ML20196L682
| ML20196L682 | |
| Person / Time | |
|---|---|
| Site: | Vogtle  |
| Issue date: | 08/11/2020 |
| From: | Jennivine Rankin NRC/NRR/VPOB |
| To: | City of Dalton, GA, Georgia Power Co, MEAG Power, Oglethorpe Power Corp, Southern Nuclear Operating Co |
| Jennivine Rankin EX. 1530 | |
| References | |
| Download: ML20196L682 (10) | |
Text
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NOS. 183 AND 181 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 February 28, 2020 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML20059N597), 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. The License Amendment Request (LAR)20-002 requested changes to the following VEGP Units 3 and 4 COL Appendix A Technical Specifications (TS):
A. Surveillance Requirement (SR) 3.7.6.3 frequency revised for Main Control Room Emergency Habitability System (VES) operation and deleted SR 3.7.6.9 which verifies the self-contained pressure regulating valve in each VES air delivery flow path is operable in accordance with the Inservice Testing (IST) Program; B. SR 3.3.8.2 for channel calibration and SR 3.3.8.3 for Engineered Safety Feature (ESF)
Response Time revised to include a Note excluding neutron detectors; C. TS 5.5.3, Inservice Testing Program, revised to replace existing detail with a reference to fulfilling the requirements of Title 10 of the Code of Federal Regulations (10 CFR) 50.55a(f);
D. TS 5.5.9, System Level OPERABILITY Testing Program, revised for appropriate wording consistency and appropriate reference to the Updated Final Safety Analysis Report (UFSAR); and E. TS 3.4.9, RCS [Reactor Coolant System] Leakage Detection Instrumentation, Applicability Note 2 revised to consistently identify the applicable power level.
2.0 REGULATORY EVALUATION
The staff considered the following regulatory requirements in reviewing the LAR that included the proposed changes.
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 TS, 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, in part, 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 governs the application for amendment of license, construction permit, or early site permit. The proposed LAR involves changes to the TS, and therefore an LAR is required to be submitted for NRC approval.
10 CFR 50.36, TS impose safety limits, limiting conditions for operation, and establish other requirements on reactor facility operation. The TS are derived from the analyses and evaluations in the safety analysis report and amendments thereto. In general, TS must contain:
(1) safety limits, limiting safety system settings, and limiting control settings; (2) limiting conditions for operation; (3) surveillance requirements; (4) design features; and (5) administrative controls.
The requirements for preservice IST of pumps and valves at nuclear power reactors are in accordance with the American Society of Mechanical Engineers (ASME) Operation and Maintenance of Nuclear Power Plants, Division 1, OM Code: Section IST (OM Code) as specified in 10 CFR 50.55a(f). The regulations in 10 CFR 50.55a(f) state, in part:
Systems and components of boiling and pressurized water-cooled nuclear power reactors must meet the requirements for preservice and inservice testing (referred to in this paragraph (f) collectively as inservice testing) of the ASME BPV [Boiler and Pressure Vessel] Code and ASME OM Code as specified in this paragraph (f). Each operating license for a boiling or pressurized water-cooled nuclear facility is subject to the following conditions [referring to 10 CFR 50.55a(f)(1) through (f)(6)].
Each combined license for a boiling or pressurized water-cooled nuclear facility is subject to the following conditions, but the conditions in paragraphs (f)(4) through (6) of this section must be met only after the Commission makes the finding under § 52.103(g) of this chapter.
The ASME OM Code is a consensus standard that is incorporated by reference into 10 CFR 50.55a. During the incorporation process, the NRC staff reviewed the ASME OM Code
requirements for technical sufficiency and found that the ASME OM Code IST program requirements were suitable for incorporation into the NRCs regulations.
The regulation in 10 CFR 50.55(a)(f)(5)(ii) states, in part, [i]f a revised inservice test program for a facility conflicts with the technical specifications for the facility, the licensee must apply to the Commission for amendment of the technical specifications to conform the technical specifications to the revised program.
