ML030770448

Inservice Inspection Relief Requests Regarding Inner Radius Examination of Class I Reactor Pressure Vessel Nozzles

ML030770448
Person / Time
Site:	Nine Mile Point
Issue date:	03/06/2003
From:	Montgomery B Constellation Energy Group
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
NMPIL 1716, TAC MB7722, TAC MB7723
Download: ML030770448 (11)

Text

"P.O.Box 63 Lycoming, New York 13093 0 Constellation Energy Group Nine Mile Point Nuclear Station March 6, 2003 NMP1L 1716 U.S. Nuclear Regulatory Commission Attn.: Document Control Desk Washington, DC 20555

SUBJECT:

Nine Mile Point Unit Nos. 1 and 2 Docket Nos. 50-220 and 50-410 Facility Operating License Nos. DPR-63 and NPF-69 Inservice Inspection Relief Requests Regarding Inner Radius Examination of Class 1Reactor Pressure Vessel Nozzles TAC Nos. MB7722 and MB7723 Gentlemen:

Pursuant to 10 CFR 50.55a(a)(3)(i), Nine Mile Point Nuclear Station, LLC (NMPNS) hereby requests approval of an alternative to certain inservice inspection requirements of the 1989 Edition (no Addenda) of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code,Section XI, for Nine Mile Point Units 1 and 2 (NM!P1 and NNMP2). The description and justification for this request are provided in the enclosed inservice inspection relief requests ISI-21A-1 for NMP1 (Enclosure 1) and ISI 21B-1 for NMP2 (Enclosure 2).

NMIPNS is requesting an alternative to the ASME Code,Section XI requirements to perform a volumetric examination of the inner radius of reactor pressure vessel (RPV) nozzles welded with full penetration welds. In lieu of volumetric examination, the relief requests propose performance of a visual examination for RPV nozzles where the plant configuration allows visual examination on essentially 100 percent of the nozzle inner radius. This proposed alternative is being requested in accordance with 10 CFR 50.55a(a)(3)(i) on the basis that it provides an acceptable level of quality and safety.

Similar requests have previously been approved for the Detroit Edison Company's Fermi Unit 2 by NRC letter dated October 5, 2001 (TAC No. MB2166), and for the Tennessee Valley Authority's Browns Ferry Nuclear Plant Unit 2 by NRC letter dated October 7, 2002 (TAC No. MB4880).

fýO4-7

Page 2 NMP1L 1716 NMPNS requests NRC approval of relief requests ISI-21A-1 and ISI-21B-1 by September 12, 2003, to support development of the NMP2 inservice inspection plans for refueling outage number 9, scheduled for the Spring of 2004.

Very truly yours, B c* . otgome Ma ager Engineeriig Service BSM/DEV/jm

Enclosures:

1. NMP1 Relief Request ISI-2 IA-1

2. NMP2 Relief Request ISI-21B-1 cc: Mr. H.J. Miller, NRC Regional Administrator, Region I Mr. G.K. Hunegs, NRC Senior Resident Inspector Mr. P.S. Tam, Senior Project Manager, NRR (2 copies)

ENCLOSURE 1 Nine Mile Point Unit 1 Third Inservice Inspection Interval Relief Request ISI-21A-1

NINE MILE POINT UNIT 1 THIRD INSERVICE INSPECTION INTERVAL RELIEF REQUEST ISI-21A-1 A. COMPONENT IDENTIFICATION System: Reactor Pressure Vessel Class: Quality Group A, ASME Code Class 1 Component

Description:

Reactor Pressure Vessel (RPV) Nozzle Inner Radius Sections ASME Section Xl, Table IWB-2500-1, Examination Category B-D, Full Penetration Welds of Nozzles In Vessels - Inspection Program B EXAM ITEM SYSTEM IDENTIFICATION NOZZLE DESCRIPTION NUMBER OF NUMBER NOZZLES B3 100 01.0 - Main Steam System Outlet Nozzles 2 32 0 - Reactor Recirculation System Loop Suction (Outlet) Nozzles 5 36 0 - RPV Closure Head Safety Valve Nozzles 18 37.0 - Reactor Head Vent Vent Nozzle 1 B. ASME SECTION XI EXAMINATION REQUIREMENTS ASME Section XI, 1989 Edition, No Addenda, Table IWB-2500-1 for Examination Category B-D, requires a volumetric examination of nozzle inner radius section of all Reactor Pressure Vessel Nozzles welded with full penetration welds as shown in Figures IWB-2500-7(a) through (d).

