ML20132A078

Safety Evaluation for Vogtle Electric Generating Plant Unit 3 (LAR 20-001)

ML20132A078
Person / Time
Site:	Vogtle
Issue date:	08/04/2020
From:	Cayetano Santos NRC/NRR/VPOB
To:	City of Dalton, GA, Georgia Power Co, MEAG Power, Oglethorpe Power Corp, Southern Nuclear Operating Co
Tanny Santos Ex 7270
References
EPID L-2020-LLE-0009, LAR 20-001
Download: ML20132A078 (12)

Text

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 182 TO THE COMBINED LICENSE NO. NPF-91 SOUTHERN NUCLEAR OPERATING COMPANY, INC.

GEORGIA POWER COMPANY OGLETHORPE POWER CORPORATION MEAG POWER SPVM, LLC MEAG POWER SPVJ, LLC MEAG POWER SPVP, LLC CITY OF DALTON, GEORGIA VOGTLE ELECTRIC GENERATING PLANT UNIT 3 DOCKET NO.52-025

1.0 INTRODUCTION

By letter dated February 7, 2020 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML20038A939), Southern Nuclear Operating Company (SNC) requested that the Nuclear Regulatory Commission (NRC or Commission) amend Vogtle Electric Generating Plant (VEGP) Unit 3 Combined License (COL) Number NPF-91. License Amendment Request (LAR)20-001 requests changes to the VEGP Units 3 and 4 Updated Final Safety Analysis Report (UFSAR) in the form of departures from the incorporated plant-specific Design Control Document (PS-DCD) Tier 2 and Tier 2* information. The LAR also requests changes to plant-specific Tier 1 information with corresponding changes to Inspections, Tests, Analyses, and Acceptance Criteria (ITAAC) in COL Appendix C. Specifically, the request proposes to change the north-south minimum gap requirement above grade1 between the Nuclear Island and the Annex Building west of Column Line I from elevation (El.) 141 feet through El. 154 feet to accommodate as-built localized nonconformances.

Pursuant to Title 10 of the Code of Federal Regulations (10 CFR) 52.63(b)(1), SNC also requested an exemption from the provisions of 10 CFR Part 52, Appendix D, Design 1 Grade is defined as elevation (El.) 100 feet.

Certification Rule for the AP1000 Design,Section III.B, Scope and Contents. The requested exemption would allow a departure from the corresponding portions of the certified information in Tier 1 of the generic DCD.2 In order to modify the UFSAR (the plant-specific DCD) Tier 1 information, the NRC must find SNCs exemption request included in its submittal for the LAR to be acceptable. The staffs review of the exemption request, as well as the LAR, is included in this safety evaluation.

On March 10, 2020, the NRC staff published a proposed no significant hazards consideration determination in the Federal Register (85 FR 13944) for the proposed amendment.

2.0 REGULATORY EVALUATION

LAR 20-001 proposes to change the minimum north-south seismic gap requirement above grade between the Nuclear Island and the Annex Building west of Column Line I from El. 141 feet through El. 154 feet in the licensing basis to accommodate as-built localized nonconformances during construction of VEGP Unit 3. The requested change in the seismic gap in this area is from 3 inches to 2-1/16 inches. The amendment requests departures from the UFSAR Unit 3 PS-DCD Tier 2* Section 3.8.5.1; Tier 2 Section 3.7.2.8.1; Tier 2 Appendix 2E Section 5.2; and VEGP Unit 3 COL Appendix C (and VEGP Unit 3 plant-specific Tier 1) Table 3.3-6, ITAAC No. 3.3.00.13. The acceptance criteria for this ITAAC would be revised to add ;

except that the minimum horizontal clearance between elevations 141'-0" and 154'-0" between structural elements of the annex building and the nuclear island between column lines I and J is 2-1/16 inches.

The staff considered the following regulatory requirements in reviewing the LAR.

