ML13119A097

Summary of Telephone Conference Call Held on April 22, 2013 Between the U.S. Nuclear Regulatory Commission and Tva,Concerning Requests for Additional Information Pertaining to the Sequoyah Nuclear Plant, License Renewal Application

ML13119A097
Person / Time
Site:	Sequoyah
Issue date:	05/21/2013
From:	Plasse R License Renewal Projects Branch 1
To:	James Shea Tennessee Valley Authority
Yoo, M 415-8583
References
TAC MF0481, TAC MF0482
Download: ML13119A097 (10)

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LICENSEE: Tennessee Valley Authority FACILITY: Sequoyah Nuclear Plant

SUMMARY

OF TELEPHONE CONFERENCE CALL HELD ON APRIL 22, 2013 BETWEEN THE U.S. NUCLEAR REGULATORY COMMISSION AND TENNESSEE VALLEY AUTHORITY, CONCERNING REQUESTS FOR ADDITIONAL INFORMATION PERTAINING TO THE SEQUOYAH NUCLEAR PLANT, LICENSE RENEWAL APPLICATION (TAC. NOS. MF0481, MF0482)

The U.S. Nuclear Regulatory Commission (NRC or the staff) and representatives of Tennessee Valley Authority held a telephone conference call on April 22, 2013, to discuss and clarify the staff's requests for additional information (RAls) concerning the Sequoyah Nuclear Plant, license renewal application. The telephone conference call was useful in clarifying the intent of the staff's RAls.

Enclosure 1 provides a listing of the participants, and Enclosure 2 contains a listing of the RAls discussed with the applicant, including a brief description on the status of the items.

The applicant had an opportunity to comment on this summary.

Richard A. PI sse, Project Manager Projects Branch 1 Division of License Renewal Office of Nuclear Reactor Regulation Docket Nos: 50-327 and 50-328

Enclosures:

1. List of Participants

2. List of Requests for Additional Information cc: Listserv

ML13119A097 OFFICE LA:DLR PM:RPB1 :DLR BC:RPB1 :DLR NAME YEdmonds R Plasse YDiaz-Sanabria DATE 5/15/2013 5/16/2013 5/21/2013

TELEPHONE CONFERENCE CALL SEQUOYAH NUCLEAR PLANT LICENSE RENEWAL APPLICATION LIST OF PARTICIPANTS APRIL 22, 2013 PARTICIPANTS AFFILIATIONS Rick Plasse Nuclear Regulatory Commission (NRC)

Mark Yoo NRC OnYee NRC Jim Medoff NRC Pat Purtscher NRC Gary Adkins Tennessee Valley Authority (TVA)

Henry Lee TVA Dennis Lundy TVA Cheryl Boggess Westinghouse Tom Hamm Westinghouse Randy Lott Westinghouse Mike Semmler Westinghouse Karli Szweda Westinghouse Dave Lach Entergy/Enercon Jacque Lingenfelter Entergy/Enercon Andy Taylor Entergy/Enercon David Wootten Entergy/Enercon Enclosure 1

REQUESTS FOR ADDITIONAL INFORMATION SEQUOYAH NUCLEAR PLANT LICENSE RENEWAL APPLICATION APRIL 22, 2013 The U.S. Nuclear Regulatory Commission (NRC or the staff) and representatives of Tennessee Valley Authority held a telephone conference call on April 22, 2013, to discuss and clarify the following requests for additional information (RAls) concerning the license renewal application (LRA).

Draft RAI B.1.34-1

Background:

LRA Section 8.1.34 provides enhancements to the "detection of aging effects" and "acceptance criteria" program elements of the Reactor Vessel Internals Program. These enhancements are associated with revising the program procedures to account for taking physical measurements, including the preload acceptance criteria, for the Type 304 stainless steel hold-down spring in Unit 1.

Applicant/Licensee Action Item (AiLAI) NO.5 of MRP-227 -A states, in part, that the applicantsllicensees shall identify plant-specific acceptance criteria to be applied when performing the physical measurements required by the NRC-approved version of MRP-227 for loss of compressibility for Westinghouse hold-down springs. It also states, in part, that the applicant/licensee shall include its proposed acceptance criteria with an explanation of how the functionality of the component being inspected will be maintained under all licensing basis conditions of operation during the period of extended operation as part of their submittal to apply the approved version of MRP-227.