C. RELIEF REQUESTED Pursuant to 10 CFR 50.55a(a)(3)(i), Nine Mile Point Nuclear Station, LLC (NMPNS) requests relief from the ASME Section XI requirements to perform volumetric examinations as described in Section B above for the individual nozzle inner radius sections identified on Attachment 1 to this request.

D. BASIS FOR RELIEF All nozzle forgings were nondestructively examined during fabrication and have previously been examined using inservice ultrasonic examination techniques specific to each nozzle configuration. No indication of fabrication defects or service related cracking has been detected by these examinations.

Attachment 1 provides the date of the last ultrasonic examination for each nozzle included within this request.

Nozzle inner radius examinations are the only non-welded areas requiring examination on the RPV. This requirement was deterministically made early in the development of the ASME Section XI Code, and applied to 100% of nozzles welded with full penetration welds.

Fatigue cracking is the only applicable degradation mechanism for the nozzle inner radius region. For all nozzles other than Feedwater, there is no significant thermal cycling during operation. Therefore, from a risk perspective there is no need to perform volumetric examination on any nozzles other than Feedwater and operational Control Rod Drive (CRD) returns. No service related cracking has ever been discovered in any Boiling Water Reactor (BWR) fleet plant nozzles other than on Feedwater or operational CRD returns.

IS121A-1-1 OF ISI21A-1-4

NINE MILE POINT UNIT 1 THIRD INSERVICE INSPECTION INTERVAL RELIEF REQUEST ISI-21A-1 The four (4) Feedwater nozzle inner radius sections will continue to be examined with ultrasonic techniques developed and qualified in accordance with GE-NE-523-A71-0594-A, Revision 1, approved by the NRC under TAC No. MA6787.

NMPNS is proposing to implement a visual examination alternative similar to the inspection alternative proposed in ASME Section Xl Code Case N-648-1. The visual examination will cover the same examination surface as specified for the volumetric examination.

NMPNS believes that application of a visual examination alternative for the listed nozzle inner radius regions ensures an acceptable level of quality and safety. Performing a visual examination has the additional benefit of reducing personnel radiation exposure, consistent with the site ALARA program.

E. ALTERNATIVE EXAMINATIONS As an alternate to the volumetric examination requirements defined in Section B above, NMPNS proposes to perform a visual examination, to include essentially 100% of the surface M-N as shown in Figures IWB-2500-7(a) through (d) in lieu of the volumetric examination required by Table IWB-2500-1, Examination Category B-D, Item B3.1 00, of Inspection Program "B", for inservice examination of reactor vessel nozzle inner radius sections listed on Attachment 1.

NMPNS proposes the direct visual (VT-1) type examination of the RPV Closure Head Safety Relief Valve (N7A thru N7L) and the RPV Head Vent (N8) nozzle inner radius sections. For the direct visual examinations, the resolution sensitivity will be established using a 1-mil (.001 inch) wire standard, or equivalent for the detection of cracking.

For the remaining nozzle inner radius sections, NMPNS proposes using the remote visual examination Enhanced VT-1 (i.e, EVT-1) type examinations of the Main Steam Outlet (N3A and N3B) and Reactor Recirculation Outlet (N1A thru N1 E) nozzle inner radius sections as described in the Electric Power Research Institute (EPRI) Technical Report entitled "TR-105696-R4 (BWRVIP-03) Revision 4: BWR Vessel and Internals Project, Reactor Pressure Vessel and Internals Examination Guidelines." The resolution sensitivity for remote in-vessel examinations will be established using the one half (1 / 2) mil wire standard as described in that report.

If crack-like surface flaws are detected by visual examination, the flaws will be characterized in accordance with Table IWB-3512-1. When applying Table IWB-3512-1 criteria, the crack depth will be assumed to be equal to one-half the measured crack length. Flaws exceeding the acceptance criteria of Table IWB-3512-1 are unacceptable for continued service unless the reactor vessel meets the requirements of IWB-3142.2, IWB-3142.3 or IWB-3142.4.

F. IMPLEMENTATION SCHEDULE Relief is requested for the remainder of the Third Ten-Year Inservice Inspection Interval (12/26/99 12/25/09).