10 CFR Part 52, Appendix D, Section VIII.A.4, states that exemptions from Tier 1 information are governed by 10 CFR 52.63(b)(1) and 52.98(f). It also states that the Commission will deny a request for an exemption from Tier 1 if it finds that the design change will result in a significant decrease in the level of safety otherwise provided by the design.

10 CFR Part 52, Appendix D, Section VIII.B.5.a allows an applicant or licensee who references this appendix to 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 Technical Specifications, or requires a license amendment under paragraphs B.5.b or B.5.c of the section.

10 CFR 52.63(b)(1) allows the licensee who references a design certification rule to request NRC approval for an exemption from one or more elements of the certification information. The Commission may grant such a request only if it determines that the exemption will comply with the requirements of 10 CFR 52.7, which, in turn, points to the requirements listed in 10 CFR 50.12 for specific exemptions. In addition to the factors listed in 10 CFR 52.7, the Commission shall consider whether the special circumstances present outweigh any decrease in safety that may result from the reduction in standardization caused by the exemption.

2 While SNC describes the requested exemption as being from Section III.B of 10 CFR Part 52, Appendix D, the entirety of the exemption pertains to proposed departures from Tier 1 information in the generic DCD. In the remainder of this evaluation, the NRC will refer to the exemption as an exemption from Tier 1 information to match the language of Section VIII.A.4 of 10 CFR Part 52, Appendix D, which specifically governs the granting of exemptions from Tier 1 information.

Therefore, any exemption from the Tier 1 information certified by Appendix D to 10 CFR Part 52 must meet the requirements of 10 CFR 50.12, 52.7, and 52.63(b)(1).

10 CFR 52.98(f) requires NRC approval for any modification to, addition to, or deletion from the terms and conditions of a COL. These activities involve a change to COL Appendix C ITAAC information with corresponding changes to the associated plant-specific DCD Tier 1 information.

Therefore, NRC approval is required prior to making these changes.

10 CFR 52.97(b) requires the ITAAC in a COL to be necessary and sufficient to provide reasonable assurance that, if the ITAAC are satisfied, the facility has been constructed and will be operated in conformity with the license, the provisions of the Atomic Energy Act of 1954, as amended, and the Commission's rules and regulations.

The specific NRC technical requirements applicable to LAR 20-001 are the general design criteria (GDC) in Appendix A, General Design Criteria for Nuclear Power Plants, to 10 CFR Part 50, Domestic Licensing of Production and Utilization Facilities. In particular, these technical requirements include the following GDC:

GDC 1, Quality standards and records, provides, in part, that structures, systems, and components (SSCs) important to safety shall be designed, fabricated, erected, and tested to quality standards commensurate with the importance of safety functions to be performed.

GDC 2, Design bases for protection against natural phenomena, provides, in part, that SSCs important to safety shall be designed to withstand the effects of natural phenomena such as earthquakes, tornadoes, hurricanes, floods, tsunami, and seiches without loss of capability to perform their safety functions.

GDC 4, Environmental and dynamic effects design bases, provides, in part, that SSCs important to safety shall be designed to accommodate the effects of and to be compatible with the environmental conditions associated with normal operation, maintenance, testing, and postulated accidents, including loss-of-coolant accidents.

Regulatory guidance referred to in this evaluation includes NUREG-0800, Standard Review Plan (SRP) Section 3.8.5, Foundation, and Section 3.7.2, Seismic System Analysis.

3.1 TECHNICAL EVALUATION

OF PROPOSED CHANGES The staff evaluated the impact of changing the minimum gap requirement above grade between the Nuclear Island and the Annex Building west of Column Line I from El. 141 feet through El.

