ML022190331
| ML022190331 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 07/29/2002 |
| From: | Mccollum W Duke Energy Corp |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| Download: ML022190331 (10) | |
Text
Pk Duke OF awr".Oconee Duke Energy Nuclear Station 7800 Rochester Highway Seneca, SC 29672 (864) 885-3107 OFFICE W R. McCollum, Jr.
(864) 885-3564 FAx Vice Presidnt July 29, 2002 U.S. Nuclear Regulatory Commission Document Control Desk Washington, DC 20555
Subject:
Duke Energy Oconee Nuclear Station, Unit 1 Docket Nos. 50-269 Third Ten Year Inservice Inspection Interval Request for Relief No.02-004 Pursuant to 10 CFR 50.55a(g) (5) (iii), attached is a Request for Relief from the requirement to examine 100% of the volume specified by the ASME Boiler and Pressure Vessel Code,Section XI, 1989 Edition with no Addenda (as modified by Code Case N-460). This request is to allow Duke Energy to take credit for twenty (20) limited ultrasonic examinations on Reactor Vessel welds described in the attached request. During examination of the subject Unit 1 welds, the ultrasonic examination coverage did not meet the 90% examination requirements of Code Case N-460. The obtainable volume coverage for each weld examination is indicated on the attached request. Achievement of greater examination coverage for these welds is impractical due to piping/valve geometry, interferences, and existing examination technology. Therefore, Duke Energy requests that the NRC grant relief as authorized under 10 CFR 50.55a(g) (6) (i).
If there are any questions or further information is needed you may contact R. P. Todd at (864) 885-3418.
Very truly yours, W. R. McCollumh Site Vice Pres' ent Attachment
U. S. Nuclear Regulatory Commission July 29, 2002 Page 2 xc w/att: L. A. Reyes, Regional Administrator U.S. Nuclear Regulatory Commission, Region II Atlanta Federal Center 61 Forsyth St., SWW, Suite 23T85 Atlanta, GA 30303 L. N. Olshan, Project Manager, Section 1 Project Directorate II Division of Licensing Project Management Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, DC 20555-0001 xc(w/o attch):
M. C. Shannon Senior NRC Resident Inspector Oconee Nuclear Station Mr. Virgil Autrey Division of Radioactive Waste Management Bureau of Land and Waste Management SC Dept. of Health & Environmental Control 2600 Bull St.
Columbia, SC 29201
Relief Request 02-004 Page 1 of 7 Proposed Relief in Accordance with 10 CFR 50.55a(g)(5)(iii)
Inservice Inspection Impracticality Duke Energy Corporation Oconee Nuclear Station - Unit 1 (EOC-20)
Third 10-Year Interval - Inservice Inspection Plan ASME Section XI Code - 1989 Edition with No Addenda I. II. & III. IV. V. VI. VII.
Limited System Code Requirement from Which Basis for Relief Alternate Justification Implementation Area/Weld Component for Which Relief is Requested: Examinations or for Granting Schedule I.D. Relief is Requested: 100% Exam Volume Coverage Testing Relief Number Area or Weld to be Exam Category Examined Item No.
Fig. No.
Limitation Percentage 1-RPV-WR34 NC System Exam Category B-A See Paragraph See Paragraph See Paragraph See Paragraph Reactor Vessel Item No. BO1.011.005 "A" "G" "H4" "K" Lower Shell to Lower Fig. IWB-2500-1 Head Ring 36% Volume Coverage Circumferential Weld 1-RPV-WR35 NC System Exam Category B-A See Paragraph See Paragraph See Paragraph See Paragraph Reactor Vessel Item No. B01.021.003 "A" "G" "H" "K" Lower Head Cap to Fig. IWB-2500-3 Lower Head Ring 42% Volume Coverage Circumferential Weld 1-RPV-WR13 NC System Exam Category B-D See Paragraph See Paragraph See Paragraph See Paragraph Reactor Vessel Item No. B03.090.001 "B" "G".I "K" Outlet Nozzle-to-Vessel Fig. IWB-2500-7(a)
Weld @ 900 82% Volume Coverage (UT from vessel I.D.)
