JPN-04-002, Alternative to Appendix Viii, Supplement 4 Requirements for Reactor Vessel Pressure Retaining Weld Inspection

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Alternative to Appendix Viii, Supplement 4 Requirements for Reactor Vessel Pressure Retaining Weld Inspection
ML040440157
Person / Time
Site: Pilgrim, FitzPatrick
Issue date: 02/11/2004
From: Kansler M
Entergy Nuclear Northeast, Entergy Operations
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
ENO Ltr 2.04.008, JPN-04-002
Download: ML040440157 (7)
v • d • e


Text

Entergy Nuclear Northeast Entergy Nuclear Operations, Inc.

i- Entergy 440 Hamilton Avenue White Plains, NY 10601 Tel 914 272 3200 Fax 914 272 3205 Michael R. Kansler President February 11, 2004 JPN-04-002 ENO Ltr 2.04.008 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Mail Stop O-Pl-17 Washington, DC 20555-0001

SUBJECT:

James A. FitzPatrick Nuclear Power Plant Docket No. 50-333 Pilgrim Nuclear Power Station Docket No. 50-293 Alternative to Appendix Vill. Supplement 4 requirements for Reactor Vessel Pressure Retaining Weld Inspection

References:

1. US NRC letter, from Anthony J. Mendiola to Howard Bergendahl, uDavis-Besse Nuclear Power Station, Unit 1 - Inservice Inspection Relief Request No. RR-A23 for the Second 10-Year Inspection Interval (TAC NO. MB1608)", dated February 13, 2002.
2. Entergy letter to USNRC, NL-03-189, 'Alternative to Appendix VIII, Supplement 4 requirements for Reactor Vessel Pressure Retaining Weld Inspection", dated December 30, 2003

Dear Sir:

Pursuant to 10CFR50.55a(a)(3)(i), Entergy Nuclear Operations, Inc. (Entergy) hereby requests the Nuclear Regulatory Commission (NRC) to approve the use of an alternative to the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code,Section XI requirements regarding the inspection of Class 1, Examination Category B-A, pressure retaining welds in the reactor vessel.

Enclosed are two (2) similar requests for relief (RRs) to use the proposed alternatives for James A. FitzPatrick Nuclear Power Plant (JAF, Enclosure 1) and Pilgrim Nuclear Power Station (PNPS, Enclosure 2). The ASME Code,Section XI, 1995 Edition, 1996 Addenda, Appendix Vil, Supplement 4, requires that performance demonstration results satisfy the statistical parameters of Paragraph 3.2(c). The statistical parameters of Paragraph 3.2(c) of Supplement 4 to Appendix Vil are inconsistent with the Performance Demonstration Initiative (PDI) program criteria.

A-O 1~

Relief is requested to use the root mean square (RMS) value of 10CFR50.55a(b)(2)(xv)(C)(1),

which modifies the depth sizing criteria of ASME Code, Section Xl, 1995 Edition, 1996 Addenda, Appendix VIII, Supplement 4, Subparagraph 3.2(a), in lieu of Subparagraph 3.2(c) of Supplement 4 to Appendix Vil. Pursuant to 10CFR50.55a(a)(3)(i), the proposed alternative maintains an acceptable level of quality and safety.

These requests for relief for JAF and PNPS are for their 3rd ISI Interval, and the applicable code of record is the 1989 Edition, No Addenda of the ASME Section Xl Code.

A similar request for relief was granted to Davis-Besse, Unit 1 (Reference 1).

Entergy requests approval of the JAF relief request (Enclosure 1) by August 2004 to support its Fall 2004 refueling outage. Since these RRs are practically identical, Entergy requests that the Pilgrim relief request (Enclosure 2) be approved at the same time. Entergy submitted in December 2003 two similar RRs for Indian Point Nuclear Generating Units 2 and 3 (Reference 2,) both of which are currently under NRC review.

There are no new commitments made in this letter. If you have any questions, please contact Ms. Charlene Faison at 914-272-3378.

Very truly yours, ichael R. Kansl

/ President /

Entergy Nu ear Operations, Inc.

