View source for TMI-09-105, Response to June 25, 2009, Public Meeting Between the U.S. Nuclear Regulatory Commission and the Pressurized Water Reactor Owners Group (PWROG) Related to PWROG Topical Report BAW-2374, Revision 2

{{Adams
| number = ML092470426
| issue date = 09/04/2009
| title = Response to June 25, 2009, Public Meeting Between the U.S. Nuclear Regulatory Commission and the Pressurized Water Reactor Owners Group (PWROG) Related to PWROG Topical Report BAW-2374, Revision 2
| author name = Cowan P B
| author affiliation = Exelon Generation Co, LLC, Exelon Nuclear
| addressee name = 
| addressee affiliation = NRC/Document Control Desk, NRC/NRR
| docket = 05000289
| license number = DPR-050
| contact person = 
| case reference number = TAC MD7404, TAC ME1797, TMI-09-105
| document type = Letter type:TMI
| page count = 4
| project = TAC:MD7404, TAC:ME1797
| stage = Meeting
}}

=Text=
{{#Wiki_filter:Exelon Nuclear 200 Exelon Way Kennett Square, PA 19348 TMI-09-105 September 4,2009 www.exeloncorp.com Nuclear U.S.Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Three Mile Island Nuclear Station, Unit 1 Facility Operating License No.DPR-50 NRC Docket No.50-289 

==Subject:==
Three Mile Island Unit 1-Response to June 25, 2009, Public Meeting Between the U.S.Nuclear Regulatory Commission and the Pressurized Water Reactor Owners Group (PWROG)Related to PWROG Topical Report BAW-2374, Revision 2.

==References:==

1.Letter from P.Bamford (U.S.Nuclear Regulatory Commission) to C.Pardee (Exelon Generation Company, LLC),"Three Mile Island, Unit No.1-Individual Plant Actions Re: Pressurized-Water Reactor Owners Group Topical Report BAW-2374, Revision 2,'Risk-Informed Assessment of Once-Through Steam Generator Tube Thermal Loads Due to Breaks in Reactor Coolant System Upper Hot Leg Large-Bore Piping'(T AC NO.ME1797)," dated August 5, 2009 2.NRC Summary of the June 25, 2009, Category 2 Public Meeting between the U.S.Nuclear Regulatory Commission and the Pressurized Water Reactor Owners Group (PWROG)to discuss issues related to PWROG Topical Report BAW-2374, Revision 2, Risk-Informed Assessment of Once-Through Steam Generator Tube Thermal Loads due to Breaks in Reactor Coolant System Upper Hot Leg Large-Bore Piping (TAC NO.MD7404), dated July 27,2009 On June 25, 2009, a public meeting was held between the U.S.Nuclear Regulatory Commission (NRC)staff and representatives of the Pressurized Water Reactor Owners Group (PWROG)at NRC Headquarters, One White Flint North, 11555 Rockville Pike, Rockville, Maryland.The purpose of the meeting was to discuss issues related to PWROG Topical Report (TR)BAW-2374, Revision 2,"Risk-informed Assessment of Once-Through Steam Generator Tube Thermal Loads Due to Breaks in Reactor Coolant System Upper Hot Leg Large-Bore Piping." A meeting summary was published by the NRC for this meeting (Reference 2).In summary, the NRC meeting with the PWROG discussed Topical Report BAW-2374, Revision 2, and the ability of the affected steam generators (SGs)to withstandloadsfrom a postulated design basis hot leg Loss-of-Coolant Accident (LOCA)on a plant-specific basis.During the meeting, the NRC staff and PWROG agreed to a number of steps to bring this issue to resolution.
Pursuant to that agreement, the NRC issued a letter to Three Mile Island Unit 1 (TMI Unit 1)(Reference 1)requesting a written response to the list of actions taken from the meeting summary of the June 25, 2009, meeting (detailed in Reference 2)within 30 days.
u.s.Nuclear Regulatory Commission September 4, 2009 Page 2 In response, TMI Unit 1 is submitting this letter confirming the following information is applicable and factual with regards to TMI Unit 1.*The justification for continued operation for addressing tube integrity following a Large Break Loss-of-Coolant Accident (LBLOCA), based on Engineering Evaluation 00184675 dated June 14, 2000, remains valid.*Operator Actions (compensatory measures), including changes to emergency operating procedures, have been incorporated into plant procedures and operator training has been performed.
*10 CFR 50.46 (a)(3)reporting requirements have been satisfied.
*All LBLOCAs (including those in the hot leg region)are considered as design basis accidents intheassessments of Steam Generator (SG)tube integrity following each SG tube inspection.
In addition to these statements, Exelon Generation Company, LLC (Exelon)makes the following two (2)new commitments involving TMI Unit1: 1.An analysis will be performed to confirm that the design of the replacement SGs is sufficient to withstand the loads associated with a LBLOCA including the thermal loads associated with a LBLOCA in the hot leg region of the reactor coolant system and the results of that analysis will be provided to the NRC by January 31, 2010.2.Provide the NRC with the structural Iimit(s)associated with the most limiting LBLOCA for the replacement SGs as part of the next SG tube inspection report (required by the technical specifications) following completion of the next inspection of the tubes in the replacement SGs, scheduled for June 30, 2012.These commitments are also identified in the Attachment to this letter.Should you have any questions concerning this letter, please contact Ms.Wendy Croft at (610)765-5726.Respectfully, Pamela B.Cowan Director-Licensing and Regulatory Affairs Exelon Generation Company, LLC 

==Attachment:==

Summary of Regulatory Commitments cc: S.J.Collins, Administrator, Region I, USNRC D.M.Kern, USNRC Senior Resident Inspector, TMI Unit 1 P.J.Bamford, USNRC Project Manager, TMI Unit 1 ATTACHMENT Three Mile Island Unit 1 Summary of Regulatory Commitments ATTACHMENT The following table identifies commitments made in this document.(Any other actions discussed in the submittal represent intended or planned actions.They are described to the NRC for the NRC's information and are not regulatory commitments.)
COMMITTED COMMITMENT TYPE COMMITMENT DATE OR One-Time Action Programmatic 1I0UTAGEII (Yes/No)(Yes/No)An analysis will be performed to confirm that January 31, 2010 Yes No the design of the replacement SGs is sufficient to withstand the loads associated with a LBLOCA including the thermal loads associated with a LBLOCA in the hot leg region of the reactor coolant system and the results of that analysis will be provided to the NRC.Provide the NRC with the structurallimit(s)
June 30, 2012 Yes No associated with the most limiting LBLOCA for the replacement SGs as part of the next SG tube inspection report (required by the technical specifications) following completionofthe next inspectionofthe tubes in the replacement SGs.Page 1 of 1}}