ML20049A808

Safety Evaluation for Vogtle Electric Generating Plant Units 3 and 4 (LAR 19-009)

ML20049A808
Person / Time
Site:	Vogtle
Issue date:	03/11/2020
From:	Alina Schiller NRC/NRR/VPOB
To:	City of Dalton, GA, Georgia Power Co, MEAG Power, Oglethorpe Power Corp, Southern Nuclear Operating Co
Alina Schiller EX. 8177
References
EPID L-2019-LLA-0213, LAR 19-009
Download: ML20049A808 (14)

Text

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NOS. 176 and 175 TO THE COMBINED LICENSE NOS. NPF-91 AND NPF-92, RESPECTIVELY 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 UNITS 3 AND 4 DOCKET NOS.52-025 AND 52-026

1.0 INTRODUCTION

By letter dated September 30, 2019 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML19273A953), as supplemented by letter dated December 19, 2019 (ADAMS Accession No. ML19353B752), Southern Nuclear Operating Company (SNC) requested that the Nuclear Regulatory Commission (NRC) amend Vogtle Electric Generating Plant (VEGP) Units 3 and 4, Combined License (COL) Numbers NPF-91 and NPF-92, respectively. SNC License Amendment Request (LAR)19-009 requested revisions of the licensing basis documents to reflect revised automatic depressurization system (ADS) and core makeup tank (CMT) parameters. The requested amendment required a departure from the Updated Final Safety Analysis Report (UFSAR) Tier 2 information that involves a change to the plant-specific Tier 1 (and associated COL Appendix C) information in Table 2.1.2-4 identifying:

a) the maximum stroke times for the ADS Stages 1, 2, and 3 valves; and b) the minimum effective flow areas for the ADS Stages 2 and 3 valves. The requested amendment also required a departure from the UFSAR Tier 2 information that involves a change to the COL Appendix A information in Technical Specification (TS) 3.5.2 and TS 3.5.3 identifying the required CMT minimum volume. Editorial changes were also provided for TS 3.5.3.

Pursuant to Section 52.63(b)(1) of Title 10 of the Code of Federal Regulations (10 CFR), SNC also requested an exemption from the provisions of 10 CFR Part 52, Appendix D, Design 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 design control document (DCD).1 In order to modify the UFSAR (plant-specific design control document (PS-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.

The supplement dated December 19, 2019, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staffs original proposed no significant hazards consideration determination as published in the Federal Register on December 3, 2019 (84 FR 66234).

2.0 REGULATORY EVALUATION

The requested amendment proposes changes to revise the UFSAR Tier 2 information, which involves a change to the TSs in COL Appendix A, and to COL Appendix C, and corresponding plant-specific Tier 1 information, as discussed in Section 1.0 of this safety evaluation. The staff considered the following regulatory requirements in reviewing the LAR that included the proposed changes:

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

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 TSs, or requires a license amendment under paragraphs B.5.b or B.5.c of the section.

10 CFR Part 52, Appendix D, Section VIII.C.6 states that after issuance of a license, Changes to the plant-specific TS will be treated as license amendments under 10 CFR 50.90. 10 CFR 50.90 states that a license holder, including a holder of a combined license, must file an application for an amendment with the Commission that fully describes the changes desired.

The proposed LAR requires changes in the TSs, and therefore a LAR is required to be submitted for NRC approval.

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 only grant such a request 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 outweigh any decrease in safety 1 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 PS-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.

that may result from the reduction in standardization caused by the exemption. 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 Inspections, Tests, Analyses, and Acceptance Criteria (ITAAC) information, with corresponding changes to the associated PS-DCD Tier 1 information. Therefore, NRC approval is required prior to making the plant-specific proposed changes in this LAR.

10 CFR 50.36 specifies that TSs impose limits, operating conditions, and other requirements upon reactor facility operation for the public health and safety. The TSs are derived from the analyses and evaluations in the safety analysis report. In general, TSs must contain: (1) safety limits and limiting safety system settings; (2) limiting conditions for operation; (3) surveillance requirements; (4) design features; and (5) administrative controls. 10 CFR 50.36(c)(3) states that surveillance requirements are requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met.