The specific NRC technical requirements applicable to LAR 20-002 are the general design criteria (GDC) in Appendix A, General Design Criteria for Nuclear Power Plants, to 10 CFR Part 50. In particular, these technical requirements include the following GDCs:
GDC 13, Instrumentation and Control, requires [i]nstrumentation shall be provided to monitor variables and systems over their anticipated ranges for normal operation, for anticipated operational occurrences, and for accident conditions as appropriate to assure adequate safety, including those variables and systems that can affect the fission process, the integrity of the reactor core, the reactor coolant pressure boundary, and the containment and its associated system. Appropriate controls shall be provided to maintain these variables and systems within prescribed operating ranges.
GDC 19, Control Room, requires, in part, [a] control room shall be provided from which actions can be taken to operate the nuclear power unit safely under normal conditions and to maintain it in a safe condition under accident conditions, including loss-of-coolant accidents.
GDC 21, Protection System Reliability and Testability, requires, in part, [t]he protection system shall be designed to permit periodic testing of its functioning when the reactor is in operation, including a capability to test channels independently to determine failures and losses of redundancy that may have occurred.
The regulatory guidance applicable to LAR 20-002 includes the following:
The NRC staffs guidance for review of TS is provided in NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [Light-Water Reactor] Edition, Chapter 16, Technical Specifications, Revision 3, dated March 2010 (ADAMS Accession No. ML100351425). As described therein, as part of the regulatory standardization effort, the staff has prepared improved Standard Technical Specifications (STS) for each of the LWR nuclear steam supply systems and associated balance-of-plant equipment systems. Part of SNCs LAR is consistent with Technical Specifications Task Force (TSTF) 545, Revision 3, TS Inservice Testing Program Removal & Clarify SR Usage Rule Application to Section 5.5 Testing, dated December 11, 2015 (ADAMS Accession No. ML15317A071),
which is an NRC-approved change to the improved STS. The staffs review includes consideration of whether the proposed changes are consistent with or meet the intent of TSTF-545, Revision 3. Special attention is given to TS provisions that depart from the improved STS, as modified by NRC-approved TSTF travelers, to determine whether proposed differences are justified by the uniqueness of the plant design or other considerations so that 10 CFR 50.36 is met. In addition, the guidance states that comparing the change to previous STS can help clarify the TS intent.
TSTF-545, Revision 3, proposes elimination of the IST Program from the Administrative Controls section of the TS. The TS contain surveillances that require testing or test intervals in accordance with the IST Program. The elimination of the IST Program from the TS could cause
uncertainty regarding the correct application of these SRs. Therefore, TSTF-545, Revision 3, also proposes adding a new defined term, INSERVICE TESTING PROGRAM, to the TS, which would be defined as the licensee program that fulfills the requirements of 10 CFR 50.55a(f).
NUREG-1482, Revision 2, Guidelines for Inservice Testing at Nuclear Power Plants: Inservice Testing of Pumps and Valves and Inservice Examination and Testing of Dynamic Restraints (Snubbers) at Nuclear Power Plants, dated October 2013 (ADAMS Accession No. ML13295A020) provides guidance for the inservice testing of pumps, valves, and dynamic restraints.
3.0 TECHNICAL EVALUATION
3.
0.1 TECHNICAL EVALUATION
OF SR 3.7.6.3 and SR 3.7.6.9 Changes In LAR 20-002, SNC explains that SR 3.7.6.9 requires verification that the self-contained pressure regulating valve in each VES air delivery flow path is OPERABLE; however, test requirements for the pressure regulating valves are not addressed in the IST program requirements of the ASME OM Code. The AP1000 Design Control Document (DCD) and the VEGP Units 3 and 4 plant-specific UFSAR were updated to reflect quarterly exercise stroke tests consisting of a pressure drop test across the valves using the downstream test connection to ensure adequate testing of the valves. In LAR 20-002, SNC states that the AP1000 Final Safety Evaluation Report (FSER) notes that the VES pressure regulating valves are exempt from the ASME OM Code; however, SR 3.7.6.9 was not revised. SNC proposes to delete SR 3.7.6.9 and renumber the subsequent SRs. In addition, SR 3.7.6.3 currently requires operation of the VES for > 15 minutes every 31 days. SNC proposes to stagger the three air delivery flow paths by changing the SR 3.7.6.3 frequency from 31 days to 31 days on a STAGGERED TEST BASIS which will result in at least quarterly exercising of each air delivery valve.
