ND-18-1006, Supplement to Request for License Amendment and Exemption: Changes to Tier 2* Departure Evaluation Process (LAR-17-037S4)

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Supplement to Request for License Amendment and Exemption: Changes to Tier 2* Departure Evaluation Process (LAR-17-037S4)
ML18215A461
Person / Time
Site: Vogtle  Southern Nuclear icon.png
Issue date: 08/03/2018
From: Whitley B
Southern Nuclear Operating Co
To:
Document Control Desk, Office of New Reactors
References
LAR-17-037S4, ND-18-1006
Download: ML18215A461 (84)
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Text

Brian H. Whitley Southern Nuclear Director, Regulatory Affairs Operating Company, Inc.

3535 Colonnade Parkway Birmingham, AL 35243 Tel 205.992.7079 August 3, 2018 Docket Nos.: 52-025 ND-18-1006 52-026 10 CFR 50.90 10 CFR 52.7 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Southern Nuclear Operating Company Vogtle Electric Generating Plant Units 3 and 4 Supplement to Request for License Amendment and Exemption:

Changes to Tier 2* Departure Evaluation Process (LAR-17-037S4)

Ladies and Gentlemen:

Pursuant to 10 CFR 52.98(c) and in accordance with 10 CFR 50.90, Southern Nuclear Operating Company (SNC) requested an amendment to the combined licenses (COLs) for Vogtle Electric Generating Plant (VEGP) Units 3 and 4 (License Numbers NPF-91 and NPF-92, respectively) by SNC letter ND-17-1726, dated December 21, 2017 [ADAMS Accession Number ML17355A416],

to apply the existing departure evaluation process for Tier 2 departures to the evaluation of certain Tier 2* departures. This license amendment request (LAR)17-037 was supplemented by SNC letters ND-18-0417 (LAR-17-037S1), dated April 6, 2018 [ML18096B328], ND-18-0608 (LAR-17-037S2), dated May 11, 2018 [ML18131A263], and ND-18-0646 (LAR-17-037S3), dated June 18, 2018 [ML18169A431]. By letter ND-17-1726, SNC also requested an exemption from certain change requirements in 10 CFR Part 52, Appendix D, consistent with the requested license amendment.

This supplement (LAR-17-037S4) provides information in response to a Request for Additional Information (RAI), identified as RAI LAR-17-037-2, from the Structural Engineering Branch (SEB) by electronic mail (email) dated April 12, 2018 [ML18102B682]. The response to RAI LAR-17-037-2 considers feedback provided by the NRC staff during multiple public meetings held between May and July 2018. This supplement also provides the response to a question provided by the NRC staff in an email dated June 29, 2018 [ML18180A374] regarding Updated Final Safety Analysis Report (UFSAR) Chapter 14 Tier 2* text addressing first-of-a-kind testing.

Enclosures 1 through 8 were provided with the original LAR-17-037, in SNC letter ND-17-1726. provided a response to RAI LAR-17-037-1 as the first supplement to LAR-17-037, in SNC letter ND-18-0417. Enclosures 10, 11, and 12 provided a revised response to RAI LAR-17-037-1, and the initial responses to RAIs LAR-17-037-3 and RAI LAR-17-037-4, respectively, in SNC letter ND-18-0608. Enclosures 13, 14, 15, 16, and 17 provided the responses to RAIs LAR-17-037-5, -6, -7, -8, and -9, respectively, in SNC letter ND-18-0646. 8, included in this letter, provides the response to RAI LAR-17-037-2. Enclosure 19, provides the response to the question from the NRC staff regarding Tier 2* text addressing first-of-a-kind testing.

U.S. Nuclear Regulatory Commission ND-18-1006 Page 2 of 6 Additionally, Enclosures 1U through 8U in this letter provide updates to the eight enclosures originally provided with the original LAR-17-037 in SNC letter ND-17-1726. The text in these updated enclosures provides the changes that were required by the supplemental information provided in the RAI responses included in Supplements 2 and 3, as well as the supplemental information provided in Enclosures 18 and 19 of this letter. The supplemental information provided in each of these eight updated (U) enclosures is identified with the use of a vertical change bar in the right-hand margin adjacent to the text that differs from that provided in the original eight enclosures (i.e., Enclosures 1 through 8) in letter ND-17-1726. (Note that the suffix U indicates the original enclosure is updated.)

The final proposed license conditions (Enclosure 3U) also reflect other editorial and non-technical text changes from the text proposed in the original LAR-17-037 and supplements that were provided by NRC staff during the various public interactions referred to previously. The exemption request has been updated (Enclosure 2U) to reflect changes that were made to the text in the final proposed license conditions.

The information provided in this LAR supplement does not impact the scope, technical content, or conclusions of the Significant Hazards Consideration Determination, or the Environmental Considerations of the original LAR-17-037 provided in Enclosure 1 of SNC letter ND-17-1726.

This letter has been reviewed and confirmed to not contain security-related information. This letter contains one new regulatory commitment that is reflected in Enclosure 8U. The commitment that was proposed in the original LAR-17-037, Enclosure 8, that required SNC to annotate Tier 2*

departures implemented without submitting a license amendment request within departure reports submitted in accordance with 10 CFR Part 52, Appendix D, paragraphs X.B.1 and X.B.3.b, has been removed because this obligation will be included as a new license condition 2.D.(13)(b).2 (refer to Enclosure 3U).

SNC now requests NRC Staff review and approval of the license amendment and exemption by September 7, 2018, and would implement the amendment within 45 days of issuance. This license amendment and exemption is not tied to any particular construction activity; however, any delay in the issuance would also delay the benefits.

In accordance with 10 CFR 50.91, SNC is notifying the State of Georgia of this LAR supplement by transmitting a copy of this letter and its enclosure to the designated State Official.

Should you have any questions, please contact Wesley Sparkman at (205) 992-5061.

U.S. Nuclear Regulatory Commission ND-18-1006 Page 3of6 I declare under penalty of perjury that the foregoing is true and correct. Executed on the 3rd of August 2018.

Respectfully submitted, Brian H. Whitley Director, Regulatory Affairs Southern Nuclear Operating Company Enclosures 1 - 8) (Previously submitted with the original LAR, LAR-17-037, in letter ND-17-1726) 1U) Vogtle Electric Generating Plant (VEGP) Units 3 and 4 - Updated Request for License Amendment: Changes to Tier 2* Departure Evaluation Process (LAR-17-037S4) 2U) Vogtle Electric Generating Plant (VEGP) Units 3 and 4 - Updated Exemption Request: Changes to Tier 2* Departure Evaluation Process (LAR-17-037S4) 3U) Vogtle Electric Generating Plant (VEGP) Units 3 and 4 - Updated Proposed Changes to Licensing Basis Documents (LAR-17-037S4) 4U) Vogtle Electric Generating Plant (VEGP) Units 3 and 4 - Updated Reviewer's Aid - Proposed Tier 2* Departure Evaluation Process (LAR-17-037S4)

SU) Vogtle Electric Generating Plant (VEGP) Units 3 and 4 - Updated Reviewer's Aid -Tier 2* Matters Analysis Summary (LAR-17-037S4) 6U) Vogtle Electric Generating Plant (VEGP) Units 3 and 4 - Updated Reviewer's Aid - Tier 2* Departure Example Not Requiring Prior NRC Approval (LAR-17-037S4) 7U) Vogtle Electric Generating Plant (VEGP) Units 3 and 4 - Updated Reviewer's Aid - Tier 2* Departure Examples Requiring Prior NRC Approval (LAR-17-037S4)

BU) Vogtle Electric Generating Plant (VEGP) Units 3 and 4- Updated Proposed Regulatory Commitment (LAR-17-037S4)

9) (Previously submitted with LAR-17-037S1, in letter ND-18-0417)

U.S. Nuclear Regulatory Commission ND-18-1006 Page 4 of 6 10 - 12) (Previously submitted with LAR-17-037S2, in letter ND-18-0608) 13 - 17) (Previously submitted with LAR-17-037S3, in letter ND-18-0646)

18) Vogtle Electric Generating Plant (VEGP) Units 3 and 4 - Response to NRC Request for Additional Information (RAI) LAR-17-037 Regarding the LAR-17-037 Review (LAR-17-037S4)
19) Vogtle Electric Generating Plant (VEGP) Units 3 and 4 - Response to NRC Question on Evaluation of Departures from Tier 2* Information Regarding First-of-a-Kind Testing (LAR-17-037S4)

U.S. Nuclear Regulatory Commission ND-18-1006 Page 5 of 6 cc:

Southern Nuclear Operating Company / Georgia Power Company Mr. S. E. Kuczynski (w/o enclosures)

Mr. D. G. Bost (w/o enclosures)

Mr. M. D. Meier (w/o enclosures)

Mr. D. H. Jones (w/o enclosures)

Mr. J. B. Klecha Mr. G. Chick Mr. D. L. McKinney (w/o enclosures)

Mr. T. W. Yelverton (w/o enclosures)

Mr. B. H. Whitley Ms. C. A. Gayheart Mr. C. R. Pierce Ms. A. G. Aughtman Mr. D. L. Fulton Mr. M. J. Yox Mr. E. W. Rasmussen Mr. J. Tupik Mr. W. A. Sparkman Ms. A. C. Chamberlain Ms. A. L. Pugh Document Services RTYPE: VND.LI.L00 File AR.01.02.06 Nuclear Regulatory Commission Mr. W. Jones (w/o enclosures)

Ms. J. Dixon-Herrity Mr. C. Patel Ms. J. M. Heisserer Mr. B. Kemker Mr. G. Khouri Ms. S. Temple Mr. F. Brown Mr. T.E. Chandler Ms. P. Braxton Mr. T. Brimfield Mr. C. J. Even Mr. A. Lerch State of Georgia Mr. R. Dunn

U.S. Nuclear Regulatory Commission ND-18-1006 Page 6 of 6 Oglethorpe Power Corporation Mr. M. W. Price Ms. A. Whaley Municipal Electric Authority of Georgia Mr. J. E. Fuller Mr. S. M. Jackson Dalton Utilities Mr. T. Bundros Westinghouse Electric Company, LLC Mr. L. Oriani (w/o enclosures)

Mr. C. Churchman (w/o enclosures)

Mr. M. Corletti Mr. M. L. Clyde Ms. L. Iller Mr. D. Hawkins Mr. J. Coward Other Mr. S. W. Kline, Bechtel Power Corporation Ms. L. A. Matis, Tetra Tech NUS, Inc.

Dr. W. R. Jacobs, Jr., Ph.D., GDS Associates, Inc.

Mr. S. Roetger, Georgia Public Service Commission Ms. S. W. Kernizan, Georgia Public Service Commission Mr. K. C. Greene, Troutman Sanders Mr. S. Blanton, Balch Bingham Mr. R. Grumbir, APOG NDDocumentinBox@duke-energy.com, Duke Energy Mr. S. Franzone, Florida Power & Light

Southern Nuclear Operating Company ND-18-1006 Enclosure 1U Vogtle Electric Generating Plant (VEGP) Units 3 and 4 Updated Request for License Amendment:

Changes to Tier 2* Departure Evaluation Process (LAR-17-037S4)

(This Enclosure consists of 28 pages, including this cover page.)

ND-18-1006 U Updated Request for License Amendment: Changes to Tier 2* Departure Evaluation Process (LAR-17-037S4)

Table of Contents

1.

SUMMARY

DESCRIPTION

2. DETAILED DESCRIPTION
3. TECHNICAL EVALUATION
4. REGULATORY EVALUATION 4.1. Applicable Regulatory Requirements/Criteria 4.2. Precedent 4.3. Significant Hazards Consideration 4.4. Conclusions
5. ENVIRONMENTAL CONSIDERATIONS
6. REFERENCES Page 2 of 28

ND-18-1006 U Updated Request for License Amendment: Changes to Tier 2* Departure Evaluation Process (LAR-17-037S4)

Pursuant to 10 CFR 52.98(c) and in accordance with 10 CFR 50.90, Southern Nuclear Operating Company (SNC) hereby requests an amendment to Combined License (COL) Nos. NPF-91 and NPF-92 for Vogtle Electric Generating Plant (VEGP) Units 3 and 4, respectively.

1.

SUMMARY

DESCRIPTION Under the current departure evaluation process applicable to Tier 2* information described in 10 CFR Part 52, Appendix D, Paragraph VIII.B, SNC must seek prior NRC approval through a License Amendment Request (LAR) for any proposed change to Tier 2* information, even if SNC can demonstrate that the change results in no more than a minimal impact to safety or improves safety. As the NRC staff has recently recognized in SECY-17-0075, Planned Improvements in Design Certification Tiered Information Designations, [ADAMS Accession Number ML16196A321], One specific lesson is that some information has been designated as Tier 2*

when other regulatory tools could have been used instead to ensure a facility is safely designed, constructed and operated. This results in licensees submitting license amendment requests (LARs) on topics that may not involve safety significant facility changes. This is consistent with SNCs experience with the Tier 2* departure evaluation process. In order to mitigate the regulatory inefficiency associated with this issue, SNC proposes a site-specific permanent exemption and license amendment that would use new screening criteria to determine whether a proposed Tier 2* departure would qualify to utilize the Tier 2 departure evaluation process.

Qualifying Tier 2* departures would be evaluated under the existing Tier 2 departure evaluation process. Non-qualifying Tier 2* departures would continue to require prior NRC approval. Thus, any safety-significant Tier 2* departure would require prior NRC approval. A diagram of the proposed process is shown in Enclosure 4U.

2. DETAILED DESCRIPTION The NRC issued the first Part 52 licenses to SNC VEGP Units 3 and 4 in February 2012. Changes to the licensing bases for those licenses are governed, in part, by 10 CFR Part 52, Appendix D, Paragraph VIII.B. This portion of the regulations specifies the change process for Tier 2 information and Tier 2* information and requires NRC approval for all departures from Tier 2*

information.

However, recent Design Certification applications do not contain Tier 2* information, in part because the level of detail contained in Tier 1 information will encompass information that might be designated as Tier 2*, and the existing Tier 2 change process requires prior NRC approval of safety-significant departures. SNC is proposing changes to the VEGP Units 3 and 4 licensing bases regarding Tier 2* change processes to make them functionally similar to the processes currently under development between Korea Hydro and Nuclear Power (KHNP) and the NRC as part of the Advanced Power Reactor 1400 (APR1400) design certification application.

SNC acknowledges that the Commission employed a Tier 2* designation to capture certain significant AP1000 design information existing in Tier 2 that the Commission did not want changed without prior approval (see 71 Fed. Reg. 4474 (Jan. 27, 2006)). In SECY-17-0075, the NRC discussed the reasons for designating some Tier 2 information as Tier 2* and indicated Page 3 of 28

ND-18-1006 U Updated Request for License Amendment: Changes to Tier 2* Departure Evaluation Process (LAR-17-037S4) that Tier 2* information is intended to have substantial safety significance, commensurate with information designated as Tier 1. However, SECY-17-0075 suggests that 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; e.g., background information and other information of minimal safety significance. Furthermore, SNCs experience has demonstrated that not every change to information designated as Tier 2* has an impact on the safety-significant nature, if any, of the information. As such, SNC proposes to invoke a process functionally consistent with departure evaluation processes applied by current applicants for the certification of designs that contain no Tier 2* information but have significant safety-related information contained in the Tier 1 design control document (DCD). Specifically, SNC proposes a site-specific amendment that would allow qualifying departures from Tier 2* information to be evaluated under the existing departure evaluation process for Tier 2 departures in 10 CFR Part 52, Appendix D, Paragraphs VIII.B.5.a through VIII.B.5.e. Qualifying departures from Tier 2*

information would be determined by applying screening criteria to proposed departures from Tier 2* information. Departures from Tier 2* information that involve safety significance commensurate with Tier 1 information would be non-qualifying Tier 2* departures and would continue to require prior NRC review and approval in accordance with 10 CFR Part 52, Appendix D, Paragraph VIII.B.6. Qualifying Tier 2* departures would be evaluated under the existing Tier 2 departure evaluation process. Thus, any safety-significant Tier 2* departures would require prior NRC approval.

Consistent with the NRCs findings in SECY-17-0075, SNC has identified several examples of departures from Tier 2* information that were not safety-significant, but nonetheless required prior NRC approval through a LAR. Application of the Tier 2 departure evaluation process to these proposed departures would have concluded with a determination that the proposed change was not safety-significant and could therefore have been processed as a departure consistent with 10 CFR Part 52, Appendix D, Paragraph VIII.B.5.

  • A figure in SNCs licensing basis included a Note specifying the design basis size and spacing of shear studs in the structural modules. However, a change to the Note was needed for consistency with design basis calculations that were previously revised and incorporated into the AP1000 generic DCD. To resolve this inconsistency, the figure needed to be changed to make the Note consistent with the design basis and clarify that spacing may be changed to satisfy the applicable codes and standards. The change had the effect of enhancing safety by reflecting the design philosophy of adherence with the specific codes and standards invoked by the licensing basis. Nevertheless, because the Note was designated as Tier 2*, prior NRC approval was required.1
  • During construction, it was discovered that the tolerances for basemat thickness would potentially not ensure a level floor. The positive tolerance needed to be expanded to improve the probability of a level surface on which to construct the Nuclear Island structures. An engineering evaluation demonstrated that the change in tolerance was 1 SNC letter ND-12-0101, Request for License Amendment: Containment Internal Structural Module Shear Stud Size and Spacing (LAR-12-001), dated February 14, 2012 [ADAMS Accession No. ML12047A067]

and SNC letter ND-12-1399, Revised Request for License Amendment: Structural Modules Shear Stud Size and Spacing (LAR-12-001S), dated March 12, 2012 [ADAMS Accession No. ML12074A180].