The applicant's response AlLAI NO.5 in LRA Appendix C states that the plant specific acceptance criteria for hold-down springs and an explanation of how the proposed acceptance criteria are consistent with the SQN licensing basis and the need to maintain the functionality of the hold-down springs under all licensing basis conditions will be developed prior to the first required physical measurement.

Issue:

Action/Licensee Action Item NO.5 requires the identification of the plant-specific acceptance criteria to be applied when performing the physical measurements and an explanation of how the functionality of the component being inspected will be maintained under all licenSing basis conditions of operation during the period of extended operation.

The applicant's proposed enhancements to revise its procedures to take physical measurements of the Type 304 stainless steel hold-down spring in Unit 1 and to include preload acceptance criteria does not adequately address Action/Licensee Action AlLAI Item NO.5 of MRP-227-A. Specifically, the applicant did not provide its plant-specific acceptance criteria for the Type 304 stainless steel hold-down spring in Unit 1 and the explanation outlined in AlLAI NO.5.

Enclosure 2

Request:

A. Define and justify the physical measurement techniques that will be used to determine RVI hold-down spring height when inspections are performed on the component in accordance with the MRP-227-A.

B. Explain and justify how the proposed acceptance criteria is consistent with the Unit 1 licensing basis and the need to maintain the functionality of the component being inspected under all licensing basis conditions of operation during the period of extended operation.

Revise the response to A1LAI No.5, as necessary.

Discussion: The staff and the applicant discussed the topic of identifying the acceptance criteria when performing physical measurements as part of A1LAI NO.5 of MRP-227-A. As a result of this discussion, no action was required.

Draft RAI 8.1.34-2

Background:

LRA Section 3.1.2.2.9.B.2 states that Unit 2 uses a Type 403 stainless steel hold-down spring.

LRA Table 3.1.2-2 indicates that the "interfacing components: internals hold-down spring" made of stainless steel (Type 403) is a "No Additional Measures" component for "loss of material wear" and "loss of preload" as part of the Reactor Vessel Internals Program.

A1LAI NO.5 of MRP-227 -A states, in part, that applicantsllicensees shall identify plant-specific acceptance criteria to be applied when performing the physical measurements required by the NRC-approved version of MRP-227 for loss of compressibility for Westinghouse hold-down springs. It also states, in part, that the applicant/licensee shall include its proposed acceptance criteria and an explanation of how the functionality of the component being inspected will be maintained under all licensing basis conditions of operation during the period of extended operation as part of their submittal to apply the approved version of MRP-227.

Issue:

The Westinghouse Type 403 stainless steel hold down spring is not specifically excluded from the scope of A1LAI NO.5 for MRP-227-A. The applicant did not address the Unit 2 hold-down spring made of Type 403 stainless steel in its response to A1LAI NO.5 in LRA Appendix C or justify that "loss of material - wear" and "loss of preload" are not applicable aging effects.

Request:

o Justify that the Unit 2 Type 403 stainless steel hold down spring is not subject to stress relaxation such that the functionality of the component will be maintained under all licensing basis conditions of operation during the period of extended operation.

o In lieu of this demonstration, revise the Reactor Vessels Internals Program, LRA Table 3.1.2-2 and LRA Table C-1 to identify that the Unit 2 hold-down springs made of Type 403 stainless steel are managed for "loss of material - wear" and "loss of preload" as a Enclosure 2

"Primary" component. In addition, provide responses to the following questions, as discussed in A/LA I NO.5 of MRP-227-A:

o Define and justify the physical measurement techniques that will be used to determine RVI hold-down spring height when inspections are performed on the component in accordance with the MRP-227-A.

o Explain and justify how the proposed acceptance criteria is consistent with the Unit 2 licensing basis and the need to maintain the functionality of the component being inspected under all licensing basis conditions of operation during the period of extended operation.

o Revise the response to AlLAI No.5, as necessary.

Discussion: The staff determined that clarification was needed on the issue section of the question and will reword the issue section to clarify why justification is needed to not manage the effects of aging on the Westinghouse Type 403 stainless steel hold down spring. The staff has replaced the issue section of the question with the following:

The Westinghouse Type 403 stainless steel hold down spring is not specifically excluded from the scope of AlLAI NO.5 for MRP-227 -A. Since the Type 403 stainless steel hold down spring was not addressed in the MRP-227-A (staff reviewed and approved), it is not clear to the staff why the applicant did not address this component in its response to NLAI NO.5 in LRA Appendix C or justify that this component does not need to managed for "loss of material - wear" and "loss of preload."

