ML13150A412

Requests for Additional Information for the Review of the Sequoyah Nuclear Plant, Units 1 and 2, License Renewal Application (TAC Nos. MF0481 and MF0482) - Set 8

ML13150A412
Person / Time
Site:	Sequoyah
Issue date:	06/24/2013
From:	Plasse R License Renewal Projects Branch 1
To:	James Shea Tennessee Valley Authority
Sayoc E, 415-1924
References
TAC MF0481, TAC MF0482
Download: ML13150A412 (13)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 June 24, 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 - SET 8 (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 Federal 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 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, outlined in the enclosure, were discussed with Henry Lee, and a mutually agreeable date for the response is within 30 days from the date of this letter. For RAI4.3.2-1, 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 or bye-mail at Richard. Plasse@nrc.gov.

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Richard A. ~asse, Projects Branch 1 Project Manager 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

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 - SET 8 (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 Federal 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 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. outlined in the enclosure. were discussed with Henry Lee, and a mutually agreeable date for the response is within 30 days from the date of this letter. For RAI 4.3.2-1. 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 or bye-mail at Richard. Plasse@nrc.gov.

Sincerely, IRA!

Richard A. Plasse, Project Manager Projects 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 ML13150A412 *concurred via email PM:RPB1 :DLR PM: RPB1:DLR ESa oc YDiaz-Sa RPlasse 6/13/13 6/24/13 6/24/13 OFFICIAL RECORD COpy

SEQUOYAH NUCLEAR PLANT, UNITS 1 AND 2 LICENSE RENEWAL APPLICATION REQUESTS FOR ADDITIONAL INFORMATION RAI4.3.1-1

Background:

Technical Specification (TS) 6.8.4.1 provides controls to track the updated final safety analysis report (UFSAR) Section 5.2.1 cyclic and transient occurrences to ensure that components are maintained within the design limit, UFSAR Table 5.2.1-1 identifies the reactor coolant system (RCS) design transients.

Issue:

1. License renewal application (LRA) Tables 4.3-1 and 4.3-2 list pressurizer heatups as a normal operating condition transient, but this transient is not defined as a design transient for normal operating conditions in UFSAR Table 5.2.1-1 for the RCS system.

2. UFSAR Table 5.2.1-1 lists the 10% step load increase and decrease transient as an applicable design basis transient; however, these transients are not listed as transients that would need to be monitored in LRA Tables 4.3-1 and 4.3-2. It is not evident why the Fatigue Monitoring Program would not need to monitor the 10% step load increase and decrease normal operating condition transient, as this would be required to be performed in accordance with the appropriate TS requirements.

3. LRA Tables 4.3-1 and 4.3-2 list 10 cycles as the deSign cycle limit for the pressurizer auxiliary spray actuations transient; however, UFSAR Table 5.2.1-1 identifies that the cycle limit for this transient is 12 cycles.

Request:

1. Provide the basis why UFSAR Table 5.2.1-1 does not list pressurizer heatups as an applicable normal operating condition transient, when this transient is listed in LRA Tables 4.3-1 and 4.3-2. Clarify and justify whether a 10 CFR 50.71(e) update of UFSAR Table 5.2.1-1 will need to be processed to add the pressurizer heatup transient as a normal operating condition transient for the Safety Class 1 or Class A components at the units.

2. Provide the basis why the Fatigue Monitoring Program would not need to monitor the 10% step load increase and decrease normal operating condition. Specifically, justify why the monitoring of these transients would not need to be performed in accordance with the applicable TS 6.8.4.1 requirements for the units.

3. Justify the basis for reporting a different value for the cycle limit for the pressurizer auxiliary spray actuations transient (Le., 12 cycles) in UFSAR Table 5.2.1-1 that is different from the cycle limit for this transient in LRA Tables 4.3-1 and 4.3-2 (i.e.,

10 cycles).

ENCLOSURE

-2 RAI4.3.2-1 8ackground:

LRA Section 4.3.2 discusses the maximum allowable stress range reduction analyses (implicit fatigue analyses for those Non-Safety Class 1 or Non Safety Class A piping systems that were designed either to the USAS 831.1 design code or those in the ASME Code Section III requirements for Class 2 or 3 components. These implicit fatigue analyses are identified as time-limited aging analyses (TLAAs) for the LRA. The evaluation of the TLAAs in accordance with 10 CFR 54.21 (c)(1)(i) is conducted by comparing the cumulative number of full thermal range transient occurrences for the components to a value of 7000 cycles in order to demonstrate that the maximum allowable stress ranges for the components would not need to be reduced.

