ML021150777

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American Society of Mechanical EngineersSection XI, Inservice Inspection Program - Requests for Relief 2-ISI-16 and 2-ISI-17
ML021150777
Person / Time
Site: Browns Ferry Tennessee Valley Authority icon.png
Issue date: 04/23/2002
From: Abney T
Tennessee Valley Authority
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
2-ISI-16, 2-ISI-17
Download: ML021150777 (42)
v • d • e


Text

R08 020423 626 April 23, 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 - REQUESTS FOR RELIEF 2-ISI-16 AND 2-ISI-17 In accordance with 10 CFR 50.55a(a)(3)(i) and (ii), TVA is requesting relief from certain inservice inspection requirements in Section XI of the ASME Boiler and Pressure Vessel Code. The enclosure to this letter contains BFN Unit 2 requests for relief 2-ISI-16 and 2-ISI-17 for NRC review and approval.

TVA is seeking relief from inservice inspection requirements of the 1995 Edition with the 1996 Addenda,Section XI of the ASME Code for the volumetric examination of Class 1, Reactor Pressure Vessel (RPV) nozzle inner radius sections for the RPV and RPV head. As an alternative, TVA proposes to use an enhanced remote visual (VT-1) examination, capable of a 1-mil resolution, for the RPV nozzles inner radius section. For the RPV head nozzles inner radius section, TVA proposes to use an enhanced direct visual (VT-1) examination capable of a 1-mil resolution.

These requests for relief are consistent with ones submitted by Detroit Edison Company for Fermi Unit 2 by letters dated June 11, and August 16, 2001. The NRC approved the requests for relief by letter dated October 5, 2001. Additionally,

U.S. Nuclear Regulatory Commission Page 2 April 23, 2002 request for relief 2-ISI-16 is consistent with one for BFN Unit 3 RPV head nozzles (3-ISI-11) submitted by TVA letters dated August 13, 2001, January 9, and February 5, 2002. NRC approved this request for relief by letter dated March 13, 2002.

TVA requests approval of this request for relief by June 28, 2002, to support resource planning for the Unit 2 Cycle 12 (Spring 2003) refueling outage.

There are no new commitments contained in this letter. 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 cc: See Page 3

U.S. Nuclear Regulatory Commission Page 3 April 23, 2002 Enclosure (Via NRC Electronic Distribution) cc (Enclosure):

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 4 April 23, 2002 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 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\Unit 2 ASME RFR 2-ISI-16 & 17.doc

ENCLOSURE 1 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 (See Attached)

TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT (BFN)

UNIT 2 ASME SECTION XI INSERVICE INSPECTION PROGRAM (THIRD TEN-YEAR INSPECTION INTERVAL)

REQUEST FOR RELIEF 2-ISI-16 Executive Summary: In accordance with 10 CFR 50.55a(a)(3)(i),

TVA is requesting relief from inservice inspection requirements of the 1995 Edition with the 1996 Addenda,Section XI of the ASME Boiler and Pressure Vessel Code for the volumetric examination of Class 1, reactor pressure vessel (RPV) and RPV head nozzles inner radius sections.

The examination requirement is for a volumetric examination of ASME Section XI, Examination Category B-D, Full Penetration Welded Nozzles in Vessels, Item No. B3.100, Reactor Vessel Nozzle Inner Radius Section.

This request for relief applies to the BFN Unit 2 Reactor Pressure Vessel and RPV Head nozzles, with the exception of the six (N4) Feedwater nozzles. The six Feedwater nozzle radius sections will continue to be examined with ultrasonic testing (UT) techniques developed and qualified using GE-NE-523-A71-0594-A, Revision 01, "Alternate BWR Feedwater Nozzle Inspection Requirements," (the NRC has approved this report in an SER transmitted by letter dated March 10, 2000), and ASME Section XI Code requirements. TVA will take both ASME Section XI Code credit and credit in accordance with GE-NE-523-A71-0594-A, Revision 01, for all six Feedwater nozzles.

TVA proposes to implement alternate visual (VT-1) requirements consistent with ASME Code Case N-648-1, Alternative E1-2

Requirements for Inner Radius Examination of Class 1 Reactor Vessel Nozzles. TVA recognizes that the original publication of Code Case N-648 references a subsurface flaw acceptance criteria table that does not exist (i.e., Table IWB-3513-3, which is a typographical error). Code Case N-648 was revised to correct the error and references Table IWB-3510-3, the correct applicable acceptance criteria.