Page 4 of 28

ND-18-1006 U Updated Request for License Amendment: Changes to Tier 2* Departure Evaluation Process (LAR-17-037S4) within the code allowance and the strength of the basemat would be maintained; however, because the tolerance was designated as Tier 2*, prior NRC approval was required.2

  • In a document incorporated by reference (IBRd) into the Updated Final Safety Analysis Report (UFSAR) and designated as Tier 2*, SNC had to obtain prior NRC approval to make a clarification that the phrase ISV Facility included identical facilities located both at Westinghouse and Vogtle, rather than just Westinghouse. Another change to this Tier 2* document requiring prior NRC permission was needed to add two questions to a survey given to students after simulator drills.3 Because the IBRd document is designated as Tier 2*, prior NRC approval was required.
  • Several editorial changes, such as typing, clerical, spelling, and consistency changes, were required to Tier 2* information to achieve consistency throughout the licensing basis.

These changes affected nothing in the physical layout of the plant nor in the design function of the plant. Safety is enhanced by these kinds of changes because electronic searches of the licensing basis become more accurate. For example, a typographical inconsistency in an acronym would impede an electronic search for that acronym, as it would not yield the portion of the licensing basis containing the inconsistency; editorial changes were needed to resolve this issue.4 These examples demonstrate that although Tier 2* information was intended to have substantial safety significance, commensurate with information designated as Tier 1, some Tier 2*

departures are not, in fact, safety-significant. This license amendment request would allow SNC to apply screening criteria to departures from Tier 2* information to determine whether such departures qualify to be evaluated under the Tier 2 departure evaluation process.

A Tier 2* departure would qualify to be evaluated under the Tier 2 departure evaluation process unless the proposed departure would:

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,
2. Result in a change to a design process described in the plant-specific DCD that is material to implementation of an industry standard or endorsed regulatory guidance, 2 SNC letter ND-12-0670, Request for License Amendment: Nuclear Island Basemat Thickness Tolerance (LAR-12-003), dated April 6, 2012 [ADAMS Accession No. ML12100A185], as supplemented by SNC letter ND-12-0809, Request for License Amendment - Supplemental Information: Nuclear Island Basemat Thickness Tolerance (LAR-12-003), dated April 12, 2012 [ADAMS Accession No. ML12104A323], and revised by SNC letter ND-12-0990, Request for License Amendment: Nuclear Island Basemat Thickness Tolerance (LAR-12-003R) Revised, dated May 7, 2012 [ADAMS Accession No. ML12130A468].

3 SNC letter ND-13-0348, Request for License Amendment: Revision to AP1000 Human Factors Engineering Integrated System Validation Plan I GEH-320 (LAR-13-001), dated February 15, 2013

[ADAMS Accession No. ML13050A214].

4 SNC letter ND-14-1045, Request for License Amendment: Tier 2* Editorial and Consistency Changes (LAR-13-033), dated July 30, 2014 [ADAMS Accession No. ML14211A666].

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ND-18-1006 U Updated Request for License Amendment: Changes to Tier 2* Departure Evaluation Process (LAR-17-037S4)

3. (i) Result in a change to the fuel criteria evaluation process, the fuel principal design requirements, or the nuclear design of the fuel and the reactivity control system that is material to a fuel or reactivity control system design function, or the evaluation methods in WCAP-12488, Westinghouse Fuel Criteria Evaluation Process, or (ii) Result in any change to the maximum fuel rod average burn-up limits or the small break LOCA analysis methodology described in UFSAR Subsections 15.6.5.4B.2.2 or 15.6.5.4B.2.3;
4. Adversely affect the containment debris limits or debris screen design criteria,
5. Change the Reactor Coolant Pump (RCP) type from a canned motor to a different type of RCP,
6. Result in a change to the Passive Residual Heat Removal Heat Exchanger natural circulation test (first plant test), the Core Makeup Tank Heated Recirculation Tests (first three plants test), or the Automatic Depressurization System Blowdown Test (first three plants test) that is material to the test objectives or test performance criteria,
7. Involve structural materials or analytical or design methods, including design codes and analytical assumptions, that deviate from those credited in the plant-specific DCD for critical sections,
8. Result in a change to the design of the steel faceplates, internal trusses, tie bars, or headed studs of the steel-concrete (SC) module walls in the Nuclear Island or the Shield Building, including SC-to-reinforced concrete (RC) connections,
9. Result in an increase in the demand to capacity (D/C) ratio of a critical section of the structure. SNC shall determine the D/C ratio under this condition for each critical section structural member including, but not limited to, wall segments, wall sections, concrete panels, slabs, or basemat sections, affected by a departure by:

(i) Using the Tier 2* information in the UFSAR Section 3.8 or Appendix 3H table that directly states the D/C ratio or states the area of steel provided and the area of steel required for the affected structural member, or (ii) Providing the same total area of steel across the entire critical section using any combination of rebar sizes and spacing allowed by the design basis codes used in the UFSAR as the total area of steel specified in UFSAR Section 3.8 and Appendix 3H tables marked Tier 2*.

If the screening criteria are all answered no, the proposed change would be considered a Qualifying Change and would be processed in accordance with the Tier 2 departure evaluation process specified in 10 CFR Part 52, Appendix D, Paragraphs VIII.B.5.a through VIII.B.5.e (See U for a proposed process diagram).

To provide the NRC the opportunity to provide additional oversight of the proposed revision to the Tier 2* departure process, SNC is proposing a new license condition [2.D.(13)(b).2] that requires Page 6 of 28

ND-18-1006 U Updated Request for License Amendment: Changes to Tier 2* Departure Evaluation Process (LAR-17-037S4)

SNC to prepare and maintain a written evaluation that provides the bases for its determinations regarding the criteria in License Condition 2.D.(13)(a). In the report that 10 CFR Part 52, Appendix D, Section X.B.1 requires SNC to submit, SNC shall include a brief description of each departure and a summary of the evaluation of the departure. This proposed requirement would be implemented coincident with the implementation of the license amendment approving this LAR, and would be applicable to Tier 2* departures identified in departure reports submitted subsequent to the implementation of this license amendment.

To ensure the proposed qualifying Criteria reliably and predictably differentiate between Tier 2*

information with safety significance commensurate with Tier 1 and other information that does not warrant the same level of control, SNC is proposing a regulatory commitment that would require SNC to develop, implement, and maintain procedural guidance with a level of detail commensurate with the detailed implementation guidance and related bases for the proposed Criteria contained in this LAR, including additional guidance provided by SNC in the supplements to this LAR. The proposed regulatory commitment would be implemented prior to the implementation of the license amendment approving this LAR. The proposed regulatory commitment is shown in Enclosure 8U.

Licensing Basis Change Descriptions:

Proposed Licensing Basis Changes COL License Condition Description of the Proposed Change 2.D.(13) Adds new license condition 2.D.(13) to document that the licensee is exempt from the requirements of 10 CFR Part 52, Appendix D, Paragraphs VIII.B.5.a and VIII.B.6 subject to the conditions and limitations set forth in Section 2.D.(13) of this license and to specify the plant-specific licensing requirements for the Tier 2*

departure evaluation process. The elements of this process are provided in sub-paragraphs (a) and (b).

2.D.(13)(a) Adds a new license condition sub-paragraph that defines the Tier 2* departure regulations from which SNC is exempt except when any of nine screening criteria are met.

2.D.(13)(b) Adds a new license condition sub-paragraph that allows Tier 2* departures to be evaluated under the provisions of 10 CFR Part 52, Appendix D, Section VIII.B.5 provided the conditions of the license condition are met, that requires the departure or change to Tier 2* information to remain Tier 2* information in the plant-specific DCD, and requires that SNC prepare and maintain a written evaluation that provides the bases for its determinations regarding the criteria in License Condition 2.D.(13)(a). In Page 7 of 28

ND-18-1006 U Updated Request for License Amendment: Changes to Tier 2* Departure Evaluation Process (LAR-17-037S4) the report that 10 CFR Part 52, Appendix D, Section X.B.1 requires SNC to submit, SNC shall include a brief description of each departure and a summary of the evaluation of the departure.

UFSAR pages with a The footer is modified to stipulate that prior NRC approval footer regarding Tier 2* of departures from Tier 2* information may be required in information accordance with the departure evaluation process specified in License Condition 2.D.(13).

3. TECHNICAL EVALUATION The NRC issued the first Part 52 licenses to VEGP Units 3 and 4 in February 2012. Changes to the licensing bases for those licenses are governed, in part, by 10 CFR Part 52, Appendix D, Paragraph VIII.B. This portion of the regulations specifies the departure and change process for Tier 2 information and Tier 2* information and requires NRC approval for all departures from Tier 2* information.

SNC acknowledges that the Commission employed a Tier 2* designation to capture certain significant AP1000 design information existing in Tier 2 that the Commission determined should not be changed without prior approval (see 71 Fed. Reg. 4474 (Jan. 27, 2006)). In SECY-17-0075, the NRC discussed the reasons for designating some Tier 2 information as Tier 2* and indicated that Tier 2* information is intended to have substantial safety significance, commensurate with information designated as Tier 1. However, SECY-17-0075 suggests that 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; e.g., background information and other information of minimal safety significance. Furthermore, SNCs experience has demonstrated that not every change to information designated as Tier 2*

has an impact on the safety-significant nature, if any, of the information. As such, SNC proposes to invoke a process functionally consistent with departure evaluation processes applied by current applicants for the certification of designs that contain no Tier 2* information. Specifically, SNC proposes a site-specific amendment that would use new screening criteria to determine whether a proposed Tier 2* departure would qualify to utilize the Tier 2 departure evaluation process.

Qualifying Tier 2* departures would be evaluated under the existing Tier 2 departure evaluation process specified in 10 CFR Part 52, Appendix D, Paragraphs VIII.B.5.a through VIII.B.5.e. Non-qualifying Tier 2* departures would continue to require prior NRC approval in accordance with 10 CFR Part 52, Appendix D, Paragraph VIII.B.6. Thus, any safety-significant Tier 2* departure would require prior NRC approval.

SECY-17-0075 provides the historical basis and origins for designating information as Tier 2*.

Citing the history of the development of Tier 2*, SECY-17-0075 explains that Tier 2* was intended to have the same safety significance as Tier 1 information. In addition, SECY-17-0075 references SECY-96-0775 which also provides insight to the origins and requirements of Tier 2* information.

Specifically, SECY-96-077 states, Also, many codes, standards, and design processes, which 5 SECY-96-077, Certification of Two Evolutionary Designs, April 15, 1996 (ADAMS Accession No. ML003708129)

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ND-18-1006 U Updated Request for License Amendment: Changes to Tier 2* Departure Evaluation Process (LAR-17-037S4) were not specified in Tier 1, that are acceptable for meeting [inspections, tests, analyses, and acceptance criteria] ITAAC were specified in Tier 2. The result of these actions is that certain significant information only exists in Tier 2 and the NRC does not want this significant information to be changed without prior NRC approval. To address the issues identified in SECY-96-077, SNC performed an analysis of the Tier 2* matters listed in 10 CFR Part 52, Appendix D, Section VIII paragraphs B.6.b and B.6.c. The analysis examined each item in terms of the following criteria:

  • Is the Tier 2* information adequately addressed in the VEGP 3 and 4 Plant-specific Tier 1 DCD or VEGP 3 and 4 Combined License (COL)? This step included a review to determine the degree to which codes, standards, and design and qualification processes, are relied upon for ITAAC acceptance criteria, but not specified in the VEGP 3 and 4 Plant-specific Tier 1 DCD.
  • Would changes in the Tier 2* information be adequately addressed by other applicable regulations, e.g., 10 CFR 50.46?
  • Would a change to the Tier 2* information have safety-significance commensurate with a change to Tier 1 information?
  • Would the evaluation process defined in 10 CFR Part 52, Appendix D, paragraph VIII.B.5 consistently and reliably require prior NRC approval of a change to the Tier 2* information?

Following the evaluation process described above, SNC made the following conclusions regarding 9 of the 24 Tier 2* matters listed in 10 CFR Part 52, Appendix D, Section VIII paragraphs B.6.b and B.6.c:

  • First, a set of Tier 2* information is already addressed in Tier 1 and thus a change to this Tier 2* information, which would involve a change to the associated Tier 1 information, would require prior NRC approval. Therefore, neither an evaluation of safety-significance nor new evaluation criteria were considered necessary to provide assurance that changes would receive prior NRC approval.
  • Second, for another set of Tier 2* information it was concluded that a change to this information would not have safety-significance commensurate with a change to Tier 1 information. Thus, new evaluation criteria were not considered necessary for this set of Tier 2* information.
  • Third, it was determined that a change to a third set of Tier 2* information would require a prior NRC approval under 10 CFR Part 52, Appendix D, paragraph VIII.B.5 or another regulation in a consistent and reliable manner. Thus, it was concluded that the evaluation criteria currently provided in 10 CFR Part 52, Appendix D, VIII.B.5.b or VIII.B.5.c are adequate to reliably and consistently address changes to this information and new evaluation criteria to address changes to this information were not necessary.

The remaining 15 of the 24 Tier 2* matters listed in 10 CFR Part 52, Appendix D, Section VIII paragraphs B.6.b and B.6.c were selected for development of additional screening criteria that would determine whether an associated Tier 2* departure qualifies for the departure evaluation process outlined in 10 CFR Part 52, Appendix D, Section VIII.B.5. A summary of the analysis is provided in Enclosure 5U. The selected matters are:

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  • Maximum fuel rod average burn-up
  • Fuel principal design requirements
  • Fuel criteria evaluation process
  • Small-break loss-of-coolant accident (LOCA) analysis methodology
  • Screen design criteria
  • Design Summary of Critical Sections
  • American Concrete Institute (ACI) 318, ACI 349, American National Standards Institute/American Institute of Steel Construction (ANSI/AISC)-690, and American Iron and Steel Institute (AISI), Specification for the Design of Cold Formed Steel Structural Members, Part 1 and 2, 1996 Edition and 2000 Supplement
  • Nuclear design of fuel and reactivity control system, except burn-up limit
  • Instrumentation and control system design processes, methods, and standards
  • Piping design acceptance criteria
  • Human factors engineering
  • Steel composite structural module details A set of criteria was then developed that would be used to determine the critical safety aspects of the above matters to determine whether a proposed departure from Tier 2* could qualify to be evaluated under the departure evaluation process for Tier 2 departures outlined in Section VIII.B.5. A proposed Tier 2* departure would not qualify to be evaluated under Section VIII.B.5, if it:
1. Involves 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,
2. Results in a change to a design process described in the plant-specific DCD that is material to implementation of an industry standard or endorsed regulatory guidance,
3. (i) Results in a change to the fuel criteria evaluation process, the fuel principal design requirements, or the nuclear design of the fuel and the reactivity control system that is material to a fuel or reactivity control system design function, or the evaluation methods in WCAP-12488, Westinghouse Fuel Criteria Evaluation Process, or (ii) Results in any change to the maximum fuel rod average burn-up limits or the small break LOCA analysis methodology described in UFSAR Subsections 15.6.5.4B.2.2 or 15.6.5.4B.2.3,
4. Adversely affects the containment debris limits or debris screen design criteria,
5. Changes to the Reactor Coolant Pump (RCP) type from a canned motor to a different type of RCP, Page 10 of 28

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6. Results in a change to the Passive Residual Heat Removal Heat Exchanger natural circulation test (first plant test), the Core Makeup Tank Heated Recirculation Tests (first three plants test), or the Automatic Depressurization System Blowdown Test (first three plants test) that is material to the test objectives or test performance criteria,
7. Involves structural materials or analytical or design methods, including design codes and analytical assumptions, that deviate from those credited in the plant-specific DCD for critical sections,
8. Results in a change to the design of the steel faceplates, internal trusses, tie bars, or headed studs of the steel-concrete (SC) module walls in the Nuclear Island or the Shield Building, including SC-to-reinforced concrete (RC) connections,
9. Results in an increase in the demand to capacity (D/C) ratio of a critical section of the structure. SNC shall determine the D/C ratio under this condition for each critical section structural member including, but not limited to, wall segments, wall sections, concrete panels, slabs, or basemat sections, affected by a departure by:

(i) Using the Tier 2* information in the UFSAR Section 3.8 or Appendix 3H table that directly states the D/C ratio or states the area of steel provided and the area of steel required for the affected structural member, or (ii) Providing the same total area of steel across the entire critical section using any combination of rebar sizes and spacing allowed by the design basis codes used in the UFSAR as the total area of steel specified in UFSAR Section 3.8 and Appendix 3H tables marked Tier 2*.

Criterion 1 represents screening criteria that were developed as a result of the analysis performed that was related to the following Tier 2* matters:

  • Design Summary of Critical Sections
  • American Concrete Institute (ACI) 318, ACI 349, American National Standards Institute/American Institute of Steel Construction (ANSI/AISC)-690, and American Iron and Steel Institute (AISI), Specification for the Design of Cold Formed Steel Structural Members, Part 1 and 2, 1996 Edition and 2000 Supplement
  • Steel composite structural module details Criterion 2 represents screening criteria that were developed as a result of the analysis performed that was related to the following Tier 2* matters:
  • Instrumentation and control system design processes, methods, and standards
  • Piping design acceptance criteria
  • Human factors engineering Criterion 3 represents screening criteria that were developed as a result of the analysis performed that was related to the following Tier 2* matters:
  • Maximum fuel rod average burn-up Page 11 of 28

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  • Fuel principal design requirements
  • Fuel criteria evaluation process
  • Nuclear design of fuel and reactivity control system, except burn-up limit
  • Small-break loss-of-coolant accident (LOCA) analysis methodology Criterion 4 represents screening criteria that were developed as a result of the analysis performed that was related to the following Tier 2* matter:
  • Screen design criteria Criterion 5 represents screening criteria that were developed as a result of the analysis performed that was related to the following Tier 2* matter:
  • Reactor coolant pump type Criterion 6 represents screening criteria that were developed as a result of the analysis performed that was related to the following Tier 2* matters:

Criteria 7, 8 and 9 represent screening criteria that were developed as a result of the analysis performed that was related to the following Tier 2* matter:

  • Design Summary of Critical Sections.