1-RPV-WR13 NC System Exam Category B-D See Paragraph See Paragraph See Paragraph See Paragraph Reactor Vessel Item No. B03.090.001A "B".G" "I", "K" Outlet Nozzle-to-Vessel Fig. IWB-2500-7(a)
Weld @ 900 82% Volume Coverage (UT from nozzle I.D.)
Relief Request 02-004 Page 2 of 7 I. II. & III. IV. V. VI. VII.
Limited System / Code Requirement from Which Basis for Relief Alternate Justification Implementation Area/Weld Component for Which Relief is Requested: Examinations or for Granting Schedule I.D. Relief is Requested: 100% Exam Volume Coverage Testing Relief Number Area or Weld to be Exam Category Examined Item No.
Fig. No.
Limitation Percentage I-RPV- NC System Exam Category B-D See Paragraph See Paragraph See Paragraph See Paragraph WR13A Reactor Vessel Item No. B03.090.002 "B" "G" .K" "I Outlet Nozzle-to-Vessel Fig. IWB-2500-7(a)
Weld @ 2700 82% Volume Coverage (UT from vessel I.D.)
1-RPV- NC System Exam Category B-D See Paragraph See Paragraph See Paragraph See Paragraph WR13A Reactor Vessel Item No. B03.090.002A "B".G" "I" "K" Outlet Nozzle-to-Vessel Fig. IWB-2500-7(a)
Weld @ 270' 82% Volume Coverage (UT from nozzle I.D.)
1-RPV-WR54 NC System Exam Category B-D See Paragraph See Paragraph See Paragraph See Paragraph Reactor Vessel Fig. IWB-2500-7(a) "C" "G" .. I,, "K" Core Flood Item No. B03.090.007 Nozzle-to-Vessel Weld (UT from vessel I.D.)
@ 00 81% Volume Coverage Item No. B03.090.007A (UT from nozzle ID) 0% Volume Coverage (not able to scan) 1-RPV- NC System Exam Category B-D See Paragraph See Paragraph See Paragraph See Paragraph WR54A Reactor Vessel Fig. IWB-2500-7(a) "C".G" "I".K" Core Flood Item No. B03.090.008 Nozzle-to-Vessel Weld (UT from vessel ID)
@ 1800 81% Volume Coverage Item No. B03.090.008A (UT from nozzle ID) 0% Volume Coverage (not able to scan)
I-RPV-WR54 NC System Exam Category B-D See Paragraph See Paragraph See Paragraph See Paragraph Reactor Vessel Item No. B03.100.007 "D".G" "I" "K" Core Flood Nozzle Fig. IWB-2500-7(a)
Inside Radius Section 52% Volume Coverage
@ 00
Relief Request 02-004 Page 3 of 7 I. II. & III. IV. V. VI. VII.
Limited System / Code Requirement from Which Basis for Relief Alternate Justification Implementation Area/Weld Component for Which Relief is Requested: Examinations or for Granting Schedule I.D. Relief is Requested: 100% Exam Volume Coverage Testing Relief Number Area or Weld to be Exam Category Examined Item No.
Fig. No.
Limitation Percentage I-RPV- NC System Exam Category B-D See Paragraph See Paragraph See Paragraph See Paragraph WR54A Reactor Vessel Item No. B03.100.008 "D" "G" ',I "K" Core Flood Nozzle Fig. IWB-2500-7(a)
Inside Radius Section 52% Volume Coverage
@ 1800 1-RPV-WR53 NC System Exam Category B-F See Paragraph See Paragraph See Paragraph See Paragraph Reactor Vessel Item No. B05.010.001 "E" "G".I" "K" Core Flood Fig. IWB-2500-8(c)
Nozzle-to-Safe-End 86% Volume Coverage Butt Weld @ 00 (UT from nozzle I.D. in lieu of PT from O.D.)