154 feet from 3 inches to 2-1/16 inches. In performing its evaluation, the staff considered SNCs design criteria described in UFSAR Section 3.7.2, Seismic System Analysis, for the Annex Building where the portion of the Annex Building adjacent to the Nuclear Island is classified as a seismic Category II building. Seismic Category II building structures are designed for the safe shutdown earthquake (SSE) using the same methodology and design criteria used for seismic Category I structures. For the LAR review, the staff considered structural design guidance in SRP Section 3.8.5, Foundation, and SRP Section 3.7.2, Seismic System Analysis, which provide acceptable methods for meeting the requirements of GDC 1, 2, and 4.

As a part of the review, the staff performed an audit in the Westinghouse Electric Company (Westinghouse) Electronic Reading Room (ERR) to review SNCs settlement information supporting LAR 20-001 (ADAMS Accession Number ML20141L698). SNC and Westinghouse made several documents available on the ERR for the NRC staff to audit. The documents included the predicted and measured settlements at different locations of the Nuclear Island and the Annex Building, including the construction sequence with time lag of construction between these buildings, trend analysis of the total and differential settlement with time, and predicted Annex Building displacement due to soil-structure interaction (SSI) during a safe-shutdown earthquake. The staff reviewed the following areas supporting the revised seismic gap identified in LAR 20-001: (1) applied loads during various phases of construction, (2) the settlement monitoring data, (3) the predicted settlement values, and (4) the stiffness of the turbine and annex buildings.

Soil-Structure Interaction SNC proposes to change the minimum gap requirement above grade between the Nuclear Island and Annex Building west of Column Line I from El. 141 feet through El. 154 feet from 3 inches to 2-1/16 inches. LAR 20-001 states the following:

Currently, the requirement in the licensing basis for the minimum gap between the nuclear island and annex building is 3 inches, as specified in COL Appendix C, ITAAC No. 3.3.00.13, UFSAR Appendix 2.5E Section 5.2, and UFSAR Subsections 3.7.2.8.1 and 3.8.5.1. UFSAR Subsection 3.8.5.1 requires that a minimum 1-inch gap be maintained between the nuclear island and annex building considering the displacements of the buildings during the SSE events. The purpose of the licensing basis requirements is to prevent interaction between the nuclear island and annex building during SSE events.

The UFSAR describes the seismic response analyses, including soil-structure interaction between the Nuclear Island and the adjacent buildings. The licensee performed soil-structure interaction analysis using the System for Analysis of Soil-Structure Interaction (SASSI) program.

As specified in UFSAR Section 3.7.2.8.4, Seismic Modeling and Analysis of Seismic Category II Building Structures, the relative displacement between the Nuclear Island and adjacent buildings is established from the SASSI two-dimensional (2D) analysis. LAR 20-001 states that previous design changes are incorporated into the latest AP1000 generic 2D SASSI analysis, including changes to the Nuclear Island (e.g., polar crane mass change) and adjacent buildings (e.g., change of structures of turbine building first bay). The licensee performed a site-specific 2D SASSI analysis for VEGP Units 3 and 4 to show the acceptability of the AP1000 plant at the VEGP site. Also, LAR 20-001 states that a study was performed to compare the deflections at the perimeter walls from the generic SASSI analysis using models that include the significant building changes and those that do not include the changes. The staff accepts this study because it was based on the approved design bases methodology. The study confirmed that the recent changes do not have a significant impact on the relative displacement between buildings during an SSE (including static and dynamic loading conditions). LAR 20-001 states that the gap between the auxiliary building and annex building at locations with nonconformances during a seismic event calculated based on the VEGP Unit 3 site-specific SASSI [analysis] is larger than 1.73 inches. The staff confirmed these results during the audit.

Therefore, the gap at locations with nonconformances is larger than the 1-inch minimum gap during an SSE that is stated in the UFSAR.