To ensure consistent application of the evaluation criteria, the following detailed guidance would be used to perform the evaluations:

Criterion 1 (Codes and Standards) detailed guidance:

  • Use of a code or standard not approved by the NRC is a deviation from a code or standard.
  • Use of a later edition of a code or standard than the edition approved by the NRC is a deviation from a code or standard.
  • Use of an equivalent code or standard is a deviation from a code or standard.
  • Changes to design output using the approved standards and codes (e.g., structural dimensions) are not deviations provided the standard or code limit is met.
  • Editorial and grammatical corrections are not deviations.
  • Corrections required to achieve consistency within the document are not deviations.

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Criterion 1 (Codes and Standards) Bases:

It is noted that some, but not all, codes and standards credited in the plant-specific DCD for the design or construction of the AP1000 are referenced in the VEGP Plant-specific Tier 1 DCD.

Therefore, this screening criterion assures that Tier 2* departures involving deviations from codes and standards will be submitted for prior NRC approval. It should be noted that the detailed guidance examples for Criterion 1 are more conservative than criteria that would be applied to a departure evaluated under the requirements of 10 CFR Part 52, Appendix D, Section VIII paragraph B.5 because the proposed criteria require prior NRC approval for deviations from codes and standards while regulatory guidance related to the application of paragraph B.5 allows some flexibility when evaluating deviations from codes and standards.

Applicable codes and standards are designated in the plant-specific Tier 2 DCD portion of the VEGP Units 3 and 4 Updated Final Safety Analysis Report (UFSAR). Departures from the plant-specific Tier 2 DCD are controlled by 10 CFR Part 52, Appendix D, Section VIII.B.5.

Regulatory guidance for the evaluation of departures from the UFSAR is contained in NEI 96-07, Guidelines for 10 CFR 50.59 Implementation, Revision 1, and NEI 96-07, Appendix C, Guideline for Implementation of Change Processes for New Nuclear Power Plants Licensed Under 10 CFR Part 52, Revision 0. NEI 96-07, Sections 4.3.1 and 4.3.2 state in part, Although this criterion allows minimal increases, licensees must still meet applicable regulatory requirements and other acceptance criteria to which they are committed (such as contained in regulatory guides and nationally recognized industry consensus standards; e.g., the ASME B&PV Code and IEEE standards). Further, departures from the design, fabrication, construction, testing and performance standards as outlined in the General Design Criteria (Appendix A to Part 50) are not compatible with a no more than minimal increase standard Because safety-significant departures from codes and standards would require prior NRC approval, the expectation for safety-significant information changes related to codes and standards to require prior NRC approval continues to be met.

Criterion 2 (Design Processes) detailed guidance:

A material change affects a design process output, or method of performing a design process, or method of controlling the design process.

  • The following are examples of material changes:

o The addition, deletion, or alteration of a design process step o Reconfiguration of design process steps o Departures from regulatory guidance related to the design process o Alteration of a detail that serves as the basis for acceptance in an NRC Final Safety Evaluation Report (FSER) related to the affected design process

  • The following examples are not material changes:

o Editorial changes o Clarifications to improve reader understanding o Correction of inconsistencies within the document which are clearly discernible (e.g.,

between sections) o Changes that do not change the meaning or substance of information presented (e.g.,

reformatting or removing detail as described in NEI 98-03, Revision 1, Guidelines for Updating Final Safety Analysis Reports, Section A4 [ADAMS Accession Number ML003779028])

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Criterion 2 (Design Processes) Bases:

The design processes addressed in the VEGP 3 and 4 Plant-specific Tier 1 DCD and for which some Tier 2* information is contained in the VEGP 3 and 4 plant-specific Tier 2 DCD are:

  • Diverse Actuation System (Plant-specific Tier 1 DCD, Section 2.5.1; Plant-specific Tier 2 DCD, Chapter 7);
  • Protection and Safety Monitoring System (Plant-specific Tier 1 DCD, Section 2.5.2; Plant-specific Tier 2 DCD, Chapter 7);
  • Component Interface Module (Plant-specific Tier 1 DCD, Section 2.5.2; Plant-specific Tier 2 DCD, Chapter 7);
  • Piping design acceptance criteria (multiple system sections in the plant-specific Tier 1; plant-specific Tier 2 DCD, Subsections 3.6.2 and 3.9.3);
  • Human Factors Engineering (Plant-specific Tier 1 DCD, Section 3.2; Plant-specific Tier 2 DCD, Chapter 18);

Diverse Actuation System (DAS)

While paragraph B.5.b allows changes to design processes without prior NRC approval provided that the design function is not more than minimally adversely6 affected, this new criterion does not allow any material change to a design process.

Protection and Safety Monitoring System (PMS)

UFSAR Tier 2* information related to PMS is contained in Westinghouse WCAP reports that are incorporated by reference into the UFSAR. For the PMS, departures related to a design process as described in Westinghouse WCAPs may not be easily evaluated against the eight criteria of paragraph B.5.b; therefore, some departures may not receive prior NRC approval as required. The application of proposed Criterion 2 assures that any material departure related to PMS design processes receives prior NRC approval.

Component Interface Module (CIM)

UFSAR Tier 2* information related to the CIM is contained in WCAP-17179-P (Proprietary) and WCAP-17179-NP (Non-Proprietary), which are incorporated by reference into the UFSAR. For the CIM, departures related to a design process as described in WCAP-17179-P/NP may not be easily evaluated against the eight criteria of paragraph B.5.b; therefore, some departures may not receive prior NRC approval as required. The application of proposed Criterion 2 assures that any material departure related to the CIM design processes receives prior NRC approval.

Piping Design Acceptance Criteria (DAC)

This UFSAR Tier 2* text describes a design process for piping design that is used to implement an industry standard (e.g., ASME Code) or endorsed regulatory guidance. For example, as explained in UFSAR Section 3.6.2.1.1, this text defines the process for 6 The use of the terms adverse/adversely and design function, as used in the guidance discussions, is derived from the use of the same terms in NEI 96-07, Revision 1, Guidelines For 10 CFR 50.59 Implementation [ADAMS Accession Number ML003771157]. This NEI guidance provides an extensive discussion regarding how to evaluate whether a change adversely affects a design function. The term design function is defined in NEI 96-07, Section 3.3.

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ND-18-1006 U Updated Request for License Amendment: Changes to Tier 2* Departure Evaluation Process (LAR-17-037S4) determining pipe break locations in piping designed and constructed to the requirements for Class 1 piping in the ASME Code,Section III, Division 1. Departures related to this design process may not be easily evaluated against the eight criteria of paragraph B.5.b; therefore, some departures may not receive prior NRC approval as required. The application of proposed Criterion 2 assures that any material departure related to piping DAC receives prior NRC approval.

Human Factors Engineering (HFE)

The UFSAR Tier 2* information related to HFE is contained in the Westinghouse documents that are incorporated by reference into the UFSAR. For HFE, departures related to a design process as described in Westinghouse documents may not be easily evaluated against the eight criteria of paragraph B.5.b; therefore, some departures may not receive prior NRC approval as required. The application of proposed Criterion 2 assures that any material departure related to HFE design processes receives prior NRC approval.

Regulatory assurance related to design processes is assured through the inclusion of key design processes in the VEGP Units 3 and 4 Plant-specific Tier 1 DCD because 10 CFR Part 52, Section VIII, paragraph B.5.a requires that Tier 2 departures involving Tier 1 information receive prior NRC approval. The key design processes included in the VEGP 3 and 4 Plant-specific Tier 1 DCD that have information designated as Tier 2* in the VEGP 3 and 4 plant-specific Tier 2 DCD are related to the Diverse Actuation System (DAS), Protection and Safety Monitoring System (PMS), piping design acceptance criteria, and Human Factors Engineering (HFE).

VEGP 3 and 4 Plant-specific Tier 1 DCD, Section 2.5.1, Diverse Actuation System, contains a description as to how the associated hardware and software is to be designed during the following life cycle stages:

a) Development phase for hardware and any software b) System test phase c) Installation phase Details of the design process are verified in associated inspections, tests, analyses, and acceptance criteria (ITAAC).

VEGP 3 and 4 Tier 1 Plant -specific DCD, Section 2.5.2, Protection and Safety Monitoring System, contains a description of the requirements for the development of associated hardware and software during the following life cycle stages.

a) Design requirements phase, may be referred to as conceptual or project definition phase (Complete) b) System definition phase c) Hardware and software development phase, consisting of hardware and software design and implementation d) System integration and test phase e) Installation phase Additional requirements listed for software design, testing and maintenance include:

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ND-18-1006 U Updated Request for License Amendment: Changes to Tier 2* Departure Evaluation Process (LAR-17-037S4) a) Software management including documentation requirements, standards, review requirements, and procedures for problem reporting and corrective action.

b) Software configuration management including historical records of software and control of software changes.

c) Verification and validation including requirements for reviewer independence.

Details of the design process are verified in associated inspections, tests, analyses, and acceptance criteria (ITAAC).

Various system ITAAC in the plant-specific Tier 1 DCD address piping design. For example, plant-specific Tier 1 Section 2.1.2, Reactor Coolant System, item 6 reads as follows:

Each of the as-built lines identified in Table 2.1.2-2 as designed for leak before break (LBB) meets the LBB criteria, or an evaluation is performed of the protection from the dynamic effects of a rupture of the line.

The Tier 2* text in plant-specific Tier 2 DCD Subsections 3.6.2 and 3.9.3 define the processes (i.e., piping design acceptance criteria) necessary to implement the Tier 1 requirement. These processes define, for example, how to determine pipe break locations for ASME Code Class 1, 2 and 3 piping systems.

Tier 1 DCD, Section 3.2, Human Factors Engineering, contains a description of the process to be used when designing the operation and control centers system (OCS). The design description for the HFE program states in part, The AP1000 human-system interface (HSI) will be developed and implemented based upon a human factors engineering (HFE) program. Figure 3.2-1 illustrates the HFE program elements. The HSI scope includes the design of the operation and control centers system (OCS) and each of the HSI resources. For the purposes of the HFE program, the OCS includes the main control room (MCR), the remote shutdown workstation (RSW), the local control stations, and the associated workstations for each of these centers. The HSI resources include the wall panel information system, alarm system, plant information system (nonsafety-related displays), qualified safety-related displays, and soft and dedicated controls.

Minimum inventories of controls, displays, and visual alerts are specified as part of the HSI for the MCR and the RSW Criterion 3 (Nuclear Fuel) detailed guidance:

  • A material change is any change in a method of evaluation or calculation. Note that WCAP-12488, Westinghouse Fuel Criteria Evaluation Process provides the fuel criteria evaluation process. This WCAP topical report describes the process and criteria that applies to changes in existing fuel designs that will not require NRC review and approval as long as these criteria are satisfied. Changes made in accordance with this WCAP are not considered material changes to the fuel criteria evaluation process, the fuel principal design requirements, the maximum fuel rod average burn-up limits, or the nuclear design of fuel and reactivity control system. The proposed Tier 2* screening and evaluation process criterion would not permit material changes to the WCAP-12488.
  • A material change to a design would be any change that has an adverse effect on a design function.
  • Any change to the maximum fuel rod average burn-up limits requires prior NRC approval.

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  • The following examples are not material changes:

o Editorial changes o Clarifications to improve reader understanding o Correction of inconsistencies within the document which are clearly discernible (e.g., between sections) o Changes that do not change the meaning or substance of information presented (e.g., reformatting or removing detail as described in NEI 98-03, Revision 1, Guidelines for Updating Final Safety Analysis Reports, Section A4 [ADAMS Accession Number ML003779028])

Criterion 3 (Nuclear Fuel) Bases:

The VEGP 3 and 4 Plant-specific Tier 1 DCD does not contain information related to nuclear fuel.

Proposed Criterion 3 would provide assurance that a change to the fuel criteria evaluation process, the fuel principal design requirements, or the nuclear design of the fuel and the reactivity control system that is material to a fuel or reactivity control system design function or the evaluation methods in WCAP-12488, Westinghouse Fuel Criteria Evaluation Process,, or any change to the maximum fuel rod average burn-up limits would receive prior NRC approval. It should be noted that the proposed Criterion 3 is more conservative than criteria that would be applied to a departure evaluated under 10 CFR Part 52, Appendix D, Section VIII Paragraph B.5 because proposed Criterion 3 does not apply the no more than minimal standard. In addition, Criterion 3 does not allow changes to methods of evaluation.

Due to the uniqueness of the AP1000 design, the use of the NOTRUMP code is considered acceptable, in part, because of the identified Tier 2* information in Chapter 15 (two paragraphs in Subsections 15.6.5.4B.2.2 and 15.6.5.4B.2.3). The Tier 2* information associated with NOTRUMP homogeneous sensitivity model and critical heat flux assessment during accumulator injection is considered to be safety-significant and an integral aspect of the methodology as approved for the AP1000. Therefore, any changes to that information would involve a departure from a method of evaluation described in the FSAR and require prior NRC review and approval.

Criterion 4 (Debris Screen) detailed guidance:

  • An adverse change is any change that would be considered a non-conservative change of a debris value established in the UFSAR.
  • An adverse change would be any change that changes any element of the evaluations used to determine the design of the debris screens.
  • Containment resident debris limit is defined in UFSAR Subsection 6.3.2.2.7.1 (item 12).
  • Fibrous debris limit is defined in UFSAR Subsection 6.3.2.2.7.1 (item 12).
  • The criteria apply to departures affecting the In-Containment Refueling Water Storage Tank (IRWST) Screens and the Containment Recirculation Screens.

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Application of the criteria related to debris values is demonstrated by the following examples:

Example 1:

Following a refueling outage, a review of the containment closeout inspection results reveals that the calculated total amount of resident containment debris is 135 pounds. An engineering analysis determines that the screens would still be able to meet their safety function. Consequently, a change is proposed to raise the limit in the UFSAR to 135 pounds.

The proposed departure would be evaluated against all nine proposed Qualifying Criteria, and when evaluated against Qualifying Criterion 4, the evaluation would determine that the proposed departure is considered an adverse effect on containment debris limits and would require prior NRC approval before implementation. The condition would be considered adverse because any relaxation of the limit (increase in value) would be considered adverse.

Example 2:

A design change is proposed that improves the effectiveness of the screens. The engineering evaluation, using the methodology described in the UFSAR, demonstrates that the fibrous debris limit could be raised to 10 pounds. As a result, it is proposed to raise the limit in the UFSAR to 10 pounds.

The proposed departure would be evaluated against all nine proposed Qualifying Criteria, and when evaluated against Qualifying Criterion 4, the evaluation would determine that the proposed departure is considered an adverse effect on the containment debris limits and would require prior NRC approval before implementation. The condition would be considered adverse because any relaxation of the limit (increase in value) would be considered adverse.

Example 3:

A design change to the containment screens is proposed which would alter the size of the screens slightly. An engineering evaluation determines the screens would continue to meet their design function if the fibrous debris limit were set at 6.0 pounds.

Consequently, it is proposed to revise the UFSAR to change the limit from 6.6 pounds to 6.0 pounds.

The proposed departure would be evaluated against all nine proposed Qualifying Criteria, and when evaluated against Qualifying Criterion 4, the evaluation would determine that the proposed debris limit departure is not considered an adverse effect on the containment debris limits and would not require prior NRC approval before implementation. The condition would not be considered adverse because the revised limit is more restrictive and continues to ensure the screens meet their design function. However, the proposed departure and associated screen design change would also be evaluated against the criteria of 10 CFR Part 52, Appendix D, paragraph VIII.B.5 and it may be determined that prior NRC approval is required.

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Application of the criteria related to debris screens is demonstrated by the following examples:

Example 1:

A design change is proposed to relocate a stairwell inside containment. An evaluation of the potential impacts of the design change reveals that the stairwell is credited as an intervening structure in the LOCA pipe break analysis, and a ventilation filter (which contains fibrous material) is located 40 inside diameters from the break along an axis that is a continuation of the pipe axis. Per UFSAR Subsection 6.3.2.2.7.1, the ZOI in the absence of intervening components, supports, structures, or other objects includes insulation in a cylindrical area extending out a distance equal to 45 inside diameters from the break along an axis that is a continuation of the pipe axis and up to 5 inside diameters in the radial direction from the axis. The 5 inside diameter limit in the radial direction from the pipe axis continues to be met. An engineering evaluation and testing demonstrate that the non-qualifying insulation material will not be adversely affected by the assumed pipe break. As a result, a change is proposed to revise the UFSAR ZOI limit from 45 inside diameters to 40 inside diameters in this area.

The proposed departure would be evaluated against all nine proposed Qualifying Criteria, and when evaluated against Qualifying Criterion 4, the evaluation would determine that the proposed departure is considered an adverse effect on the debris screen design criteria and require prior NRC approval before implementation. The condition would be considered adverse because any relaxation of the ZOI distance (decrease in value) would be considered adverse.

Example 2:

A design change is proposed that would add a structure in the lower regions of the containment. The impact of the change would be that the maximum post-design basis accident (DBA) LOCA floodup water level would be raised to plant elevation 111.0 feet.

Per UFSAR Subsection 6.3.2.2.7.1, the maximum post-DBA LOCA floodup water level is plant elevation 110.2 feet. Additional analysis reveals that no non-qualifying insulation is located below 111.0 feet. As a result, a change is proposed to revise the UFSAR maximum post-DBA LOCA floodup value to 111.0 feet.

The proposed departure would be evaluated against all nine proposed Qualifying Criteria, and when evaluated against Qualifying Criterion 4, the evaluation would determine that the proposed departure is not considered an adverse effect on the containment debris screen design criteria and would not require prior NRC approval before implementation.

The condition would not be considered adverse because the revised post-DBA LOCA floodup water level is more restrictive and continues to ensure fibrous insulation material will not be introduced following a DBA LOCA. However, the proposed departure and associated design change would also be evaluated against the criteria of 10 CFR Part 52, Appendix D, paragraph VIII.B.5, and this evaluation may determine that prior NRC approval is required.

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Criterion 4 (Debris Screens) Bases:

It is noted that the VEGP 3 and 4 Plant-specific Tier 1 DCD does not contain design description information related to debris screens, but detailed design information is extensively covered in ITAAC (Table 2.2.3-4), which will no longer be part of the licensing basis after the 10 CFR 52.103(g) finding. Screening Criterion 4 provides assurance that departures from Tier 2*

information that adversely affect debris screen design criteria would receive prior NRC approval.