In addition, an editorial clarification was added to N-648 to provide guidance for determining component thickness to use with this table. A copy of Code Case N-648-1, that passed ASME Code Subcommittee is included in Attachment A of this request.

In summary TVAs proposed alternative examinations are as follows: (1) For the Reactor pressure vessel nozzle inner radius sections, TVA will perform an enhanced remote visual (VT-1) examination, capable of a 1-mil wire resolution, in accordance with ASME Section XI, VT-1 requirements, (2) For the RPV head nozzles, TVA will perform an enhanced direct visual (VT-1) examination of the nozzle inner radius sections, capable of a 1-mil wire resolution, in accordance with ASME Section XI VT-1 requirements.

TVA considers the above proposed alternative examinations will provide an acceptable level of quality and safety.

The proposed alternatives will also provide a significant savings in examination resources and radiation exposure.

This request for relief is consistent with one submitted by Detroit Edison for Fermi Unit 2 by letters dated June 11, 2001, and August 16, 2001. NRC approved this request for relief by letter dated October 5, 2001.

This request for relief is also consistent with one for the BFN Unit 3 RPV head E1-3

nozzles (3-ISI-11) submitted by TVA letters dated August 13, 2001, January 9, and February 5, 2002. NRC approved this request for relief by letter dated March 13, 2002.

Unit: Two (2)

ISI Interval: ASME Section XI, Third Ten-Year ISI Interval (May 25, 2001 to May 24, 2011)

System(s): Reactor Pressure Vessel (RPV)

Components: RPV and RPV Head Nozzles, Inner Radius Sections:

Reactor Recirculation Suction Loop, N1A and N1B (Total of 2 nozzles)

Main Steam Nozzles, N3A, N3B, N3C, and N3D (Total of 4 nozzles)

Reactor Pressure Vessel Head Nozzles, N6A, N6B, and N7 (Total of 3 nozzles)

Control Rod Drive Return Line Nozzle, (capped) N9 (Total of 1 nozzle)

ASME Code Class: ASME Code Class 1 ASME Section XI Code Edition: 1995 Edition with 1996 Addenda Code Table: IWB-2500-1 Examination Category: B-D, Full Penetration Welded Nozzles In Vessels Examination Item Number: B3.100, Nozzle Inner Radius Sections, (Nozzles N1A, N1B, N3A, N3B, N3C, N3D, N6A, N6B, N7, and N9)

Code Requirement: The 1995 Edition with 1996 Addenda, ASME Section XI, Table IWB-2500-1, Examination Category B-D, Item B3.100, requires a volumetric examination of the reactor E1-4

pressure vessel (RPV) and RPV head nozzles inner radius section.

Code Requirements From Which Relief Is Requested: Relief is requested from the requirement to perform the volumetric examination of the reactor pressure vessel and RPV head nozzle inner radius sections (Nozzles N1A, N1B, N3A, N3B, N3C, N3D, N6A, N6B, N7, and N9)

List Of Items Associated With The Relief Request: Reactor Pressure Vessel and RPV Head Nozzles, N1A, N1B, N3A, N3B, N3C, N3D, N6A, N6B, N7, and N9 (Total of 10 nozzles)

Note: TVAs proposed alternative examination is consistent with Code Case N-648-1 for the nozzle inner radius section of the RPV and RPV head nozzles.

The six Feedwater nozzle inner radius sections will continue to be examined with ultrasonic examination (UT) techniques developed and qualified with GE-NE-523-A71-0594-A, Revision 01 (NRC has approved this report by letter dated March 10, 2000), and ASME Section XI Code requirements. TVA will take both ASME Section XI Code credit for the nozzle to vessel weld for Code Category B-D, Item No. B3.90 and the inner radius sections Code Category B-D, Item No. B3.100, and credit in accordance with GE-NE-523-A71-0594-A, Revision 01, simultaneously for all six Feedwater nozzles.

Basis For Relief Request: Pursuant to 10 CFR 50. 55a(a)(3)(i) TVA is requesting relief from ASME Section XI requirements to perform the volumetric E1-5

examination described above. TVA is proposing to implement a visual examination alternative. This examination is consistent with the alternative proposed in ASME Code Case N-648-1. The visual examination will cover the same inspection surface as specified for the volumetric examination.

