ML14232A023
| ML14232A023 | |
| Person / Time | |
|---|---|
| Site: | Waterford  |
| Issue date: | 10/21/2014 |
| From: | Wang A Plant Licensing Branch IV |
| To: | Entergy Operations |
| Wang A | |
| References | |
| TAC MF3247 | |
| Download: ML14232A023 (5) | |
Text
UNITED STATES
. NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 Vice President, Operations Entergy Operations, Inc.
Waterford Steam Electric Station, Unit 3 17265 River Road Killona, LA 70057-3093 October 2~. 2014
SUBJECT:
WATERFORD STEAM ELECTRIC STATION, UNIT 3-REQUEST FOR ADDITIONAL INFORMATION REGARDING THE REACTOR VESSEL INTERNALS AGING MANAGEMENT PROGRAM (TAC NO. MF3247)
Dear Sir or Madam:
On February 5, 2005, Entergy Operations Inc., the licensee, as part of its implementation of the extended power uprate, made a commitment to submit an aging management program (AMP) for the reactor vessel internals (RVI) components at Waterford Steam Electric Station, Unit 3 (WF3).
Consistent with this commitment, by letter dated December 16, 2013, Entergy submitted an AMP for the RVI components at WF3. The Electric Power Research Institute's technical report MRP-227-A, "Materials Reliability Program: Pressurized Water Reactor (PWR) Internals Inspection and Evaluation Guidelines," December 2008, and its supporting reports were used as the technical bases for developing WF3's AMP.
After reviewing your request, the U.S. Nuclear Regulatory Commission staff has determined that additional information is required to complete the review. During a phone call on September 4, 2014, the licensee agreed to provide the additional information requested on two schedules.
RAI-1 and RAI-3 are to be provided within 90 days of receipt of this letter and RAI-2 and RAI-4 are to be provided by May 31, 2015.
If you have any questions, please contact me at 301-415-1445 or via e-mail at alan.wang@nrc.gov.
Docket No. 50-382
Enclosure:
Request for Additional Information cc w/encl: Distribution via Listserv Sincerely, Alan B. Wang, Project Plant Licensing IV-2 a r Transition Branch.
Division of Operating Reactor Licensing Office ofNuclear Reactor Regulation
REQUEST FOR ADDITIONAL INFORMATION AGING MANAGEMENT PROGRAM FOR THE REACTOR VESSEL INTERNALS ENTERGY OPERATIONS, INC.
WATERFORD STEAM ELECTRIC STATION, UNIT 3 DOCKET NO. 50-382 On February 5, 2005 (Agencywide Documents Access and Management System (ADAMS)
Accession No. ML050400463), Entergy Operations Inc. (Entergy, the licensee), as part of its implementation of the extended power uprate, made a commitment to submit an aging management program (AMP) for the reactor vessel internals (RVI) components at Waterford Steam Electric Station, Unit 3 (WF3). Consistent with this commitment, by letter dated December 16, 2013 (ADAMS Accession No. ML13352A041 ), Entergy submitted an AMP for the RVI components at WF3. The Electric Power Research Institute's technical report MRP-227-A, "Materials Reliability Program: Pressurized Water Reactor (PWR) Internals Inspection and Evaluation Guidelines," December 2008 (non-proprietary version at ADAMS Accession No. ML090160205), and its supporting reports were used as the technical bases for developing WF3'sAMP.
Based on its review of WF3's AMP conducted thus far, the U.S. Nuclear Regulatory Commission (NRC) staff has determined the following information is needed to complete its review:
RAI-1: Historically, the following materials used in the PWR RVI components are known to be susceptible to some of the aging degradation mechanisms that are identified in the MRP-227 -A report. In this context, the NRC staff requests that the licensee provide a list of any additional.
RVI components (not listed in MRP-227-A and MRP-191, Revision 0, "Materials Reliability Program: Screening, Categorization, and Ranking of Reactor Internals Components for Westinghouse and Combustion Engineering PWR Design," November 2006 (ADAMS Accession No. ML091910130)) that are manufactured from the following materials. If any additional RVI components at WF3 are fabricated from these materials, please identify the aging effect/mechanism combination to which the material is susceptible, and the type of aging management that will be implemented on these components.
(1)
Nickel base alloys-lnconel 600; Weld Metals-Alloy 82 and 182; Alloy X-750, (2)
Stainless steel type 347 material (excluding baffle-former bolts),
(3)
Precipitation hardened (PH) stainless ste~l materials-17 -4 and 15-5, (4)
Type 431 stainless steel material, (5)
Alloy A-286, ASTM A 453 Grade 660, Condition A orB.