List of

Enclosures:

1. James A. FitzPatrick Nuclear Power Plant, RR-34
2. Pilgrim Nuclear Power Station, PRR-32 cc:

Mr. Hubert J. Miller Mr. Travis Tate, Project Manager License Regional Administrator, Region I Project Directorate I U.S. Nuclear Regulatory Commission Division of Licensing Project Management 475 Allendale Road Office of Nuclear Reactor Regulation King of Prussia, PA 19406-1415 U.S. Nuclear Regulatory Commission Mail Stop 7-D-1 Mr. Guy S. Vissing, Project Manager Washington, DC 20555-0001 Project Directorate I Division of Licensing Project Management Resident Inspector's Office Office of Nuclear Reactor Regulation James A. FitzPatrick U.S. Nuclear Regulatory Commission U.S. Nuclear Regulatory Commission Mail Stop 8-C2 P.O. Box 136 Washington, DC 20555-0001 Lycoming, NY 13093-0136 2

Senior Resident Inspector Pilgrim Nuclear Power Station U.S. Nuclear Regulatory Commission 600 Rocky Hill Road Mail Stop 66 Plymouth, MA 02360 Mr. Paul Eddy New York State Department of Public Service 3 Empire State Plaza Albany, NY 12223 Mr. Peter R. Smith, Acting President New York State Energy, Research, and Development Authority Corporate Plaza West 286 Washington Avenue Extension Albany, NY 12203-6399 3

JPN-04-002 I ENO Ltr 2.04.008 Enclosure I JAMES A. FITZPATRICK THIRD TEN-YEAR INTERVAL INSERVICE INSPECTION PROGRAM RELIEF REQUEST RR-34 Proposed Alternative In Accordance with 10CFR50.55a(a)(3)(i)

--Alternative Provides Acceptable Level of Quality and Safety--

1. ASME Code Component(s) Affected Component Numbers: ASME Section Xl Class 1 Pressure Retaining Welds in Reactor Vessel Examination Category: B-A Item Number: B1.10 Circumferential and Longitudinal Shell Welds B1.20 Head Welds
2. Applicable Code Edition and Addenda The Code of Record for the third Inservice Inspection Interval is ASME Section XI Code, 1989 Edition, No Addenda.
3. Applicable Code Requirements ASME Code,Section XI, Appendix VilI, Supplement 4, Subparagraph 3.2(c), states that the UT performance demonstration results must be plotted on a two-dimensional plot with the measured depth plotted along the ordinate axis and the true depth plotted along the abscissa axis. For qualification, the plot must satisfy the following statistical parameters: (1) slope of the linear regression line is not less than 0.7, (2) the mean deviation of flaw depth is less than 0.25 inches, and (3) correlation coefficient is not less than 0.70.
4. Reason for Request The statistical parameters of Paragraph 3.2(c) of Supplement 4 to Appendix VIII are inconsistent with the PDI program criteria.
5. Proposed Alternative Pursuant to 10CFR50.55a(a)(3)(i), relief is requested to use the root mean square (RMS) value of 10CFR50.55a(b)(2)(xv)(C)(1), which modifies the depth sizing criteria of ASME Code, Section Xl, 1995 Edition, 1996 Addenda, Appendix Vil, Supplement 4, Subparagraph 3.2(a), in lieu of Subparagraph 3.2(c) of Supplement 4 to Appendix Vil.

I

JPN-04-002 / ENO Ltr 2.04.008 Enclosure I Basis for Use ASME Code, Section Xl, Appendix VilI, supplement 4, Subparagraph 3.2(c) imposes three statistical parameters for depth sizing. The first parameter, 3.2(c)(1), pertains to the scope of a linear regression line. The linear regression line is the difference between actual versus true value plotted along a through-wall thickness. For Supplement 4 performance demonstrations, a linear regression line of the data is not applicable because the performance demonstrations are performed on test specimens with flaws located in the inner15 percent through-wall. The differences between the measured versus true value produce a tight grouping of results, which resemble a shotgun pattern. The slope of a regression line from such data is extremely sensitive to small variations, thus making the parameter of 3.2(c)(1), an inappropriate criterion. The second parameter, 3.2(c)(2), pertains to the mean deviation of flaw depth. The value used in the Code is too lax with respect to evaluating flaw depths within the inner 15 percent of wall thickness.

Therefore, Entergy Nuclear Operations, Inc. (Entergy) proposes to use the more appropriate criterion of 0.15 inch RMS of 10CFR50.55a(b)(2)(xv)(C)(1), which modifies Subparagraph 3.2(a),

as the acceptance criterion. The third parameter, 3.2(c)(3), pertains to a correction coefficient.

The value of the correction coefficient in Subparagraph 3.2(c)(3) is inappropriate for this application since it is based on the linear regression from Subparagraph 3.2(c)(1).

Entergy believes the proposed alternative to use the RMS value of I0CFR50.55a(b)(2)(xv)(C)(1),

which modifies the criterion of ASME Code, Appendix VIII, Supplement 4, Subparagraph 3.2(a), in lieu of Subparagraph 3.2(c), will provide an acceptable level of quality and safety.

