Letter Sequence Other |
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MONTHYEARLIC-14-0099, Revised Fourth Ten-Year Interval Inservice Inspection (ISI) Interval2014-11-0303 November 2014
Revised Fourth Ten-Year Interval Inservice Inspection (ISI) Interval
Project stage: Request
LIC-15-0066, Relief Request Number RR-14, Request for Relief from Paragraph -3142.1(c) of ASME Code Case N-729-1 for Reactor Vessel Head Penetration Nozzle Welds2015-05-0909 May 2015
Relief Request Number RR-14, Request for Relief from Paragraph -3142.1(c) of ASME Code Case N-729-1 for Reactor Vessel Head Penetration Nozzle Welds
Project stage: Request
ML15131A4252015-05-11011 May 2015
NRR E-mail Capture - FW: Ft Calhoun--relief Request 14 Reactor Vessel Head Inspections (MF6206)
Project stage: Other
LIC-15-0069, Response to NRC Request for Additional Information Regarding Relief Request RR-142015-05-13013 May 2015
Response to NRC Request for Additional Information Regarding Relief Request RR-14
Project stage: Response to RAI
ML15137A0042015-05-15015 May 2015
NRR E-mail Capture - FCS: RAI Number 2 for RR-14, Rvh Leakage/Asme Code Case N-729-1
Project stage: RAI
LIC-15-0070, Response to NRC Request for Additional Information Regarding Relief Request RR-142015-05-16016 May 2015
Response to NRC Request for Additional Information Regarding Relief Request RR-14
Project stage: Response to RAI
LIC-15-0071, Supplemental Information Associated with Omaha Public Power District (OPPD) Relief Request RR-142015-05-17017 May 2015
Supplemental Information Associated with Omaha Public Power District (OPPD) Relief Request RR-14
Project stage: Supplement
ML15138A0122015-05-17017 May 2015
NRR E-mail Capture - Verbal Authorization for Relief Request RR-14
Project stage: Other
ML15147A1552015-05-18018 May 2015
Reactor Vessel Head Penetration Photographs Associated with Omaha Public Power District (OPPD) Relief Request RR-14
Project stage: Other
ML15139A0102015-05-26026 May 2015
Summary of 5/17/15 Telephone Call, Verbal Authorization of Relief Request RR-14, Relief from ASME Code Case N-729-1 Requirements for Reactor Vessel Head Penetration Nozzle Welds, 4th 10-Year Inservice Inspection Interval
Project stage: Other
ML15232A0032015-08-21021 August 2015
Relief Request RR-14, Relief from ASME Code Case N-729-1 Requirements for Reactor Vessel Head Penetration Nozzle Welds, Fourth 10-Year Inservice Inspection Interval
Project stage: Acceptance Review
2015-05-16
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Category:Code Relief or Alternative
MONTHYEARML16104A0742016-04-15015 April 2016
Relief Request, Request to Use a Portion of a Later Edition of the ASME B&PV Code, Section XI, Fourth 10-Year Inservice Inspection Interval
ML16041A3082016-02-19019 February 2016
Relief Requests P-1 - LPSI and CS Pumps; P-2 - Adjusting Hydraulic Parameters Consistent W/Code Case OMN-21; G-1 - Test Frequency Consistent W/Code Case OMN-20, Fifth 10-Year Inservice Testing Interval
LIC-15-0114, Request to Use a Portion of a Later Edition of the ASME B&PV Code, Section XI2015-11-24024 November 2015
Request to Use a Portion of a Later Edition of the ASME B&PV Code, Section XI
LIC-15-0089, Submits Relief Requests Associated with the Fifth Inservice Testing Interval2015-08-27027 August 2015
Submits Relief Requests Associated with the Fifth Inservice Testing Interval
ML15232A0032015-08-21021 August 2015
Relief Request RR-14, Relief from ASME Code Case N-729-1 Requirements for Reactor Vessel Head Penetration Nozzle Welds, Fourth 10-Year Inservice Inspection Interval
LIC-15-0086, Supplement to License Amendment Request (LAR) 14-04, Request to Adopt ASME Code, Section III, 1980 Edition (No Addenda) as an Alternative to Current Code of Record2015-07-0202 July 2015
Supplement to License Amendment Request (LAR) 14-04, Request to Adopt ASME Code, Section III, 1980 Edition (No Addenda) as an Alternative to Current Code of Record