NRC Evaluation of SR 3.7.6.3 and SR 3.7.6.9 Changes The NRC staff reviewed LAR 20-002, the AP1000 DCD, and UFSAR Section 3.9, Mechanical Systems and Components, with respect to the proposed SR 3.7.6.3 and SR 3.7.6.9 changes.
SNC indicates that the main control room emergency habitability system pressure regulating valves VES-PL-V002A and V002B are not part of the ASME OM Code IST Program.
Table 3.9-16, Valve Inservice Test Requirements, Note 38, in the VEGP Units 3 and 4 UFSAR states that the exercise stroke test for the VES pressure regulating valves will consist of a pressure drop test across the valve using the downstream test connection. Note 38 also states that this method ensures adequate testing of the valves. The staff notes that these valves will be tested as part of the augmented IST program. The staff finds that the testing described in Note 38 of Table 3.9-16 provides reasonable assurance of the capability of the valves to perform their safety function.
In addition to deleting SR 3.7.6.9, SNC proposes to revise existing SR 3.7.6.3, which requires monthly operation of VES for 15 minutes from a frequency of 31 days to 31 days on a STAGGERED TEST BASIS.
UFSAR Section 6.6, Inservice Inspection of Class 2, 3, and MC Components, describes the VES, which provides emergency ventilation and pressurization for the main control room if alternating current power is unavailable or control room air supply radioactivity is high. The main control room emergency habitability system also provides emergency passive heat sinks
for the main control room, instrumentation and control rooms, and direct current equipment rooms. The VES system is tested and inspected at appropriate intervals, as defined by the TS.
UFSAR Figure 6.4-2 shows the simplified piping and instrumentation diagram (P&ID). The TS Bases for SR 3.7.6.3 state, [s]tandby systems should be checked periodically to ensure that they function properly. As the environment and normal operating conditions on this system are not too severe, testing VES once every month provides an adequate check of the system.
The NRC staff notes in the revised TS Bases for SR 3.7.6.3 that this monthly VES operation requires operation of one of the two pressure regulating valves, which each require quarterly testing (per UFSAR Table 3.9-16 and Note 38). Additionally, the main air delivery line has two parallel air delivery flow paths via automatically actuated isolation valves (VES-V005A/B) that are also required to be exercised quarterly. Monthly testing of the three air delivery flow paths on a STAGGERED TEST BASIS will result in at least quarterly exercising of each air delivery valve. The 31 day on a STAGGERED TEST BASIS Frequency is based on the reliability of the equipment and the availability of system redundancy. The NRC staff verified from the simplified P&ID that component redundancy exists such that this testing plan satisfies the quarterly testing requirement.
The NRC staff notes that the effect of deleting SR 3.7.6.9 and revising SR 3.7.6.3 is consistent with the requirements provided in UFSAR Table 3.9-16 without referencing the ASME OM Code IST Program. The NRC staff concludes that the testing provision for the VES pressure regulating valves provide reasonable assurance of their capability to perform the specified safety functions and does not result in a change to the VES design or radiological dose rates to the main control room operators during a design basis accident. Therefore, the NRC staff finds that GDC 19 is satisfied and the proposed TS changes continue to meet the requirements of 10 CFR 50.36.
3.
0.2 TECHNICAL EVALUATION
OF SR 3.3.8.2 and SR 3.3.8.3 Changes The Nuclear Instrumentation System (NIS) is a subsystem of the Protection and Safety Monitoring System (PMS) that measures neutron leakage from the reactor core over the full range of reactor power by using source range, intermediate range (IR), and power range (PR) neutron detectors. The NIS provides an Engineered Safety Feature Actuation System (ESFAS) protection function against unplanned reactor criticality resulting from inadvertent boron dilution of the RCS. SR 3.3.8.2 requires performance of channel calibration in accordance with the Setpoint Program and SR 3.3.8.3 requires verification that ESF response time is within limit for the ESFAS system.
