ML093240496
| ML093240496 | |
| Person / Time | |
|---|---|
| Site: | Susquehanna  |
| Issue date: | 11/20/2009 |
| From: | Pickett D Plant Licensing Branch 1 |
| To: | Crowthers M PP&L |
| Pickett , NRR/DLPM, 415-1364 | |
| References | |
| RR-07, TAC ME2629 | |
| Download: ML093240496 (4) | |
Text
From: Pickett, Douglas Sent: Friday, November 20, 2009 3:53 PM To: 'Crowthers, Michael H' Cc: Tingen, Steve; Salgado, Nancy; Kim, James; Jackson, Donald; Nelson, Robert; Vaidya, Bhalchandra
Subject:
Susquehanna Unit No. 1 Verbal Authorization for RR-07 Mr. Crowthers:
Earlier today, the Nuclear Regulatory Commission provided verbal authorization for Relief Request RR-07 for the Susquehanna Steam Electric Station Unit No. 1 as requested in your letter dated November 20, 2009. Our letter transmitting the written safety evaluation will be issued in the near future.
Authorizing the approval was Steven Tingen, Acting Branch Chief for the Component Performance & Testing Branch of the Division of Component Integrity of the Office of Nuclear Reactor Regulation and myself as Acting Branch Chief of Plant Licensing Branch I-1, Division of Operating Reactor Licensing of the Office of Nuclear Reactor Regulation.
As part of documenting the verbal authoriation, I am including the script that Mr. Tingen and I read. Please let me know if you have any questions.
A copy of this email will be made publicly available in ADAMS.
Douglas V. Pickett, Senior Project Manager Calvert Cliffs Nuclear Power Plant, R.E. Ginna Nuclear Power Plant U.S. Nuclear Regulatory Commission 301-415-1364 Email: douglas.pickett@nrc.gov This is Steven Tingen, Acting Branch Chief for the Component Performance & Testing Branch of the Division of Component Integrity of the Office of Nuclear Reactor Regulation.
By letter dated November 20, 2009, PPL Susquehanna, LLC (the licensee), requested that the Nuclear Regulatory Commission (NRC) authorize the use of Alternative Request RR-07 for the Susquehanna Steam Electric Station, Unit 1 (Susquehanna SES). The licensee requested authorization to use an alternative inservice test (IST) provision in the American Society of Mechanical Engineers (ASME), Code for Operation and Maintenance of Nuclear Power Plants (OM Code). The licensee requested authorization to extend the test interval for Susquehanna SES Unit 1 main steam safety relief valves (SRVs) PSV141F013A and PSV141F013C beyond six years on a one-time basis until the March 2010 refueling outage.
ASME OM Code Mandatory Appendix I, Paragraph I-1330(a), Test Frequencies, Class 1 Pressure Relief Valves, 1998 Edition with the 1999 and 2000 Addenda, requires that Class 1 pressure relief valves be tested at least once every five years. ASME Code Interpretation 01-18, AASME OM Code-1995 With OMa ASME Code-1996 Addenda, Appendix I,@ dated June 26, 2003, clarifies that the 5-year test interval starts when the valve is tested.
In a letter dated March 10, 2005, the NRC authorized the alternative in Request RR-02 for the Susquehanna SES third 10-year IST program interval. The alternative in Request RR-02
extended the test interval for Unit 1 main steam SRVs to six years or every three refueling outages. Refueling outages are every 24 months at Susquehanna SES.
The 6-year test interval for SRVs PSV141F013A and PSV141F013C will expire in November 2009. The licensee requested authorization in Alternative Request RR-07 to extend the test interval beyond six years on a one-time basis until the March 2010 refueling outage for these SRVs. The licensee=s practice was to test a SRV and then place it in storage before installing the SRV in the unit. The licensee considered the start of the 6-year test interval to begin once the SRV was installed and storage time was not included in the 6-year interval.