Issue:

LRA Section 4.3.2 does not identify which Non-Safety Class 1 or Non-Safety Class A piping systems in the engineered safety feature (ESF) systems, auxiliary (AUX) systems, or steam and power conversion (SPC) systems were within the scope of the applicable implicit fatigue analysis requirements in either the USAS 831.1 design code or in the ASME Section III provisions for Class 2 or 3 components.

Also, LRA Section 4.3.2 does not identify the type of piping components and piping elements that are within the scope of these analyses or identify which design transients are characterized as full thermal range transients for the implicit fatigue analyses of these non-Safety Class 1/non Safety Class A piping components and elements.

Request:

1. Identify all non-Safety Class 1/non-Safety Class A ESF, AUX, and SPC systems, and the piping components and elements in these systems, that are within the scope of the applicable implicit fatigue analysis requirements in the USAS 831.1 design code or the ASME Code Section III provisions for Class 2 or 3 components. For these systems, identify the design basis transients that constitute "full thermal range" transients for the implicit fatigue analysis of these non-Class 1/non-Class A systems. Justify that the total number of the cycles for the "full thermal range" transients will remain less than or equal to the limit of 7000 cycles during the period of extended operation.

2. Compare the systems and components in the response to Part 1. of this request for additional information (RAI) to the list of components in the "Table 2" AMR tables for those ESF, AUX, and SPC systems. Amend the LRA accordingly if it is determined that additional aging management review (AMR) items on "cracking - fatigue" need to be identified for the LRA's AMR tables for ESF, AUX, and SPC systems.

3. Revise LRA Appendix A as appropriate based on the response.

-3 RAI4.3.3-1

Background:

LRA Section 4.3.3 provides the applicant's environmentally-assisted fatigue evaluations for Safety Class 1 or Safety Class A locations in the reactor coolant pressure boundary. The applicant provides its environmentally-assisted fatigue results (Le., Fen-adjusted cumulative usage factor results or CU-Fen results) for these components in LRA Table 4.3-12. The applicant identifies that the CU-Fen results were calculated using the recommended formulas in NUREG/CR-6583 for carbon or low-alloy steel components and in NUREG/CR-5704 for those stainless steel components.

Issue:

The Fen values that are derived in accordance with the NUREG report formulas are dependent on plant parameter inputs, such sulfur content and dissolved oxygen impurity contents for the reactor coolant, the operating temperature of the coolant, and strain-rates for the components.

It is not evident to the staff which plant parameter assumptions were used to establish the Fen value of 2.45 for Safety Class 1/Safety Class A components made from low alloy steel or carbon steel materials or the Fen value of 15.36 for Safety Class 1/Safety Class A components made from stainless steel materials.

Request:

Clarify how the Fen values for the low-alloy steel or carbon steel components and for stainless steel components were derived in accordance with applicable NUREG methodology for the respective material type. Identify and justify any assumptions on the plant parameter inputs (e.g., sulfur content, temperature, dissolved oxygen, and strain rate parameters) that were used to derive the Fen factors for these Safety Class 1/Safety Class A components.

RAI 3.1.2.2.1-1

Background:

The LRA includes the following AMR items to manage "cracking - fatigue" in the RCS components during the period of extended operation: (a) an AMR item in LRA Table 3.1.2-1 for reactor vessel components; (b) an AMR item LRA Table 3.1.2-1 on reactor vessel closure flange (closure stud assembly) components; (c) an AMR item in LRA Table 3.1.2-2 for reactor vessel internal (RVI) components; (d) an AMR item in LRA Table 3.1.2-3 for reactor coolant system (RCS) components; (e) an AMR item for reactor coolant pressure boundary components; (f) three AMR items in LRA Table 3.1.2-4 for steam generator components; and (g) AMR items in LRA Table 3.1.2-5 on management of "cracking - fatigue" in non-Class 1 or non-Class A piping, piping components, and piping elements, rupture discs, thermowells, tubing, and valve bodies.

