
This letter contains no new commitments nor revises any previous commitments.
If you have any questions, please contact Steve Catron at (319) 851-7234.
Gary Van Middlesworth Site Vice President, Duane Arnold Energy Center FPL Energy Duane Arnold, LLC .
Enclosure.-.*, ,..: , * ,
cc: Administrator, Region Ill,IUSNRC,
Project Manager, .
Senior Resident Inspector, DAEC,,USNRC -.
Enclosure to NG-07-0368 Page 1 of 6 10 CFR 50.55a REQUEST IN ACCORDANCE WITH 10 CFR 50.55a(g)(5)(iii)
INSERVICE INSPECTION IMPRACTICALITY Component Identification Code Class: Class 1 and 2
References:
ASME Code,Section XI, Subarticle IWB-2500 ASME Code,Section XI, Subarticle IWC-2500 Table IWB-2500-1 Table IWC-2500-1 Code Case N-613-1 Examination Categories: B-D, B-J, C-A Item Numbers: B3.90, B3.100, B9.11, and C1.20
==
Description:==
Nozzle-to-Vessel Welds and a pressure retaining pipe weld.
Component Numbers: See Table A for Component Identification
Applicable Code Edition and Addenda
l l ASME Section XI 1989 Edition, No Addenda Code Requirement Section XI (1989 Edition, no addenda), Subarticle IWB-2500 states, in part "Components shall be examined and tested as specified in Table IWB-2500-1."
Subarticle IWC-2500 states, in part, "Components shall be examined and tested as specified in Table IWC-2500-1." Table IWB-2500-1 Category B-D, Item B3.90 and Category B-J, Item B9.11 require a volumetric examination of applicable Class 1 pressure retaining-welds, which includes essentially.100% of weld length once during the ten year interval. Table IWB-2500-1 Category B-D, Item B3.100 requires examining nozzle inner radius volume region as defined by Figures IWB-2500-7(a-d).Section XI (1989 Edition), Table IWC-2500-1 Category C-A, Items C1.20 requires a volumetric examination of applicable Class 2 pressure retaining welds, which includes essentially 100% of weld length once during the ten year interval.
Code Case N-460 permits a reduction in examination coverage of Class 1 and Class 2 welds provided the coverage reduction is less than 10%. The Duane Arnold Energy Center (DAEC) has adopted Code Case N-460 in the Inservice Inspection (ISI) Program Plan, as permitted by USNRC Regulatory Guide 1.147, Revision 14.
In lieu of the examination volume depicted in Figure IWB-2500-7(b), the United States Nuclear Regulatory Commission (NRC) has authorized the DAEC to use the alternative examination volume requirements of Code Case N-613-1 (Reference 1) for the Nozzle-to-Vessel Shell welds listed in this request.
Enclosure to NG-07-0368 Page 2 of 6
Reason for Request
The DAEC construction permit was issued in 1970 and the operating license was issued in 1974. The reactor vessel was designed and installed to ASME Section III, 1965 Edition, with the Summer 1967 Addenda. The parameters for accessibility for Inservice Inspection were not requirements at that time, and therefore not necessarily factored into component and system configurations, thereby creating conditions where ASME Section Xl Code required examination coverage of Class 1 and 2 welds cannot be obtained.
10 CFR 50.55a recognizes the limitations to in-service inspection of components in accordance with Section XI of the ASME Code that are imposed due to early plants' design and construction, as follows:
10 CFR 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.
10 CFR 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 provisions and pre-service 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.
10 CFR 50.55a(g)(5)(iii): If the licensee has determined that conformance with certain code requirements is impractical for its facility, the licensee shall notify the Commission and submit, as specified in § 50.4, information to support' the determinations.
Enclosure to NG-07-0368 Page 3 of 6 Table A - Limited Examination Weld List Examination Item Component Period Code Description Category Number Number Examined Coverage MSC-DO01 3 79.38% Main Steam Nozzle to Vessel Weld MSD-D001 3 79.38% Main VeslWdSteam Nozzle to Vessel Weld Feedwater Nozzle to Vessel 3 87.67% Wel FWB-DO01 Weld Feedwater Nozzle to Vessel FWC-DO01 3 87.67% Feed Weld RRG-D001 3 80.44% tRecirculation eslWl Riser Nozzle to Vessel Weld Jet Pump Instrumentation Nozzle to Vessel Weld LCA-DO01 3 63.71% Standby Liquid Control Nozzle to Vessel Weld Vessel Instrumentation Nozzle to Vessel Weld LCA-DO01 Standby Liquid Control B-D B3.100 INNER 3 71.69% Inter-Radius RAD B-J B9.1 1 *RBB-J001 2 35.56% Recirculation System Weld Residual Heat Removal C-A C1.20 HEA-CA-05 2 78.07% Heat Exchanger Head to Shell weld
- Note that this weld examination was performed prior to the implementation of the Risk Informed Program.
Nozzle to Vessel Welds -
MSC-DO01, MSD-DO01, FWB-DO01, FWC-DO01, FWD-DOO1, RRG-DO01, JPB-D001, LCA-DO01, and VIB-DO01 The nozzle-to-vessel welds are accessible from the vessel side, but examination cannot be performed from the nozzle side because of forging curvature. In addition to component configuration, certain nozzle-to-vessel weld examinations are further limited by reactor pressure vessel design obstructions.