Draft RAI 8.1.34-3

Background:

ApplicanULicensee Action Item NO.8 states, in part, for those cumulative usage factor (CUF) analyses that are time-limited aging analysis (TLAAs) for reactor vessel internals, the acceptance of these TLAAs may be done in accordance with either 10 CFR 54.21 (c)(1)(i) or (ii),

or in accordance with 10 CFR 54.21 (c)(1)(iii) using the applicant's program that corresponds to NUREG-1801, Revision 2, AMP X.M1, "Metal Fatigue of Reactor Coolant Pressure Boundary Program." To satisfy the evaluation requirements of ASME Code,Section III, Subsection NG- 2160 and NG-3121, the existing fatigue cumulative usuage factor (CUF) analyses shall include the effects of the reactor coolant system water environment. The applicant's response to Part 5 of NLAI NO.8 in LRA Appendix C states that TLAAs are identified in LRA Section 4.

LRA Section 4.3.1.2 provides the applicant's TLAA for reactor vessel internal components with CUF values, which include the lower core plate and control rod drive guide tube pins. The applicant dispositioned this TLAA in accordance with 10 CFR 54.21 (c)(1 ) (iii) such that the Fatigue Monitoring Program will manage the effects of aging due to fatigue on the reactor vessel internals.

Issue:

Since the TLAA is managed with the Fatigue Monitoring Program, the staff noted that LRA Section 4.3 and LRA Appendix C do not address the aspect in Part 5 of NLAI NO.8 that states "the existing fatigue CUF analyses shall include the effects of the reactor coolant system water environment. "

Enclosure 2

"the existing fatigue CUF analyses shall include the effects of the reactor coolant system water environment. "

Reguest:

• Since the Fatigue Monitoring Program will manage the effects of aging due to fatigue on the reactor vessel internals in accordance with 10 CFR 54.21 (c)(1 )(iii), justify how the existing fatigue CUF analyses will include the effects of the reactor coolant system water environment as discussed in Part 5 of AILA I NO.8. Revise the response to AILA I NO.8 and LRA Sections A 1.11 and B.1.11 to explicitly describe how the effects of the reactor coolant system water environment for the reactor vessel internals TLAA will be managed.

Discussion: The staff and the applicant discussed the topic of managing the effects of reactor coolant system water environment on reactor vessel internals. As a result of this discussion, no action was required.

Draft RAJ 8.1.34-6

Background:

AILA I No. 7 states, in part, the applicants/licensees of Westinghouse reactors are required to develop plant-specific analyses to be applied for their facilities to demonstrate that Westinghouse lower support column bodies will maintain their functionality during the period of extended operation or for additional RVI components that may be fabricated from CASS, martensitic stainless steel, or precipitation hardened stainless steel materials.

AILA I No. 7 continues to state that these analyses should also consider the possible loss of fracture toughness in these components due to thermal and irradiation embrittlement, and may also need to consider limitations on accessibility for inspection and the resolution/sensitivity of the inspection techniques. Furthermore, it states, in part, that this would apply to components fabricated from materials susceptible to thermal and/or irradiation embrittlement for which an individual licensee has determined aging management is required, for example during their review performed in accordance with AlLAI NO.2.

For Unit 2, the applicant stated in LRA Appendix C that the hold-down spring is fabricated of Type 403 stainless steel, which is a martensitic stainless steel. Table 5-1 of MRP-191 a/so indicates that the Type 403 stainless steel hold-down spring may be subject to thermal embrittlement.

Issue:

Since AlLAI No. 7 specifically discusses the performance of a plant-specific analysis for reactor vessel internal components fabricated from martensitic stainless steel materials, it is not clear whether the applicant has performed this analysis for the Unit 2 Type 403 hold-down spring to consider the possible loss of fracture toughness due to thermal and irradiation embrittlement.

Enclosure 2

Request:

• Clarify whether the Unit 2 Type 403 stainless steel hold down springs were evaluated in response to AlLAI NO.7.

• If yes, describe and justify the evaluation performed to consider the possible loss of fracture toughness due to thermal and irradiation embrittlement.

• If not, justify that the Unit 2 Type 403 stainless steel hold down spring is not applicable to the evaluation discussed in AILA I NO.7.