10 CFR 50.46 provides acceptance criteria for emergency core cooling systems for light-water nuclear power reactors.

10 CFR 52.97(b) requires that the Commission identify within the combined license the inspections, tests, and analyses, including those applicable to emergency planning, that the licensee shall perform, and the acceptance criteria that, if met, are necessary and sufficient to provide reasonable assurance that the facility has been constructed and will be operated in conformity with the license, the provisions of the Atomic Energy Act, and the Commission's rules and regulations. Consequently, proposed changes to the ITAAC should continue to meet the requirements of 10 CFR 52.97(b).

The specific NRC technical requirements applicable to LAR 19-009 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 GDCs:

GDC 10, Reactor Design, states that the reactor core and associated coolant, control, and protection systems shall be designed with appropriate margin to assure that specified acceptable fuel design limits are not exceeded during any condition of normal operation, including the effects of anticipated operational occurrences (AOOs).

GDC 15, Reactor coolant system design, states that the reactor coolant system (RCS) and associated auxiliary, control, and protection systems shall be designed with sufficient margin to assure that the design conditions of the reactor coolant pressure boundary are not exceeded during any condition of normal operation, including AOOs.

GDC 29, Protection against anticipated operational occurrences, states that the protection and reactivity control systems shall be designed to assure an extremely high probability of accomplishing their safety functions in the event of AOOs.

GDC 34, Residual heat removal, states that a system to remove residual heat shall be provided. The system safety function shall be to transfer fission product decay heat and other residual heat from the reactor core at a rate such that specified acceptable fuel design limits and the design conditions of the reactor coolant pressure boundary are not exceeded.

GDC 35, Emergency core cooling, states, in relevant part, that a system to provide abundant emergency core cooling shall be provided. The system safety function shall be to transfer heat from the reactor core following any loss of reactor coolant at a rate such that (1) fuel and clad damage that could interfere with continued effective core cooling is prevented and (2) clad metal-water reaction is limited to negligible amounts.

The following regulatory guidance was referred to in the technical evaluation:

Regulatory Guide 1.100, Seismic Qualification of Electrical and Active Mechanical Equipment and Functional Qualification of Active Mechanical Equipment for Nuclear Power Plants, Revision 3, September 2009 (ADAMS Accession No. ML091320468).

3.1 TECHNICAL EVALUATION

OF THE REQUESTED CHANGES In LAR 19-009, SNC requests a COL amendment for VEGP Units 3 and 4 to depart from the UFSAR Tier 2 information (which includes the PS-DCD Tier 2 information), to make related changes to plant-specific Tier 1 information, with corresponding changes to the associated COL Appendix C information, and to make related changes to COL Appendix A, Technical Specifications. SNC also requests an exemption for elements of the design as certified in the 10 CFR Part 52, Appendix D, design certification rule for the PS-DCD Tier 1 material departures. The LAR proposes changes to reflect revisions in the design parameters for: (a) the maximum stroke times for ADS Stages 1, 2, and 3 valves; (b) the minimum effective flow areas for ADS Stages 2 and 3 valves; and (c) the CMT minimum volume.

Changes Associated with ADS Stages 1, 2, and 3 The ADS Stages 1, 2, and 3 control valves and isolation valves open sequentially to provide rapid pressure reduction of the RCS prior to opening of the ADS Stage 4 valves to allow the passive core cooling system to deliver cooling water flow to the reactor core in the event of a loss-of-coolant accident (LOCA). The ADS Stage 1 control valves are two normally closed dc-powered motor-operated 4-inch globe valves each in series with a normally closed, dc-powered, motor-operated 4-inch gate isolation valve. The ADS Stages 2 and 3 control valves are four normally closed, dc-powered, motor-operated 8-inch globe valves each in series with a normally closed, dc-powered, motor-operated 8-inch gate isolation valve.