It should be noted that proposed Criterion 4 is more conservative than criteria that would be applied to a departure evaluated under 10 CFR Part 52, Appendix D, Section VIII paragraph B.5 because the proposed criterion does not allow any adverse change7 versus the no more than minimal8 standard used in paragraph B.5.b.

Criterion 5 (Reactor Coolant Pump Type) detailed guidance:

Tier 2* information regarding RCP type is contained in UFSAR Subsection 5.4.1.2.2, Design Description. Any departure from the design of the RCP that would not utilize the canned motor design would meet Criterion 5 and the departure would not qualify for evaluation under paragraph B.5.b.

Criterion 5 (Reactor Coolant Pump Type) Bases The VEGP 3 and 4 Plant-specific Tier 1 DCD does not contain information related to the canned motor design attributes of the RCP. Proposed Criterion 5 would provide assurance that departures from Tier 2* information related to RCP type would receive prior NRC approval.

Criterion 6 (Initial Test Program) detailed guidance:

A change that is material to the test objectives or test performance criteria influences the outcome of the test such that it would affect whether the test objectives or performance criteria would be met.

The following examples are material changes:

  • The addition, deletion, or alteration of a test step
  • Alteration of a detail that serves as the basis for acceptance in an NRC Final Safety Evaluation Report (FSER) related to the affected test The following examples are not material changes:
  • Editorial changes
  • Clarifications to improve reader understanding
  • Correction of inconsistencies within the document which are clearly discernible (e.g.,

between sections)

  • Changes that do not change the meaning or substance of information presented (e.g.,

reformatting or removing detail as described in NEI 98-03, Revision 1, Guidelines for 7 Adverse effects are described in NEI-96-07, Guidelines for 10 CFR 50.59 Implementation, Revision 1

[ADAMS Accession Number ML003771157]

8 The no more than minimal standard is described in NEI-96-07, Guidelines for 10 CFR 50.59 Implementation, Revision 1 [ADAMS Accession Number ML003771157]

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Updating Final Safety Analysis Reports, Section A4 [ADAMS Accession Number ML003779028])

Criterion 6 (Initial Test Program) Bases:

The VEGP 3 and 4 Plant-specific Tier 1 DCD does not contain information related to the special tests that establish a unique phenomenological performance parameter of the AP1000 design features beyond testing performed for Design Certification for the AP600 and that will not change from plant to plant. Proposed Criterion 6 would provide assurance that a change that is material to the test objectives or test performance criteria of the Tier 2* information related to these special tests would receive prior NRC approval. The special tests (i.e., first plant tests and first three plant tests) for which some Tier 2* information is contained in the VEGP 3 and 4 plant-specific Tier 2 DCD are:

  • Core Makeup Tank Heated Recirculation Tests (first three plants test) identified in UFSAR Subsection 14.2.9.1.3 Items (k) and (w)), and

Criteria 7, 8, and 9 (Critical Sections) detailed guidance:

The design of critical sections is important in assuring the integrity of the buildings which house safety related systems and components. The proposed Criteria 7, 8, and 9 rely on a combination of requirements to ensure that safety significant changes to critical section design will require prior NRC review and approval. These requirements address changes to critical section structural materials or analytical or design methods, including design codes and analytical assumptions, changes to critical sections that use steel concrete sandwich construction, and increases in demand to capacity ratios of critical sections using reinforced concrete. Furthermore, changes to the design of the steel faceplates, internal trusses, tie bars, or headed studs of the steel-concrete (SC) module walls in the Nuclear Island or the Shield Building, including SC-to-reinforced concrete (RC) connections will require prior NRC review and approval.

Criteria 7, 8, and 9 (Critical Sections) Bases:

It is noted that the UFSAR provides requirements regarding the structural materials and, analytical and design methods for critical sections including design codes and analytical assumptions.

Changes to these requirements are included in this Criterion and prior NRC approval is required if deviations are proposed to these requirements.

Furthermore, the critical sections criterion contains restrictions regarding the ability to increase demand-to-capacity (D/C) ratios for reinforcement in critical sections without prior NRC staff approval. The Criterion requires prior NRC approval for any increase in the D/C ratio of a critical section of the structure beyond the ratio specified in or calculated from the applicable UFSAR Section 3.8 or Appendix 3H Table marked Tier 2*. SNC shall determine the D/C ratio under this condition for each critical section structural member affected by a departure by either:

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  • Using the Tier 2* information in the UFSAR Section 3.8 or Appendix 3H table that directly states the D/C ratio or states the area of steel provided and the area of steel required for the affected structural member, or
  • Providing the same total area of steel across the entire critical section using any combination of rebar sizes and spacing allowed by the design basis codes used in the UFSAR as the total area of steel specified in UFSAR Section 3.8 and Appendix 3H tables marked Tier 2*.

The term structural member as used in the critical sections criterion refers to a segment of a nuclear island structure including, but not limited to, wall segments, wall sections, concrete panels, slabs, or basemat sections. For example, UFSAR Tier 2* Table 3H.5-3, describes one segment as Wall Section 1, 6, Elevation 180-0 to 135-3, and defines the required and minimum provided area of steel for that segment/structural member.

An example of a common situation during construction is when a reinforcing bar is moved or cannot be placed due to interference. To address this situation and ensure that the critical section structure continues to perform as expected, the design authority may choose to add reinforcement at a nearby location. Thus, the design authority examines the ratio of required reinforcement to provided reinforcement using the total area of steel across the entire critical section using any combination of rebar sizes and spacing allowed by the design basis codes used in the UFSAR as the total area of steel specified in UFSAR Section 3.8 and Appendix 3H tables marked Tier 2*,

and verifies that the resulting ratio does not exceed that ratio value stated in or calculated from UFSAR Section 3.8 and Appendix 3H tables Tier 2* values for the applicable structural member.

Provided that all Code requirements are satisfied and the D/C ratio of the member based on the applicable Tier 2* values are not exceeded, then the change would qualify for the application of the Tier 2 change process.

Potential changes in reinforcement that need to be accommodated due to construction are generally localized (e.g., a missing reinforcing bar due to an interference) and are readily accommodated within design basis code commitments. Changes that affect the overall capacity of a structural member (e.g., changing the overall size and spacing of the reinforcement in a reinforced concrete wall) are not anticipated, and NRC staff review and approval of such a change is required.

Additionally, due to the unique non-symmetries of the shield building with regard to its location within the nuclear island footprint and the varying elevation of the SC-to-RC connection around its circumference, no changes are permitted to the steel faceplates, internal trusses, tie bars, or headed studs of SC module walls in the Nuclear Island or Shield Building, or Shield Building SC-to-RC connections without prior NRC staff review and approval. Proposed local changes such as attachment of miscellaneous low energy small bore piping hanger steel to SC modules would typically be acceptable to perform under the Tier 2 change process. Other proposed changes such as systematically eliminating tie bars in the shield building SC modules would require prior NRC approval.

Because the screening criteria would provide assurance that departures from Tier 2* matters that are safety-significant and would meet criteria for inclusion in a Tier 1 design control document, would require prior NRC approval, the underlying intent of the Tier 2* designation is maintained.

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ND-18-1006 U Updated Request for License Amendment: Changes to Tier 2* Departure Evaluation Process (LAR-17-037S4)

Should a proposed Tier 2* departure meet any of the nine criteria outlined above, then it would not qualify for application of the Tier 2 departure evaluation process and would require prior NRC approval.

Should a Tier 2* departure qualify for evaluation under 10 CFR Part 52, Appendix D, Section VIII.B.5, and be determined to involve more than a minimal safety significance, it would continue to require prior NRC approval through the analysis in Section VIII.B.5.b through VIII.B.5.e for the reasons outlined below.

Because departures from Tier 1 information require prior NRC approval via a license amendment request and an exemption request, and involved Tier 2 departures require prior NRC approval via a license amendment, the expectation for safety-significant information changes related to design processes to require NRC prior approval continues to be met.

The proposed license condition [2.D(13).2(b)] requires SNC to prepare and maintain a written evaluation that provides the bases for its determinations regarding the criteria in License Condition 2.D.(13)(a). In the report that 10 CFR Part 52, Appendix D, Section X.B.1 requires SNC to submit, SNC shall include a brief description of each departure and a summary of the evaluation of the departure. This requirement provides additional assurance that the revised departure process will be implemented correctly.

The proposed regulatory commitment that would require SNC to develop, implement, and maintain detailed procedural guidance related to how the qualifying criteria would be applied to proposed Tier 2* departures ensures that departures from Tier 2* information with a safety significance commensurate with Tier 1 will require prior NRC approval. This procedural guidance will be maintained in accordance with SNCs Commitments Management Program for as long as the license condition remains in effect.

As a result, the proposed change would continue to meet NRC requirements and expectations regarding designation of safety-significant Tier 2 information as Tier 2* and would require NRC review and approval of departures from Tier 2* information that meet the safety significance standard.

The proposed changes do not affect any function or feature used for the prevention or mitigation of accidents or their safety analyses. No safety-related structure, system, component (SSC) or function is involved. The proposed changes neither involve nor interface with any SSC accident initiator or initiating sequence of events related to the accidents evaluated in the UFSAR, and therefore, do not have an adverse effect on any SSC design function.

The proposed changes do not affect the radiological source terms (i.e., amounts and types of radioactive materials released, their release rates, and release durations) used in the accident analyses. The equipment involved in these proposed changes does not affect safety-related equipment or any fission product barrier. No system or design function or equipment qualification is adversely affected by the proposed changes. The changes do not result in a new failure mode, malfunction, or sequence of events that could adversely affect a radioactive material barrier or safety-related equipment. The proposed changes do not allow for a new fission product release path, result in a new fission product barrier failure mode, or create a new sequence of events that would result in significant fuel cladding failures.

Page 23 of 28

ND-18-1006 U Updated Request for License Amendment: Changes to Tier 2* Departure Evaluation Process (LAR-17-037S4)

This license amendment request does not affect SSCs that are used to contain, control, channel, monitor, process or release radioactive or non-radioactive materials. The types and quantities of expected effluents are not changed, and no effluent release path is adversely affected by the proposed changes. Therefore, radioactive and non-radioactive material effluents are not affected by the proposed changes.

Plant radiation zones (as described in UFSAR Section 12.3), controls under 10 CFR Part 20, and expected amounts and types of radioactive materials are not affected by the proposed changes.

Therefore, individual and cumulative radiation exposures do not change.

The change activity has no adverse impact on the emergency plan or the physical security plan implementation, because there are no changes to physical access to credited equipment inside the Nuclear Island (including containment or the auxiliary building) and no adverse impact to plant personnels ability to respond to any plant operations or security event.

4. REGULATORY EVALUATION 4.1 Applicable Regulatory Requirements/Criteria 10 CFR 52.98(f) requires NRC approval for any modification to, addition to, or deletion from the terms and conditions of a combined license (COL). The proposed change involves the addition of a new COL License Condition 2.D.(13) to specify the regulatory process for evaluating departures from Plant-specific Tier 2* matters and Tier 2 information that involves a change to or departure from Tier 2* information, Paragraphs VIII.B.5 and VIII.B.6 subject to the conditions and limitations set forth in new License Condition 2.D.(13). Therefore, NRC approval is required prior to making the plant-specific proposed change in this license amendment request.

4.2 Precedent None.

4.3 Significant Hazards Consideration The requested license amendment would amend, for Southern Nuclear Operating Companys (SNCs) Vogtle Electric Generating Plant (VEGP) Units 3 and 4, Combined License (COL) Numbers NPF-91 (Unit 3) and NPF-92 (Unit 4), the departure evaluation process for qualifying departures from Tier 2* information.

An evaluation to determine whether a significant hazards consideration is involved with the proposed amendment was completed by focusing on the three standards set forth in 10 CFR 50.92, Issuance of amendment, as discussed below:

4.3.1 Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

Page 24 of 28

ND-18-1006 U Updated Request for License Amendment: Changes to Tier 2* Departure Evaluation Process (LAR-17-037S4)

Response: No.

The proposed changes would add a license condition that would allow use of the Tier 2 departure evaluation process for Tier 2* departures, where such departures would not have more than a minimal impact to safety. Changing the criteria by which departures from Tier 2* information are evaluated to determine if NRC approval is required does not affect the plant itself. Changing these criteria does not affect prevention and mitigation of abnormal events, e.g., accidents, anticipated operational occurrences, earthquakes, floods and turbine missiles, or their safety or design analyses. No safety-related structure, system, component (SSC) or function is adversely affected. The changes neither involve nor interface with any SSC accident initiator or initiating sequence of events, and thus, the probabilities of the accidents evaluated in the Updated Final Safety Analysis Report (UFSAR) are not affected. Because the changes do not involve any safety-related SSC or function used to mitigate an accident, the consequences of the accidents evaluated in the UFSAR are not affected.

Therefore, the proposed amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated.

4.3.2 Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

The proposed changes would add a license condition that would allow use of the Tier 2 departure evaluation process for Tier 2* departures, where such departures would not have more than a minimal impact to safety. The changes do not affect the safety-related equipment itself, nor do they affect equipment which, if it failed, could initiate an accident or a failure of a fission product barrier. No analysis is adversely affected. No system or design function or equipment qualification is adversely affected by the changes. This activity does not allow for a new fission product release path, result in a new fission product barrier failure mode, or create a new sequence of events that would result in significant fuel cladding failures. In addition, the changes do not result in a new failure mode, malfunction or sequence of events that could affect safety or safety-related equipment.

Therefore, the proposed amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated.

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ND-18-1006 U Updated Request for License Amendment: Changes to Tier 2* Departure Evaluation Process (LAR-17-037S4) 4.3.3 Does the proposed amendment involve a significant reduction in a margin of safety.

Response: No.

The proposed changes would add a license condition that would allow use of the Tier 2 departure evaluation process for Tier 2* departures, where such departures would not have more than a minimal impact to safety.

The proposed change is not a modification, addition to, or removal of any plant SSCs. Furthermore, the proposed amendment is not a change to procedures or method of control of the nuclear plant or any plant SSCs. The only impact of this activity is the application of the current Tier 2 departure evaluation process to Tier 2* departures.

Therefore, the proposed amendment does not involve a significant reduction in a margin of safety.

Based on the above, it is concluded that the proposed amendment does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of no significant hazards consideration is justified.

4.4 Conclusions In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commissions regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public. Pursuant to 10 CFR 50.92, the requested change does not involve a Significant Hazards Consideration.

5. ENVIRONMENTAL CONSIDERATIONS The proposed changes would add a license condition that would allow use of the Tier 2 departure evaluation process for Tier 2* departures, where such departures would not have more than a minimal impact to safety.

A review has determined that the proposed license condition requires an amendment to the COLs; however, a review of the anticipated construction and operational effects of the proposed amendment and exemption has determined that it meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9), in that:

(i) There is no significant hazards consideration.

As documented above, an evaluation was completed to determine whether or not a significant hazards consideration is involved by focusing on the three standards set forth Page 26 of 28

ND-18-1006 U Updated Request for License Amendment: Changes to Tier 2* Departure Evaluation Process (LAR-17-037S4) in 10 CFR 50.92, Issuance of amendment. The Significant Hazards Consideration determined that (1) the proposed amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated; (2) the proposed amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated; and (3) the proposed amendment does not involve a significant reduction in a margin of safety. Therefore, it is concluded that the proposed amendment does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and accordingly, a finding of no significant hazards consideration is justified.

(ii) There is no significant change in the types or significant increase in the amounts of any effluents that may be released offsite.

The proposed changes would add a license condition that establishes a departure evaluation process to determine whether site-specific departures from Tier 2* information would have more than a minimal impact to safety. The proposed changes are unrelated to any aspect of plant construction or operation that would introduce any change to effluent types (e.g., effluents containing chemicals or biocides, sanitary system effluents, and other effluents), or affect any plant radiological or non-radiological effluent release quantities.

Furthermore, the proposed changes do not affect any effluent release path or diminish the functionality of any design or operational features that are credited with controlling the release of effluents during plant operation. Therefore, it is concluded that the requested amendment does not involve a significant change in the types or a significant increase in the amounts of any effluents that may be released offsite.

(iii) There is no significant increase in individual or cumulative occupational radiation exposure.

The proposed changes would add a license condition that would allow use of the Tier 2 departure evaluation process for Tier 2* departures, where such departures would not have more than a minimal impact to safety. Plant radiation zones (addressed in UFSAR Section 12.3) are not affected, and controls under 10 CFR Part 20 preclude a significant increase in occupational radiation exposure. Therefore, the requested amendment does not involve a significant increase in individual or cumulative occupational radiation exposure.

Based on the above review of the requested amendment, it has been determined that anticipated construction and operational effects of the requested amendment do not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluents that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the requested amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), an environmental impact statement or environmental assessment of the proposed amendment and exemption is not required.

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ND-18-1006 U Updated Request for License Amendment: Changes to Tier 2* Departure Evaluation Process (LAR-17-037S4)

6. REFERENCES None.

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Southern Nuclear Operating Company ND-18-1006 Enclosure 2U Vogtle Electric Generating Plant (VEGP) Units 3 and 4 Updated Exemption Request:

Changes to Tier 2* Departure Evaluation Process (LAR-17-037S4)

(This Enclosure consists of nine pages, including this cover page.)

ND-18-1006 U Updated Exemption Request: Changes to Tier 2* Departure Evaluation Process (LAR-17-037S4) 1.0 Purpose Southern Nuclear Operating Company (SNC), the licensee for Vogtle Electric Generating Plant (VEGP) Units 3 and 4, requests a permanent exemption from certain provisions of 10 CFR Part 52, Appendix D, Design Certification Rule for AP1000 Design, to allow plant-specific departures from Tier 2* matters, and from Tier 2 information that involves a change to or departure from Tier 2* matters, identified in 10 CFR Part 52, Appendix D, Section VIII.B.6, without prior NRC approval for qualifying Tier 2* departures.

