Defective power operated relief valve
By letter dated August 10, 2018 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML18222A599), revised October 11, 2018 (ADAMS Accession No. ML18284A447), and supplemented February 15, 2019 (ADAMS Accession No. ML19046A172), the Southern Nuclear Operating Company (SNC) requested that the U.S. Nuclear Regulatory Commission (NRC or the Commission) amend Vogtle Electric Generating Plant (VEGP) Units 3 and 4, Combined License (COL) Numbers NPF-91 and NPF-92, respectively. License Amendment Request (LAR) 18-021 seeks departures from the generic AP1000 Design Control Document (DCD) Tier 1 in the VEGP COL plant-specific DCD (PS-DCD) with corresponding changes to the associated COL Appendix C, and the Updated Final Safety Analysis Report (UFSAR) to relocate the power operated relief valve (PORV) branch lines upstream of the main steam safety valves (MSSVs) in the main steam (MS) lines. In addition to the relocation of the PORV branch lines, the LAR seeks to change the PORV block valves from gate valves to globe valves in the UFSAR. Specifically, this amendment results in changes to COL Appendix C (plant-specific Tier 1) Figure 2.2.1-1 and Figure 2.2.4-1 (Sheets 1 and 2), and UFSAR Figure 10.3.2-1 (Sheets 1 and 2), UFSAR Figure 3E-1 (Sheets 1 and 2), UFSAR Table 3.9-16, UFSAR Table 6.2.3-1, and UFSAR Table 10.3.3-1. These changes are sought by SNC to reduce the noise contribution to the main control room (MCR) and improve human factors when the PORVs are in operation.
2.0 REGULATORY EVALUATION
SNC summarized the proposed changes to plant-specific Tier 1 (and COL Appendix C) and the UFSAR related to relocating the PORV branch lines upstream of the MSSV branch connections, and changing the PORV block valves from gate valves to globe valves, as follows:
• COL Appendix C (plant-specific Tier 1) Figure 2.2.1-1 and Figure 2.2.4-1 (Sheets 1 and 2) are revised to move the PORV branch line upstream of the MSSVs. • UFSAR Table 3.9-16, Table 10.3.3-1, Figure 10.3.2-1 (Sheets 1 and 2), and Figure 3E-1 (Sheets 1 and 2) are revised to change the MS line PORV block valves (SGS-PL-V027A/B) from gate valves to globe valves, to relocate the branch lines to the PORV and PORV block valve to upstream of the MSSVs, to resize the line from 6 to 12 inches, and to remove the reducer from downstream of the PORV block valve. • UFSAR Table 6.2.3-1 is revised to change the pipe length from each containment penetration to valves SGS-PL-V027A and SGS-PL-V027B to 26 feet. • Technical Specifications (TS) Bases B 3.7.102 is revised to remove the size of the branch line in which the PORV is installed.
The staff considered the following regulatory requirements in reviewing the LAR that included the proposed changes:
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. 2 The staff notes that changes to TS Bases are not required to be reviewed and approved by the staff but are mentioned here for completeness.
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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 TS, or requires a license amendment under paragraphs B.5.b or B.5.c of the section.
10 CFR 50.55a, “Codes and standards,” incorporates by reference the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (BPV Code), and ASME Operation and Maintenance of Nuclear Power Plants, Division 1, OM Code: Section IST (ASME OM Code), including specific editions, addenda, and codes cases, for the design, inservice inspection, and inservice testing of nuclear power plant components. As guidance, the NRC endorses ASME Standard QME-1-2007, “Qualification of Active Mechanical Equipment Used in Nuclear Power Plants,” in Regulatory Guide (RG) 1.100 (Revision 3), “Seismic Qualification of Electrical and Active Mechanical Equipment and Functional Qualification of Active Mechanical Equipment for Nuclear Power Plants.”
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 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.
