L-13-196, Supplemental Reply to Request for Additional Information for the Review of the License Renewal Application and License Renewal Application Amendment No. 43
| ML13156A388 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 06/03/2013 |
| From: | Lieb R FirstEnergy Nuclear Operating Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| L-13-196 | |
| Download: ML13156A388 (8) | |
Text
FENOCTM 5501 North State Route 2 Oak Harbor,Ohio 43449 FirstEnergyNuclearOperating Company Raymond A. Ueb 419-321-7676 Vice President,Nuclear Fax: 419-321-7582 June 3, 2013 L-1 3-196 10 CFR 54 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001
SUBJECT:
Davis-Besse Nuclear Power Station, Unit No. 1 Docket No. 50-346, License Number NPF-3 Supplemental Reply to Request for Additional Information for the Review of the Davis-Besse Nuclear Power Station, Unit No. 1, License Renewal Application (TAC No. ME4640) and License Renewal Application Amendment No. 43 By letter dated August 27, 2010 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML102450565), FirstEnergy Nuclear Operating Company (FENOC) submitted an application pursuant to Title 10 of the Code of Federal Regulations, Part 54 for renewal of Operating License NPF-3 for the Davis-Besse Nuclear Power Station, Unit No. 1 (Davis-Besse). During a telephone conference call on May 21, 2013, the Nuclear Regulatory Commission (NRC) staff requested clarification regarding the supplemental response to NRC request for additional information (RAI) 4.2.4-1 related to pressure-temperature limits provided by FENOC letter dated November 2, 2012 (ML12311A024).
The Attachment provides the FENOC supplemental reply to the NRC request. The NRC request is shown in bold text followed by the FENOC response. The Enclosure provides Amendment No. 43 to the Davis-Besse License Renewal Application.
A-1(45
Davis-Besse Nuclear Power Station, Unit No. 1 L-1 3-196 Page 2 There are no regulatory commitments contained in this letter. If there are any questions or if additional information is required, please contact Mr. Clifford I. Custer, Fleet License Renewal Project Manager, at 724-682-7139.
I declare under penalty of perjury that the foregoing is true and correct. Executed on June _*?*, 2013.
Sincerely, iaym dA Liebe
Attachment:
Supplemental Reply to Request for Additional Information for the Review of the Davis-Besse Nuclear Power Station, Unit No. 1 (Davis-Besse), License Renewal Application, Section 4.2.4
Enclosure:
Amendment No. 43 to the Davis-Besse License Renewal Application cc: NRC DLR Project Manager NRC Region III Administrator cc: w/o Attachment or Enclosure NRC DLR Director NRR DORL Project Manager NRC Resident Inspector Utility Radiological Safety Board
Attachment L-1 3-196 Supplemental Reply to Request for Additional Information for the Review of the Davis-Besse Nuclear Power Station, Unit No. 1 (Davis-Besse),
License Renewal Application (LRA),
Section 4.2.4 Page 1 of 1 Supplemental Question RAI 4.2.4-1 The Nuclear Regulatory Commission (NRC) staff initiated a telephone conference call with FirstEnergy Nuclear Operating Company (FENOC) on May 21, 2013, to discuss the FENOC supplemental response to NRC request for additional information (RAI) 4.2.4-1 submitted by FENOC letter dated November 2, 2012 (ML12311A024).
In LRA Sections and 4.2.4 and A.2.2.4, a list of the 60-year reactor vessel beltline materials is provided. The NRC staff identified a concern that, due to plant modifications in the future (e.g., power uprate), other materials could experience 52 effective full power years (EFPY) inside surface fluence greater than 1.0E17 neutrons per centimeter squared (n/cm 2 ).
FENOC proposed a wording change to LRA Sections 4.2.4 and A.2.2.4 to read as follows (deleted text is lined-out and added text is underlined):
The 60-year reactor vessel beltline materials are those listed below as feGlows-plus any other that could experience 52 EFPY inside surface fluence greater than 1.0E17 n/cm2 .
Following discussions, both parties agreed that FENOC would submit a supplemental response to RAI 4.2.4-1 to incorporate the proposed changes into LRA Sections 4.2.4 and A.2.2.4, both titled "Pressure-Temperature Limits."
SUPPLEMENTAL RESPONSE RAI 4.2.4-1 LRA Sections 4.2.4 and A.2.2.4 are revised to address the NRC identified concern that, due to plant modifications in the future (e.g., power uprate), other materials beyond those listed could experience 52 EFPY inside surface fluence greater than 1.0E1 7 n/cm 2 .
See the Enclosure to this letter for the revision to the Davis-Besse LRA.
Enclosure Davis-Besse Nuclear Power Station, Unit No. I (Davis-Besse)
Letter L-13-196 Amendment No. 43 to the Davis-Besse License Renewal Application Page 1 of 5 License Renewal Application Sections Affected Section 4.2.4 Section A.2.2.4 The Enclosure identifies the change to the License Renewal Application (LRA) by Affected LRA Section, LRA Page No., and Affected Paragraph and Sentence. The count for the affected paragraph, sentence, bullet, etc. starts at the beginning of the affected Section or at the top of the affected page, as appropriate. Below each section the reason for the change is identified, and the sentence affected is printed in italics with deleted text fined-ou and added text underlined.
