ML12270A249
| ML12270A249 | |
| Person / Time | |
|---|---|
| Site: | Grand Gulf  |
| Issue date: | 10/03/2012 |
| From: | Ferrer N License Renewal Projects Branch 1 |
| To: | Mike Perito Entergy Operations |
| Ferrer N, 415-1045 | |
| References | |
| TAC ME7493 | |
| Download: ML12270A249 (7) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 October 3, 2012 Mr. Michael Perito Vice President, Site Entergy Operations, Inc.
P.O. Box 756 Port Gibson, MS 39150
SUBJECT:
REQUESTS FOR ADDITIONAL INFORMATION FOR THE REVIEW OF THE GRAND GULF NUCLEAR STATION LICENSE RENEWAL APPLICATION (TAC NO. ME7493)
Dear Mr. Perito:
By letter dated October 28, 2011, Entergy Operations, Inc., submitted an application pursuant to Title 10 of the Code of Federal Regulations Part 54, to renew the operating license for Grand Gulf Nuclear Station, Unit 1, for review by the U.S. Nuclear Regulatory Commission (NRC or the staff). The staff is reviewing the information contained in the license renewal application and has identified, in the enclosure, areas where additional information is needed to complete the review.
These requests for additional information were discussed with Jeff Seiter, and a mutually agreeable date for the response is within 30 days from the date of this letter. If you have any questions, please contact me at 301-415-1045 or bye-mail at nathanieIJerrer@nrc.gov.
Sincerely,
~2 Nathaniel Ferrer, Project Manager Projects Branch 1 Division of License Renewal Office of Nuclear Reactor Regulation Docket No. 50-416
Enclosure:
Requests for Additional Information cc w/encl: Listserv
GRAND GULF NUCLEAR STATION LICENSE RENEWAL APPLICATION REQUESTS FOR ADDITIONAL INFORMATION - SET 38 RAI4.3-5b Background. In its response to follow-up request for additional information (RAI) 4.3-5a dated September 4, 2012, Entergy Operations, Inc. (the applicant) provided a table that lists its non-Class 1 expansion joints. For several of these expansion joints under the evaluation column the applicant provided the usage factor associated with the design cycles for the non-Class 1 expansion joints. These expansion joints include the standby liquid control (at the pump discharge), standby diesel generators (turbocharger water outlets), high pressure core spray (standby service water supply to diesel, diesel air start and diesel fuel oil) and the leakage detection and controls system (exhaust blower inlet). As part of the evaluation, the applicant stated that based on the usage factor associated with the design cycles these expansion joints are acceptable for many more cycles than specified.
Furthermore, for several other expansion joints under the design cycles column the applicant provided the number of cycles for which the specific expansion joint was designed. In addition, under the evaluation column the applicant provided the number of cycles that a particular expansion joint was "qualified" for. These expansion joints include the standby liquid control (at tank outlet and pump inlets and at test tank), high pressure core spray (diesel exhaust) and compressed air (air accumulators). As part of the evaluation, the applicant stated that these expansion joints were qualified or determined to a number of cycles that is greater than the design cycles; therefore, the expansion joints are acceptable for many more cycles than specified.
Issue. Although the expansion joints may have a "small" usage factor when compared to the design limit of 1.0, the U.S. Nuclear Regulatory Commission (NRC or the staff) noted that a "small" usage factor does not support the disposition that the analysis remains valid for the period of extended operation (i.e., 10 CFR 54.21 (c)(1)(i)). Similarly, the staff noted a component being "qualified" for many more cycles compared to the design cycles does not support the disposition that the analysis, which is based on the number of design cycles, remains valid for the period of extended operation (Le., 10 CFR 54.21(c)(1)(i)).
In order for the time-limited aging analyses of these expansion joints to remain valid for the period of extended operation in accordance with 10 CFR 54.21(c)(1)(i), the applicant must demonstrate that the number of "design cycles" used in the current design analysis will not be exceeded after 60 years of operation (e.g., claiming a design cycle limit of 1200 thermal cycles for the high pressure core spray diesel exhaust joint without demonstrating that plant operation would not exceed this 1200 cycles is not an adequate justification under 10 CFR 54.21(c)(1 )(i)).