SNC proposes to add Notes to SR 3.3.8.2 and 3.3.8.3 to exclude neutron detectors from channel calibration and response time testing, respectively. LAR 20-002 states that the NIS source range neutron detectors in TS 3.3.2 and 3.9.3 have a note that excludes neutron detectors from both channel calibration and response time testing. In addition, the same exception is applied for PR and IR neutron detectors TS 3.3.1 and TS 3.3.3.
NRC Evaluation of SR 3.3.8.2 and SR 3.3.8.3 Changes The NRC staff reviewed LAR 20-002 and the associated TS and TS Bases. LAR 20-002 states that the proposed neutron detector exclusions to SR 3.3.8.2 and SR 3.3.8.3 achieve consistency with other source range neutron flux Surveillances. The NRC staff verified that current SR 3.3.1.6 and SR 3.3.1.8 also include a Note to exclude neutron detectors from both channel calibration and response time testing for the reactor trip system. Current SR 3.3.3.2 and SR 3.9.3.2 exclude neutron detectors from channel calibration for the post accident
monitoring system and nuclear instrumentation, respectively. In addition, the NRC staff verified that SR 3.3.1.11 and SR 3.3.1.16 in the Standard Technical Specifications for Westinghouse Plants, NUREG-1431, Revision 4.0 (ADAMS Accession No. ML12100A222) also contain the exclusion of neutron detectors from both channel calibration and response time testing. The current TS Bases include similar statements and reasons to exclude neutron detectors for the above corresponding SRs. Specifically, this is because their calibration is instead performed by obtaining the neutron detector plateau or preamp discriminator curves, evaluating those curves, and comparing the curves to the manufacturers data. In addition, the measuring principles for the neutron detectors provide a virtually instantaneous response, which would have no measurable or adverse impact on the response time performance of the NIS protection function for the ESFAS system.
From the above evaluation, the NRC staff finds that the proposed Notes to SR 3.3.8.2 and 3.3.8.3 maintain consistency with other relevant SRs. Additionally, the NRC staff finds that the proposed changes to add Notes to SR 3.3.8.2 and 3.3.8.3 do not have any impact on the NIS design, PMS design, or its capability for periodic testing during reactor power operation.
Therefore, the NRC staff concludes that the regulatory requirements of GDC 13 regarding instrumentation and control and GDC 21 regarding protection system reliability and testability are satisfied, and the proposed TS changes continue to meet the requirements of 10 CFR 50.36.
3.
0.3 TECHNICAL EVALUATION
OF TS 5.5.3 CHANGES In LAR 20-002, SNC proposes a change to TS Section 5.5.3, Inservice Testing Program, to remove requirements duplicated in, or are in conflict with the ASME OM Code. SNC notes that the proposed change is consistent with TSTF-545, Revision 3, TS Inservice Testing Program Removal & Clarify SR Usage Rule Application to Section 5.5 Testing, with the following variations: (1) replace the removed details with The Inservice Testing Program is the licensee program that fulfills the requirements of 10 CFR 50.55a(f) instead of the word DELETED, and (2) not adding INSERVICE TESTING PROGRAM as a new defined term.
NRC Evaluation of TS 5.5.3 Changes Presently, TS 5.5.3 provides control for IST of ASME Code Class 1, 2 and 3 components including applicable supports. TS 5.5.3 addresses four items. First, TS 5.5.3.a describes how to interpret monthly, quarterly, and other time periods used in testing. Second, TS 5.5.3.b describes when a licensee can avail itself of the provisions of SR 3.0.2 and thereby change the required testing schedule. Third, TS 5.5.3.c allows usage of SR 3.0.3, which describes how a licensee is to treat a situation where it discovers that a SR was not performed within its specified frequency. Fourth, TS 5.5.3.d states that nothing in the ASME OM Code shall be construed to supersede the requirements of any TS.