G. ATTACHMENTS Attachment 1 List of Applicable RPV Nozzle Inner Radius Sections ISI21A-1-2 OF ISI21A-1-4

NINE MILE POINT UNIT 1 THIRD INSERVICE INSPECTION INTERVAL RELIEF REQUEST ISI-21A-1 ATTACHMENT 1 LIST OF APPLICABLE RPV NOZZLE INNER RADIUS SECTIONS NOZZLE REACTOR VESSEL RPV NOZZLE INNER SIZE LAST UT UT ESTIMATED VISUAL DESIGNATION NOZZLE DESCRIPTION RADIUS EXAMINATION EXAM COVERAGE VISUAL LIMITATIONS IDENTIFICATION COVERAGE NUMBER N3A 01.0 Main Steam 01-WD-001-IR 24.0" 12/88 74% 100% None Nozzle I N3B 01.0 Main Steam 01-WD-033-IR 240" 05/99 100% 100% None Nozzle N1A 32.0 Reactor 32-WD-001-IR 28.0" 03/95 95% 100% None Recirculation Loop Outlet Nozzle NIB 32.0 Reactor 32-WD-044-IR 28 0" 04/99 100% 100% None Recirculation Loop Outlet Nozzle I NiC 32 0 Reactor 32-WD-084-IR 28 0" 04/99 100% 100% None Recirculation Loop Outlet Nozzle NII 32 0 Reactor 32-WD-124-IR 28 0" 04/99 100% 100% None Recirculation Loop Outlet Nozzle NIE 32.0 Reactor 32-WD-166-IR 28 0" 03/95 95% 100% None Recirculation Loop Outlet Nozzle N7A 36 0 1 Reactor Vessel 36-WD-012-IR 6.0" 03/01 90 2% 100% None Head Safety Valve Nozzle N7B 36.0 ' Reactor Vessel 36-WD-014-IR 6 0" 03/01 902% 100% None Head Safety Valve Nozzle N7C 36 0 1 Reactor Vessel 36-WD-016-IR 6.0" 03/01 902% 100% None Head Safety Valve Nozzle N7D 36 0 1 Reactor Vessel 36-WD-018-IR 6 0" 03/01 90.2% 100% None Head Safety Valve Nozzle I N7E 36 0 1 Reactor Vessel 36-WD-020-IR 6 0" 03/01 902% 100% None Head Blind Flange Nozzle N7F 36 0 1 Reactor Vessel 36-WD-022-IR 6 0" 11/88 64% 100% None Head Safety Valve Nozzle N7G 36.0 1 Reactor Vessel 36-WD-024-1R 6 0" 03/97 100% 100% None Head Safety Valve Nozzle N7H 36 0 1 Reactor Vessel 36-WD-026-IR 60" 06/88 64% 100% None Head Safety Valve Nozzle ISI21A-1-3 OF ISI21A-1-4

NINE MILE POINT UNIT 1 THIRD INSERVICE INSPECTION INTERVAL RELIEF REQUEST ISI-21A-1 ATTACHMENT 1 LIST OF APPLICABLE RPV NOZZLE INNER RADIUS SECTIONS NOZZLE REACTOR VESSEL RPV NOZZLE INNER SIZE LAST UT UT ESTIMATED VISUAL DESIGNATION NOZZLE DESCRIPTION RADIUS EXAMINATION EXAM COVERAGE VISUAL LIMITATIONS IDENTIFICATION COVERAGE NUMBER N7J 36 0 1 Reactor Vessel 36-WD-028-IR 6 0" 03/97 100% 100% None Head Safety Valve Nozzle N7K 36.0 1 Reactor Vessel 36-WD-030-IR 6 0" 03/97 100% 100% None Head Blind Flange Nozzle N7L 36 0 1 Reactor Vessel 36-WD-1073-IR 60" 03/97 100% 100% None Head Blind Flange Nozzle N7M 36 0 1 Reactor Vessel 36-WD-032-IR 60" 03/97 100% 100% None Head Safety Valve Nozzle N7N 36 0 1 Reactor Vessel 36-WD-034-IR 60" 03/97 100% 100% None Head Blind Flange Nozzle N7P 36 0 1 Reactor Vessel 36-WD-036-IR 6.0" 03/97 100% 100% None Head Blind Flange Nozzle N7R 36 0 1 Reactor Vessel 36-WD-038-IR 6.0" 03/97 100% 100% None Head Blind Flange Nozzle N7S 36.0 1 Reactor Vessel 36-WD-040-IR 6 0" 03/97 100% 100% None Head Blind Flange Nozzle N7T 36 0 1 Reactor Vessel 36-WD-042-IR 6.0" 03/97 100% 100% None Head Blind Flange Nozzle N7U 36.0 ' Reactor Vessel 36-WD-044-IR 6 0" 03/97 100% 100% None Head Blind Flange Nozzle N8 37.0 1 Reactor Head 37-WD-001-IR 40" 03/97 100% 100% None Vent Nozzle There are a total of nineteen (19) Closure Head Nozzles Five (5) of the nineteen (19) nozzle inner radius sections were examined by the volumetric (ultrasonic) examination method during refueling outage sixteen (RFO-1 6), in the First Inservice Inspection Period. The remaining fourteen (14) nozzle inner radius sections are scheduled for completion in the Second and Third Inservice Inspection periods and are included within this request for relief as identified above.

