ML15033A373
| ML15033A373 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 01/26/2015 |
| From: | Mark D. Sartain Dominion, Dominion Nuclear Connecticut |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| 14-623, TAC MF3402 | |
| Download: ML15033A373 (6) | |
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VDominion Dominion Nuclear Connecticut, Inc.
5000 Dominion Boulevard, Glen Allen, VA 23060 Web Address: www.dom.com January 26, 2015 U. S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555 Serial No.
NSSLIWDC Docket No.
License No.14-623 RO 50-336 DPR-65 DOMINION NUCLEAR CONNECTICUT, INC.
MILLSTONE POWER STATION UNIT 2 RESPONSE TO SECOND REQUEST FOR ADDITIONAL INFORMATION REGARDING AGING MANAGEMENT PROGRAM DESCRIPTION: INSERVICE INSPECTION - REACTOR VESSEL INTERNALS, LICENSE RENEWAL COMMITMENT #13 (MF3402)
By letter dated July 31, 2013, Dominion Nuclear Connecticut, Inc. (DNC) submitted the "Aging Management Program
Description:
Inservice Inspection -
Reactor Vessel Internals" to address License Renewal Commitment #13 for Millstone Power Station Unit 2 (MPS2).
The submittal contains an updated Reactor Vessel Internals (RVI)
Aging Management Program and RVI Inspection Plan in accordance with topical report "Materials Reliability Program: Pressurized Water Reactor Inspection and Evaluation Guidelines" (MRP-227-A).
In an email dated May 14, 2014, the Nuclear Regulatory Commission transmitted a request for additional information (RAI) related to the submittal. DNC responded to the RAI in letters dated July 21 and December 19, 2014. In an email dated December 15, 2014, the NRC transmitted a second RAI related to the submittal. Attachment 1 to this letter contains DNC's response to the RAI.
If you have any questions or require additional information, please contact Wanda Craft at (804) 273-4687.
Sincerely, wba21*a Mark D. Sartain Vice President - Nuclear Engineering CRAIG D SLY 4
Notary Public I
Commonwealth of Virginia Reg. # 7518653
,04 My Commission Expires December 31, 20_
COMMONWEALTH OF VIRGINIA
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COUNTY OF HENRICO The foregoing document was acknowledged before me, in and for the County and Commonwealth aforesaid, today by Mark D. Sartain, who is Vice President - Nuclear Engineering of Dominion Nuclear Connecticut, Inc. He has affirmed before me that he is duly authorized to execute and file the foregoing document in behalf of that company, and that the statements in the document are true to the best of his knowledge and belief.
Acknowledged before me this ____
day of."J a
,r* 2015.
)Ct-7 My Commission Expires: "t ece44_
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Notary Public I
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Serial No.14-623 Docket No. 50-336 Page 2 of 2 Commitments made in this letter: None
Attachment:
1.
Response to Second Request for Additional Information Regarding License Renewal Commitment #13 cc:
U.S. Nuclear Regulatory Commission Region I 2100 Renaissance Blvd Suite 100 King of Prussia, PA 19406-2713 Mohan C. Thadani Senior Project Manager U.S. Nuclear Regulatory Commission One White Flint North, Mail Stop 08 B1 11555 Rockville Pike Rockville, MD 20852-2738 NRC Senior Resident Inspector Millstone Power Station
Serial No 14-623 Docket No. 50-336 Response to Second Request for Additional Information Regarding License Renewal Commitment #13 Dominion Nuclear Connecticut, Inc.
Millstone Power Station Unit 2
Serial No 14-623 Docket No. 50-336, Page 1 of 3 Response to Second Request for Additional Information Regarding License Renewal Commitment #13 By letter dated July 31, 2013, Dominion Nuclear Connecticut, Inc. (DNC) submitted the "Aging Management Program
Description:
Inservice Inspection - Reactor Vessel Internals" to address License Renewal Commitment #13 for Millstone Power Station Unit 2 (MPS2).
The submittal contains an updated Reactor Vessel Internals (RVI) Aging Management Program and RVI Inspection Plan in accordance with topical report "Materials Reliability Program: Pressurized Water Reactor Inspection and Evaluation Guidelines" (MRP-227-A).
In an email dated May 14, 2014, the Nuclear Regulatory Commission transmitted a request for additional information (RAI) related to the submittal.
DNC responded to the RAI in letters dated July 21 and December 19, 2014.
In an email dated December 15, 2014, the NRC transmitted a second RAI related to the submittal. The response to this RAI is as follows:
RAI 2-1 For the cast austenitic stainless steel (CASS) core support columns at Millstone Power Station, Unit 2 (MPS2), the licensee was able to screen out thermal embrittlement (TE) for 63 of 68 columns for which certified material test reports (CMTRs) were available. The columns would therefore be susceptible to irradiation embrittlement (IE) and possibly TE for those columns lacking CMTRs. Therefore, in a request for additional information (RAI),
the staff requested the licensee to clarify the scope of the core support column weld inspection, and to provide either a functionality analysis for the columns or modify its reactor vessel internals (RVI) Inspection Plan to add an inspection of the columns as a Primary or Expansion category if no part of the core support columns other than the welds is covered by the current inspection. The licensee provided the results of its statistical analysis showing those columns for which the ferrite content was not available were extremely unlikely to exceed the TE screening criteria; therefore, the licensee concluded the columns were susceptible to IE only. The licensee's July 21, 2014 response to RAI 2 indicates that the licensee considers that the "Primary" inspection of the core support column welds provides adequate management for IE because a portion of the CASS material is exposed in the weld area to a bounding level of irradiation. However, the licensee's response did not discuss actions to be taken if cracking is found in these welds.