The ADS Stage 1 control valves and isolation valves actuate at a discrete CMT level, as either tanks level decreases during injection or from spilling out of a broken injection line. The ADS Stages 2 and 3 control and isolation valves actuate based upon a timed delay after actuation of the preceding stage. The valve opening sequence prevents simultaneous opening of more than one stage to allow sequential opening for the valves. During the actuation of each stage, the isolation valve is sequenced to open before the control valve.

The ADS Stage 4 valves are four 14-inch pyrotechnic-actuated (squib) valves each in series with a normally open, dc-powered, motor-operated gate isolation valve. The ADS Stage 4 squib valves are interlocked so they cannot be opened until the RCS pressure has been substantially reduced.

As described in COL Appendix C, Table 2.1.2-4, the design commitment for ITAAC 2.1.02.08d.i (Index No. 32) is that the RCS provides automatic depressurization during design-basis events.

ITAAC 2.1.02.08d.iv (Index No. 35) requires type tests and analysis to determine the minimum effective flow area through the ADS Stages 1, 2, and 3 control valves, with the minimum effective flow area through each ADS Stages 2 and 3 control valve verified to be greater than or equal to 19 square inches (in2). ITAAC 2.1.02.11a.ii (Index No. 47) and ITAAC 2.1.02.11b.iii (Index No. 50) together require that testing be performed to demonstrate that the Stages 1, 2, and 3 ADS valves open within the required response time, with the maximum opening time for ADS Stage 1 control valves verified to be less than or equal to 40 seconds and the maximum opening time for the ADS Stages 2 and 3 control valves verified to be less than or equal to 100 seconds.

With respect to the ADS valves, SNC proposes the following licensing basis changes to COL Appendix C (and corresponding plant-specific Tier 1) Table 2.1.2-4, RCS ITAAC:

Revise ITAAC 2.1.02.08d.iv (Index No. 35) for ADS Stages 2 and 3 control valves minimum effective flow area.

Revise ITAAC 2.1.02.11a.ii (Index No. 47) and ITAAC 2.1.02.11.b.iii (Index No. 47) for both ADS Stage 1 control valves and ADS Stages 2 and 3 control valves maximum opening times.

Corresponding changes are proposed for UFSAR Table 14.3-2 (Sheet 17 of 17), Design Basis Accident Analysis, and Table 15.6.5-10, ADS Parameters Used in Small-Break LOCA Analysis.

With respect to the ADS Stage 1 control valves (RCS-PL-V001A/B), SNC states that it had determined that these valves might not meet their design specification requirements at elevated design temperatures in that the motor might not meet the maximum valve opening time requirement specified in the licensing basis. Therefore, SNC proposes that the maximum opening time for these valves be changed from 40 seconds to 48 seconds. SNC also plans to change the limit switch setting to indicate that each valve is open between 85 and 90% of valve stem travel and to shut off the motor approximately +3% after the open indication signal. SNC indicates that the ADS Stage 1 isolation valves will meet their stroke-time opening requirements such that no changes are necessary for those valves. SNC notes that no changes to the effective flow area of the ADS Stage 1 valves are necessary.

With respect to the ADS Stages 2 and 3 control valves (RCS-PL-V002A/B and V003A/B), SNC states that the effective flow area for these valves was not appropriately calculated for the anticipated flow conditions. SNC indicates that new testing was performed to determine the effective flow area with compressible flow for these valves. As a result, SNC proposes to update the analyzed effective flow area for the ADS Stages 2 and 3 control valves based on the new test data. The minimum effective flow area will be changed from 19 in2 to 16 in2 based on the test data for the 8-inch globe valves used for the ADS Stages 2 and 3 control valves. The limit switch settings for the ADS Stages 2 and 3 control valves will limit the valve travel to the point at which the full effective flow area is reached to allow slower opening characteristics and to reduce hydrodynamic loading. SNC also proposes to increase the maximum opening stroke times for the ADS Stages 2 and 3 control valves from 100 seconds to 120 seconds. SNC indicates that no changes are necessary to the ADS Stage 2 and 3 isolation valves.

The staff notes that LAR 19-009 specifies that the proposed changes to the valve parameters do not involve any physical changes to the valve body or internals for the ADS valves. SNC will accomplish the longer stroke times for the ADS valves requested in LAR 19-009 by gearing adjustments to the actuators. These gearing adjustments will result in increased actuator output capability for opening their valves against the fluid flow forces.