6. Duration of Proposed Alternative It is proposed to use the alternative for the remainder of the Third Inservice Inspection Interval for James A. FitzPatrick Nuclear Power Plant.
7. Precedents A similar request for relief was granted to Davis-Besse, Unit 1 (Docket No. 50-346, TAC NO.

MB1608, dated February 13, 2002).

8. Attachment None 2

JPN-04-002 1 ENO 2.04.008 Enclosure 2 PILGRIM NUCLEAR POWER STATION THIRD TEN-YEAR INTERVAL INSERVICE INSPECTION PROGRAM RELIEF REQUEST PRR-32 Proposed Alternative In Accordance with 10CFR50.55a(a)(3)(i)

-Alternative Provides Acceptable Level of Quality and Safety-

1. ASME Code Component(s) Affected Component Numbers: ASME Section Xl Class 1 Pressure Retaining Welds in Reactor Vessel Examination Category: B-A Item Number: B1.10 Circumferential and Longitudinal Shell Welds B13.20 Head Welds
2. Applicable Code Edition and Addenda The Code of Record for the third Inservice Inspection Interval is ASME Section Xl Code, 1989 Edition, No Addenda.
3. Applicable Code Requirements ASME Code, Section Xl, Appendix Vill, Supplement 4, Subparagraph 3.2(c), states that the UT performance demonstration results must be plotted on a two-dimensional plot with the measured depth plotted along the ordinate axis and the true depth plotted along the abscissa axis. For qualification, the plot must satisfy the following statistical parameters: (1) slope of the linear regression line is not less than 0.7, (2) the mean deviation of flaw depth is less than 0.25 inches, and (3) correlation coefficient is not less than 0.70.
4. Reason for Request The statistical parameters of Paragraph 3.2(c) of Supplement 4 to Appendix Vil are inconsistent with the PDI program criteria.
5. Proposed Alternative Pursuant to 10CFR50.55a(a)(3)(i), relief is requested to use the root mean square (RMS) value of 10CFR50.55a(b)(2)(xv)(C)(1), which modifies the depth sizing criteria of ASME Code, Section Xl, 1995 Edition, 1996 Addenda, Appendix VilI, Supplement 4, Subparagraph 3.2(a), in lieu of Subparagraph 3.2(c) of Supplement 4 to Appendix Vil.

1

JPN-04-002 / ENO 2.04.008 Enclosure 2 Basis for Use ASME Code, Section Xl, Appendix VIII, supplement 4, Subparagraph 3.2(c) imposes three statistical parameters for depth sizing. The first parameter, 3.2(c)(1), pertains to the scope of a linear regression line. The linear regression line is the difference between actual versus true value plotted along a through-wall thickness. For Supplement 4 performance demonstrations, a linear regression line of the data is not applicable because the performance demonstrations are performed on test specimens with flaws located in the 15 percent through-wall. The differences between the actual versus true value produce a tight grouping of results, which resemble a shotgun pattern. The slope of a regression line from such data is extremely sensitive to small variations, thus making the parameter of 3.2(c)(1), an inappropriate criterion. The second parameter, 3.2(c)(2), pertains to the mean deviation of flaw depth. The value used in the Code is too lax with respect to evaluating flaw depths within the inner 15 percent of wall thickness.

Therefore, Entergy Nuclear Operations, Inc. (Entergy) proposes to use the more appropriate criterion of 0.15 inch RMS of 10CFR50.55a(b)(2)(xv)(C)(1), which modifies Subparagraph 3.2(a),

as the acceptance criterion. The third parameter, 3.2(c)(3), pertains to a correction coefficient.

The value of the correction coefficient in Subparagraph 3.2(c)(3) is inappropriate for this application since it is based on the linear regression from Subparagraph 3.2(c)(1).

Entergy believes the proposed alternative to use the RMS value of 10CFR50.55a(b)(2)(xv)(C)(1),

which modifies the criterion of ASME Code, Appendix VIII, Supplement 4, Subparagraph 3.2(a), in lieu of Subparagraph 3.2(c), will provide an acceptable level of quality and safety.

6. Duration of Proposed Alternative It is proposed to use the alternative for the remainder of the Third Inservice Inspection Interval for Pilgrim Station.
7. Precedents A similar request for relief was granted to Davis-Besse, Unit 1 (Docket No. 50-346, TAC NO.

MB1608, dated February 13, 2002).

8. Attachment None 2
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