In LRA Section 4.3, the applicant identifies that the fatigue-related TLAA analyses for RCS components fall into one of three different categories: (a) RCS components analyzed in accordance with a fatigue usage factor analysis; (b) RCS components analyzed in accordance

-4 with a fatigue waiver analysis; and (c) RCS piping components and piping elements designed to the USAS B31.1 design code requirements and analyzed in accordance with a maximum allowable stress range reduction analysis (i.e., an implicit fatigue analysis).

Issue:

TheAMR items on "cracking -fatigue" in LRA Tables 3.1.2-1,3.1.2-2,3.1.2-3, and 3.1.2-4 do not clearly identify which components are specifically within the scope of the commodity groups in the AMR items. In addition, in LRA Table 3.1.2-5, the applicant identifies that plant analyses for the non-Safety Class 1/non-Safety Class A rupture discs, tubing and valve bodies in the RCS include fatigue-related TLAAs; yet, these components are not identified as being within the scope of any of the fatigue-related TLAAs that are discussed in the subsections of LRA Section

4.3. Request

1. Identify all components that are within the scope of the commodity groups in the AMR items on "cracking - fatigue" in LRA Tables 3.1.2-1, 3.1.2-2, 3.1.2-3, and 3.1.2-4.

2. For those AMR items included in LRA Table 3.1.2-5 on "cracking - fatigue" of non-Safety Class 1/non-Safety Class A rupture discs, tubing and valve bodies, provide a basis for why LRA Section 4.3 does not mention that these non-Safety Class 1Jnon-Safety Class A RCS components were within the scope of an applicable fatigue usage factor analysis, fatigue waiver analysis, or maximum allowable stress range reduction analysis.

RAI B.1.33-1

Background:

LRA Section 8.1.33 describes the existing Reactor Head Closure Stud Bolting Program as consistent, with enhancements with GALL AMP XI.M3, "Reactor Head Closure Stud Bolting."

The LRA also states that one of the studs has measured yield strength of 150.7 ksi. The applicant's enhancement states that replacement closure bolting will be fabricated from bolting material with actual measured yield strength less than 150 ksi, consistent with the recommendations of the GALL AMP. The staff noted that the applicant did not state any exceptions to the GALL Report AMP.

Issue:

LRA Section 8.1.33 states that preventive actions include use of bolting material that has actual yield strength of less than 150 ksi for all studs. However, one stud has measured yield strength of 150.7 ksi.

Request:

1. Clarify if the stud with the measured yield strength of 150.7 ksi will be replaced prior to the start of the period of extended operation; otherwise provide a basis for not taking an exception to the GALL Report AMP XI.M3 for use of bolting with greater than 150 ksi measured yield strength.

2. Revise the LRA as necessary and consistent with the response.

-5 RAI B.1.40-6

Background:

LRA Section B.1.40, Structures Monitoring, states an enhancement to the "scope of program" program element. In this enhancement, the applicant stated that the Structures Monitoring Program procedures will be revised to specify each of the in-scope structures and structural components for each of the Structures Monitoring, Regulatory Guide (RG) 1.127, Inspection of Water-Control Structures Associated with Nuclear Power Plants, and Masonry Wall AMPs.

Issue:

The staff understands this revision to the Structures Monitoring Program procedures to be an enhancement, describe in LRA Appendix A (UFSAR Supplement), to the existing program in order to make the program consistent with the GALL Report; however, because this is an existing program. it is not clear which structures and structural components are being added to the scope of the program for license renewal. that are not already within the existing program.

Request:

Identify the structures and structural components and commodities that are being added to the scope of the Structures Monitoring Program for license renewal, that are not currently listed in the existing Structures Monitoring Program.

RAI 8.1.40-7

Background:

LRA Section A.1.40, Structures Monitoring Program, provides a summary description of the program to be incorporated as part of the UFSAR Supplement. The GALL Report provides an option to include the inspection of masonry walls and water-control structures within the scope of the Structures Monitoring Program provided all the attributes of GALL Report AMP XI.S5, "Masonry Wall" and XI.S7, "RG 1.127, Inspection of Water-Control Structures Associated with Nuclear Power Plants" are incorporated in the attributes of the Structures Monitoring Program.

