PLA-7154, Response to Request for Additional Information on Fourth Ten-Year Inservice Inspection Interval Proposed Relief Requests

Response to Request for Additional Information on Fourth Ten-Year Inservice Inspection Interval Proposed Relief Requests

ML14118A443
Person / Time
Site:	Susquehanna
Issue date:	04/28/2014
From:	Franke J Susquehanna
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
PLA-7154
Download: ML14118A443 (9)

Text

Jon A. Franke PPL Susquehanna, LLC Site Vice President 769 Salem Boulevard Berwick, PA 18603 Tel. 570.542.2904 Fax 570.542.1504 jfranke@ pplweb.com APR 2 8 2014 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001 SUSQUEHANNA STEAM ELECTRIC STATION RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION ON FOURTH TEN-YEAR INSERVICE INSPECTION INTERVAL PROPOSED RELIEF REQUESTS Docket No 50-387 PLA-7154 and No. 50-388

References:

1. PPL Letter (PIA-7052), "Proposed Relief Requests for the Fourth Ten- Year Inservice Inspection Interval for Susquehanna Units 1 and 2," dated August 30, 2013 (Accession ML13247A167).

2. NRC Letter, "Request for Additional Information Regarding Relief Requests for the Fourth 10-Year Inservice Inspection Interval (TAC Nos. MF2705, MF2706, MF2709, and MF2710)," dated December 19, 2013 (Accession ML13338A442).

3. PPL Letter (PIA-7126), "Response to Request for Additional Information on Fourth Ten-Year Inservice Inspection Interval Proposed Relief Requests," dated January 31, 2014 (Accession ML14031A081).

The purpose of this letter is for PPL Susquehanna, LLC (PPL) to provide clarification to the information previously provided in response to NRC question RAI 4RR-06 1 from Reference 2. The NRC requested this clarification in a discussion on March 28, 2014, at which time PPL agreed to provide the additional information.

Additionally, PPL determined on April 3, 2014, that the requested alternative 4RR-07 for the reactor pressure vessel head flange seal leak detection line visual examination incorrectly described the line as Code Class 2. This line is in fact a Code Class 1 component.

The changes necessary to 4RR-06 1 in response to the RAI that have been discussed with the NRC are described in Attachment 1. The change in Code Class described in 4RR-07 is provided in Attachment 2 as 4RR-07, Revision 1, which replaces the original revision in its entirety. The station's Corrective Action Program addresses the condition in 4RR-07 for evaluation, and this response supports the current NRC review of the proposed alternative requests in Reference 1.

Document Control Desk PLA-7154 The attachments to this provide a response to the NRC RAI 4RR-06 1, which replaces the original response provided in Reference 3. Additionally, the revised 4RR-07, Revision 1, is provided in Attachment 2.

There are no new regulatory commitments associated with this response.

If you have any questions or require additional information, please contact Mr. Duane L. Filchner (570) 542-6501.

J. * .Franke : Response to Request for Additional Information : Relief Request 4RR-07, Revision 1 Copy: NRC Region I Mr. J. Greives, NRC Sr. Resident Inspector Mr. J. Whited, NRC Project Manager Mr. L. Winker, PA DEP/BRP

Attachment 1 to PLA-7154 Response to Request for Additional Information

Attachment 1 to PLA-7154 Page 1 of2 Response to Request for Additional Information By letter dated August 30, 2013, (I) PPL Susquehanna, LLC (PPL the licensee), submitted a set of Alternative Requests from the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),Section XI for the Fourth Ten-Year Inservice Inspection (lSI)

Interval at Susquehanna Steam Electric Station (SSES) Units 1 and 2. The NRC requested additional information (RAI) in a letter dated December 19, 2013. (Z) PPL provided the additional information requested in a letter dated January 31, 2014. (3) The additional information provided in this attachment is for the purpose of clarifying the response previously provided in Reference 3 to the NRC question 4RR-06 1. This will replace the response in its entirety, related to the requirements for inspection of snubbers, supports and attachments.

