ML18116A009

COL Docs - (External_Sender) Draft RAI Responses for Thursday, May 3, Public Meeting

ML18116A009
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Site:	Vogtle
Issue date:	04/25/2018
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1 Vogtle PEmails From:

Haggerty, Neil <X2NHAGGE@SOUTHERNCO.COM>

Sent:

Wednesday, April 25, 2018 8:56 PM To:

Habib, Donald Cc:

Patel, Chandu; Hoellman, Jordan; Sparkman, Wesley A.; Redd, Jason P.; Hicks, Thomas E.; Amundson, Theodore Edwin; Aughtman, Amy G.; Haggerty, Neil

Subject:

[External_Sender] Draft RAI responses for Thursday, May 3, Public Meeting Attachments:

2018-05-03_RAI LAR-17-037-2_SEB - DRAFT RAI 9530 response for Public Mtg.pdf; 2018-05-03_RAI LAR-17-037-3_MCB - Draft RAI 9477 response for Public Mtg.pdf

Don, Attached are draft responses to the following Requests for Additional Information for discussion in the Public Meeting on Thursday, May 3:

RAI LAR-17-037-2, eRAI 9530, Structural Engineering Branch (SEB) questions regarding Critical Sections RAI LAR-17-037-3, eRAI 9477, Materials & Chemistry Branch (MCB) questions regarding Reactor Coolant Pumps Please provide these draft responses to the technical staff responsible for your reviews, to allow them sufficient time to familiarize with our responses.

The information provided in these responses do NOT contain Sensitive Unclassified Non-Safeguards Information (SUNSI),

and may be released to the Public.

Please contact me or any personnel on cc: for this message if you have any questions regarding this information.

Thank you, Neil Haggerty Neil Haggerty l Southern Nuclear Operating Company Nuclear Development Regulatory Affairs - VEGP 3&4 Licensing SNC-Inverness: 205.992.7047 l office: 301.874.8537 l mobile: 240.566.2442 x2nhagge@southernco.com neil.haggerty@excelservices.com This e-mail and any attachments thereto are intended only for the use by the addressee(s) named herein and may contain proprietary and confidential information. If you have received this e-mail in error, please immediately notify me by telephone and permanently delete the original and any copy of any e-mail and any printout thereof.

Hearing Identifier:

Vogtle_COL_Docs_Public Email Number:

265 Mail Envelope Properties (22CCC8EC7A80FC4189AACACBEBCAC516F03FC4)

Subject:

[External_Sender] Draft RAI responses for Thursday, May 3, Public Meeting Sent Date:

4/25/2018 8:56:00 PM Received Date:

4/25/2018 8:56:48 PM From:

Haggerty, Neil Created By:

X2NHAGGE@SOUTHERNCO.COM Recipients:

"Patel, Chandu" <Chandu.Patel@nrc.gov>

Tracking Status: None "Hoellman, Jordan" <Jordan.Hoellman2@nrc.gov>

Tracking Status: None "Sparkman, Wesley A." <WASPARKM@southernco.com>

Tracking Status: None "Redd, Jason P." <JPREDD@southernco.com>

Tracking Status: None "Hicks, Thomas E." <X2TEHICK@southernco.com>

Tracking Status: None "Amundson, Theodore Edwin" <X2TAMUNS@southernco.com>

Tracking Status: None "Aughtman, Amy G." <AGAUGHTM@SOUTHERNCO.COM>

Tracking Status: None "Haggerty, Neil" <X2NHAGGE@SOUTHERNCO.COM>

Tracking Status: None "Habib, Donald" <Donald.Habib@nrc.gov>

Tracking Status: None Post Office:

ALXAPEX242.southernco.com Files Size Date & Time MESSAGE 1533 4/25/2018 8:56:48 PM 2018-05-03_RAI LAR-17-037-2_SEB - DRAFT RAI 9530 response for Public Mtg.pdf 823875 2018-05-03_RAI LAR-17-037-3_MCB - Draft RAI 9477 response for Public Mtg.pdf 685160 Options Priority:

Standard Return Notification:

No Reply Requested:

No Sensitivity:

Normal Expiration Date:

Recipients Received:

[DRAFT 4/25/18]

Page 1 of 6 The following is a question provided by the NRC Staff [Request for Additional Information (RAI)

LAR 17-037-2] regarding the review of Southern Nuclear Operating Company (SNC) License Amendment Request (LAR)17-037, which was submitted by SNC letter ND-17-1726 on December 21, 2017 [ADAMS Accession No. ML17355A416].