Discussion: The staff and the applicant discussed the topic managing the effects of aging on Westinghouse Type 403 stainless steel hold-down springs. As a result of this discussion, no action was required.

Draft RAI 6.1.34-7

Background:

During its audit, the staff noted that the "detection of aging effects" program element for the Reactor Vessel Internals Program indicates that the "Existing Programs" components were taken from Table 4-9 in MRP-227-A. Section 3.3 of MRP-227-A defines "Existing Programs" components as those PWR internals that generic and plant-specific existing aging management program elements are capable of managing aging effects. Section 4.4 of MRP-227 -A states the following for "Existing Programs" components:

Included in the Existing Programs are PWR internals that are classified as removable core support structures. ASME Section XI, IWB-2S00, Examination Category 8-N-3 does not list component specific examination requirements for removable core support structures. Accordingly, factors such as original design, licensing and code of construction variability could result in significant differences in an individual plant's current 8-N-3 requirements. These guidelines credit specific components contained within the general B-N-3 classification for maintaining functionality.

Issue:

As an example, MRP-227 -A noted that the "Existing Programs" components managed by ASME Section XI may vary from plant to plant based on original design, licensing and code of construction. Thus, since "Existing Programs" components may vary from plant to plant, it is not appropriate to rely on the list in Table 4-9 of MRP-227-A to determine the components that are classified as "Existing Programs" components. It is not clear whether the applicant confirmed during the Integrated Plant Assessment process when developing the LRA whether the components listed in Table 4-9 in MRP-227-A encompass all plant-specific "Existing Programs" components at Sequoyah Units 1 and 2.

In addition, during its review the staff noted that LRA Table 3.1.2-2 indicates the following:

• Stainless steel "Interfacing components: Upper core plate alignment pins" is subject to cracking, which is managed by the Reactor Vessels Internals Program as an "existing programs" component Enclosure 2

• Stainless steel "Control rod guide tube assembly and downcomer: Guide tube support pins (split pins)" is subject to cracking and loss of material - wear, which is managed by the Reactor Vessels Internals Program as an "existing programs" component.

However, LRA Table C-3, "Existing Program Components at SQN Units 1 and 2" indicates that the "Alignment and interfacing components: Upper core plate alignment pins" are managed for loss of material (wear) by ASME Code Section XI but is silent about cracking. Furthermore, LRA Table C-3 does not identify the Control rod guide tube assembly and downcomer: Guide tube support pins (split pins) as an item or component that is managed by an existing program.

Request:

• Clarify the discrepancies identified above between LRA Table 3.1.2-2 and LRA Table C-3 for the "Interfacing components: Upper core plate alignment pins" and "Control rod guide tube assembly and downcomer: Guide tube support pins (split pins)." If revisions to the LRA are necessary, justify any revisions that are made.

• Confirm that a review was performed to determine that the components in Table 4-9 of MRP-227 -A encompass the plant-specific "Existing Programs" components at Sequoyah Units 1 and 2.

o If not, justify that the sole use of Table 4-9 in MRP-227-A to determine the "Existing Programs" components is applicable to Sequoyah Units 1 and 2.

Discussion: The staff and the applicant discussed the discrepancies in identifying "Existing Programs" components in the LRA. As a result of this discussion, no action was required.

Enclosure 2

SUBJECT:

Summary of Telephone Conference Call conducted on April 22, 2013 DISTRIBUTION:

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RidsNrrDlr Resource RidsNrrDlrRpb1 Resource RidsNrrDlrRpb2 Resource RidsNrrDlrRerb Resource RidsNrrDlrRarb Resource RidsNrrDlrRasb Resource beth.mizuno@nrc.gov brian.harris@nrc.gov john. pelchat@nrc.gov <gov't liaison officer>

gena. woodruff@nrc.gov <gov't liaison officer>

siva.lingam@nrc.gov <dorl PM>

wesley.deschaine@nrc.gov <res insp> \

galen.smith@nrc.gov <res insp>

scott.shaeffer@nrc.gov (RII) jeffrey. hamman@nrc.gov (RII) craig. kontz@nrc.gov (RII) caudle.julian@nrc.gov {RII}

generette.lloyd@epa.gov {RIV ~-

gmadkins@tva.gov clwilson@tva.gov hleeO@tva.gov dllundy@tva.gov Enclosure 2