ML15139A0102015-05-26026 May 2015
Summary of 5/17/15 Telephone Call, Verbal Authorization of Relief Request RR-14, Relief from ASME Code Case N-729-1 Requirements for Reactor Vessel Head Penetration Nozzle Welds, 4th 10-Year Inservice Inspection Interval
LIC-15-0066, Relief Request Number RR-14, Request for Relief from Paragraph -3142.1(c) of ASME Code Case N-729-1 for Reactor Vessel Head Penetration Nozzle Welds2015-05-0909 May 2015
Relief Request Number RR-14, Request for Relief from Paragraph -3142.1(c) of ASME Code Case N-729-1 for Reactor Vessel Head Penetration Nozzle Welds
ML14323A5992014-12-0202 December 2014
Relief Request RR-13, Relief from Inservice Testing Requirements to Perform Testing of 4 Valves During the April 2015 Refueling Outage
ML14316A1672014-11-19019 November 2014
Relief Request RR-14, Proposed Alternative for Temporary Acceptance of a Pin Hole Leak in Raw Water System 20-Inch Elbow Located in Room 19 of Auxiliary Building
ML1022101332010-08-20020 August 2010
Request for Use of Alternative to Depth-Sizing Qualification for Volumetric Examinations of Reactor Pressure Vessel Welds for 4th 10-year Inservice Inspection Interval
ML0712300902007-04-27027 April 2007
Part 21 Interim Report, Dresser Investigation File No. 2007-02, Interim Reporting of a Potential Defect Involving Power Actuated Pressure Relief Valves Supplied to Calvert Cliffs, Fort Calhoun and Oconee Plants
ML0634504072007-01-0505 January 2007
Request for Relief Use of Later Edition and Addenda of ASME Code for Examination of Cast Austenitic Stainless Steel Piping
ML0609701242006-04-0606 April 2006
Relief E-2, 4-th 10-year Pump and Valve Inservice Testing Program
ML0519600742005-10-0303 October 2005
Relief Request - Alternative Test Requirements for Containment Repairs
ML0514407352005-05-24024 May 2005
5/24/05 Fort Calhoun - Relaxation Request from U.S. NRC Order EA-03-009 for the Control Element Drive Mechanism Nozzles
ML0501203572005-02-28028 February 2005
Request for Relief from ASME Code Repair Requirements and Using an Alternative for the Pressurizer Nozzle Repair.
LIC-04-0008, Relief Request Pertaining to Reactor Vessel Nozzle Inspections for the Third 10-Year Interval, Revision2004-04-0202 April 2004
Relief Request Pertaining to Reactor Vessel Nozzle Inspections for the Third 10-Year Interval, Revision
ML0326000132003-09-12012 September 2003
Relief Request - Third and Fourth 10-Year Interval Inservice Inspection Program Plan - Request for Relief RR-8
LIC-03-0062, Relief Request Pertaining to Visual Inspection of Inaccessible Piping & Components2003-05-0101 May 2003
Relief Request Pertaining to Visual Inspection of Inaccessible Piping & Components
LIC-02-0142, Relief Requests Pertaining to the Fort Calhoun Inservice Inspection (ISI) of the Reactor Pressure Vessel (RPV) for the Third Ten Year ISI Interval (1993-2003)2002-12-20020 December 2002
Relief Requests Pertaining to the Fort Calhoun Inservice Inspection (ISI) of the Reactor Pressure Vessel (RPV) for the Third Ten Year ISI Interval (1993-2003)
2016-04-15
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Text
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 May 26, 2015 LICENSEE: Omaha Public Power District FACILITY: Fort Calhoun Station, Unit 1
SUBJECT:
SUMMARY
OF TELEPHONE CONFERENCE ON MAY 17, 2015, TO AUTHORIZE VERBAL RELIEF FOR RELIEF REQUEST 14 (TAC NO. MF6206)
This memorandum summarizes the telephone discussion on May 17, 2015, between the U.S. Nuclear Regulatory Commission (NRC) staff and Omaha Public Power District (OPPD, the licensee) staff regarding the licensee's request for relief RR-14 for Fort Calhoun Station, Unit 1 (FCS). Participants in the discussion included L. Cortopassi, J. Wilson, W. Hansher, et al.
(OPPD), and M. Markley, D. Alley, F. Lyon, et al. (NRC).