Question The final safety analysis report of the Vogtle Electric Generating Plant (VEGP) Units 3 and 4 references the Westinghouse AP1000 certified design. Appendix D to 10 CFR Part 52, Design Certification Rule for the AP1000 Design, provides the regulatory requirements for the AP1000 design. 10 CFR Part 52, Appendix D, Section VIII.B.6.c provides a list of Tier 2* matters, including a design summary of critical sections, that a licensee who references this appendix may not depart from without NRC approval.

Furthermore, SECY-17-0075, Planned Improvements in Design Certification Tiered Information Designations, described the staffs approach to using the Tier 2*

designation for safety significant information. The SECY noted that if Tier 2* were to be eliminated, certain safety-significant information currently in Tier 2* should be included in Tier 1 rather than in Tier 2. The staff considers that a critical section has attributes that make it safety significant in maintaining the integrity of the plant structure. The designed capacity of the critical sections support the reasonable assurance of safety determination for the AP1000 DCD, Rev.19 design in the staff safety evaluation.

The staff reviewed the LAR and noted that the criteria for screening Tier 2* information pertaining to critical sections is not well defined.

In Enclosure 3, Proposed Changes to Licensing Basis Documents, of the LAR, the licensee proposed to revise its combined license (COL) to include a new license condition to address the Tier 2* change process. The licensee included a new license condition, proposed License Condition 13, Departures from Plant-Specific DCD Tier 2*

Information. The proposed license condition states that the licensee

... is exempt from the requirements of 10 CFR Part 52, Appendix D, Paragraphs II.F and VIII.B.6 that invoke the Tier 2* change process that requires prior NRC approval via a license amendment for departures from Tier 2* information; and Paragraph VIII.B.5.a for Tier 2 information that involves a change to, or departure from, Tier 2* information; except for departures from Tier 2* information that:

1. Involve design methodology or construction materials that deviate from a code or standard credited in the plant-specific DCD for establishing the criteria for the design or construction of a structure, system, or component (SSC) important to safety.

The proposed license condition is not clear as to how the critical sections associated with the steel-concrete (SC) modular construction would be screened using the above DRAFT C)

C) 726 on 726 o g Plant (VEGP) Units 3 g Plant (VEGP) U

n. Appendix D to 10 CFR

n. Appendix D to 10 CF n, provides the regulatory n, provides the regulatory Appendix D,Section VIII.B.6.c Appendix D,Section VIII.B.6.c summary of critical sections, that a summary of critical sections, that a epart from without NRC approval.

epart from without N ements in Design Certification Tiered ments in Design Certi ffs approach to using the Tier 2*

pproach to using t on. The SECY noted that if Tier 2* were ECY noted t ormation currently in Tier 2* should be in tly in Ti aff considers that a critical section has aff considers that a critical aintaining the integrity of the plant struct ntaining the integrity of the plant al sections support the reasonable assu tions support the reasonable ass 0 DCD, Rev.19 design in the staff safety ev.19 design in the staff safety LAR and noted that the criteria for scree AR and noted that the criteria for scree sections is not well defined.

sections is not well defined.

Proposed Changes to Licensi Proposed Changes to Licensing Basis ng posed to revise its combined license (CO posed to revise its combined license (

o address the Tier 2* change process. T address the Tier 2* change process.

n, proposed License Condition 13, Dep posed License Condition 13, Dep mation. The proposed license condition he proposed license condition

... is exempt from the requireme

... is exempt from the Paragraphs II.F and VIII.B.6 th agraphs II.F and VIII.

requires prior NRC approva prior NRC approv Tier 2* information; and P ormation; and P involves a change to, o chan departures from Tie from 1.

1. Involve des lve des a code o a code o the cri the cri com com

[DRAFT 4/25/18]

Page 2 of 6 criteria because information from analysis and tests were used in conjunction with codes and standards for the design of SC modules. As approved in the certified design, linear analysis, nonlinear analysis, and testing of the SC module design were performed and the results were compared to provisions of two different codes in order to validate the use of the codes.

The staff considers that the critical sections have safety significance in assuring the integrity of the building which house safety related systems and components. The proposed Criterion 1 relies on code compliance in the design and detailing of the critical sections to screen out details that are code controlled. The application of this criteria may lead the applicant to conclude that the parameters of the critical sections can be modified in the field using available NRC change processes without resorting to the license amendment process. The staff finds instances where the application of this criterion will not yield the desired results. The staff has identified the following cases as exceptions to the Criterion 1:

x Critical sections using steel concrete sandwich construction, and x

the capacity aspects such as area of steel provided or the demand to capacity ratio of critical sections using reinforced concrete In both cases, the staff has determined that neither the design nor the cited attributes of the critical sections are code defined, making Criterion 1 in-applicable in these instances. The staff requests the applicant to revise the Criterion 1 such that the conditions identified above are screened in and a license amendment process followed for any changes to these cases, or that the applicant provide additional explanation as to why the proposed criteria would not need to be revised in order to maintain a reasonable assurance of safety.