ITAAC 2.1.02.12a.i (Index No. 53) requires SNC to demonstrate that the ADS valves are capable of satisfying their performance parameters in accordance with the valve qualification requirements. In LAR 19-009, SNC states that there will be no changes to the qualification of the ADS valves in accordance with American Society of Mechanical Engineers (ASME)

Standard QME-1-2007, Qualification of Active Mechanical Equipment Used in Nuclear Power Plants. The PS-DCD in Section 3.9.3.2.2, Valve Operability, specifies that AP1000 valves will be qualified to perform their safety functions in accordance with ASME Standard QME-1-2007.

The NRC accepts ASME Standard QME-1-2007 in Regulatory Guide 1.100 (Revision 3),

Seismic Qualification of Electrical and Active Mechanical Equipment and Functional Qualification of Active Mechanical Equipment for Nuclear Power Plants, with conditions. ASME Standard QME-1-2007 specifies that an Application Report must be prepared for each production valve to demonstrate its design-basis capability consistent with the qualified valve assembly. For VEGP Units 3 and 4, SNC will prepare a QME-1 Application Report for each ADS valve to address the motor actuator output capability and stroke time to satisfy the flow operating requirements as updated in LAR 19-009.

In addition, SNC states that the small break loss-of-coolant-accident (SBLOCA) analysis was updated to account for the accumulated design changes described above. This resulted in a 3 degree Fahrenheit (°F) peak cladding temperature (PCT) increase to the limiting inadvertent ADS transient for a total licensing basis PCT of 1099 °F. The staff finds the resultant increase to the PCT for a postulated SBLOCA acceptable because the total PCT remains below the maximum allowed cladding temperature specified in 10 CFR 50.46 (2200 °F).

NRC staff performed an audit as part of its review of LAR 19-009. A summary of the audit is provided in an audit report dated March 11, 2020 (ADAMS Accession No. ML20065M803).

During this audit, NRC staff reviewed the shutdown evaluation for loss of Normal Residual Heat Removal System (RNS) Cooling Analysis during Mode 4 through Mode 5 with the RCS intact and with automatic Safeguards System actuation. This review confirmed the conclusions documented in the FSAR Chapter 19E, Shutdown Evaluation, remain valid given the proposed design changes for the ADS Stages 1, 2, and 3 valves. The analyses showed that with loss of RNS being initiated at 2.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> post shutdown, the resultant core levels remained above the Top of Active Fuel (TAF). The acceptance criterion for the evaluation is that active fuel must be covered by a two-phase mixture during the duration of the event. Assumptions for equipment availability were consistent with TSs for the applicable plant modes and plant conditions.

As part of its review, the NRC staff evaluated whether VEGP Units 3 and 4 will continue to satisfy GDC 10 for reactor design, GDC 15 for RCS design, GDC 29 for AOO protection, GDC 34 for residual heat removal capability, and GDC 35 for emergency core cooling performance, with the valve modifications described in LAR 19-009. Based on the considerations stated above, the staff determined that the valve modifications will not adversely impact the reactor design, RCS design, AOO protection, residual heat removal capability, and emergency core cooling performance for VEGP Units 3 and 4. Therefore, the valves modifications described in LAR 19-009 will not affect the plant design features associated with GDC 10, 15, 29, 34, and 35.

The staff also finds that the revised ITAAC continue to be sufficient to verify that the facility has been constructed and will operate in accordance with the license, the provisions of the Act, and the Commissions rules and regulations. Therefore, within the scope of this license amendment, the NRC finds that the revised ITAAC meet the requirements of 10 CFR 52.97(b).

