ML022490301

Clarification & Changes to Requests for Relief 2-ISI-16 & 2-ISI-17 ASME,Section XI, Inservice Inspection Program

ML022490301
Person / Time
Site:	Browns Ferry
Issue date:	09/05/2002
From:	Abney T Tennessee Valley Authority
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
Download: ML022490301 (9)

Text

September 5, 2002 10 CFR 50.55a(a)(3)

U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Mail Stop: OWFN P1-35 Washington, D.C. 20555-0001 Gentlemen:

In the Matter of ) Docket No. 50-260 Tennessee Valley Authority )

BROWNS FERRY NUCLEAR PLANT (BFN) - UNIT 2 - AMERICAN SOCIETY OF MECHANICAL ENGINEERS (ASME) SECTION XI, INSERVICE INSPECTION (ISI) PROGRAM - CLARIFICATION AND CHANGES TO REQUESTS FOR RELIEF 2-ISI-16 AND 2-ISI-17 This letter provides clarification and changes to BFN Unit 2 requests for relief 2-ISI-16 and 2-ISI-17. TVA submitted, by letter dated April 23, 2002, requests for relief 2-ISI-16 and 2-ISI-17 for the Unit 2 ASME Section XI Inservice Inspection Program for the Browns Ferry Nuclear Plant.

Requests for relief 2-ISI-16 and 2-ISI-17 provide alternative requirements for the examination of the BFN Unit 2 reactor pressure vessel nozzles using a direct enhanced visual examination for the RPV head nozzles and a remote enhanced visual examination for the RPV nozzles (except feedwater nozzles) rather than the Code prescribed volumetric examination. The feedwater nozzles will continue to receive a volumetric examination. During its review of the BFN requests for relief, the NRC staff identified questions regarding TVAs proposed alternate examinations.

TVA and the NRC staff held teleconferences on June 17, and August 21, 2002, to discuss the NRC questions. As a result of those teleconferences, TVA is providing clarification and changes to its proposed alternative examinations stated in requests for relief 2-ISI-16 and 2-ISI-17. The enclosure to this letter lists the six NRC questions and provides the corresponding TVA response.

U.S. Nuclear Regulatory Commission Page 2 September 5, 2002 TVA seeks review of these requests for relief by September 30, 2002, to support resource planning for the Unit 2 Cycle 12 (Spring 2003) refueling outage.

If you have any questions, please contact me at (256) 729-2636.

Sincerely, original signed by T. E. Abney Manager of Licensing and Industry Affairs Enclosure cc (Enclosure):

(Via NRC Electronic Distribution)

Mr. Paul E. Fredrickson, Branch Chief U.S. Nuclear Regulatory Commission Region II Sam Nunn Atlanta Federal Center 61 Forsyth Street, SW, Suite 23T85 Atlanta, Georgia 30303-8931 NRC Resident Inspector Browns Ferry Nuclear Plant P.O. Box 149 Athens, Alabama 35611 Mr. Kahtan N. Jabbour, Senior Project Manager U.S. Nuclear Regulatory Commission One White Flint, North (MS 08G9) 11555 Rockville Pike Rockville, Maryland 20852-2739

U.S. Nuclear Regulatory Commission Page 3 September 5, 2002 DTL:JWD:BAB Enclosure cc (Enclosure):

A. S. Bhatnagar, PAB 1E-BFN M. J. Burzynski, BR 4X-C R. G. Jones, POB 2C-BFN A. L. Ladd, PEC-2A-BFN J. E Maddox, LP 6A-C D. C. Olcsvary, LP 6A-C C. M. Root, PAB 1G-BFN J. R. Rupert, LP 6A-C K. W. Singer, LP 6A-C E. J. Vigluicci, ET 11A-K R. E. Wiggall, PEC 2A-BFN NSRB Support, LP 5M-C EDMS-K s:\lic\submit\subs\U2-ISI-16 &-17 RAI.doc

ENCLOSURE TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT (BFN)

UNIT 2 AMERICAN SOCIETY OF MECHANICAL ENGINEERS (ASME) SECTION XI, INSERVICE INSPECTION (ISI) PROGRAM (THIRD TEN-YEAR INSPECTION INTERVAL)

REQUEST FOR RELIEF 2-ISI-16 AND 2-ISI-17 RESPONSE TO NRC QUESTIONS TVA submitted BFN requests for relief 2-ISI-16 and 2 -ISI-17 allowing the use of an enhanced visual examination for the reactor pressure vessel (RPV) in lieu of the ASME Section XI Code specified ultrasonic examination. With regards to these relief requests, the staff has the following questions/requests:

NRC Question No. 1 Provide a list/table of the nozzles, nozzle sizes, percent coverage from prior examinations, and when (month/year) the examinations occurred.