SNC Response to RAI Question 1 Tier 2* information is intended to have substantial safety significance, commensurate with information designated as Tier 1. As noted in SECY-17-0075, Planned Improvements in Design Certification Tiered Information Designations, [ADAMS Accession Number ML16196A321], the Tier 2* scope identified in previous design certifications, such as AP1000, may be broader than necessary, and includes information more appropriately designated as Tier 2. SNC proposes to invoke a process whereby VEGP 3 and 4 Tier 2* departures would be submitted to the NRC for prior approval when the safety level rose to that which is commensurate with the safety level of Tier 1 information.

While Updated Final Safety Analysis Report (UFSAR) Appendix 3H, Auxiliary and Shield Building Critical Sections, contains significant critical section detailed design, including detailed figures of critical sections, the majority of the AP1000 structural design requirements are derived from applicable codes. SNC acknowledges that for the shield building design nonlinear analysis and testing were performed to validate the use of applicable codes. The performance of these activities, however, does not invalidate SNCs position that the design of the shield building and other critical sections is based in large part on meeting applicable industry codes. Hence, the to v to v e in assuring the e in assu components. The components.

and detailing of the critical and detailing of the application of this criteria application of this criter f the critical sections can be f the critical sections can be esses without resorting to the esses without resorting to the ces where the application of this ces whe aff has identified the following cases as aff has identified the following cases as ete sandwich construction, and h constructi area of steel provided or the demand to area of steel provided or the ng reinforced concrete reinforced con determined that neither the desig d that neither the design nor th n nor th ode defined, making Criterion 1 in making Criterion 1 in-applic applic quests the applicant to revise the Criteri quests the applicant to revise the Criteri above above are screened in and a license a are screened in and a m

o these cases, or that the applicant prov o these cases, or that the appl posed criteria would not need to be revis posed criteria would not need to be assurance of safety.

assurance of safety.

Response to RAI Question to RAI Questi 1

DR Tier 2* information is intended to have subst ier 2* information is intended information designated as Tier 1. As noted information designated as Tier 1 Certification Tiered Information Designa Certification Tiered Information Des Tier 2* scope identified in previous d Tier 2* scope identified in previous d necessary, and includes informatio ludes informatio invoke a process whereby invoke a process whereb VEG prior approval when the safe prior approval when the s Tier 1 information Tier 1 i e Updated Final Updated Final g C g Critical Se ritical Se critica critica ab ab

[DRAFT 4/25/18]

Page 3 of 6 proposed evaluation Criterion 1, which requires prior NRC approval for any Tier 2* text, table and figure change that deviates from these codes, combined with the 10 CFR Part 52, Appendix D, Section VIII.B.5.b evaluation criteria, does provide a reasonable assurance of safety.

Nevertheless, SNC has identified enhancements to Criterion 1 that expand the scope of this Criterion to include requirements described in the UFSAR that supplement code requirements for critical sections (discussed below).

In addressing the staffs question, this response is divided into three areas: reinforced concreted (RC) design; concrete-filled steel plate construction (SC) module design; and shield building design.

RC Design Design requirements for RC structures are governed by accepted industry codes as described in the following UFSAR subsections:

x UFSAR Subsection 3H.3.1, Governing Codes and Standards, describes the primary codes and standards used in the design of the auxiliary and shield buildings: American Concrete Institute (ACI) standard ACI 349-01, Code Requirements for Nuclear Safety Related Concrete Structures (and Subsection 3.8.4.5 for supplementary requirements and Subsection 3.8.4.4.1 for alternative requirements); American National Standards Institute (ANSI) / American Institute of Steel Construction (AISC) standard ANSI/AISC N690-1994, Specification for the Design, Fabrication and Erection of Safety-Related Steel Structures for Nuclear Facilities (and Subsection 3.8.4.5 for supplemental requirements); American Welding Society (AWS), Structural Welding Code - Steel, AWS D1.1-2000 (provides an acceptable alternative for AISC N690 weld requirements as described in Subsections 3.8.3.2 and 3.8.4.2).

x UFSAR Subsection 3H.5.1, Shear Walls, states that the wall sections are designed in accordance with the requirements of ACI 349-01.