Settlement In LAR 20-001, SNC states that differential settlement of foundations may impact the gap between the Nuclear Island and adjacent buildings. SNC measured the settlement at different selected survey points on the Nuclear Island and adjacent buildings of VEGP Unit 3 during all construction stages as a part of the settlement monitoring program. Based on the VEGP Unit 3 settlement survey data, LAR 20-001 states that the walls of the Nuclear Island tend to tilt away from the Annex Building and, thereby, increase the gap between structures. In addition, the deflection contour of the Annex Building is uniform near the Nuclear Island. Consequently, the Annex Building is not showing any tendency to tilt towards the Nuclear Island. Therefore, SNC concluded in the LAR that differential settlement of the Nuclear Island and the Annex Building does not have an adverse impact on the gap between these two structures.

The staff reviewed settlement information to determine whether the measured differential settlement between the foundations of the Nuclear Island and the Annex Building would reduce the current gap between these two structures further and, therefore, would result in an adverse effect on these structures during an SSE.

UFSAR Section 2.5.4, Stability of Subsurface Materials and Foundations, describes the subsurface materials and foundation at the VEGP site. The staff notes the following information from this UFSAR Section:

Section 2.5.4.2, Properties of Subsurface Materials, provides the description and characteristics of the subsurface materials below the foundation of the buildings.

Section 2.5.4.2.2.2, Blue Bluff Marl (Lisbon Formation), describes the characteristics of the Blue Bluff Marl layer at the site.

Section 2.5.4.5, Excavation and Backfill, describes the properties and design of the backfill placed underneath the foundations replacing the Upper Sand Stratum.

Section 2.5.4.8, Liquefaction Potential, describes the possibility of liquefaction of the subsurface materials and the factor of safety against it.

Section 2.5.4.10.1, Bearing Capacity, describes the allowable bearing capacity of the subsurface materials below the foundations, including backfill.

Section 2.5.4.10.2, Settlement Analysis, describes the acceptable total and differential settlements between the buildings and across the Nuclear Island foundation mat and the settlement monitoring plan throughout construction.

LAR 20-001 states the following:

The long-term (consolidation) settlement is expected to be relatively small because the Vogtle site has thick engineered compacted backfill and over-consolidated Blue Bluff Marl overlying the lower sand stratum. Based on the site-specific settlement data through 2019, no significant changes are anticipated to the aforementioned short-term and long-term settlement trends.

The Nuclear Island of VEGP Unit 3 is constructed over a 90-foot-thick compacted Seismic Category 1 backfill. The backfill is placed at the top of Blue Bluff Marl layer having a thickness of approximately 76 feet. The Lower Sand stratum is 900 to 1,000 feet thick and lies between the Blue Bluff Marl Layer and the rock layer. Based on the laboratory and field-measured experimental results and analyses, the UFSAR concludes that both the compacted backfill and the Blue Bluff Marl have an adequate factor of safety against liquefaction during an SSE.

Laboratory tests have established that the Blue Bluff Marl and the Lower Sand stratum behave elastically and are very dense and stiff (i.e., very high modulus). In addition, the Blue Bluff Marl is an overconsolidated material with an overconsolidated ratio of approximately 8. Therefore, the settlement from construction of the structures would be small, and a significant portion of it would occur immediately as the load is applied or within a short time thereafter due to this high overconsolidation stress in addition to the high modulus of the materials and slow loading rate.

During the audit, the staff reviewed the VEGP Unit 3 settlement survey data and related documents. The audited documents contain measured settlement data at different locations around the Nuclear Island and adjacent buildings from the beginning of construction through the current state. Plots of the settlement survey data present the total settlement at the various measurement points as a function of time from the start of construction. In addition, several settlement profiles in the east-west and north-south directions of the foundations of the Nuclear Island and adjacent buildings were provided.

The staff notes that both the actual settlement at the survey points and the predicted settlement values in the UFSAR are within the acceptable limits given in Table 5.0-1, Site Parameters, of the AP1000 DCD and UFSAR. Consequently, the staff expects that settlement will be well controlled within the acceptable settlement limits throughout the entire construction sequence and through plant operation.

