ML13109A515

Requests for Additional Information for the Review of the Sequoyah Nuclear Plant, Units 1 and 2, License Renewal Application (TAC Nos. MF0481 and MF0482)

ML13109A515
Person / Time
Site:	Sequoyah
Issue date:	04/26/2013
From:	Plasse R License Renewal Projects Branch 1
To:	James Shea Tennessee Valley Authority
Yoo M, 415-8583
References
TAC MF0481, TAC MF0482
Download: ML13109A515 (11)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON. D.C. 20555-0001 Aprif 26, 2013 Mr. Joe W. Shea Vice President, Nuclear Licensing Tennessee Valley Authority P.O. Box 2000 Soddy-Daisy. TN 37384

SUBJECT:

REQUESTS FOR ADDITIONAL INFORMATION FOR THE REVIEW OF THE SEQUOYAH NUCLEAR PLANT, UNITS 1 AND 2, LICENSE RENEWAL APPLICATION (TAC NOS. MF0481 AND MF0482)

Dear Mr. Shea:

By letter dated January 7, 2013, Tennessee Valley Authority submitted an application pursuant to Title 10 of the Code of Regulations (CFR) Part 54, to renew the operating license DPR-77 and DPR-79 for Sequoyah Nuclear Plant, Units 1 and 2, for review by the U.S. Nuclear Regulatory Commission (NRC) staff. The NRC 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 Gary Adkins, and a mutually agreeable date for the response is within 60 days from the date of this letter. If you have any questions, please contact me at 301-415-1427 ore-mail at Richard.Plasse@nrc.gov.

Sincerely, Richard A. lasse, Project Manager Project Branch 1 Division of License Renewal Office of Nuclear Reactor Regulation Docket Nos. 50-327 and 50-328

Enclosure:

Requests for Additional Information cc w/encl: Listserv

April 26, 2013 Mr. Joe W. Shea Vice President, Nuclear Licensing Tennessee Valley Authority P.O. Box 2000 Soddy-Daisy, TN 37384

SUBJECT:

REQUESTS FOR ADDITIONAL INFORMATION FOR THE REVIEW OF THE SEQUOYAH NUCLEAR PLANT, UNITS 1 AND 2, LICENSE RENEWAL APPLICATION (TAC NOS. MF0481 AND MF0482)

Dear Mr. Shea:

By letter dated January 7,2013, Tennessee Valley Authority submitted an application pursuant to Title 10 of the Code of Regulations (CFR) Part 54, to renew the operating license DPR-77 and DPR-79 for Sequoyah Nuclear Plant, Units 1 and 2, for review by the U.S. Nuclear Regulatory Commission (NRC) staff. The NRC 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 Gary Adkins, 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 301-415-1427 or e-mail at Richard.Plasse@nrc.gov.

Sincerely, IRA!

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

Enclosure:

Requests for Additional Information cc w/encl: Listserv DISTRIBUTION: See following pages ADAMS Accession No'.. ML13109A515 *concurred via email OFFIC :RPB2:DLR PM:RPB1 :DLR BC:RPB1 :DLR PM: RPB1 :DLR I King R Plasse D Morey R Plasse NAME DATE 4/24/13 4/24/13 4/25/13 4/26/13 OFFICIAL RECORD COpy

SEQUOYAH NUCLEAR PLANT, UNITS 1 AND 2 LICENSE RENEWAL APPLICATION REQUESTS FOR ADDITIONAL INFORMATION LRA Section 8.1.34 - Reactor Vessel Internals RAI8.1.34-1

Background:

License renewal application (LRA) Section B.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 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 Sequoyah Nuclear Plant (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:

AlLAI 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 AlLAI 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:

• Define and justify the physical measurement techniques that will be used to determine reactor vessel internal (RVI) hold-down spring height when inspections are performed on the component in accordance with the MRP-227-A. 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 AlLAI No.5, as necessary.

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.

AlLAI No.5 of MRP-227 -A states, in part, that applicants/licensees 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 its 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 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 be managed for "loss of material wear" and "loss of preload."

Request:

• 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.

• 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

-3

• "Primary" component. In addition, provide responses to the following questions, as discussed in AILA 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.

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

RAI 8.1.34-3

Background:

AILA I No.8 states, in part, for those cumulative usage factor (CUF) analyses that are time limited aging analyses (TLAAs) for RVls, 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 CUF analyses shall include the effects of the reactor coolant system water environment. The applicant's response to Part 5 of AlLAI 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 RVI 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 RVls.

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 AlLAI No.8 that states "the existing fatigue CUF analyses shall include the effects of the reactor coolant system water environment."

Request:

• Since the Fatigue Monitoring Program will manage the effects of aging due to fatigue on the RVls 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 AlLAI No.8. Revise the response to AlLAI 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 RVls TLAA will be managed.

-4 RAI B.1.34-4

Background:

NLAI NO.3 of MRP-227-A states, in part, that applicants/licensees of Westinghouse are required to perform plant-specific analysis either to justify the acceptability of an existing program, or to identify changes to the programs that should be implemented to manage the aging of Westinghouse guide tube support pins. Section 3.2.5.3 of the staff's safety evaluation (SE), Rev.1 for MRP-227 clarifies, in part, that the evaluation consider the need to inspect the replacement Type 316 stainless steel support pins to ensure that cracking has been mitigated and that aging degradation is adequately monitored during the extended period of operation.