1-RPV-WR53 NC System Exam Category B-F See Paragraph See Paragraph See Paragraph See Paragraph Reactor Vessel Item No. B05.010.001A "E" "G" "I" "K" Core Flood Fig. IWB-2500-8(c)
Nozzle-to-Safe-End 86% Volume Coverage Butt Weld @ 00 (UT from nozzle side)
I-RPV-WR53 NC System Exam Category B-F See Paragraph See Paragraph See Paragraph See Paragraph Reactor Vessel Item No. B05.010.001B "E" "." "I .K" Core Flood Fig. IWB-2500-8(c)
Nozzle-to-Safe-End 86% Volume Coverage Butt Weld @ 00 (UT from safe-end side) 1-RPV- NC System Exam Category B-F See Paragraph See Paragraph See Paragraph See Paragraph WR53A Reactor Vessel Item No. B05.010.002 "E" "G" .K" "I Core Flood Fig. IWB-2500-8(c)
Nozzle-to-Safe-End 81% Volume Coverage Butt Weld @ 1800 (UT from nozzle I.D. in lieu of PT from O.D.)
1-RPV- NC System Exam Category B-F See Paragraph See Paragraph See Paragraph See Paragraph WR53A Reactor Vessel Item No. B05.010.002A "E" "." "" "K" Core Flood Fig. IWB-2500-8(c)
Nozzle-to-Safe-End 81% Volume Coverage Butt Weld @ 1800 (UT from nozzle side)
Relief Request 02-004 Page 4 of 7 I. II. & III. IV. V. VI. VII.
Limited System / Code Requirement from Which Basis for Relief Alternate Justification Implementation Area/Weld Component for Which Relief is Requested: Examinations or for Granting Schedule I.D. Relief is Requested: 100% Exam Volume Coverage Testing Relief Number Area or Weld to be Exam Category Examined Item No.
Fig. No.
Limitation Percentage I-RPV- NC System Exam Category B-F See Paragraph See Paragraph See Paragraph See Paragraph WR53A Reactor Vessel Item No. B05.010.002B "E" "G" "K" "I
Core Flood Fig. IWB-2500-8(c)
Nozzle-to-Safe-End 81% Volume Coverage Butt Weld @ 1800 (UT from safe-end side) 1-53A NC System Exam Category B-J See Paragraph See Paragraph See Paragraph See Paragraph 43L Reactor Vessel Item No. B09.011.090 "F" "G" "I "K" Core Flood Fig. IWB-2500-8(c)
Safe-End to Pipe 76% Volume Coverage Circumferential Weld
@ 00 1-53A NC System Exam Category B-J See Paragraph See Paragraph See Paragraph See Paragraph 43L Reactor Vessel Item No. B09.011.090A "F' "G" .. r "K" Core Flood Fig. IWB-2500-8(c)
Safe-End to Pipe 76% Volume Coverage Circumferential Weld (UT from nozzle I.D. in lieu of
@ 00 PT from O.D.)
1-53A-01-IL NC System Exam Category B-J See Paragraph See Paragraph See Paragraph See Paragraph Reactor Vessel Item No. B09.01 1.100 "F' "G" "I", "K" Core Flood Fig. IWB-2500-8(c)
Safe-End to Pipe 83% Volume Coverage Circumferential Weld
@ 1800 1-53A-01-IL NC System Exam Category B-J See Paragraph See Paragraph See Paragraph See Paragraph Reactor Vessel Item No. B09.01 1. 1OOA "F' "G" "I "K" Core Flood Fig. IWB-2500-8(c)
Safe-End to Pipe 83% Volume Coverage Circumferential Weld (UT from nozzle I.D. in lieu of
@ 1800 PT from O.D.)
Note: See Attachment A for a drawing on all the welds listed above.
Relief Request 02-004 Page 5 of 7 IV. Basis for Relief (See Attachment A for area/weld locations.)
Paragraph A:
During the ultrasonic examination of welds 1-RPV-WR34 and 1-RPV-WR35, 100% coverage of the required examination volume could not be obtained. The examination coverage was limited to 36% and 42% respectively.
