ML031400574

Inservice Inspection Program, Revision to Relief Request HC-RR-B11, Hope Creek Generating Station

ML031400574
Person / Time
Site:	Hope Creek
Issue date:	05/09/2003
From:	Salamon G Public Service Enterprise Group
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
LRN-03-0215
Download: ML031400574 (6)

Text

PSEG Nuclear LLC P.O. Box 236, Hancocks Bridge, New Jersey 08038-0236 r

MAY.0 9 2003 0 Nuclear PSEG LLC LRN-03-0215 United States Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 INSERVICE INSPECTION PROGRAM REVISION TO RELIEF REQUEST HC-RR-B1I HOPE CREEK GENERATING STATION FACILITY OPERATING LICENSES NPF-57 DOCKET NOS. 50-354 On April 14, 2003, PSEG Nuclear requested relief (HC-RR-Bl 1) from the requirements of ASME Section Xl 1989 Edition, Table IWB-2500-1, Examination Category B-D, Item B3.100, in order to implement the alternative requirements of an enhanced VT-1, visual examination technique of the surface M-N shown in ASME Section Xl, Figures IWB-2500-7 (a) through (d) as an alternative to ASME Section Xl Table IWB-2500-1, Examination Category B-D, Item B3.100 that required volumetric examination (Ultrasonic, UT) of the Inner Radius of Class 1 Reactor Vessel Nozzles. The request for relief is for the second 10-year inservice inspection (ISI) interval.

During a teleconference on May 5, 2003, PSEG provided a response to the NRC Staff's question concerning the relief request. As a result of this telecon, PSEG is submitting a revision to relief request HC-RR-B 1 for the NRC Staffs action. The relief request contained in this submiffal replaces the relief request submitted on April 14, 2003 in its entirety.

Pursuant to 10 CFR 50.55a(a)(3), PSEG Nuclear, LLC (PSEG Nuclear) requests approval of the enclosed relief request. Approval for relief is requested in accordance with the alternative examination provisions of 10CFR50.55a(a)(3)(i). PSEG Nuclear proposes to use an alternative to the required volumetric examination for nozzles where plant configuration is such that visual examination of the inner radius may be performed on essentially 100 percent of the inner radius in lieu of the existing ASME Section Xl Table IWB-2500-1, Examination Category B-D, Item B3.100. Compliance with the proposed alternatives will provide an adequate level of quality and safety for examination of the affected areas.

The attachment to this letter includes the proposed alternative and supporting justification for the relief. Based on the evaluation contained in the attachment, PSEG Nuclear has concluded that the proposed alternative provides an acceptable level of quality and safety. Accordingly, this proposal satisfies the requirements of 10 CFR 50.55a(a)(3)(i).

- 4r) 95-2168 REV. 7/99

Document Control Desk MAY 0 9 2003 LRN-03-0215 Should you have any questions regarding this request, please contact Mr. Brian Thomas at 856-339-2022.

Sincerely, G. Salamon Manager - Nuclear Safety and Licensing Atachment:

ISI Relief Request HC-RR-B 1 C Mr. H. Miller, Administrator Regional Administrator - Region I U. S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 U. S. Nuclear Regulatory Commission ATTN: Mr. G. Wunder Licensing Project Manager - Hope Creek Mail Stop 08B1 Washington, DC 20555-001 USNRC Senior Resident Inspector - Hope Creek (X24)

Mr. K. Tosch, Manager IV Bureau of Nuclear Engineering P. O. Box 415 Trenton, NJ 08625

Document Control Desk Attachment I LRN-03-021 5 Relief Request HC-RR-BIl ASME Code Component Affected Alternative Exam Requirements for Inner Radius Examination of Class 1 Reactor Pressure Vessel Nozzles. [See Table 1 below]

Applicable ASME Code Edition and Addenda ASME Section Xl, 1989 Edition, is the code of record for PSEG Nuclear LLC (PSEG Nuclear) Hope Creek Nuclear Generating Station's Second Ten-Year ISI Program Interval.

Applicable Code Requirements Conduct ultrasonic examinations of Hope Creek Nuclear Generating Station Reactor Pressure Vessel (RPV) Nozzle Inside Radius Sections in accordance with ASME Section Xl 1989 Edition IWB-2500-1 requirements for Class 1 Examination Category B-D, Item B3.100, Figures IWB-2500-7 (a) through (d).

Proposed Alternative Pursuant to 10 CFR 50.55a(a)(3)(i), approval is requested to use the proposed alternative to the required volumetric examination for nozzles where plant configuration is such that visual examination of the inner radius may be performed on essentially 100 percent of the inner radius in lieu of the existing ASME Section Xl Table IWB-2500-1, Examination Category B-D, Item B3.1 00.

