BSEP 10-0058, Request for Alternative to Nozzle-to-Vessel Shell Weld and Inner Radius Section Examinations

Request for Alternative to Nozzle-to-Vessel Shell Weld and Inner Radius Section Examinations

ML101310390
Person / Time
Site:	Brunswick
Issue date:	04/29/2010
From:	Mentel P Progress Energy Carolinas
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
BSEP 10-0058
Download: ML101310390 (11)

Text

&j2Progress Energy 10 CFR 50.55a(a)(3)(i)

APR 2 9 2010 SERIAL: BSEP 10-0058 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001

Subject:

Brunswick Steam Electric Plant, Unit Nos. 1 and 2 Renewed Facility Operating License Nos. DPR-71 and DPR-62 Docket Nos. 50-325 and 50-324 Request for Alternative to Nozzle-to-Vessel Shell Weld and Inner Radius Section Examinations Ladies and Gentlemen:

In accordance with 10 CFR 50.55a(a)(3)(i), Carolina Power & Light Company (CP&L),

now doing business as Progress Energy Carolinas, Inc., requests NRC approval of a proposed alternative to the applicable edition of the American Society of Mechanical Engineers (ASME) Code,Section XI, for the Brunswick Steam Electric Plant (BSEP),

Units 1 and 2. The proposed alternative will allow reduced requirements for reactor pressure vessel nozzle-to-vessel weld and inner radius examinations. The alternative is requested for the fourth 10-year inservice inspection interval of the Inservice Inspection Program for both BSEP Unit 1 and Unit 2.

Details of the 10 CFR 50.55a proposed alternative are provided in the enclosure of this letter. The proposed alternative is consistent with ASME Code Case N-702, "Alternative Requirements for Boiling Water Reactor (BWR) Nozzle Inner Radius and Nozzle-to-Shell Welds,Section XI, Division 1," as described in Enclosure 1.

The Electric Power Research Institute (EPRI) proprietary report, "BWR Vessel and Internals Project, Technical Basis for the Reduction of Inspection Requirements for the Boiling Water Reactor Nozzle-to-Vessel Shell Welds and Nozzle Inner Radius (BWRVIP-108)," dated October 2002, is the technical basis document for ASME Code Case N-702. CP&L has performed an evaluation to demonstrate the plant-specific '

applicability of the BWRVIP- 108 report as technical basis for the proposed alternative for BSEP Units 1 and 2. The plant-specific applicability evaluation is provided in Enclosure 2.

CP&L requests approval of this relief request by February 1, 2011, to support planning for the Brunswick Unit 2 refueling outage, which is currently scheduled to begin March 5, 2011.

Progress Energy Carolinas, Inc.

Brunswick Nuclear Plant PO Box 10429 Southport, NC28461

Document Control Desk BSEP 10-0058 / Page 2 No regulatory commitments are contained in this letter. Please refer any questions regarding this submittal to Ms. Annette Pope, Supervisor - Licensing/Regulatory Programs, at (910) 457-2184.

Sincerely, Phyllis N. Mentel Manager - Support Services Brunswick Steam Electric Plant WRM/wrm

Enclosures:

1. 10 CFR 50.55a Alternative Request Number ISI-05

2. Plant-specific Applicability Evaluation of BWRVIP- 108

Document Control Desk BSEP 10-0058 / Page 3 cc (with enclosures):

U. S. Nuclear Regulatory Commission, Region II ATTN: Mr. Luis A. Reyes, Regional Administrator 245 Peachtree Center Ave., NE, Suite 1200 Atlanta, GA 30303-1257 U. S. Nuclear Regulatory Commission ATTN: Mr. Philip B. O'Bryan, NRC Senior Resident Inspector 8470 River Road Southport, NC 28461-8869 U. S. Nuclear Regulatory Commission (Electronic Copy Only)

ATTN: Mrs. Farideh E. Saba (Mail Stop OWFN 8G9A) 11555 Rockville Pike Rockville, MD 20852-2738 Chair - North Carolina Utilities Commission P.O. Box 29510 Raleigh, NC 27626-0510 Mr. Jack M. Given, Jr., Bureau Chief North Carolina Department of Labor Boiler Safety Bureau 1101 Mail Service Center Raleigh, NC 27699-1101