x UFSAR Subsection 3H.5.2, Composite Structures (Floors and Roof), states that the designs of the floors are in conformance with AISC N690 and ACI 349. This section also requires that the reinforcement size and spacing are based on loads and spans for this type of floor and are determined at each location based on the requirements in ACI 349 and ACI 318-11. The slab concrete and the reinforcement is designed to meet the requirements of ACI 349-01.

x UFSAR Subsection 3H.5.3, Reinforced Concrete Slabs, states that the design of these floors is in conformance with AISC N690 and ACI 349. The reinforcement size and spacing are determined for each location, based on specific loads and spans, and satisfy the requirements in ACI 349 and ACI 318-11. The precast panels are connected to the concrete placed above them by shear reinforcement which satisfies the requirements of ACI 349.

x UFSAR Subsection 3H.5.4, Concrete Finned Floors, states that the finned floors are designed as reinforced concrete slabs in accordance with ACI 349. Composite section properties, based on an all steel-transformed section, as detailed in Section Q1.11 of ANSI/AISC N690-94 are used to design the weld strength between stiffener and the steel plate and the spacing of the shear studs for the composite action. The plate is designed against the criteria for bending and shear, specified in ANSI/AISC N690-94.

DRAFT e scope of th e scope cod code requirements e re areas: reinforced concreted areas: reinforced c design; and shield building design; and shield buildin by accepted industry codes by accepted industry c as described g Codes and Standards d Standards, describes the prima d

esign of the auxiliary and shield buildings:

ary and A

ACI AC 349-01, Code Requirements for Nucl equirem (and and Subsection 3.8.4.5 for supplementary Subsection 3.8.4.5 for supplem r alternative ernative requirements);

requireme American Natio rican Na an Institute of Steel Construction (AISC) sta te of Steel Construction (AISC) sta tion for the Design, Fabrication and Erectio Design, Fabrication and Erec Nuclear Facilities es (and S (and Subsection 3.8.4.5 ubsection 3.8.4.5 merican Welding Society (AWS), Structural merican Welding Society (AWS), Structural ovides an acceptable alternative for AISC N ovides an acceptable alternative Su Subsections 3.8.3.2 and 3.8.4.2).

bsections 3.8.3.2 and 3.8.4.2 ubs ubsection ec 3H.5.11, Shear Walls S

, states that stat ance with t ance w he requirements of ACI 349 uirements of ACI 349-01.

0 SAR S AR Subs ub ection 3H.5.2 H.5.2, Composite Structu Composite Structu esigns of the floors are in conformance with s of the floors are in conformance with requires that t hat the reinforcement size and s he reinforcement size and s type of floor and are determined at eac type of floor and are d and ACI 318 and ACI 3

-11.

1 The slab concrete a The s requirements of ACI 349 requirements of ACI 349-01.

x UFSAR S R Subs ubsection e

3H.5.3 3H.5.3, Re floors is in conformance with conformance with spacing are determined dete fo the requirements in AC ents concrete placed abov ced a ACI 349.

x UFSAR S UFSAR Subs ubsec ec designed as designed as properties properties SI/A SI/A

[DRAFT 4/25/18]

Page 4 of 6 These codes provide requirements for, in part, capacity aspects such as the demand-to-capacity ratio of walls and floors using reinforced concrete. Hence, no new evaluation Criterion is considered necessary to address demand-to-capacity ratios for RC structures. However, SNC acknowledges that the UFSAR describes requirements for the design of critical sections that supplement the code requirements in some areas, as follows:

x UFSAR Subsection 3.8.4.5.1, provides supplemental requirements for concrete structures beyond ACI 349.

Therefore, proposed evaluation Criterion 1 will be revised to add a requirement that prior NRC staff review is needed to deviate from these supplemental requirements in addition to the existing Criterion 1 topic of code provisions.

SC (Sandwich) Design Design requirements for SC module design are primarily governed by accepted industry codes as described in the following UFSAR subsections:

x UFSAR Subsection 3.8.3.5.3, Structural Wall Modules, states that structural modules without concrete fill, such as the west wall of the in-containment refueling water storage tank, are designed as steel structures, according to the requirements of AISC N690.