Under the current departure evaluation process applicable to Tier 2* information, SNC must seek prior NRC approval through a License Amendment Request (LAR) for any proposed change to Tier 2* information, even if SNC can demonstrate that the change results in no more than a minimal impact to safety or improves safety. As stated in SECY-17-0075, Planned Improvements in Design Certification Tiered Information Designations, [ADAMS Accession Number ML16196A321] One specific lesson is that some information has been designated as Tier 2* when other regulatory tools could have been used instead to ensure a facility is safely designed, constructed and operated. This results in licensees submitting license amendment requests (LARs) on topics that may not involve safety significant facility changes. Thus, although the Tier 2* designation and associated departure evaluation process was intended to require NRC approval for safety-significant information, in practice, the Tier 2* departure evaluation process has resulted in LARs for departures that are not safety-significant. This is consistent with SNCs experience with the Tier 2* departure evaluation process. In order to mitigate the regulatory inefficiency associated with this issue, SNC proposes a site-specific permanent exemption and license amendment that would apply the existing Tier 2 departure evaluation process to some proposed Tier 2* departures and Tier 2 departures that involve a change to or departure from Tier 2* information, provided the proposed Tier 2* departure does not meet any of the proposed screening criteria which would exclude Tier 2*

departures of high safety significance. Application of the Tier 2 departure evaluation process and the proposed screening criteria would ensure that any safety-significant departures from Tier 2* information would continue to require prior NRC approval, while departures that would improve safety or would result in no more than a minimal impact to safety could proceed as a departure without prior NRC approval.

The specific provisions of Appendix D from which SNC requests an exemption are:

  • Section VIII, Processes for Changes and Departures, Subsection B, Tier 2 Information, paragraph 6.b (VIII.B.6.b):

Paragraph VIII.B.6.b requires a licensee who references 10 CFR Part 52, Appendix D to obtain NRC approval prior to departing from the eight identified categories of Tier 2*

matters. (SNC was previously granted an exemption from Criterion (4), regarding Fire Areas [ADAMS Accession Number ML15191A128].) The requested exemption would allow application of the Tier 2 change process outlined in VIII.B.5 for qualifying Tier 2*

departures for the remaining seven identified categories of Tier 2* matters based on new screening criteria. The requested exemption does not change the list of categories of Tier 2* matters provided in B.6.b.

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ND-18-1006 U Updated Exemption Request: Changes to Tier 2* Departure Evaluation Process (LAR-17-037S4)

  • Section VIII, Processes for Changes and Departures, Subsection B, Tier 2 Information, paragraph 6.c (VIII.B.6.c):

Paragraph VIII.B.6.c refers to paragraph VIII.B.6.b for the departure process which requires a licensee who references 10 CFR Part 52, Appendix D to obtain NRC approval prior to departing from the 16 identified categories of Tier 2* matters that will revert to Tier 2 status after the plant first achieves full power. The requested exemption would allow application of the Tier 2 change process outlined in VIII.B.5 for qualifying Tier 2* departures based on new screening criteria. The requested exemption does not change the list of categories of Tier 2* matters provided in paragraph B.6.c.

  • Section VIII, Processes for Changes and Departures, Subsection B, Tier 2 Information, paragraph 5.a (VIII.B.5.a)

Paragraph VIII.B.5.a specifies that a licensee who references 10 CFR Part 52, Appendix D may depart from Tier 2 information, without prior NRC approval, unless the proposed departure involves a change to or departure from Tier 1 information, Tier 2*

information, or the TS, or requires a license amendment under paragraphs B.5.b or B.5.c of this section. The requested exemption would allow application of the Tier 2 change process outlined in VIII.B.5 for qualifying Tier 2* departures based on new screening criteria.

This request for exemption provides the technical and regulatory basis to demonstrate that 10 CFR 52.7 and §50.12 requirements are met with regards to the Tier 2* departure evaluation process changes identified above.

2.0 Background The Licensee is the holder of Combined License Nos. NPF-91 and NPF-92, which authorize construction and operation of two Westinghouse Electric Company AP1000 nuclear plants, named Vogtle Electric Generating Plant (VEGP) Units 3 and 4, respectively.

The NRC issued the first Part 52 licenses to SNCs VEGP Units 3 and 4 in February 2012.

Changes to the licensing bases for those licenses are governed, in part, by 10 CFR Part 52, Appendix D, Paragraph VIII.B. This portion of the regulations specifies the change process for Tier 2* information and requires NRC approval for all departures from Tier 2* information.

SNC was the first applicant to receive 10 CFR Part 52 licenses and begin construction under the 10 CFR Part 52 regulatory processes. Prior to the associated construction experience, the impact of departures to Tier 2* information during construction could not be entirely understood. Experience has shown that more departures are needed than were initially expected.

SNC has identified several examples of departures from Tier 2* information that were not safety-significant, but nonetheless required prior NRC approval through a LAR.

Application of the Tier 2 departure evaluation process to these proposed departures would Page 3 of 9

ND-18-1006 U Updated Exemption Request: Changes to Tier 2* Departure Evaluation Process (LAR-17-037S4) have concluded with a determination that the proposed change was not safety-significant and could therefore have been processed as a departure not requiring prior NRC approval consistent with 10 CFR Part 52, Appendix D, Section VIII paragraphs B.5.b and B.5.c.

Specific details of the examples supporting this request for exemption are provided in Section 2 of the associated License Amendment Request provided in Enclosure 1U of this letter.

SNC acknowledges that the Commission employed a Tier 2* designation to capture certain significant AP1000 design information existing in Tier 2 that the Commission did not want changed without prior approval (see 71 Fed. Reg. 4474 (Jan. 27, 2006)). In SECY-17-0075, the NRC discussed the reasons for designating some Tier 2 information as Tier 2* and indicated that Tier 2* information is intended to have substantial safety significance, commensurate with information designated as Tier 1. However, SECY 0075 suggests that 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; e.g., background information and other information of minimal safety significance. Furthermore, SNCs experience has demonstrated that not every change to information designated as Tier 2* has an impact on the safety-significant nature, if any, of the information. As such, SNC proposes to invoke a process functionally consistent with departure evaluation processes applied by current applicants for the certification of designs that contain no Tier 2* information. Specifically, SNC is requesting a site-specific amendment that would allow qualifying departures from Tier 2* information to be evaluated under the existing departure evaluation process for Tier 2 departures in 10 CFR Part 52, Appendix D, Paragraphs VIII.B.5.a through VIII.B.5.e. Qualification of Tier 2* changes for the Tier 2 departure process would be determined by applying screening criteria to proposed changes to Tier 2* information in order to exclude changes to Tier 2* information that involve safety significance commensurate with Tier 1 information. Such non-qualifying Tier 2* changes would continue to require prior NRC approval in accordance with 10 CFR Part 52, Appendix D, Paragraph VIII.B.6. Implementation of the proposed license condition requires a permanent exemption from the current provisions of 10 CFR Part 52, Appendix D, to allow plant-specific departures from Tier 2* matters, and from Tier 2 information that involves a change to or departure from Tier 2* matters, identified in 10 CFR Part 52, Appendix D, Paragraph VIII.B.6, without prior NRC approval.

3.0 Technical Justification of Acceptability The departure evaluation process proposed by SNC would apply screening criteria to proposed Tier 2* departures to determine if the departure would qualify to be evaluated using the Tier 2 departure evaluation criteria in 10 CFR Part 52, Appendix D, Paragraphs VIII.B.5.a through VIII.B.5.e to identify those departures that require prior NRC approval.

Tier 2* departures that do not qualify because they meet one or more of the screening criteria would continue to be submitted for prior NRC approval in accordance with 10 CFR Part 52, Appendix D, Paragraph VIII.B.6.

The new departure evaluation process for Tier 2* departures that do not meet any of the proposed new screening criteria would be the same as existing processes governing Tier 2 information. In general, current regulations allow Tier 2 information to be changed if a departure evaluation determines that the change only results in a minimal increase Page 4 of 9

ND-18-1006 U Updated Exemption Request: Changes to Tier 2* Departure Evaluation Process (LAR-17-037S4) in the frequency or severity of an adverse event. Regulations governing Tier 2 departures are contained in 10 CFR Part 52, Appendix D, Paragraph VIII.B.5, and provide the departure evaluation method used to determine if Tier 2 departures require prior NRC approval. SNC proposes to use this same departure evaluation process for qualifying Tier 2* departures that do not meet any of the new screening criteria.

The requested exemption would only allow a change to the departure screening and evaluation process for Tier 2* departures and would not actually implement any changes to the design, construction, or operation of the plant. The proposed change does not affect any function or feature used for the prevention and mitigation of accidents or their safety analyses. No safety-related structure, system, component (SSC) or function is involved.

The requested exemption would accomplish the goal of focusing licensee and regulator resources on the more safety-significant change activities by expanding the scope of the existing departure evaluation process in 10 CFR Part 52, Appendix D, paragraph VIII.B.5, for Tier 2 departures to apply to qualifying Tier 2* departures and Tier 2 departures that involve departures from qualifying Tier 2* matters. The proposed exemption and amendment also address issues discussed in the NRCs Part 52 Implementation Self-Assessment Review Report (July 2013) [ADAMS Accession No. ML13196A403], SECY-17-0075, Planned Improvements in Design Certification Tiered Information Designations

[ADAMS Accession Number ML16196A321], and SECY-96-077, Certification of Two Evolutionary Designs, April 15, 1996 [ADAMS Accession No. ML003708129].

Detailed technical justification supporting this request for exemption is provided in Section 3 of the associated License Amendment Request in Enclosure 1U of this letter.

4.0 Justification of Exemption 10 CFR 52.7 governs the granting of exemptions from the requirements of 10 CFR Part 52, with consideration governed by the requirements of 10 CFR 50.12. Since SNC has identified a need to deviate from 10 CFR Part 52, Appendix D regulations as discussed in Enclosure 1U of the accompanying License Amendment Request, an exemption from the regulations is needed.

10 CFR 52.7 and §50.12 state that the NRC may grant exemptions from the requirements of the regulations provided four conditions are met: 1) the exemption is authorized by law

[§50.12(a)(1)]; 2) the exemption will not present an undue risk to the health and safety of the public [§50.12(a)(1)]; 3) the exemption is consistent with the common defense and security [§50.12(a)(1)]; and 4) special circumstances are present [§50.12(a)(2)].

The requested exemption satisfies the criteria for granting specific exemptions, as described below.

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ND-18-1006 U Updated Exemption Request: Changes to Tier 2* Departure Evaluation Process (LAR-17-037S4)

1. This exemption is authorized by law The NRC has authority under 10 CFR 52.7 and §50.12 to grant exemptions from the requirements of NRC regulations. Specifically, 10 CFR 50.12 and §52.7 state that the NRC may grant exemptions from the requirements of 10 CFR Part 52 upon a proper showing. No law exists that would preclude the changes covered by this exemption request. Additionally, granting of the proposed exemption does not result in a violation of the Atomic Energy Act of 1954, as amended, or the Commissions regulations.

Accordingly, this requested exemption is authorized by law, as required by 10 CFR 50.12(a)(1).

2. This exemption will not present an undue risk to the health and safety of the public The proposed exemption would allow departures from Tier 2* information using the Tier 2 departure process when those departures do not meet any of the new screening criteria. The exemption would only authorize departures from Tier 2*

information without NRC approval when those departures are determined to have no more than a minimal impact to safety. Because the exemption would allow departures from Tier 2* information without NRC approval only after evaluation against the screening criteria defined in a new License Condition and application of the Tier 2 departure evaluation criteria, any safety-significant departures would continue to require prior NRC approval.

Therefore, the requested exemption from 10 CFR 52, Appendix D, Section VIII, paragraphs B.5.a, B.6.b, and B.6.c would not present an undue risk to the health and safety of the public.

3. The exemption is consistent with the common defense and security The exemption would allow a departure from Tier 2* information without prior NRC approval only if: a. the change is qualified for the revised change process by the application of the new screening criteria; and b. it were determined that NRC approval is not required by existing departure evaluation criteria for Tier 2 information. The exemption would not alter the design, function, or operation of any plant equipment that is necessary to maintain a safe and secure status of the plant.

The proposed exemption has no impact on plant security or safeguards procedures, systems, or equipment.

Therefore, the requested exemption is consistent with the common defense and security.

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ND-18-1006 U Updated Exemption Request: Changes to Tier 2* Departure Evaluation Process (LAR-17-037S4)

4. Special circumstances are present 10 CFR 50.12(a)(2) lists six special circumstances for which an exemption may be granted. Only one of these special circumstances need be present before granting an exemption request. In this case, two of the six special circumstances are present, specifically 10 CFR 50.12(a)(2)(ii) and (iii).

4.1 Application would not serve the underlying purpose of the rule 10 CFR 50.12(a)(2)(ii) defines special circumstances as when [a]pplication of the regulation in the particular circumstances would not serve the underlying purpose of the rule or is not necessary to achieve the underlying purpose of the rule. The rule under consideration is 10 CFR Part 52, Appendix D, specifically Section VIII, the departure evaluation process. Certain Tier 2 information is identified as Tier 2* to reflect the potential safety significance of the Tier 2* information. The NRC was specifically concerned with certain significant information [that] only exists in Tier 2 [that] the Commission does not want [] to be changed without prior NRC approval. 71 Fed. Reg. at 4474. Accordingly, the underlying purpose of requiring prior NRC approval for departures from Tier 2* information is to prevent potentially safety-significant changes to plant-specific DCD Tier 2 information without prior NRC review and approval. However, compliance with 10 CFR Part 52, Appendix D, Section VIII, B.6, currently requires the licensee to obtain NRC approval for any change to Tier 2* information - even those having no more than a minimal impact to safety.

Because the exemption would allow departures from Tier 2* without NRC approval only after evaluation against the screening criteria defined in a new License Condition and application of the Tier 2 departure screening and evaluation criteria for Tier 2* departures, any safety-significant departures would continue to require prior NRC approval.

Furthermore, the requested exemption would only allow a change to the departure screening and evaluation process for Tier 2* departures and would not actually implement any changes to the design, construction, or operation of the plant. The proposed change does not affect any function or feature used for the prevention and mitigation of accidents or their safety analyses. No safety-related structure, system, component (SSC) or function is involved.

Therefore, application of 10 CFR 52, Appendix D, Section VIII, paragraphs B.5.a, B.6.b, and B.6.c, in the particular circumstances discussed in this request, is not necessary to achieve the underlying purpose of the rule.

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ND-18-1006 U Updated Exemption Request: Changes to Tier 2* Departure Evaluation Process (LAR-17-037S4) 4.2 Compliance would result in undue hardship 10 CFR 50.12(a)(2)(iii) defines special circumstances as when [c]ompliance would result in undue hardship or other costs that are significantly in excess of those contemplated when the regulation was adopted, or that are significantly in excess of those incurred by others similarly situated. The NRCs goal was to identify only that significant information for which prior approval was required if changes were proposed to be made. It follows then that the NRCs assessment of the compliance obligation for licensees related to the Tier 2*

designation would be limited to those instances involving safety-significant departures from this Tier 2* information.

The current departure evaluation process has no mechanism whereby departures that have minimal bearing on the safety-significant nature of Tier 2*

information can be made without NRC approval. Compliance imposes significant costs and delay, both to SNC and to the NRC, without a corresponding benefit. The requested exemption would accomplish the goal of focusing licensee and regulator resources on the more safety-significant change activities by expanding the scope of the existing departure evaluation process in 10 CFR Part 52, Appendix D, VIII.B.5, for Tier 2 departures to apply to qualifying Tier 2* departures and Tier 2 departures that involve departures from qualifying Tier 2* matters; thereby minimizing any undue hardship associated with the current requirement for prior NRC approval of any change to Tier 2* information, regardless of its safety-significance.

Therefore, compliance with 10 CFR Part 52, Appendix D, Section VIII, paragraphs B.5.a, B.6.b, and B.6.c, in the particular circumstances discussed in this request, would result in undue hardship.

5.0 Risk Assessment A risk assessment was determined to be not applicable to address the acceptability of this request.

6.0 Precedent Exemptions The NRC has long used screenings and evaluations as a regulatory tool; e.g., 10 CFR 50.59. The change process for Tier 2 information has been effective at ensuring that departures that would result in more than a minimal impact to safety require prior NRC approval through an LAR. SNC proposes to use the same departure evaluation process for departures from Tier 2* information that do not meet any of the criteria specified in new License Condition 2.D.(13). SNCs proposal is consistent with the statement in the NRCs Principles of Good Regulation, Regulatory activities should be consistent with the degree of risk reduction they achieve.

Page 8 of 9

ND-18-1006 U Updated Exemption Request: Changes to Tier 2* Departure Evaluation Process (LAR-17-037S4) 7.0 Environmental Consideration The Licensee requests a permanent exemption from certain provisions of 10 CFR Part 52, Appendix D, Design Certification Rule for AP1000 Design, to allow the application of screening and the existing Tier 2 departure evaluation process to proposed Tier 2*

departures and departures from Tier 2 information that involves a change to or departure from Tier 2* matters, and that do not meet any of the criteria of new License Condition 2.D.(13). However, the Licensee evaluation of the proposed exemption has determined that the proposed exemption meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).

Based on the above review of the proposed exemption, the Licensee has determined that the proposed activity does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluents that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed exemption meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), an environmental impact statement or environmental assessment of the proposed exemption is not required.

Specific details of the environmental considerations supporting this request for exemption are provided in Section 5 of the associated License Amendment Request provided in Enclosure 1U of this letter.

8.0 Conclusion The proposed changes to the COL allow the application of screening criteria to identify Tier 2* departures that require prior NRC approval, followed by application of the existing Tier 2 departure evaluation process to the remaining plant-specific Tier 2* departures that do not meet any of the screening criteria defined in new License Condition 2.D.(13) and departures from Tier 2 information that involve departures from this same Tier 2* matter.

The exemption request meets the requirements of 10 CFR 52.7, Specific exemptions and 10 CFR 50.12, Specific exemptions. Specifically, the exemption request meets the criteria of 10 CFR 50.12(a)(1) in that the request is authorized by law, presents no undue risk to public health and safety, and is consistent with the common defense and security.