Enclosure RAI-2: MRP-2013-025, "MRP-227-A Applicability Template Guidelines," dated October 14, 2013 (ADAMS Accession No. ML13322A454), was developed by Westinghouse Electric Company LLC (Westinghouse) as a part of its response to the NRC staff's Action Item 1 of the safety evaluation for the MRP-227 -A report. In MRP-2013-025, two issues were addressed (A) Cold Worked materials which is addressed in Appendix A of the report; and (B) Fuel Design or Fuel Management which is addressed in Appendix B of the report. With regard to the above, the NRC staff requests the following information:
A Please clarify if the WF3 RVI components have non-weld or bolting austenitic stainless steel components with 20 percent cold work or greater, and if so, whether the affected components have operating stresses greater than 30 ksi. In particular, please provide the plant-specific information on the extent of cold work on its RVI components. The licensee can apply "Option 1" or "Option 2," as addressed in Appendix A of the report. If "Option 2" is applicable to WF3, the licensee should list plant-specific RVI components that have been exposed to cold work equal to or greater than 20 percent. Plant-specific information related to this issue as addressed in "Option 2" in Appendix A, should be provided.
B.
- 1)
(
Please explain if WF3 has ever utilized atypical design or fuel management that
- could make the assumptions of MRP-227 -A regarding core loading/core design non-representative for that plant, including power changes/uprates such as the extended power uprate implemented in 2005. The following guidelines provided by MRP should be followed. The licensee is requested to use the MRP document dated October 14, 2013, MRP-2013-025, and it can apply "Option 1" or "Option 2," as addressed in Appendix B of the report..
Option 1 WF3 complies with the MRP-227-A assumptions regarding core loading/core design. Neutron fluence and heat generation rates are concluded to be Option A or Option B.
Option A: acceptable based on the following assessment to the limiting MRP guidance threshold values.
- Option B: unacceptable based on an assessment to the limiting MRP guidance threshold values.
If Option A as addressed under "Option 1" is applicable, the following plant-specific values should be submitted: (a) active fuel to fuel alignment plate distance; (b) average core power density; and, (c) heat generation figure of merit.
If Option B under "Option 1" is applicable to WF3, the licensee should justify the usage of its fuel management program.
Option 2 The licensee should provide a technical justification for the application of MRP-227 -A criter.ion to WF3.
- RAI-3: Chapter 7 of the MRP-227 -A report addresses how a licensee will evaluate and disposition relevant plant-specific or generic operating experience that is applicable to RVI components at its PWR facility. The NRC staff requests that the licensee identify any and all generic and plant-specific operating experience that is applicable to the design of the RVI components including, but not limited to, operating experience that is applicable to the following components at WF3: panel to former bolts, core barrel bolting, thermal shields (including positioning pins), fuel alignment pins, guide lug inserts and.bolts, guide lugs, core support barrel girth welds, in-core instrumentation 'flux thimble tubes, core barrel at the prior thermal shield bracket attachment areas, reactor vessel flow skirt, and, include the RVI components addressed in Table 2 of the December 16, 2013, submittal.
If a portion of the listed components are not applicable to WF3, please provide that information in the reply.
RAI-4: As discussed in Section 3.3.7 of Reyision 1 to the safety evaluation for MRP-227 dated December 16, 2011 (ADAMS Accession No. ML11308A770), Action Item 7 requires that the licensees of Westinghouse reactors develop plant-specific analyses to be applied for their facilities to demonstrate that lower support colunin cast austenitic stainless steel, (CASS) bodies will maintain their function during the extended period of operation. However, licensees are observing degradation in CASS bodies during the current operating license. MRP-227 -A Table 3-2 (Final disposition of Combustion Engineering internals) classifies CASS lower support columns as Primary Components based on susceptibility to irradiation embrittlement (IE) and irradiation assisted stress-corrosion cracking and thermal embrittlement (TE). After further review of the existing literature data for the thre~hold limits for IE and TE of CASS materials, the NRC staff developed a new position for screening of CASS materials for combined IE and TE.
The bases for the staff's new threshold limits are described in the document "NRC position on Aging Management of CASS Reactor Vessel Internal Components," at ADAMS Accession No. ML14163A112.
To enable an assessment of the adequacy of aging management for the lower support columns in response to TE.and IE, the NRC staff requests that the licensee address Action Item 7 of the December 16, 2011, safety evaluation for Revision 1 of MRP-227 -A under the current operating license.
- memo dated OFFICE NRR/DORLILPL4-2/PM NRR/DORLILPL4-2/PM NRR/DORL/LPL4-2/LA NAME MOrenak AWang JBurkhardt DATE 9/15/14 9/15/14 9/12/14 OFFI9E NRR/DE/EVIB/BC NRR/DORLILPL4-2/BC NRR/DORLILPL4-2/PM NAME.
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DBroaddus A Wang DATE 7/23/14 10/16/14 10/21/14