Additionally, and as described in the balance of this Attachment, PPL is correcting information previously submitted in Reference 1, in an alternative request 4RR-07 for the reactor pressure vessel head flange seal leak detection line visual examination. This is discussed further below, and Revision 1 of 4RR-07 is provided in Attachment 2.

RAI 4RR-06 1:

Based on information in the relief request, two different ASME Codes will be used. Verify that for snubbers (pin-to-pin), ASME OM Code 2004 Edition with 2005 and 2006 Addenda and for the snubber supports and attachments, ASME Section XI 2007 Edition with 2008 addenda will be used.

PPL's Response:

Examinations of snubbers and their associated supports will be administered and scheduled per the PPL Snubber Program under the requirements of the ASME OM Code 2004 Edition through the 2006 Addenda. Examination of the snubber attachments is required per ASME Section XI Code 2007 Edition with the 2008 Addenda and the PPL lSI Program. Visual examinations of the associated attachments will be performed at the same time as the required visual examination of the snubber. This is being done as a dose and time saving effort. For both examinations, a VT-3 qualified inspector will be performing the examination. Examination of 100 percent of the snubber attachments exceeds Code requirements as defined in the ASME Section XI Code.

(1) PPL Letter (PLA-7052), "Proposed Relief Requests for the Fourth Ten-Year In service Inspection Interval for Susquehanna Units 1 and 2," dated August 30,2013 (Accession ML13247A167).

(2) NRC Letter, "Request for Additional Information Regarding Relief Requests for the Fourth 10-Year Inservice Inspection Interval (TAC Nos. MF2705, MF2706, MF2709, and MF2710)," dated December 19, 2013 (Accession ML13338A442).

(3) PPL Letter (PLA-7126), "Response to Request for Additional Information on Fourth Ten-Year Inservice Inspection Interval Proposed Relief Requests," dated January 31, 2014 (Accession ML14031A081).

Attachment 1 to PLA-7154 Page 2 of2 4RR-07, Revision 1:

The station's Corrective Action Program addresses a condition in the alternative request 4RR-07, Revision 0, where the reactor pressure vessel head flange seal leak detection line is described incorrectly as a Code Class 2 component, when it is in fact a Code Class 1 component. The request has also been revised further to replace 10 CFR 50.55a(a)(3)(i) provisions for an alternative with an equivalent level of quality and safety, with a request for an alternative pursuant to 10 CFR 50.55a(a)(3)(ii), on the basis that compliance with the Code specified pressure requirement to test the leak-off lines at system operating pressure would result in hardship or unusual difficulty without a compensating increase in the level of quality or safety. provides Revision 1 of 4RR-07 to replace Revision 0, in its entirety. This corrects the information and uses the appropriate Reference to Table IWB-2500-1 and Examination Category B-P. The corrected information also results in the leakage test being required each refueling outage. Accordingly, the alternative required examination will be performed with the frequency specified in Table IWB-2500-1 for a System Leakage Test (once each refueling outage).

Attachment 2 to PLA-7154 Relief Request 4RR-07, Revision 1

RELIEF REQUEST NUMBER: 4RR-07 REVISION 1 (Page 1 of 3)

COMPONENT IDENTIFICATION Code Class: 1

Reference:

Table IWB-2500-1 Examination Category: B-P Item Number: B15.10

==

Description:==

Request for Alternative from System Leakage Testing Reactor Pressure Vessel Head Flange Seal Leak Detection System Component Number: Flange Seal Leak Detection Line Pressure Retaining Components APPLICABLE CODE EDITION AND ADDENDA The Susquehanna Steam Electric Station's will start the 4th 10-Year Inservice Inspection (IS I)

Program Interval on June 1, 2014 and is required to follow the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code,Section XI, "Rules for Inservice Inspection of Nuclear Power Plant Components," (ASME Section XI), 2007 Edition through the 2008 Addenda.