The staff also confirmed through examination of the settlement survey data and plots during the audit that the Nuclear Island basemat has deflected more in the center (where the Shield Building is located) and less at the perimeter. This pattern of basemat deflection would tend to cause the perimeter walls to lean towards the center of the Nuclear Island. Additionally, the staff confirmed that the foundation deflection of the Annex Building is quite uniform near the Nuclear Island. This observation indicates that the walls of the Annex Building near the Nuclear Island are not exhibiting a tendency to lean towards the Nuclear Island. Based on above, the staff finds that the differential settlement of the foundations of the Nuclear Island and the Annex Building will not adversely affect the seismic gap between these two structures by reducing the currently available gap, especially at the area of nonconformance.

UFSAR Section 2.5.4.8 states that [l]iquefaction is not a concern. UFSAR Section 2.5.4.10.1 states that the subsurface materials underneath the foundations of the structures have adequate bearing capacity to withstand an SSE. Therefore, the staff finds that the proposed change in the seismic gap will not result in adverse effects during an SSE.

Conclusions Based on the above, the staff finds that the proposed changes to the gap requirements between the Nuclear Island and the Annex Building at the area of nonconformance do not affect the structural integrity of the seismic Category I structures. Similarly, the structural integrity of the seismic Category II structures is not impacted. The proposed design change is consistent with the acceptance criteria specified in SRP Sections 3.7.2.8 and 3.8.5 and UFSAR Sections 3.7 and 3.8. Based on these findings and because the LAR meets the guidance in SRP Sections 3.7.2 and 3.8.5, the staff concludes that the requirements of GDC 1, 2, and 4 of Appendix A to 10 CFR Part 50 will continue to be met. Therefore, the staff finds the proposed changes to be acceptable.

The staff also finds that with the LARs proposed revision to ITAAC 3.3.00.13, the ITAAC will continue to be sufficient to verify that the facility has been constructed and will operate in accordance with the license, the provisions of the Atomic Energy Act of 1954, as amended, and the Commissions rules and regulations. This finding is based on the fact that the revised acceptance criteria in this ITAAC is consistent with the revised gap requirement between the Nuclear Island and Annex Building in the area of the nonconformance. Therefore, within the scope of this license amendment, the NRC finds that 10 CFR 52.97(b) is satisfied.

3.2 EVALUATION OF EXEMPTION The regulations in Section III.B of Appendix D to 10 CFR Part 52 require a holder of a COL referencing Appendix D to 10 CFR Part 52 to incorporate Appendix D by reference and comply with its requirements, including certified information in Tier 1 of the generic AP1000 DCD.

Exemptions from Tier 1 information are governed by the change process in Section VIII.A.4 of Appendix D of 10 CFR Part 52. Because SNC has identified changes to plant-specific Tier 1 information, with corresponding changes to the associated COL Appendix C information, resulting in the need for a departure, an exemption from the certified design information within plant-specific Tier 1 material is required to implement the LAR.

The Tier 1 information for which a plant-specific departure and exemption was requested is described above. The result of this exemption would be that SNC could implement the requested modifications to Tier 1 information, with corresponding changes to COL Appendix C.

Pursuant to the provisions of 10 CFR 52.63(b)(1), an exemption from elements of the design as certified in the 10 CFR Part 52, Appendix D, design certification rule is requested for the specified Tier 1 change described and justified in LAR 20-001. This exemption is a permanent exemption limited in scope to the particular Tier 1 information specified.

As stated in Section VIII.A.4 of Appendix D to 10 CFR Part 52, an exemption from Tier 1 information is governed by the requirements of 10 CFR 52.63(b)(1) and 52.98(f). Additionally,Section VIII.A.4 of Appendix D to 10 CFR Part 52 provides that the Commission will deny a request for an exemption from Tier 1 if it finds that the requested change will result in a significant decrease in the level of safety otherwise provided by the design. Pursuant to 10 CFR 52.63(b)(1), the Commission may grant exemptions from one or more elements of the certification information, so long as the criteria given in 10 CFR 52.7, which, in turn, references 10 CFR 50.12, are met. Also, the Commission must consider whether the special circumstances, which are defined by 10 CFR 50.12(a)(2), outweigh any potential decrease in safety due to reduced standardization caused by the exemption.