The staff noted that for the expansion jOints in the standby liquid control system and high pressure core spray system, the license renewal application (LRA) does not provide AMR results for these expansion joints subject to "cracking-fatigue."
ENCLOSURE
- 2 Request.
- a. Supplement RAI 4.3-5a response to provide, for non-Class 1 expansion joints discussed above, adequate justification for the disposition of 10 CFR 54.21 (c)(1 )(i), by demonstrating that the design number of cycles used in the original analysis will not be exceeded during the period of extended operation (Le., cycles from plant operation will not exceed the number of design cycles). Alternatively, revise the TLM disposition and include adequate justification.
- b. Provide the AMR results for all non-Class 1 expansion joints in accordance with 10 CFR 54.21(a)(1) or justify why it is not necessary to provide the AMR results.
RAI B.1.9-4a Backqround. By letter dated August 15, 2012, the applicant responded to RAI B.1.9-4 that addresses the applicant's recent operating experience related to an indication in one of the residual heat removal system to reactor pressure vessel nozzles (N06B-KB weld).
In its response, the applicant stated that due to industry concerns with cracking in dissimilar metal welds, the industry committed to an accelerated inspection program and details of this program are provided in Boiling Water Reactor Vessel and Internals Project (BWRVIP)-222, "Accelerated Inspection Program for BWRVIP-75-A Category C Dissimilar Metal Welds Containing Alloy 182," July 2009. The applicant also stated that during 2012, the nozzle discussed in RAJ B.1.9-4 received extensive weld crown reduction and surface preparation which enabled the examination to detect the subject flaw.
Issue. It is not clear to the staff how the inspection schedule for Category C dissimilar metal welds was accelerated and what guidance document is used for the inspection method. In addition, the LRA does not identify the implementation of the accelerated inspections, which are specified in BWRVIP-222, as an enhancement to the existing BWR Stress Corrosion Cracking Program.
The staff is also not clear if the applicant's program includes any other weld that has limited inspection coverage as was the case with Weld N06B-KB prior to the weld crown reduction.
Request.
- a. Describe how the inspection schedule for Category C dissimilar metal welds was accelerated. As part of the response, clarify whether the inspection schedule in the accelerated inspections meets or is more conservative than the inspection schedule that would be used in accordance with BWRVIP-75-A In addition, describe the inspection method and the reference of the guidance document for the inspection method.
- b. Justify why the LRA does not identify the implementation of the accelerated inspection program, which is described in BWRVIP-222, as an enhancement to the existing BWR Stress Corrosion Cracking Program.
- 3
- c. Clarify whether any other welds in the scope of the program have limited inspection coverage as was the case with Weld N06B-KB prior to the weld crown reduction. If so, justify why the limited inspection coverage is acceptable to manage cracking of such welds.
RAI B.1.11-6 Background. By letter dated August 15, 2012, the applicant responded to RAI B.1.9-2a that, in part, addresses the applicant's aging management for the stainless steel and nickel alloy thermal sleeves and thermal sleeve extensions of reactor vessel nozzles [recirculation inlet, core spray inlet, and residual heat removal (RHR)lIow-pressure coolant injection (LPCI) nozzles].
In its response, the applicant amended the LRA and indicated that the BWRVIP, along with the Water Chemistry Control - BWR Program, is credited to manage cracking due to stress corrosion cracking and intergranular stress corrosion cracking in the thermal sleeves and thermal sleeve extensions of the reactor vessel nozzles. The applicant also stated the recirculation inlet, core spray inlet, and RHR/LPCI nozzles are in part, formed by the internal leg of the V-shaped safe ends for those nozzles.
As described in Appendix C of the LRA, the applicant's response to Action Item NO.5 of BWRVIP-42-A states that the BWRVIP has developed strategies to ensure the integrity of inaccessible welds [associated with the LPCI nozzle and thermal sleeve]. The applicant also stated that these strategies are included in Section 3 of BWRVIP-42, Revision 1 and it has committed to programs described as necessary in the BWRVIP reports to manage the effects of aging during the period of extended operation.