Based on its review, the staff finds that the proposed changes in LAR 19-009 for VEGP Units 3 and 4 to the stroke time for the ADS Stage 1 control valves, and the stroke time and minimum effective flow area for the ADS Stages 2 and 3 control valves, to be acceptable. SNC is required to demonstrate that the updated valve performance parameters in ITAAC 2.1.02.08d.iv (Index No. 35) and ITAAC 2.1.02.11a.ii (Index No. 47) with the acceptance criteria 11.b.iii (Index No. 47) are satisfied in completing those ITAAC. In addition, SNC is required to demonstrate that the ADS valves are capable of satisfying their updated performance parameters in accordance with the valve qualification requirements in ITAAC 2.1.02.12a.i (Index No. 53). During ITAAC inspections, the staff might verify completion of the applicable ITAAC for the ADS valves to demonstrate that their qualification is consistent with ASME Standard QME-1-2007 as specified in the PS-DCD. The staffs inspections will also cover selected valves with high safety significance (such as ADS valves) for VEGP Units 3 and 4 in accordance with NRC Inspection Procedure (IP) 73758, Part 52, Functional Design and Qualification, and Preservice and Inservice Testing Programs for Pumps, Valves and Dynamic Restraints.

Changes associated with CMT minimum volume The CMTs are cylindrical tanks with hemispherical upper and lower heads. They are made of carbon steel and clad on the internal surfaces with stainless steel. They are located in containment at an elevation slightly above the reactor coolant loops. Each tank is full of borated water at > 3400 ppm. During normal operation, the CMTs are maintained at RCS pressure through a normally open pressure balance line from the cold leg. The outlet line from each CMT is connected to one of two direct vessel injection lines, which provides an injection path for the water supplied by the CMT. The outlet line from each CMT is isolated by parallel, normally closed, fail open valves. Upon receipt of a safeguards actuation signal, these four valves open to align the CMTs to the RCS. Each of the two redundant CMTs provides sufficient borated water to assure RCS reactivity and inventory control for all design basis accidents, including both LOCA events and non-LOCA events.

In LAR 19-009, SNC requests changes to VEGP Units 3 and 4 COL Appendix A TS LCO 3.5.2, Core Makeup Tanks (CMTs) - Operating. Specifically, SNC proposes to delete Surveillance Requirement (SR) 3.5.2.2 which requires verification of borated water volume in each CMT 2487 ft3 when in Modes 1, 2, and 3, and in Mode 4 with the RCS not being cooled by RNS every seven days. This change also affects LCO 3.5.3, Core Makeup Tanks (CMTs) - Shutdown, Reactor Coolant System (RCS) Intact, when in Mode 4 with RCS cooling provided by RNS since SR 3.5.3.1 currently refers to SR 3.5.2.2 as an applicable surveillance. In the LAR, SNC stated that SBLOCA analysis is performed with a CMT volume of 2487 ft3. In this LAR, SNC also stated that a method for confirming the minimum volume requirement is through the use of the wide range (WR) level instrumentation. However, due to the configuration of the CMT WR level tap locations, and instrument uncertainty considerations, the CMT level reflecting 2487 ft3 cannot be measured. SNC stated that VEGP Unit 3 CMTs were fabricated close to the minimum volume and additional deviations to the location of nozzles used have reduced the measurable range, which precludes confirmation of measuring 2487 ft3 of borated water volume with the existing WR level instrumentation. This issue is specific to VEGP Unit 3, however, SNC is proposing this change to both Units 3 and 4 TS for consistency.

According to SNC, the only way to verify the CMT contains the necessary water volume is with high-point sensors which are used to verify compliance with existing TS SR 3.5.2.4 (TS SR 3.5.3.2 for TS 3.5.3). Existing TS SR 3.5.2.4 requires verification of the volume of noncondensible gases in each CMT inlet line has not caused the high-point water level to drop below the sensor every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The corresponding volume verified in TS SR 3.5.2.4 bounds TS SR 3.5.2.2 and therefore, in the unlikely event of loss of borated water volume in the CMTs, TS SR 3.5.2.4 will be affected first.