BSEP 10-0058 Enclosure 1 Page 1 of 6 10 CFR 50.55a Alternative Request Number ISI-05 Proposed Alternative In Accordance with 10 CFR 50.55a(a)(3)(i)

- Alternative Provides Acceptable Level of Quality and Safety -

1. ASME Code Components Affected Code Class: ASME Code Class 1

References:

1. ASME Code,Section XI, Subarticle IWB-2500, Table IWB-2500-1

2. ASME Code Case N-702 Examination Categories: B-D, Full Penetration Welded Nozzles in Vessels - Inspection Program B Item Numbers: B3.90, "Nozzle-to-Vessel Welds" B3.100, "Nozzle Inner Radius Section" Component Numbers: Reactor Pressure Vessel Nozzles N2, N3, N5, N6, N8, N 11, N12, and N16 (See Attachment 1 for the list of component identifications)

Description:

Alternative to ASME Code,Section XI, Table IWB-2500-1

2. Applicable Code Edition and Addenda

The Inservice Inspection Program for the fourth 10-year inservice inspection interval is based on the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code,Section XI, 2001 Edition with 2003 Addenda.

Additionally, for ultrasonic examinations, the ASME Code,Section XI, Appendix VIII, "Performance Demonstration for Ultrasonic Examination Systems," of the 1995 Edition is implemented, as required and modified by 10 CFR 5055a(b)(2)(xv).

The Brunswick Steam Electric plant (BSEP) is currently in the fourth 10-year Inservice Inspection (ISI) Program interval.

3. Applicable Code Renuirement The applicable Code requirement is the ASME Code,Section XI, 2001 Edition, 2003 Addenda, Table IWB-2500-1, Examination Category B-D, Full Penetration Welded Nozzles in Vessels - Inspection Program B. Item B3.90 requires volumetric examination of all

BSEP 10-0058 Enclosure 1 Page 2 of 6 nozzle-to-vessel welds. Item B3.100 requires volumetric examination of all nozzle inside radii sections.

All of the nozzle assemblies listed above are full penetration welds. Attachment 1 provides a complete list of the affected nozzle welds.

4. Reason for Request

The proposed alternative provides an acceptable level of quality and safety. The reduction in inspection scope could provide a radiological exposure savings of as much as 3.5 person-rem for Unit 1 and 4.0 person-rem for Unit 2, over the remainder of the interval.

5. Proposed Alternative and Basis for Use Proposed Alternative In accordance with 10 CFR 50.55a(g)(5)(iii), CP&L requests an alternative from performing the ASME Code-required examinations of 100 percent of the reactor pressure vessel (RPV) nozzle assemblies at BSEP Units 1 and 2, as identified in Attachment 1. As an alternative, CP&L proposes the use of ASME Code Case N-702 (i.e., Reference 1), which would require examination of a minimum of 25 percent of the nozzle-to-vessel shell welds and nozzle inner blend radius sections, including at least one nozzle from each system and nominal pipe size.

Table 1 summarizes the proposed inspection requirements for the affected vessel nozzle assemblies listed in Attachment 1. Both the nozzle inner radius and the nozzle-to-vessel shell weld would be examined for each of the identified nozzle assemblies.

The ASME Code requires that the 100 percent of the RPV nozzle assemblies be examined each ISI interval. Under the proposed alternative, CP&L would examine a minimum of 25 percent of the RPV nozzle assemblies, as identified in the table below.