Concrete-filled structural wall modules are designed as reinforced concrete structures in accordance with the requirements of ACI 349 and other code requirements as detailed in this UFSAR subsection. The reinforcing steel used to anchor the modules to the concrete has a development that satisfies the requirements of ACI 349.

x UFSAR Subsection 3H.3.1, Governing Codes and Standards, describes the primary codes and standards used in the design of the auxiliary and shield buildings: ACI 349-01, Code Requirements for Nuclear Safety Related Concrete Structures (and Subsection 3.8.4.5 for supplementary requirements and Subsection 3.8.4.4.1 for alternative requirements); ANSI/AISC N690-1994, Specification for the Design, Fabrication and Erection of Safety-Related Steel Structures for Nuclear Facilities (and Subsection 3.8.4.5 for supplemental requirements); American Welding Society (AWS),

Structural Welding Code - Steel, AWS D1.1-2000 (provides an acceptable alternative for AISC N690 weld requirements as described in Subsections 3.8.3.2 and 3.8.4.2).

x UFSAR Subsection 3H.5.5, Structural Modules, states that the design methodology of these modules in the auxiliary building is similar to the design of the structural modules in the containment internal structures described in Subsection 3.8.3.5.3. These modules include the spent fuel pool, fuel transfer canal, and cask loading and cask washdown pits.

o UFSAR Subsection 3H.5.5.1, West Wall of Spent Fuel Pool, states that the concrete filled structural wall modules are designed as reinforced concrete structures in accordance with the requirements of ACI 349. The face plates are treated as reinforcing steel.

These codes provide a comprehensive set of requirements for SC structures. However, SNC acknowledges that the UFSAR describes requirements for the design of SC modules that supplement the code requirements in some areas, as follows:

x UFSAR Subsection 3.8.4.5.2, provides supplemental requirements for steel structures beyond AISC N690.

DRAFT alal for or concrete conc equirement that prior NRC equirement that prio ments ments in addition to the n to the arily arily governed by accepted industry codes governed by accepted industry codes al Wall Modules dules, states that structural mod states t st wall of the in-containment refueling wate ontainm tures, according to the requirements of tures, according to the require AIS modules are designed as reinforced concre odules are designed as reinforced ements of ts of ACI 349 ACI 349 and other code requirem and other code require The reinforcing steel used to anchor the mo forcing steel used to anchor the mo pment that satisfies the requirements of AC atisfies the requirements of A 3H.3.1, Governing ng Codes and Standards, d Codes and Standards, d ards used in the design of the auxiliary and ards used in the design of the a quirements for Nuclear Safety Related Conc quirements for Nuclear Safety R 3.8.4.5 for supplementary requirements 3.8.4.5 for supplementary require an e requirements); ANSI/AISC N690 e requirements); ANSI/AISC N690-1994, S 1

ation and Erection of Safety ation and Erection of S

-Related Steel S d Ste section 3.8.4.5 for supplemental requireme ection 3.8.4.5 for supplemental requireme Structural Welding Code ral Welding Code - Steel, AWS D1.1 Steel, AWS D1.1 AISC N690 weld requirements as describe 90 weld requirements as describe x

UFSAR Subsection 3H.5.5 UFSAR Subsection 3

, Structural M these modules in the auxiliary buildin these modules in the au in the c the containment internal structu ontainment interna include the spent fuel pool, fuel the spent fuel pool, f pits.

o UFSAR Subsection AR Subsection concrete filled st ete f structures in a tures treated as ated a se codes provide se codes provide wledges tha wledges tha nt the nt the

[DRAFT 4/25/18]

Page 5 of 6 x

UFSAR Subsection 3H.3.4, provides supplemental requirements to AISC N690 for load combinations and stress limit coefficients.

Therefore, proposed evaluation Criterion 1 will be revised to add a requirement that prior NRC staff review is needed to deviate from these supplemental requirements in addition to the existing Criterion 1 topic of code provisions.

Shield Building Design The shield building uses SC as well as RC construction. As described in the UFSAR, Subsection 3.8.4.1.1, Shield Building, and Appendix 3H the design of much of the shield building is based on compliance to codes. This point is stated in NUREG-1793, Final Safety Evaluation Related to Certification of the AP1000 Standard Design, Supplement 2, Subsection 3.8.4.1.1.3.1, Design Methodology and Process for Shield Building Design, [ADAMS Accession No. ML112061231] which states:

the concrete design of the following areas of the AP1000 shield building falls directly within the scope of ACI 349:

x shield building roof x

knuckle region of the roof near the PCCWST wall x

compression ring x

PCCWST The applicant designed these areas in accordance with the provisions in the established design codes by using linear elastic analysis methods. Specifically, the design for the sections in these areas is based on compliance with the ACI 349 Code, as supplemented with guidance in NRC Regulatory Guide (RG) 1.142 for concrete structures. The design of the sections in these areas, which uses established design codes and analysis methods listed in Section 3.8.4 of NUREG-0800, satisfies the regulatory basis listed above and is, therefore, acceptable to the staff.