Enclosure L-13-196 Page 2 of 5 Affected LRA Section LRA Page No. Affected Paragraph and Sentence 4.2.4 Page 4.2-11 5 th Paragraph Based on the supplemental response to request for additional information (RAI) 4.2.4-1, the 5 th paragraph of LRA Section 4.2.4, "Pressure-Temperature Limits,"
previously revised in FENOC letter dated November 2, 2012 (ML12311A024), is revised to read as follows:
The current P-T limits, generated consistent with the requirements of 10 CFR 50 Appendix G and Regulatory Guide 1.99 Revision 2, are valid until 32 EFPY, or April 22, 2017, whichever occurs first. A revised pressure and temperature limits report will be submitted to the NRC, in accordance with Technical Specification 5.6.4, before Davis-Besse operates beyond 32 EFPY, or April 22, 2017, whichever occurs first, in accordance with the requirements of 10 CFR 50, Appendix G. The revised P-T limits for the period of extended operation will be based on an evaluation of the effects of neutron embrittlement for the 60-year beltline materials, the stresses in the closure head region of the reactor vessel (subject to significant stresses due mechanical loads resulting from bolt preload) and the stresses in the reactor vessel outlet nozzles (largest nozzles in the RCS and the inside corners of the nozzles are subjected to high local stresses produced by pressure). The 60-year reactor vessel beltline materials are those listedas-afelow_&below. plus any other that could experience 52 EFPY inside surface fluence greaterthan 1.0E17 n/cm2 .
- Nozzle Belt Forging (ADB 203)
- Upper Shell Forging (AKJ 233)
- Lower Shell Forging (BCC 241)
" Nozzle Belt Forging to Upper Shell Forging Circumferential Weld (Inside 9%) (WF-232) / (Outside 91%) (WF-233)
- Upper Shell Forging to Lower Shell Forging Circumferential Weld (WF-182-1)
- Reactor Vessel Inlet Nozzle Forgings (BSS 270)
- Reactor Vessel Outlet Nozzle Forgings (ATS 239)
" Dutchman Forging (122Y384VA1)
" Nozzle Belt Forging to Bottom of Reactor Vessel Inlet Nozzle Forging Weld (WF-233 / WF-232)
Enclosure L-13-196 Page 3 of 5 0 Nozzle Belt Forging to Bottom of Reactor Vessel Outlet Nozzle Forging Weld (WF-233) 0 Lower Shell Forging to Dutchman Forging Circumferential Weld (Inside 12%) (WF-232) / (Outside 88%) (WF-233)
Revisions to the P-T limits will be managed as part of the Reactor Vessel Surveillance Program for the period of extended operation.
Enclosure L-13-196 Page 4 of 5 Affected LRA Section LRA Page No. Affected Paragraph and Sentence A.2.2.4 Page A-33 3 rd Paragraph Based on the supplemental response to RAI 4.2.4-1, the 3 rd paragraph of LRA Section A.2.2.4, "Pressure-Temperature Limits," previously revised in FENOC letter dated November 2, 2012 (ML12311A024), is revised to read as follows:
The current P-T limits, generated consistent with the requirements of 10 CFR 50 Appendix G and Regulatory Guide 1.99 Revision 2, are valid until 32 EFPY, or April 22, 2017, whichever occurs first. A revised pressure and temperature limits report (PTLR) will be submitted to the NRC, in accordance with Technical Specification 5.6.4, before Davis-Besse operates beyond 32 EFPY, or April 22, 2017, whichever occurs first, in accordance with the requirements of 10 CFR 50, Appendix G. The revised P-T limits for the period of extended operation will be based on an evaluation of the effects of neutron embrittlement for the 60-year beltline materials, the stresses in the closure head region of the reactor vessel (subject to significant stresses due mechanical loads resulting from bolt preload) and the stresses in the reactor vessel outlet nozzles (largest nozzles in the RCS and the inside corners of the nozzles are subjected to high local stresses produced by pressure). The 60-year reactor vessel be/tline materials are those listed-ais-fl9lows. below, plus any other that could experience 52 EFPY inside surface fluence greaterthan 1.OE1 7 n/cm2 .
" Nozzle Belt Forging (ADB 203)
" Upper Shell Forging (AKJ 233)
" Lower Shell Forging (BCC 241)
" Nozzle Belt Forging to Upper Shell Forging Circumferential Weld (Inside 9%) (WF-232) / (Outside 91 %) (WF-233)
- Upper Shell Forging to Lower Shell Forging Circumferential Weld (WF-1 82-1)
" Reactor Vessel Inlet Nozzle Forgings (BSS 270)
- Reactor Vessel Outlet Nozzle Forgings (ATS 239)
- Dutchman Forging (122Y384VA1)
- Nozzle Belt Forging to Bottom of Reactor Vessel Inlet Nozzle Forging Weld (WF-233 / WF-232)
Enclosure L-13-196 Page 5 of 5
- Nozzle Belt Forging to Bottom of Reactor Vessel Outlet Nozzle Forging Weld (WF-233) 0 Lower Shell Forging to Dutchman Forging Circumferential Weld (Inside 12%) (WF-232) / (Outside 88%) (WF-233)