In the LAR, SNC stated that currently, failure of SR 3.5.2.2 would lead to entry into condition E (condition D for TS 3.5.3) with a restore action and completion time (CT) of eight hours. SNC also stated that as a consequence of deleting SR 3.5.2.2, failure of the existing SR 3.5.2.4 would lead to entry into condition D (condition C for TS 3.5.3) with a CT of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The NRC staff agrees with SNCs position that failure to meet SR 3.5.2.4 leads to entry into condition D (condition C for TS 3.5.3). However, since SNC is unable to verify minimum volume of borated water in the CMTs, there will be an indeterminate volume in the CMT which may lead to a potential noncompliance with the safety analysis. In its letter dated December 19, 2019 (ADAMS Accession No. ML19353B752), SNC stated that the amount of measurable volume with the WR level, accounting for error and uncertainty, is approximately 2475 ft3 for the most limiting as-built tank. Therefore, even though operators may not get timely indication of minimum volume of borated water not within limits until the indicated volume is 2475 ft3, the moment the volume of borated water is less than 2487 ft3, entry into condition E is required.

The NRC staff finds this proposed change acceptable for the following reasons:

As SNC stated, in part, in its letter dated December 19, 2019:

The high point gas accumulation instrumentation is approximately 28" above the inlet nozzle of the CMT and, therefore, when a gas accumulation alarm is not present, the entire CMT is full. If gas were to accumulate in the inlet line high point, it would generate an alarm. The expected operator action is to vent the gas to clear the alarm and re-confirm adequate CMT volume. The CMTs have a normally open inlet line connected to the RCS cold legs which maintains each CMT at RCS pressure. The CMT outlets have parallel, normally closed, air-operated valves. There is no credible mechanism for gross leakage from the CMT and therefore, when an alarm is present, it is not expected that the volume of the CMT would decrease beneath the tee connection of the high point.

Additionally, for the volume of the CMT to be affected, not only does the high point gas accumulation pipe stub require voiding, but also the inlet piping in the high point above the CMT inlet nozzle would need to be voided. With no gas voids in the inlet piping, RCS water would flow into the CMT from the balance line to maintain the CMT water volume.

In addition, SNC stated that because this action (Condition D as a result of not meeting SR 3.5.2.4) is entered with greater remaining volume in the CMT, restoration within the 24-hour CT would be expected in a similar time frame as waiting to enter until the lower volume is reached and providing for restoration in eight hours.

Based upon the information provided regarding the physical arrangement of CMT gas accumulation instrumentation and CMT WR level taps, when a high point gas alarm is present and the CMT wide range level is at 100%, the volume of the CMT is indeterminate, but there is some certainty that the CMT water volume is greater than 2475 ft3. In addition, without a significant amount of accumulated gas and the lack of a mechanism that could drain water below the top nozzle of the CMT, the water volume would not fall below the inlet nozzle (or vertical inlet piping before the tee). Thus, a reduction of volume in the CMT inlet line due to accumulation of non-condensable gas would occur slowly and venting of the gas intrusion volume within the 24-hour CT of Condition D would most likely result in restoration of CMT level prior to reduction below the safety analysis assumed volume of 2487 ft3.

In the unlikely event of a significant RCS leak leading to a decreased water volume in the CMTs, as stated by SNC in its letter dated December 19, 2019, TS 3.4.7, RCS Operational LEAKAGE, would not be met and the resulting required actions would be more restrictive than TS 3.5.2 required action E.1.

The NRC staff finds the proposed deletion of SR 3.5.2.2 acceptable since there are no credible leaking mechanisms for the CMT and therefore, water level is not expected to drop below the minimum required. In addition, in the unlikely event of leakage leading to a decreased water volume in the CMT, other TS actions would require more restrictive actions than those in TS 3.5.2 and TS 3.5.3.

In LAR 19-009, SNC also requests changes to VEGP Units 3 and 4 COL Appendix A TS 3.5.3.