Furthermore, approval of this request does not result in a significant decrease in the level of safety, satisfies the underlying purpose of the AP1000 Design Certification Rule, and would not perpetuate undue hardship to the Licensee.

9.0 References.

None.

Page 9 of 9

Southern Nuclear Operating Company ND-18-1006 Enclosure 3U Vogtle Electric Generating Plant (VEGP) Units 3 and 4 Updated Proposed Changes to Licensing Basis Documents (LAR-17-037S4)

Insertions Denoted by Blue Underline and Deletions by Red Strikethrough Omitted text is identified by three asterisks ( * * * )

(This Enclosure consists of four pages, including this cover page.)

ND-18-1006 U Updated Proposed Changes to Licensing Basis Documents (LAR-17-037S4)

Revise Combined License (COL) License Condition 2.D, by adding new condition (13), to address the Tier 2* Change Process, as follows:

D. The license is subject to, and SNC shall comply with, the conditions specified and incorporated below:

(13) Departures from Plant-specific DCD Tier 2* Information (a) SNC is exempt from the requirements of 10 CFR Part 52, Appendix D, Paragraphs VIII.B.6 and VIII.B.5.a for prior NRC approval of departures from Tier 2* information involving a change to or departure from Tier 2* information; except for departures 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,
2. Result in a change to a design process described in the plant-specific DCD that is material to implementation of an industry standard or endorsed regulatory guidance,
3. (i) Result in a change to the fuel criteria evaluation process, the fuel principal design requirements, or the nuclear design of the fuel or the reactivity control system that is material to a fuel or reactivity control system design function, or the evaluation methods in WCAP-12488, Westinghouse Fuel Criteria Evaluation Process, or (ii) Result in any change to the maximum fuel rod average burn-up limits or the small break LOCA analysis methodology described in UFSAR Subsections 15.6.5.4B.2.2 or 15.6.5.4B.2.3,
4. Adversely affect the containment debris limits or debris screen design criteria,
5. Change the Reactor Coolant Pump (RCP) type from a canned motor to a different type of RCP,
6. Result in a change to the Passive Residual Heat Removal Heat Exchanger natural circulation test (first plant test), the Core Makeup Tank Heated Recirculation Tests (first three plants test), or the Automatic Depressurization System Blowdown Test (first three plants test) that is material to the test objectives or test performance criteria,
7. Involve structural materials or analytical or design methods, including design codes and analytical assumptions, that deviate from those credited in the plant-specific DCD for critical sections,
8. Result in a change to the design of the steel faceplates, internal trusses, tie bars, or headed studs of the steel-concrete (SC) module walls in the Page 2 of 4

ND-18-1006 U Updated Proposed Changes to Licensing Basis Documents (LAR-17-037S4)

Nuclear Island or the Shield Building, including SC-to-reinforced concrete (RC) connections,

9. Result in an increase in the demand to capacity (D/C) ratio of a critical section of the structure. SNC shall determine the D/C ratio under this condition for each critical section structural member including, but not limited to, wall segments, wall sections, concrete panels, slabs, or basemat sections, affected by a departure by:

(i) Using the Tier 2* information in the UFSAR Section 3.8 or Appendix 3H table that directly states the D/C ratio or states the area of steel provided and the area of steel required for the affected structural member, or (ii) Providing the same total area of steel across the entire critical section using any combination of rebar sizes and spacing allowed by the design basis codes used in the UFSAR as the total area of steel specified in UFSAR Section 3.8 and Appendix 3H tables marked Tier 2*;

(b) For a departure from Tier 2* information that does not require prior NRC approval under the exemption in License Condition 2.D.(13)(a), SNC may take the departure under and in compliance with the Tier 2 change processes in 10 CFR Part 52, Appendix D, Paragraph VIII.B.5, as modified by the exemption in License Condition 2.D.(13)(a). For each departure authorized by this License Condition:.

1. The departure or change to Tier 2* information shall remain Tier 2*

information in the plant-specific DCD.

2. SNC shall prepare and maintain a written evaluation that provides the bases for its determinations regarding the criteria in License Condition 2.D.(13)(a). In the report that 10 CFR Part 52, Appendix D, Section X.B.1 requires SNC to submit, SNC shall include a brief description of each departure and a summary of the evaluation of the departure.

Page 3 of 4

ND-18-1006 U Updated Proposed Changes to Licensing Basis Documents (LAR-17-037S4)

Revise Updated Final Safety Analysis Report page footers that contain a Tier 2* note as follows:

  • In accordance with the departure evaluation process specified in License Condition 2.D.(13), NRC Staff approval is may be required prior to implementing a change in this information.

Page 4 of 4

Southern Nuclear Operating Company ND-18-1006 Enclosure 4U Vogtle Electric Generating Plant (VEGP) Units 3 and 4 Updated Reviewers Aid Proposed Tier 2* Departure Evaluation Process (LAR-17-037S4)

(This Enclosure consists of five pages, including this cover page.)

ND-18-1006 U Updated Reviewers Aid - Proposed Tier 2* Departure Evaluation Process (LAR-17-037S4)

Proposed Departure Evaluation Process Proposed Tier 2*

Departure Involve design methodology or construction materials that deviate from a code or standard credited in the plant-Yes Qualifying Criterion specific DCD for establishing the criteria for the design or construction of a structure, system, or component (SSC) 1 important to safety No Yes Result in a change to a design process described in the Qualifying Criterion plant-specific DCD that is material to implementation of an 2

industry standard or endorsed regulatory guidance No (i) Result in a change to the fuel criteria evaluation process, the fuel principal design requirements, or nuclear design of fuel and reactivity control system that is material to a fuel or reactivity control system design function or the evaluation methods in WCAP- Yes 12488, Westinghouse Fuel Criteria Evaluation Qualifying Criterion Process, or 3 (ii) result in any change to the maximum fuel rod average burn-up limits or the small break LOCA analysis methodology described in UFSAR Subsections No 15.6.54B.2.2 or 15.6.5.4B.2.3 Adversely affect the containment debris limits or debris Qualifying Criterion Yes screen design criteria 4 No Go to Submit license Go to Qualifying amendment request Criterion 5 (Paragraph VIII.B.6.b or VIII.B.6.c)

Page 2 of 5

ND-18-1006 U Updated Reviewers Aid - Proposed Tier 2* Departure Evaluation Process (LAR-17-037S4)

Proposed Departure Evaluation Process (cont.)

From Qualifying Criterion 4 Yes Change the Reactor Coolant Pump (RCP) type from a Qualifying Criterion canned motor to a different type of RCP 5 No Result in a change to the Passive Residual Heat Removal Heat Exchanger natural circulation test (first plant test), Yes the Core Makeup Tank Heated Recirculation Tests (first Qualifying Criterion three plants test), or the Automatic Depressurization 6 System Blowdown Test (first three plants test) that is material to the test objectives or test performance criteria, No Yes Involve structural materials or analytical or design methods, including design codes and analytical Qualifying Criterion assumptions, that deviate from those credited in the plant- 7 specific DCD for critical sections, No Result in a change to the design of the steel faceplates, Yes internal trusses, tie bars, or headed studs of the steel- Qualifying Criterion concrete (SC) module walls in the Nuclear Island or the 8

Shield Building, including SC-to-reinforced concrete (RC) connections, No Go to Submit license Go to Qualifying amendment request Criterion 9 (Paragraph VIII.B.6.b or VIII.B.6.c)

Page 3 of 5

ND-18-1006 U Updated Reviewers Aid - Proposed Tier 2* Departure Evaluation Process (LAR-17-037S4)

Proposed Departure Evaluation Process (cont.)

From submit license From Qualifying amendment request Criterion 8 (Paragraph VIII.B.6.b or VIII.B.6.c)

Result in an increase in the demand to capacity (D/C) of a critical section of the structure. SNC shall determine the D/C ratio under this condition for each critical section structural member including, but not limited to, wall segments, wall sections, concrete panels, slabs, or basement sections, affected by a departure by:

(i) Using the Tier 2* information in the UFSAR Yes Submit license Section 3.8 or Appendix 3H table that directly states the D/C ratio or states the are of steel Qualifying Criterion amendment request provided and the area of steel required for the 9 (Paragraph VIII.B.6.b or affected structural member, or VIII.B.6.c)

(ii) Providing the same total area of steel across the entire critical section using any combination of No rebar sizes and spacing allowed by the design basis codes used in the UFSAR as the total area of steel specified in UFSAR Section 3.8 and Appendix 3H tables marked as Tier 2*;

Go to Paragraph VIII.B.5.a Page 4 of 5

ND-18-1006 U Updated Reviewers Aid - Proposed Tier 2* Departure Evaluation Process (LAR-17-037S4)

Proposed Departure Evaluation Process (cont.)

From go to Paragraph VIII.B.5.a Yes Involve Tier 1 or TS (Paragraph VIII.B.5.a)

No More than minimal Yes impact (Paragraph VIII.B.5.b)

No Substantial increase Yes Submit license Ex-vessel severe amendment request accident (Paragraph VIII.B.5.e)

(Paragraph VIII.B.5.c)

No Yes No Aircraft impact Document compliance Implement assessment affected with 10 CFR 50.150 departure (Paragraph VIII.B.5.d)

Page 5 of 5

Southern Nuclear Operating Company ND-18-1006 Enclosure 5U Vogtle Electric Generating Plant (VEGP) Units 3 and 4 Updated Reviewers Aid Tier 2* Matters Analysis Summary (LAR-17-037S4)

(This Enclosure consists of seven pages, including this cover page.)

ND-18-1006 U Updated Reviewers Aid - Tier 2* Matters Analysis Summary (LAR-17-037S4)

Tier 2* Analysis ResultsSection VIII.B.6.b Selected for (Tier 2* Matters that additional Do Not Expire at Full Power) screening Basis Associated Criteria 1 Maximum fuel rod average burn- Yes Not addressed in Result in any change to the maximum fuel rod up. Tier 1 average burn-up limits; 2 Fuel principal design Yes Not addressed in Result in a change to the fuel principal design requirements. Tier 1 requirements, that is material to the fuel or reactivity control system design function 3 Fuel criteria evaluation process. Yes Not addressed in Result in a change to the fuel criteria evaluation Tier 1 process, that is material to the fuel or reactivity control system design function or the evaluation methods in WCAP-12488, Westinghouse Fuel Criteria Evaluation Process, 4 Fire areas. N/A Previous exemption N/A re-designated VEGP 3 and 4 fire area figures as Tier 2 5 Reactor coolant pump type. Yes Not addressed in Change the Reactor Coolant Pump (RCP) type from Tier 1 a canned motor to a different type of RCP.

6 Small-break loss-of-coolant Yes Safety significance Result in any change to small break LOCA accident (LOCA) analysis analysis methodology described in UFSAR methodology. Subsections 15.6.5.4B.2.2 or 15.6.5.4B.2.3.

7 Screen design criteria. Yes Paragraph VIII.B.5 Adversely affect the containment debris limits or may not work well in debris screen design criteria.

Page 2 of 7

ND-18-1006 U Updated Reviewers Aid - Tier 2* Matters Analysis Summary (LAR-17-037S4)

Section VIII.B.6.b Selected for (Tier 2* Matters that additional Do Not Expire at Full Power) screening Basis Associated Criteria all cases; safety significance 8 Heat sink data for containment No Adequately N/A pressure analysis. addressed by paragraph VIII.B.5 Section VIII.B.6.c Selected for (Tier 2* Matters that additional Expire at Full Power) screening Basis Associated Criteria 1 Nuclear Island Structural No Adequately addressed in N/A Dimensions. Tier 1 2 American Society of Mechanical No Adequately addressed in N/A Engineers Boiler & Pressure Tier 1; 10 CFR 50.55a; Vessel Code (ASME Code) Paragraph VIII.B.5 piping design and welding restrictions, and ASME Code Cases.

Page 3 of 7

ND-18-1006 U Updated Reviewers Aid - Tier 2* Matters Analysis Summary (LAR-17-037S4) 3 Design Summary of Critical Yes Safety significance Involve structural materials or analytical or Sections. design methods, including design codes and analytical assumptions, that deviate from those credited in the plant-specific DCD for critical sections, or Result in a change to the design of the steel faceplates, internal trusses, tie bars, or headed studs of the steel-concrete (SC) module walls in the Nuclear Island or the Shield Building, including SC-to-reinforced concrete (RC) connections, or Result in an increase in the demand to capacity (D/C) ratio of a critical section of the structure.

SNC shall determine the D/C ratio under this condition for each critical section structural member including, but not limited to, wall segments, wall sections, concrete panels, slabs, or basemat sections, affected by a departure by:

(i) Using the Tier 2* information in the UFSAR Section 3.8 or Appendix 3H table that directly states the D/C ratio or states the area of steel provided and the area of steel required for the affected structural member, or (ii) Providing the same total area of steel across the entire critical section using any combination of rebar sizes and spacing allowed by the design basis codes used in the UFSAR as the total area of steel specified in UFSAR Section 3.8 and Appendix 3H tables marked Tier 2*;

Page 4 of 7

ND-18-1006 U Updated Reviewers Aid - Tier 2* Matters Analysis Summary (LAR-17-037S4)

Section VIII.B.6.c Selected for (Tier 2* Matters that additional Expire at Full Power) screening Basis Associated Criteria 4 American Concrete Institute Yes Safety significance Involve design methodology or construction (ACI) 318, ACI 349, American materials that deviate from a code or standard National Standards credited in the plant-specific DCD for Institute/American Institute of establishing the criteria for the design or Steel Construction construction of a structure, system, or (ANSI/AISC)-690, and component (SSC) important to safety.

American Iron and Steel Institute (AISI), Specification for the Design of Cold Formed Steel Structural Members, Part 1 and 2, 1996 Edition and 2000 Supplement.

5 Definition of critical locations No Adequately addressed in N/A and thicknesses. Tier 1 6 Seismic qualification methods No Adequately addressed N/A and standards. by paragraph VIII.B.5 7 Nuclear design of fuel and Yes Not addressed in Tier 1 Result in a change to the nuclear design of reactivity control system, except and safety significance fuel and reactivity control system that is material burn-up limit. to a fuel or reactivity control system design function; 8 Motor-operated and power- No Adequately addressed in N/A operated valves. Tier 1 and by paragraph VIII.B.5 Page 5 of 7

ND-18-1006 U Updated Reviewers Aid - Tier 2* Matters Analysis Summary (LAR-17-037S4)

Section VIII.B.6.c Selected for (Tier 2* Matters that additional Expire at Full Power) screening Basis Associated Criteria 9 Instrumentation and control Yes Safety Significance Result in a change to a design process system design processes, described in the plant-specific DCD that is methods, and standards. material to implementation of an industry standard or endorsed regulatory guidance.

10 Passive residual heat removal Yes Not addressed in Tier 1 Results in a change to the Passive Residual (PRHR) natural circulation test Heat Removal Heat Exchanger natural (first plant only). circulation test (first plant test)that is material to the test objectives or test performance criteria 11 Automatic depressurization Yes Not addressed in Tier 1 Results in a change to the Core Makeup system (ADS) and core make- Tank Heated Recirculation Tests (first three up tank (CMT) verification tests plants test), or the Automatic Depressurization (first three plants only). System Blowdown Test (first three plants test) that is material to the test objectives or test performance criteria, 12 Polar crane parked orientation. No Does not meet criteria for N/A Tier 1; therefore, paragraph VIII.B.5 will adequately address 13 Piping design acceptance Yes Safety Significance Result in a change to a design process criteria. described in the plant-specific DCD that is material to implementation of an industry standard or endorsed regulatory guidance.

Page 6 of 7

ND-18-1006 U Updated Reviewers Aid - Tier 2* Matters Analysis Summary (LAR-17-037S4)

Section VIII.B.6.c Selected for (Tier 2* Matters that additional Expire at Full Power) screening Basis Associated Criteria 14 Containment vessel design No Adequately addressed in N/A parameters, including ASME Tier 1 and paragraph Code,Section III, Subsection VIII.B.5 NE.

15 Human factors engineering. Yes Paragraph VIII.B.5 may Result in a change to a design process not work well and safety described in the plant-specific DCD that is significance material to implementation of an industry standard or endorsed regulatory guidance.

16 Steel composite structural Yes Safety significance Involve design methodology or construction module details. 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.

Page 7 of 7

Southern Nuclear Operating Company ND-18-1006 Enclosure 6U Vogtle Electric Generating Plant (VEGP) Units 3 and 4 Updated Reviewers Aid Tier 2* Departure Example Not Requiring Prior NRC Approval (LAR-17-037S4)

(This Enclosure consists of three pages, including this cover page.)

ND-18-1006 U Updated Reviewers Aid - Tier 2* Departure Example Not Requiring Prior NRC Approval (LAR-17-037S4)

Example application of the LAR-17-037 proposed process to a proposed change to Tier 2*

material which results in a determination that prior NRC approval is not required:

In Vogtle 3&4 LAR-13-006R [ML13240A217], SNC proposed the following change in Enclosure 1, page 3 of 15, Summary

Description:

The proposed changes in the requirements for detailed design of structural wall modules used to construct containment internal structures and portions of the auxiliary building are necessary to address regulatory compliance for design of shear studs and internal trusses.

The proposed changes would depart from plant-specific Design Control Document (DCD) Tier 2* and associated Tier 2 material incorporated into the Updated Final Safety Analysis Report (UFSAR) by revising requirements for design spacing of shear studs and wall module trusses and the design of structural elements of the trusses such as angles and channels. These revisions are to address interferences and obstructions that may cause a change to the design spacing in a local area. In each case where the spacing exceeds the design spacing, an evaluation supporting the increase will be completed to demonstrate that the revised spacing is in conformance with design and analysis requirements identified in the UFSAR. The designation of maximum design spacing is revised for the stud spacing and truss spacing to reduce the potential confusion about the application of fabrication tolerances.

The proposed changes include revising a note on UFSAR Figure 3.8.3-8, Sheet 1 to clarify that the stud spacing specified is a design value not an exact dimension.

