ML031900071
| ML031900071 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 07/02/2003 |
| From: | Rosalyn Jones Duke Energy Corp, Duke Power Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| Download: ML031900071 (6) | |
Text
_ Duke R. A. JONES
~SPower. Vice President A Duke Energy Company Duke Power 29672 / Oconee Nuclear Site 7800 Rochester Highway Seneca, SC 29672 864 885 3158 864 885 3564 fax July 2, 2003 U.S. Nuclear Regulatory Commission Document Control Desk Washington, DC 20555
Subject:
Duke Energy Corporation Oconee Nuclear Station, Unit 3 - Docket No. 50-287 Third Ten-year Inservice Inspection Interval Request for Alternate (Relief Request No.03-004)
Duke Energy Corporation (Duke) hereby submits a Request for Altemate for the Oconee Unit 3 third 10-year reactor vessel examination per 10 CFR 50.55a(g)(5)(iii). This examination is planned for the next Unit 3 refueling outage scheduled for the fall of 2004.
The reactor pressure vessel nozzle inner radius sections must be volumetrically examined per ASME Section Xi, Appendix VilI, 1995 Edition through the 1996 Addenda. Duke uses the Performance Demonstration Initiative (PDI) to qualify vendor and in-house ultrasonic test personnel. However, the PDI test set for nozzle inner radius qualification from the inside surface does not adequately model the B&W 177 reactor vessel core flood nozzle. In addition, design features (permanently installed flow restrictors) prevent full coverage of the examination surface.
Therefore, Duke proposes to use a remote enhanced VT-1 (visual) examination as described in the attachment as an alternate to the specified volumetric examination, Approval is requested per 10 CFR 50.55a(a)(3)(i) because the alternative provides an acceptable level of quality and safety.
This request is similar to requests submitted by Florida Power and Light for Turkey Point Units 3 and 4, and Public Service Electric and Gas for Salem Units 1 and 2.
Please direct any questions to R. P.Todd at (864) 885-3418.
Ve yyours, nes, Site Vice President, Oconee Nuclear Station
Attachment:
ISI Relief Request 03-04 4 www. duke-energy. corn
U. S. Nuclear Regulatory Commission Page 2 July 2, 2003 xc wlatt: L. A. Reyes, Regional Administrator U.S. Nuclear Regulatory Commission, Region II Atlanta Federal Center 61 Forsyth St., SW, 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 att):
M. C. Shannon Senior NRC Resident Inspector Oconee Nuclear Station
Page of3 DUKE ENERGY CORPORATION Oconee Nuclear Station Unit 3 10-YEAR INTERVAL REQUEST FOR RELIEF NO.03-004 Pursuant to 10CFR50.55a (a) (3) (i), Duke Energy Corporation (Duke) proposes an alternative to the requirements of ASME Section XI, Ta-ble IWB-2500-1, Examination Category B-D, Item No. B3.100, Nozzle Inner Radius Sections.
I. System/Components for Which the Alternative Applies:
Category B-D Full Penetration Welds of Nozzles in Vessels Item Numbers:
B03.100.007 Core Flood Nozzle Inner Radius, 3-RPV-WR54 B03.100.008 Core Flood Nozzle Inner Radius, 3-RPV-WR54A II. Code Requirement: ASME Section XI, Table IWB-2500-1 Examina-tion Category B-D, Full Penetration Welds of Nozzles In Ves-sels, Volumetric Examination, Figure IWB-2500-7(b), Examina-tion Volume M-N-O-P of ASME Section XI, 1989 Edition with no addenda.
III. Code Requirement for Which the Alternative is Requested:
Relief is requested to perform a remote VT-1 examination in lieu of the volumetric examination required in Table IWB-2500-1 Examination Category B-D, Full Penetration Welds of Nozzles In Vessels, Item No. B3.100, Nozzle Inner Radius Sections.
IV. Basis for Relief: Pursuant to 10CFR50.55a (a) (3) (i), Relief is requested to perform a remote VT-1 examination in lieu of the required volumetric examination on the basis that the pro-posed alternative provides an acceptable level of quality and safety.