Specifically, SNC proposes to add the words required CMT to required action B.1 and required to required action C.1. In the LAR, SNC stated these proposed changes are editorial and provide consistency between the required action and its entry condition statement. SNC also proposes to delete the words For the CMT required to be OPERABLE from SR 3.5.3.1 and SR 3.5.3.2. In the LAR, SNC stated that this phrase is unnecessary since the LCO clearly identifies the components to which the SRs are to be applied. In addition, SNC proposes to delete the term Normal Residual Heat Removal System from SR 3.5.3.2 since this term has been spelled out previously in the document and the acronym remains in the text of the surveillance required. SNC also proposed to add new SR 3.5.3.3 requiring a verification of borated water volume 2450 ft3 every seven days. This SR is modified by a note stating that it is only required to be met in Mode 5 with the RCS not vented. In the LAR, SNC stated that loss of RNS analysis for Mode 5 with RCS not vented is performed with a CMT volume of 2450 ft3 which is consistent with new SR TS 3.5.3.3. Finally, under LCO 3.5.3, SNC proposes to make conforming renumbering changes to the referenced SRs under SR 3.5.3.1 and SR 3.5.3.2.

During the audit conducted as part of this LAR review, the audit team reviewed the analysis of loss of RNS during Mode 5 with the RCS not vented with (1) reduced CMT volume of 2450 ft3 and (2) the proposed design changes for ADS Stages 1, 2, and 3. The audit team reviewed the calculation for Automatic Safeguards system actuation. The loss of RNS is assumed to occur 2.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> post shutdown. The analyses showed that the resultant core levels remained above the TAF. The acceptance criterion for the evaluation assumes active fuel must be covered by a two-phase mixture during the duration of the event. Assumptions for equipment availability were consistent with TSs for the applicable plant modes and plant conditions. Therefore, the staff found the licensees calculations acceptable.

The NRC staff finds the proposed changes to TS 3.5.3 acceptable since they are editorial and because the staff has determined that the changes do not alter the intent of the technical requirements. In addition, the NRC staff finds new SR 3.5.3.3 acceptable since it requires verification of a value assumed in the safety analysis at an appropriate frequency.

In LAR 19-009, SNC also requests conforming changes to TS 3.3.20, ADS and IRWST Injection Blocking Device. Specifically, SNC proposes to renumber the referred SRs under SR 3.3.20.7 as a result of the SR renumbering which occurred from the deletion of SR 3.5.2.2.

The NRC staff finds the proposed changes to TS 3.3.20 acceptable because they are editorial and because staff has determined that the changes do not alter the intent of the technical requirements.

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 by reference and comply with the requirements of Appendix D, 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 the licensee 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 the licensee could implement modifications to a COL Appendix C table and corresponding plant-specific Tier 1 information.

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 involved Tier 1 information described and justified in LAR 19-009. 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 and that the special circumstances, which are defined by 10 CFR 50.12(a)(2), outweigh any potential decrease in safety due to reduced standardization.

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 bases to be present in order for the NRC to consider granting an exemption request. The licensee 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 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 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 a plant based on the certified information found in the AP1000 DCD, which was incorporated by reference into the VEGP Units 3 and 4 licensing basis. The proposed changes described in the above technical evaluation do not impact the ability of any systems, structures or components to perform their functions or negatively impact safety.

Special circumstances are present in the particular circumstances discussed in LAR 19-009 because the application of the specified Tier 1 information is not necessary to achieve the underlying purpose of the rule. The proposed changes are equal or provide additional clarity to the existing requirement. The proposed changes do not affect any function or feature used for the prevention and mitigation of accidents or their safety analyses, and no safety-related SSC or function is involved. This exemption request and associated revisions to the Tier 1 information and corresponding changes to Appendix C demonstrate that the applicable regulatory requirements will continue to be met. 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 PS-DCD and corresponding changes to COL Appendix C that are being proposed in the LAR. The justification provided in LAR 19-009, 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. 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 ADS. Consequently, the safety impact that may result from any reduction in standardization is minimized, because the proposed design change does not result in a reduction in the level of safety. Based on the foregoing reasons, as required by 10 CFR Part 52.63(b)(1), 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 This exemption would allow the implementation of changes discussed above. The exemption request proposes to depart from the certified design by allowing changes discussed above in the technical evaluation. The changes for consistency will not impact the functional capabilities of this system. The proposed changes will not adversely affect the ability of the ADS 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 52.7, 10 CFR 52.98(f), and 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.