The applicants integrated design process also makes use of the design process for structural steel components in certain areas of the shield building. Specifically, it uses ANSI/AISC N690 in designing structural steel components of seismic Category I structures. The applicant used ANSI/AISC N690 in designing the following areas of the shield building:

x the steel roof that supports the concrete roof slab x

tension ring x

SC/RC connection The design process uses provisions from two different design codes: ACI 349 Code for RC components, which uses an ultimate strength design approach and ANSI/AISC N690 Standard for steel and composite components, which uses an allowable stress design approach.

The proposed evaluation Criterion 1, which addresses deviations from codes, does provide sufficient restrictions (i.e., obtain prior NRC staff approval) on shield building design changes involving these above areas when the change deviates from these codes.

DRAFT t pp on to the on to th d in the UFSAR, d in the UFSAR, n of much of the shield n of much of the shield n NUREG NUR

-1793, Final Safety al Safety Design, Supplement 2

Design,

, Subsection bsection d Building Design d Buildin

, [ADAMS Accession ession as of the AP1000 shield building falls directl of the AP1000 shield b f near the PCCWST wall f near the PCCWST wall d these areas in accordance with the provis s in accordance with the prov ing linear elastic analysis methods. Specific analysis methods. Specific areas is based on compliance with the ACI areas is based on compliance with guidance in with guidance NRC Regulatory Guide (

Regulat R

The desig The des n of the sections in these areas, w of the sections in thes d analysis methods listed in Section 3.8.4 o d analysis methods listed in Section tory basis listed above and is, therefore, acc tory basis listed above and is, therefore, The applicants integrated design pr applicants integrated design process ocess structural steel components in certain area steel components in certain area ANSI/AISC N690 in designing structura ANSI/AISC N690 in d structures. The applicant used ANSI/

structures. The applica shield building:

shield buildin x

the steel roof that sup e steel roof that sup x

tension ring ion rin x

SC/RC connectio C co The design process proce RC components RC components Standard for s Standard for s approach.

approach.

ed ed

[DRAFT 4/25/18]

Page 6 of 6 However, SNC acknowledges that ACI 349 and ANSI/AISC N690 are not exclusively applicable to the shield building SC wall modules, including connections to RC. For example, there is significant design requirement information beyond code requirements in UFSAR Subsection 3.8.7, Reference 57, APP-GW-GLR-602, Revision 5 (Proprietary) and APP-GW-GLR-603, Revision 5 (Non-Proprietary), "AP1000 Shield Building Design Details for Select Wall and RC/SC Connections" [ADAMS Accession No. ML110910541]. Furthermore, there are additional UFSAR sections that provide supplemental design requirements beyond code requirements that SNC acknowledges are important to the design of the shield building. For example, UFSAR Subsection 3.8.4.5.5, Shield Building Structural Wall Modules, states that design requirements for shield building concrete-filled structural wall modules are addressed in UFSAR referenced codes and supplemental requirements not addressed in codes:

[Concrete-filled structural wall modules are designed as reinforced concrete structures in accordance with the requirements of ACI 349, and supplemented with additional requirements discussed in subsection 3.8.3.5.3 and below]*

[Note that UFSAR Subsection 3.8.3.5.3 is Tier 2 text.] Within UFSAR Subsection 3.8.4.5.5, supplemental Tier 2* design requirements for the shield building are addressed in Subsection 3.8.4.5.5.5, Design of Shear Studs and Tie Bars.

By including these additional shield building requirements within the scope of proposed evaluation Criterion 1, combined with the 10 CFR Part 52, Appendix D, Section VIII.B.5.b evaluation criteria, the new evaluation process will provide a reasonable assurance of safety.

Proposed Change to New Evaluation Criterion 1 To address the concerns addressed by the NRC staff in the question, SNC agrees to broaden the scope of proposed evaluation Criterion 1 to address additional requirements. Proposed Criterion 1 is revised to address RC design, SC module design, and shield building design as follows:

1. Involve design methodology or construction materials that deviate from a code or standard credited in the plant-specific DCD for establishing the criteria for the design or construction of a structure, system, or component (SSC) important to safety, deviate from the design methodology described in UFSAR Subsection 3.8.7, Reference 57, or deviate from the supplemental design requirements described in UFSAR Subsections 3.8.4.5.1, 3.8.4.5.2, 3.8.4.5.5.5, or 3H.3.4.