Pursuant to 10 CFR 52.7, the Commission may, upon application by any interested person or upon its own initiative, grant exemptions from the requirements of 10 CFR Part 52. As 10 CFR 52.7 further states, the Commissions consideration will be governed by 10 CFR 50.12, Specific exemptions, which states that an exemption may be granted when: (1) the exemptions are authorized by law, will not present an undue risk to the public health and safety, and are consistent with the common defense and security; and (2) special circumstances are present. Specifically, 10 CFR 50.12(a)(2) lists six circumstances for which an exemption may be granted. It is necessary for one of these circumstances to be present in order for the NRC to consider granting an exemption request. SNC stated that the requested exemption meets the special circumstances of 10 CFR 50.12(a)(2)(ii). That subparagraph 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 staffs analysis of these findings is presented below.

3.2.1 AUTHORIZED BY LAW The requested exemption would allow SNC to implement the amendment described above.

This exemption is a permanent exemption limited in scope to particular Tier 1 information.

Subsequent changes to this plant-specific Tier 1 information, and corresponding changes to Appendix C, or any other Tier 1 information would be subject to the exemption process specified in Section VIII.A.4 of Appendix D to 10 CFR Part 52 and the requirements of 10 CFR 52.63(b)(1). As stated above, 10 CFR Part 52, Appendix D, Section VIII.A.4 allows the NRC to grant exemptions from one or more elements of the Tier 1 information. The NRC staff has determined that granting of SNCs proposed exemption will not result in a violation of the Atomic Energy Act of 1954, as amended, or the Commissions regulations. Therefore, as required by 10 CFR 52.7 and 10 CFR 50.12(a)(1), the exemption is authorized by law.

3.2.2 NO UNDUE RISK TO PUBLIC HEALTH AND SAFETY As discussed above in the technical evaluation, the proposed changes comply with the NRCs substantive safety regulations. Therefore, there is no undue risk to the public health and safety.

3.2.3 CONSISTENT WITH COMMON DEFENSE AND SECURITY The proposed exemption would allow changes as described above in the technical evaluation, thereby departing from the AP1000 certified (Tier 1) design information. The change does not alter or impede the design, function, or operation of any plant structures, systems, or components associated with the facilitys physical or cyber security and, therefore, does not affect any plant equipment that is necessary to maintain a safe and secure plant status. In addition, the changes have no impact on plant security or safeguards. Therefore, as required by 10 CFR 50.12(a)(1), the staff finds that the common defense and security is not impacted by this exemption.

3.2.4 SPECIAL CIRCUMSTANCES Special circumstances, in accordance with 10 CFR 50.12(a)(2), are present, in part, whenever application 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 underlying purpose of the Tier 1 information is to ensure that a licensee will safely construct and operate the plant based on the certified information found in the AP1000 DCD, which was incorporated by reference into the VEGP Unit 3 licensing basis. The proposed changes described in the above technical evaluation do not impact the ability of any SSCs to perform their functions or negatively impact safety.

Special circumstances are present in the particular circumstances discussed in LAR 20-001 because the application of the specified Tier 1 information is not necessary to achieve the underlying purpose of the rule. The proposed changes do not adversely affect any function or feature used for the prevention and mitigation of accidents or their safety analyses. The LAR and exemption request demonstrate that the proposed revisions to the Tier 1 information and corresponding changes to Appendix C will continue to meet applicable regulatory requirements.

Therefore, for the above reasons, the staff finds that the special circumstances required by 10 CFR 50.12(a)(2)(ii) for the granting of an exemption from the Tier 1 information exist.

3.2.5 SPECIAL CIRCUMSTANCES OUTWEIGH REDUCED STANDARDIZATION This exemption would allow the implementation of changes to Tier 1 information in the plant-specific DCD and corresponding changes to Appendix C that are being proposed in the LAR.