The Nondestructive Examination (NDE) procedures used at DAEC incorporate examination techniques qualified under Appendix VIII of the ASME Section XI Code by the Performance Demonstration Initiative (PDI) for examination of the subject nozzle-to-vessel shell welds. The Electric Power Research Institute (EPRI) computer modeling report (Reference 2) was generated to assist DAEC in
Enclosure to NG-07-0368 Page 4 of 6 developing and qualifying Ultrasonic Test (UT) examination techniques for the nozzle inner corner regions and nozzle-to-vessel shell welds. The examinations were performed using a manual contact method from the nozzle outside blend radius and vessel shell surfaces as discussed in the EPRI modeling report and as stated in DAEC procedures. The UT scanning methodology modeled in the EPRI modeling report was applicable to the coverage for the inner corner regions and for the inner 15 percent volume of the nozzle-to-vessel shell welds when scanning parallel to the weld. The examination of the remaining outer 85 percent volume of the nozzle-to-vessel shell welds was based on a separate PDI qualified technique and procedure which did not require use of the EPRI computer modeling report to validate.
Standby Liquid Control Nozzle Inner Radius Examination LCA-DO01 The proximity of the vessel skirt weld obstructed access for performing the Standby Liquid Control Nozzle Inner Radius examination. The NDE procedures used at DAEC incorporate examination techniques qualified under Appendix VIII of the ASME Section XI Code by the PDI for examination of the subject nozzle inner radius. The EPRI computer modeling report (Reference 2) was generated to assist DAEC in developing and qualifying UT examination techniques for the nozzle inner corner regions. The examinations were performed using a manual contact method from the nozzle outside blend radius and vessel shell surfaces as discussed in the EPRI modeling report and as stated in DAEC procedures. The UT scanning methodology modeled in the EPRI report was applicable to the coverage for the inner corner regions volume of the shell nozzle when scanning parallel to the weld.
The examination of the remaining outer 85 percent volume of the nozzle inner radius was based on a separate PDI qualified technique and procedure which did not require use of the EPRI computer model to validate.
Recirculation System Class I Weld RBB-JO01 This weld is the branch connection of the recirculation bypass line to the recirculation discharge line. The weldolet side is not accessible for scanning due to geometry.
Residual Heat Removal Heat Exchanger Head to Shell Weld HEA-CA-05 The heat exchanger tie-down brackets obstruct access to the head-to shell weld, limiting examination coverage.
Enclosure to NG-07-0368 Page 5 of 6 Proposed Alternative and Basis for Use In accordance with 10 CFR 50.55a(g)(5)(iii), relief is requested for the components listed in Table A on the basis that the required examination coverage of "essentially 100 percent" is impractical due to physical obstructions and the limitations imposed by design, geometry, and materials of construction. DAEC performed qualified examinations that achieved the maximum practical amount of coverage obtainable within the limitations imposed by the design of the components.
Additionally, for the Class 1 examination Category B-P components, VT-2 examinations are performed on the subject components of the Reactor Coolant Pressure Boundary during system pressure tests on a refueling outage frequency.
Those examinations were completed during the 2007 refueling outage and no evidence of leakage was identified for these components.
The UT examinations of the nozzle to vessel welds and the one nozzle inner radius examination which were limited in coverage involved the remaining outer 85 percent of the required volume when scanning parallel to the weld, and the examination volume required when scanning normal to the weld. The examination volume described in the EPRI computer modeling report for DAEC is not applicable to the UT examination limitations included in the requested relief. That volume was applicable to the coverage for the inner corner regions and for the inner 15 percent volume of the nozzle-to-vessel shell welds when scanning parallel to the weld.
Based on the above, with due consideration of the earlier plant 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. VT-2 examinations performed on the subject Class 1 components during system pressure testing each refueling outage (in accordance with examination Category B-P) provide continued assurance that the structural integrity of the subject components is maintained. Additionally, the DAEC Water Chemistry Program and inerted primary containment environment provide added measures of protection for the component materials.
Duration of Proposed Alternative Relief is requested for the Third Ten year Interval of the Inservice Inspection Program for the DAEC, which ended on October 31, 2006.
Precedents
- 1) NRC Letter dated October 18, 1999, "Safety Evaluation of Third 10-Year Interval Inservice Inspection Program Plan Requests for Relief, Duane Arnold Energy Center (TAC No. MA4151)" (specifically for relief request NDE-R028)
Enclosure to NG-07-0368 Page 6 of 6
- 2) NRC Letter dated March 23, 1998, "Safety Evaluation of Third 10-Year Interval Inservice Inspection Program Plan Requests for Relief, Duane Arnold Energy Center (TAC No. M95412)," (specifically for relief requests NDE-R006, NDE-R007, NDE-R008, NDE-R009, and NDE-RO10)
References
- 1) NRC Letter dated October 6, 2004, "Duane Arnold Energy Center and Monticello Nuclear Generating Plant re: Request for Authorization to Utilize Code Case N-613-1, TAC Nos. MC2374 and MC2375" (ML042380089)
- 2) EPRI Modeling Reports, IR-2004-62 for N2, N3, N4, N8, N10, N11, IR-2006-103 is for N1, N5, and IR-2006-231 for N6, N7, N9