10 CFR 50.55a(f) specifies the requirements for the IST of pumps and valves. Therefore, requiring the licensee to have an IST program in the TS is duplicative. Thus, with the proposed TS changes, the licensee will continue to be required to maintain an IST program in accordance with the ASME OM Code, as specified in 10 CFR 50.55a(f). For the reasons explained below, it is not necessary to have additional administrative controls in the TS relating to the IST program to assure operation of the facility in a safe manner.
Consideration of TS 5.5.3.a
The ASME OM Code requires testing to normally be performed within certain time periods. TS 5.5.3.a sets IST frequencies more precisely than those specified in the ASME OM Code and applicable addenda (e.g., at least once per 31 days contrasted with monthly). The NRC staff determined that the more precise IST frequencies are not necessary to assure operation of the facility in a safe manner. Therefore, the NRC staff concludes that deletion of TS 5.5.3.a is acceptable.
Consideration of TS 5.5.3.b TS 5.5.3.b allows the licensee to extend, by up to 25 percent, the interval between IST activities, as required by TS 5.5.3.a and for other normal and accelerated frequencies specified as 2 years or less in the IST program. Similar to TS 5.5.3.b, the NRC has approved the use of ASME Code Case OMN-20, Inservice Test Frequency, to allow licensees to extend the IST intervals specified in the ASME OM Code by up to 25 percent with the acceptance of Code Case OMN-20 in NRC Regulatory Guide (RG) 1.192, Revision 3, Operation and Maintenance Code Case Acceptability, ASME OM Code, as incorporated by reference in 10 CFR 50.55a. ASME Code Case OMN-20 can be used with editions and addenda of the OM Code incorporated by reference in 10 CFR 50.55a. The NRC staff determined that the TS 5.5.3.b allowance to extend IST intervals is not needed to assure operation of the facility in a safe manner. Therefore, the NRC staff concludes that deletion of TS 5.5.3.b is acceptable.
Consideration of TS 5.5.3.c TS 5.5.3.c allows the licensee to use SR 3.0.3 when it discovers that an SR associated with an inservice test was not performed within its specified frequency. SR 3.0.3 allows the licensee to delay declaring a limiting condition for operation not met in order to perform the missed surveillance. The use of SR 3.0.3 for inservice tests is limited to those inservice tests required by an SR. In accordance with 10 CFR 50.55a, the licensee may also request relief from the ASME OM Code requirements to address issues associated with a missed inservice test.
Deletion of TS 5.5.3.c does not change any of these requirements, and SR 3.0.3 will continue to apply to those inservice tests required by SRs. Based on the above, the NRC staff concludes that deletion of TS 5.5.3.c is acceptable.
Consideration of TS 5.5.3.d TS 5.5.3.d states that nothing in the ASME OM Code shall be construed to supersede the requirements of any TS. However, the regulations in 10 CFR 50.55a(f)(5)(ii) specify the requirements if a revised IST program for a facility conflicts with the TS for the facility. In particular, the regulations require the licensee to apply for an amendment to the TS to conform the TS to the revised IST program at least 6 months prior to the start of the period for which the provisions become applicable. Accordingly, there is no need for the TS to address conflicts between the TS and the IST program, because the regulations specify how conflicts must be resolved. Based on the above, the NRC staff concludes that deletion of TS 5.5.3.d is acceptable.
Conclusion Regarding Deletion of TS 5.5.3 The NRC staff determined that the requirements currently in TS 5.5.3 are not necessary to assure operation of the facility in a safe manner. Based on this evaluation, the staff concludes that deletion of TS 5.5.3 from the licensees TS is acceptable because TS 5.5.3 is not required
by 10 CFR 50.36(c)(5). Moreover, the sentence added in its place is acceptable to the NRC staff because it correctly refers to the IST requirements in 10 CFR 50.55a(f).
3.
0.4 TECHNICAL EVALUATION
OF TS 5.5.9 CHANGES The current language in the second sentence of TS 5.5.9 reads as follows:
The System Level Inservice Tests specified in FSAR Section 3.9.6 and FSAR Table 3.9-17 apply when specified by individual Surveillance Requirements.