DRAFT R

Wall an Wall a here are add here are ode requirements th ode requ r example, r example UFSAR hat hat design requirements design requ sed in UFSAR referenced sed in UFSAR refer ed as reinforced concret ed as re e structures ructures and supplemented with additional and sup and below and be

]*

ext.] ] Within UFSAR Subsection 3.8.4.5.5, Within UFSAR Subs he shield building building are addressed in Subsect are add Bars.

ding requ ding req irements within the scope of n the sc propo the 10 CFR Part 52, Appendix D, Section V he 10 CFR Part 52, Appendix D, S tion process will provide rocess will provide a reasonable assu a reasonable ass valuation Criterion riteri 1

addressed by the NRC staff addressed by the NRC staff in the question in the question evaluat evaluation Criterion 1 to address ion Criterion 1 to addre additiona d

d to address to address RC RC design, d

SC module design C mod

. Involve design methodolo Involve design methodology or construct gy or construc standard credited in the plant standard credited in the plant-specific specific design or construction of a structure esign or construction of a structure safetyty, deviate from the design m

, deviate DR 3.8.7, Reference 57 3.8.7, Referen

, or deviat DR described in UFSAR Subse described in UFS DR

Page 1 of 2

[DRAFT 4/25/18]

Request for Additional Information LAR 17-037-3 Issue Date: 04/12/2018 Application

Title:

Vogtle Nuclear Site, Units 3 and 4, LAR 17-037 Operating Company: Southern Nuclear Operating Co.

Docket No. 52-0025 and 52-0026 Review/Application Section: Not Applicable By letter dated December 21, 2017, Southern Nuclear Operating Company, Inc. (SNC),

submitted License Amendment Request (LAR) No.17-037 to the U. S. Nuclear Regulatory Commission (NRC) for Vogtle Electric Generating Plant Units 3 and 4, Combined License Nos. NPF-91 and NPF-92 (ADAMS Accession No. ML17355A416). The LAR requests NRC approval of a proposed license condition that would allow departures from Tier 2* information without a license amendment by allowing the licensee to evaluate a prospective departure against criteria stated in the proposed license condition.

Enclosure (5) of the referenced LAR provides a summary of an analysis of Tier 2* matters using the proposed license condition (screening criteria) presented in enclosure (3) of your letter dated December 21, 2017. For Section VIII.B.6.b (Tier 2* Matters that Do Not Expire at Full Power), Item 5, the enclosure states that the reactor coolant pump type is adequately addressed in Tier 1, and therefore no additional screening criteria is required. Also, note that since the enclosure only states that it is adequately addressed by Tier 1, paragraph VIII.B.5 would not be needed since a change in Tier 1 information requires NRC approval. Therefore, per the enclosure, no additional screening criteria is needed since the information is in Tier 1.

The NRC staff notes that the analysis provided in Enclosure (5) of the SNC submittal is not correct, since Tier 1 does not specify the type of RCP; but only specifies "sealless reactor coolant pumps." "Sealless reactor coolant pumps" is a generic term that only states that the pump does not have seals, which makes seal failure not a concern. Using the proposed license condition, a different sealless pump type could be used because the pump still meets the sealless (shaft seal failure) requirement that is specified as Tier 1. The Tier 2* information is canned motor type RCP, which is not addressed in Tier 1. Therefore, Enclosure (5) is not correct since Tier 1 does not adequately address all the essential attributes of the type of pump (i.e., canned motor, wet winding, etc.). However, the essential attributes of the type of pump are included as Tier 2* information which states "centrifugal sealless pump of canned motor design".

Therefore, since all of the essential attributes of the RCP (i.e., that it be of a canned motor design) are not addressed in Tier 1, the staff requests that the licensee revise the proposed License Condition 2.d (13)(a) in Enclosure (3) to address this essential attribute by adding the following, or explain why it is not necessary:

"5. Results in a change to the RCP type (canned motor design)."

SNC Response to RAI Question LAR 17-037-3 As discussed in Section 3 of Enclosure 1 of this LAR, SNC evaluated existing Tier 2* material for safety significance. Tier 2* material that was deemed to be of safety significance not warranted to be included in Tier 1 was not selected for development of a non-qualifying criteria. With regard to Reactor Coolant Pumps (RCPs), SNC reviewed relative information regarding the safety significant attributes of the RCP design. Information reviewed included Vogtle Electric Generating Plant (VEGP) licensing documents including the VEGP 3 and 4 Tier 1 Plant-Specific Design Control Document and the VEGP 3 and 4 DRAFT mpany, Inc. (SNC),

mpany, Inc. (SNC U. S. Nuclear Regulatory S. Nuclear Regulato and 4, Co and 4, C mbined License icense A416).