By letter to the NRC dated May 9, 2015 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML15129A004), as supplemented by letters dated May 13, 16, and 17, 2015 (ADAMS Accession Nos. ML15135A387, ML15136A002, and ML15142A411, respectively), OPPD submitted RR-14 for the inspection of reactor vessel head nozzles at FCS.
In RR-14, the licensee proposed to use alternative inspection requirements for reactor vessel head nozzles with respect to American Society of Mechanical Engineers (ASME) Code Case N-729-1, "Alternative Examination Requirements for PWR [Pressurized-Water Reactor]
Reactor Vessel Upper Heads With Nozzles Having Pressure-Retaining Partial-Penetration WeldsSection XI, Division 1," as conditioned in Title 10 of the Code of Federal Regulations (10 CFR) paragraph 50.55a(g)(6)(ii)(D) until the end of operating cycle 28 or until a degraded reactor vessel head nozzle is detected.
The NRC staff reviewed the licensee's results of current bare metal visual examinations of the reactor vessel head nozzles, the potential of nozzle ejection and reactor vessel head degradation resulting from the boric acid corrosion, and the proposed inspection plan that the licensee will be performing in the next refueling outage in fall 2016.
The NRC staff concludes that:
- 1. The licensee has demonstrated that nozzle ejection and reactor vessel head degradation are not likely in the next fuel cycle.
- 2. The licensee's bare metal visual examinations did not identify any areas of significant corrosion,
- 3. The licensee has demonstrated that there was an alternate possible source other than nozzle leakage for the relevant condition for each of the nozzles for which relief is requested.
- 4. The licensee's chemistry analysis provided some additional supporting information for leakage sources other than possible nozzle leakage
- 5. The licensee will use administrative controls such that at an unidentified leak rate increase of greater than 0.1 gallons per minute above stable baseline, actions will be taken to identify the source of leakage. If the source is not identified within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, actions will be taken to shut down the plant.
- 6. The licensee will perform a bare metal visual examination of all reactor vessel head nozzles in accordance with ASME Code Case N-729-1 on the first cold shutdown of greater than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> that occurs after at least 4 months of operation.
- 7. For the fall 2016 inspection, the licensee will perform bare metal visual examinations in combination with ultrasonic examinations or surface examinations of all reactor vessel head nozzles in accordance with ASME Code Case N-729-1 as conditioned in 10 CFR 50.55a(g)(6)(ii)(D).
Based on the above, the NRC staff has determined that the proposed alternative inspection performed in spring 2015 provides reasonable assurance that the structural integrity of the reactor vessel head and attached nozzles will be maintained until the next refueling outage, which is scheduled for fall 2016.
The NRC staff concludes that RR-14 will provide reasonable assurance of the structural integrity of the reactor vessel head and attached nozzles. The NRC staff concludes that complying with the specified inspection in accordance with ASME Code Case N-729-1 would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.
Accordingly, the NRC staff concludes that the licensee has adequately addressed all of the regulatory requirements set forth in 10 CFR 50.55a(z)(2) and is in compliance with the requirements of the ASME Code,Section XI, Code Case N-729-1 as conditioned by 10 CFR 50.55a(g)(6)(ii)(D). Therefore, on May 17, 2015, as authorized by David Alley, Chief, Component Performance, Non-Destructive Examination, and Testing Branch, and Michael Markley, Chief, Plant Licensing Branch IV-1, Office of Nuclear Reactor Regulation, the NRC authorizes the use of RR-14 at FCS until the end of operating cycle 28, scheduled for fall 2016, or until a degraded reactor vessel head nozzle is detected.
All other requirements of ASME Code,Section XI, and 1O CFR 50.55a(g)(6)(ii)(D) for which relief was not specifically requested and authorized by the NRC staff remain applicable, including the third-party review by the Authorized Nuclear lnservice Inspector.
This verbal authorization does not preclude the NRC staff from asking additional clarification questions regarding RR-14 while preparing the subsequent written safety evaluation.
If you have any questions, please contact me at (301) 415-2296 or by e-mail at fred.lyon@nrc.gov.
Carl F. Lyon, Project Manager Plant Licensing Branch IV-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-285 cc: Distribution via Listserv
ML15139A010 OFFICE NRR/DORL/LPL4-1 /PM NRR/DORL/LPL4-1/LA NRR/DORL/LPL4-1 /BC NRR/DORL/LPL4-1/PM NAME Fl yon JBurkhardt MMarkley FL yon DATE 5/26/15 5/19/15 5/26/15 5/26/15