The applicant's response to NLAI No.3 in LRA Appendix C states that third generation split pins, which were qualified for 40 years from the time of installation, were installed in the fall of 2001 for Unit 1 and spring of 2002 for Unit 2. It further states that potential aging effects were evaluated, including those identified in MRP-191 Table 5-1, and no additional inspection requirements were established for the control rod guide tube support pins in the deSign change packages that installed them. LRA Appendix C states that the basis for not establishing additional inspection requirements is the following: (1) cold-worked Type 316 stainless steel split pins have been installed at other plants since 1997 and none of these plants have experienced any failures and (2) since other plants have installed split pins since 1997 and SQI\I did not install them until 2001 for Unit 1 and 2002 for Unit 2, the other plants will provide a leading indicator. Thus, the effects of aging on these components will be managed in the period of extended operation based on operating experience.

The Standard Review Plan for License Renewal (SRP-LR) Section A.1.2.3.4 states, in part, the effects of aging on a structure or component should be managed to ensure its availability to perform its intended function(s) as designed when called upon and that a program based solely on detecting structure and component failure should not be considered as an effective AMP for license renewal.

Issue:

The staff's SE, Rev.1, for MRP-227, specifically discusses the inspection of replacement Type 316 stainless steel support pins to ensure that cracking has been mitigated and that aging degradation is adequately monitored during the period of extended operation. Whereas, the applicant has stated that no additional inspection requirements were established for the control rod guide tube support pins and the effects of aging on these components will be managed in the period of extended operation based on operating experience from other plants.

The applicant's approach for aging management is not appropriate based on (1) the staff's SE, Rev. 1, for MRP-227, (2) NLAI NO.3 and (3) SRP-LR Section A.1.2.3.4. It is not clear that it is appropriate for the applicant to rely solely on the operating experience at other plants as a means of aging management for its control rod guide tube support pins.

-5 Request:

• Justify that age-related degradation of the Type 316 stainless steel control rod guide tube support pins is adequately monitored during the period of extended operation in response to AlLAI NO.3 and Section 3.2.5.3 of the staff's SE, Rev.1 for MRP-227. As part of the justification, provide the inspection category, techniques, frequency and coverage for the replacement Type 316 stainless steel control rod guide tube support pins to ensure that age-related degradation is adequately monitored during the extended period of operation. Revise the LRA, as needed, to provide program enhancements/augmentations.

RAI B.1.34-5

Background:

The applicant's response to AlLAI No.2 in LRA Appendix C states "SQN reviewed the information in Table 4-4 of IVIRP-191 and determined that there are no additional components contained in the SQN design. Table 4-4 of MRP-191 contains all of the RVI components that are within the scope of license renewal for SQN Units 1 and 2."

The staff noted that Table 4-4 of MRP-191 indicates that the "Control Rod Guide Tube Assemblies and Flow Downcomers: Guide plates/cards" were evaluated as Type 304 stainless steel. Similarly, Table 3-3 in MRP-227-A indicates that the guide plates (cards) were evaluated as Type 304 stainless steel. However, LRA Table 3.1.2-2 indicates that the "control rode guide tube assembly and downcomer: guide cards and plates" are fabricated from cast austenitic stainless steel (CASS) and are considered a "No Additional Measures" component for managing reduction of fracture toughness.

AlLAI No.7 states, in part, that it 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. The plant-specific analysis shall be consistent with the plant's licensing basis and the need to maintain the functionality of the components being evaluated under all licensing basis conditions of operation. The applicantsllicensees shall include the plant-specific analysis as part of their submittal to apply the approved version of MRP-227.

Issue:

It is not clear whether the results in Table 4-4 in MRP-191 and Table 3-3 in MRP-227-A for the Type 304 stainless steel guide plates (cards) is applicable to the applicant's CASS guide plates/cards and whether the applicant's response to AlLAI No.2 is accurate. In addition, considering that the applicant's guide plate (cards) are fabricated of CASS, it appears that this component should be evaluated as part of AlLAI No. 7 to consider the possible loss of fracture toughness due to thermal and irradiation embrittlement and, if applicable, the limitations on accessibility for inspection and the resolution/sensitivity of the inspection techniques.

-6 Request:

• Considering that the applicant's guide plates (cards) are fabricated of CASS, clarify the applicability of the MRP-191 and MRP-227-A that evaluated the guide plates (cards) as Type 304 stainless steel.

• Confirm that there are no other discrepancies in material fabrication of components evaluated in MRP-191 and MRP-227-A with those at the applicant's site. Ifthere are other discrepancies, provide the component (including material) and justify the aging effects and the inspection category, techniques, coverage, and frequency to account for the material differences. Revise the LRA and the response to AlLAI No.2, as needed.

• Since the guide plates (cards) are fabricated from CASS, describe and justify the plant specific analysis performed in response to AlLAI No. 7 that considers the possible loss of fracture toughness in these components due to thermal and irradiation embrittlement, and, if applicable, the limitations on accessibility for inspection and the resolution/sensitivity of the inspection techniques.

• Revise the response to AlLAI Nos. 2 and 7, as necessary.

RAI B.1.34-6

Background:

AlLAI NO.7 states, in part, the applicantsllicensees 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 also 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 RVI components fabricated from martensitic stainless steel materials, it is not clear whether the

-7 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.

Request:

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

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

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

RAI B.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 pressurized-water reactor (PWR) internals that generic and plant-specific existing aging management program elements are capable of managing aging effects. Section 4.4 of IVIRP-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-2500, Examination Category B-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 B-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 ASIVIE 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 SQN, Units 1 and 2.

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

- 8 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.

• Stainless steel "Control rod guide tube assembly and down comer: 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 SON 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 SON, 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 SON, Units 1 and 2.

Letter to J. Shea from R. Plasse dated April 26,2013

SUBJECT:

REQUESTS FOR ADDITIONAL INFORMATION FOR THE REVIEW OF THE SEQUOYAH NUCLEAR PLANT, UNITS 1 AND 2, LICENSE RENEWAL APPLICATION (TAC NOS. MF0481 AND MF0482)

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