The volumetric examinations (ultrasonic) conducted from the outside surfaces are difficult and time consuming due to the asymmetrical configuration of both the nozzle outside surface (where the transducers are manipulated) and the inner radius section of the nozzle being interrogated. Examination of the asymmetrical surfaces may require several different transducer/wedge angle combinations and these are applied at certain azimuths around the nozzle weld blend area of the vessel surface.

Different size nozzles usually require a separate set of transducer/wedge angle combinations and calibrations. Several hours may be required for the calibrations and examination of one typical 6-inch diameter nozzle inner radius section.

An enhanced visual (VT-1) examination of the nozzle inner radius sections would provide assurance of the required coverage and indicate the presence of surface flaws. The option to perform an enhanced visual (VT-1) examination will provide an acceptable examination without compromising the level of quality and safety.

The proposed alternative will also provide a significant savings in examination resources and radiation exposure to examination and support personnel.

Alternate Examination: In accordance with 10 CFR 50.55a(a)(3)(i)

TVA will perform the following alternate examinations:

E1-6

TVA will perform an enhanced direct visual (VT-1) examination, capable of a 1-mil wire resolution, of the reactor pressure vessel (RPV) head nozzles (Nozzles N6A, N6B, and N7) inner radius sections, in accordance with ASME Section XI, VT-1 requirements. Essentially 100 percent Code coverage will be attained.

For the Reactor pressure vessel nozzles inner radius sections, (N1A, N1B, N3A, N3B, N3C, N3D, and N9), TVA will perform an enhanced remote visual examination (VT-1), capable of a 1-mil wire resolution, in accordance with ASME Section XI, VT-1 requirements. This examination will be a remote visual exam utilizing cameras.

Essentially 100 percent Code coverage will be attained.

Crack-like surface 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.

Justification For The Granting Of Relief: The RPV nozzles were nondestructively examined during fabrication and have previously been examined using inservice ultrasonic techniques specific to the nozzle configuration. No indication of fabrication defects or service related cracking has been detected by these examinations.

The RPV and RPV head nozzles inner radius sections are the only non-welded areas (excluding the RPV head bolts) requiring examination on the reactor vessel. This requirement was deterministically made early in the development of ASME Section XI. For all nozzles, other than Feedwater, there is no significant thermal cycling during operation. From a E1-7

risk perspective, there is no need to perform a volumetric examination on any nozzle other than the Feedwater and CRD return nozzles. No service related cracking has ever been discovered in any of the BWR fleet nozzles other than on Feedwater or CRD return lines.

Development of Code Case N-648 was coordinated with the Westinghouse Owners Group (WOG), ASME, and the NRC. On May 9, 2000, the WOG met with NRC to discuss issues related to the proposed inspection elimination for reactor pressure vessel inner radius regions. Although justification was presented to eliminate any examination of RPV nozzle inner radius sections (excluding BWR Feedwater and CRD nozzles), a consensus was reached between the WOG and the NRC, to replace the volumetric examination of the other RPV nozzles (non Feedwater or CRD) with a visual (VT-1) examination.

Implementation Schedule: This request for relief is applicable to the BFN Unit 2, Third Ten-Year ASME Section XI Inservice Inspection Interval (May 25, 2001 to May 24, 2011).

Attachments: Attachment A ASME Code Case N-648-1, Alternative Requirements for Inner Radius Examinations of Class 1 Reactor Vessel Nozzles,Section XI, Division 1 Attachment B - (7 sketches)

Sketch SK-B2001, Reactor Pressure Vessel Assembly Sketch SK-B2017, N1 Recirculation Nozzles Sketch SK-B2018, N3,Main Steam Nozzles E1-8

Sketch SK-B2020, N9, Control Rod Drive Return Line Nozzle Sketch SK-B2011, Attachments Map Top Head Assembly Sketch SK-B2015, Weld Detail Vent Nozzle N7 Sketch SK-B2016, Weld Detail Head Spray/Instrumentation Nozzles N6 E1-9