Limitations were caused by the core guide lugs & flow stabilizers for WR34 and incore nozzles & flow stabilizers for WR35 that restrict the scanning surface. The percentage of coverage reported represents the aggregate coverage.
In order to achieve more coverage the core guide lugs, incore nozzles and flow stabilizers would have to be moved to allow greater access for scanning, which is impractical.
Paragraph B:
During the ultrasonic examination of welds 1-RPV-WR13 and I-RPV-WR13A, 100% coverage of the required examination volume could not be obtained. The examination coverage was limited to 82%. Limitations were caused by the outlet nozzle boss that restricts the scanning surface. The percentage of coverage reported represents the aggregate coverage. In order to achieve more coverage, the outlet nozzle boss would have to be moved to allow greater access for scanning, which is impractical.
Paragraph C:
During the ultrasonic examination of welds 1-RPV-WR54 and 1-RPV-WR54A, 100% coverage of the required examination volume could not be obtained. The examination coverage was limited to 81% of the required volume from one side of the weld. Limitations were caused by the flange taper and inlet nozzles that restrict the scanning surface. The percentage of coverage reported represents the aggregate coverage. In order to achieve more coverage, the inlet nozzles would have to be moved and the taper on the flange would have to be redesigned to allow greater access for scanning, which is impractical. In addition, because of the proximity of the flow restrictors no scanning was performed from the nozzle I.D. (0% examination coverage). In order to achieve more coverage, the flow restrictor would have to be moved to allow access for scanning, which is impractical.
Paragraph D:
During the ultrasonic examination of inside radius sections 1-RPV-WR54 and I-RPV-WR54A, 100% coverage of the required examination volume could not be obtained. The examination coverage was limited to 52%. Limitations were caused by the flow restrictor that prevents scanning the surface. The percentage of coverage reported represents the aggregate coverage. In order to achieve more coverage, the flow restrictor would have to be moved to allow greater access for scanning, which is impractical.
Paragraph E:
During the ultrasonic examination of welds I-RPV-WR53 and I-RPV-WR53A, 100% coverage of the required examination volume could not be obtained. The examination coverage was limited to 86% and 81%, respectively.
Limitations were caused by air at the top of nozzle that prevents the transducer from making contact for scanning the surface. The percentage of coverage reported represents the aggregate coverage. In order to achieve more coverage, the reactor coolant pumps would have to be in operation to permit reactor coolant flow which would remove the air at the top of the nozzle, which is impractical.
Paragraph F:
During the ultrasonic examination of welds 1-53A-02-43L and 1-53A-01-IL, 100% coverage of the required examination volume could not be obtained. The examination coverage was limited to 76% and 83%, respectively.
Limitations were caused by air at the top of nozzle that prevents the transducer from making contact for scanning the surface. The percentage of coverage reported represents the aggregate coverage. In order to achieve more coverage, the reactor coolant pumps would have to be in operation to permit reactor coolant flow which would remove the air at the top of the nozzle, which is impractical.
Relief Request 02-004 Page 6 of 7 V. Alternate Examinations or Testing Paragraph G:
The scheduled 10-year code examination was performed on the referenced area/weld and it resulted in the noted limited coverage of the required ultrasonic volume. No additional examinations are planned for the area/weld during the current inspection interval.
VI. Justification for Granting Relief Paragraph H:
Ultrasonic examination of welds for item numbers B01.011 and B01.021 were conducted using personnel, equipment and procedures qualified in accordance with ASME Section XI, Appendix VIII, 1995 Edition with the 1996 Addenda as administered through the Performance Demonstration Initiative (PDI) Program. Although 100%
coverage of the examination volume could not be achieved, the amount of coverage obtained for this examination provides an acceptable level of quality and integrity. (See Paragraph J for additional justification.)