TVA Response See table below. Note that enhanced visual examination coverage percentages are estimates only.

Ultrasonic Examination Enhanced Visual Examination Date Actual Nozzle Size Examined Report No.  % Estimated %

Coverage Coverage N1A-IR 28 10/94 BF2-1002 100 100 N1B-IR 28 10/97 R-207A 100 100 N2A-IR 12 11/97 R-208A 100 50 N2B-IR 12 10/94 BF2-1022 100 50 N2C-IR 12 11/97 R-209A 100 50 N2D-IR 12 04/01 R-126A 100 50 N2E-IR 12 04/01 R-127A 100 50 N2F-IR 12 10/94 BF2-1024 100 50 N2G-IR 12 10/97 R-210A 100 50 N2H-IR 12 10/97 R-211A 100 50 N2J-IR 12 10/94 BF2-1026 100 50 N2K-IR 12 04/01 R-128A 100 50 N3A-IR 26 04/01 R-129A 100 100 N3B-IR 26 11/97 R-212A 100 100 N3C-IR 26 04/01 R-130A 100 100 N3D-IR 26 10/94 BF2-1004 100 100 N5A-IR 10 10/94 BF2-1028 100 40 N5B-IR 10 10/97 R-215A 100 40 N6A-IR 6 02/93 R-0063 100 100 N6B-IR 6 4/99 R-331 100 100 N7-IR 4 4/99 R-335 100 100 N8A-IR 4 10/94 BF2-1011 73 60 N8B-IR 4 04/01 R-133A 100 60

• N9-IR 4 10/97 R-216A 100 100

• Non-operational/capped NRC Question No. 2 Discuss the effects on coverage as a result of changing the examination method from UT to VT.

TVA Response For 2-ISI-17, the visual examination coverage will be approximately the percentages shown in the table above. As stated in the relief request, limitations are due to the reactor internal piping configuration preventing placement of the camera in all positions necessary for 100 percent coverage.

E-2

Visual examination of the inner radius region is limited because reactor internal piping configuration prevents placement of the camera in all positions necessary to examine the M-N surface over the full circumference. (See TVA response to NRC Question number six for specific description of the obstructions.)

The requirement for inner radius examinations was deterministically made early in the development of ASME Section XI, and applied to 100 percent of nozzles welded with full penetration welds. Fatigue cracking is the only applicable degradation mechanism for the nozzle inner radius region. For all nozzles, except feedwater and operational CRD, there is no significant thermal cycling during operation. In addition, no service related cracking has ever been discovered in any of the BWR plant nozzles other than feedwater and operational CRD returns.

Reference:

White Paper ISI-99-26, Technical Basis For Elimination of Reactor Vessel Nozzle Inner Radius Inspections.

NRC Question No. 3 Describe the direct/enhanced visual examination systems and the resolution sensitivity that will be used during the examinations. Discuss how the 1-mil resolution sensitivity will be demonstrated for each system.

TVA Response The direct enhanced visual examination of the reactor pressure vessel (RPV) head nozzles (N6A, N6B, and N7) inner radius sections, will be performed in accordance with ASME Section XI VT-1 requirements. Resolution sensitivity for the direct enhanced visual examination will be demonstrated utilizing a vision test chart containing a 1-mil wire at a maximum distance of 2 feet, in lieu of the 0.044 inch character required by IWA-2210 of ASME Section XI 1995 Edition, 1996 Addenda.

Remote in-vessel enhanced visual examinations will be performed for all other nozzles (except feedwater) with approved procedures requiring that the resolution sensitivity be established using a 1-mil wire standard. This is consistent with that used for IVVI (reactor pressure vessel internal examinations) intended to detect cracking.

Note: IVVI examinations typically utilize a Sensitivity, Resolution and Contrast Standard (SRCS) which is fabricated with a surface texture representative of the surface to be E-3

examined or the actual surface to be examined may be used.

A target (1-mil wire) is superimposed over the SRCS or surface to be examined. Equipment resolution and sensitivity is demonstrated prior to performing examinations. Resolution and sensitivity of the examination equipment and technique is considered adequate when the system is capable of discerning the required target.

Listed below are the RPV nozzles within the scope of this request for relief and the type of enhanced visual (direct or remote) examination to be performed.