Attachment A 2-ISI-16 ASME Code Case N-648-1 E1-10

E1-11 Attachment B 2-ISI-16 Seven (7) Sketches SK-B2001, RPV Assembly SK-B2017, Recirc. Nozzles SK-B2018, Main Steam Nozzles SK-B2020, CRD, Return Line Nozzle SK-B2011, RPV Top Head Assembly SK-B2015, RPV Head, Weld Detail Vent Nozzle SK-B2016, RPV Head, Weld Detail Head Spray/Inst. Nozzle E1-12

Attachment B Sketch SK-B2001 E1-13

Attachment B Sketch SK-B2017 E1-14

Attachment B Sketch SK-B2018 E1-15

Attachment B Sketch SK-B2020 E1-16

Attachment B Sketch SK-B2011 E1-17

Attachment B Sketch SK-B2015 E1-18

Attachment B Sketch SK-B2016 E1-19

ENCLOSURE 2 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-17 (See Attached)

TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT (BFN)

UNIT 2 ASME SECTION XI INSERVICE INSPECTION PROGRAM (THIRD TEN-YEAR INSPECTION INTERVAL)

REQUEST FOR RELIEF 2-ISI-17 Executive Summary: In accordance with 10 CFR 50.55a(a)(3)(ii), TVA is requesting relief from inservice inspection requirements of the 1995 Edition with the 1996 Addenda,Section XI of the ASME Boiler and Pressure Vessel Code for the volumetric examination of Class 1, reactor pressure vessel (RPV) nozzle inner radius sections for the Recirculation Inlet loop, Jet Pump Instrumentation, and Core Spray nozzles.

The examination requirement is for a volumetric examination of ASME Section XI, Examination Category B-D; Full Penetration Welded Nozzles in Vessels, Item No.

B3.100, Reactor Vessel Nozzle Inner Radius Section.

TVA is proposing to perform a remote visual examination of the accessible surface (reactor internal piping configuration prevents placement of camera to obtain essentially 100 percent coverage) of the nozzle inner radius region for the specified nozzles. TVAs proposal is consistent with the alternative visual (VT-1), requirements of ASME Code Case N-648-1, Alternative Requirements for Inner Radius Examination of Class 1 Reactor Vessel Nozzles. TVA recognizes that the original publication of Code Case N-648 references a subsurface flaw acceptance criteria table that does not exist (i.e., Table IWB-3513-3, which is a typographical error). Code Case N-648 was revised to correct the error and references Table IWB-3510-3, the correct applicable acceptance criteria. In E2-2

addition, an editorial clarification was added to provide guidance for determining component thickness to use with this table. A copy of Code E2-3

Case N-648-1, that passed ASME Section XI Subcommittee is included in Attachment A of this request.

TVA considers that the proposed enhanced visual (VT-1) examination, capable of a 1-mil resolution, of the accessible portions of the nozzle inner radius region for the specified nozzles will provide an acceptable level of quality and safety.

Further, compliance with the existing ASME Section XI requirement would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.

Performance of the volumetric examination requires the examiner to enter and remain inside the biological shield penetration area around the nozzle for the duration of the ultrasonic examination, which takes approximately one hour. Dose rates for the specified RPV nozzles are in the range of 500 to 1200 millirem per hour, with shielding in place. Performance of these examinations results in an estimated personnel exposure of about 4.5 Rem per inspection interval. Performance of a visual examination using remote cameras essentially eliminates any personnel exposure.

This request for relief is consistent with one submitted by Detroit Edison for Fermi Unit 2 by letters dated June 11, 2001, and August 16, 2001. NRC approved this request for relief by letter dated October 5, 2001.

Unit: Two (2)

ISI Interval: ASME Section XI, Third Ten-Year ISI Interval (May 25, 2001 to May 24, 2011)

System(s): Reactor Pressure Vessel (RPV)

Components: RPV Nozzles Inner Radius Sections:

E2-4

Reactor Recirculation Inlet Loop Nozzles, N2A, N2B, N2C, N2D, N2E, N2F, N2G, N2H, N2J, and N2K (Total of 10 nozzles)

Core Spray Nozzles, N5A and N5B (Total of 2 nozzles)

Jet Pump Instrumentation Nozzles, N8A and N8B (Total of 2 nozzles)

ASME Code Class: ASME Code Class 1 ASME Section XI Code Edition: 1995 Edition with 1996 Addenda Code Table: IWB-2500-1 Examination Category: B-D, Full Penetration Welded Nozzles in Vessels Examination Item Number: B3.100, Nozzle Inner Radius Sections, (Nozzles N2A, N2B, N2C, N2D, N2E, N2F, N2G, N2H, N2J, N2K, N5A, N5B, N8A, and N8B)

Code Requirement: The 1995 Edition with 1996 Addenda, ASME Section XI, Table IWB-2500-1, Examination Category B-D, Item B3.100, requires a volumetric examination of all reactor pressure vessel nozzles inner radius section welded with full penetration welds as shown in Figures IWB-2500-7(a) through (d).