Compliance with the proposed alternatives will provide an adequate level of quality and safety for examination of the affected areas.

PSEG Nuclear proposes to perform an enhanced VT-1 (EVT-1), visual examination technique of the surface M-N shown in ASME Section XI, Figures IWB-2500-7 (a) through (d) as an altemative to ASME Section Xi Table IWB-2500-1, Examination Category B-D, Item B3.100 requiring volumetric examination (Ultrasonic, UT) of the Inner Radius of Class 1 Reactor Vessel Nozzles.

The enhanced remote visual examination (EVr-1) will be performed upon the examination surface M-N to achieve essentially 100% coverage using 8x magnification video equipment to examine the inner radii. The resolution sensitivity for this remote examination will be established using a 1-mil diameter wire standard similar to that used for other reactor pressure vessel intemal examinations intended to detect cracking.

Proposed Altemative InAccordance with 10 CFR 50.55a(a)(3)(i)

- Alternative Provides Acceptable Level of Quality and Safety --

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714, Document Control Desk Aftachment I LRN-03-0215 Relief Request HC-RR-BII 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.

Examinations proposed would be performed during the following refueling outages RFO1 (Spring 2003), RFO12 (Fall 2004), and RFO13 (Spring 2005).

Basis for Relief The following Hope Creek RPV Nozzle Inner radius exams listed below do not contain configurations that would impede visual examination of the nozzle inner radius area surface M-N.

Table 1 Hope Creek RPV Nozzle Inner Radius Exams Summary Number Examination Area Configuration Identification 100408 RPVI-NIAIR 0O- Recirculation Outlet Nozzle 100409 RPVI-NIBIR 1800- Recirculation Outlet Nozzle 100460 RPV1 -N3AIR 72 0-Main Steam Nozzle 100465 RPV1-N3BAIR 108 0-Main Steam Nozzle 100470 RPV1-N3CAIR 2520-Main Steam Nozzle 100475 RPV1-N3DAIR 288 0-Main Steam Nozzle 100520 RPV1-N6AIR Spray Head Nozzle 100525 RPVI-N6BIR Spare Spray Head Nozzle 100530 RPVI-N71R Head Vent Nozzle 100330 RPV1-N9A Capped CRD Hydraulic Retum Nozzle All nozzle forgings were nondestructively examined during fabrication and have been previously examined using ultrasonic techniques specific to the nozzle Proposed Alternative InAccordance with 10 CFR 50.55a(a)(3)(i)

- Alternative Provides Acceptable Level of Quality and Safety --

Page 2 of 4

Document Control Desk Aftachment I LRN-03-021 5 Relief Request HC-RR-B1I configuration. No indications of fabrication or service related cracking have been observed as result of these exams.

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 ASME Section Xl, 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 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 or operational CRD returns.

No service related cracking has been discovered in any of the BWR (boiling water reactor) fleet plant nozzles other than feedwater and operational CRD returns. The six feedwater nozzle inner radius sections will continue to be examined in accordance with UT techniques developed and qualified with GE-NE-523-A71-0594-A Revision 1 (the NRC has approved this report under TAC No. MA6787). PSEG Nuclear believes that application of a visual examination alternative for the listed nozzle inner radius regions ensures an acceptable level of quality and safety.

According to the NRC memorandum (Reference No. 1), the staff indicated that an ultrasonic examination could be replaced by a Vr-1 visual examination of the proposed nozzle inspections on the basis that the surveillance is being maintained and a VT-1 visual examination is completed.

The implementation of this relief request should reduce vessel examination time by approximately 18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br />, which translates to significantly reduced personnel radiation exposure and cost savings.

Note: 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. As previously stated, crack-like surface flaws found 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.

Duration of Proposed Alternative Hope Creek - Second Ten-Year Interval (ASME Xl 1989 Edition)

Precedence Previous relief has been granted to Detroit Edison, Fermi Unit 2 [NRC Safety Evaluation Reports TAC No. MB2166 and MB2755 dated October 5, 2001].

Proposed Altemative InAccordance with 10 CFR 50.55a(a)(3)(i)

- Altemative Provides Acceptable Level of Quality and Safety -

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Document Control Desk Attachment 1 LRN-03-021 5 Relief Request HC-RR-B1I References

1. NRC Internal memorandum from K.R. Wichman (NRC) to W.H. Bateman (NRC) dated May 25, 2000; Subject The Third Meeting with the Industry to discuss the elimination of RPV Inner Radius Inspection (ML003718630).

2. Code Case N-648, Alternative Requirements for Inner Radius Examinations of Class 1 Reactor Vessel Nozzles Section Xl, Division 1.

Proposed Alternative In Accordance with 10 CFR 50.55a(a)(3)(i)

- Alternative Provides Acceptable Level of Quality and Safety -

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