The justification provided in LAR 20-001, the exemption request, and the associated licensing basis mark-ups demonstrate that there is a limited change from the standard information provided in the generic AP1000 DCD that does not reduce the level of safety in the design. The design functions of the system associated with this request will continue to be maintained because the associated revisions to the Tier 1 information support the design function of the Nuclear Island and Annex Building. Consequently, the safety impact that may result from any reduction in standardization is minimized. Based on the foregoing reasons, the staff finds that the special circumstances outweigh any decrease in safety that may result from the reduction of standardization of the AP1000 design.

3.2.6 NO SIGNIFICANT REDUCTION IN SAFETY The exemption request proposes to depart from the certified design by allowing changes discussed above in the technical evaluation. The proposed changes will not adversely affect the ability of the Nuclear Island and Annex Building to perform its design functions, and the level of safety provided by the current systems and equipment therein is unchanged. Therefore, based on the foregoing reasons and as required by 10 CFR Part 52, Appendix D, Section VIII.A.4, the staff finds that granting the exemption would not result in a significant decrease in the level of safety otherwise provided by the design.

For the reasons given above, the standards for an exemption from the specified Tier 1 information have been satisfied.

4.0 FINAL NO SIGNIFICANT HAZARDS CONSIDERATION

The NRC staff published its proposed no significant hazards consideration determination in the Federal Register on March 10, 2020 (85 FR 13944). A request for a hearing was filed on May 11, 2020, by the Blue Ridge Environmental Defense League and its chapter, Concerned Citizens of Shell Bluff (ADAMS Accession No. ML20132D299). An Atomic Safety and Licensing Board will rule on whether to grant the hearing request.

Under its regulations, the Commission may issue an amendment and make it immediately effective, notwithstanding the pendency before it of a request for a hearing from any person, where it has made a final determination that no significant hazards consideration is involved.

The NRCs regulation in 10 CFR 50.92(c) states that the NRC may make a final determination, under the procedures in 10 CFR 50.91, that a license amendment involves no significant hazards consideration if operation of the facility, in accordance with the amendment, would not:

(1) involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) involve a significant reduction in a margin of safety.

An evaluation of the issue of no significant hazards consideration is presented below:

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

Response: No.

The proposed change would revise the COL and licensing basis for VEGP Unit 3 to locally modify the seismic gap requirement above grade between the nuclear island and portions of the annex building adjacent to the nuclear island.

The proposed change to the gap requirement does not affect the structural integrity requirements on seismic Category I structures. The safety functions of the seismic Category I structures are not impacted. The performance of the seismic Category II structures is not impacted, and the change will not degrade the function of a seismic Category I structure, system, or component (SSC). The proposed change does not involve a significant change to the design of the nuclear island or annex building, and no SSC function is affected. No design or safety analysis is affected. The proposed change does not affect any accident initiating event or component failure, thus the probabilities of the accidents previously evaluated are not affected. No function used to mitigate a radioactive material release and no radioactive material release source term is involved; thus the radiological releases in the accident analyses are not affected.

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

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 change would revise the COL and licensing basis for VEGP Unit 3 to locally modify the seismic gap requirement above grade between the nuclear island and portions of the annex building adjacent to the nuclear island.

The proposed change does not involve a significant change to the design of the nuclear island or annex building, and no SSC function is affected. The performance of the seismic Category II structures is not impacted and will not degrade the function of a seismic Category I SSC. The proposed change would not introduce a new failure mode, fault or sequence of events that could result in a radioactive material release.

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

3. Does the proposed amendment involve a significant reduction in a margin of safety?

Response: No.

The proposed change would revise the COL and licensing basis for VEGP Unit 3 to locally modify the seismic gap requirement above grade between the nuclear island and portions of the annex building adjacent to the nuclear island.