TABLE 1 Total Minimum Number Group Number to be Examined Recirculation Inlet Nozzles 10 3 N2A, N2B, N2C, N2D, N2E, N2F, N2G, N2H, 2J, and N2K Main Steam Nozzles 4 1 N3A, N3B, N3C, and N3D Core Spray Nozzles 2 1 N5A and N5B Reactor Pressure Vessel Head 2 1 Spray Nozzles N6A and N6B (Head Spray)

BSEP 10-0058 Enclosure 1 Page 3 of 6 Total Minimum Number Group Number to be Examined Jet Pump Instrumentation nozzles 2 1 N8A and N8B Instrumentation Nozzles 6 2 N11A, N11B, N12A, N12B, N16A and N16 BWRVIP-108 eliminated from consideration instrumentation nozzles with partial penetration welds (i.e., see Section 3.4, "Selection of Nozzles for Evaluation"). However, for BSEP Units 1 and 2, the N11, N12, and N16 instrumentation nozzle welds are full penetration welds and thus are being included in the scope of this request.

ASME Code Case N-702 stipulates that a VT-I examination method may be used in lieu of the volumetric examination method for the inner radius sections (i.e., Item B3.100). CP&L has adopted ASME Code Case N-648-1, "Alternative Requirements for Inner Radius Examination of Class 1 Reactor Vessel Nozzles," (i.e., Reference 2) with the provisions stipulated in Regulatory Guide 1.147 (i.e., Reference 3), in the BSEP Inservice Inspection Program for the fourth inspection interval.

Basis for Use Electric Power Research Institute (EPRI) proprietary report, "BWR Vessel and Internals Project, Technical Basis for the Reduction of Inspection Requirements for the Boiling Water Reactor Nozzle-to-Vessel Shell Welds and Nozzle Inner Radius (BWRVIP- 108)," dated October 2002, provides the technical basis for ASME Code Case N-702. The evaluation found that failure probability at the nozzle blend radius region and nozzle-to-vessel shell welds is very low (i.e., < 1 x 10-6 for 40 years) with or without inservice inspection. The report concludes that inspection of 25 percent of each nozzle type is technically justified.

The EPRI report was approved by the NRC in a Safety Evaluation dated December 19, 2007 (i.e., ADAMS Accession Number ML073600374). Section 5.0, "Plant-Specific Applicability," of the Safety Evaluation indicates that each licensee who plans to request relief from the ASME Code nozzle-to-vessel shell welds and nozzle inner radius sections may reference the BWRVIP- 108 report as the technical basis for the use of ASME Code Case N-702 as an alternative. However, each licensee should demonstrate the plant-specific applicability criteria from the BWRVIP-108 report to its units by showing that all the general and nozzle-specific criteria addressed below are satisfied. The BSEP-specific applicability is demonstrated in Enclosure 2.

Based on these evaluations, the proposed alternative use of ASME Code Case N-702, in lieu of ASME Code,Section XI requirements, provides an acceptable level of quality and safety in accordance with 10 CFR 50.55a(a)(3)(i) for the applicable RPV nozzle-to-vessel shell welds and nozzle inner radii sections.

I BSEP 10-0058 Enclosure 1 Page 4 of 6

6. Duration of the Proposed Alternative CP&L proposes the use of the proposed alternative for the fourth 10-year inservice inspection interval at BSEP, Units 1 and 2, or until such time as ASME Code Case N-702 is published in a future revision of Regulatory Guide 1.147, "Inservice Inspection Code Case Acceptability, ASME Section XI, Division 1." The fourth 10-year interval began on May 11, 2008, and will end on May 10, 2018, for BSEP, Units 1 and 2.

7. Precedents Similar requests were granted for the Clinton Power Station, Columbia Generating Station, Dresden Nuclear Power Station, and Quad Cities Nuclear Power Station, as listed in References 4 through 7, respectively.

8. References

1. ASME Code Case N-702, Alternate Requirements for Boiling Water Reactor (BWR)

Nozzle Inner Radius and Nozzle-to-Shell WeldsSection XI, Division 1.

2. ASME Code Case N-648-1, Alternate Requirements for Inner Radius Examinations of Class 1 Reactor Vessel NozzlesSection XI, Division 1.

3. Regulatory Guide 1.147, Revision 15, Inservice Inspection Code Case Acceptability, ASME Section XI, Division 1.

4. Letter from U.S. Nuclear Regulatory Commission (USNRC) to Exelon Nuclear, Clinton Power Station, Unit No. 1 - Proposed Alternative to 10 CFR 50.55a Examination Requirements for Reactor Pressure Vessel Weld Inspections (TAC No. ME0218), dated August 24, 2009, ADAMS Accession Number ML092300394.

