ML031130007

Code Relief, Alternative to Required Volumetric Examination for Nozzles Where Plant Configuration Is Such That Visual Examination of Inner Radius May Be Performed on Essentially 100 Percent of Inner Radius

ML031130007
Person / Time
Site:	Hope Creek
Issue date:	04/14/2003
From:	Salamon G Public Service Enterprise Group
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
HC-RR-B11, LRN-03-0081
Download: ML031130007 (6)

Text

PSEG Nuclear LLC P.O Box 236, Hancocks Bndge, New Jersey 08038-0236 APR 1 4 2003 0 PSEG NuclearLLC LRN-03-0081 United States Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 INSERVICE INSPECTION PROGRAM RELIEF REQUEST HC-RR-B11I HOPE CREEK GENERATING STATION FACILITY OPERATING LICENSES NPF-57 DOCKET NOS. 50-354 Pursuant to 10CFR50.55a(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 I OCFR50.55a(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).

This relief request is applicable to PSEG Nuclear Hope Creek Generating Station.

PSEG Nuclear requests that the NRC approve this request by April 2003 in order to support Hope Creek refueling outage RFO 1 scheduled to commence April 12, 2003.

Should you have any questions regarding this request, please contact Mr. Howard Berrick at 856-339-1862.

S3 r elI G.

Salamon Manager - Nuclear Safety and Licensing

Attachment:

go4 ISI Relief Request HC-RR-B11 95-2168 REV 7/99

Document Control Desk APR 1 4 2003 LRN-03-0081 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. Box415 Trenton, NJ 08625

Document Control Desk Attachment 1 LRN-03-0081 Relief Request HC-RR-B1 I ASME Code Component Affected Alternative Exam Requirements for Inner Radius Examination of Class 1 Reactor Pressure Vessel Nozzles. [See Table I 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 Xi 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 B3.100.

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, 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 requiring volumetric examination (Ultrasonic, UT) of the Inner Radius of Class 1 Reactor Vessel Nozzles.

The enhanced remote visual examination 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 internal examinations intended to detect cracking.

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

-- Alternative Provides Acceptable Level of Quality and Safety -

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Document Control Desk Attachment I LRN-03-0081 Relief Request HC-RR-B I Reactor vessel closure head vent and spray nozzles inner radii will receive direct visual examinations (VT-1) conducted in accordance with ASME Xl requirements, while the other remaining aforementioned components will receive enhanced visual examinations using the 1-mil diameter wire standard.

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 RFO11 (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 I Hope Creek RPV Nozzle Inner Radius Exams Summary Number Examination Area Configuration 100408 RPVI-NIAIR - 0°- Recirculation Outlet Nozzle 100409 RPV1-NIBIR 180°- Recirculation Outlet Nozzle 100460 RPV1-N3AIR 720 -Main Steam Nozzle 100465 RPVI-N3BAIR 1080-Main Steam Nozzle 100470 RPV1-N3CAIR 252 0-Main Steam Nozzle 100475 RPV1-N3DAIR 288 0-Main Steam Nozzle 100520 RPVI-N6AIR Spray Head Nozzle 100525 RPV1-N6BIR Spare Spray Head Nozzle 100530 RPVI-N71R Head Vent Nozzle 100330 RPVI-N9A Capped CRD Hydraulic Return Nozzle Proposed Alternative In Accordance with 10 CFR 50.55a(a)(3)(i)

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

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Document Control Desk Attachment I LRN-03-0081 Relief Request HC-RR-B11I All nozzle forgings were nondestructively examined during fabrication and have been previously examined using ultrasonic techniques specific to the nozzle 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 tnan 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 VT-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)

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

- Alternative Provides Acceptable Level of Quality and Safety -

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Document Control Desk Attachment I LRN-03-0081 Relief Request HC-RR-B11I 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].

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 NozzlesSection XI, 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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