ISI21A-1-4 OF ISI21A-1-4

ENCLOSURE 2 Nine Mile Point Unit 2 Second Inservice Inspection Interval Relief Request ISI-211B-1

NINE MILE POINT UNIT 2 SECOND INSERVICE INSPECTION INTERVAL RELIEF REQUEST ISI-21 B-1 A. COMPONENT IDENTIFICATION System: Reactor Pressure Vessel Class: ASME Code Class 1 Component

Description:

Reactor Pressure Vessel (RPV) Nozzle Inner Radius Sections ASME Section Xl, Table IWB-2500-1, Examination Category B-D, Full Penetration Welds of Nozzles In Vessels - Inspection Program B EXAM ITEM NOZZLE IDENTIFICATION NOZZLE DESCRIPTION NUMBER OF NUMBER NOZZLES B3.100 N1 - Reactor Recirculation Outlet Nozzles 2 N3 - Main Steam Outlet Nozzles 4 N7 - Closure Head RCIC Spray Nozzle 1 N8 - Reactor Head Vent Vent Nozzle 1 N1 8 - Closure Head Spare Nozzle 1 B. ASME SECTION XI EXAMINATION REQUIREMENTS ASME Section XI, 1989 Edition, No Addenda, Table IWB-2500-1 for Examination Category B-D, requires a volumetric examination of nozzle inner radius section of all Reactor Pressure Vessel nozzles welded with full penetration welds as shown in Figures IWB-2500-7(a) through (d).

C. RELIEF REQUESTED Pursuant to 10 CFR 50.55a(a)(3)(i), Nine Mile Point Nuclear Station, LLC (NMPNS) requests relief from the ASME Section XI requirements to perform volumetric examinations as described in Section B above for the individual nozzle inner radius sections identified on Attachment 1 to this request.

D. BASIS FOR RELIEF All nozzle forgings were nondestructively examined during fabrication and have previously been examined using inservice ultrasonic examination techniques specific to each nozzle configuration. No indication of fabrication defects or service related cracking has been detected by these examinations.

Attachment 1 provides the date of the last ultrasonic examination for each nozzle included within this request.

Nozzle inner radius examinations are the only non-welded areas requiring examination on the RPV. This requirement was deterministically made early in the development of the ASME Section XI Code, and applied to 100% of nozzles welded with full penetration welds.

Fatigue cracking is the only applicable degradation mechanism for the nozzle inner radius region. For all nozzles other than Feedwater, there is no significant thermal cycling during operation. Therefore, from a risk perspective there is no need to perform volumetric examination on any nozzles other than Feedwater and operational Control Rod Drive (CRD) returns. No service related cracking has ever been discovered in any Boiling Water Reactor (BWR) fleet plant nozzles other than on Feedwater or operational CRD returns.

ISI21B-1-1 OF ISI21B-1-3

NINE MILE POINT UNIT 2 SECOND INSERVICE INSPECTION INTERVAL RELIEF REQUEST ISI-21 B-1 NMPNS is proposing to implement a visual examination alternative similar to the inspection alternative proposed in ASME Section XI Code Case N-648-1. The visual examination will cover the same examination surface as specified for the volumetric examination.

NMPNS believes that application of a visual examination alternative for the listed nozzle inner radius regions ensures an acceptable level of quality and safety. Performing a visual examination has the additional benefit of reducing personnel radiation exposure, consistent with the site ALARA program.