A416). The LAR requests NRC The LAR requests NRC epartures from Tier 2* information epartures from Tier 2* information evaluate evalua a prospective departure

n.

n.

summary of an analysis of Tier 2* matters u mary of an analysis o eria) presented in enclosure (

nted in enclosu

3) of your lette II.B.6.b (Tier 2* Matters that Do Not Expire a Matters t t the reactor coolant pump type is adequate t the reactor coolant pump ty additional screenin additional sc g criteria is required.

a is requ Als it is adequately addressed by Tier 1, parag adequately addressed by Tier 1, pa nge in Tier 1 information requires NRC app Tier 1 information requires NRC app al screening criteria is needed since the info g criteria is needed since the info e analysis provided in Enclosure (5) of the S ided in Enclosure (5) of the S s not specify the type of RCP; but only spec not specify the type of RCP; but only spec less reactor coolant pumps" is a generic ter less reactor coolant pumps" is e seals, e seals, which makes which ma seal failure not a conc failur ent sealless pump type could be used beca ent sealless pump type could be us ft seal failure ft seal f

) requirement that is specified uirement that is specif tor type RCP, whi or type ch is not addressed in Tie ch is not addressed in T nce Tier 1 does not adequately Tier 1 does not adequately address all address all anned motor, wet winding, etc.).

tor, wet winding, etc.). However, t However, t uded as Tier 2* information which states "ce informatio esign".

esign".

Therefore, since all of the essential attri Therefore, since all of the essentia design) are not addressed in Tier 1, design) are not addressed in Tier 1, t License Condition 2.d (13)(a) in En 2.d (13)(a) in En following fo

, or explain why it is no why "5. Results in a c ts in a Response to RA Response to RA D

sed in sed in

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[DRAFT 4/25/18]

Updated Final Safety Analysis Report (UFSAR) which contains the plant-specific Tier 2 Design Control Document (DCD). Also reviewed were Final Safety Evaluation Reports for the AP1000 and VEGP 3 and 4. Finally, transcripts and letters from the Advisory Committee on Reactor Safeguards (ACRS) were reviewed. Based on these reviews, SNC determined that the principal safety feature of the RCP design is the sealless feature. As a result, this design feature is captured in Tier 1 information. In addition, Tier 1 information identifies RCP safety-related safety functions as 1) rotating inertia to provide Reactor Coolant System (RCS) flow coastdown on loss of power to the pumps and 2) that each RCP flywheel assembly can withstand a design overspeed condition. Details of these key safety design features including the flywheel, the flywheel retaining ring, and the flywheel enclosure which acts as a missile barrier, are described in Tier 2 information of the UFSAR. The SNC review did not identify the canned motor design feature as being a principal safety design feature. Thus, the existing Tier 2* material regarding the canned motor feature was not selected for the development of a non-qualifying criterion.

SNC also reviewed what the impact of a potential change to Tier 2* information would be with regard to existing change processes and regulations. Should a fundamental change in the RCP be proposed that would involve the change from a canned motor design, the change would be evaluated against the criteria of 10 CFR Part 52, Appendix D, paragraph VIII.B.5. This evaluation would address impacts to the Chapter 15 safety analyses. SNC concluded that a postulated change to use a RCP other than a canned motor design would most likely result in a conclusion that prior NRC approval would be required for the change.

Therefore, SNC concludes that no new evaluation Criterion for License Condition 2.d (13)(a) is necessary to address the canned-motor feature of the RCP.

Based on the above discussion, change the RCP entry for Enclosure 5 of the original LAR-17-037to read:

5 Reactor coolant pump type.

No Adequately addressed in Tier 1 and paragraph VIII.B.5 N/A DRAFT n add n add provide Re provide ach RCP flywhee ach RCP y design features y design hich acts as a missile hich acts as did not did not identify identify the canned th e existing Tier 2* material existing Tier 2* mater ent of ent of a non-qualifying criterion.

ng criter

• information would be

• inform with regard to egard to ntal change in the RCP be proposed that ntal change in the RCP be proposed that hange woul hange w d be evaluated against the

. This evaluation would

. This evaluation would address a

impacts to the he stulated ulated change to use change to use a RCP other than a RC a cann that prior NRC approval would be required for t rior NRC approval wo ation Criterion for License Condition 2.d (13)(a) nse Con RCP.

RCP.

nge the RCP entry for Enclosure 5 e RCP entry for Enclos of the origina he origin pe.

No No Adequately addr Adequately add Tier 1 T

and par and par A

VIII.B.5 A