5. Letter from U.S. Nuclear Regulatory Commission (USNRC) to Energy Northwest, Columbia Generating Station - Request for Relief No. 3ISI-09 for the Third 10-Year Inservice Inspection Program Interval (TAC No. MD9850), dated April 8, 2009, ADAMS Accession Number ML090790588.

6. Letter from U.S. Nuclear Regulatory Commission (USNRC) to Exelon Nuclear, Dresden Nuclear Power Station, Units 2 and 3 - Alternative to nozzle-to-Vessel Weld and inner Radius Examinations (TAC.Nos. ME0882 and ME0883), dated November 3, 2009, ADAMS Accession Number ML092940436.

7. Letter from U.S. Nuclear Regulatory Commission (USNRC) to Exelon Nuclear, Quad Cities Nuclear Power Station, Unit Nos. 1 and 2 - Alternative to Nozzle to Vessel Weld and Inner Radius Examinations (TAC Nos. ME0765 and ME0766), dated February 2, 2010, ADAMS Accession Number ML092860259.

BSEP 10-0058 Enclosure 1 Page 5 of 6 Attachment 1 Applicable Nozzles Noninal Category Item Pipe Size Component ID Number Number System (Inches)

N2A Nozzle B-D B3.90 Recirc Inlet 12 N2A Inner Radius Section B-D B3.100 Recirc Inlet 12 N2B Nozzle B-D B3.90 Recirc Inlet 12 N2B Inner Radius Section B-D B3.100 Recirc Inlet 12 N2C Nozzle B-D B3.90 Recirc Inlet 12 N2C Inner Radius Section B-D B3.100 Recirc Inlet 12 N2D Nozzle B-D B3.90 Recirc Inlet 12 N2D Inner Radius Section B-D B3.100 Recirc Inlet 12 N2E Nozzle B-D B3.90 Recirc Inlet 12 N2E Inner Radius Section B-D B3.100 Recirc Inlet 12 N2F Nozzle B-D B3.90 Recirc Inlet 12 N2F Inner Radius Section B-D B3.100 Recirc Inlet 12 N2G Nozzle B-D B3.90 Recirc Inlet 12 N2G Inner Radius Section B-D B3.100 Recirc Inlet 12 N2H Nozzle B-D B3.90 Recirc Inlet 12 N2H Inner Radius Section B-D B3.100 Recirc Inlet 12 N2J Nozzle B-D B3.90 Recirc Inlet 12 N2J Inner Radius Section B-D B3.100 Recirc Inlet 12 N2K Nozzle B-D B3.90 Recirc Inlet 12 N2K Inner Radius Section B-D B3.100 Recirc Inlet 12 N3A Nozzle B-D B3.90 Main Steam 28 N3A Inner Radius Section B-D B3. 100 Main Steam 28 N3B Nozzle B-D B3.90 Main Steam 28 N3B Inner Radius Section B-D B3.100 Main Steam 28 N3C Nozzle B-D B3.90 Main Steam 28 N3C Inner Radius Section B-D B3.100 Main Steam 28 N3D Nozzle B-D B3.90 Main Steam 28 N3D Inner Radius Section B-D B3.100 Main Steam 28 N5A Nozzle B-D B3.90 Core Spray 10 N5A Inner Radius Section B-D B3.100 Core Spray 10 N5B Nozzle B-D B3.90 Core Spray 10 N5B Inner Radius Section B-D B3.100 Core Spray 10 N6A Nozzle B-D B3.90 Head Spray 6 N6A Inner Radius Section B-D B3.100 Head Spray 6 N6B Nozzle B-D B3.90 Head Spray 6 N6B Inner Radius Section B-D B3.100 Head Spray 6 N8A Nozzle B-D B3.90 Jet Pump 4 Instrumentation N8A Inner Radius Section B-D B3.100 Jet Pump 4 Instrumentation N8B Nozzle B-D B3.90 Jet Pump 4 Instrumentation N8B Inner Radius Section B-D B3. 100 Jet Pump 4 Instrumentation N 11ANozzle B-D B3.90 Instrumentation 2 N IIA Inner Radius Section B-D B3.100 Instrumentation 2 N 11B Nozzle B-D B3.90 Instrumentation 2 N l1B Inner Radius Section B-D B3.100 Instrumentation 2