E. ALTERNATIVE EXAMINATIONS As an alternate to the volumetric examination requirements defined in Section B above, NMPNS proposes to perform a visual examination, to include essentially 100% of the surface M-N as shown in Figures IWB-2500-7(a) through (d) in lieu of the volumetric examination required by Table IWB-2500-1, Examination Category B-D, Item B3.1 00, of Inspection Program "B", for inservice examination of reactor vessel nozzle inner radius sections listed in Attachment 1.

NMPNS proposes the direct visual (VT-1) type examination of the RPV Closure Head RCIC Spray (N7),

Closure Head Vent (NB) and Closure Head Spare (N1 8) nozzle inner radius sections. For the direct visual examinations, the resolution sensitivity will be established using a 1-mil (.001 inch) wire standard, or equivalent for the detection of cracking.

For the remaining nozzle inner radius sections, NMPNS proposes using the remote visual examination Enhanced VT-1 (i.e., EVT-1) type examinations of the Main Steam Outlet (N3A thru N3D) and Reactor Recirculation Outlet (N1A and N1B) nozzle inner radius sections as described in the Electric Power Research Institute (EPRI) Technical Report entitled "TR-105696-R4 (BWRVIP-03) Revision 4: BWR Vessel and Internals Project, Reactor Pressure Vessel and Internals Examination Guidelines." The resolution sensitivity for remote in-vessel examinations will be established using the one half (1 / 2) mil wire standard as described in that report.

Ifcrack-like surface flaws are detected by visual examination, the flaws will be characterized in accordance with Table IWB-3512-1. When applying Table IWB-3512-1 criteria, the crack depth will be assumed to be equal to one-half the measured crack length. Flaws exceeding the acceptance criteria of Table IWB-3512-1 are unacceptable for continued service unless the reactor vessel meets the requirements of IWB-3142.2, IWB-3142.3 or IWB-3142.4.

F. IMPLEMENTATION SCHEDULE Relief is requested for the remainder of the Second Ten-Year Inservice Inspection Interval (04/05/98 04/04/08).

G. ATTACHMENTS Attachment 1 List of Applicable RPV Nozzle Inner Radius Sections ISI21B-1-2 OF ISI21B-1-3

NINE MILE POINT UNIT 2 SECOND INSERVICE INSPECTION INTERVAL RELIEF REQUEST ISI-21 B-1 ATTACHMENT 1 LIST OF APPLICABLE RPV NOZZLE INNER RADIUS SECTIONS NOZZLE REACTOR VESSEL RPV NOZZLE INNER SIZE LAST UT UT ESTIMATED VISUAL DESIGNATION NOZZLE DESCRIPTION RADIUS EXAMINATION EXAM COVERAGE VISUAL LIMITATIONS IDENTIFICATION COVERAGE NUMBER NIA Reactor Recirculation 2RPV-ACC 24.0" 3/92 100% 100% None Outlet Nozzle N1B Reactor Recirculation 2RPV-ACF 24.0" 3/92 100% 100% None Outlet Nozzle N3A Main Steam Outlet 2RPV-ACR 26 0" 5/98 100% 100% None Nozzle N3B Main Steam Outlet 2RPV-ACU 26.0" 3/02 100% 100% None Nozzle N3C Main Steam Outlet 2RPV-ACY 26 0" 4/95 100% 100% None Nozzle N3D Main Steam Outlet 2RPV-AEB 26 0* 4/95 100% 100% None Nozzle N7 Closure Head RCIC 2RPV-AFX 6 0" 5/95 85% 100% None Spray Nozzle N8 Closure Head Vent 2RPV-AGA 4.0" 11/93 100% 100% None Nozzle N18 Closure Head Spare 2RPV-AGS 6 0" 10/96 85% 100% None Nozzle 4- + 9 I 4- 4-

1. 4- 4 I 4- 4 4-4- + I 1 1- 1 1-4- + 9 4 4- 4 4-A. 4. I 1 + 4 +

4- + I 4 4- 4 4-

+ I I t t 4- 9 4 4- 4 4-A. -I- I I + 4 .9-4- + I I 4- t t 4- + 4 4 4- 1 7 4- + I I 4- t t

j. +/- ____________ I ____ L ______ .L _______ L ________ J.

IS121 B-1 -3 OF IS121 B-1 -3