Code Requirements From Which Relief Is Requested: Relief is requested from the requirement to perform the volumetric examination of the reactor pressure vessel nozzles inner radius section (Nozzles N2A, N2B, N2C, N2D, N2E, N2F, N2G, N2H, N2J, N2K, N5A, N5B, N8A, and N8B)

List Of Items Associated With E2-5

The Relief Request: Reactor Pressure Vessel Nozzles, N2A, N2B, N2C, N2D, N2E, N2F, N2G, N2H, N2J, N2K, N5A, N5B, N8A, and N8B (Total of 14 nozzles)

Basis For Relief Request: Pursuant to 10 CFR 50.55a(a)(3)(ii) TVA is requesting relief from ASME Section XI requirements to perform the volumetric examination described above. Performance of the volumetric examination results in significant personnel radiation exposure without a compensating increase in the level of plant quality or safety. TVA is proposing to perform a remote enhanced visual (VT-1) examination, capable of a 1-mil resolution, of the accessible portions of the nozzle inner radius region for the specified RPV nozzles.

The volumetric examinations (ultrasonic) conducted from the outside surfaces are difficult and time consuming due to the asymmetrical configuration of both the nozzle outside surface (where the transducers are manipulated) and the inner radius section of the nozzle being interrogated. Examination of the asymmetrical surfaces may require several different transducer/wedge angle combinations and these are applied at certain azimuths around the nozzle weld blend area of the vessel surface.

Different size nozzles usually require a separate set of transducer/wedge angle combinations and calibrations. Several hours may be required for the calibrations and examination of one typical 6-inch diameter nozzle inner radius section.

TVAs proposed examination is consistent with the visual (VT-1), requirements of ASME Code Case N-648-1, Alternative Requirements for Inner Radius Examination of Class 1 Reactor Vessel Nozzles.

Development of Code Case N-648 was coordinated with the Westinghouse Owners E2-6

Group (WOG), ASME, and the NRC. On May 9, 2000, the WOG met with NRC to discuss issues related to the proposed inspection elimination for reactor pressure vessel inner radius regions. Although justification was presented to eliminate any examination of RPV nozzle inner radius sections (excluding BWR Feedwater and CRD nozzles), a consensus was reached between the WOG and the NRC, to replace the volumetric RPV nozzle inner radius section examination with a visual examination (VT-1).

An enhanced remote visual (VT-1) examination of the accessible regions of the nozzle inner radius sections would provide assurance of the required coverage and indicate the presence of surface flaws. The performance of an enhanced remote visual (VT-1) examination would provide an acceptable examination without compromising the level of quality or safety.

The proposed examination will also provide a significant savings in examination resources and radiation exposure to examination and support personnel.

Alternate Examination: TVA will perform the following examination on the specified nozzles:

For the reactor pressure vessel nozzles inner radius sections, (N2A, N2B, N2C, N2D, N2E, N2F, N2G, N2H, N2J, N2K, N5A, N5B, N8A, and N8B), TVA will perform an enhanced remote visual examination (VT-1),

capable of a 1-mil wire resolution, in accordance with ASME Section XI, VT-1 requirements. This examination will be a remote visual examination utilizing cameras. Visual examination of the inner radius section for the above nozzles is limited because the reactor internal piping configuration prevents placement of E2-7

the camera in all positions necessary to examine the surface M-N [See Figures IWB-2500-7(a) through (d)] over the full circumference.

Crack-like surface 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.

The specific limitations and estimated examination coverage for each nozzle are provided below.