Issue. In its review, the staff noted that the applicant's AMR line items for the thermal sleeve components in LRA Table 3.1.2-1 do not include the thermal sleeve and thermal sleeve extension of the reactor vessel feedwater nozzle. In addition, neither the LRA or applicant's RAI response provide a clear description of the "strategies" in BWRVIP-42, Revision 1 to manage the inaccessible thermal sleeve welds of the LPCI nozzle and other reactor vessel nozzles as applicable.
The staff also needs additional information to clarify how the applicant will inspect the thermal sleeves and thermal sleeve extensions. In case inspections are not performed on the thermal sleeve components, leakage analyses are necessary to ensure that the intended functions of the thermal sleeve components are adequately maintained in a consistent manner with the guidance in Section 3.2.4 of BWRVIP-18-A.
Request.
- a. Clarify why the applicant's AMR line items for the thermal sleeve components in LRA Table 3.1.2-1 do not include the thermal sleeve and thermal sleeve extension of the reactor vessel feedwater nozzle.
- b. Describe the "strategies" in BWRVIP-42, Revision 1 to clarify how the implementation of the strategies will manage cracking of the inaccessible thermal sleeve components of
-4 the LPCI nozzle and other reactor vessel nozzles as applicable. Include the following in the description, as applicable:
- 1. If the program includes leakage analyses to manage cracking of the thermal sleeve components (e.g., for inaccessible locations), describe the results of the leakage analyses to demonstrate that cracking of the thermal sleeve components does not affect the intended functions of these components.
- 2. If the program includes inspections to manage cracking of the thermal sleeve components, describe the method and frequency of the inspections to demonstrate the adequacy of the inspections. As part of the response, clarify whether any of the thermal sleeve welds of the recirculation inlet, core spray, RHR/LPCI and feedwater nozzles can be examined using ultrasonic testing that is applied on the outer surface of the associated piping and safe ends.
- c. Ensure that the LRA is consistent with the response.
October 3,2012 Mr. Michael Perito Vice President, Site Entergy Operations, Inc.
P.O. Box 756 Port Gibson, MS 39150
SUBJECT:
REQUESTS FOR ADDITIONAL INFORMATION FOR THE REVIEW OF THE GRAND GULF NUCLEAR STATION LICENSE RENEWAL APPLICATION (TAC NO. ME7493)
Dear Mr. Perito:
By letter dated October 28, 2011, Entergy Operations, Inc., submitted an application pursuant to Title 10 of the Code of Federal Regulations, Part 54, to renew the operating license for Grand Gulf Nuclear Station, Unit 1, for review by the U.S. Nuclear Regulatory Commission (NRC or the staff). The staff is reviewing the information contained in the license renewal application and has identified, in the enclosure, areas where additional information is needed to complete the review.
These requests for additional information were discussed with Jeff Seiter, and a mutually agreeable date for the response is within 30 days from the date of this letter. If you have any questions, please contact me at 301-415-1045 or bye-mail at nathaniel.ferrer@nrc.gov.
Sincerely, IRA!
Nathaniel Ferrer, Project Manager Projects Branch 1 Division of License Renewal Office of Nuclear Reactor Regulation Docket No. 50-416
Enclosure:
Requests for Additional Information cc w/encl: Listserv DISTRIBUTION: See next page ADAMS Accession No' ML12270A249 OFFICE PM:RPB1 :DLR LA:RPB1 :DLR BC:RPB1 :DLR PM: RPB1:DLR NAME NFerrer IKing DMorey NFerrer DATE 10102/12 09/28/12 10103/12 10103/12 i
OFFICIAL RECORD COpy
Letter to Michael Perito from Nathaniel Ferrer dated October 3, 2012
SUBJECT:
REQUESTS FOR ADDITIONAL INFORMATION FOR THE REVIEW OF THE GRAND GULF NUCLEAR STATION, LICENSE RENEWAL APPLICATION DISTRIBUTION:
E-MAIL:
PUBLIC RidsNrrDlr Resource RidsNrrDlrRpb1 Resource RidsNrrDlrRpb2 Resource RidsNrrDlrRarb Resource RidsNrrDlrRapb Resource RidsNrrDlrRasb Resource RidsNrrDlrRerb Resource RidsNrrDlrRsrg Resource NFerrer DDrucker DWrona DMorey AWang RSmith, RIV BRice, RIV GPick, RIV DMclntyre,OPA