The proposed change does not involve a significant change to the design of the nuclear island or annex building, and no SSC function is affected. The performance of the seismic Category II structures is not impacted, and the change will not degrade the function of a seismic Category I SSC and would not affect any design function or analysis. There would be no change to an existing design basis, design function, regulatory criterion, or analysis. No safety analysis or design basis acceptance limit/criterion is involved.

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

Based on the above evaluation, the staff concludes that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the staff has made a final determination that no significant hazards consideration is involved for the proposed amendment and that the amendment should be issued as allowed by the criteria contained in 10 CFR 50.91.

5.0 STATE CONSULTATION

In accordance with the Commission's regulations, the Georgia State official was notified of the proposed issuance of the amendment on July 7, 2020. The State official had no comments.

6.0 ENVIRONMENTAL CONSIDERATION

The amendment changes a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20, Standards for Protection Against Radiation. The staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite. Also, there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration, and there has been no public comment on such finding as published in the Federal Register on March 10, 2020 (85 FR 13944). Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Under 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.

Because the exemption is necessary to allow the changes proposed in this LAR, and because the exemption does not authorize any activities other than those proposed in this LAR, the environmental consideration for the exemption is identical to that of the license amendment.

Accordingly, the 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), the staff finds that no environmental impact statement or environmental assessment needs to be prepared in connection with the issuance of the exemption.

7.0 CONCLUSION

The staff has determined that pursuant to Section VIII.A.4 of Appendix D to 10 CFR Part 52, the exemption proposed in this LAR (1) is authorized by law; (2) presents no undue risk to the public health and safety; (3) is consistent with the common defense and security; (4) presents special circumstances; and (5) does not significantly reduce the level of safety at the licensees facility. Also, the staff has determined that the special circumstances for the exemption outweigh any decrease in safety that may result from a reduction in standardization caused by the exemption. Therefore, the NRC staff grants the exemption from the Tier 1 information requested by SNC.

The staff has also concluded, based on the technical evaluation presented in Section 3.1 that there is reasonable assurance that: (1) the health and safety of the public will not be endangered by operation in the proposed manner; (2) there is reasonable assurance that 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. Therefore, the NRC staff finds the changes proposed in this LAR acceptable.

8.0 REFERENCES

1. Southern Nuclear Operating Company, Vogtle Electric Generating Plant Unit 3, Request for License Amendment and Exemption: Unit 3 Auxiliary Building Wall 11 Seismic Gap Requirements (LAR-20-001), February 7, 2020 (ADAMS Accession No. ML20038A939).

2. Vogtle Electric Generating Plant Units 3 and 4, Updated Final Safety Analysis Report, Revision 8, June 14, 2019 (ADAMS Accession No. ML19171A096).

3. U.S. Nuclear Regulatory Commission, Audit Plan for Vogtle Electric Generating Plant Unit 3, Request for License Amendment and Exemption: Unit 3 Auxiliary Building Wall 11 Seismic Gap Requirements (LAR 20-001), dated March 20, 2020 (ADAMS Accession No. ML20063H206).

4. U.S. Nuclear Regulatory Commission, Audit Summary for Vogtle Electric Generating Plant Unit 3, Request for License Amendment and Exemption: Unit 3 Auxiliary Building Wall 11 Seismic Gap Requirements (LAR 20-001), dated May 26, 2020 (ADAMS Accession No. ML20141L698).

5. AP1000 Design Control Document, Revision 19, dated June 13, 2011 (ADAMS Accession No. ML11171A500).

6. Vogtle Electric Generating Plant, Unit 3 Current Facility Combined License NPF-91, (ADAMS Accession No. ML14100A106).

7. U.S. Nuclear Regulatory Commission, NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition, Section 3.7.2, Seismic System Analysis, (ADAMS Accession No. ML13198A223).

8. U.S. Nuclear Regulatory Commission, NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition, Section 3.8.5, Foundations (ADAMS Accession No. ML13198A267).