V. Alternative Method for Ultrasonic Examination:
Duke Energy Corporation proposes to use a remote VT-1 examina-tion of surface M-N as shown in ASME Section XI, Figure IWB-2500-7(b) of the 1989 Edition with no addenda. The remote T-1 equipment will have sufficient magnification and sensitivity to resolve a 0.001 inch wire in lieu of the sensitivity re-quired for an ultrasonic examination. The examination results will be evaluated in accordance with ASME Section XI, IWB-
Page 2 of 3 3140, 1989 Edition with no addenda. Crack-like surface flaws exceeding the acceptance criteria of Table IWB-3512-1 are un-acceptable for continued service unless the reactor vessel meets the requirements of IWB-3142.2, IWB-2142.3 or IWB-3142.4. When applying Table IWB-3512-1 criteria, crack depth will be assumed to be equal to one-half of the measured crack length.
It must be noted that because of the permanent attachment of flow guides in the core flood nozzles, full coverage of the examination surface will not be possible (See attached draw-ing). During ultrasonic examinations performed in the first and second Inspection Intervals 50% coverage of the examina-tion volume was achieved. Relief from the volumetric examina-tion coverage requirements was sought under Oconee Relief Re-quest 94-01 submitted by letter of April 4, 1994 and supple-ments dated April 14, 1994, and March 16, 1995. Relief was granted by NRC in a letter of June 12, 1995. The VT-i exami-nation coverage is expected be no less than the ultrasonic ex-amination.
VI. Justification for the Granting of Relief:
The Core Flood (CF) piping from each of the two RV Nozzles-to the first upstream check valve is Duke Class A (ASME Code Class 1). The Oconee Unit 3 core flood nozzles in the reactor pressure vessel are made of forged SA-508 ferritic steel with stainless steel cladding. The piping is stagnant during all normal and upset operating conditions. These piping compo-nents have been analyzed to Section III of the 1983 Edition of the ASME B&PV Code. The results of this analysis show that the CF piping components meet all Code requirements and allow-able stresses.
The primary degradation mode in reactor vessel nozzles is thermal fatigue. The thermal fatigue cumulative usage factor (including the effects of all applicable thermal cycling events and thermal stratification) is less than 0.033 (the Code limit is 1.00) for all Duke Class A piping components.
Therefore, thermal fatigue degradation is unlikely.
Thermal fatigue in this application may produce hairline sur-face indications along the nozzle's inner radius section. The intent of the Code requirement for an ultrasonic examination is that the examination would detect such surface indications.
Page 3of 3 These nozzles were nondestructively examined during fabrica-tion and subsequently examined inservice twice using the ul-trasonic method, There were no examination findings and no flaws were detected in any of the Oconee reactor pressure ves-sel nozzles.
According to a NRC memorandum['], the NRC staff has indicated that an ultrasonic examination could be replaced by an en-hanced VT-1 visual examination for the proposed nozzle inspec-tions. Subsequent to that memorandum, the NRC granted re-quests similar to this request to Florida Power and Light for Turkey Point Units 3 & 4 (submitted May 6, 2002 and approved August 15, 2002) and Public Service Electric and Gas for Salem Units 1 & 2 (submitted February 11, 2002 and approved March 21, 2002).
Like Florida Power and Light and Public Service Electric and Gas, Duke proposes to use high magnification cameras to give 1-mil resolution capability for the remote VT-1 examination of the accessible portion of the nozzle inner radius section sur-face area. With this resolution, it is highly likely that Duke would detect and disposition flaws using the allowable flaw length criteria in Table IWB-3512-1 of the ASME Code,Section XI, for the disposition of any linear flaws. There-fore, the proposed alternative provides reasonable assurance of structural integrity.
VII. Implementation: Duke Energy will perform the remote VT-1 ex-amination of the core flood nozzle inner radii in conjunction with the ONS 3rd 10 year reactor pressure vessel examination.
The ONS 3rd 10 year interval started 7-15-1994, and ends 12-16-2004.
Sponsored By i. C Date &2f-403 Approved By .Date /2/
'NRC Memorandum from K.R. Wichman to W.H. Bateman dated May 25, 2000;
Subject:
The Third Meeting with the Industry to Discuss the Elimination of RPV Inner Radius Inspection (ADAMS Accession No. ML003718630).