A tolerance for stud spacing, consistent with American Welding Society (AWS)

D1.1 requirements, is added to the note.

The proposed changes include revision of the weld symbol on a Tier 2* figure to change the symbol to a symbol that indicates complete joint penetration and change to the associated Tier 2 text to clarify that the weld symbol used in the figure indicates complete joint penetration.

Consider the proposed process flow chart provided in Enclosure 4U of this LAR, with revising requirements for design spacing of shear studs and wall module trusses as the Proposed Tier 2* Departure entry oval. The details of the proposed changes to spacing of shear studs and wall module trusses are found in LAR-13-006 Enclosure 1, Sections 2 and 3.

Criterion 1 asks, Yes/No, does the Proposed Tier 2* Departure 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.

In this example, the answer to this question would be No. For this proposed change to Tier 2* material, SNC requested flexibility in the placement of studs and trusses within the bounds of the AP1000 DCD-endorsed ACI 349-01, AISC N690-94, and AWS D1.1 standards as described by UFSAR 3.8.3 et. al. The requested change eliminated potential confusion by deleting the adjective maximum from Tier 2* text and Figures describing Page 2 of 3

ND-18-1006 U Updated Reviewers Aid - Tier 2* Departure Example Not Requiring Prior NRC Approval (LAR-17-037S4) the design spacing of the shear studs and trusses inside the steel-concrete composite modules. No deviation from a code or standard was requested.

Of particular note is that the published NRC SER for LAR-13-006 [ML13266A164]

concurred that no deviation from a code or standard was requested by this proposed change and that the proposed changes were within the scope and technical requirements of the applicable codes, standards, and design methodology of the steel-concrete composite structures as described by the AP1000 DCD and Vogtle UFSAR.

Qualifying Criterion 2 asks if the proposed Tier 2* departure will, Result in a material change to a design process described in the plant-specific DCD that is used to implement an industry standard or endorsed regulatory guidance. Since no change was proposed to a design process addressed by Criterion 2 this Criterion does not apply.

Qualifying Criterion 3 and 4 pose questions regarding impacts to nuclear fuel and the in-containment refueling water storage tank (IRWST) screen design and containment recirculation screen design; neither of those topics are affected by the proposed Tier 2*

change. Thus, the answer to both questions is No and the process shown in LAR-17-037 continues to satisfy 10 CFR 52 Appendix D, Paragraph VIII.B.5.a.

LAR-17-037 Enclosure 4U, page 3 of 3 steps the user through the 10 CFR 52 Appendix D, Paragraph VIII.B.5.a and subsequent questions.

First, the LAR-13-006 proposed Tier 2* change does not involve Tier 1 material nor the Vogtle Technical Specifications (TS); thus, the answer to this question is No and the screening continues.

Second, the LAR-13-006 proposed Tier 2* change does not prompt a Yes answer to any of the eight questions posed in VIII.B.5.b regarding the increased likelihood of an accident, increased effects of accidents, or creation of a new type of accident; thus, the answer to this question is No and the screening continues.

Third, the LAR-13-006 proposed Tier 2* change does not prompt a Yes answer to either of the two questions posed in VIII.B.5.c regarding ex-vessel accidents; thus, the answer to this question is No and the screening continues.

Fourth, the LAR-13-006 proposed Tier 2* change does not prompt a Yes answer to the questions posed in VIII.B.5.d regarding the Aircraft Impact Assessment; thus, the answer to this question is No and the screening continues.

As a result of the above screening, the Enclosure 4U flow chart directs the user to implement the departure. The departure is then implemented in accordance with the applicable 10 CFR 50 Appendix B procedures and processes and reported accordingly.

In this example, the LAR-17-037 proposed screening process challenged the proposed Tier 2* change appropriately and came to the correct conclusion that nuclear safety, regulatory compliance, and the public health and safety would not be adversely impacted by the licensee implementing the proposed change without requiring prior NRC approval.

Page 3 of 3

Southern Nuclear Operating Company ND-18-1006 Enclosure 7U Vogtle Electric Generating Plant (VEGP) Units 3 and 4 Updated Reviewers Aid Tier 2* Departure Example Requiring Prior NRC Approval (LAR-17-037S4)

(This Enclosure consists of two pages, including this cover page.)

ND-18-1006 U Updated Reviewers Aid - Tier 2* Departure Example Requiring Prior NRC Approval (LAR-17-037S4)

Example application of the LAR-17-037 proposed process to a proposed change to Tier 2*

material which would result in a determination that prior NRC approval is required:

In Vogtle 3&4 LAR-13-004 [ML13022A254], SNC proposed the following change in Enclosure 1, page 2 of 10, Summary

Description:

The proposed changes would depart from plant-specific Design Control Document (DCD) Tier 2* and associated Tier 2 material incorporated into the Updated Final Safety Analysis Report (UFSAR) by revising the structural analysis requirements to provide alternative requirements for development of shear reinforcement bars within the nuclear island basemat concrete.

The proposed changes revise the requirements for development of basemat shear reinforcement in the licensing basis from ACI 349 Appendix B to ACI 318-11, Section 12.6. The use of ACI 318 criteria for headed reinforcement results in longer shear ties and thicker concrete in areas below the elevator pits and a sump in the nuclear island basemat. The thicker concrete is accomplished by raising the floor of the elevator pits and sump in the nuclear island basemat resulting in a minor reduction in volume of the sump. The requirements for concrete cover over the reinforcement bars are also changed.

This enclosure requests approval of the license amendment necessary to implement the proposed changes to the Tier 2* and associated Tier 2 material.

Consider the proposed process flow chart provided in Enclosure 4U of this LAR, with revise the requirements for development of basemat shear reinforcement in the licensing basis from ACI 349 Appendix B to ACI 318-11, Section 12.6 as the Proposed Tier 2*

Departure entry oval.

Criterion 7 asks, Yes/No, does the Proposed Tier 2* Departure Involve structural materials or analytical or design methods, including design codes and analytical assumptions, that deviate from those credited in the plant-specific DCD for critical sections.

In this example, the answer to this question would be Yes. For this proposed change to Tier 2* material, SNC requested to deviate from the use of the ACI 349-01 concrete code and instead use ACI 318-11 for the design of headed shear reinforcement in the Nuclear Island basemat. ACI 349-01 was endorsed for use in the design and construction of concrete structures by the NRC in the approval of the AP1000 Design Certification.

Conversely, ACI 318-11 was not utilized in the AP1000 Design Certification nor is ACI 318-11 generically endorsed for the proposed use by the NRC via other regulatory means such as an issued Regulatory Guide. In short, the proposed Tier 2* change deviates from the design method that was credited in the plant-specific DCD for the design and construction of an SSC important to safety.

While other components of the LAR-13-004 proposed change, such as a minor change to the floor elevation of the bottom of an elevator pit, may not require prior NRC approval under the LAR-17-037 change process, the deviation from the approved code would be properly categorized by Criterion 1 which immediately directs the user to Submit license amendment request (Paragraph VIII.B.6.b or VIII.B.6.c).

Page 2 of 2

Southern Nuclear Operating Company ND-18-1006 Enclosure 8U Vogtle Electric Generating Plant (VEGP) Units 3 and 4 Updated Proposed Regulatory Commitment (LAR-17-037S4)

(This Enclosure consists of two pages, including this cover page.)

ND-18-1006 U Updated Proposed Regulatory Commitment (LAR-17-037S4)

The following table identifies the regulatory commitment in this document. Any other statements in this submittal represent intended or planned actions. They are provided for information purposes and are not considered to be regulatory commitments.

REGULATORY COMMITMENT DUE DATE / EVENT Develop, implement, and maintain procedural guidance that Implemented prior to the contains a description of the qualifying criteria contained in implementation of the License Condition 2.D(13) and the supporting detailed guidance license amendment and bases contained in the Technical Evaluation section of the approving this LAR approved LAR-17-037, including additional guidance provided by SNC in the supplements to the LAR. This procedural guidance will be maintained in accordance with SNCs Commitments Management Program for as long as the license condition remains in effect.

Page 2 of 2

Southern Nuclear Operating Company ND-18-1006 Enclosure 18 Vogtle Electric Generating Plant (VEGP) Units 3 and 4 Response to NRC Request for Additional Information (RAI) LAR 17-037-2 Regarding the LAR-17-037 Review (LAR-17-037S4)

Supplement 4 changes to the original LAR text are shown as blue-underlined text; deletions of original LAR text are shown as red strikethrough text.

(This Enclosure consists of eleven pages, including this cover page.)

ND-18-1006 8 Response to NRC Request for Additional Information (RAI) LAR 17-037-2 Regarding the LAR-17-037 Review (LAR-17-037S4)

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 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.

Page 2 of 11

ND-18-1006 8 Response to NRC Request for Additional Information (RAI) LAR 17-037-2 Regarding the LAR 17-037 Review (LAR-17-037S4)

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:

  • Critical sections using steel concrete sandwich construction, and
  • 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 Tier 2* information is intended to have substantial safety significance, commensurate with information designated as Tier 1. However, 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 of the change rises 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 steel-concrete composite regions of the nuclear island and 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.

Nevertheless, SNC has identified enhancements to the critical section criterion that expand the scope of the initially proposed critical sections criterion to address changes to structural materials or analytical or design methods, changes to steel-concrete (SC) module walls in the Nuclear Island or the Shield Building, and changes to reinforcement steel in critical sections.

Page 3 of 11

ND-18-1006 8 Response to NRC Request for Additional Information (RAI) LAR 17-037-2 Regarding the LAR 17-037 Review (LAR-17-037S4)

Structural Design Change Process Westinghouse is the design authority for the Vogtle 3&4 AP1000 units; conducting such activities under the detailed supervision of SNC as the Licensee. Any change to the design of Vogtle 3&4, regardless of its scope or potential impact, is subjected to a rigorous process(es) under 10 CFR 50 Appendix B Quality Assurance Program controls regarding both technical and regulatory aspects.

Proposed changes may enter the design control and licensing control processes in several ways; the most common are field-identified changes and those changes identified in advance during work planning (including by use of clash-identifying modeling software), procurement, and impact reviews from other changes. Those deviations from Tier 2* details of critical sections discovered in the field are documented per the site 10 CFR 50 Appendix B design control and corrective action programs, dispositioned via the design control and licensing control processes, and tracked for consideration in future changes and the as-built reconciliation report. Changes may also be initiated via field change requests and design change requests when a challenge is identified prior to field construction.

Regardless of the design control process entry method or originating location (i.e. at Vogtle site or offsite), proposed design changes proceed through detailed technical and regulatory reviews.

WEC Engineering performs a technical review of the proposed change to first determine whether applicable design requirements (e.g., codes, seismic analysis) are satisfied. This effort includes evaluating the stress at the proposed change location, the impact of prior nearby design changes on the proposed change, and the impact of the proposed change on the structural member (and by extension, entire Nuclear Island) in consideration of the overall cumulative impact of the proposed and prior design changes. This review of impacts both locally and cumulatively occurs with each proposed change, thus precluding a finding at the end of the construction effort that the structure no longer conforms to the approved design or will not withstand the design basis loads specified in the Design Description without loss of structural integrity or the safety related functions. In close coordination, the proposed change is evaluated for regulatory impacts and the appropriate licensing change process entered as necessary. The license conditions proposed by this LAR ensure that key critical section structural attributes of the Vogtle units are maintained and that adverse changes to those attributes undergo prior NRC review and approval. Examples of non-adverse changes include minor deviations in rebar spacing or changes in weld details which dont materially alter the load distribution through the structure; adverse changes affect the strength, stiffness, ductility, or dynamic response of an element.

If a proposed design change is found to be acceptable technically and has received the proper regulatory approvals, the design change is implemented through updates to drawings and associated documents for field construction. Within the design control software program employed by Westinghouse, Intergraph SmartPlant, changes to the design are tracked against the relevant drawings, calculations, and computer models. This linking and updating of changes to the design documents both ensures that future changes to the same structural members consider the implemented changes, and that when the as-built design reconciliation report (see Tier 1 Table 3.3-6 Design Commitment 2a) is prepared such changes to the COL-approved design are considered.

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

Page 4 of 11

ND-18-1006 8 Response to NRC Request for Additional Information (RAI) LAR 17-037-2 Regarding the LAR 17-037 Review (LAR-17-037S4)

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

  • 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 3.8.4.5.1 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 3.8.4.5.2 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).
  • UFSAR Subsection 3H.5.1, Shear Walls, states that the wall sections are designed in accordance with the requirements of ACI 349-01.
  • 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.
  • 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.
  • 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.

These codes provide requirements for design and construction of reinforced concrete structures, including allowable capacity aspects such as the demand-to-capacity ratio of walls and floors using reinforced concrete. Nevertheless, to address the NRC staffs concern regarding the ability to change the demand-to-capacity ratios for reinforcement in critical sections, SNC proposes to add restrictions to the critical sections criterion regarding changes to these ratios as described below.

Potential changes in reinforcement that need to be accommodated due to construction are generally localized (e.g., a missing reinforcing bar due to an interference) and are readily accommodated within design basis code commitments. Changes that affect the overall capacity of a structural member (e.g., changing the typical size and spacing of the reinforcement in a reinforced concrete wall) are not anticipated, and NRC staff prior review and approval of such a Page 5 of 11

ND-18-1006 8 Response to NRC Request for Additional Information (RAI) LAR 17-037-2 Regarding the LAR 17-037 Review (LAR-17-037S4) change aligns with the spirit of the Tier 2* designation. Therefore, it is proposed to add language to exclude from the critical sections criterion changes in steel area for structural members provided that the change satisfies code requirements and the total area of steel across the entire critical section remains within the values described in UFSAR Section 3.8 or Appendix 3H Tier 2*

tables (e.g., area of reinforcing steel in a wall face).

The term structural member as used in the critical sections criterion refers to a segment of a nuclear island structure including, but not limited to, wall segments, wall sections, concrete panels, slabs, or basemat sections. For example, UFSAR Tier 2* Table 3H.5-3, describes one segment as Wall Section 1, 6, Elevation 180-0 to 135-3, and defines the required and minimum provided area of steel for that segment/structural member.

An example of a common situation during construction is when a reinforcing bar is moved or cannot be placed due to interference. To address this situation and ensure that the critical section structure continues to perform as expected, the design authority may choose to add reinforcement at a nearby location. Thus, the design authority examines the ratio of required reinforcement to provided reinforcement using the total area of steel across the entire critical section using any combination of rebar sizes and spacing allowed by the design basis codes used in the UFSAR as the total area of steel specified in UFSAR Section 3.8 and Appendix 3H Tables marked Tier 2*,

and verifies that the resulting ratio does not exceed that ratio value stated in or calculated from UFSAR Section 3.8 and Appendix 3H Tables Tier 2* values for the applicable structural member.

Provided that all Code requirements are satisfied and the demand-to-capacity ratio of the member based on the applicable Tier 2* values are not exceeded, then the change would qualify for the application of the Tier 2 change process.

Accordingly, SNC proposes to add a new criterion for critical sections that addresses RC demand-to-capacity ratios and requires that a departure obtain prior NRC review and approval when the change:

Results in an increase in the demand to capacity (D/C) ratio of a critical section of the structure. SNC shall determine the D/C ratio under this condition for each critical section structural member including, but not limited to, wall segments, wall sections, concrete panels, slabs, or basemat sections, affected by a departure by:

i. Using the Tier 2* information in the UFSAR Section 3.8 or Appendix 3H table that directly states the D/C ratio or states the area of steel provided and the area of steel required for the affected structural member, or ii. Providing the same total area of steel across the entire critical section using any combination of rebar sizes and spacing allowed by the design basis codes used in the UFSAR as the total area of steel specified in UFSAR Section 3.8 and Appendix 3H tables marked Tier 2*.

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

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ND-18-1006 8 Response to NRC Request for Additional Information (RAI) LAR 17-037-2 Regarding the LAR 17-037 Review (LAR-17-037S4)

  • 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.

  • 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).
  • 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.
  • 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.

In recognition of the nonlinear analysis and testing that was performed to support the design and licensing of the SC modules employed in the AP1000 design, the critical section criterion is modified to require prior NRC approval for any changes to the design of the steel faceplates, internal trusses, tie bars, or headed studs of the steel-concrete (SC) module walls in the Nuclear Island or the Shield Building including SC-to-reinforced concrete (RC) connections.

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:

  • knuckle region of the roof near the PCCWST wall
  • compression ring
  • PCCWST Page 7 of 11

ND-18-1006 8 Response to NRC Request for Additional Information (RAI) LAR 17-037-2 Regarding the LAR 17-037 Review (LAR-17-037S4)

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:

  • the steel roof that supports the concrete roof slab
  • tension ring
  • 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 critical sections criterion, which addresses deviations from design methodology including 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.

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. Note that the UFSAR specifically identifies the structural steel rods between shield building SC faces as tie bars, not traditional deformed Page 8 of 11

ND-18-1006 8 Response to NRC Request for Additional Information (RAI) LAR 17-037-2 Regarding the LAR 17-037 Review (LAR-17-037S4) reinforcing bar performing common reinforcing functions, thus discussions of treatment of reinforcing bars should not be directly applied to those tie bars.

SNC recognizes that due to the unique non-symmetries of the shield building with regard to its location within the nuclear island footprint and the varying elevation of the SC-to-RC connection around its circumference, that great diligence must be applied to any proposed change.

Therefore, the critical sections criterion is revised to require that any changes to the design of the steel faceplates, internal trusses, tie bars, or headed studs of the SC module walls in the Nuclear Island or the Shield Building including SC-to-reinforced concrete (RC) connections shall require prior NRC review and approval.

By including these additional shield building requirements within the scope of the proposed critical sections criterion, 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.