CODE REQIDRE:MENT Table IWB-2500-1 requires a VT-2 visual examination to be performed during a system leakage test each refueling outage. Subparagraph IWB-5221(a) requires that the system leakage test be performed at a pressure not less than the pressure corresponding to 100 percent rated reactor power.

BASIS FOR RELIEF Pursuant to 10CFR 50.55a(a)(3)(ii), relief is requested on the basis that compliance with the Code specified pressure requirement to test the leak-off lines at system operating pressure would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.

The Reactor Pressure Vessel Head Flange Leak Detection Line is separated from the reactor pressure boundary by one passive membrane, a silver plated 0-ring located on the vessel flange. A second 0-ring is located on the opposite side of the tap in the vessel flange (See Figure 4RR-07.1).

This line is required during plant operation in order to indicate failure of the inner flange seal 0-ring. Failure of the inner 0-ring is the only condition under which this line is pressurized.

The configuration of this system precludes manual testing while the vessel head is removed because the odd configuration of the vessel tap (See Figure 4RR-07.1), combined with the small size of the tap and the high test pressure requirement (1035 psig minimum), prevents the tap in the flange from being temporarily plugged. The opening in the flange is only 3116 of an inch in diameter and is smooth walled making a high pressure temporary seal very difficult. Failure of this seal could possibly cause ejection of the device used for plugging into the vessel.

A pneumatic test performed with the head installed is precluded due to the configuration of the top head. The top head of the vessel contains two grooves that hold the 0-rings. The 0-rings are held in place by a series of retainer clips. The retainer clips are contained in a recessed cavity in the top

RELIEF REQUEST NUMBER: 4RR-07 REVISION 1 (Page 2 of 3)

BASIS FOR RELIEF (Continued) head (See Figure 4RR-07.1). If a pressure test was performed from the leak-off line side with the head on, the inner 0-ring would be pressurized in a direction opposite to what it would see in normal operation. This test pressure would result in a net inward force on the 0-ring that would tend to push it into the recessed cavity that houses the retainer clips. The 0-ring material is a thin silver plating and could very likely be damaged by this deformation into the recessed areas on the top head.

In addition to the problems associated with the 0-ring design that preclude this testing it is also questionable whether a pneumatic test is appropriate for this line. Although the line will initially contain steam if the inner 0-ring leaks, the system actually detects leakage rate by measuring the level of condensate in a collection chamber. This would make the system medium water at the level switch. Finally, the use of a pneumatic test performed at a minimum of 1035 psig would represent an unnecessary risk in safety for the inspectors and test engineers in the unlikely event of a test failure, due to the large amount of stored energy contained in air pressurized to 1035 psig.

System leakage testing of this line is precluded because the line will only be pressurized in the event of a failure of the inner 0-ring. It is extremely impractical to purposely fail the inner 0-ring in order to perform a test.

Based on the above, SSES requests relief from the ASME Section XI requirements for system leakage testing of the Reactor Pressure Vessel Head Flange Seal Leak Detection System.

PROPOSED ALTERNATE EXAMINATIONS A VT-2 visual examination will be performed on the accessible portions of the line after the refueling cavity has been filled to its normal refueling water level for at least 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. For sections of the line that are inaccessible for direct VT-2 visual examination, examination will include the surrounding area underneath the piping for evidence of leakage, as permitted by IW A-5241(b). The static head developed due to the water above the vessel flange during flood-up will allow for the detection of any gross indications in the line. This examination will be performed with the frequency specified by Table IWB-2500-1 for a System Leakage Test (once each refueling outage).

DURATION OF PROPOSED ALTERNATIVE This relief will remain in effect for the duration of the Fourth 10 year interval of the lSI Program for SSES Units 1 and 2 (June 1, 2024).

PRECEDENTS None

RELIEF REQUEST NUMBER: 4RR-07 REVISION 1 (Page 3 of 3)

FIGURE 4RR-07.1 FLANGE SEAL LEAK DETECTION LINE DETAIL METALLIC "0" RINGS TO DRYWELL--+

DRILLED PASSAGE REACTOR VESSEL---+