As such, the applicant has identified the enhancements to the Structures Monitoring Program and UFSAR supplement, which incorporate the attributes necessary for making the programs consistent with the GALL Report.

Issue:

The enhancements, described in LRA Appendix A (UFSAR Supplement). to the Structures Monitoring Program primarily describe revisions to the Structures Monitoring Program procedures, which will be completed prior to the period of extended operation (Sequoyah Nuclear Power Station (SQN), Unit 1: 9/17/2020 and SQN, Unit 2: 9/15/2021). However, it is not clear that the implementation of the activities associated with the revision to the procedures (e.g., inspection of additional structures, groundwater sampling and chemical analysis, verification of acceptance criteria, etc.) will be completed prior to the period of extended operation.

- 6 Request:

Revise the UFSAR supplement and associated commitment(s) to clarify when the implementation of the activities associated with the revisions to the Structures Monitoring Program procedures, which have been identified as enhancements, will take place.

RAI3.6-3

Background:

In LRA Section 3.6.2.2.3, the applicant states that the design of switchyard bus bolted connections precludes torque relaxation as confirmed by plant-specific operating experience.

The design of switchyard bolted connections includes Bellville washers. The type of bolting plate and the use of Bellville washers is the industry standard to preclude torque relaxation.

Issue:

EPRI document TR-104213, "Bolted Joint Maintenance & Application Guide," identifies a special problem with Belleville washers. It states that hydrogen embrittlement is a recurring problem with Belleville washers and other springs. When springs are electroplated, the plating process forces hydrogen into the metal grain boundaries. If the hydrogen is not removed, the spring may spontaneously fail at any time while in service.

Request:

Identify if electroplated Belleville washers are currently used at SQN. If they are, explain why hydrogen embrittlement is not a concern for switchyard bus bolted connections at SQN.

RAI 3.5.2.2.2.1-1

Background:

SRP-LR Sections 3.5.3.2.2.1.1 and 3.5.3.2.2.3.1 state that loss of material and cracking due to freeze-thaw could occur in inaccessible concrete areas of Groups 1-3, 5, and 7-9 as well as Group 6 structures. Further evaluation is needed for plants located in moderate to severe weathering conditions. The SRP-LR further states that a plant-specific program is not required if documented evidence confirms that the concrete has an air content between 3 and 8 percent and inspections have not identified degradation related to freeze-thaw. SRP-LR Table 3.5-1, items ID 42 and ID 49 address freeze-thaw in inaccessible concrete areas.

Issue:

In LRA Table 3.5.1, items 42 and 49, the applicant stated that freeze-thaw does not require management. In the associated further evaluation sections (3.5.2.2.2.1, item 1 and 3.5.2.2.2.3, item 1), the applicant stated that TVA's construction specifications require all concrete to contain air-entraining agent in sufficient quantity to maintain specified percentages; therefore, loss of material and cracking due to freeze-thaw in inaccessible concrete areas are not aging effects that require aging management. However, the applicant does not state the actual air content in the concrete, or discuss results of past inspections that demonstrate freeze-thaw degradation is not an issue.

-7 Request:

1. Provide the air content values of the concrete in the Groups 1-3, 5, 7-9 and Group 6 structures.

2. Explain whether or not past inspections have identified degradation that was attributed to freeze-thaw degradation.

RAI 3.5.2.3.1-1

Background:

LRA Table 3.5.2-1 states that for fiber reinforced polyester (FRP) lower inlet doors exposed to uncontrolled indoor air, aging effects are not applicable and no AMP is proposed. The AMR item cites generic note J. The AMR also cites plant-specific note 502 which states that the "material is encapsulated within a stainless steel sheet steel paneL" UFSAR Section 6.5.9 describes each lower inlet door as consisting of a 0.5 in. thick FRP plate stiffened by six steel ribs, bolted to the plate. Seven inches of urethane foam are bonded to the back of the FRP plate to provide thermal insulation, and the front and back surfaces of the door are protected with 26 gauge stainless steel covers which provide a complete vapor barrier around the insulation.

Issue:

The staff notes that fiber reinforced polyester can be constructed with different bonding agents/resins which may respond differently to environmental factors. UFSAR section 6.5.9 states that the maximum radiation at the inlet doors is 5 r/hr gamma during normal operations, and there is no secondary radiation due to neutron exposure. The staff does not have sufficient information to conclude that there would be no aging effect requiring management (AERM) for the environments to which the FRP is exposed.