The licensees justification for extending the test interval for PSV141F013A and PSV141F013C beyond 6 years is that SRV as-found set pressure test data demonstrates the current maintenance practices have been effective, and that storage has had no significant impact on SRV test results.
The ASME developed Code Case OMN-17, Alternative Rules for Testing ASME Class 1 Pressure Relief/Safety Valves. The ASME has approved and plans to publish OMN-17 in the upcoming edition/addenda of the ASME OM Code. Code Case OMN-17 allows extension of the test frequency for SRVs from five years to six years plus a 6-month grace period. The code case imposes a special maintenance requirement to disassemble and inspect each SRV to verify that parts are free from defects resulting from the time-related degradation or maintenance-induced wear prior to the start of the extended test frequency. Similar to the special maintenance requirement in Code Case OMN-17, the licensee stated that main steam SRVs PSV141F013A and PSV141F013C were disassembled to perform inspection and maintenance activities prior to the start of the current 6-year test interval. All adverse conditions were corrected and each SRV was reassembled.
The NRC staff finds that extending the 6-year test interval for main steam SRVs PSV141F013A and PSV141F013C for up to six months is acceptable on a one time-basis until the March 2010 refueling outage. This test interval is consistent with the provisions of ASME Code Case OMN-17 which allows a test interval of six years plus a 6-month grace period. Extending the test interval for up to approximately six months until the March 2010 refueling outage should not adversely affect the operational readiness of the SRVs.
The licensee has adequately addressed all the regulatory requirements set forth in 10 CFR 50.55a(a)(3)(i). The NRC staff authorizes the alternative in Request RR-07 for Unit 1 Susquehanna SES main steam SRVs PSV141F013A and PSV141F013C on a one-time basis until the March 2010 refueling outage.
This is Douglas Pickett, Acting Branch Chief of Plant Licensing Branch I-1 of the Division of Operating Reactor Licensing of the Office of Nuclear Reactor Regulation.
Reading from NRR Office Instruction LIC-102, Relief Request Reviews:
On rare occasions, the NRC staff may grant verbal authorizations as an alternative under 10 CFR 50.55a(a)(3) when, due to unforeseen circumstances, licensees need NRC authorization before the staff is able to issue its written SE. Temporary verbal authorization for an alternative under 10 CFR 50.55a(a)(3) is subject to the following:
- The proposed alternative is in writing and all information that the staff requires to write the SE has been docketed,
- An identified need for the verbal authorization is recognized given the circumstances of the licensees request,
- The NRC technical staff has completed its review and determined that the proposed alternative is technically justified, but has not yet formally documented it in an SE, and
- The technical branch and the Division of Operating Reactor Licensing branch chiefs have agreed to the verbal authorization.
The above items from LIC-102 have been satisfied. Therefore, Mr. Tingen and I both concur that Alternative Request RR-07 is authorized.
E-mail Properties Mail Envelope Properties ()
Subject:
Susquehanna Unit No. 1 Verbal Authorization for RR-07 Sent Date: 11/20/2009 3:38:01 PM Received Date: 11/20/2009 3:53:00 PM From: Pickett, Douglas Created By: Douglas.Pickett@nrc.gov Recipients:
mhcrowthers@pplweb.com ('Crowthers, Michael H')
Tracking Status: None Steve.Tingen@nrc.gov (Tingen, Steve)
Tracking Status: None Nancy.Salgado@nrc.gov (Salgado, Nancy)
Tracking Status: None James.Kim@nrc.gov (Kim, James)
Tracking Status: None Donald.Jackson@nrc.gov (Jackson, Donald)
Tracking Status: None Robert.Nelson@nrc.gov (Nelson, Robert)
Tracking Status: None Bhalchandra.Vaidya@nrc.gov (Vaidya, Bhalchandra)
Tracking Status: None Post Office:
Files Size Date & Time MESSAGE 25480 11/20/2009
Options Expiration Date:
Priority: olImportanceNormal ReplyRequested: False Return Notification: False Sensitivity: olNormal Recipients received: