ML082880120
| ML082880120 | |
| Person / Time | |
|---|---|
| Site: | Peach Bottom  (DPR-044, DPR-056) |
| Issue date: | 10/09/2008 |
| From: | Cowan P Exelon Generation Co, Exelon Nuclear |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| TAC MD8296, TAC MD8297 | |
| Download: ML082880120 (5) | |
Text
Exelon Nuclear 200 Exelon Way Kennett Square, PA 19348 October 9,2008 www.exeloncorp.com Exel n
Nuclear 10 CFR 50.55a U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555 Peach Bottom Atomic Power Station, Units 2 and 3 Renewed Facility Operating License Nos. DPR-44 and DPR-56 NRC Docket Nos. 50-277 and 50-278
Subject:
Relief Requests Associated with the Third and Fourth Inservice Inspection (lSI) Intervals and the First and Second Containment Inservice Inspection (CISI) Intervals - Response to Request for Additional Information Concerning Relief Request CRR-13 and 14R-47, and Withdrawal of Relief Request 14R-08
References:
- 1) Letter from P. B. Cowan (Exelon Generation Company, LLC) to U. S. Nuclear Regulatory Commission, "Submittal of Relief Requests Associated with the Third and Fourth Inservice Inspection (lSI) Intervals and the First and Second Containment Inservice Inspection (CISI) Intervals," dated February 29, 2008 2)
Letter from J. D. Hughey (U. S. Nuclear Regulatory Commission) to C. G.
Pardee (Exelon Generation Company, LLC), "Peach Bottom Atomic Power Station, Units 2 and 3 - Request for Supplemental Information Regarding Relief Request 14R-44 (TAC NOS. MD8296 and MD8297)," dated May 7, 2008 3)
Letter from P. B. Cowan (Exelon Generation Company, LLC) to U. S. Nuclear Regulatory Commission, "Response to Request for Supplemental Information Associated with Relief Request 14R-44," dated May 13, 2008 4)
Letter from J. D. Hughey (U. S. Nuclear Regulatory Commission) to C. G.
Pardee (Exelon Generation Company, LLC), "Peach Bottom Atomic Power Station, Units 2 and 3: Request for Additional Information Regarding Relief Request CRR-13 Associated with the First and Second Containment Inservice Inspection Intervals (TAC NOS. MD8308 and MD8309)," dated July 9,2008 5)
Letter from P. B. Cowan (Exelon Generation Company, LLC) to U. S. Nuclear Regulatory Commission, "Relief Requests Associated with the Third and Fourth Inservice Inspection (lSI) Intervals and the First and Second Containment Inservice Inspection (CISI) Intervals - Response to Request for Additional Information Concerning Relief Request CRR-13," dated August 4, 2008
Relief Requests Associated with Third and Fourth lSI Intervals and First and Second CISllntervals October 9,2008 Page 2 6)
Letter from J. D. Hughey (U. S. Nuclear Regulatory Commission) to C. G.
Pardee (Exelon Generation Company, LLC), "Peach Bottom Atomic Power Station, Units 2 and 3: Request for Additional Information Regarding Relief Request CRR-13 Associated with the First and Second Containment Inservice Inspection Intervals (TAC NOS. MD8308 AND MD8309)," dated October 2,2008 7)
Letter from J. D. Hughey (U. S. Nuclear Regulatory Commission) to C. G.
Pardee (Exelon Generation Company, LLC), "Peach Bottom Atomic Power Station, Units 2 and 3: Request for Additional Information Regarding Relief Request 14R-47 Associated with the Fourth Inservice Inspection Interval (TAC NOS. MD8304 AND MD8305)," dated October 2, 2008 In the Reference 1 letter, Exelon Generation Company, LLC (EGC) submitted for your review and approval relief requests associated with the third and fourth Inservice Inspection (lSI) intervals for Peach Bottom Atomic Power Station (PBAPS), Units 2 and 3. Also included for your review and approval were relief requests associated with the first and second Containment Inservice Inspection (CIS!) intervals for PBAPS, Units 2 and 3.
As a result of a conversation with the NRC staff on September 9,2008, the U. S. Nuclear Regulatory Commission Staff requested additional information as discussed in the Reference 6 and 7 letters. Attached is our response to these requests.
Additionally, EGC is withdrawing Relief Request 14R-08 as requested by the NRC, who has verbally informed EGC that its policy is to no longer review this type of relief request in advance of the interval. This withdrawal does not impact the other relief requests.
There are no regulatory commitments contained in this letter.
Should you have any questions concerning this letter, please contact Tom Loomis at (610) 765-5510.
Respectfully, Pamela B. Cowan Director - Licensing & Regulatory Affairs Exelon Generation Company, LLC
Attachment:
1)
Response to Request for Additional Information - Relief Requests CRR-13 and 14R-47 cc:
S. J. Collins, Regional Administrator, Region I, USNRC F. Bower, USNRC Senior Resident Inspector, PBAPS J. Hughey, Project Manager, USNRC S. T. Gray, State of Maryland R. R. Janati, Commonwealth of Pennsylvania Response to Request for Additional Information Relief Requests CRR-13 and 14R-47
RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION REGARDING RELIEF REQUESTS CRR-13 AND 14R-47 PEACH BOTTOM ATOMIC POWER STATION, UNITS 2 AND 3 Question 1:
By letter dated February 29, 2008, (Agencywide Documents Access and Management System (ADAMS) Accession No. ML080640587), as supplemented by letter dated August 4, 2008, (ADAMS Accession No. ML082200279), Exelon Generation Company, LLC, submitted Relief Request CRR-13 associated with the first and second Containment Inservice Inspection intervals for Peach Bottom Atomic Power Station (PBAPS), Units 2 and 3. The Nuclear Regulatory Commission (NRC) staff has identified an apparent discrepancy between the detail drawings provided with the relief request dated February 29, 2008, and the Request for Additional Information (RAI) response dated August 4,2008. Specifically, the note next to the seal weld on Figure CRR-13-2 (Detail E dwg S-53) in relief request CRR-13 states:
SEAL WELD AFTER CONCRETE HAS BEEN POURED OUTSIDE THE DRYWELL (WELDS BY OTHERS)
However, in the RAI response, Note 1 with regard to "Sequence of Concrete Pours Under Drywell Shell" shown in Section A on Drawing S-188 states:
AFTER COMPLETION OF PNEUMATIC TESTS ON THE DRYWELL, SEAL WELD THE MANHOLE IN THE BOTTOM OF THE DRYWELL, INSTALL REBARS AND PLACE CONCRETE POUR #1 INSIDE THE DRYWELL UP TO EL 116 1-0".
Per a similar Note 2 on drawing S-188, Concrete Pour # 2 under (Outside) the drywell is made after Pour #1.
Thus, drawing S-53 indicates that the seal weld was performed after concrete was poured outside the drywell and drawing S-188 indicates that the seal weld was performed before concrete was poured. (Note that dwg. S-53 is also referenced in Section 1 of relief request CRR-13, but drawing S-188 is not).
In order for the NRC staff to complete its evaluation, response to the following RAI question is requested.
RAI-1)
Please explain the discrepancy described above with regard to the seal weld information contained in Figure CRR-13-2 (Detail E drawing S-53) in relief request CRR-13 and that on PBAPS drawing S-188.
Response
A review of the Specification for Reactor Drywell and Suppression Chamber Containment Vessels (6280-C-2) was completed and Section 8.2.1.1 indicates that the note on the drawing S-53 is incorrect. However, neither configuration relieves the station from the code inspection requirements from which Peach Bottom Atomic Power Station, Units 2 and 3, is seeking relief.
As discussed in Relief Request CRR-13, the N-3 manhole was seal welded and cannot meet the IWE-1232(a)(2) code requirement for a double butt weld. Specification 6280-C-2, Section
Response to Request for Additional Information Regarding Relief Requests CRR-13 and 14R-47 Page 2 8.2.1.1 states that the leak and pressure testing for tightness is completed before concrete fill is placed under the drywell, which supports the note on S-188. In addition, drawing S-188 is considered the final as-built drawing for the drywell.
Question 2:
By letter dated February 29, 2008, (Agencywide Documents Access and Management System (ADAMS) Accession No. ML080640587), Exelon Generation Company, LLC, submitted Relief Request (RR) No. 14R-47, related to the Fourth 10-Year lntervallnservice Inspection Program for the Peach Bottom Atomic Power Station, Units 2 and 3. In RR No. 14R-47, the licensee proposed an alternative to perform the Code-required end-of-interval system leakage test of the Control Rod Drive (CRD) pressure boundary during pressurization resulting from Scram Time testing of each CRD. Scram Time testing is routinely completed prior to achieving 40%
reactor power.
The Nuclear Regulatory Commission (NRC) staff has reviewed the request for relief the licensee provided in the February 29,2008, submittal. In order for the NRC staff to complete its evaluation, response to the following request for additional information (RAI) questions is requested.
RAI-1)
During Scram Time testing of the CRD, how long will the CV-2(3)-03A-13-127 valve for each CRD remain open? Specifically, how long will the CRD pressure boundary between the CV-2(3)-03A-13-127 valve and the HV-2(3)-03A-13112 valve remain pressurized at reactor coolant system pressure while a VT-2 visual examination of the pressure boundary is performed?
RAI-2)
Provide an estimate of the rate of pressure loss following pressurization of the pressure boundary between the CV-2(3)-03A-13-127 valve and the HV-2(3)-03A-13112 valve during Scram Time testing.
Response
RAI-1) During scram time testing, the normal length of time that the CV-2(3)-03A-13-127 valve for each CRD will remain open is less than approximately 60 seconds. However, during scram time testing, CV-2(3)-03A-13-127 remains open and the piping between the CV-2(3)-03A 127 valve and the HV-2(3)-03A-13112 valve will have flow through it from the reactor. The reactor will be at normal operating pressure during this time. Once the scram time test is complete the test switch is returned to normal and the CV-2(3)-03A-13-127 valve shuts and the portion of piping depressurizes.
RAI-2) During scram time testing, when the visual (VT-2) inspection is being performed, the CV-2(3)-03A-13-127 valve is open and the HV-2(3)-03A-13112 valve is open. The HV-2(3)-
03A-13112 valves are always open to the Scram Discharge Volume unless the hand valve is manually shut. While the CV-2(3)-03A-13-127 valve is open, the piping being inspected will be at approximately normal reactor pressure with negligible pressure loss since water from the reactor will be flowing through that section of piping until the CV-2(3)-03A-13-127 is manipulated. Once the scram time test is complete, the test switch is returned to normal, the CV-2(3)-03A-13-127 valve shuts, and the line immediately depressurizes as the water volume in the piping drains to the scram discharge volume.