Nozzle Size Exam Type Nozzle Size Exam Type N1A-IR 28 Enhanced N6A-IR 6 Enhanced Remote Direct N1B-IR 28 Enhanced N6B-IR 6 Enhanced Remote Direct N2A-IR 12 Enhanced N7-IR 4 Enhanced Remote Direct N2B-IR 12 Enhanced N8A-IR 4 Enhanced Remote Remote N2C-IR 12 Enhanced N8B-IR 4 Enhanced Remote Remote N2D-IR 12 Enhanced *N9-IR 4 Enhanced Remote Remote N2E-IR 12 Enhanced N3A-IR 26 Enhanced Remote Remote N2F-IR 12 Enhanced N3B-IR 26 Enhanced Remote Remote N2G-IR 12 Enhanced N3C-IR 26 Enhanced Remote Remote N2H-IR 12 Enhanced N3D-IR 26 Enhanced Remote Remote N2J-IR 12 Enhanced N5A-IR 10 Enhanced Remote Remote N2K-IR 12 Enhanced N5B-IR 10 Enhanced Remote Remote

• Non-operational/capped Note: The RPV feedwater nozzles are not within the scope of this request for relief and will continue to receive a volumetric examination in accordance with ASME Section XI, Table IWB-2500.

E-4

NRC Question No. 4 The bases for these relief requests is ASME Code Case N-648-1, which references Table-IWB 3510-3 as the allowable linear crack acceptance criteria. This appears to be in conflict with the acceptance criteria in Table IWB-3512-1 for the inner radius.

Discuss using Table IWB-3510-3 as the acceptance criteria, and include any data supporting your position.

TVA Response TVA will perform enhanced visual examinations for the following RPV nozzles inner radius sections as an alternative to the ASME Section XI Code required volumetric examinations. (Specific nozzles are listed in TVAs response to NRC question number 3.)

• RPV head nozzles (3 nozzles) will be examined using a direct enhanced visual (VT-1) examination, with sufficient lighting, capable of a 1-mil resolution at a maximum distance of 2 feet.

• The remaining RPV nozzles (21 nozzles), will be examined using a remote enhanced visual (VT-1) examination, with sufficient magnification and lighting, capable of a 1-mil resolution.

TVAs original submittal referenced ASME Section XI Code Case N-648-1 and Table IWB-3510-3. Instead, TVA will utilize the acceptance criteria of Table IWB-3512-1 of the 1995 Edition, 1996 Addenda of Section XI for the examination. The conservative approach is to use Table IWB-3512-1 utilizing an aspect ratio of 0.50 and surface flaw depth of 2.5 percent for calculating the flaw acceptance criteria. For Table IWB-3512-1, the depth of a crack indication is assumed to be one half of the measured length of the crack indication.

Crack-like surface flaws exceeding the acceptance criteria of ASME Section XI Code, Table IWB-3512-1 are considered unacceptable for continued service unless the reactor pressure vessel meets the requirements of IWB-3142.2, IWB-3142.3 or IWB-3142.4.

NRC Question No. 5 Provide a description for the performance of the proposed enhanced direct visual examination of the RPV head nozzles that will assure a 1-mil resolution.

E-5

TVA Response The direct enhanced visual examination of the reactor pressure vessel (RPV) head nozzles (N6A, N6B, and N7) inner radius sections, will be performed in accordance with ASME Section XI VT-1 requirements. Resolution sensitivity for the direct enhanced visual examination will be demonstrated utilizing a vision test chart containing a 1-mil wire at a maximum distance of 2 feet, in lieu of the 0.044 inch character required by IWA-2210 of ASME Section XI 1995 Edition, 1996 Addenda.

NRC Question No. 6 Discuss the obstructions that will limit full coverage for the remote enhanced visual examination of the Recirculation, Core Spray, and Jet-Pump Instrumentation nozzles inner radius section.

TVA Response Visual examination of the inner radius region is limited because reactor internal piping configuration prevents placement of the camera in all positions necessary to examine the M-N surface over the full circumference. The limitations are described below.

N-5A and N-5B, Core Spray nozzles - Core spray thermal sleeve and tee box and feed water sparger. Feedwater spargers are located above the core spray nozzle and the configuration of the core spray thermal sleeve and tee box prohibits placement of the camera 360 degrees around the nozzle. The limitations are at the top position at approximate clock positions 10:00 to 2:00, and the bottom position at approximate clock positions 4:00 to 8:00.

N-2 nozzles (10), Recirculation Inlet - Thermal sleeve and jet pump riser piping. The inaccessible area is the inside bend radius of elbow at approximate clock positions 11:00 to 1:00, and at the bottom outside diameter bend of the elbow at approximate clock positions 5:00 to 7:00. The limitations are due to the proximity of the of the jet pump risers.

N-8A and N-8B nozzles, Jet-Pump Instrumentation - 12 Instrumentation lines pass through the vessel wall into the vessel. The core shroud support plate is located directly beneath the nozzle preventing placement of the camera from approximate clock positions 4:00 to 8:00.

E-6