The specific NRC technical requirements applicable to LAR 18-021 are the general design criteria (GDC) in Appendix A, “General Design Criteria for Nuclear Power Plants,” to 10 CFR Part 50. These technical requirements include the following GDC:
10 CFR Part 50, Appendix A, GDC 1, “Quality standards and records,” requires that structures, systems, and components (SSCs) important to safety shall be designed, fabricated, erected, and tested to quality standards commensurate with the importance of the safety functions to be performed. A quality assurance program shall be established and implemented to provide adequate assurance that these SSCs will satisfactorily perform their safety functions. Appropriate records of the design, fabrication, erection, and testing of SSCs important to safety shall be maintained by or under the control of the nuclear power unit licensee throughout the life of the unit.
10 CFR Part 50, Appendix A, GDC 2, “Design Bases for Protection Against Natural Phenomena,” requires that SSCs important to safety shall be designed to withstand the effects
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of natural phenomena such as earthquakes, tornadoes, hurricanes, floods, tsunami, and seiches without loss of capability to perform their safety functions.
10 CFR Part 50, Appendix A, GDC 4, “Environmental and dynamic effects design bases,” requires that nuclear power plant SSCs important to safety be designed to accommodate the effects of, and be compatible with, environmental conditions associated with normal operation, maintenance, testing, and postulated accidents, including loss-of-coolant accidents. These SSCs shall be appropriately protected against dynamic effects, including the effects of missiles, pipe whipping and discharging fluids, that may result from equipment failures and from events and conditions outside the nuclear power unit. However, dynamic effects associated with postulated pipe ruptures in nuclear power units may be excluded from the design basis when analyses reviewed and approved by the Commission demonstrate that the probability of fluid system piping rupture is extremely low under conditions consistent with the design basis for the piping.
3.0 TECHNICAL EVALUATION
VEGP Units 3 and 4 COL, Appendix C, Subsection 2.2.4, contains ITAAC for the steam generator system (SGS). COL Appendix C, Figure 2.2.4-1 shows the SGS piping from the steam generators (SGs) through the auxiliary building and depicts the PORV as located downstream of the MSSVs. During plant operations, the PORVs are automatically controlled by steam line pressure, and modulate open and exhaust to the atmosphere whenever the steam line pressure exceeds an established setpoint. When needed for plant cooldown, the PORVs are automatically controlled by steam line pressure with remote manual adjustment of the pressure setpoint from the control room or remote shutdown workstation. To cool down the plant, the reactor operator manually adjusts the pressure setpoint downward in discrete steps or takes manual control of the valve position. Each PORV is installed in a branch line off the safety-related portion of the MS line upstream of the main steam isolation valve (MSIV). The PORV block valves perform the safety-related functions of containment isolation, SG isolation, and SG relief isolation. The PORV block valves also provide the capability to isolate a leaking or stuck-open PORV. The PORV block valves are AP1000 Safety Class B, and close automatically on a Low-2 steam line pressure signal generated in the Protection and Safety Monitoring System.
To reduce noise due to acoustic resonance, SNC proposes in LAR 18-021 to increase the size of the PORV block valves from 6 inches to 12 inches in diameter. SNC also proposes to increase the branch line in which each PORV block valve is installed from 6 inches to 12 inches. These changes will reduce the flow velocity in the PORV branch lines. In addition, SNC proposes to change the PORV block valve type from a gate valve to a globe valve to mitigate the potential for a Helmholtz or standing wave source developing in the valve body or seat. The globe valve will be qualified for the same environmental, pressure, and temperature conditions as the current valve type. Due to layout constraints, SNC proposes to move the location of the branch line for PORV block valves upstream of the MSSV branch connections.
3.1 TECHNICAL EVALUATION OF THE REQUESTED CHANGES
The staff conducted a regulatory audit from November 5, 2018, to January 31, 2019, to review applicable documents provided by SNC in its electronic reading room (eRR) in support of the proposed changes described in LAR 18-021. As part of that audit, the staff conducted telephone conferences with SNC to clarify information in specific documents. The staff’
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