Changes to Original LAR-17-037 in response to RAI LAR-17-037-2 Changes to Enclosure 1:

Revise the Criterion detailed guidance as follows:

Add new Criteria 7, 8, and 9 (critical sections) detailed guidance The design of critical sections is important in assuring the integrity of the buildings which house safety related systems and components. The proposed Criteria 7, 8, and 9 rely on a combination of requirements to ensure that safety significant changes to critical section design will require prior NRC review and approval. These requirements address changes critical section structural materials or analytical or design methods, including design codes and analytical assumptions, changes to critical sections that use steel concrete sandwich construction, and increases in demand to capacity ratios of critical sections using reinforced concrete. Furthermore, changes to the design of the steel faceplates, internal trusses, tie bars, or headed studs of the steel-concrete (SC) module walls in the Nuclear Island or the Shield Building, including SC-to-reinforced concrete (RC) connections will require prior NRC review and approval.

Revise the critical sections Criteria Bases discussion by adding the following paragraphs after the existing paragraph:

It is noted that the UFSAR provides requirements regarding the structural materials and, analytical and design methods for critical sections including design codes and analytical assumptions. Changes to these requirements are included in this Criterion and prior NRC approval is required if deviations are proposed to these requirements.

Furthermore, the critical sections criterion contains restrictions regarding the ability to increase demand-to-capacity ratios for reinforcement in critical sections without prior NRC staff approval. The Criterion requires prior NRC approval for any increase in the demand to capacity (D/C) ratio of a critical section of the structure beyond the ratio specified in or Page 9 of 11

ND-18-1006 8 Response to NRC Request for Additional Information (RAI) LAR 17-037-2 Regarding the LAR 17-037 Review (LAR-17-037S4) calculated from the applicable UFSAR Section 3.8 or Appendix 3H Table marked Tier 2*.

SNC shall determine the D/C ratio under this condition for each critical section structural member affected by a departure by either:

  • Using the Tier 2* information in the UFSAR Section 3.8 or Appendix 3H Table that directly states the D/C ratio or states the area of steel provided and the area of steel required for the affected structural member, or
  • Providing the same total area of steel across the entire critical section using any combination of rebar sizes and spacing allowed by the design basis codes used in the UFSAR as the total area of steel specified in UFSAR Section 3.8 and Appendix 3H Tables marked Tier 2*.

The term structural member as used in the critical sections criterion refers to a segment of a nuclear island structure including, but not limited to, wall segments, wall sections, concrete panels, slabs, or basemat sections. For example, UFSAR Tier 2* Table 3H.5-3, describes one segment as Wall Section 1, 6, Elevation 180-0 to 135-3, and defines the required and minimum provided area of steel for that segment/structural member.

An example of a common situation during construction is when a reinforcing bar is moved or cannot be placed due to interference. To address this situation and ensure that the critical section structure continues to perform as expected, the design authority may choose to add reinforcement at a nearby location. Thus, the design authority examines the ratio of required reinforcement to provided reinforcement using the total area of steel across the entire critical section using any combination of rebar sizes and spacing allowed by the design basis codes used in the UFSAR as the total area of steel specified in UFSAR Section 3.8 and Appendix 3H tables marked Tier 2*, and verifies that the resulting ratio does not exceed that ratio value stated in or calculated from UFSAR Section 3.8 and Appendix 3H tables Tier 2* values for the applicable structural member. Provided that all Code requirements are satisfied and the demand-to-capacity ratio of the member based on the applicable Tier 2* values are not exceeded, then the change would qualify for the application of the Tier 2 change process.

Potential changes in reinforcement that need to be accommodated due to construction are generally localized (e.g., a missing reinforcing bar due to an interference) and are readily accommodated within design basis code commitments. Changes that affect the overall capacity of a structural member (e.g., changing the overall size and spacing of the reinforcement in a reinforced concrete wall) are not anticipated, and NRC staff review and approval of such a change is required.

Additionally, due to the unique non-symmetries of the shield building with regard to its location within the nuclear island footprint and the varying elevation of the SC-to-RC connection around its circumference, no changes are permitted to the steel faceplates, internal trusses, tie bars, or headed studs of SC module walls in the Nuclear Island or Page 10 of 11

ND-18-1006 8 Response to NRC Request for Additional Information (RAI) LAR 17-037-2 Regarding the LAR 17-037 Review (LAR-17-037S4)

Shield Building, or Shield Building SC-to-RC connections without prior NRC staff review and approval. Proposed local changes such as attachment of miscellaneous low energy small bore piping hanger steel to SC modules would typically be acceptable to perform under the Tier 2 change process. Other proposed changes such as systematically eliminating tie bars in the shield building SC modules would require prior NRC approval.

Changes to Enclosure 3:

To address the concerns addressed by the NRC staff in the question, SNC agrees to revise the proposed evaluation Criterion for critical sections to address additional requirements. Proposed Criteria 7, 8 and 9 are as follows:

7. Involve structural materials or analytical or design methods, including design codes and analytical assumptions, that deviate from those credited in the plant-specific DCD for critical sections,
8. Result in a change to the design of the steel faceplates, internal trusses, tie bars, or headed studs of the steel-concrete (SC) module walls in the Nuclear Island or the Shield Building, including SC-to-reinforced concrete (RC) connections,
9. Result in an increase in the demand to capacity (D/C) ratio of a critical section of the structure. SNC shall determine the D/C ratio under this condition for each critical section structural member including, but not limited to, wall segments, wall sections, concrete panels, slabs, or basemat sections affected by a departure by:
i. Using the Tier 2* information in the UFSAR Section 3.8 or Appendix 3H table that directly states the D/C ratio or states the area of steel provided and the area of steel required for the affected structural member, or ii. Providing the same total area of steel across the entire critical section using any combination of rebar sizes and spacing allowed by the design basis codes used in the UFSAR as the total area of steel specified in UFSAR Section 3.8 and Appendix 3H tables marked Tier 2*.

Similar conforming changes are also applicable to Enclosures 1, 4, 5 and 7 of the original LAR 17-037.

Page 11 of 11

Southern Nuclear Operating Company ND-18-1006 Enclosure 19 Vogtle Electric Generating Plant (VEGP) Units 3 and 4 Response to NRC Question on Evaluation of Departures from Tier 2* Information Regarding First-of-a-Kind Testing (LAR-17-037S4)

Supplement 4 changes to the original LAR text are shown as blue-underlined text; deletions of original LAR text are shown as red strikethrough text.

(This Enclosure consists of seven pages, including this cover page.)

ND-18-1006 9 Response to NRC Question on Evaluation of Departures from Tier 2* Information Regarding First-of-a-Kind Testing (LAR-17-037S4)

The following is a question provided by the NRC Staff [email dated June 29, 2018] 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 In Vogtle Electric Generating Plants (VGEPs) Units 3 and 4 Final Safety Analysis Report (FSAR),

Chapter 14, Initial Test Program, Revision 6, it states that the overall objective of the initial test program is to demonstrate that the plant has been constructed as designed, that the systems perform consistent with the plant design, and that activities culminating in operation at full licensed power including initial fuel load, initial criticality, and power ascension are performed in a controlled and safe manner.

Special tests to further establish a unique phenomenological performance parameter of the AP1000 design features beyond testing performed for Design Certification of the AP600 and that will not change from plant to plant, are performed for the first plant only. Because of the standardization of the AP1000 design, these special tests (designated as first plant only tests) are not required on follow plants. These first plant only tests are identified in the individual test descriptions (See Subsections 14.2.9 and 14.2.10 of the VEGP 3&4 FSAR). The following is a listing of the first plant only tests, and the corresponding section in which they appear:

First Plant Only Test Section IRWST Heatup Test 14.2.9.1.3 Item (h)

Pressurizer Surge Line Stratification Evaluation 14.2.9.1.7 Item (d)

Reactor Vessel Internals Vibration Testing 14.2.9.1.9 - Prototype Test

[Natural Circulation Tests]* 14.2.10.3.6, [14.2.10.3.7]*

Rod Cluster Control Assembly Out of Bank Measurements 14.2.10.4.6 Load Follow Demonstration 14.2.10.4.22 Other special tests which further establish a unique phenomenological performance parameter of the AP1000 design features beyond testing performed for Design Certification for the AP600 and that will not change from plant to plant, are performed for the first three plants. Because of the standardization of the AP1000 design, once these special tests have affirmed consistent passive system function they are not required on follow plants. These tests required on the first three plants are identified in the individual test descriptions (See Subsection 14.2.9). The following is a listing of the tests required on the first three plants, and the corresponding section in which they appear.

First Three Plant Tests Section Core Makeup Tank Heated Recirculation Tests 14.2.9.1.3 Items (k) and (w)

ADS Blowdown Test 14.2.9.1.3 Item (s)

These tests are Tier 2*, as described in the VEGP 3&4 FSAR.

Page 2 of 7

ND-18-1006 9 Response to NRC Question on Evaluation of Departures from Tier 2* Information Regarding First of a Kind Testing (LAR-17-037S4)

In letter dated December 21, 2017, Southern Nuclear Operating Company Vogtle Electric Generating Plant Units 3 and 4 Request for License Amendment and Exemption: Changes to Tier 2* Departure Evaluation Process (LAR17037), the licensee requests to apply the existing departure evaluation process for Tier 2 departures in the evaluation of certain Tier 2* departures.

This would entail an exemption from certain change requirements in 10 CFR Part 52, Appendix D, to allow departures form Tier 2* information evaluated against criteria proposed in a new License Condition without a license amendment by expanding the applicability of the existing Tier 2 evaluation process.

The staff did not identify in the VEGP 3&4 FSAR chapter, Tier 1 Revision 5, any Tier 1 Inspection, Tests, Analyses and Acceptance Criteria (ITAAC), with the exception of the natural circulation test, that address the first plant only and third plant only tests described to be performed in accordance with Tier 2.

The staff is concerned that SNC would be able to make certain changes to the Tier 2* test descriptions without NRC approval. In particular, where a change to a Tier 2* test description would influence the outcome of the test such that it would affect whether the corresponding Tier 1 acceptance criteria would be met, the change should require NRC approval.

The staff would like SNC to discuss either revising SNCs proposed screening criteria license condition to assure that such a change would come to NRC for approval, or explaining why it is not necessary.

SNC Response to NRC Question regarding Initial Test Program:

SNC proposes to add a sixth non-qualifying criterion that reads:

6. Result in a change to the Passive Residual Heat Removal Heat Exchanger natural circulation test (first plant test), the Core Makeup Tank Heated Recirculation Tests (first three plants test), or the Automatic Depressurization System Blowdown Test (first three plants test) that is material to the test objectives or test performance criteria, The following changes to the original LAR-17-037 are proposed.

Page 3 of 7

ND-18-1006 9 Response to NRC Question on Evaluation of Departures from Tier 2* Information Regarding First of a Kind Testing (LAR-17-037S4)

Changes to Original LAR-17-037:

Changes to Enclosure 1:

Add a sixth screening criterion described on page 6 of 19 to read:

6. Result in a change to the Passive Residual Heat Removal Heat Exchanger natural circulation test (first plant test), the Core Makeup Tank Heated Recirculation Tests (first three plants test), or the Automatic Depressurization System Blowdown Test (first three plants test) that is material to the test objectives or test performance criteria, Revise the first two paragraphs and supporting bullets on Page 8 of 19 of the original LAR-17-037 to read as follows:

[Note: This new text also reflects text that is added by the response to RAI LAR-17-037-3 in SNC letter ND-18-0608, Enclosure 11, page 3 of 4 (LAR-17-037S2) and revised by the response to RAI LAR-17-037-9 in SNC letter ND-18-0646, Enclosure 17, page 8 of 9 (LAR-17-037S3). Text that was added by LAR-17-037S2 or LAR-17-037S3 is shown in underlined black font.]

SNC performed an analysis of the Tier 2* matters listed in 10 CFR Part 52, Appendix D, Section VIII paragraphs B.6.b and B.6.c. The analysis examined each item in terms of the following criteria:

  • Is the Tier 2* information adequately addressed in the VEGP 3 and 4 Plant-specific Tier 1 DCD or VEGP 3 and 4 Combined License (COL)? This step included a review to determine the degree to which codes, standards, and design and qualification processes, are relied upon for ITAAC acceptance criteria, but not specified in the VEGP 3 and 4 Plant-specific Tier 1 DCD.
  • Would changes in the Tier 2* information be adequately addressed by other applicable regulations, e.g., 10 CFR 50.46?
  • Would a change to the Tier 2* information have safety-significance commensurate with a change to Tier 1 information?
  • Would the evaluation process defined in 10 CFR Part 52, Appendix D, paragraph VIII.B.5 consistently and reliably require prior NRC approval of a change to the Tier 2* information?

Following the evaluation process described above, SNC made the following conclusions regarding 11 9 of the 24 Tier 2* matters listed in 10 CFR Part 52, Appendix D, Section VIII paragraphs B.6.b and B.6.c:

  • First, a set of Tier 2* information is already addressed in Tier 1 and thus a change to this Tier 2* information, which would involve a change to the associated Tier 1 information, would require prior NRC approval. Therefore, neither an evaluation of safety-significance nor new evaluation criteria were considered necessary to provide assurance that changes would receive prior NRC approval.
  • Second, for another set of Tier 2* information it was concluded that a change to this information would not have safety-significance commensurate with a change Page 4 of 7

ND-18-1006 9 Response to NRC Question on Evaluation of Departures from Tier 2* Information Regarding First of a Kind Testing (LAR-17-037S4) to Tier 1 information. Thus, new evaluation criteria were not considered necessary for this set of Tier 2* information.

  • Third, it was determined that a change to a third set of Tier 2* information would require a prior NRC approval under 10 CFR Part 52, Appendix D, paragraph VIII.B.5 or another regulation in a consistent and reliable manner. Thus, it was concluded that the evaluation criteria currently provided in 10 CFR Part 52, Appendix D, VIII.B.5.b or VIII.B.5.c are adequate to reliably and consistently address changes to this information and new evaluation criteria to address changes to this information were not necessary.

The remaining 15 11 of the 24 Tier 2* matters listed in 10 CFR Part 52, Appendix D, Section VIII paragraphs B.6.b and B.6.c were selected for development of additional screening criteria that would determine whether an associated Tier 2* departure qualifies for the departure evaluation process outlined in 10 CFR Part 52, Appendix D, Section VIII.B.5. A summary of the analysis is provided in Enclosure 5. The selected matters are:

  • Instrumentation and control system design processes, methods, and standards
  • Piping design acceptance criteria Add a sixth screening criterion on page 9 of 19 to read:
6. Result in a change to the Passive Residual Heat Removal Heat Exchanger natural circulation test (first plant test), the Core Makeup Tank Heated Recirculation Tests (first three plants test), or the Automatic Depressurization System Blowdown Test (first three plants test) that is material to the test objectives or test performance criteria, Add the following detailed description and bases for Criterion 6 following the Criterion 4 bases on page 13 of 19:

Criterion 6 (Initial Test Program) detailed guidance:

A material change influences the outcome of the test such that it would affect whether the test objectives or performance criteria would be met.

  • The following examples are material changes:

o The addition, deletion, or alteration of a test step o Alteration of a detail that serves as the basis for acceptance in an NRC Final Safety Evaluation Report (FSER) related to the affected test

  • The following examples are not material changes:

o Editorial changes o Clarifications to improve reader understanding o Correction of inconsistencies within the document which are clearly discernible (e.g., between sections)

Page 5 of 7

ND-18-1006 9 Response to NRC Question on Evaluation of Departures from Tier 2* Information Regarding First of a Kind Testing (LAR-17-037S4) o Changes that do not change the meaning or substance of information presented (e.g., reformatting or removing detail as described in NEI 98-03, Revision 1, Guidelines for Updating Final Safety Analysis Reports, Section A4 [ADAMS Accession Number ML003779028])

Criterion 6 (Initial Test Program) Bases:

The VEGP 3 and 4 Plant-specific Tier 1 DCD does not contain information related to the special tests that establish a unique phenomenological performance parameter of the AP1000 design features beyond testing performed for Design Certification for the AP600 and that will not change from plant to plant. Proposed Criterion 6 would provide assurance that material changes to Tier 2* information related to these special tests would receive prior NRC approval. The special tests (i.e., first plant tests and first three plant tests) for which some Tier 2* information is contained in the VEGP 3 and 4 plant-specific Tier 2 DCD are:

  • Core Makeup Tank Heated Recirculation Tests (first three plants test) identified in UFSAR Subsection 14.2.9.1.3 Items (k) and (w)), and

Changes to Enclosure 3:

Add a sixth screening criterion for proposed License Condition 2.D(13) to read:

6. Result in a change to the Passive Residual Heat Removal Heat Exchanger natural circulation test (first plant test), the Core Makeup Tank Heated Recirculation Tests (first three plants test), or the Automatic Depressurization System Blowdown Test (first three plants test) that is material to the test objectives or test performance criteria, Changes to Enclosure 4:

Add a decision box after the decision box for Criterion 6 on page 2 of 3 with adjacent text to read:

6. Result in a change to the Passive Residual Heat Removal Heat Exchanger natural circulation test (first plant test), the Core Makeup Tank Heated Recirculation Tests (first three plants test), or the Automatic Depressurization System Blowdown Test (first three plants test) that is material to the test objectives or test performance criteria, Page 6 of 7

ND-18-1006 9 Response to NRC Question on Evaluation of Departures from Tier 2* Information Regarding First of a Kind Testing (LAR-17-037S4)

Changes to Enclosure 5 Change the table entries for first plant only and first three plants only tests on page 5 of 6 to read:

Section VIII.B.6.c Selected for (Tier 2* Matters that additional Expire at Full Power) screening Basis Associated Criteria 10 Passive residual heat removal No Yes Adequately N/A Result in a change to (PRHR) natural circulation test addressed in Tier 1 the Passive Residual Heat (first plant only). and COL Not Removal Heat Exchanger addressed in Tier 1 natural circulation test (first plant test)that is material to the test objectives or test performance criteria 11 Automatic depressurization No Yes Adequately N/A Result in a change to system (ADS) and core make- addressed in Tier 1 the Core Makeup Tank up tank (CMT) verification tests and COL Not Heated Recirculation Tests (first three plants only). addressed in Tier 1 (first three plants test), or the Automatic Depressurization System Blowdown Test (first three plants test) that is material to the test objectives or test performance criteria Page 7 of 7

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