NOZZLE TYPE/NO. LIMITATION ESTIMATED COVERAGE Recirculation Thermal Sleeve/ 50%

Inlet, N2 Jet-Pump Riser (10 nozzles)

Core Spray, N5 Thermal Sleeve 40%

(2 Nozzles) and Sparger Jet-Pump Instrumentation 60%

Instrumentation, Lines N8 (2 Nozzles)

Justification For The Granting Of Relief: The limited remote visual examination of the RPV nozzles inner radius section does not significantly reduce the level of plant quality and safety for the following reasons:

  • There are no mechanisms of damage other than fatigue for the nozzle inner radius, and for other than Feedwater nozzles, there is no cause for significant thermal cycling. Therefore, E2-8

the primary flaw of concern would be a flaw that was not detected during the manufacturing process. The BFN Unit 2 RPV nozzles were examined during and after manufacturing by surface and volumetric techniques. Additionally preservice and inservice ultrasonic examinations have detected no flaws. It is unlikely that flaws would be initiated by the fatigue mechanism.

  • After approximately 20 years of operation (Unit 2 was shut down from September 1984 to May 1991), no cracking in the subject BFN Unit 2 RPV nozzles inner radius region has been found.
  • Fracture toughness tests performed at Oakridge National Laboratories1 indicate there is a large flaw tolerance for BWR Nozzle inner radius regions. Even if flaw propagation was assumed, test results indicate a leak before break scenario would occur, which would not result in a significant increase in core damage frequency2. In addition, system pressure testing continues to be performed each refueling outage, and during plant operation the containment is monitored for changes in unidentified leakage.
  • More than 50 percent of the total RPV nozzle population receives a complete (i.e., essentially 100 percent) nozzle inner radius examination.
  • Visual examination of the accessible nozzle inner radius surface (zone M-N) provides reasonable assurance that deep flaws are not present. Additionally, when flaws are initiated by the fatigue mechanism, they typically are encountered over a significant portion of the circumference as was the case for cracking of Feedwater nozzles addressed in NUREG-0619.

E2-9

In summary, fatigue cracking is the only applicable degradation mechanism for the RPV nozzle inner radius region, and 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 CRD return lines.

This is supported by the fact that no service related cracking has ever been discovered in any of the BWR fleet RPV nozzles other than on Feedwater and CRD return lines.

The BFN Unit 2 Feedwater nozzles 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. The NRC accepted this 1 & 2: Conclusions made in ASME NDE subcommittee report ISI-99-26, Technical Basis for Elimination of Reactor Vessel Nozzle Inner Radius Inspections E2-10

methodology by letter to the Boiling Water Reactor Owners Group (BWROG) dated March 10, 2000. TVA notified the NRC by letter dated October 23, 2000 (TAC Nos. M08436 and M08437) that it was adopting the BWROG methodology.

The Feedwater nozzles alone represent 20 percent of all RPV nozzles currently requiring volumetric inner radius examination, which is more than industry accepted risk sampling requirements for similar items. Additionally, BFN request for relief 2-ISI-16 provides for a full (i.e., essentially 100 percent) visual examination of 10 other RPV nozzles, resulting in a complete examination of more than 50 percent of the total BFN Unit 2 RPV nozzle population. Therefore, TVA believes that the remote visual examination of the accessible regions of the 14 specified RPV nozzle inner radius sections will ensure an acceptable level of quality and safety while providing a significant reduction in personnel dose.

Implementation Schedule: This request for relief is applicable to the BFN Unit 2, Third Ten-Year ASME Section XI Inservice Inspection Interval (May 25, 2001 to May 24, 2011).

Attachments: Attachment A ASME Code Case N-648-1, Alternative Requirements for Inner Radius Examinations of Class 1 Reactor Vessel Nozzles,Section XI, Division 1 Attachment B - (3 sketches)

Sketch SK-B2001, Reactor Pressure Vessel Assembly Sketch SK-B2018, Recirculation N2 and Core Spray Nozzles N5 E2-11

Sketch SK-B2019, Jet-Pump Nozzles N8 E2-12

Attachment A 2-ISI-17 ASME Code Case N-648-1 E2-13

E2-14 Attachment B 2-ISI-17 Three (3) Sketches SK-B2001, RPV Assembly SK-B2018, Recirculation N2 and Core Spray N5 Nozzles SK-B2019, Jet-Pump Nozzles N8 E2-15

Attachment B Sketch SK-B2001 E2-16

Attachment B Sketch SK-B2018 E2-17

Attachment B Sketch SK-B2019 E2-18

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