SNC proposes changing the language to accurately reflect the title of FSAR Table 3.9-17 and to remove an incorrect reference to FSAR Section 3.9.6. The proposed change reads as follows:
The System Level OPERABILITY Test Requirements specified in FSAR Table 3.9-17 apply when specified by individual Surveillance Requirements.
NRC Evaluation of TS 5.5.9 Changes In LAR 20-002, SNC states that UFSAR Section 3.9.6, Inservice Testing of Pumps and Valves, include periodic system level tests and inspections that demonstrate the capability of safety-related features to perform their safety-related functions such as passing flow or transferring heat. SNC also states that these tests may be performed in conjunction with inservice tests conducted to exercise check valves or to perform power-operated valve operability tests. In addition, SNC states that alternate means of performing these tests and inspections that provide equivalent demonstration may be developed in the IST program.
UFSAR Table 3.9-17, System Level Operability Test Requirements, identifies the system level tests which are not governed by the ASME OM Code.
In LAR 20-002, SNC states that TS 5.5.9 provides the TS reference to these tests as addressed in individual Surveillances. SNC also states that the specific wording in the second sentence of TS 5.5.9 contains a misleading reference to System Level Inservice Tests [emphasis added].
SNC proposed this sentence in TS 5.5.9 be revised to reflect the actual title of Table 3.9-17 as System Level OPERABILITY Test Requirements. Additionally, SNC states that TS 5.5.9 specifies that these tests are contained in UFSAR Section 3.9.6 and UFSAR Table 3.9-17; however, the tests are only found in Table 3.9-17.
The NRC staff confirmed the system level operability test requirements are contained in the UFSAR Table 3.9-17. In addition, the NRC staff confirmed that these tests are only found in Table 3.9-17 and not contained in FSAR Section 3.9.6. The staff concludes that the proposed changes to TS 5.5.9 are administrative in nature and continue to meet the requirements of 10 CFR 50.36(c)(5).
3.
0.5 TECHNICAL EVALUATION
OF TS 3.4.9 CHANGES The current language TS 3.4.9 Applicability Note 2 reads as follows:
The containment atmosphere F18 particulate monitor is only required to be OPERABLE in MODE 1 with RTP [RATED THERMAL POWER] >20%.
SNC proposes changing the language to consistently identify the applicable power level. The proposed change reads as follows:
The containment atmosphere F18 particulate monitor is only required to be OPERABLE in MODE 1 with THERMAL POWER >20% RTP.
NRC Evaluation of TS 3.4.9 Changes The NRC staff reviewed the acronym definitions and confirms that the RTP is expressed as a fixed value and thermal power is expressed as a percentage of RTP. Thus, using THERMAL POWER >20% RTP is a better clarification of RTP >20%. The NRC staff concludes the proposed changes to TS 3.4.9 consistently identify the applicable power level, are administrative in nature, and continue to meet the requirements of 10 CFR 50.36.
4.0 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.
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 and changes SRs.
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 (85 FR 16684) dated March 24, 2020. 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.
6.0 CONCLUSION
The staff has concluded, based on the considerations discussed in Section 3.0 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: Main Control Room Emergency Habitability System (VES) Pressure Regulating Valve and Source Range Neutron Flux Doubling Surveillance Changes, and Other Miscellaneous Technical Specification Changes (LAR-20-002),
February 28, 2020 (ADAMS Accession No. ML20059N597).
- 2. 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).
- 3. AP1000 Design Control Document, Revision 19, June 13, 2011 (ADAMS Accession No. ML11171A500).
- 4. Combined License NPF-91 for Vogtle Electric Generating Plant Unit 3, Southern Nuclear Operating Company (ADAMS Accession No. ML14100A106).
- 5. Combined License NPF-92 for Vogtle Electric Generating Plant Unit 4, Southern Nuclear Operating Company (ADAMS Accession No. ML14100A135).
- 6. TSTF-545, Revision 3, Final Model Safety Evaluation of Technical Specifications Task Force Traveler TSTF-545, Revision 3, TS Inservice Testing Program Removal & Clarify SR Usage Rule Application to Section 5.5 Testing, December 11, 2015 (ADAMS Accession No ML15317A071).
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