BSEP 10-0058 Enclosure 1 Page 6 of 6 Noninal Category Item Pipe Size Component ID Number Number System Inches)

N12ANozzle B-D B3.90 Instrumentation 2 N I2A Inner Radius Section B-D B3.100 Instrumentation 2 N12B Nozzle B-D B3.90 Instrumentation 2 N 12B Inner Radius Section B-D B3.100 Instrumentation 2 N16ANozzle B-D B3.90 Instrumentation 2 N 16A Inner Radius Section B-D B3.100 Instrumentation 2 N I6B Nozzle B-D B3.90 Instrumentation 2 N 16B Inner Radius Section B-D B3.100 Instrumentation 2

BSEP 10-0058 Enclosure 2 Page 1 of 2 Plant-Specific Applicability Evaluation of BWRVIP-108 Electric Power Research Institute (EPRI) proprietary report, "BWR Vessel and Internals Project, Technical Basis for the Reduction of Inspection Requirements for the Boiling Water Reactor Nozzle-to-Vessel Shell Welds and Nozzle Inner Radius (BWRVIP-108)," dated October 2002, was approved by the NRC in a Safety Evaluation dated December 19, 2007 (i.e., ADAMS Accession Number ML073600374). Section 5.0, "Plant-Specific Applicability," of the Safety Evaluation indicates that each licensee who plans to request relief from the ASME Code nozzle-to-vessel shell welds and nozzle inner radius sections may reference the BWRVIP-108 report as the technical basis for the use of ASME Code Case N-702 as an alternative. However, each licensee should demonstrate the plant-specific applicability criteria from the BWRVIP-108 report to its units by showing that all the general and nozzle-specific criteria addressed below are satisfied. The plant-specific applicability criteria are addressed below.

(1) Criterion 1 from the BWRVIP-108 report specifies that the maximum RPV heatup/cooldown rate should be less than 11 5°F per hour. As specified by Brunswick Technical Specification 3.4.5, "RCS Pressure and Temperature (P/T) Limits," the maximum Reactor Pressure Vessel (RPV) heatup/cooldown rate is limited to less than 100°F per hour.

For Recirculation Inlet Nozzles (N2)

(2) (pr/t)/CRPV < 1. 15 p = RPV Normal Operating Pressure 1030 r = RPV Inner Radius 110.2 t = RPV Wall Thickness 5.68 CRPV 19332 Result: 1.033698 < 1.15 PASS (3) [p(r 0 2 + ri2 ) / (ro 2 - ri2 )] / CNOZZLE < 1 .15 p = RPV Normal Operating Pressure 1030 ro= Nozzle Outer Radius 14.125 ri= Nozzle Inner Radius 7.0625 CNOZZLE 1637 Result: 1.048666 < 1.15 PASS Based on the results of Criteria 2 and 3 above, the recirculation inlet nozzles are included in the scope of this request.

BSEP 10-0058 Enclosure 2 Page 2 of 2 For Recirculation Outlet Nozzles (NI)

(4) (pr/t)/CRpv < 1.15 p RPV Normal Operating Pressure 1030 r = RPV Inner Radius 110.2 t = RPV Wall Thickness 5.68 CRPV 16171 Result: 1.235759 < 1.15 FAIL (5) [p(ro2 + ri2 ) / (r0 2 - ri2 )] / CNOZZLE < 1.15 p = RPV Normal Operating Pressure 1030 ro = Nozzle Outer Radius 24.375 ri Nozzle Inner Radius 13.0625 CNOZZLE 1977 Result: 0.940797 < 1.15 PASS Based on the result of Criteria 4 above, the recirculation outlet nozzles are excluded from the scope of this request.