The staff also notes that the urethane foam described in UFSAR Section 6.5.9, which provides an insulation function for the doors, has not been evaluated in the LRA.

Reguest:

1. Provide the technical basis for concluding that there is no AERM, or identify the potential aging effects and propose an AMP to manage the aging effects for the FRP in the lower inlet doors.

2. Identify potential aging effects associated with the urethane foam used for thermal insulation and propose an AMP to manage the aging effects, or provide the technical justification for why there are no AERM.

RAI 3.5.2.3.4-2

Background:

LRA Table 3.5.2-4 states that for fiberglass seismic/expansion joint exposed to outdoor air, aging effects are not applicable and no AMP is proposed. The AMR items cite generic note J.

- 8 Regulatory Issue Summary 2012-02, "Insights Into Recent License Renewal Application Consistency with the Generic Aging Lessons Learned Report," states that when an applicant states that there is no AERM and no proposed AMP, the application should state the specific material and provide greater detail on the specific environment (e.g., ultraviolet light, ozone).

Issue:

The staff notes that fiberglass seismic/expansion joint can be constructed with different bonding materials which may respond differently to environmental factors. The staff does not have sufficient information to conclude that there would be no AERM for the environments to which the fiberglass seismic/expansion joint is exposed.

Request:

Provide the technical basis for concluding that there is no AERM, or identify the potential aging effects and propose an AMP to manage the aging effects for fiberglass seismic/expansion joint.

RAI3.5.1-1

Background:

SRP-LR Section 3.5 includes several subsections (e.g., 3.5.2.2.2.1, 3.5.2.2.2.3) which identify aging effects (loss of material due to freeze-thaw, cracking due to reaction with aggregates, etc.) that do not require additional plant specific aging management for inaccessible concrete areas if certain conditions can be met. SRP-LR Table 3.5-1 includes line items for the same aging effects for accessible concrete areas and recommends GALL Report AMPs to manage the effects of aging, with no associated conditions that can be met to consider the aging effect not applicable. The staff expects these aging effects to be included within the recommended structural AMP.

Issue:

1. The Discussion column of several items associated with inaccessible concrete (42, 47, 49, and 51) in LRA Table 3.5.1 states that "listed aging effects do not require aging management at SQN." The staff does not agree, and has not been provided adequate plant-specific technical basis to support that statement. As noted above, pending an acceptable further evaluation, the staff believes that the listed aging effects do not require additional plant specific management if the aging effects are addressed appropriately for accessible areas.

2. LRA Table 3.5.1, Items 43,50 and 54 address cracking due to reaction with aggregates in accessible and inaccessible concrete. For the inaccessible concrete (Items 43 and

50) the Discussion column states that "listed aging effects do not require aging management at SON," while Items 54 states " ... the design and construction of these groups of structures at SQN prevents the effect of this aging form occurring; therefore, this aging effect does not require management ..." The staff does not agree, and has not been provided adequate plant-specific technical basis to support that statement.

Regardless of the design and construction of the concrete, the staff believes all aging effects could occur in accessible areas and, therefore, require management. The discussion in the LRA states that the components are included in the Structures

- 9 Monitoring Program; however, the associated line items do not appear in any of the LRA "Table 2's."

Request:

1. For each item that states "listed aging effects do not require aging management at SON" clarify whether this means no additional plant specific aging management is required for the inaccessible areas or if this means no aging management is required. If it is the latter, provide a technical justification for why that aging effect does not require management.

2. Provide a technical justification for why cracking due to reaction with aggregates does not require management in accessible or inaccessible areas or identify a program to manage this aging effect. If a program is identified to manage this aging effect. updated the LRA accordingly.

RAI4.7.2-1

Background:

SRP-LR 4.7.3.1.1 provides the NRC's review procedures for reviewing plant-specific TLAAs that are accepted in accordance with 10 CFR 54.21 (c)(1 )(i). The SRP-LR states that the existing analyses must be verified to be valid and bounding for the period of extended operation.