3.3

SUMMARY

In LAR 19-009, SNC proposed to make changes to the UFSAR Tier 2 information, which involve changes to COL Appendix C and corresponding PS-DCD Tier 1 information, and to the TSs in COL Appendix A, in order to reflect revisions to the ADS and CMT design parameters. The NRC staff documented its review of these changes in Section 3.1 of this safety evaluation.

Based on the reasons stated above, the staff finds the requested changes to the maximum stroke times for the ADS Stages 1, 2, and 3 valves; the minimum effective flow areas for the ADS Stages 2 and 3 valves; and the CMT minimum volume, to be acceptable. The NRC staff also finds that VEGP Units 3 and 4 will continue to satisfy 10 CFR 52.97(b) for ITAAC, GDC 10 for reactor design, GDC 15 for RCS design, GDC 29 for AOO protection, GDC 34 for residual heat removal capability, and GDC 35 for emergency core cooling performance, with the modifications described in LAR 19-009.

4.0 STATE CONSULTATION

In accordance with the Commission's regulations in 10 CFR 50.91(b)(2), the Georgia State official was notified of the proposed issuance of the amendment on February 14, 2020. The State official had no comments.

5.0 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, and changes surveillance requirements. 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, and that 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 December 3, 2019 (84 FR 66234). Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).

Pursuant to 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 the license amendment, and because the exemption does not authorize any activities other than those proposed in the license amendment, 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), no environmental impact statement or environmental assessment needs to be prepared in connection with the issuance of the exemption.

6.0 CONCLUSION

The staff has determined that pursuant to Section VIII.A.4 of Appendix D to 10 CFR Part 52, the exemption (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 reduce the level of safety at the licensees facility. Therefore, the staff grants the licensee an exemption from the Tier 1 information requested by the licensee.

The staff has concluded, based on the considerations discussed 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 staff finds the changes proposed in this license amendment acceptable.

7.0 REFERENCES

1. Southern Nuclear Operating Company, Vogtle Electric Generating Plant Units 3 and 4, Request for License Amendment and Exemption: Automatic Depressurization System (ADS) and Core Makeup Tank (CMT) Design Parameters (LAR 19-009), September 30, 2019 (ADAMS Accession No. ML19273A953).

2. Southern Nuclear Operating Company, Vogtle Electric Generating Plant Units 3 and 4, Supplement to Request for License Amendment and Exemption: Automatic Depressurization System (ADS) and Core Makeup Tank (CMT) Design Parameters (LAR 19-009S1), December 19, 2019 (ADAMS Accession No. ML19353B752).

3. Vogtle Electric Generating Plant Units 3 and 4, Updated Final Safety Analysis Report, Revision 8, and Tier 1, Revision 7, June 14, 2019 (ADAMS Accession No. ML19171A096).

4. Westinghouse Electric Companys AP1000 Design Control Document, Revision 19, June 13, 2011 (ADAMS Accession No. ML11171A500).

5. Vogtle Electric Generating Plant Unit 3, Current Facility Combined License NPF-91, Revised June 25, 2019 (ADAMS Accession No. ML14100A106).

6. Vogtle Electric Generating Plant Unit 4, Current Facility Combined License NPF-92, Revised June 25, 2019 (ADAMS Accession No. ML14100A135).

7. NRC Inspection Procedure (IP) 73758, Part 52, Functional Design and Qualification, and Preservice and Inservice Testing Programs for Pumps, Valves and Dynamic Restraints (ADAMS Accession No. ML18222A281).

8. Regulatory Guide 1.100, Revision 3, Seismic Qualification of Electrical and Active Mechanical Equipment and Functional Qualification of Active Mechanical Equipment for Nuclear Power Plants, September 2009 (ADAMS Accession No. ML091320468).

9. Vogtle Electric Generating Plant Units 3 and 4, Audit Report: Automatic Depressurization System (ADS) and Core Makeup Tank (CMT) Design Parameters (LAR 19-009),

March 11, 2020 (ADAMS Accession No. ML20065M803).

10. ASME Standard QME-1-2007, Qualification of Active Mechanical Equipment Used in Nuclear Power Plants.