Paragraph I:
Ultrasonic examination of areas/welds for item numbers B03.090, B03. 100, B05.010 and B09.011 were conducted using personnel, equipment and procedures qualified in accordance with ASME Section XI, Appendix I, 1989 Edition with no Addenda. Although 100% coverage of the examination volume could not be achieved, the amount of coverage obtained for this examination provides an acceptable level of quality and integrity. (See Paragraph J for additional justification.)
Paragraph J:
Duke Energy will use pressure testing and VT-2 visual examination to compliment the limited examination coverage.
The Code requires (reference Table IWB-2500-1, item numbers B 15.010 and B 15.050) that a system leakage test be performed after each refueling outage for Class 1. Additionally a system hydrostatic test (reference Table IWB 2500-1, item numbers B 15.011 and B 15.051) is required once during each 10-year inspection interval. These tests require a VT-2 visual examination for evidence of leakage. This testing provides adequate assurance of pressure boundary integrity.
Duke Energy will use VT-3 visual examination to compliment the limited examination coverage. The Code requires (reference Table IWB-2500-1, item number B 13.010) that a VT-3 examination be performed after the first refueling outage and subsequent refueling outages at approximately 3 year periods. During the first and second periods of an interval a VT-3 examination is performed on areas above and below the reactor core that are made accessible for examination by removal of components during normal refueling outages. During the third period of an interval the VT-3 examination is performed on all of the reactor vessel interior surfaces at the same time that the automated UT exams are performed on the reactor vessel welds. This examination provides adequate assurance of pressure boundary integrity.
In addition to the above Code required examinations (volumetric, pressure test, and VT-3), there are other activities which provide a high level of confidence that, in the unlikely case that leakage did occur through these welds, it would be detected and isolated. Specifically, leakage from these welds would be detected by monitoring of the Reactor Coolant System (RCS), which is performed once each shift under procedure PT/i,2,3/A/0600/10, "RCS Leakage". This RCS leakage monitoring is a requirement of Technical Specification 3.4.13, "Reactor Coolant System Leakage". Leakage is also evaluated in accordance with this Technical Specification. The leakage could also be detected through several other methods. One is the RCS mass balance calculation. A second is the Reactor Building air particulate monitor. This monitor is sensitive to low leak rates; the iodine monitor, gaseous monitor and area monitor are capable of detecting any fission products in the coolant and will make these monitors sensitive to
Relief Request 02-004 Page 7 of 7 coolant leakage. A third is the level indicator in the Reactor Building normal sump. A fourth is a loss of level in the Letdown Storage Tank.
Due to the design of the reactor vessel and location of the core guide lugs, flow stabilizers, outlet nozzle boss, flow restrictors and inlet nozzles and air in the top of some of the nozzles; it is not feasible to obtain the examination coverage required for all of the welds listed in this request for relief. Duke Energy has examined the welds/components referenced in this request to the maximum extent possible utilizing the latest in examination techniques and equipment. These welds were rigorously inspected by volumetric NDE methods during construction and verified to be free from unacceptable fabrication defects. Based on the portions and results of the required volumetric and visual examinations performed during this outage, it's our opinion that this combination of examinations provides a reasonable assurance of component integrity. Thus, an acceptable level of quality and safety will have been achieved and allowing relief from the aforementioned Code requirements will not endanger public health and safety.
VII.Implementation Schedule Paragraph K The scheduled third 10-year interval plan code examination was performed on the referenced area/weld resulting in limited volumetric coverage. No additional examinations are planned for the area/weld during the current inspection interval. The same area/weld may be examined again as part of the next (fourth) 10-year interval plan, depending on the applicable code year edition and addenda requirements adopted in the future.
VIII. Other Information The following individuals contributed to the development of this relief request:
James J. McArdle (NDE Level III Inspector) provided Sections II through V and part of Section VI.
B. W. Carney, Jr. (Oconee Engineering) provided part of Section VI.
Larry C. Keith (Oconee ISI Plan Manager) compiled the remaining sections.
Sponsored By: (OJV C. V,1tat, Date w_3 - 6 D Approved By: #ý - Date 7/43
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