SRP-LR Section 4.7.3.1.1 instructs the NRC reviewer to review the TLAA justification provided by the applicant in order to verify that the existing analyses are valid for the period of extended operation. SRP-LR Section 4.7.3.1.1 states that the existing analyses should be shown to be bounding even during the period of extended operation. Cranes built to Crane Manufacturer's Association of America Specification #70 (CMAA-70) are qualified for 100,000 load cycles. LRA Section 4.7.2 states that the manipulator cranes are the only cranes that included CMAA-70 in their design specifications.

Issues:

1. The staff reviewed the SON UFSAR - 23 and did not find information related to the applicable codes, standards, and specifications for the design or analysis of the manipulator cranes.

2. In LRA Section 4.7.2 the applicant stated that the number of lifts each manipulator crane would experience in 60 years, assuming a 1.25 multiplier for safety margin, is -20,500 lift cycles; far below the 100,000 cycle limit. However. the applicant did not provide any information on how that estimate was developed.

Requests:

1. Explain how the Manipulator Cranes at SON. Units 1 and 2, were determined to meet the design specifications of CMAA-70.

2. Explain and justify how the 20.500 estimated lift value was determined for 60 years.

- 10 RAI4.7.2-2 8ackground:

LRA Section 4.7.2 states that "No other cranes [besides the manipulator crane] at SQN were built to CMAA-70 requirements ... The SQN responses to NUREG-0612 and the review of the site cranes identified that the reactor building polar crane and the auxiliary building crane were not built to the structural fatigue requirements of CMAA-70."

UFSAR Section 3.12.4.1 contains SQN commitments in response to NUREG 0612, which recommends compliance with seven guidelines to ensure the Control of Heavy Loads Program is adequate. Guideline 7 states that 'The crane should be designed to meet the applicable criteria and guidelines of Chapter 2-1 of 8301.2-1976 and CMAA-70." The applicant's response states "The actual design data for the auxiliary building crane and the reactor building crane were compared with the guidelines of CMAA-70 and ANSI (ASME) 830.2." Where specific compliance was not evident by review, an evaluation was made by imposing these guidelines on the actual design ... this was the approach used for evaluating the design of major structural components by using load combinations and allowable stresses given in CMAA-70. The results of this review and analysis indicate that both cranes meet or exceed the requirements of CMAA 70 and ANSI (ASME) B30.2."

UFSAR Section 3.8.6.2.2 "Applicable Codes, Standards, and Specifications" of the SQN UFSAR states that the requirements of CMAA-70 were used to upgrade the Auxiliary Building Crane to single failure proof crane systems.

Issue:

While the original design of the reactor building and auxiliary building cranes may not have directly incorporated the guidelines of CMAA-70 and ANSI 830.2, several analyses have been done to compare the design of the auxiliary building and reactor building cranes to CMAA-70 and ANSI B30.2 to demonstrate compliance with the guidance outlined in NUREG 0612. In addition, CMAA-70 was used in an analysis to upgrade the Auxiliary Building Crane design.

The staff believes that since these analyses and comparisons to the criteria and guidelines of CMAA-70 and ANSI 830.2 for the auxiliary building crane and reactor building crane are outlined in the UFSAR, the applicant's review of the compliance of the auxiliary and reactor building cranes to the CMAA -70 standard meets the criteria for a TLAA.

Request:

Provide basis for the conclusion that current licensing basis does not incorporate the applicable design specifications of CMAA-70 and ANSI 830.2, and does not consider the TLAA analyses for the Auxiliary and Reactor Building Cranes.

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) - SET 8 DISTRIBUTION:

HARDCOPY:

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PUBLIC RidsNrrDlr Resource RidsNrrDlrRpb1 Resource RidsNrrDlrRpb2 Resource RidsNrrDlrRerb Resource RidsNrrDlrRarb Resource RidsNrrDlrRasb Resource beth.mizuno@nrc.gov brian. harris@nrc.gov john.pelchat@nrc.gov gena.woodruff@nrc.gov siva.lingam@nrc.gov wesley.deschaine@nrc.gov galen.smith@nrc.gov scott.shaeffer@nrc.gov jeffrey. hamman@nrc.gov craig. kontz@nrc.gov caudle.julian@nrc.gov generette.lloyd@epa.gov gmadkins@tva.gov clwilson@tva.gov hleeO@tva.gov dllundy@tva.gov