Per MRP-191, "Materials Reliability Program: Screening, Categorization, and Ranking of Reactor Internals Components for Westinghouse and Combustion Engineering PWR Design," (Reference 1) and MRP-227-A, the core support column welds and the CASS core support columns are susceptible to the same set of aging mechanisms: cracking due to fatigue, irradiation-assisted stress corrosion cracking (IASCC) and stress corrosion cracking (SCC), and loss of fracture toughness due to TE and IE. Since the weld area includes some of the CASS core support column material, it represents a leading indicator for degradation of the remainder of the core support column. The NRC staff is concerned that since MRP-227-A does not formalize this relationship by classification of the core support columns as an Expansion category component, cracking that could cause loss of
Serial No 14-623 Docket No. 50-336, Page 2 of 3 function of the columns which have reduced fracture toughness due to IE will not be adequately managed.
What actions would be taken to ensure the integrity of the core support columns if evidence of service-induced cracking exceeding the acceptance criteria of Table 5-2 of MRP-227-A is detected during the "Primary" inspection of the core support column welds?
DNC Response Sections 2.4, 2.5, 3.6 and 3.7 of the aging management program (AMP) submitted July 31, 2013, address the actions and process that would be followed in the event of identification of relevant conditions. Specifically, if service-induced cracking was detected by the core support column weld inspection, the results would be entered into the corrective action program and a condition report (CR) would be generated. The CR process initiates the operability determination process that prompts an engineering evaluation that would address the effect of the condition on the overall integrity of the core support system and impact on functionality of affected components. The CR process also identifies any extent of condition actions required to form the basis for conclusions reached in the engineering evaluation. Also, additional characterization of the subject indication may be performed to support the engineering evaluation. Core reload is contingent upon achieving satisfactory conclusions regarding structural integrity of the core support system and its capability to perform safety related functions.
The engineering evaluation process to determine structural integrity is described in AMP Sections 2.5 and 3.7. The evaluation could result in determination of acceptability for a limited duration and subject to subsequent confirmatory activities. In general, however, it is not possible to describe the details of the process or its potential outcomes unless the specifics of the actual occurrence are known and the process itself has been entered.
As stated in the July 31, 2013 submittal, AMP Section 3.7, Corrective Actions, regarding engineering evaluations, "A more current acceptance criteria methodology document, WCAP-17096 Rev. 2, has been proposed, and to the extent it is subsequently approved for use, it will be the preferred methodology for determining engineering acceptance or corrective action for items requiring disposition as a result of MRP-227-A inspections.
MPS2 will comply with the requirements and limitations of the WCAP document as stipulated in the NRC SER [Safety Evaluation Report]." Although both the WCAP-17096 document and the NRC SER are not yet final, DNC understands that, as discussed in "Summary of March 21, 2013 Teleconference" (ADAMS Accession No. ML13084A013),
the current SER draft contains a condition that licensees submit their engineering evaluations within one year of discovery, for review by the NRC staff.
Serial No 14-623 Docket No. 50-336, Page 3 of 3 RAI 2-2 In RAI 8, Item 1 the staff requested, that the licensee provide the plant-specific fatigue evaluations for the RVI components for which fatigue evaluations are being credited in lieu of inspections.
In its July 21, 2014, response to RAI 8, Item 1, the licensee indicated a plant-specific fatigue evaluation was not performed but that the evaluation for MPS2 was developed based on comparisons and scaling of analyses performed by the vendor for another plant. The response also provided an outline of the methodology of the evaluation.
However, the staff requires more detail of the fatigue evaluation to complete its review.
The staff requests the licensee provide the following:
- 1. Provide a reference for the fatigue analysis methodology, such as an [American Society of Mechanical Engineers Boiler and Pressure Vessel Code] ASME Code subsection. Specifically, was the reference plant methodology per the ASME Code, Section Xl, Subsection NG?
- 2. Describe the method by which the effects of the reactor coolant system water environment were considered in the fatigue analysis, or provide a justification for not considering environmental effects.
- 3. Provide the exact CUF [Cumulative Usage Factor] calculated for the Core Support Barrel Assembly - Lower Flange Weld and the Lower Support Structure - Core Support Plate at MPS2.
DNC Response
- 1. For the fatigue evaluations performed by the vendor, the ASME III 1971 Edition with the Winter 1973 Addendum are referenced. The Winter 1973 Addendum incorporated the draft Subsection NG.
- 2. The fatigue evaluation described in the RAI 8, Item 1 response was based on traditional ASME Code calculations and did not include environmental effects. The MPS2 licensing basis does not require consideration of environmental effects for the evaluated components. However, as described in MRP-175, Appendix D, the 0.1 screening value used in the fatigue evaluation includes appropriate consideration of potential environmental effects.
The screening value was not exceeded for the analyzed components.
- 3. The evaluated fatigue usage for 60 years of expected plant operation, using the fatigue analysis methodology described in (1) above, was determined to be 0.052 for the core support plate and 0.005 for the core support barrel/lower support structure flexure weld.
Thus, both components satisfy the MRP-175 screening criteria of 0.1 with significant margin.